Canadian Fusion Fuels Technology Project annual report 93/94

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-31

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

Canadian Fusion Fuels Technology Project annual report 93/94

International Nuclear Information System (INIS)

1994-01-01

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

BWR fuel performance

International Nuclear Information System (INIS)

Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

1979-01-01

The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

Third annual report of the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Dixon, R.S.; Rosinger, E.L.J.

1981-12-01

This report, the third of a series of annual reports, reviews the progress that has been made in the research and development program for the safe management and disposal of Canada's nuclear fuel wastes. The report summarizes activities over the past year, in the areas of public interaction, irradiated fuel storage and transportation, immobilization of irradiated fuel and fuel recycle wastes, research and development associated with deep underground disposal, and environmental and safety assessment

Fourth annual report of the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Rosinger, E.L.J.; Dixon, R.S.

1982-12-01

This report, the fourth of a series of annual reports, reviews the progress that has been made in the research and development program for the safe management and disposal of Canada's nuclear fuel waste. The report summarizes activities over the past year in the following areas: public interaction, used fuel storage and transportation, immobilization of used fuel and fuel recycle waste, geoscience research associated with deep underground disposal, environmental research, environmental and safety assessment

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.

2016 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

Energy Technology Data Exchange (ETDEWEB)

None, None

2017-03-09

The 2016 Annual Progress Report summarizes fiscal year 2016 activities and accomplishments by projects funded by the DOE Hydrogen and Fuel Cells Program. It covers the program areas of hydrogen production; hydrogen delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes and standards; systems analysis; market transformation; and Small Business Innovation Research projects.

Second annual report of the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Boulton, J.

1980-12-01

This report, the second of a series of annual reports, reviews in general terms the progress which has been achieved in the research and development program for the safe, permanent disposal of Canada's nuclear fuel wastes. The report summarizes activities over the past year, in the areas of public interaction, used fuel storage and transportation, immobilization of used fuel and fuel reprocessing wastes, research and development associated with deep underground disposal, and environmental and safety assessment. (auth)

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

International Nuclear Information System (INIS)

Salmon, R.; Kee, C.W.

1977-06-01

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

CANDU fuel performance

International Nuclear Information System (INIS)

Ivanoff, N.V.; Bazeley, E.G.; Hastings, I.J.

1982-01-01

CANDU fuel has operated successfully in Ontario Hydro's power reactors since 1962. In the 19 years of experience, about 99.9% of all fuel bundles have performed as designed. Most defects occurred before 1979 and subsequent changes in fuel design, fuel management, reactor control, and manufacturing quality control have reduced the current defect rate to near zero. Loss of power production due to defective fuel has been negligible. The outstanding performance continues while maintaining a low unit energy cost for fuel

First annual report of the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Boulton, J.; Gibson, A.R.

1979-12-01

The research and development program for the safe, permanent disposal of Canada's nuclear fuel wastes has been established. This report, the first of a series of annual reports, reviews in general terms the progress which has been achieved. After briefly reviewing the rationale and organization of the program, the report summarizes activities in the area of public information, used fuel storage and transportation, immobilization of used fuel and fuel reprocessing wastes, research and development associated with deep underground disposal, and environmental and safety assessment. (auth) [fr

Dry Process Fuel Performance Evaluation

Energy Technology Data Exchange (ETDEWEB)

Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

2005-04-15

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-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 II are summarized as follows : - 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 (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

Dry Process Fuel Performance Evaluation

International Nuclear Information System (INIS)

Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

2005-04-01

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-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 II are summarized as follows : - 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 (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

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.

Regulatory activities in the area of fuel safety and performance

International Nuclear Information System (INIS)

Viktorov, A.; Couture, M.

2005-01-01

Generic Action Item 94G02 'Impact of Fuel Bundle Condition on Reactor Safety' in many ways determined the present priorities in regulatory activities related to fuel performance. As one of the closure criteria it required that all licensees establish 'an effective formal and systematic process for integrating fuel design, fuel and channel inspection, laboratory examination, research, operating limits and safety analysis'. To date, such a process has been, to a large extent, put in place by all licensees. To assure that such processes remain operational and effective after the GAI closure, CNSC required, through S-99, to report annually on fuel performance and major activities in the fuel safety area. The scope of reported information has been defined to allow CNSC staff evaluation of key events and trends in fuel performance. To compliment reporting by the industry, CNSC staff has conducted targeted inspections of fuel compliance programs at all sites. Combined together, these activities provide the regulator with the confidence that CANDU fuel is robust and operates with safety margins. The scrutiny, to which fuel performance has been subjected lately, has allowed identification of certain programmatic weaknesses and gaps in the knowledge concerning the fuel behaviour under various conditions. It has become apparent that top-level strategies for assessment of fuel performance may have been inadequate and far from systematic; fuel inspection practices and capabilities have varied significantly from site to site; certain issues were identified but remained unaddressed for significant time; priorities in experimental or design support activities were not assigned consistently. The presentation gives examples of areas where, in the opinion of the CNSC staff, further work is required to support fuel design and safety envelopes. The implementation of new CANFLEX fuel designs is currently being considered by the industry and CNSC staff has been engaged in the review

Safety analysis of MOX fuels by fuel performance code

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

2013 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

Energy Technology Data Exchange (ETDEWEB)

none,

2013-12-01

The 2013 Annual Progress Report summarizes fiscal year 2013 activities and accomplishments by projects funded by the DOE Hydrogen Program. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing; technology validation; safety, codes and standards; market transformation; and systems analysis.

2014 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

Energy Technology Data Exchange (ETDEWEB)

none,

2014-11-01

The 2014 Annual Progress Report summarizes fiscal year 2014 activities and accomplishments by projects funded by the DOE Hydrogen Program. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing; technology validation; safety, codes and standards; market transformation; and systems analysis.

2011 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

Energy Technology Data Exchange (ETDEWEB)

Satyapal, Sunita [Office of Energy Efficiency and Renewable Energy (EERE), Washington, DC (United States)

2011-11-01

The 2011 Annual Progress Report summarizes fiscal year 2011 activities and accomplishments by projects funded by the DOE Hydrogen Program. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing; technology validation; safety, codes and standards; education; market transformation; and systems analysis.

Performance and fuel cycle cost analysis of one Janus 30 conceptual design for several fuel element design options

Energy Technology Data Exchange (ETDEWEB)

Nurdin, Martias [Research Centre for Nuclear Techniques, National Atomic Energy Agency (Indonesia); Matos, J E; Freese, K E [RERTR Program, Argonne National Laboratory (United States)

1983-09-01

The performance and fuel cycle costs for a 25 MW, JANUS 30 reactor conceptual design by INTERATOM, Federal Republic of Germany, for BATAN, Republic of Indonesia have been studied using 19.75% enriched uranium in four fuel element design options. All of these fuel element designs have either been proposed by INTERATOM for various reactors or are currently in use with 93% enriched uranium in reactors in the Federal Republic of Germany. Aluminide, oxide, and silicide fuels were studied for selected designs using the range of uranium densities that are either currently qualified or are being developed and demonstrated internationally. These uranium densities include 1.7-2.3 g/cm{sup 3} in aluminide fuel, 1.7-3.2 g/cm{sup 3} in oxide fuel, and 2.9-6.8 g/cm{sup 3} in silicide fuel. As of November 1982) both the aluminide and the oxide fuels with about 1.7 g U/cm{sup 3} are considered to be fully-proven for licensing purposes. Irradiation screening and proof testing of fuels with uranium densities greater than 1.7 g/cm{sup 3} are currently in progress, and these tests need to be completed in order to obtain licensing authorization for routine reactor use. To assess the long-term fuel adaptation strategy as well as the present fuel acceptance, reactor performance and annual fuel cycle costs were computed for seventeen cases based on a representative end-of-cycle excess reactivity and duty factor. In addition, a study was made to provide data for evaluating the trade-off between the increased safety associated with thicker cladding and the economic penalty due to increased fuel consumption. (author)

Fuel performance experience

International Nuclear Information System (INIS)

Sofer, G.A.

1986-01-01

The history of LWR fuel supply has been characterized by a wide range of design developments and fuel cycle cost improvements. Exxon Nuclear Company, Inc. has pursued an aggressive fuel research and development program aimed at improved fuel performance. Exxon Nuclear has introduced many design innovations which have improved fuel cycle economics and operating flexibility while fuel failures remain at very low levels. The removable upper tie plate feature of Exxon Nuclear assemblies has helped accelerate this development, enabling repeated inspections during successive plant outages. Also, this design feature has made it possible to repair damaged fuel assemblies during refueling outages, thereby minimizing the economic impact of fuel failure from all causes

Fuel cells for transportation program: FY1997 national laboratory annual report

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

The Department of Energy (DOE) Fuel Cells for Transportation Program is structured to effectively implement the research and development (R and D) required for highly efficient, low or zero emission fuel cell power systems to be a viable replacement for the internal combustion engine in automobiles. The Program is part of the Partnership for a New Generation of Vehicles (PNGV), a government-industry initiative aimed at development of an 80 mile-per-gallon vehicle. This Annual Report summarizes the technical accomplishments of the laboratories during 1997. Participants include: Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Lawrence Berkeley National Laboratory (LBNL), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), Pacific Northwest National Laboratory (PNNL), and the National Renewable Energy Laboratory (NREL). During 1997, the laboratory R and D included one project on solid oxide fuel cells; this project has since been terminated to focus Department resources on PEM fuel cells. The technical component of this report is divided into five key areas: fuel cell stack research and development; fuel processing; fuel cell modeling, testing, and evaluation; direct methanol PEM fuel cells; and solid oxide fuel cells.

TACO: fuel pin performance analysis

International Nuclear Information System (INIS)

Stoudt, R.H.; Buchanan, D.T.; Buescher, B.J.; Losh, L.L.; Wilson, H.W.; Henningson, P.J.

1977-08-01

The thermal performance of fuel in an LWR during its operational lifetime must be described for LOCA analysis as well as for other safety analyses. The determination of stored energy in the LOCA analysis, for example, requires a conservative fuel pin thermal performance model that is capable of calculating fuel and cladding behavior, including the gap conductance between the fuel and cladding, as a function of burnup. The determination of parameters that affect the fuel and cladding performance, such as fuel densification, fission gas release, cladding dimensional changes, fuel relocation, and thermal expansion, should be accounted for in the model. Babcock and Wilcox (B and W) has submitted a topical report, BAW-10087P, December 1975, which describes their thermal performance model TACO. A summary of the elements that comprise the TACO model and an evaluation are presented

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)

Fuel management for off-load annual refuelling of the D-HHT 600 MW(e) reference core

Energy Technology Data Exchange (ETDEWEB)

Hansen, U

1973-03-16

The reference design for the Dragon-HHT reactor has been optimised for on-load continuous refuelling. The possiblity to operate the reactor on a discontinuous annual reloading schedule might prove of interest and/or necessity. In this paper the influence of an annual 4-batch fuel management scheme on the core physics and fuel cycle economics is investigated. The results of the present investigation give a good indication of the relative merits of the two fuel management schemes. Although a broader parameter survey and a more detailed scrutinising of special cases would be desirable, we feel that the main conclusions are correct and that the principle differences have been elicited.

40 CFR 80.604 - What are the annual reporting requirements for refiners and importers of NRLM diesel fuel?

Science.gov (United States)

2010-07-01

... Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel... requirements for refiners and importers of NRLM diesel fuel? 80.604 Section 80.604 Protection of Environment... importers of NRLM diesel fuel? Beginning with the annual compliance period that begins June 1, 2007, or the...

Handbook of fuel cell performance

Energy Technology Data Exchange (ETDEWEB)

Benjamin, T.G.; Camara, E.H.; Marianowski, L.G.

1980-05-01

The intent of this document is to provide a description of fuel cells, their performances and operating conditions, and the relationship between fuel processors and fuel cells. This information will enable fuel cell engineers to know which fuel processing schemes are most compatible with which fuel cells and to predict the performance of a fuel cell integrated with any fuel processor. The data and estimates presented are for the phosphoric acid and molten carbonate fuel cells because they are closer to commercialization than other types of fuel cells. Performance of the cells is shown as a function of operating temperature, pressure, fuel conversion (utilization), and oxidant utilization. The effect of oxidant composition (for example, air versus O/sub 2/) as well as fuel composition is examined because fuels provided by some of the more advanced fuel processing schemes such as coal conversion will contain varying amounts of H/sub 2/, CO, CO/sub 2/, CH/sub 4/, H/sub 2/O, and sulfur and nitrogen compounds. A brief description of fuel cells and their application to industrial, commercial, and residential power generation is given. The electrochemical aspects of fuel cells are reviewed. The phosphoric acid fuel cell is discussed, including how it is affected by operating conditions; and the molten carbonate fuel cell is discussed. The equations developed will help systems engineers to evaluate the application of the phosphoric acid and molten carbonate fuel cells to commercial, utility, and industrial power generation and waste heat utilization. A detailed discussion of fuel cell efficiency, and examples of fuel cell systems are given.

Advanced fuel technology and performance

International Nuclear Information System (INIS)

1985-10-01

The purpose of the Advisory Group Meeting on Advanced Fuel Technology and Performance was to review the experience of advanced fuel fabrication technology, its performance, peculiarities of the back-end of the nuclear fuel cycle with regard to all types of reactors and to outline the future trends. As a result of the meeting recommendations were made for the future conduct of work on advanced fuel technology and performance. A separate abstract was prepared for each of the 20 papers in this issue

Nuclear fuel elements design, fabrication and performance

CERN Document Server

Frost, Brian R T

1982-01-01

Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

Fuel performance, design and development

International Nuclear Information System (INIS)

Prasad, P.N.; Tripathi, Rahul Mani; Soni, Rakesh; Ravi, M.; Vijay Kumar, S.; Dwivedi, K.P.; Pandarinathan, P.R.; Neema, L.K.

2006-01-01

The normal fuel configurations for operating 220 MWe and 540 MWe PHWRs are natural uranium dioxide 19-element and 37- element fuel bundle types respectively. The fuel configuration for BWRs is 6 x 6 fuel. So far, about 330 thousand PHWR fuel bundles and 3500 number of BWR bundles have been irradiated in the 14 PHWRs and 2 BWRs. Improvements in fuel design, fabrication, quality control and operating practices are continuously carried out towards improving fuel utilization as well as reducing fuel failure rate. Efforts have been put to improve the fuel bundle utilization by increasing the fuel discharge burnup of the natural uranium bundles The overall fuel failure rate currently is less than 0.1 % . Presently the core discharge burnups in different reactors are around 7500 MWD/TeU. The paper gives the fuel performance experience over the years in the different power reactors and actions taken to improve fuel performance over the years. (author)

2015 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

none,

2015-10-01

This report summarizes comments from the Peer Review Panel at the 2015 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on June 8-12, 2015, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; market transformation; and systems analysis.

2011 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

none,

2011-09-01

This report summarizes comments from the Peer Review Panel at the 2011 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on May 9-13, 2011, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; education; market transformation; and systems analysis.

2013 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

none,

2013-10-01

This report summarizes comments from the Peer Review Panel at the 2013 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on May 13-17, 2013, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; market transformation; and systems analysis.

2014 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

none,

2014-10-01

This report summarizes comments from the Peer Review Panel at the 2014 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on June 16-20, 2014, in Washington, DC. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; market transformation; and systems analysis.

2012 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

none,

2012-09-01

This report summarizes comments from the Peer Review Panel at the 2012 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on May 14-18, 2012, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; education; market transformation; and systems analysis.

Low-enriched fuel particle performance review

International Nuclear Information System (INIS)

Homan, F.; Nabielek, H.; Yang, L.

1978-08-01

The available data on low-enriched uranium (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance, and fuel particle chemical performance

Proceedings of the 2006 International Meeting on LWR fuel performance 'Nuclear Fuel: Addressing the future' - TopFuel 2006 Transactions

International Nuclear Information System (INIS)

2006-01-01

From 22-26 October, 340 researchers, nuclear engineers and scientists from across Europe and beyond congregated in the ancient university city of Salamanca, Spain, to discuss the challenges facing the developers and manufacturers of new high-performance nuclear fuels-fuels that will help meet current and future energy demand and reduce man's over dependence upon CO 2 -emitting fossil fuels. TopFuel is an annual topical meeting organised by ENS, the American Nuclear Society and the Atomic Energy Society of Japan. This year it was co-sponsored by the IAEA, the OECD/NEA and the Spanish Nuclear Society (SNE). TopFuel's primary objective was to bring together leading specialists in the field from around the world to analyse advances in nuclear fuel management technology and to use the findings of the latest cutting-edge research to help manufacture the high performance nuclear fuels of today and tomorrow. The TopFuel 2006 agenda revolved around ten technical sessions dedicated to priority issues such as security of supply, new fuel and reactor core designs, fuel cycle strategies and spent fuel management. Among the many topics under discussion were new developments in fuel performance modelling, advanced fuel assembly design and the improved conditioning and processing of spent fuel. During the week, a poster exhibition also gave delegates the opportunity to display and discuss the results of their latest work and to network with fellow professionals. One important statement to emerge from TopFuel 2006 was that the world has enough reserves of uranium to support the large-scale and long-term production of nuclear energy. The OECD/NEA and the IAEA recently published a report entitled Uranium 2005: Resources, Production and Demand (the Red Book). The report, which makes a comprehensive assessment of uranium supplies and projected demand up until the year 2025, concludes by saying 'the uranium resource base is adequate to meet projected future requirements'. With the

Annual Performance Report - FY 2011

Science.gov (United States)

This report summarizes OIG activity, performance, results, and challenges, and provides a financial accounting of resources for fiscal year (FY) 2011 compared to our FY 2011 annual performance targets.

High energy-density liquid rocket fuel performance

Science.gov (United States)

Rapp, Douglas C.

1990-01-01

A fuel performance database of liquid hydrocarbons and aluminum-hydrocarbon fuels was compiled using engine parametrics from the Space Transportation Engine Program as a baseline. Propellant performance parameters are introduced. General hydrocarbon fuel performance trends are discussed with respect to hydrogen-to-carbon ratio and heat of formation. Aluminum-hydrocarbon fuel performance is discussed with respect to aluminum metal loading. Hydrocarbon and aluminum-hydrocarbon fuel performance is presented with respect to fuel density, specific impulse and propellant density specific impulse.

Irradiation performance of metallic fuels

International Nuclear Information System (INIS)

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

1989-01-01

Argonne National Laboratory has been working for the past five years to develop and demonstrate the Integral Fast Reactor (IFR) concept. The concept involves a closed system for fast-reactor power generation and on-site fuel reprocessing, both designed specifically around the use of metallic fuel. The Experimental Breeder Reactor-II (EBR-II) has used metallic fuel for all of its 25-year life. In 1985, tests were begun to examine the irradiation performance of advanced-design metallic fuel systems based on U-Zr or U-Pu-Zr fuels. These tests have demonstrated the viable performance of these fuel systems to high burnup. The initial testing program will be described in this paper. 2 figs

Fuel performance analysis code 'FAIR'

International Nuclear Information System (INIS)

Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

1994-01-01

For modelling nuclear reactor fuel rod behaviour of water cooled reactors under severe power maneuvering and high burnups, a mechanistic fuel performance analysis code FAIR has been developed. The code incorporates finite element based thermomechanical module, physically based fission gas release module and relevant models for modelling fuel related phenomena, such as, pellet cracking, densification and swelling, radial flux redistribution across the pellet due to the build up of plutonium near the pellet surface, pellet clad mechanical interaction/stress corrosion cracking (PCMI/SSC) failure of sheath etc. The code follows the established principles of fuel rod analysis programmes, such as coupling of thermal and mechanical solutions along with the fission gas release calculations, analysing different axial segments of fuel rod simultaneously, providing means for performing local analysis such as clad ridging analysis etc. The modular nature of the code offers flexibility in affecting modifications easily to the code for modelling MOX fuels and thorium based fuels. For performing analysis of fuel rods subjected to very long power histories within a reasonable amount of time, the code has been parallelised and is commissioned on the ANUPAM parallel processing system developed at Bhabha Atomic Research Centre (BARC). (author). 37 refs

CANDU reactor experience: fuel performance

International Nuclear Information System (INIS)

Truant, P.T.; Hastings, I.J.

1985-07-01

Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

The Fuel Performance Analysis of LWR Fuel containing High Thermal Conductivity Reinforcements

International Nuclear Information System (INIS)

Kim, Seung Su; Ryu, Ho Jin

2015-01-01

The thermal conductivity of fuel affects many performance parameters including the fuel centerline temperature, fission gas release and internal pressure. In addition, enhanced safety margin of fuel might be expected when the thermal conductivity of fuel is improved by the addition of high thermal conductivity reinforcements. Therefore, the effects of thermal conductivity enhancement on the fuel performance of reinforced UO2 fuel with high thermal conductivity compounds should be analyzed. In this study, we analyzed the fuel performance of modified UO2 fuel with high thermal conductivity reinforcements by using the FRAPCON-3.5 code. The fissile density and mechanical properties of the modified fuel are considered the same with the standard UO2 fuel. The fuel performance of modified UO2 with high thermal conductivity reinforcements were analyzed by using the FRAPCON-3.5 code. The thermal conductivity enhancement factors of the modified fuels were obtained from the Maxwell model considering the volume fraction of reinforcements

Water Reactor Fuel Performance Meeting 2008

International Nuclear Information System (INIS)

2008-10-01

This meeting contains articles of the Water Reactor Fuel Performance Meeting 2008 of Korean Nuclear Society, Atomic Energy Society of Japan, Chinese Nuclear Society, European Nuclear Society and American Nuclear Society. It was held on Oct. 19-23, 2008 in Seoul, Korea and subject of Meeting is 'New Clear' Fuel - A green energy solution. This proceedings is comprised of 5 tracks. The main topic titles of track are as follows: Advances in water reactor fuel technology, Fuel performance and operational experience, Transient fuel behavior and safety-related issues, Fuel cycle, spent fuel storage and transportations and Fuel modeling and analysis. (Yi, J. H.)

HTGR Fuel performance basis

International Nuclear Information System (INIS)

Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

1982-05-01

The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents

Low-enriched fuel particle performance review

International Nuclear Information System (INIS)

Homan, F.; Nabielek, H.; Yang, L.

1978-08-01

The available data on low-enriched (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance and fuel particle chemical performande. (orig.) [de

Structure of fuel performance audit code for SFR metal fuel

Energy Technology Data Exchange (ETDEWEB)

Yang, Yong Sik; Kim, Hyo Chan [KAERI, Daejeon (Korea, Republic of); Jeong, Hye Dong; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2012-10-15

A Sodium Cooled Fast Reactor (SFR) is a promising option to solve the spent fuel problems, but, there are still much technical issues to commercialize a SFR. One of issues is a development of advanced fuel which can solve the safety and the economic issues at the same time. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured. In Korea Institute of Nuclear Safety (KINS), the new project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. To develop the new code system, the code structure design and its requirements need to be studied. Various performance models and code systems are reviewed and their characteristics are analyzed in this paper. Based on this study, the fundamental performance models are deduced and basic code requirements and structure are established.

Fuel performance-REP, Seminars on nuclear fuel performance based on basic underlining phenomena, proceedings

International Nuclear Information System (INIS)

2008-01-01

Description: The need for further improving the understanding of basic phenomena underlying nuclear fuel behaviour has been recognised both by fuel vendors, experts in fuel research in the different laboratories and committees and working groups coordinating international activities. The OECD/NEA Nuclear Science Committee has established an Experts Group addressing this issue. This has led to establishing an International Fuel Performance Experiments Database (IFPE) that should help model evaluation and validation. Many years ago the IAEA established an International Working Group on Fuel Performance and Technology (IWGFPT) that led to the FUMEX-I and FUMEX-II (Fuel Modelling Exercise) which has had an important impact on code improvements. Both international organisations, with the support of national organisations, co-operate in establishing and maintaining the Database and to build confidence in the predictive power of the models through international comparison exercises. But above all the different parties have agreed that seminars focussed on specific phenomena would be beneficial to exchange current knowledge, identify outstanding problems and agree on common action that would lead to improved understanding of the phenomena. A series of three seminars has been initiated by the Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), Framatome and Cogema under the aegis of the OECD/NEA and the IAEA. 1. Thermal Performance of High Burn-Up LWR Fuel at Cadarache, France, from 3 to 6 of March 1998. Thermal performance occupies the most important aspect of the fuel performance modelling. Not only is it extremely important from a safety point of view, but also many of the material properties of interest and behaviour, such as transport properties like fuel creep and fission gas release are thermally activated processes. Thus, in order to model these processes correctly, it is critical to calculate temperatures and their distribution as accurately as

Fuel performance of DOE fuels in water storage

International Nuclear Information System (INIS)

Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

1993-01-01

Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory. In April of 1992, the U.S. Department of Energy (DOE) decided to end the fuel reprocessing mission at ICPP. Fuel performance in storage received increased emphasis as the fuel now needs to be stored until final dispositioning is defined and implemented. Fuels are stored in four main areas: an original underwater storage facility, a modern underwater storage facility, and two dry fuel storage facilities. As a result of the reactor research mission of the DOE and predecessor agencies, the Energy Research and Development Administration and the Atomic Energy Commission, many types of nuclear fuel have been developed, used, and assigned to storage at the ICPP. Fuel clad with stainless steel, zirconium, aluminum, and graphite are represented. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels, resulting in 55 different fuel types in storage. Also included in the fuel storage inventory is canned scrap material

Nondestructive analysis of irradiated fuels

International Nuclear Information System (INIS)

Dudey, N.D.; Frick, D.C.

1977-01-01

The principal nondestructive examination techniques presently used to assess the physical integrity of reactor fuels and cladding materials include gamma-scanning, profilometry, eddy current, visual inspection, rod-to-rod spacing, and neutron radiography. LWR fuels are generally examined during annual refueling outages, and are conducted underwater in the spent fuel pool. FBR fuels are primarily examined in hot cells after fuel discharge. Although the NDE techniques are identical, LWR fuel examinations emphasize tests to demonstrate adherence to technical specification and reliable fuel performance; whereas, FBR fuel examinations emphasize aspects more related to the relative performance of different types of fuel and cladding materials subjected to variable irradiation conditions

DOE Hydrogen and Fuel Cells Program 2017 Annual Merit Review and Peer Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

None, None

2017-10-16

The fiscal year 2017 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting (AMR), in conjunction with DOE's Vehicle Technologies Office AMR, was held from June June 5-9, 2017, in Washington, D.C. This report is a summary of comments by AMR peer reviewers about the hydrogen and fuel cell projects funded by DOE's Office of Energy Efficiency and Renewable Energy.

DOE Hydrogen and Fuel Cells Program 2016 Annual Merit Review and Peer Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

None, None

2016-11-01

The fiscal year 2016 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting (AMR), in conjunction with DOE's Vehicle Technologies Office AMR, was held from June 6-10, 2016, in Washington, D.C. This report is a summary of comments by AMR peer reviewers about the hydrogen and fuel cell projects funded by DOE's Office of Energy Efficiency and Renewable Energy.

Nitride fuels irradiation performance data base

International Nuclear Information System (INIS)

Brozak, D.E.; Thomas, J.K.; Peddicord, K.L.

1987-01-01

An irradiation performance data base for nitride fuels has been developed from an extensive literature search and review that emphasized uranium nitride, but also included performance data for mixed nitrides [(U,Pu)N] and carbonitrides [(U,Pu)C,N] to increase the quantity and depth of pin data available. This work represents a very extensive effort to systematically collect and organize irradiation data for nitride-based fuels. The data base has many potential applications. First, it can facilitate parametric studies of nitride-based fuels to be performed using a wide range of pin designs and operating conditions. This should aid in the identification of important parameters and design requirements for multimegawatt and SP-100 fuel systems. Secondly, the data base can be used to evaluate fuel performance models. For detailed studies, it can serve as a guide to selecting a small group of pin specimens for extensive characterization. Finally, the data base will serve as an easily accessible and expandable source of irradiation performance information for nitride fuels

Annual report 2003

International Nuclear Information System (INIS)

2003-01-01

Delivering products and services to nuclear power plants operators, AREVA operates in every sector of the civilian nuclear power and fuel cycle industry. This annual report 2003 provides, in seven chapters, information on persons responsible for the annual report and for auditing the financial statements, general information on the company and share capital (statute, capital, share trading, dividends), information on company operations, changes and future prospects, assets, financial position and financial performance, corporate governance, recent developments and future prospects. (A.L.B.)

Fuels Performance | Transportation Research | NREL

Science.gov (United States)

chemistry, combustion, and emissions to answer two primary questions: How can improved fuel chemistry lead initiative Combustion and engines Fuel chemistry Emissions and fuel economy Find out about related NREL Chemistry. (2014) Saturated Monoglyceride Effects on Low-Temperature Performance of Biodiesel Blends. G.M

Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

International Nuclear Information System (INIS)

Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

2005-10-01

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor

Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

2005-10-15

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.

The UK fuel poverty strategy: 5th annual progress report 2007

International Nuclear Information System (INIS)

2007-12-01

This fifth annual progress report details government progress in 2007 in tackling fuel poverty and movement towards targets. The United Kingdom were the first country in the world to recognise the issue of fuel poverty and to put in place measures to tackle the issue, including spending 20 billion pounds sterling on benefits and programmes since the year 2000. The report covers progress to date, schemes and initiatives to tackle fuel poverty, the energy market and looks ahead to the future. Progress and development of the schemes across the devolved nations are also considered. This report is the first to publish the Government's proposals for the Carbon Emissions Reduction Target (CERT) priority group, which were laid before Parliament on 5th December 2007. This report is the first to present the fuel poverty figures for 2005, and shows the effects of rising energy prices. The Government continues to take action to ensure that the energy market is working properly, and to encourage reform in the EU on energy market liberalisation - this should reduce pressure on prices. Those in fuel poverty have much to gain by switching supplier and this report outlines the action taken by Ofgem and Energywatch to encourage this

Review on Fuel Loading Process and Performance for Advanced Fuel Handling Equipment

International Nuclear Information System (INIS)

Chang, Sang-Gyoon; Lee, Dae-Hee; Kim, Young-Baik; Lee, Deuck-Soo

2007-01-01

The fuel loading process and the performance of the advanced fuel handling equipment for OPR 1000 (Optimized Power Plant) are analyzed and evaluated. The fuel handling equipment, which acts critical processes in the refueling outage, has been improved to reduce fuel handling time. The analysis of the fuel loading process can be a useful tool to improve the performance of the fuel handling equipment effectively. Some recommendations for further improvement are provided based on this study

Alternate-Fueled Combustor-Sector Performance

Science.gov (United States)

Thomas, Anna E.; Saxena, Nikita T.; Shouse, Dale T.; Neuroth, Craig; Hendricks, Robert C.; Lynch, Amy; Frayne, Charles W.; Stutrud, Jeffrey S.; Corporan, Edwin; Hankins, Terry

2013-01-01

In order to realize alternative fueling for military and commercial use, the industry has set forth guidelines that must be met by each fuel. These aviation fueling requirements are outlined in MIL-DTL-83133F(2008) or ASTM D 7566 Annex (2011) standards, and are classified as "drop-in" fuel replacements. This report provides combustor performance data for synthetic-paraffinic-kerosene- (SPK-) type (Fischer-Tropsch (FT)) fuel and blends with JP-8+100, relative to JP-8+100 as baseline fueling. Data were taken at various nominal inlet conditions: 75 psia (0.52 MPa) at 500 degF (533 K), 125 psia (0.86 MPa) at 625 degF (603 K), 175 psia (1.21 MPa) at 725 degF (658 K), and 225 psia (1.55 MPa) at 790 degF (694 K). Combustor performance analysis assessments were made for the change in flame temperatures, combustor efficiency, wall temperatures, and exhaust plane temperatures at 3, 4, and 5 percent combustor pressure drop (DP) for fuel:air ratios (F/A) ranging from 0.010 to 0.025. Significant general trends show lower liner temperatures and higher flame and combustor outlet temperatures with increases in FT fueling relative to JP-8+100 fueling. The latter affects both turbine efficiency and blade and vane lives.

Irradiation test and performance evaluation of DUPIC fuel

International Nuclear Information System (INIS)

Yang, Myung Seung; Song, K. C.; Moon, J. S.

2002-05-01

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-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 II are summarized as follows : - 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 (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

Surveillance of WWER-440 fuel performance

International Nuclear Information System (INIS)

Simko, J.; Urban, P.

1999-01-01

In this lecture next problems of surveillance of WWER-440 fuel performance are presented: surveillance of WWER-440 fuel performance at Mochovce NPP; basic data of WWER-440 reactor; in-core reactor measuring system 'SVRK'; basic level of SVRK; information output of basic level of SVRK; surveillance of fuel performance; table of permissible operation conditions of the reactor; limitation of the unit 1 power at the beginning of the operation; cyclic changes of power; future perspectives

MONJU fuel pin performance analysis

International Nuclear Information System (INIS)

Kitagawa, H.; Yamanaka, T.; Hayashi, H.

1979-01-01

Monju fuel pin has almost the same properties as other LMFBR fuel pins, i.e. Phenix, PFR, CRBR, but would be irradiated under severe conditions: maximum linear heat rate of 381 watt/cm, hot spot cladding temperature of 675 deg C, peak burnup of 131,000 MWd/t, peak fluence (E greater than 0.1 MeV) of 2.3 10 23 n/cm 2 . In order to understand in-core performance of Monju fuel pin, its thermal and mechanical behaviour was predicted using the fast running performance code SIMPLE. The code takes into account pellet-cladding interaction due to thermal expansion and swelling, gap conductance, structural changes of fuel pellets, fission product gas release with burnup and temperature increase, swelling and creep of fuel pellets, corrosion of cladding due to sodium flow and chemical attack by fission products, and cumulative damage of the cladding due to thermal creep

Reactor fuel performance data file, 1985 edition

International Nuclear Information System (INIS)

Harayama, Yasuo; Fujita, Misao; Watanabe, Kohji.

1986-07-01

In safety evaluation and integrity studies of reactor fuel, data on fuel performance are the most basic materials. The Fuel Reliability Laboratory No.1 has obtained the fuel performance data by joining in some international programs to study the safety and integrity of fuel. Those data have only used for the studies in the above two fields. However, if the data are rearranged and compiled in a easily usable form, they can be utilized in other field of studies. Then, a 'data file' on fuel performance is beeing compiled by adding data from open literatures to those obtained in international programs. The present report is prepared on the basis of the data file compiled by March in 1986. (author)

Commercial waste and spent fuel packaging program. Annual report

International Nuclear Information System (INIS)

Hakl, A.R.

1981-10-01

This document is a report of activities performed by Westinghouse Advanced Energy Systems Division - Nevada Operations in meeting subtask objectives described in the Nevada Nuclear Waste Storage Investigations (NNWSI) Project Plan and revised planning documentation for Fiscal Year (FY) 1981. Major activities included: completion of the first fuel exchange in the Spent Fuel Test - Climax program; plasma arc welder development; modification and qualification of a canister cutter; installation, and activation of a remote area monitor, constant air monitor and an alpha/beta/gamma counting system; qualification of grapples required to handle pressurized water reactor or boiling water reactor fuel and high level waste (HLW) logs; data acquisition from the 3 kilowatt soil temperature test, 2 kw fuel temperature test, and 2 kw drywell test; calorimetry of the fuel assembly used in the fuel temperature test; evaluation of moisture accumulation in the drywells and recommendations for proposed changes; revision of safety assessment document to include HLW log operations; preparation of quality assurance plan and procedures; development and qualification of all equipment and procedures to receive, handle and encapsulate both the HLW log and spent fuel for the basalt waste isolation program/near surface test facility program; preliminary studies of both the requirements to perform waste packaging for the test and evaluation facility and a cask storage program for the DOE Interim Spent Fuel Management program; and remote handling operations on radioactive source calibration in support of other contractors

U.S. report on fuel performance and technology

Energy Technology Data Exchange (ETDEWEB)

Cook, T [Department of Energy, Washington, DC (United States). Office of Engineering and Technology Development

1997-12-01

The report reviews the following aspects of fuel performance and technology: increased demand on fuel performance;improved fuel failure rate; operating fuel cycles; capacity factor for US nuclear electric generating plants; potential reduction of SNF due to improved fuel burnup.

High Burnup Fuel Performance and Safety Research

Energy Technology Data Exchange (ETDEWEB)

Bang, Je Keun; Lee, Chan Bok; Kim, Dae Ho (and others)

2007-03-15

The worldwide trend of nuclear fuel development is to develop a high burnup and high performance nuclear fuel with high economies and safety. Because the fuel performance evaluation code, INFRA, has a patent, and the superiority for prediction of fuel performance was proven through the IAEA CRP FUMEX-II program, the INFRA code can be utilized with commercial purpose in the industry. The INFRA code was provided and utilized usefully in the universities and relevant institutes domesticallly and it has been used as a reference code in the industry for the development of the intrinsic fuel rod design code.

Nuclear fuel performance in boiling water reactors

International Nuclear Information System (INIS)

Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

1981-01-01

A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs

Code structure for U-Mo fuel performance analysis in high performance research reactor

Energy Technology Data Exchange (ETDEWEB)

Jeong, Gwan Yoon; Cho, Tae Won; Lee, Chul Min; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

A performance analysis modeling applicable to research reactor fuel is being developed with available models describing fuel performance phenomena observed from in-pile tests. We established the calculation algorithm and scheme to best predict fuel performance using radio-thermo-mechanically coupled system to consider fuel swelling, interaction layer growth, pore formation in the fuel meat, and creep fuel deformation and mass relocation, etc. In this paper, we present a general structure of the performance analysis code for typical research reactor fuel and advanced features such as a model to predict fuel failure induced by combination of breakaway swelling and pore growth in the fuel meat. Thermo-mechanical code dedicated to the modeling of U-Mo dispersion fuel plates is being under development in Korea to satisfy a demand for advanced performance analysis and safe assessment of the plates. The major physical phenomena during irradiation are considered in the code such that interaction layer formation by fuel-matrix interdiffusion, fission induced swelling of fuel particle, mass relocation by fission induced stress, and pore formation at the interface between the reaction product and Al matrix.

Irradiation performance of HTGR recycle fissile fuel

International Nuclear Information System (INIS)

Homan, F.J.; Long, E.L. Jr.

1976-08-01

The irradiation performance of candidate HTGR recycle fissile fuel under accelerated testing conditions is reviewed. Failure modes for coated-particle fuels are described, and the performance of candidate recycle fissile fuels is discussed in terms of these failure modes. The bases on which UO 2 and (Th,U)O 2 were rejected as candidate recycle fissile fuels are outlined, along with the bases on which the weak-acid resin (WAR)-derived fissile fuel was selected as the reference recycle kernel. Comparisons are made relative to the irradiation behavior of WAR-derived fuels of varying stoichiometry and conclusions are drawn about the optimum stoichiometry and the range of acceptable values. Plans for future testing in support of specification development, confirmation of the results of accelerated testing by real-time experiments, and improvement in fuel performance and reliability are described

Semi-annual status report of the Canadian Nuclear Fuel Waste Management Program, April 1--September 30, 1991

Energy Technology Data Exchange (ETDEWEB)

Wright, E. D. [comp.

1992-02-01

This report is the eleventh in a series of semi-annual status reports on the research and development program for the safe management and disposal of Canada's nuclear fuel waste. it describes progress achieved in the three major subprograms, engineered systems, natural systems and performance assessment, from 1991 April 1 to September 30. It also gives a brief description of the activities being carried out in preparation for the public and governmental review of the disposal concept. Since 1987, this program has been jointly funded by AECL and Ontario Hydro under the auspices of the CANDU Owners Group (COG).

HTGR fuel performance basis

International Nuclear Information System (INIS)

Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

1982-01-01

The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents. The slow release of fission products over hundreds of hours allows for decay of short-lived isotopes. The slow and limited release of fission products under HTGR accident conditions results in very low off-site doses. The slow nature of the accident provides more time for operator action to mitigate the accident and for local and state authorities to respond. These features can be used to take advantage of close-in siting for process applications, flexibility in site selection, and emergency planning

Design and Performance of LPG Fuel Mixer for Dual Fuel Diesel Engine

Science.gov (United States)

Desrial; Saputro, W.; Garcia, P. P.

2018-05-01

Small horizontal diesel engines are commonly used for agricultural machinery, however, availability of diesel fuel become one of big problems especially in remote area. Conversely, in line with government policy for conversion of kerosene into LPG for cooking, then LPG become more popular and available even in remote area. Therefore, LPG is potential fuel to replace the shortage of diesel fuel for operating diesel engine in remote area. The purpose of this study was to design mixing device for using dual fuel i.e. LPG and diesel fuel and evaluate its performance accordingly. Simulation by using CFD was done in order to analyze mixture characteristics of LPG in air intake manifold. The performance test was done by varying the amount of LPG injected in intake air at 20%, 25%, 30%, 35%, until 40%, respectively. Result of CFD contour simulation showed the best combination when mixing 30% LPG into the intake air. Performance test of this research revealed that mixing LPG in air intake can reduce the diesel fuel consumption about 0.7 l/hour (without load) and 1.14 l/hour (with load). Diesel engine revolution increases almost 300 rpm faster than when using diesel fuel only. Based on economic analysis, using the fuel combination (diesel fuel – LPG) is not recommended in the area near SPBU where the price of diesel fuel is standard. However, using the fuel combination LPG-diesel fuel is highly recommended in the remote areas in Indonesia where price of diesel fuel is comparatively expensive which will provide cheaper total fuel cost for diesel engine operation.

Delivering high performance BWR fuel reliably

International Nuclear Information System (INIS)

Schardt, J.F.

1998-01-01

Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

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

International Nuclear Information System (INIS)

THIELGES, J.R.

2000-01-01

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

Delivering high performance BWR fuel reliably

Energy Technology Data Exchange (ETDEWEB)

Schardt, J.F. [GE Nuclear Energy, Wilmington, NC (United States)

1998-07-01

Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

DUPIC fuel irradiation test and performance evaluation; the performance analysis of pellet-cladding contact fuel

Energy Technology Data Exchange (ETDEWEB)

Ho, K. I.; Kim, H. M.; Yang, K. B.; Choi, S. J. [Suwon University, Whasung (Korea)

2002-04-01

Thermal and mechanical models were reviewed, and selected for the analysis of nuclear fuel performance in reactor. 2 dimensional FEM software was developed. Thermal models-gap conductances, thermal conductivity of pellets, fission gas release, temperature distribution-were set and packaged into a software. Both thermal and mechanical models were interrelated to each other, and the final results, fuel performance during irradiation is obtained by iteration calculation. Also, the contact phenomena between pellet and cladding was analysed by mechanical computer software which was developed during this work. dimensional FEM program was developed which estimate the mechanical behavior and the thermal behaviors of nuclear fuel during irradiation. Since there is a importance during the mechanical deformation analysis in describing pellet-cladding contact phenomena, simplified 2 dimensional calculation method is used after the contact. The estimation of thermal fuel behavior during irradiation was compared with the results of other. 8 refs., 17 figs. (Author)

Irradiated fuel performance evaluation technology development

International Nuclear Information System (INIS)

Koo, Yang Hyun; Bang, J. G.; Kim, D. H.

2012-01-01

Alpha version performance code for dual-cooled annular fuel under steady state operation, so called 'DUOS', has been developed applying performance models and proposed methodology. Furthermore, nonlinear finite element module which could be integrated into transient/accident fuel performance code was also developed and evaluated using commercial FE code. The first/second irradiation and PIE test of annular pellet for dual-cooled annular fuel in the world have been completed. In-pile irradiation test DB of annular pellet up to burnup of 10,000 MWd/MTU through the 1st test was established and cracking behavior of annular pellet and swelling rate at low temperature were studied. To do irradiation test of dual-cooled annular fuel under PWR's simulating steady-state conditions, irradiation test rig/rod design/manufacture of mock-up/performance test have been completed through international collaboration program with Halden reactor project. The irradiation test of large grain pellets has been continued from 2002 to 2011 and completed successfully. Burnup of 70,000 MWd/MTU which is the highest burnup among irradiation test pellets in domestic was achieved

Transient performance of EBR-II driver fuel

International Nuclear Information System (INIS)

Buzzell, J.A.; Hudman, G.D.; Porter, D.L.

1981-01-01

The first phases of qualification of the EBR-II driver fuel for repeated transient overpower operation have recently been completed. The accomplishments include prediction of the transient fuel and cladding performance through ex-core testing and fuel-element modeling studies, localized in-core power testing during steady-state operation, and whole-core multiple transient testing. The metallic driver fuel successfully survived 56 transients, spaced over a 45-day period, with power increases of approx. 160% at rates of approx. 1%/s with a 720-second hold at full power. The performance results obtained from both ex-core and n-core tests indicate that the fuel is capable of repeated transient operation

Irradiation and performance evaluation of DUPIC fuel

International Nuclear Information System (INIS)

Bae, Ki Kwang; Yang, M. S.; Song, K. C.

2000-05-01

The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis

Irradiation and performance evaluation of DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Bae, Ki Kwang; Yang, M S; Song, K C [and others

2000-05-01

The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis.

Fuel performance evaluation for the CAFE experimental device

International Nuclear Information System (INIS)

Giovedi, Claudia; Hirota, Leandro T.

2011-01-01

Fuel rod cladding material is the second barrier to prevent the release of radioactive inventories in a PWR reactor. In this sense, an important safety aspect is to assess the fuel behavior under operational conditions. This can be made by means of fuel performance codes and confirmed by experimental measurements. In order to evaluate the fuel behavior of fuel rods in steady-state conditions, it was designed an experimental irradiation device, the Nuclear Fuel Irradiation Circuit (CAFE-Mod1). This device will allow controlling the surface rod temperature, to measure the power associated to the rod and the evolution of fission gas release for a typical PWR fuel pin. However, to support the experimental irradiation program, it is extremely important to simulate the experimental conditions using a fuel performance code. The aim of this paper is to evaluate some parameters and aspects related to the fuel rod behavior during the irradiation program. This evaluation was carried out by means of an adapted fuel performance code. Obtained results have shown that besides of the variation observed for parameters, such as, fuel temperature and fission gas release as a function of fuel enrichment level, the fuel rod integrity was preserved in all studied conditions. (author)

FFTF fuel pin design bases and performance

International Nuclear Information System (INIS)

Cox, C.M.; Hanson, J.E.; Roake, W.E.; Slember, R.J.; Weber, C.E.; Millunzi, A.C.

1975-04-01

The FFTF fuel pin was conservatively designed to meet thermal and structural performance requirements in the categories normal operation, upset events, emergency events, and hypothetical, faulted events. The fuel pin operating limits consistent with these requirements were developed from a strong fuel pin irradiation testing program scoped to define the performance capability under relevant steady state and transient conditions. Comparison of the results of the irradiation testing program with design requirements indicates that the FFTF fuel pin can exceed its goal burnup of 80,000 MWd/MTM. (U.S.)

Alternate-Fueled Combustor-Sector Performance: Part A: Combustor Performance Part B: Combustor Emissions

Science.gov (United States)

Shouse, D. T.; Neuroth, C.; Henricks, R. C.; Lynch, A.; Frayne, C.; Stutrud, J. S.; Corporan, E.; Hankins, T.

2010-01-01

Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F(2008) or ASTM D 7566 (2010) standards, respectively, and are classified as drop-in fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are certified individually on the basis of feedstock. Adherence to alternate fuels and fuel blends requires smart fueling systems or advanced fuel-flexible systems, including combustors and engines without significant sacrifice in performance or emissions requirements. This paper provides preliminary performance (Part A) and emissions and particulates (Part B) combustor sector data for synthetic-parafinic-kerosene- (SPK-) type fuel and blends with JP-8+100 relative to JP-8+100 as baseline fueling.

H2 and Fuel cell annual activity report - France 2011

International Nuclear Information System (INIS)

Lucchese, Paul; Julien, Marianne

2012-01-01

This report aims making better known the hydrogen and fuel cell technologies in France and their main actors (large groups, small and medium-sized enterprises, start-ups, research centres). After a presentation of the French energy context, it presents the national programme and strategic actions, and local programmes and initiatives. The next chapter presents the main results and events for the different fields of application: leading edge markets, transports, decentralized or residential stationary applications, hydrogen and renewable energies, portable applications, transverse domains. The annual activity and main results of the different actors are then presented: research and development, small or medium sized enterprises and start-ups, large groups

Performance of candu-6 fuel bundles manufactured in romania nuclear fuel plant

International Nuclear Information System (INIS)

Bailescu, A.; Barbu, A.; Din, F.; Dinuta, G.; Dumitru, I.; Musetoiu, A.; Serban, G.; Tomescu, A.

2013-01-01

The purpose of this article is to present the performance of nuclear fuel produced by Nuclear Fuel Plant (N.F.P.) - Pitesti during 1995 - 2012 and irradiated in units U1 and U2 from Nuclear Power Plant (N.P.P.) Cernavoda and also present the Nuclear Fuel Plant (N.F.P.) - Pitesti concern for providing technology to prevent the failure causes of fuel bundles in the reactor. This article presents Nuclear Fuel Plant (N.F.P.) - Pitesti experience on tracking performance of nuclear fuel in reactor and strategy investigation of fuel bundles notified as suspicious and / or defectives both as fuel element and fuel bundle, it analyzes the possible defects that can occur at fuel bundle or fuel element and can lead to their failure in the reactor. Implementation of modern technologies has enabled optimization of manufacturing processes and hence better quality stability of achieving components (end caps, chamfered sheath), better verification of end cap - sheath welding. These technologies were qualified by Nuclear Fuel Plant (N.F.P.) - Pitesti on automatic and Computer Numerical Control (C.N.C.) programming machines. A post-irradiation conclusive analysis which will take place later this year (2013) in Institute for Nuclear Research Pitesti (the action was initiated earlier this year by bringing a fuel bundle which has been reported defective by pool visual inspection) will provide additional information concerning potential damage causes of fuel bundles due to manufacturing processes. (authors)

Thorium fuel performance assessment in HTRs

Energy Technology Data Exchange (ETDEWEB)

Allelein, H.-J. [Forschungszentrum Jülich, D-52425 Jülich (Germany); RWTH Aachen, D-52072 Aachen (Germany); Kania, M.J.; Nabielek, H. [Forschungszentrum Jülich, D-52425 Jülich (Germany); Verfondern, K., E-mail: k.verfondern@fz-juelich.de [Forschungszentrum Jülich, D-52425 Jülich (Germany)

2014-05-01

Thorium as a nuclear fuel is receiving renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTR development employed thorium together with high-enriched uranium. After 1980, most HTR fuel systems switched to low-enriched uranium. After completing fuel development for AVR and THTR with BISO coated particles, the German program expanded efforts on a new program utilizing thorium and high-enriched uranium TRISO coated particles for advanced HTR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of LTI inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with HTI-BISO coatings. The improved performance of the HEU (Th,U)O{sub 2} TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 °C in normal operations and 1600 °C in accidents, with burnups up to 13% FIMA and fast fluences to 8 × 10{sup 25} m{sup −2} (E > 16 fJ), the results exceed the design limits on manufacturing and operational requirements for the German HTR Modul concept, which were: <6.5 × 10{sup −5} for manufacturing; <2 × 10{sup −4} for normal operating conditions; and <5 × 10{sup −4} for accident conditions. These

Fuel performance update at Japanese BWRs

International Nuclear Information System (INIS)

Otsuka, Yasuyuki; Abe, Moriyasu

2008-01-01

Fuel Performance of Japanese BWRs has shown excellency with a very low fuel failure rate over the past couple of decades. In order to eliminate debris-fretting failures, which is considered as a dominant cause, efforts have been made to work out strict FME program and introduce lower tie plate with debris filters. Regarding a measure to detect failures without delay, an online off-gas monitor is installed, and there occurred no significant secondary failures leading to unnecessary plant shutdowns thanks to the monitor. Current standard fuel designs are 9x9 lattice configuration with the design maximum burnup set at 55 GWd/t and the bundle average discharge burnup at 45 GWd/t. Satisfactory performance of many fuels has gained so far to the burnup level. However recent data show an increase in hydrogen pickup of fuel claddings and spacers in the region of high burnup or high irradiation period, and this is one of the issues that should be addressed to maintain and improve fuel reliability for the reasons that hydrogen concentration may affect ductility of zirconium alloy. Furthermore, a recent study showed abnormally high hydrogen concentration in fuel cladding. Comprehensive root cause analyses were made, which include a poolside measurement for a number of high burnup fuels and development of new technique for tenacious crud removal, with the result that it is confirmed that a unique combination of cladding material, irradiation period and environment caused the phenomenon and it is not expected to occur again. New iron enhanced zirconium alloys, which have been investigated in for a long time, are expected to have greater resistance to hydrogen pickup, and introduction of the alloys is considered as the future materials. Regarding extensive inspections of fuel assemblies and channel boxes at Kashiwazaki-Kariwa NPS, no abnormality due to Niigata-ken Chuetsu-oki earthquake was observed for all plants. This paper deals with the current fuel performance at Japanese

Probabilistic assessment of light water reactor fuel performance

International Nuclear Information System (INIS)

Misfeldt, I.

1978-10-01

A computer system for the statistical evaluation of LWR fuel performance has been developed. The computer code FRP, Fuel Reliability Predictor, calculates the distributions for parameters characterizing the fuel performance and failure probability. The statistical methods employed are either Monte Carlo simulations or low order Taylor approximation. Included in the computer system is a deterministic fuel performance code, which has been verified by comparison with data from irradiation experiments. The distributions for all material data utilized in the fuel simulations are estimations from the best available information in the literature. For the failure prediction, a stress corrosion failure criterion has been derived. The failure criterion is based on data from out-of-reactor stress corrosion experiments performed on unirradiated and irradiated zircaloy with iodine present. By means of an example the typical distributions of the variables characterizing the fuel performance and the accuracy of the methods themselves have been investigated. The application of the computer system is illustrated by a number of examples, these include the evaluation of irradiation experiments, design comparisons, and analyses of minor accidents. (author)

Water reactor fuel element fabrication, with special emphasis on its effects on fuel performance

International Nuclear Information System (INIS)

1979-01-01

Full text: The performance of nuclear fuel has improved over the years and is now a minor cause of outages and of power limitations in nuclear power plants. On the other hand, an increasing number of countries are in the process of developing or implementing their own capability for manufacturing fuel elements. In this context, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) advised that a symposium be organized devoted to the relationship between fuel fabrication and performance The Czechoslovak Atomic Energy Commission agreed to co-operate in the organization of this symposium and to host it in Prague. Those factors which influence fuel fabrication requirements are now well ascertained: as little reactor primary circuit contamination as possible, the tendency to increased burnups, reactor manoeuverability to match power grid demands, the desirability of an autonomous fabrication capability. It is the general experience of fuel element suppliers that fuel quality and performance has increased over the years, the importance of quality assurance and process monitoring has been decisive in this respect The ever increasing mass-production aspect of nuclear fuel leads to some processing steps being revised and alternatives being developed. The relation between fabrication processes and fuel performance characteristics, although generally well perceived, are still the subject of a large amount of experiment and assessment in most countries, both industrial and developing This evidence is most encouraging; it means indeed that nuclear power, which is already amongst the cheapest and safest sources of energy, will continue to be improved. The performance of Zircaloy fuel cladding - presently the material used in most water reactors - is under particular consideration. Better understanding of this quite recent alloy will pave the way for broader fuel utilization limits in the future. The panel discussion, which noted some

The fuel performance code future

International Nuclear Information System (INIS)

Ronchi, C.; Van de Laar, J.

1988-01-01

The paper describes the LWR version of the fuel performance code FUTURE, which was recently developed to calculate the fuel response (swelling, cladding deformation, release) to reactor transient conditions, starting from a broad-based description of the processes of major concern. The main physical models assumed are presented together with the scheme of the computer program

20. Annual report. OECD Halden reactor project. 1979

International Nuclear Information System (INIS)

1981-01-01

This is the Twentieth Annual Report on the OECD Halden Reactor Project, describing activities during 1979, the first year of the 1979-1981 Halden Agreement. Research work at the project is focussed on three areas: 1) In-core behaviour of reactor fuel, particularly reliability and safety aspects, which is studied through irradiation of test fuel elements. 2) Prediction, surveillance and control of fuel and core performance, for which models of fuel and core behaviour are developed. 3) Applications of process computers to power plant control, for which prototype software systems and hardware arrangements are developed

Performance tests for integral reactor nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

2006-02-15

An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

76 FR 37703 - Regulation of Fuels and Fuel Additives: 2012 Renewable Fuel Standards; Public Hearing

Science.gov (United States)

2011-06-28

... Regulation of Fuels and Fuel Additives: 2012 Renewable Fuel Standards; Public Hearing AGENCY: Environmental... hearing to be held for the proposed rule ``Regulation of Fuels and Fuel Additives: 2012 Renewable Fuel... be proposing amendments to the renewable fuel standard program regulations to establish annual...

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)

Effect of Fuel Additives on Spray Performance of Alternative Jet Fuels

Science.gov (United States)

Kannaiyan, Kumaran; Sadr, Reza

2015-11-01

Role of alternative fuels on reducing the combustion pollutants is gaining momentum in both land and air transport. Recent studies have shown that addition of nanoscale metal particles as fuel additives to liquid fuels have a positive effect not only on their combustion performance but also in reducing the pollutant formation. However, most of those studies are still in the early stages of investigation with the addition of nanoparticles at low weight percentages. Such an addition can affect the hydrodynamic and thermo-physical properties of the fuel. In this study, the near nozzle spray performance of gas-to-liquid jet fuel with and without the addition of alumina nanoparticles are investigated at macro- and microscopic levels using optical diagnostic techniques. At macroscopic level, the addition of nanoparticles is seen to enhance the sheet breakup process when compared to that of the base fuel. Furthermore, the microscopic spray characteristics such as droplet size and velocity are also found to be affected. Although the addition of nanoscale metal particles at low weight percentages does not affect the bulk fluid properties, the atomization process is found to be affected in the near nozzle region. Funded by Qatar National Research Fund.

Performance optimization of a PEM hydrogen-oxygen fuel cell

Energy Technology Data Exchange (ETDEWEB)

Sadiq Al-Baghdadi, Maher A.R. [Fuel Cell Research Center, International Energy and Environment Foundation, Al-Najaf, P.O.Box 39 (Iraq)

2013-07-01

The objective was to develop a semi-empirical model that would simulate the performance of proton exchange membrane (PEM) fuel cells without extensive calculations. A fuel cell mathematical module has been designed and constructed to determine the performance of a PEM fuel cell. The influence of some operating parameters on the performance of PEM fuel cell has been investigated using pure hydrogen on the anode side and oxygen on the cathode side. The present model can be used to investigate the influence of process variables for design optimization of fuel cells, stacks, and complete fuel cell power system. The possible mechanisms of the parameter effects and their interrelationships are discussed. In order to assess the validity of the developed model a real PEM fuel cell system has been used to generate experimental data. The comparison shows good agreements between the modelling results and the experimental data. The model is shown a very useful for estimating the performance of PEM fuel cell stacks and optimization of fuel cell system integration and operation.

Probabilistic assessments of fuel performance

International Nuclear Information System (INIS)

Kelppe, S.; Ranta-Puska, K.

1998-01-01

The probabilistic Monte Carlo Method, coupled with quasi-random sampling, is applied for the fuel performance analyses. By using known distributions of fabrication parameters and real power histories with their randomly selected combinations, and by making a large number of ENIGMA code calculations, one expects to find out the state of the whole reactor fuel. Good statistics requires thousands of runs. A sample case representing VVER-440 reactor fuel indicates relatively low fuel temperatures and mainly athermal fission gas release if any. The rod internal pressure remains typically below 2.5 MPa, which leaves a large margin to the system pressure of 12 MPa Gap conductance, an essential parameter in the accident evaluations, shows no decrease from its start-of-life value. (orig.)

Emission and economic performance assessment of a solid oxide fuel cell micro-combined heat and power system in a domestic building

International Nuclear Information System (INIS)

Elmer, Theo; Worall, Mark; Wu, Shenyi; Riffat, Saffa B.

2015-01-01

Combined heat and power (CHP) is a promising technological configuration for reducing energy consumption and increasing energy security in the domestic built environment. Fuel cells, on account of their: high electrical efficiency, low emissions and useful heat output have been identified as a key technological option for improving both building energy efficiency and reducing emissions in domestic CHP applications. The work presented in this paper builds upon results currently reported in the literature of fuel cells operating in domestic building applications, with an emission and economic performance assessment of a real, commercially available SOFC mCHP system operating in a real building; under a UK context. This paper aims to assess the emission and economic performance of a commercially available solid oxide fuel cell (SOFC) mCHP system, operating at The University of Nottingham's Creative Energy Homes. The performance assessment evaluates, over a one year period, the associated carbon (emission assessment) and operational costs (economic assessment) of the SOFC mCHP case compared to a ‘base case’ of grid electricity and a highly efficient gas boiler. Results from the annual assessment show that the SOFC mCHP system can generate annual emission reductions of up to 56% and cost reductions of 177% compared to the base case scenario. However support mechanisms such as; electrical export, feed in tariff and export tariff, are required in order to achieve this, the results are significantly less without. A net present value (NPV) analysis shows that the base case is still more profitable over a 15 year period, even though the SOFC mCHP system generates annual revenue; this is on account of the SOFC's high capital cost. In summary, grid interaction and incubator support is essential for significant annual emission and cost reductions compared to a grid electricity and gas boiler scenario. Currently capital cost is the greatest barrier to the economic

Advances in HTGR fuel performance models

International Nuclear Information System (INIS)

Stansfield, O.M.; Goodin, D.T.; Hanson, D.L.; Turner, R.F.

1985-01-01

Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10 -4 fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW/t HTGR. (author)

State of art in FE-based fuel performance codes

International Nuclear Information System (INIS)

Kim, Hyo Chan; Yang, Yong Sik; Kim, Dae Ho; Bang, Je Geon; Kim, Sun Ki; Koo, Yang Hyun

2013-01-01

Fuel performance codes approximate this complex behavior using an axisymmetric, axially-stacked, one-dimensional radial representation to save computation cost. However, the need for improved modeling of PCMI and, particularly, the importance of multidimensional capability for accurate fuel performance simulation has been identified as safety margin decreases. Finite element (FE) method that is reliable and proven solution in mechanical field has been introduced into fuel performance codes for multidimensional analysis. The present state of the art in numerical simulation of FE-based fuel performance predominantly involves 2-D axisymmetric model and 3-D volumetric model. The FRAPCON and FRAPTRAN own 1.5-D and 2-D FE model to simulate PCMI and cladding ballooning. In 2-D simulation, the FALCON code, developed by EPRI, is a 2-D (R-Z and R-θ) fully thermal-mechanically coupled steady-state and transient FE-based fuel behavior code. The French codes TOUTATIS and ALCYONE which are 3-D, and typically used to investigate localized behavior. In 2008, the Idaho National Laboratory (INL) has been developing multidimensional (2-D and 3-D) nuclear fuel performance code called BISON. In this paper, the current state of FE-based fuel performance code and their models are presented. Based on investigation into the codes, requirements and direction of development for new FE-based fuel performance code can be discussed. Based on comparison of models in FE-based fuel performance code, status of art in the codes can be discussed. A new FE-based fuel performance code should include typical pellet and cladding models which all codes own. In particular, specified pellet and cladding model such as gaseous swelling and high burnup structure (HBS) model should be developed to improve accuracy of code as well as consider AC condition. To reduce computation cost, the approximated gap and the optimized contact model should be also developed

Validation of Fuel Performance Uncertainty for RIA Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Nam-Gyu; Yoo, Jong-Sung; Jung, Yil-Sup [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of)

2016-10-15

To achieve this the computer code performance has to be validated based on the experimental results. And for the uncertainty quantification, important uncertainty parameters need to be selected, and combined uncertainty has to be evaluated with an acceptable statistical treatment. And important uncertainty parameters to the rod performance such as fuel enthalpy, fission gas release, cladding hoop strain etc. were chosen through the rigorous sensitivity studies. And their validity has been assessed by utilizing the experimental results, which were tested in CABRI and NSRR. Analysis results revealed that several tested rods were not bounded within combined fuel performance uncertainty. Assessment of fuel performance with an extended fuel power uncertainty on tested rods in NSRR and CABRI has been done. Analysis results showed that several tested rods were not bounded within calculated fuel performance uncertainty. This implies that the currently considered uncertainty range of the parameters is not enough to cover the fuel performance sufficiently.

Annual report 2003; Rapport annuel 2003

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

Delivering products and services to nuclear power plants operators, AREVA operates in every sector of the civilian nuclear power and fuel cycle industry. This annual report 2003 provides, in seven chapters, information on persons responsible for the annual report and for auditing the financial statements, general information on the company and share capital (statute, capital, share trading, dividends), information on company operations, changes and future prospects, assets, financial position and financial performance, corporate governance, recent developments and future prospects. (A.L.B.)

Developments of fuel performance analysis codes in KEPCO NF

International Nuclear Information System (INIS)

Han, H. T.; Choi, J. M.; Jung, C. D.; Yoo, J. S.

2012-01-01

The KEPCO NF has developed fuel performance analysis and design code named as ROPER, and utility codes of XGCOL and XDNB in order to perform fuel rod design evaluation for Korean nuclear power plants. The ROPER code intends to cover full range of fuel performance evaluation. The XGCOL code is for the clad flattening evaluation and the XDNB code is for the extensive DNB propagation evaluation. In addition to these, the KEPCO NF is now in the developing stage for 3-dimensional fuel performance analysis code, named as OPER3D, using 3-dimensional FEM for the nest generation within the joint project CANDU ENERGY in order to analyze PCMI behavior and fuel performance under load following operation. Of these, the ROPER code is now in the stage of licensing activities by Korean regulatory body and the other two are almost in the final developing stage. After finishing the developing, licensing activities are to be performed. These activities are intending to acquire competitiveness, originality, vendor-free ownership of fuel performance codes in the KEPCO NF

Performance evaluation of an open-cathode PEM fuel cell stack under ambient conditions: Case study of United Arab Emirates

International Nuclear Information System (INIS)

Al-Zeyoudi, Hend; Sasmito, Agus P.; Shamim, Tariq

2015-01-01

Highlights: • Performance evaluation of open-cathode PEM fuel cell stacks with forced air-convection. • Stack performance can vary up to 40% from winter to summer. • Hot and arid condition leads to membrane drying and performance deterioration. • Anode humidification improves the stack performance up to 40% during summer. - Abstract: The open-cathode polymer electrolyte membrane (PEM) fuel cell stack has been a promising candidate as a sustainable energy conversion system for replacing fossil fuel-based energy conversion devices in portable and automotive applications. As the ambient air is directly used to provide both oxidant and cooling, the complex cooling loop can be avoided which reduces the complexity and cost. However, the stack performance is highly affected by ambient conditions, i.e., ambient temperature and humidity. In this study, the effect of monthly ambient air conditions (temperature and humidity) is evaluated with respect to the stack’s power production performance as well as thermal, water and gas management by employing a validated three-dimensional open-cathode PEM fuel cell stack model. The annual climate data from the hot and arid environment of Abu Dhabi, United Arab Emirates (UAE) are used as a case study. The objective is to develop a better fundamental understanding of the interactions of physical phenomena in a fuel cell stack, which can assist in improving the performance and operation of an open-cathode PEM fuel cell-powered vehicle. The results indicate that the stack performance can vary significantly (up to 40%) from winter to summer, especially at high operating currents, with significant changes in the stack temperature and the water content at the membrane. Moreover, the anode humidification results in a significant improvement in the stack performance (up to 40%) in hot and dry conditions. However, a careful balance has to be struck between the humidifier parasitic load and the stack power.

Design of JMTR high-performance fuel element

International Nuclear Information System (INIS)

Sakurai, Fumio; Shimakawa, Satoshi; Komori, Yoshihiro; Tsuchihashi, Keiichiro; Kaminaga, Fumito

1999-01-01

For test and research reactors, the core conversion to low-enriched uranium fuel is required from the viewpoint of non-proliferation of nuclear weapon material. Improvements of core performance are also required in order to respond to recent advanced utilization needs. In order to meet both requirements, a high-performance fuel element of high uranium density with Cd wires as burnable absorbers was adopted for JMTR core conversion to low-enriched uranium fuel. From the result of examination of an adaptability of a few group constants generated by a conventional transport-theory calculation with an isotropic scattering approximation to a few group diffusion-theory core calculation for design of the JMTR high-performance fuel element, it was clear that the depletion of Cd wires was not able to be predicted accurately using group constants generated by the conventional method. Therefore, a new generation method of a few group constants in consideration of an incident neutron spectrum at Cd wire was developed. As the result, the most suitable high-performance fuel element for JMTR was designed successfully, and that allowed extension of operation duration without refueling to almost twice as long and offer of irradiation field with constant neutron flux. (author)

U.S. Department of Energy Hydrogen and Fuel Cells Program: 2017 Annual Merit Review and Peer Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

Popovich, Neil A [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

2017-10-18

The fiscal year 2017 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting (AMR), in conjunction with DOE's Vehicle Technologies Office AMR, was held from June June 5-9, 2017, in Washington, D.C. This report is a summary of comments by AMR peer reviewers about the hydrogen and fuel cell projects funded by DOE's Office of Energy Efficiency and Renewable Energy.

PWR and WWER fuel performance. A comparison of major characteristics

International Nuclear Information System (INIS)

Weidinger, H.

2006-01-01

PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

Review of CAGR fuel performance

Energy Technology Data Exchange (ETDEWEB)

Boocock, P.M. (South of Scotland Electricity Board, Glasgow (UK)); Eldred, V.W. (UKAEA Windscale Nuclear Power Development Labs.); Hines, G.F. (UKAEA Atomic Energy Establishment, Winfrith); Seeley, T.A. (British Nuclear Fuels Ltd., Springfields); Shaw, R.A. (UKAEA Springfields Nuclear Power Development Labs.); Waddington, J.S. (Central Electricity Generating Board, Berkeley (UK). Berkeley Nuclear Labs.)

1983-04-01

The performance of CAGR fuel in the Electricity Board's reactors is reviewed in the light of operating experience and the post-irradiation examination assessments which have been carried out. The results show that the extrapolation of the design from the Windscale AGR prototype has been very successful. Future trends in fuel design in order to achieve the higher burn-up targets now being specified are discussed.

State of art in FE-based fuel performance codes

International Nuclear Information System (INIS)

Kim, Hyo Chan; Yang, Yong Sik; Kim, Dae Ho; Bang, Je Geon; Kim, Sun Ki; Koo, Yang Hyun

2013-01-01

Finite element (FE) method that is reliable and proven solution in mechanical field has been introduced into fuel performance codes for multidimensional analysis. The present state of the art in numerical simulation of FE-based fuel performance predominantly involves 2-D axisymmetric model and 3-D volumetric model. The FRAPCON and FRAPTRAN own 1.5-D and 2-D FE model to simulate PCMI and cladding ballooning. In 2-D simulation, the FALCON code, developed by EPRI, is a 2-D (R-Z and R-θ) fully thermal-mechanically coupled steady-state and transient FE-based fuel behavior code. The French codes TOUTATIS and ALCYONE which are 3-D, and typically used to investigate localized behavior. In 2008, the Idaho National Laboratory (INL) has been developing multidimensional (2-D and 3-D) nuclear fuel performance code called BISON. In this paper, the current state of FE-based fuel performance code and their models are presented. Based on investigation into the codes, requirements and direction of development for new FE-based fuel performance code can be discussed. Based on comparison of models in FE-based fuel performance code, status of art in the codes can be discussed. A new FE-based fuel performance code should include typical pellet and cladding models which all codes own. In particular, specified pellet and cladding model such as gaseous swelling and high burnup structure (HBS) model should be developed to improve accuracy of code as well as consider AC condition. To reduce computation cost, the approximated gap and the optimized contact model should be also developed. Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena, occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. This multiphysics behavior is often tightly coupled, a well known example being the thermomechanical behavior. Adding to this complexity, important aspects of fuel behavior are inherently

Fuel performance experience at TVO nuclear power plant

International Nuclear Information System (INIS)

Patrakka, E.T.

1985-01-01

TVO nuclear power plant consists of two BWR units of ASEA-ATOM design. The fuel performance experience extending through six cycles at TVO I and four cycles at TVO II is reported. The experience obtained so far is mainly based on ASEA-ATOM 8 x 8 fuel and has been satisfactory. Until autumn 1984 one leaking fuel assembly had been identified at TVO I and none at TVO II. Most of the problems encountered have been related to leaf spring screws and channel screws. The experience indicates that satisfactory fuel performance can be achieved when utilizing strict operational rules and proper control of fuel design and manufacture. (author)

Fuels Performance | Transportation Research | NREL

Science.gov (United States)

Biodiesel Performance in Modern Engines NREL is working cooperatively with the National Biodiesel Board on research that assesses several aspects of biodiesel compatibility with engines, vehicles, and fuel-handling infrastructure. Several cooperative studies were recently conducted that examined how biodiesel performs with new

Performance optimization of a PEM hydrogen-oxygen fuel cell

OpenAIRE

Maher A.R. Sadiq Al-Baghdadi

2013-01-01

The objective was to develop a semi-empirical model that would simulate the performance of proton exchange membrane (PEM) fuel cells without extensive calculations. A fuel cell mathematical module has been designed and constructed to determine the performance of a PEM fuel cell. The influence of some operating parameters on the performance of PEM fuel cell has been investigated using pure hydrogen on the anode side and oxygen on the cathode side. The present model can be used to investigate t...

An annual performance appraisal for CERN

CERN Multimedia

2012-01-01

The March Council session is the occasion for CERN’s annual performance appraisal, but instead of the MARS form familiar to CERN staff, the lab’s working document is an Annual Progress Report, linked to the Medium-Term Plan, matching achievements to objectives. This year, I think it’s fair to say, we were firmly able to tick the ‘achieved’ box.   Top of the list of objectives was the LHC, which exceeded expectations in 2011 by delivering over five times the anticipated luminosity for protons while improving the lead-ion integrated luminosity by an order of magnitude compared to 2010.  As a result, the experiments published over 190 papers and made a staggering 1900 conference presentations. Underpinning this was the Worldwide LHC Computing Grid, which routinely performs so well that we hardly notice it's there. Non-LHC physics also had its day in the sun in 2011, with experiments at the PS, SPS, AD, n_TOF and ISOLDE, as well ...

Probabilistic evaluation of fuel element performance by the combined use of a fast running simplistic and a detailed deterministic fuel performance code

International Nuclear Information System (INIS)

Misfeldt, I.

1980-01-01

A comprehensive evaluation of fuel element performance requires a probabilistic fuel code supported by a well bench-marked deterministic code. This paper presents an analysis of a SGHWR ramp experiment, where the probabilistic fuel code FRP is utilized in combination with the deterministic fuel models FFRS and SLEUTH/SEER. The statistical methods employed in FRP are Monte Carlo simulation or a low-order Taylor approximation. The fast-running simplistic fuel code FFRS is used for the deterministic simulations, whereas simulations with SLEUTH/SEER are used to verify the predictions of FFRS. The ramp test was performed with a SGHWR fuel element, where 9 of the 36 fuel pins failed. There seemed to be good agreement between the deterministic simulations and the experiment, but the statistical evaluation shows that the uncertainty on the important performance parameters is too large for this ''nice'' result. The analysis does therefore indicate a discrepancy between the experiment and the deterministic code predictions. Possible explanations for this disagreement are discussed. (author)

2017 Annual Technology Baseline

Energy Technology Data Exchange (ETDEWEB)

Cole, Wesley J [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Hand, M. M [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Eberle, Annika [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Beiter, Philipp C [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Kurup, Parthiv [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Turchi, Craig S [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Feldman, David J [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Margolis, Robert M [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Augustine, Chad R [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Maness, Michael [Formerly NREL; O' Connor, Patrick [Oak Ridge National Laboratory

2018-03-26

Consistent cost and performance data for various electricity generation technologies can be difficult to find and may change frequently for certain technologies. With the Annual Technology Baseline (ATB), the National Renewable Energy Laboratory annually provides an organized and centralized set of such cost and performance data. The ATB uses the best information from the Department of Energy national laboratories' renewable energy analysts as well as information from the Energy Information Administration for fuel-based technologies. The ATB has been reviewed by experts and it includes the following electricity generation technologies: land-based wind, offshore wind, utility-scale solar photovoltaics (PV), commercial-scale solar PV, residential-scale solar PV, concentrating solar power, geothermal power, hydropower, coal, natural gas, nuclear, and conventional biopower. This webinar presentation introduces the 2017 ATB.

Technical specifications and performance of CANDU fuel

International Nuclear Information System (INIS)

Sejnoha, R.

1997-01-01

The relations between Technical Specifications and fuel performance are discussed in terms of design limits and margins. The excellent performance record of CANDU reactor fuel demonstrates that the fuel design defined in the Technical Specifications (and with it other components of the procurement cycle, such as fuel manufacturing), satisfy the requirements. New requirements, changing conditions of fuel application and accumulating experience make periodic updates of the Technical Specifications necessary. Under the CANDU Owners Group (COG) Working Party 9, a Work Package has been conducted to support the review of the Specifications and the documentation of the rationales for their requirements. So far, the review has been completed for 4 Specifications: 1 for Zircaloy tubing, and 3 for uranium dioxide powder. It is planned to complete the review of all 11 currently used specifications by 1999. The paper summarizes the results achieved to mid 1997. (author)

What Happens Inside a Fuel Cell? Developing an Experimental Functional Map of Fuel Cell Performance

KAUST Repository

Brett, Daniel J. L.

2010-08-20

Fuel cell performance is determined by the complex interplay of mass transport, energy transfer and electrochemical processes. The convolution of these processes leads to spatial heterogeneity in the way that fuel cells perform, particularly due to reactant consumption, water management and the design of fluid-flow plates. It is therefore unlikely that any bulk measurement made on a fuel cell will accurately represent performance at all parts of the cell. The ability to make spatially resolved measurements in a fuel cell provides one of the most useful ways in which to monitor and optimise performance. This Minireview explores a range of in situ techniques being used to study fuel cells and describes the use of novel experimental techniques that the authors have used to develop an \\'experimental functional map\\' of fuel cell performance. These techniques include the mapping of current density, electrochemical impedance, electrolyte conductivity, contact resistance and CO poisoning distribution within working PEFCs, as well as mapping the flow of reactant in gas channels using laser Doppler anemometry (LDA). For the high-temperature solid oxide fuel cell (SOFC), temperature mapping, reference electrode placement and the use of Raman spectroscopy are described along with methods to map the microstructural features of electrodes. The combination of these techniques, applied across a range of fuel cell operating conditions, allows a unique picture of the internal workings of fuel cells to be obtained and have been used to validate both numerical and analytical models. © 2010 Wiley-VCH Verlag GmbH& Co. KGaA, Weinheim.

Hydrogen engine performance analysis project. Second annual report

Energy Technology Data Exchange (ETDEWEB)

Adt, Jr., R. R.; Swain, M. R.; Pappas, J. M.

1980-01-01

Progress in a 3 year research program to evaluate the performance and emission characteristics of hydrogen-fueled internal combustion engines is reported. Fifteen hydrogen engine configurations will be subjected to performance and emissions characterization tests. During the first two years, baseline data for throttled and unthrottled, carburetted and timed hydrogen induction, Pre IVC hydrogen-fueled engine configurations, with and without exhaust gas recirculation (EGR) and water injection, were obtained. These data, along with descriptions of the test engine and its components, the test apparatus, experimental techniques, experiments performed and the results obtained, are given. Analyses of other hydrogen-engine project data are also presented and compared with the results of the present effort. The unthrottled engine vis-a-vis the throttled engine is found, in general, to exhibit higher brake thermal efficiency. The unthrottled engine also yields lower NO/sub x/ emissions, which were found to be a strong function of fuel-air equivalence ratio. (LCL)

Performance of Combustion Engineering fuel in operating PWRs

International Nuclear Information System (INIS)

Andrews, M.G.; Freeburn, H.R.; Wohlsen, W.D.

1979-01-01

Performance data on fuel assembly components from seven (7) operating reactors are presented, and discussed in detail where potential problems have occurred and been resolved. Fuel rod performance has continually improved over the last four (4) years with the gradual changeover to the current C-E fuel design. The reliability level is estimated at better than 99.99%, based on activity levels obtained through January 1979 at each plant. Control rod guide tubes have shown various degrees of wear caused by vibration of the control rods in their fully-withdrawn position. The retrofit of wear sleeves within the top portion of the affected guide tubes during routine refueling has permitted the use of these fuel assemblies with no significant loss in performance or safety margins

Irradiation performance of uranium-molybdenum alloy dispersion fuels

International Nuclear Information System (INIS)

Almeida, Cirila Tacconi de

2005-01-01

The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm 3 were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm 3 showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

Fuel Combustion and Engine Performance | Transportation Research | NREL

Science.gov (United States)

Fuel Combustion and Engine Performance Fuel Combustion and Engine Performance Photo of a gasoline emissions in advanced engine technologies. Photo by Dennis Schroeder, NREL NREL's combustion research and combustion and engine research activities include: Developing experimental and simulation research platforms

Fuel performance of rod-type research reactor fuel using a centrifugally atomized U-Mo powder

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Lee, Yoon Sang; Kim, Chang Kyu

2009-01-01

A low enriched uranium nuclear fuel for research reactors has been developed in order to replace a highly enriched uranium fuel according to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program. In KAERI, a rod-type U 3 Si dispersion fuel has been developed for a localization of the HANARO fuel and a U 3 Si/Al dispersion fuel of 3.15 gU/cc has been used at HANARO as a driver fuel since 2005. Although uranium silicide dispersion fuels such as U 3 Si 2 /Al and U 3 Si/Al are being used widely, high uranium density dispersion fuels (8-9 g/cm 3 ) are required for some high performance research reactors. U-Mo alloys have been considered as one of the most promising uranium alloys for a dispersion fuel due to their good irradiation performance. An international qualification program on U-Mo fuel to replace a uranium silicide dispersion fuel with a U-Mo dispersion fuel has been carried out

Effects of variations in fuel pellet composition and size on mixed-oxide fuel pin performance

International Nuclear Information System (INIS)

Makenas, B.J.; Jensen, B.W.; Baker, R.B.

1980-10-01

Experiments have been conducted which assess the effects on fuel pin performance of specific minor variations from nominal in both fuel pellet size and pellet composition. Such pellets are generally referred to in the literature as rogue pellets. The effect of these rogue pellets on fuel pin and reactor performance is shown to be minimal

Development of LWR fuel performance code FEMAXI-6

International Nuclear Information System (INIS)

Suzuki, Motoe

2006-01-01

LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

A comparison of thermal algorithms of fuel rod performance code systems

International Nuclear Information System (INIS)

Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

2003-11-01

The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

A comparison of thermal algorithms of fuel rod performance code systems

Energy Technology Data Exchange (ETDEWEB)

Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

2003-11-01

The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.

Anode-Engineered Protonic Ceramic Fuel Cell with Excellent Performance and Fuel Compatibility

NARCIS (Netherlands)

Hua, B.; Yan, N.; Li, M.; Sun, Y.-F.; Zhang, Y.-Q.; Li, J.; Etsell, T.; Sarkar, P.; Luo, J.L.

2016-01-01

Directly utilizing hydrocarbon fuels, particularly methane, is advantageous yet challenging in high-performance protonic ceramic fuel cells. In this work, this technological hurdle is well addressed by selective deposition of secondary electrocatalysts within the porous Ni-cermet anode. This novel

Manufacture of fuel and fuel channels and their performance in Indian PHWRs'

International Nuclear Information System (INIS)

Kalidas, R.

2005-01-01

Nuclear Fuel Complex (NFC) at Hyderabad is conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried our several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

SGHWR fuel performance, safety and reliability

International Nuclear Information System (INIS)

Pickman, D.O.; Inglis, G.H.

1977-01-01

The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of performance, safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith S.G.H.W.R. that the necessary bias toward safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications and has to be paid for. By a detailed evaluation of S.G.H.W.R. fuel defects it is shown that very few defects can be shown to be related to design, rating or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that must be met to achieve the desirable reliability level. It is possible that large-scale experience with S.G.H.W.R. fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to S.G.H.W.R. fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. The philosophy and supporting experimental work programme are outlined and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to new 60 pin fuel element to be used in the commercial S.G.H.W.R.'s and how it compares in design and performance aspects with the 36 pin element that has been used to date in the Winfrith S.G.H.W.R

Effect of different fuel options on performance of high-temperature PEMFC (proton exchange membrane fuel cell) systems

International Nuclear Information System (INIS)

Authayanun, Suthida; Saebea, Dang; Patcharavorachot, Yaneeporn; Arpornwichanop, Amornchai

2014-01-01

High-temperature proton exchange membrane fuel cells (HT-PEMFCs) have received substantial attention due to their high CO (carbon monoxide) tolerance and simplified water management. The hydrogen and CO fractions affect the HT-PEMFC performance and different fuel sources for hydrogen production result in different product gas compositions. Therefore, the aim of this study is to investigate the theoretical performance of HT-PEMFCs fueled by the reformate gas derived from various fuel options (i.e., methane, methanol, ethanol, and glycerol). Effects of fuel types and CO poisoning on the HT-PEMFC performance are analyzed. Furthermore, the necessity of a water-gas shift (WGS) reactor as a CO removal unit for pretreating the reformate gas is investigated for each fuel type. The methane steam reforming shows the highest possibility of CO formation, whereas the methanol steam reforming produces the lowest quantity of CO in the reformate gas. The methane fuel processing gives the maximum fraction of hydrogen (≈0.79) when the WGS reactor is included. The most suitable fuel is the one with the lowest CO poisoning effect and the maximum fuel cell performance. It is found that the HT-PEMFC system fueled by methanol without the WGS reactor and methane with WGS reactor shows the highest system efficiency (≈50%). - Highlights: • Performance of HT-PEMFC run on different fuel options is theoretically investigated. • Glycerol, methanol, ethanol and methane are hydrogen sources for the HT-PEMFC system. • Effect of CO poisoning on the HT-PEMFC performance is taken into account. • The suitable fuel for HT-PEMFC system is identified regarding the system efficiency

Influence of oxygen-metal ratio on mixed-oxide fuel performance

International Nuclear Information System (INIS)

Lawrence, L.A.; Leggett, R.D.

1979-04-01

The fuel oxygen-to-metal ratio (O/M) is recognized as an important consideration for performance of uranium--plutonium oxide fuels. An overview of the effects of differing O/M's on the irradiation performance of reference design mixed-oxide fuel in the areas of chemical and mechanical behavior, thermal performance, and fission gas behavior is presented. The pellet fuel has a nominal composition of 75 wt% UO 2 + 25 wt% PuO 2 at a pellet density of approx. 90% TD. for nominal conditions this results in a smeared density of approx. 85%. The cladding in all cases is 20% CW type 316 stainless steel with an outer diameter of 5.84 to 6.35 mm. O/M has been found to significantly influence fuel pin chemistry, mainly FCCI and fission product and fuel migration. It has little effect on thermal performance and overall mechanical behavior or fission gas release. The effects of O/M (ranging from 1.938 to 1.984) in the areas of fuel pin chemistry, to date, have not resulted in any reduction in fuel pin performance capability to goal burnups of approx. 8 atom% or more

Performance of metallic fuels in liquid-metal fast reactors

International Nuclear Information System (INIS)

Seidel, B.R.; Walters, L.C.; Kittel, J.H.

1984-01-01

Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

Effect of pilot fuel quantity on the performance of a dual fuel engine

Energy Technology Data Exchange (ETDEWEB)

Abd Alla, G.H.; Soliman, H.A.; Badr, O.A.; Abd Rabbo, M.F. [Zagazig University, Cairo (Egypt). Shoubra Faculty of Engineering

2000-04-01

It is well known that the operation of dual fuel engines at lower loads suffers from lower thermal efficiency and higher unburned percentages of fuel. To rectify this problem, tests have been conducted on a special single cylinder compression ignition research engine (Ricardo E6) to investigate the effect of pilot fuel quantity on the performance of an indirect injection diesel engine fuelled with gaseous fuel. Diesel fuel was used as the pilot fuel and methane or propane was used as the main fuel which was inducted into the intake manifold to mix with the intake air. Through experimental investigations, it is shown that, the low efficiency and excess emissions at light loads can be improved significantly by increasing the amount of pilot fuel, while increasing the amount of pilot fuel at high loads led to early knocking. (author)

Spent fuel and fuel pool component integrity. Annual report, FY 1979

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

1980-05-01

International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-μm) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report

Thermodynamic properties of the DUPIC fuel and its performance

Energy Technology Data Exchange (ETDEWEB)

Park, Kwang Heon; Kim, Hee Moon [Kyung Hee Univ., Seoul (Korea, Republic of)

1997-07-01

This study describes thermodynamic properties of DUPIC fuel and performance. In initial state, DUPIC fuel which contains fissile materials is different from general nuclear fuel. So this study analyzed oxygen potential, thermal conductivity and specific heat of the DUPIC fuel.

SKB annual report 1992

International Nuclear Information System (INIS)

1993-05-01

This is the annual report on the activities of the Swedish Nuclear Fuel and Waste Management Co, SKB. It contains in part 1 an overview of SKB activities in different fields. Part 2 gives a description of the research and development work on nuclear waste disposal performed during 1992. Lectures and publications during 1992 as well as reports issued in the SKB technical report series are listed in part 4. Part 5 contains the summaries of all technical reports issued during 1992. SKB is the owner of CLAB, the Central Facility for Interim Storage of Spent Nuclear Fuel, located at Oskarshamn. CLAB was taken into operation in July 1985 and to the end of 1992 in total 1684 tonnes of spent fuel (measured as uranium) has been received. Transportation from the nuclear site to CLAB is made by a special ship, M/S Sigyn. At Forsmark the final repository for Radioactive Waste -SFR- was taken in operation in April 1988. At the end of 1992 a total of 11000 m 3 of waste have been deposited in SFR. The total cost for R and D during 1992 was 192.3 MSEK of which 24.8 MSEK came from participants outside Sweden. Some of the main areas for SKB research are: groundwater movements, bedrock stability, groundwater chemistry and nuclide migration, method and instruments for in situ characterization of crystalline bedrock, characterization and leaching of spent nuclear fuel, properties of bentonite for buffer, backfilling and sealing, radionuclide transport in biosphere and dose evaluations, development of performance and safety assessment methodology and assessment models, construction of an underground research laboratory. Cost calculations for the total nuclear waste management system, including decommissioning of all reactors, are updated annually. The total cost is estimated to 55 billion SEK

What Happens Inside a Fuel Cell? Developing an Experimental Functional Map of Fuel Cell Performance

KAUST Repository

Brett, Daniel J. L.; Kucernak, Anthony R.; Aguiar, Patricia; Atkins, Stephen C.; Brandon, Nigel P.; Clague, Ralph; Cohen, Lesley F.; Hinds, Gareth; Kalyvas, Christos; Offer, Gregory J.; Ladewig, Bradley; Maher, Robert; Marquis, Andrew; Shearing, Paul; Vasileiadis, Nikos; Vesovic, Velisa

2010-01-01

Fuel cell performance is determined by the complex interplay of mass transport, energy transfer and electrochemical processes. The convolution of these processes leads to spatial heterogeneity in the way that fuel cells perform, particularly due

WWER reactor fuel performance, modelling and experimental support. Proceedings

International Nuclear Information System (INIS)

Stefanova, S.; Chantoin, P.; Kolev, I.

1994-01-01

This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: 1) WWER Fuel Performance and Economics: Status and Improvement Prospects: 2) WWER Fuel Behaviour Modelling and Experimental Support; 3) Licensing of WWER Fuel and Fuel Analysis Codes; 4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items

WWER reactor fuel performance, modelling and experimental support. Proceedings

Energy Technology Data Exchange (ETDEWEB)

Stefanova, S; Chantoin, P; Kolev, I [eds.

1994-12-31

This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: (1) WWER Fuel Performance and Economics: Status and Improvement Prospects: (2) WWER Fuel Behaviour Modelling and Experimental Support; (3) Licensing of WWER Fuel and Fuel Analysis Codes; (4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items.

Advanced fuel technology and performance: Current status and trends

International Nuclear Information System (INIS)

1990-11-01

During the last years the Nuclear Fuel Cycle and Waste Management Division of the IAEA has been giving great attention to the collection, analysis and exchange of information in the field of reactor fuel technology. Most of these activities are being conducted in the framework of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT). The purpose of this Advisory Group Meeting on Advanced Fuel Technology and Performance was to update and to continue the previous work, and to review the experience of advanced fuel technology, its performance with regard to all types of reactors and to outline the future trends on the basis of national experience and discussions during the meeting. As a result of the meeting a Summary Report was prepared which reflected the status of the advanced nuclear fuel technology up to 1990. The 10 papers presented by participants of this meeting are also published here. A separate abstract was prepared for each of these papers. Refs, figs and tabs

High-quality thorium TRISO fuel performance in HTGRs

Energy Technology Data Exchange (ETDEWEB)

Verfondern, Karl [Forschungszentrum Juelich GmbH (Germany); Allelein, Hans-Josef [Forschungszentrum Juelich GmbH (Germany); Technische Hochschule Aachen (Germany); Nabielek, Heinz; Kania, Michael J.

2013-11-01

Thorium as a nuclear fuel has received renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTGR development employed thorium together with high-enriched uranium (HEU). After 1980, HTGR fuel systems switched to low-enriched uranium (LEU). After completing fuel development for the AVR and the THTR with BISO coated particles, the German program expanded its efforts utilizing thorium and HEU TRISO coated particles in advanced HTGR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of a low-temperature isotropic (LTI) inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with high-temperature isotropic (HTI) BISO coatings. The improved performance of the HEU (Th, U)O{sub 2} TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTGR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 C in normal operations and 1600 C in accidents, with burnups to 13% FIMA and fast fluences to 8 x 10{sup 25} n/m{sup 2} (E> 16 fJ), the performance results exceed the design limits on manufacturing and operational requirements for the German HTR-Modul concept, which are 6.5 x 10{sup -5} for manufacturing, 2 x 10{sup -4} for normal operating conditions, and 5 x 10{sup -4

High-quality thorium TRISO fuel performance in HTGRs

International Nuclear Information System (INIS)

Verfondern, Karl; Allelein, Hans-Josef; Nabielek, Heinz; Kania, Michael J.

2013-01-01

Thorium as a nuclear fuel has received renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTGR development employed thorium together with high-enriched uranium (HEU). After 1980, HTGR fuel systems switched to low-enriched uranium (LEU). After completing fuel development for the AVR and the THTR with BISO coated particles, the German program expanded its efforts utilizing thorium and HEU TRISO coated particles in advanced HTGR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of a low-temperature isotropic (LTI) inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with high-temperature isotropic (HTI) BISO coatings. The improved performance of the HEU (Th, U)O 2 TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTGR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 C in normal operations and 1600 C in accidents, with burnups to 13% FIMA and fast fluences to 8 x 10 25 n/m 2 (E> 16 fJ), the performance results exceed the design limits on manufacturing and operational requirements for the German HTR-Modul concept, which are 6.5 x 10 -5 for manufacturing, 2 x 10 -4 for normal operating conditions, and 5 x 10 -4 for accident conditions. These

SP-100 Fuel Pin Performance: Results from Irradiation Testing

Science.gov (United States)

Makenas, Bruce J.; Paxton, Dean M.; Vaidyanathan, Swaminathan; Marietta, Martin; Hoth, Carl W.

1994-07-01

A total of 86 experimental fuel pins with various fuel, liner, and cladding candidate materials have been irradiated in the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF) reactor as part of the SP-100 fuel pin irradiation testing program. Postirradiation examination results from these fuel pins are key in establishing performance correlations and demonstrating the lifetime and safety of the reactor fuel system. This paper provides a brief description of the in-reactor fuel pin tests and presents the most recent irradiation data on the performance of wrought rhenium (Re) liner material and high density UN fuel at goal burnup of 6 atom percent (at. %). It also provides an overview of the significant variety of other fuel/liner/cladding combinations which were irradiated as part of this program and which may be of interest to more advanced efforts.

Performance of PARR-1 with LEU Fuel

International Nuclear Information System (INIS)

Pervez, S.; Latif, M.; Bokhari, I.H.; Bakhtyar, S.

2005-01-01

Pakistan Research Reactor (PARR-1) went critical in 1965 with HEU fuel. The reactor core was converted to LEU fuel with power upgradation from 5 MW to 10 MW in 1992. The reactor has been operated with LEU fuel for about 10,000 hours and has produced about 66,000 MWh energy up to now. Average burn up of the irradiated fuel is about 42 %. The fuel performance during the last 12 years has been excellent. Post irradiation visual inspection of the fuel has revealed no abnormality. During operation there have been no signs of releases in the pool water establishing the full integrity of this fuel. The reactor has been mainly utilized for radioisotope production, beam tube experiments including neutron diffraction studies, neutron radiography etc. Studies have been completed to operate the reactor with a mixed core (HEU + LEU) to utilize the less burned HEU fuel elements. A major project of production of fission Moly using PARR-1 is in the final stages. (author)

Fuel performance-experience to date and future potential

International Nuclear Information System (INIS)

Proebstle, R.A.; Klepfer, H.H.

1987-01-01

The experience in the USA to date, as reported in the Federal Energy Regulatory Commission data, conforms a very favorable cost trend for nuclear fuel costs relative to fossil fuel costs. The nuclear fuel cost promose relative to other fuels looks even better in future. Uranium supply surplus and advances in enrichment technology suggest that this trend should continue. Threats to the economic potential for nuclear fuel costs include unexpected problems in actural versus projected core and fuel technical performance. The New designs for BWR's nuclear fuel are extended to 38,000 MWd/MTU and the fuel rod reliabilities of 0.999994 are achievable. This reliability is equivalent to less than 3 fuel rod failures over the 40 year life of a reactor. (Liu)

Isotopic modelling using the ENIGMA-B fuel performance code

International Nuclear Information System (INIS)

Rossiter, G.D.; Cook, P.M.A.; Weston, R.

2001-01-01

A number of experimental programmes by BNFL and other MOX fabricators have now shown that the in-pile performance of MOX fuel is generally similar to that of conventional UO 2 fuel. Models based on UO 2 fuel experience form a good basis for a description of MOX fuel behaviour. However, an area where the performance of MOX fuel is sufficiently different from that of UO 2 to warrant model changes is in the radial power and burnup profile. The differences in radial power and burnup profile arise from the presence of significant concentrations of plutonium in MOX fuel, at beginning of life, and their subsequent evolution with burnup. Amongst other effects, plutonium has a greater neutron absorption cross-section than uranium. This paper focuses on the development of a new model for the radial power and burnup profile within a UO 2 or MOX fuel rod, in which the underlying fissile isotope concentration distributions are tracked during irradiation. The new model has been incorporated into the ENIGMA-B fuel performance code and has been extended to track the isotopic concentrations of the fission gases, xenon and krypton. The calculated distributions have been validated against results from rod puncture measurements and electron probe micro-analysis (EPMA) linescans, performed during the M501 post irradiation examination (PIE) programme. The predicted gas inventory of the fuel/clad gap is compared with the isotopic composition measured during rod puncture and the measured radial distributions of burnup (from neodymium measurements) and plutonium in the fuel are compared with the calculated distributions. It is shown that there is good agreement between the code predictions and the measurements. (author)

Use of advanced simulations in fuel performance codes

International Nuclear Information System (INIS)

Van Uffelen, P.

2015-01-01

The simulation of the cylindrical fuel rod behaviour in a reactor or a storage pool for spent fuel requires a fuel performance code. Such tool solves the equations for the heat transfer, the stresses and strains in fuel and cladding, the evolution of several isotopes and the behaviour of various fission products in the fuel rod. The main equations along with their limitations are briefly described. The current approaches adopted for overcoming these limitations and the perspectives are also outlined. (author)

A statistical approach to nuclear fuel design and performance

Science.gov (United States)

Cunning, Travis Andrew

As CANDU fuel failures can have significant economic and operational consequences on the Canadian nuclear power industry, it is essential that factors impacting fuel performance are adequately understood. Current industrial practice relies on deterministic safety analysis and the highly conservative "limit of operating envelope" approach, where all parameters are assumed to be at their limits simultaneously. This results in a conservative prediction of event consequences with little consideration given to the high quality and precision of current manufacturing processes. This study employs a novel approach to the prediction of CANDU fuel reliability. Probability distributions are fitted to actual fuel manufacturing datasets provided by Cameco Fuel Manufacturing, Inc. They are used to form input for two industry-standard fuel performance codes: ELESTRES for the steady-state case and ELOCA for the transient case---a hypothesized 80% reactor outlet header break loss of coolant accident. Using a Monte Carlo technique for input generation, 105 independent trials are conducted and probability distributions are fitted to key model output quantities. Comparing model output against recognized industrial acceptance criteria, no fuel failures are predicted for either case. Output distributions are well removed from failure limit values, implying that margin exists in current fuel manufacturing and design. To validate the results and attempt to reduce the simulation burden of the methodology, two dimensional reduction methods are assessed. Using just 36 trials, both methods are able to produce output distributions that agree strongly with those obtained via the brute-force Monte Carlo method, often to a relative discrepancy of less than 0.3% when predicting the first statistical moment, and a relative discrepancy of less than 5% when predicting the second statistical moment. In terms of global sensitivity, pellet density proves to have the greatest impact on fuel performance

Spent fuel metal storage cask performance testing and future spent fuel concrete module performance testing

International Nuclear Information System (INIS)

McKinnon, M.A.; Creer, J.M.

1988-10-01

REA-2023 Gesellshaft fur Nuklear Service (GNS) CASTOR-V/21, Transnuclear TN-24P, and Westinghouse MC-10 metal storage casks, have been performance tested under the guidance of the Pacific Northwest Laboratory to determine their thermal and shielding performance. The REA-2023 cask was tested under Department of Energy (DOE) sponsorship at General Electric's facilities in Morris, Illinois, using BWR spent fuel from the Cooper Reactor. The other three casks were tested under a cooperative agreement between Virginia Power Company and DOE at the Idaho National Engineering Laboratory (INEL) by EGandG Idaho, Inc., using intact spent PWR fuel from the Surry reactors. The Electric Power Research Institute (EPRI) made contributions to both programs. A summary of the various cask designs and the results of the performance tests is presented. The cask designs include: solid and liquid neutron shields; lead, steel, and nodular cast iron gamma shields; stainless steel, aluminum, and copper baskets; and borated materials for criticality control. 4 refs., 8 figs., 6 tabs

Modeling of high-density U-MO dispersion fuel plate performance

International Nuclear Information System (INIS)

Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Rest, J.; Snelgrove, J.L.

2002-01-01

Results from postirradiation examinations (PIE) of highly loaded U-Mo/Al dispersion fuel plates over the past several years have shown that the interaction between the metallic fuel particles and the matrix aluminum can be extensive, reducing the volume of the high-conductivity matrix phase and producing a significant volume of low-conductivity reaction-product phase. This phenomenon results in a significant decrease in fuel meat thermal conductivity during irradiation. PIE has further shown that the fuel-matrix interaction rate is a sensitive function of irradiation temperature. The interplay between fuel temperature and fuel-matrix interaction makes the development of a simple empirical correlation between the two difficult. For this reason a comprehensive thermal model has been developed to calculate temperatures throughout the fuel plate over its lifetime, taking into account the changing volume fractions of fuel, matrix and reaction-product phases within the fuel meat owing to fuel-matrix interaction; this thermal model has been incorporated into the dispersion fuel performance code designated PLATE. Other phenomena important to fuel thermal performance that are also treated in PLATE include: gas generation and swelling in the fuel and reaction-product phases, incorporation of matrix aluminum into solid solution with the unreacted metallic fuel particles, matrix extrusion resulting from fuel swelling, and cladding corrosion. The phenomena modeled also make possible a prediction of fuel plate swelling. This paper presents a description of the models and empirical correlations employed within PLATE as well as validation of code predictions against fuel performance data for U-Mo experimental fuel plates from the RERTR-3 irradiation test. (author)

Transmutation Fuel Performance Code Thermal Model Verification

Energy Technology Data Exchange (ETDEWEB)

Gregory K. Miller; Pavel G. Medvedev

2007-09-01

FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.

Space reactor fuels performance and development issues

International Nuclear Information System (INIS)

Wewerka, E.M.

1984-01-01

Three compact reactor concepts are now under consideration by the US Space Nuclear Power Program (the SP-100 Program) as candidates for the first 100-kWe-class space reactor. Each of these reactor designs puts unique constraints and requirements on the fuels system, and raises issues of fuel systems feasibility and performance. This paper presents a brief overview of the fuel requirements for the proposed space reactor designs, a delineation of the technical feasibility issues that each raises, and a description of the fuel systems development and testing program that has been established to address key technical issues

U.S. Department of Energy Hydrogen and Fuel Cells Program 2011 Annual Merit Review and Peer Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

Satypal, S.

2011-09-01

This document summarizes the comments provided by peer reviewers on hydrogen and fuel cell projects presented at the FY 2011 U.S. Department of Energy (DOE) Hydrogen Program and Vehicle Technologies Program Annual Merit Review and Peer Evaluation Meeting (AMR), held May 9-13, 2011 in Arlington, Virginia

Investigation of microalgae HTL fuel effects on diesel engine performance and exhaust emissions using surrogate fuels

International Nuclear Information System (INIS)

Hossain, Farhad M.; Nabi, Md. Nurun; Rainey, Thomas J.; Bodisco, Timothy; Rahman, Md. Mostafizur; Suara, Kabir; Rahman, S.M.A.; Van, Thuy Chu; Ristovski, Zoran; Brown, Richard J.

2017-01-01

Highlights: • Development of a microalgae HTL surrogate of biocrude fuel using chemical compounds. • Physiochemical properties of surrogate blends were analysed. • Experimentally investigated diesel engine performance and emissions using surrogate fuels. • No significant changes in engine performance were observed with HTL surrogate blends. • Major emissions including PM, PN and CO were reduced significantly with increasing of NOx emission. - Abstract: This paper builds on previous work using surrogate fuel to investigate advanced internal combustion engine fuels. To date, a surrogate fuel of this nature has not been used for microalgae hydrothermal liquefaction (HTL) biocrude. This research used five different chemical groups found in microalgae HTL biocrude to design a surrogate fuel. Those five chemical groups constitute around 65% (by weight) of a microalgae biocrude produced by HTL. Weight percentage of the microalgae HTL biocrude chemical compounds were used to design the surrogate fuel, which was miscible with diesel at all percentages. The engine experiments were conducted on a EURO IIIA turbocharged common-rail direct-injection six-cylinder diesel engine to test engine performance and emissions. Exhaust emissions, including particulate matter and other gaseous emissions, were measured with the surrogate fuel and a reference diesel fuel. Experimental results showed that without significantly deteriorating engine performance, lower particulate mass, particulate number and CO emissions were observed with a penalty in NOx emissions for all surrogate blends compared to those of the reference diesel.

Spent Fuel Transportation Package Performance Study - Experimental Design Challenges

International Nuclear Information System (INIS)

Snyder, A. M.; Murphy, A. J.; Sprung, J. L.; Ammerman, D. J.; Lopez, C.

2003-01-01

Numerous studies of spent nuclear fuel transportation accident risks have been performed since the late seventies that considered shipping container design and performance. Based in part on these studies, NRC has concluded that the level of protection provided by spent nuclear fuel transportation package designs under accident conditions is adequate. [1] Furthermore, actual spent nuclear fuel transport experience showcase a safety record that is exceptional and unparalleled when compared to other hazardous materials transportation shipments. There has never been a known or suspected release of the radioactive contents from an NRC-certified spent nuclear fuel cask as a result of a transportation accident. In 1999 the United States Nuclear Regulatory Commission (NRC) initiated a study, the Package Performance Study, to demonstrate the performance of spent fuel and spent fuel packages during severe transportation accidents. NRC is not studying or testing its current regulations, a s the rigorous regulatory accident conditions specified in 10 CFR Part 71 are adequate to ensure safe packaging and use. As part of this study, NRC currently plans on using detailed modeling followed by experimental testing to increase public confidence in the safety of spent nuclear fuel shipments. One of the aspects of this confirmatory research study is the commitment to solicit and consider public comment during the scoping phase and experimental design planning phase of this research

Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

2007-07-01

Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

2007-01-01

Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

Metal fuel manufacturing and irradiation performance

International Nuclear Information System (INIS)

Pedersen, D.R.; Walters, L.C.

1992-01-01

The advances in metal fuel by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, and improved passive safety. The goals and the safety philosophy of the Integral Fast Reactor Program are stressed

Effect of Biodiesel Fuel Injection Timing and Venture for Gaseous Fuel Induction on the Performance, Emissions and Combustion Characteristics of Dual Fuel Engine

Directory of Open Access Journals (Sweden)

Mallikarjun Bhovi

2018-02-01

Full Text Available Advancing or retarding pilot fuel injection timing in a diesel engine provided with either conventional mechanical fuel injection (CMFIS or high pressure injection as in common rail fuel injection (CRDI systems can significantly affect its performance and tail pipe emissions. Performance of diesel engine when fueled with various biofuels as well as gaseous fuels tends to vary with subsequent changes in pilot fuel injection timings. Biodiesel derived from rubber seed oil called Rubber Seed Oil Methyl Ester (RuOME and hydrogen (H2 and hydrogen enriched compressed natural gas called (HCNG both being renewable fuels when used in diesel engines modified to operate in dual fuel mode can provide complete replacement for fossil diesel. In the present study, effect of injection timings and venture design for gas mixing on the performance, combustion and emission characteristics of dual fuel engine fitted with both CMFIS and CRDI injection systems and operated on RuOME and HCNG/hydrogen has been investigated. Results showed that high pressure CRDI assisted injection of RuOME with optimized mixing chamber (carburetor for hydrogen induction in dual fuel engine performed improved compared to that with CMFIS. In addition, for the same fuel combinations, CRDI resulted in lower biodiesel consumption, lower carbon monoxide (BSCO and hydrocarbon (BSHC emissions and increased NOx emissions than CMFIS operation.

Alternate-Fueled Combustor-Sector Performance. Parts A and B; (A) Combustor Performance; (B) Combustor Emissions

Science.gov (United States)

Shouse, D. T.; Hendricks, R. C.; Lynch, A.; Frayne, C. W.; Stutrud, J. S.; Corporan, E.; Hankins, T.

2012-01-01

Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F(2008) or ASTM D 7566 (2010) standards, respectively, and are classified as "drop-in" fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are certified individually on the basis of processing and assumed to be feedstock agnostic. Adherence to alternate fuels and fuel blends requires "smart fueling systems" or advanced fuel-flexible systems, including combustors and engines, without significant sacrifice in performance or emissions requirements. This paper provides preliminary performance (Part A) and emissions and particulates (Part B) combustor sector data. The data are for nominal inlet conditions at 225 psia and 800 F (1.551 MPa and 700 K), for synthetic-paraffinic-kerosene- (SPK-) type (Fisher-Tropsch (FT)) fuel and blends with JP-8+100 relative to JP-8+100 as baseline fueling. Assessments are made of the change in combustor efficiency, wall temperatures, emissions, and luminosity with SPK of 0%, 50%, and 100% fueling composition at 3% combustor pressure drop. The performance results (Part A) indicate no quantifiable differences in combustor efficiency, a general trend to lower liner and higher core flow temperatures with increased FT fuel blends. In general, emissions data (Part B) show little differences, but with percent increase in FT-SPK-type fueling, particulate emissions and wall temperatures are less than with baseline JP-8. High-speed photography illustrates both luminosity and combustor dynamic flame characteristics.

The METEOR/TRANSURANUS fuel performance code

International Nuclear Information System (INIS)

Struzik, C.; Guerin, Y.

1996-01-01

The first calculations for the FUMEX exercise were performed using version 1.1 of the METEOR/TRANSURANUS code. Since then, important improvements have been implemented on several models. In its present state, the code describes fuel rod behaviour in standard PWR conditions. Its validity extends to UO 2 and MOX fuels clad in Zircaloy-4. Power transient calculations for UO 2 and Gd doped fuel calculations are possible, but further developments are in progress, and the applications will be fully qualified in version 2.0. A considerable effort is made to replace semi-empirical models with models that have a sounder physical basis. (authors). 14 refs

Empirical Study on Annual Energy-Saving Performance of Energy Performance Contracting in China

Directory of Open Access Journals (Sweden)

Hongquan Ruan

2018-05-01

Full Text Available A lack of trust in Energy Service Company (ESCo is the most critical factor affecting the development of Energy Performance Contracting (EPC in China, compared with other constraints. One cannot easily estimate the energy-saving performance of an EPC project. Under that condition, lack of trust may cause the Energy-Consuming Unit (ECU to suspect the energy-saving performance promised by the ESCo, thus leaving potentially profitable projects without necessary funding. Currently, specific studies taking an across-projects viewpoint on annual energy-saving performance of EPC projects in multiple subsectors, objectively and quantitatively, are lacking. This paper studies the regression relationships of annual energy-saving quantity in terms of revamping cost and the regression relationships of annual cost saving in terms of revamping cost. The regression results show that there are statistically significant correlations in the above relationships in the nine subsectors investigated. This is significant for ESCos and ECUs, because knowledge on energy-saving performance could contribute to EPC investment decisions and trust relationships between ESCos and ECUs. Then, a multiple linear regression model of revamping cost is set up to analyze its influencing factors. The model indicates that the subsector the sample belongs to, financing, registered capital of the ESCo, and contract period have significant effects on revamping cost. Thus, policy implications regarding innovation of EE promotion technology, clarifying ESCos’ exit mechanism, innovation of financing mechanism, and improving the market credit environment for promoting investment in EPC projects, are provided.

IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

Directory of Open Access Journals (Sweden)

M.K. MEYER

2014-04-01

Full Text Available High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

Irradiation performance of U-Mo monolithic fuel

Energy Technology Data Exchange (ETDEWEB)

Meyer, M. K.; Gan, J.; Jue, J. F.; Keiser, D. D.; Perez, E.; Robinson, A.; Wachs, D. M.; Woolstenhulme, N. [Idaho National Laboratory, Idaho (Korea, Republic of); Kim, Y.S.; Hofman, G. L. [Argonne National Laboratory, Lemont (United States)

2014-04-15

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

Performance of the Westinghouse WWER-1000 fuel design

International Nuclear Information System (INIS)

Hoglund, J.; Riznychenko, O.; Latorre, R.; Lashevych, P.

2011-01-01

In 2005 six (6) Westinghouse WWER-1000 Lead Test Assemblies (LTAs) were loaded in the South Ukraine Unit 3. This design has demonstrated full compatibility with resident fuel designs and all associated fuel handling and reactor components. Operations have further demonstrated adequacy of performance margins and the reliability requirements for multiple cycles of operation. The LTA's have now been discharged after completing the planned four cycles of operation and having reached an average assembly burnup in excess of 43 MWd/kgU. Post Irradiation Examinations were performed after completion of each cycle. The final LTA inspection program at end of Cycle 20 in 2010 yielded satisfactory results on all counts, and it was concluded that the 6 Westinghouse LTA's performed as expected during their operational regimes. Very good performance was demonstrated in the WWER-1000 reactor environment for the Zr-1%Nb as grid material, and ZIRLO fuel cladding and structural components. Control Rod Assemblies drop times and drag forces were all within the accepted values. The LTA program demonstrated that this fuel design is suitable for full core applications. However, the topic of fuel assembly distortion resistance was re-visited and Westinghouse therefore considered operational experience and design features from multiple development programs to enhance the basic Westinghouse WWER-1000 fuel design for Ukrainian reactors. The design now includes features that further mitigate assembly bow while at the same time improving the fuel cycle economy. This paper describes briefly the development of the Westinghouse WWER-1000 fuel design and how test results and operational experiences from multiple sources have been utilized to produce a most suitable fuel design. Early in 2011 a full region of the Westinghouse WWER-1000 design completed another full cycle of operation at South Ukraine Unit 3, all with excellent results. All 42 fuel assemblies were examined for visible damage or non

Pushing back the boundaries of PWR fuel performance

International Nuclear Information System (INIS)

Sofer, G.A.; Skogen, F.B.; Brown, C.A.; Fresk, Y.U.

1985-01-01

In today's fiercely competitive PWR reload market utilities are benefiting from a variety of design innovations which are helping to cut fuel cycle costs and to improve fuel performance. An advanced PWR fuel design from Exxon, for example, currently under evaluation at the Ginna plant in the United States, offers higher burn-up and greater power cycling. (author)

Dissolution performance of plutonium nitride based fuel materials

Energy Technology Data Exchange (ETDEWEB)

Aneheim, E.; Hedberg, M. [Nuclear Chemistry, Chemistry and Chemical Engineering, Chalmers University of Technology, Kemivaegen 4, Gothenburg, SE41296 (Sweden)

2016-07-01

Nitride fuels have been regarded as one viable fuel option for Generation IV reactors due to their positive features compared to oxides. To be able to close the fuel cycle and follow the Generation IV concept, nitrides must, however, demonstrate their ability to be reprocessed. This means that the dissolution performance of actinide based nitrides has to be thoroughly investigated and assessed. As the zirconium stabilized nitrides show even better potential as fuel material than does the pure actinide containing nitrides, investigations on the dissolution behavior of both PuN and (Pu,Zr)N has been undertaken. If possible it is desirable to perform the fuel dissolutions using nitric acid. This, as most reprocessing strategies using solvent-solvent extraction are based on a nitride containing aqueous matrix. (Pu,Zr)N/C microspheres were produced using internal gelation. The spheres dissolution performance was investigated using nitric acid with and without additions of HF and Ag(II). In addition PuN fuel pellets were produced from powder and their dissolution performance were also assessed in a nitric acid based setting. It appears that both PuN and (Pu,Zr)N/C fuel material can be completely dissolved in nitric acid of high concentration with the use of catalytic amounts of HF. The amount of HF added strongly affects dissolution kinetics of (Pu, Zr)N and the presence of HF affects the 2 solutes differently, possibly due to inhomogeneity o the initial material. Large additions of Ag(II) can also be used to facilitate the dissolution of (Pu,Zr)N in nitric acid. PuN can be dissolved by pure nitric acid of high concentration at room temperature while (Pu, Zr)N is unaffected under similar conditions. At elevated temperature (reflux), (Pu,Zr)N can, however, also be dissolved by concentrated pure nitric acid.

Effect of compressive force on PEM fuel cell performance

Science.gov (United States)

MacDonald, Colin Stephen

Polymer electrolyte membrane (PEM) fuel cells possess the potential, as a zero-emission power source, to replace the internal combustion engine as the primary option for transportation applications. Though there are a number of obstacles to vast PEM fuel cell commercialization, such as high cost and limited durability, there has been significant progress in the field to achieve this goal. Experimental testing and analysis of fuel cell performance has been an important tool in this advancement. Experimental studies of the PEM fuel cell not only identify unfiltered performance response to manipulation of variables, but also aid in the advancement of fuel cell modelling, by allowing for validation of computational schemes. Compressive force used to contain a fuel cell assembly can play a significant role in how effectively the cell functions, the most obvious example being to ensure proper sealing within the cell. Compression can have a considerable impact on cell performance beyond the sealing aspects. The force can manipulate the ability to deliver reactants and the electrochemical functions of the cell, by altering the layers in the cell susceptible to this force. For these reasons an experimental study was undertaken, presented in this thesis, with specific focus placed on cell compression; in order to study its effect on reactant flow fields and performance response. The goal of the thesis was to develop a consistent and accurate general test procedure for the experimental analysis of a PEM fuel cell in order to analyse the effects of compression on performance. The factors potentially affecting cell performance, which were a function of compression, were identified as: (1) Sealing and surface contact; (2) Pressure drop across the flow channel; (3) Porosity of the GDL. Each factor was analysed independently in order to determine the individual contribution to changes in performance. An optimal degree of compression was identified for the cell configuration in

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2015

Energy Technology Data Exchange (ETDEWEB)

Eudy, Leslie [National Renewable Energy Lab. (NREL), Golden, CO (United States); Post, Matthew [National Renewable Energy Lab. (NREL), Golden, CO (United States); Gikakis, Christina [Federal Transit Administration, Washington, DC (United States)

2015-12-11

This report, published annually, summarizes the progress of fuel cell electric bus (FCEB) development in the United States and discusses the achievements and challenges of introducing fuel cell propulsion in transit. Various stakeholders, including FCEB developers, transit agencies, and system integrators, have expressed the value of this annual status report, which provides a summary of results from evaluations performed by the National Renewable Energy Laboratory. The annual status report tracks the progress of the FCEB industry toward meeting technical targets, documents the lessons learned, and discusses the path forward for commercial viability of fuel cell technology for transit buses. The 2015 summary results primarily focus on the most recent year for each demonstration, from August 2014 through July 2015. The results for these buses account for more than 1,045,000 miles traveled and 83,000 hours of fuel cell power system operation. The primary results presented in the report are from two demonstrations of fuel-cell-dominant bus designs: the Zero Emission Bay Area Demonstration Group led by Alameda-Contra Costa Transit District (AC Transit) in California and the American Fuel Cell Bus Project at SunLine Transit Agency in California.

Performance of commercially produced mixed-oxide fuels in EBR-II

International Nuclear Information System (INIS)

Hales, J.W.; Lawrence, L.A.

1980-11-01

Commercially produced fuels for the Fast Flux Test Facility (FFTF) were irradiated in EBR-II under conditions of high cladding temperature (approx. 700 0 C) and low power (approx. 200 W/cm) to verify that manufacturing processes did not introduce variables which significantly affect general fuel performance. Four interim examinations and a terminal examination were completed to a peak burnup of 5.2 at. % to provide irradiation data pertaining to fuel restructuring and dimensional stability at low fuel temperature, fuel-cladding reactions at high cladding temperature and general fuel behavior. The examinations indicate completely satisfactory irradiation performance for low heat rates and high cladding temperatures to 5.2 at. % burnup

Experience of safety and performance improvement for fuel handling equipment

International Nuclear Information System (INIS)

Gyoon Chang, Sang; Hee Lee, Dae

2014-01-01

The purpose of this study is to provide experience of safety and performance improvement of fuel handling equipment for nuclear power plants in Korea. The fuel handling equipment, which is used as an important part of critical processes during the refueling outage, has been improved to enhance safety and to optimize fuel handling procedures. Results of data measured during the fuel reloading are incorporated into design changes. The safety and performance improvement for fuel handling equipment could be achieved by simply modifying the components and improving the interlock system. The experience provided in this study can be useful lessons for further improvement of the fuel handling equipment. (authors)

Fuel performance analysis for the HAMP-1 mini plate test

Energy Technology Data Exchange (ETDEWEB)

Cho, Byoung Jin; Tahka, Y. W.; Yim, J. S.; Lee, B. H. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

U-7wt%Mo/Al- 5wt%Si dispersion fuel with 8gU/cm{sup 3} is chosen to achieve more efficiency and higher performance than the conventional U{sub 3}Si{sub 2} fuel. As part of the fuel qualification program for the KiJang research reactor (KJRR), three irradiation tests with mini-plates are on the way at the High-flux Advanced Neutron Application Reactor (HANARO). The first test among three HANARO Mini-Plate Irradiation tests (HAMP-1, 2, 3) has completed. PLATE code has been initially developed to analyze the thermal performance of high density U-Mo/Al dispersion fuel plates during irradiation [1]. We upgraded the PLATE code with the latest irradiation results which were implemented by corrosion, thermal conductivity and swelling model. Fuel performance analysis for HAMP-1 was conducted with updated PLATE. This paper presents results of performance evaluation of the HAMP-1. Maximum fuel temperature was obtained 136 .deg., which is far below the preset limit of 200 .deg. for the irradiation test. The meat swelling and corrosion thickness was also confirmed that the developed fuel would behave as anticipated.

Review of cell performance in anion exchange membrane fuel cells

Science.gov (United States)

Dekel, Dario R.

2018-01-01

Anion exchange membrane fuel cells (AEMFCs) have recently received increasing attention since in principle they allow for the use of non-precious metal catalysts, which dramatically reduces the cost per kilowatt of power in fuel cell devices. Until not long ago, the main barrier in the development of AEMFCs was the availability of highly conductive anion exchange membranes (AEMs); however, improvements on this front in the past decade show that newly developed AEMs have already reached high levels of conductivity, leading to satisfactory cell performance. In recent years, a growing number of research studies have reported AEMFC performance results. In the last three years, new records in performance were achieved. Most of the literature reporting cell performance is based on hydrogen-AEMFCs, although an increasing number of studies have also reported the use of fuels others than hydrogen - such as alcohols, non-alcohol C-based fuels, as well as N-based fuels. This article reviews the cell performance and performance stability achieved in AEMFCs through the years since the first reports in the early 2000s.

U.S. Department of Energy Hydrogen and Fuel Cells Program, 2013 Annual Merit Review and Peer Evaluation Report (Book)

Energy Technology Data Exchange (ETDEWEB)

None, None

2013-10-01

The fiscal year (FY) 2013 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting (AMR), in conjunction with DOE's Vehicle Technologies Office AMR, was held from May 13-16, 2013, at the Crystal City Marriott and Crystal Gateway Marriott in Arlington, Virginia. This report is a summary of comments by AMR peer reviewers about the hydrogen and fuel cell projects funded by DOE's Office of Energy Efficiency and Renewable Energy (EERE).

The irradiation performance of austenitic stainless steel clade PWR fuel rods

International Nuclear Information System (INIS)

Teixeira e Silva, A.; Esteves, A.M.

1988-01-01

The steady state irradiation performance of austenitic stainless steel clad pressurized water reactor fuel rods is modeled with fuel performance codes of the FRAP series. These codes, originally developed to model the thermal-mechanical behavior of zircaloy clad fuel rods, are modified to model stainless steel clad fuel rods. The irradiation thermal-mechanical behavior of type 348 stainless steel and zircaloy fuel rods is compared. (author) [pt

Key differences in the fabrication of US and German TRISO-coated particle fuel, and their implications on fuel performance

International Nuclear Information System (INIS)

Petti, D.A.; Buongiorno, J.; Maki, J.T.; Miller, G.K.; Hobbins, R.R.

2002-01-01

Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the US. German fuel generally displayed in-pile gas release values that were three orders of magnitude lower than US fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the US and German and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the US fuel has not faired as well, and what process/production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer US irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, degree of acceleration, power per particle) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance. (author)

Fuel Performance Calculations for FeCrAl Cladding in BWRs

Energy Technology Data Exchange (ETDEWEB)

George, Nathan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Sweet, Ryan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Maldonado, G. Ivan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Wirth, Brian D. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; 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)

2015-01-01

This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.

Performance and Exhaust Emissions in a Natural-Gas Fueled Dual-Fuel Engine

Science.gov (United States)

Shioji, Masahiro; Ishiyama, Takuji; Ikegami, Makoto; Mitani, Shinichi; Shibata, Hiroaki

In order to establish the optimum fueling in a natural gas fueled dual fuel engine, experiments were done for some operational parameters on the engine performances and the exhaust emissions. The results show that the pilot fuel quantity should be increased and its injection timing should be advanced to suppress unburned hydrocarbon emission in the middle and low output range, while the quantity should be reduced and the timing retarded to avoid onset of knock at high loads. Unburned hydrocarbon emission and thermal efficiency are improved by avoiding too lean natural gas mixture by restricting intake charge air. However, the improvement is limited because the ignition of pilot fuel deteriorates with excessive throttling. It is concluded that an adequate combination of throttle control and equivalence ratio ensures low hydrocarbon emission and the thermal efficiency comparable to diesel operation.

Fuel Performance Modeling of U-Mo Dispersion Fuel: The thermal conductivity of the interaction layers of the irradiated U-Mo dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Mistarhi, Qusai M.; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

2016-05-15

U-Mo/Al dispersion fuel performed well at a low burn-up. However, higher burn-up and higher fission rate irradiation testing showed enhanced fuel meat swelling which was caused by high interaction layer growth and pore formation. The performance of the dispersion type fuel in the irradiation and un-irradiation environment is very important. During the fabrication of the dispersion type fuel an Interaction Layer (IL) is formed due to the inter-diffusion between the U-Mo fuel particles and the Al matrix which is an intermetallic compound (U,Mo)Alx. During irradiation, the IL becomes amorphous causing a further decrease in the thermal conductivity and an increase in the centerline temperature of the fuel meat. Several analytical models and numerical methods were developed to study the performance of the unirradiated U-Mo/Al dispersion fuel. Two analytical models were developed to study the performance of the irradiated U-Mo/Al dispersion fuel. In these models, the thermal conductivity of the IL was assumed to be constant. The properties of the irradiated U-Mo dispersion fuel have been investigated recently by Huber et al. The objective of this study is to develop a correlation for IL thermal conductivity during irradiation as a function of the temperature and fission density from the experimentally measured thermal conductivity of the irradiated U-Mo/Al dispersion fuel. The thermal conductivity of IL during irradiation was calculated from the experimentally measured data and a correlation was developed from the thermal conductivity of IL as a function of T and fission density.

Heterogeneous fuels for minor actinides transmutation: Fuel performance codes predictions in the EFIT case study

Energy Technology Data Exchange (ETDEWEB)

Calabrese, R., E-mail: rolando.calabrese@enea.i [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Vettraino, F.; Artioli, C. [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Sobolev, V. [SCK.CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium); Thetford, R. [Serco Technical and Assurance Services, 150 Harwell Business Centre, Didcot OX11 0QB (United Kingdom)

2010-06-15

Plutonium recycling in new-generation fast reactors coupled with minor actinides (MA) transmutation in dedicated nuclear systems could achieve a decrease of nuclear waste long-term radiotoxicity by two orders of magnitude in comparison with current once-through strategy. In a double-strata scenario, purpose-built accelerator-driven systems (ADS) could transmute minor actinides. The innovative nuclear fuel conceived for such systems demands significant R and D efforts in order to meet the safety and technical performance of current fuel systems. The Integrated Project EUROTRANS (EUROpean research programme for the TRANSmutation of high level nuclear waste in ADS), part of the EURATOM Framework Programme 6 (FP6), undertook some of this research. EUROTRANS developed from the FP5 research programmes on ADS (PDS-XADS) and on fuels dedicated to MA transmutation (FUTURE, CONFIRM). One of its main objectives is the conceptual design of a small sub-critical nuclear system loaded with uranium-free fuel to provide high MA transmutation efficiency. These principles guided the design of EFIT (European Facility for Industrial Transmutation) in the domain DESIGN of IP EUROTRANS. The domain AFTRA (Advanced Fuels for TRAnsmutation system) identified two composite fuel systems: a ceramic-ceramic (CERCER) where fuel particles are dispersed in a magnesia matrix, and a ceramic-metallic (CERMET) with a molybdenum matrix in the place of MgO matrix to host a ceramic fissile phase. The EFIT fuel is composed of plutonium and MA oxides in solid solution with isotopic vectors typical of LWR spent fuel with 45 MWd/kg{sub HM} discharge burnup and 30 years interim storage before reprocessing. This paper is focused on the thermomechanical state of the hottest fuel pins of two EFIT cores of 400 MW{sub (th)} loaded with either CERCER or CERMET fuels. For calculations three fuel performance codes were used: FEMALE, TRAFIC and TRANSURANUS. The analysis was performed at the beginning of fuel life

High performance direct methanol fuel cell with thin electrolyte membrane

Science.gov (United States)

Wan, Nianfang

2017-06-01

A high performance direct methanol fuel cell is achieved with thin electrolyte membrane. 320 mW cm-2 of peak power density and over 260 mW cm-2 at 0.4 V are obtained when working at 90 °C with normal pressure air supply. It is revealed that the increased anode half-cell performance with temperature contributes primarily to the enhanced performance at elevated temperature. From the comparison of iR-compensated cathode potential of methanol/air with that of H2/air fuel cell, the impact of methanol crossover on cathode performance decreases with current density and becomes negligible at high current density. Current density is found to influence fuel efficiency and methanol crossover significantly from the measurement of fuel efficiency at different current density. At high current density, high fuel efficiency can be achieved even at high temperature, indicating decreased methanol crossover.

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Choi, Jin Ha; Kim, Myung Hyun

2016-01-01

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Transparency associated with the nuclear fuel cycle

International Nuclear Information System (INIS)

2009-01-01

This document presents the different fuel cycle stages with which the CEA is associated, the annual flow of materials and wastes produced at these different stages, and the destiny of these produced materials and wastes. These information are given for the different CEA R and D activities: experimentation hot laboratories (activities, fuel cycle stages, list of laboratories, tables giving annual flows for each of them), research reactors (types of reactors, fuel usage modes, annual flows of nuclear materials for each reactor), spent fuel management (different types of used materials), spent fuels and radioactive wastes with a foreign origin (quantities, processes)

Preferences for Alternative Fuel Vehicles of Lease Car Drivers in The Netherlands

Energy Technology Data Exchange (ETDEWEB)

Koetse, M.J.; Hoen, A.

2012-04-15

In this paper we aim to get insight into preferences of Dutch lease car drivers for alternative fuel vehicles (AFVs) and their characteristics. Since AFVs are either not yet available on the market or have only very limited market shares, we have to rely on stated preference research. We perform a state-of-the-art conjoint analysis, based on data obtained through an online choice experiment among Dutch lease car drivers. Results show that under current tax regulations the average lease car driver is indifferent between the conventional technology, flexifuel and the hybrid car, while negative preferences exist for the plug-in hybrid, the electric and the fuel cell car. When current tax regulations would be abolished, strong negative preferences would result for all AFCs, and especially for the electric and fuel cell car. Increases in driving range, reductions in refuelling time, and reductions in additional detour time for reaching an appropriate fuel station, increase AFV preferences substantially. On average the gap between conventional technologies and AFVs remains large, however. We also find that there is considerable heterogeneity in preferences of lease car drivers, and that various market segments and potential early adopters can be identified. In this respect the most interesting finding is that preferences for electric and fuel cell cars decrease substantially, and willingness to pay for driving range increases substantially, when annual mileage increases. Annual mileage also has a substantial impact on sensitivity to monthly costs. We therefore use simulations to assess market shares of electric and fuel cell cars for different annual mileage categories. We find that people with a relatively low annual mileage are more likely to adopt than people with a relatively high annual mileage, regardless of driving range and monthly costs. For the fuel cell car we find similar results, although when driving range is high and cost differences are large, lease car

Impact on burnup performance of coated particle fuel design in pebble bed reactor with ROX fuel

International Nuclear Information System (INIS)

Ho, Hai Quan; Obara, Toru

2015-01-01

The pebble bed reactor (PBR), a kind of high-temperature gas-cooled reactor (HTGR), is expected to be among the next generation of nuclear reactors as it has excellent passive safety features, as well as online refueling and high thermal efficiency. Rock-like oxide (ROX) fuel has been studied at the Japan Atomic Energy Agency (JAEA) as a new once-through type fuel concept. Rock-like oxide used as fuel in a PBR can be expected to achieve high burnup and improve chemical stabilities. In the once-through fuel concept, the main challenge is to achieve as high a burnup as possible without failure of the spent fuel. The purpose of this study was to investigate the impact on burnup performance of different coated fuel particle (CFP) designs in a PBR with ROX fuel. In the study, the AGR-1 Coated Particle design and Deep-Burn Coated Particle design were used to make the burnup performance comparison. Criticality and core burnup calculations were performed by MCPBR code using the JENDL-4.0 library. Results at equilibrium showed that the two reactors utilizing AGR-1 Coated Particle and Deep-Burn Coated Particle designs could be critical with almost the same multiplication factor k eff . However, the power peaking factor and maximum power per fuel ball in the AGR-1 coated particle design was lower than that of Deep-Burn coated particle design. The AGR-1 design also showed an advantage in fissions per initial fissile atoms (FIFA); the AGR-1 coated particle design produced a higher FIFA than the Deep-Burn coated particle design. These results suggest that the difference in coated particle fuel design can have an effect on the burnup performance in ROX fuel. (author)

Effects of ambient conditions on fuel cell vehicle performance

Science.gov (United States)

Haraldsson, K.; Alvfors, P.

Ambient conditions have considerable impact on the performance of fuel cell hybrid vehicles. Here, the vehicle fuel consumption, the air compressor power demand, the water management system and the heat loads of a fuel cell hybrid sport utility vehicle (SUV) were studied. The simulation results show that the vehicle fuel consumption increases with 10% when the altitude increases from 0 m up to 3000 m to 4.1 L gasoline equivalents/100 km over the New European Drive Cycle (NEDC). The increase is 19% on the more power demanding highway US06 cycle. The air compressor is the major contributor to this fuel consumption increase. Its load-following strategy makes its power demand increase with increasing altitude. Almost 40% of the net power output of the fuel cell system is consumed by the air compressor at the altitude of 3000 m with this load-following strategy and is thus more apparent in the high-power US06 cycle. Changes in ambient air temperature and relative humidity effect on the fuel cell system performance in terms of the water management rather in vehicle fuel consumption. Ambient air temperature and relative humidity have some impact on the vehicle performance mostly seen in the heat and water management of the fuel cell system. While the heat loads of the fuel cell system components vary significantly with increasing ambient temperature, the relative humidity did not have a great impact on the water balance. Overall, dimensioning the compressor and other system components to meet the fuel cell system requirements at the minimum and maximum expected ambient temperatures, in this case 5 and 40 °C, and high altitude, while simultaneously choosing a correct control strategy are important parameters for efficient vehicle power train management.

Electric power annual 1992

Energy Technology Data Exchange (ETDEWEB)

1994-01-06

The Electric Power Annual presents a summary of electric utility statistics at national, regional and State levels. The objective of the publication is to provide industry decisionmakers, government policymakers, analysts and the general public with historical data that may be used in understanding US electricity markets. The Electric Power Annual is prepared by the Survey Management Division; Office of Coal, Nuclear, Electric and Alternate Fuels; Energy Information Administration (EIA); US Department of Energy. ``The US Electric Power Industry at a Glance`` section presents a profile of the electric power industry ownership and performance, and a review of key statistics for the year. Subsequent sections present data on generating capability, including proposed capability additions; net generation; fossil-fuel statistics; retail sales; revenue; financial statistics; environmental statistics; electric power transactions; demand-side management; and nonutility power producers. In addition, the appendices provide supplemental data on major disturbances and unusual occurrences in US electricity power systems. Each section contains related text and tables and refers the reader to the appropriate publication that contains more detailed data on the subject matter. Monetary values in this publication are expressed in nominal terms.

U.S. Department of Energy Hydrogen and Fuel Cells Program 2012 Annual Merit Review and Peer Evaluation Report: May 14-18, 2012, Arlington, VA

Energy Technology Data Exchange (ETDEWEB)

2012-09-01

This document summarizes the comments provided by peer reviewers on hydrogen and fuel cell projects presented at the fiscal year (FY) 2012 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program and Vehicle Technologies Program Annual Merit Review and Peer Evaluation Meeting (AMR), held May 14-18, 2012, in Arlington, VA.

Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

Energy Technology Data Exchange (ETDEWEB)

Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

1976-04-01

Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

Development of MHI PWR fuel assembly with high thermal performance

International Nuclear Information System (INIS)

Yasushi Makino; Masaya Hoshi; Masaji Mori; Hidetoshi Kido; Kazuo Ikeda

2005-01-01

Mitsubishi Heavy Industries, Ltd. (MHI) has been developing a PWR fuel assembly to meet the needs of Japanese fuel market with mainly improving its reliability such as a mechanical strength, a seismic strength and endurance. For burn-up extension of the fuel to 55 GWd/t, MHI has introduced a Zircaloy spacer grid with better neutron economics with retaining the reliability in an operating core. However, for a future power up-rating and a longer cycle operation, a higher thermal performance is required for PWR fuel assembly. To meet the needs of fuel market, MHI has developed an advanced type of Zircaloy spacer grid with a greater DNB performance while retaining the reliability of a fuel and a relatively low pressure drop. For the greater DNB performance, MHI optimized geometrical shape of mixing vane to promote a fluid mixing performance. In this report, higher DNB performance provided by the advanced Zircaloy spacer grid is presented. The results of 3D simulation for the flow behavior in 5 x 5 partial assembly, a mixing test and a water DNB test were compared between the current and the advanced spacer grids. Consequently, it was confirmed that a crossover vane enhanced a fluid mixing and the advanced spacer grid could significantly improve DNB performance compared with the current design of spacer grids. (authors)

MOX and UOX PWR fuel performances EDF operating experience

International Nuclear Information System (INIS)

Provost, Jean-Luc; Debes, Michel

2005-01-01

Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

The neutronic and fuel cycle performance of interchangeable 3500 MWth metal and oxide fueled LMRs

International Nuclear Information System (INIS)

Fujita, E.K.; Wade, D.C.

1990-01-01

This study summarizes the neutronic and fuel cycle analysis performed at Argonne National Laboratory for an oxide and a metal fueled 3500 MWth LMR. These reactor designs formed the basis for a joint US/European study of LMR ATWS events. The oxide and metal core designs were developed to meet reactor performance specifications that are constrained by requirements for core loading interchangeability and for a small burnup reactivity swing. Differences in the computed performance parameters of the oxide and metal cores, arising from basic differences in their neutronic characteristics, are identified and discussed. It is shown that metal and oxide cores designed to the same ground rules exhibit many similar performance characteristics; however, they differ substantially in reactivity coefficients, control strategies, and fuel cycle options. 12 refs., 2 figs., 12 tabs

Features of fuel performance at high fuel burnups

International Nuclear Information System (INIS)

Proselkov, V.N.; Scheglov, A.S.; Smirnov, A.V.; Smirnov, V.P.

2001-01-01

Some features of fuel behavior at high fuel burnups, in particular, initiation and development of rim-layer, increase in the rate of fission gas release from the fuel and increase in the inner gas pressure in the fuel rod are briefly described. Basing on the analysis of the data of post-irradiation examinations of fuel rods of WWER-440 working FA and CR fuel followers, that have been operated for five fuel cycles and got the average fuel burnup or varies as 50MW-day/kgU, a conclusion is made that the WWER-440 fuel burnup can be increased at least to average burnups of 55-58 MW-day/kgU per fuel assembly (Authors)

Fuel oil and kerosene sales 1994

International Nuclear Information System (INIS)

1995-01-01

This publication contains the 1994 survey results of the ''Annual Fuel Oil and Kerosene Sales Report'' (Form EIA-821). This is the sixth year that the survey data have appeared in a separate publication. Prior to the 1989 report, the statistics appeared in the Petroleum Marketing Annual (PMA)for reference year 1988 and the Petroleum Marketing Monthly (PMM) for reference years 1984 through 1987. The 1994 edition marks the 11th annual presentation of the results of the ongoing ''Annual Fuel Oil and Kerosene Sales Report'' survey. Distillate and residual fuel oil sales continued to move in opposite directions during 1994. Distillate sales rose for the third year in a row, due to a growing economy. Residual fuel oil sales, on the other hand, declined for the sixth year in a row, due to competitive natural gas prices, and a warmer heating season than in 1993. Distillate fuel oil sales increased 4.4 percent while residual fuel oil sales declined 1.6 percent. Kerosene sales decreased 1.4 percent in 1994

Westinghouse fuel manufacturing systems: a step change in performance improvements

International Nuclear Information System (INIS)

Mutyala, Meena

2009-01-01

Today's competitive electrical generation industry demands that nuclear power plant operators minimize total operating costs, including fuel cycle cost while maintaining flawless fuel performance. The mission of Westinghouse Nuclear Fuel is to be the industry's most responsive supplier of flawless, value added fuel products and services, as judged by our customers. As nuclear is fast becoming the choice of many countries, existing manufacturing plants and facilities are once again running at full capacity. In this context Westinghouse Nuclear Fuel is committed to deliver a step change in performance improvement worldwide through its manufacturing operations by the introduction of a set of fundamentals collectively named the 'Westinghouse Fuel Manufacturing System' (WFMS), whose key principles are discussed in this paper. (author)

Burnup performance of rock-like oxide (ROX) fuel in small pebble bed reactor with accumulative fuel loading scheme

International Nuclear Information System (INIS)

Simanullang, Irwan Liapto; Obara, Toru

2017-01-01

Highlights: • Burnup performance using ROX fuel in PBR with accumulative fuel loading scheme was analyzed. • Initial excess reactivity was suppressed by reducing 235 U enrichment in the startup condition. • Negative temperature coefficient was achieved in all condition of PBR with accumulative fuel loading scheme using ROX fuel. • Core lifetime of PBR with accumulative fuel loading scheme using ROX fuel was shorter than with UO 2 fuel. • In PBR with accumulative fuel loading scheme using ROX fuel, achieved discharged burnup can be as high as that for UO 2 fuel. - Abstract: The Japan Atomic Energy Agency (JAEA) has proposed rock-like oxide (ROX) fuel as a new, once-through type fuel concept. Here, burnup performance using ROX fuel was simulated in a pebble bed reactor with an accumulative fuel loading scheme. The MVP-BURN code was used to simulate the burnup calculation. Fuel of 5 g-HM/pebble with 20% 235 U enrichment was selected as the optimum composition. Discharged burnup could reach up to 218 GWd/t, with a core lifetime of about 8.4 years. However, high excess reactivity occurred in the initial condition. Initial fuel enrichment was therefore reduced from 20% to 4.65% to counter the initial excess reactivity. The operation period was reduced by the decrease of initial fuel enrichment, but the maximum discharged burnup was 198 GWd/t. Burnup performance of ROX fuel in this reactor concept was compared with that of UO 2 fuel obtained previously. Discharged burnup for ROX fuel in the PBR with an accumulative fuel loading scheme was as high as UO 2 fuel. Maximum power density could be lowered by introducing ROX fuel compared to UO 2 fuel. However, PBR core lifetime was shorter with ROX fuel than with UO 2 fuel. A negative temperature coefficient was achieved for both UO 2 and ROX fuels throughout the operation period.

Annual report 1990-91

International Nuclear Information System (INIS)

1991-01-01

The annual report of the Department of Atomic Energy (DAE), India, for the fiscal year 1990-91 describes: (1) the research and development work carried out by DAE's various research establishments and by fully funded research institutes, (2) performance of nuclear power stations, nuclear fuel complex, fuel reprocessing plants and heavy water plants, (3) exploration work for atomic minerals, (5) regulatory and safety functions, (6) performance of public sector undertakings, namely Indian Rare Earths Ltd., Electronics Corporation of India Ltd., and (6) other activities like international relations, emergency planning, financial assistance to research projects, etc. During the year, 6182 MUs electricity was generated, Unit-1 of Narora Atomic Power Station commenced commercial operation, Hazira Heavy Water Plant was commissioned, Sewage Hygienisation Research Irradiator at Baroda and Heavy Rare Earths Oxide Plant at Udyogmandal were also commissioned, promising geologic deposits of uranium were found at Domiasat (Meghalaya), and PURNIMA-III attained criticality. (M.G.B.)

Fuel oil and kerosene sales 1992

International Nuclear Information System (INIS)

1993-01-01

This publication contains the 1992 survey results of the ''Annual Fuel Oil and Kerosene Sales Report'' (Form EIA-821). This is the fourth year that the survey data have appeared in a separate publication. Prior to the 1989 report, the statistics appeared in the Petroleum Marketing Annual (PMA) for reference year 1988 and the Petroleum Marketing Monthly (PMM for reference years 1984 through 1987. The 1992 edition marks the ninth annual presentation of the results of the ongoing ''Annual Fuel Oil and Kerosene Sales Report'' survey. Except for the kerosene and on-highway diesel information, data presented in Tables 1 through 12 (Sales of Fuel Oil and Kerosene) present results of the EIA-821 survey. Tables 13 through 24 (Adjusted Sales of Fuel Oil and Kerosene) include volumes that are based on the EIA-821 survey but have been adjusted to equal the products supplied volumes published in the Petroleum Supply Annual (PSA)

SGHWR fuel performance, safety and reliability

International Nuclear Information System (INIS)

Pickman, D.O.; Inglis, G.H.

1977-05-01

The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith SGHWR that the necessary bias towards safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications. By a detailed evaluation of SGHWR fuel defects it is shown that very few defects can be shown to be related to design, rating, or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that most be met to achieve the desirable reliability level. It is possible that large scale experience on SGHWR fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to SGHWR fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. the philosophy and supporting experimental work programme are outlines and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to the new 60-pin fuel element to be used in the commercial SGHWRs and to its comparison in design and performance aspects with the 36-pin element that has been used to date in the Winfrith SGHWR. (author)

TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

Energy Technology Data Exchange (ETDEWEB)

Powers, Jeffrey James [Univ. of California, Berkeley, CA (United States)

2011-11-30

This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated

Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu

2008-01-01

U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during irradiation tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particle size was increases. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth. (author)

BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

Energy Technology Data Exchange (ETDEWEB)

Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sweet, Ryan T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling. As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations

2011 Annual Criticality Safety Program Performance Summary

Energy Technology Data Exchange (ETDEWEB)

Andrea Hoffman

2011-12-01

The 2011 review of the INL Criticality Safety Program has determined that the program is robust and effective. The review was prepared for, and fulfills Contract Data Requirements List (CDRL) item H.20, 'Annual Criticality Safety Program performance summary that includes the status of assessments, issues, corrective actions, infractions, requirements management, training, and programmatic support.' This performance summary addresses the status of these important elements of the INL Criticality Safety Program. Assessments - Assessments in 2011 were planned and scheduled. The scheduled assessments included a Criticality Safety Program Effectiveness Review, Criticality Control Area Inspections, a Protection of Controlled Unclassified Information Inspection, an Assessment of Criticality Safety SQA, and this management assessment of the Criticality Safety Program. All of the assessments were completed with the exception of the 'Effectiveness Review' for SSPSF, which was delayed due to emerging work. Although minor issues were identified in the assessments, no issues or combination of issues indicated that the INL Criticality Safety Program was ineffective. The identification of issues demonstrates the importance of an assessment program to the overall health and effectiveness of the INL Criticality Safety Program. Issues and Corrective Actions - There are relatively few criticality safety related issues in the Laboratory ICAMS system. Most were identified by Criticality Safety Program assessments. No issues indicate ineffectiveness in the INL Criticality Safety Program. All of the issues are being worked and there are no imminent criticality concerns. Infractions - There was one criticality safety related violation in 2011. On January 18, 2011, it was discovered that a fuel plate bundle in the Nuclear Materials Inspection and Storage (NMIS) facility exceeded the fissionable mass limit, resulting in a technical safety requirement (TSR) violation. The

Manufacture of fuel and fuel channels and their performance in Indian PHWRS - an overview

International Nuclear Information System (INIS)

Kalidas, R.

2005-01-01

Nuclear Fuel Complex (NFC) at Hyderabad is a conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried out several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport- Demonstration of Approach and Results on Used Fuel Performance Characterization

Energy Technology Data Exchange (ETDEWEB)

Adkins, Harold [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Ken [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Koeppel, Brian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bignell, John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Flores, Gregg [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Jy-An [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sanborn, Scott [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Spears, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Klymyshyn, Nick [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2013-09-30

This document addresses Oak Ridge National Laboratory milestone M2FT-13OR0822015 Demonstration of Approach and Results on Used Nuclear Fuel Performance Characterization. This report provides results of the initial demonstration of the modeling capability developed to perform preliminary deterministic evaluations of moderate-to-high burnup used nuclear fuel (UNF) mechanical performance under normal conditions of storage (NCS) and normal conditions of transport (NCT) conditions. This report also provides results from the sensitivity studies that have been performed. Finally, discussion on the long-term goals and objectives of this initiative are provided.

Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

International Nuclear Information System (INIS)

McKinnon, M.A.; Doherty, A.L.

1997-06-01

This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release from the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors. Refs, figs, tabs.

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

DUPIC fuel performance from reactor physics viewpoint

International Nuclear Information System (INIS)

Choi, H.; Rhee, B.W.; Park, H.

1995-01-01

A preliminary study was performed for the evaluation of Stress Corrosion Cracking (SCC) parameters of nominal DUPIC fuel in CANDU reactor. For the reference 2-bundle shift refueling scheme, the predicted ramped power and power increase of the 43-element DUPIC fuel in the equilibrium core are below the SCC thresholds of CANDU natural uranium fuel. For 4-bundle shift refueling scheme, the envelope of element ramped power and power increase upon refueling are 8% and 44% higher than those of 2-bundle shift refueling scheme on the average, respectively, and both schemes are not expected to cause SCC failures. (author)

Influence of performance characteristic of a gaseous fuel supply system on hydrocarbon emissions of a dual-fuel engine

Energy Technology Data Exchange (ETDEWEB)

Ren, J.; Wang, Z.Y.; Zhong, H.; Hao, S.H. [Xi' an Jiaotong Univ., Dept. of Automobile Engineering, Xi' an (China)

2000-11-01

The performance of the gaseous fuel supply and its influence on hydrocarbon (HC) emissions of dual-fuel engines have been investigated. A new design of manifold respirators with mixers is also presented in the paper. The design of the gaseous fuel supply system has a great influence on HC emissions in the dual-fuel engine at light load. The problem of scavenging is discussed and solved by using the manifold respirators in the dual-fuel engine. It performs the function of retarding the gaseous fuel entry timing from the moment of intake valve opening, and its delaying effects have been measured and tested. Experimental results show that the manifold respirator gives the best performance in reducing HC emissions compared with a common pipe mixer and a respirator with bo miser. In addition, the mixing effects are sensitive to the mixer configuration. (Author)

Performance and fuel conversion efficiency of a spark ignition engine fueled with iso-butanol

International Nuclear Information System (INIS)

Irimescu, Adrian

2012-01-01

Highlights: ► Iso-butanol use in a port injection spark ignition engine. ► Fuel conversion efficiency calculated based on chassis dynamometer measurements. ► Combined study of engine efficiency and air–fuel mixture temperature. ► Excellent running characteristics with minor fuel system modifications. ► Up to 11% relative drop in part load efficiency due to incomplete fuel vaporization. -- Abstract: Alcohols are increasingly used as fuels for spark ignition engines. While ethanol is most commonly used, long chain alcohols such as butanol feature several advantages like increased heating value and reduced corrosive action. This study investigated the effect of fueling a port injection engine with iso-butanol, as compared to gasoline operation. Performance levels were maintained within the same limits as with the fossil fuel without modifications to any engine component. An additional electronic module was used for increasing fuel flow by extending the injection time. Fuel conversion efficiency decreased when the engine was fueled with iso-butanol by up to 9% at full load and by up to 11% at part load, calculated as relative values. Incomplete fuel evaporation was identified as the factor most likely to cause the drop in engine efficiency.

performance calculations of gadolinium oxide and boron nitride coated fuel

International Nuclear Information System (INIS)

Tanker, E.; Uslu, I.; Disbudak, H.; Guenduez, G.

1997-01-01

A comparative study was performed on the behaviour of natural uranium dioxide-gadolinium oxide mixture fuel and boron nitride coated low enriched fuel in a pressurized water reactor. A fuel element containing one burnable poison fuel pins was modeled with the computer code WIMS, and burn-up dependent critically, fissile isotope inventory and two dimensional power distribution were obtained. Calculations were performed for burnable poison fuels containing 5% and 10% gadolinium oxide and for those coated with 1μ,5μ and 10μ of boron nitride. Boron nitride coating was found superior to gadolinium oxide on account of its smoother criticality curve, lower power peaks and insignificant change in fissile isotope content

Addressing forecast uncertainty impact on CSP annual performance

Science.gov (United States)

Ferretti, Fabio; Hogendijk, Christopher; Aga, Vipluv; Ehrsam, Andreas

2017-06-01

This work analyzes the impact of weather forecast uncertainty on the annual performance of a Concentrated Solar Power (CSP) plant. Forecast time series has been produced by a commercial forecast provider using the technique of hindcasting for the full year 2011 in hourly resolution for Ouarzazate, Morocco. Impact of forecast uncertainty has been measured on three case studies, representing typical tariff schemes observed in recent CSP projects plus a spot market price scenario. The analysis has been carried out using an annual performance model and a standard dispatch optimization algorithm based on dynamic programming. The dispatch optimizer has been demonstrated to be a key requisite to maximize the annual revenues depending on the price scenario, harvesting the maximum potential out of the CSP plant. Forecasting uncertainty affects the revenue enhancement outcome of a dispatch optimizer depending on the error level and the price function. Results show that forecasting accuracy of direct solar irradiance (DNI) is important to make best use of an optimized dispatch but also that a higher number of calculation updates can partially compensate this uncertainty. Improvement in revenues can be significant depending on the price profile and the optimal operation strategy. Pathways to achieve better performance are presented by having more updates both by repeatedly generating new optimized trajectories but also more often updating weather forecasts. This study shows the importance of working on DNI weather forecasting for revenue enhancement as well as selecting weather services that can provide multiple updates a day and probabilistic forecast information.

High Performance Fuel Desing for Next Generation Pressurized Water Reactors

International Nuclear Information System (INIS)

Mujid S. Kazimi; Pavel Hejzlar

2006-01-01

The use of internally and externally cooled annular fuel rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and economic assessment. The investigation was conducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperature. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasibility issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density

Performance of bio fuels in diesel engines

International Nuclear Information System (INIS)

Nunez I, Manuel L; Prada V, Laura P

2007-01-01

This paper shows the preliminary results of pilot plant tests developed in oil catalytic hydrotreating process, where the crude palm oil or a mixture of crude palm oil and mineral diesel is treated with an injection of 99% pure hydrogen flux, in a fixed bed reactor at high pressures and temperatures, in a presence of Nickel Molybdenum catalyst supported on alumina bed. The main product of this process is a fuel (bio diesel) which has the same or better properties than the diesel obtained by petroleum refining. It has been made some performance fuel tests in diesel engine? with good results in terms of power, torque and fuel consumption, without any changes in engine configuration. Considering the characteristics of the Catalytic hydrotreated bio diesel compare to conventional diesel, both fuels have similar distillation range? however, bio diesel has better flash point, cetane index and thermal stability. Gas fuels (methane, ethane, and propane) CO 2 and water are the secondary products of the process.

Improving performance with accident tolerant-fuels

International Nuclear Information System (INIS)

Gomes, Daniel S.; Muniz, Rafael O.R.; Giovedi, Claudia; Universidade de São Paulo

2017-01-01

After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO 2 - Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density - above that supported by UO 2 - and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U 3 Si 2 , UN, and UC contain a higher uranium density and thermal conductivity than UO 2 , providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U 3 Si 2 , UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO 2 . These ATFs also have thermal conductivities approximately four times higher than that of UO 2 . A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO 2 - Zr as the fuel system of choice. (author)

Improving performance with accident tolerant-fuels

Energy Technology Data Exchange (ETDEWEB)

Gomes, Daniel S.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: rafael.orm@gmail.com, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Departamento de Engenharia Naval e Oceânica

2017-07-01

After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO{sub 2} - Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density - above that supported by UO{sub 2} - and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U{sub 3}Si{sub 2}, UN, and UC contain a higher uranium density and thermal conductivity than UO{sub 2}, providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U{sub 3}Si{sub 2}, UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO{sub 2}. These ATFs also have thermal conductivities approximately four times higher than that of UO{sub 2}. A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO{sub 2} - Zr as the fuel system of choice. (author)

Effects of UO2 fuel microstructure and density on fuel in-reactor performance

International Nuclear Information System (INIS)

Hansson, L.

1988-02-01

The volume changes of UO 2 fuel pellets, produced by neutron irradiation, can be characterized by two processes: fission spike induced densification through pore skrinkage and later fission produced induced swelling of UO 2 matrix. In-pile densification is controlled by the initial density and microstructure of the fuel, particularly by the pore size distribution. The extent of swelling depends mainly on the amount of fission products produced, but the fission gas release as well as the swelling may be reduced by increasing the grain size of UO 2 . Fabrication of fuel pellets having certain in-reactor properties requires detailed knowledge of the effects of individual fabrication parameters. The irradiation experience of fuels fabricated by using different conversion and pelletizing methods is extensive. Based on this experience, some general characteristics of stable/well-performing fuel microstructures have been summarized

Modelling fuel cell performance using artificial intelligence

Science.gov (United States)

Ogaji, S. O. T.; Singh, R.; Pilidis, P.; Diacakis, M.

Over the last few years, fuel cell technology has been increasing promisingly its share in the generation of stationary power. Numerous pilot projects are operating worldwide, continuously increasing the amount of operating hours either as stand-alone devices or as part of gas turbine combined cycles. An essential tool for the adequate and dynamic analysis of such systems is a software model that enables the user to assess a large number of alternative options in the least possible time. On the other hand, the sphere of application of artificial neural networks has widened covering such endeavours of life such as medicine, finance and unsurprisingly engineering (diagnostics of faults in machines). Artificial neural networks have been described as diagrammatic representation of a mathematical equation that receives values (inputs) and gives out results (outputs). Artificial neural networks systems have the capacity to recognise and associate patterns and because of their inherent design features, they can be applied to linear and non-linear problem domains. In this paper, the performance of the fuel cell is modelled using artificial neural networks. The inputs to the network are variables that are critical to the performance of the fuel cell while the outputs are the result of changes in any one or all of the fuel cell design variables, on its performance. Critical parameters for the cell include the geometrical configuration as well as the operating conditions. For the neural network, various network design parameters such as the network size, training algorithm, activation functions and their causes on the effectiveness of the performance modelling are discussed. Results from the analysis as well as the limitations of the approach are presented and discussed.

Modelling fuel cell performance using artificial intelligence

Energy Technology Data Exchange (ETDEWEB)

Ogaji, S.O.T.; Singh, R.; Pilidis, P.; Diacakis, M. [Power Propulsion and Aerospace Engineering Department, Centre for Diagnostics and Life Cycle Costs, Cranfield University (United Kingdom)

2006-03-09

Over the last few years, fuel cell technology has been increasing promisingly its share in the generation of stationary power. Numerous pilot projects are operating worldwide, continuously increasing the amount of operating hours either as stand-alone devices or as part of gas turbine combined cycles. An essential tool for the adequate and dynamic analysis of such systems is a software model that enables the user to assess a large number of alternative options in the least possible time. On the other hand, the sphere of application of artificial neural networks has widened covering such endeavours of life such as medicine, finance and unsurprisingly engineering (diagnostics of faults in machines). Artificial neural networks have been described as diagrammatic representation of a mathematical equation that receives values (inputs) and gives out results (outputs). Artificial neural networks systems have the capacity to recognise and associate patterns and because of their inherent design features, they can be applied to linear and non-linear problem domains. In this paper, the performance of the fuel cell is modelled using artificial neural networks. The inputs to the network are variables that are critical to the performance of the fuel cell while the outputs are the result of changes in any one or all of the fuel cell design variables, on its performance. Critical parameters for the cell include the geometrical configuration as well as the operating conditions. For the neural network, various network design parameters such as the network size, training algorithm, activation functions and their causes on the effectiveness of the performance modelling are discussed. Results from the analysis as well as the limitations of the approach are presented and discussed. (author)

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2012

Energy Technology Data Exchange (ETDEWEB)

Eudy, Leslie [National Renewable Energy Lab. (NREL), Golden, CO (United States); Chandler, Kevin [Battelle, Columbus, OH (United States); Gikakis, Christina [Federal Transit Administration, Washington, DC (United States)

2012-11-01

This report is the sixth in an annual series of reports that summarize the progress of fuel cell electric bus (FCEB) development in the United States and discuss the achievements and challenges of introducing fuel cell propulsion in transit. The report also provides a snapshot of current FCEB performance results over the last year.

Performance of a Fuel-Cell-Powered, Small Electric Airplane Assessed

Science.gov (United States)

Berton, Jeffrey J.

2004-01-01

Rapidly emerging fuel-cell-power technologies may be used to launch a new revolution of electric propulsion systems for light aircraft. Future small electric airplanes using fuel cell technologies hold the promise of high reliability, low maintenance, low noise, and - with the exception of water vapor - zero emissions. An analytical feasibility and performance assessment was conducted by NASA Glenn Research Center's Airbreathing Systems Analysis Office of a fuel-cell-powered, propeller-driven, small electric airplane based on a model of the MCR-01 two-place kitplane (Dyn'Aero, Darois, France). This assessment was conducted in parallel with an ongoing effort by the Advanced Technology Products Corporation and the Foundation for Advancing Science and Technology Education. Their project - partially funded by a NASA grant - is to design, build, and fly the first manned, continuously propelled, nongliding electric airplane. In our study, an analytical performance model of a proton exchange membrane (PEM) fuel cell propulsion system was developed and applied to a notional, two-place light airplane modeled after the MCR-01 kitplane. The PEM fuel cell stack was fed pure hydrogen fuel and humidified ambient air via a small automotive centrifugal supercharger. The fuel cell performance models were based on chemical reaction analyses calibrated with published data from the fledgling U.S. automotive fuel cell industry. Electric propeller motors, rated at two shaft power levels in separate assessments, were used to directly drive a two-bladed, variable-pitch propeller. Fuel sources considered were compressed hydrogen gas and cryogenic liquid hydrogen. Both of these fuel sources provided pure, contaminant-free hydrogen for the PEM cells.

Operational limitations of light water reactors relating to fuel performance

International Nuclear Information System (INIS)

Cheng, H.S.

1976-07-01

General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed

Sensitivity assessment of fuel performance codes for LOCA accident scenario

Energy Technology Data Exchange (ETDEWEB)

Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia; Martins, Marcelo, E-mail: ayabe@ipen.br, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LABRISCO/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco

2017-07-01

FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

Sensitivity assessment of fuel performance codes for LOCA accident scenario

International Nuclear Information System (INIS)

Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R.; Giovedi, Claudia; Martins, Marcelo

2017-01-01

FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

Proceedings of the Canadian Nuclear Society 12. annual conference

International Nuclear Information System (INIS)

1991-01-01

This volume contains the Proceedings of the seventeen Technical Sessions from the Twelfth Annual Conference of the Canadian Nuclear Society held in Saskatoon, Saskatchewan, June 9 to 12, 1991. As in previous years, the Annual Conference of the Canadian Nuclear Society was held in conjunction with the Annual Conference of the Canadian Nuclear Association. The major topics of discussion included: reactor physics; thermal hydraulics; industrial irradiation; computer applications; fuel channel analysis; small reactors; severe accidents; fuel behaviour under accident conditions; reactor components; safety related computer software; nuclear fuel management; nuclear waste management; and, uranium mining processing

High Performance Fuel Desing for Next Generation Pressurized Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Mujid S. Kazimi; Pavel Hejzlar

2006-01-31

The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

Performance of fuel system at different diesel temperature

Science.gov (United States)

Xu, Xiaoyong; Li, Xiaolu; Sun, Zai

2010-08-01

This paper presents the findings about performance of the fuel system of a diesel engine at different diesel temperature obtained through simulation and experiment. It can be seen from these findings that at the same rotational speed of fuel pump, the initial pressure in the fuel pipe remain unchanged as the fuel temperature increases, the peak pressure at the side of fuel pipe near the injector delays, and its largest value of pressure decreases. Meanwhile, at the same temperature, as the rotational speed increases, the initial pressure of fuel pipe is also essentially the same, the arrival of its peaks delays, and its largest value of pressure increases. The maximum fuel pressure at the side of fuel pipe near the injector has an increase of 28.9 %, 22.3%, and 13.9% respectively than the previous ones according to its conditions. At the same rotational speed, as the temperature increases, the injection quantity through the nozzle orifice decreases. At the same temperature, as the rotational speed increases, the injection quantity through the nozzle orifice increases. These experimental results are consistent with simulation results.

An innovative fuel design concept for improved Light Water Reactor performance and safety

International Nuclear Information System (INIS)

Tulenko, J.S.; Connell, R.G.

1993-01-01

The primary goal of this research is to develop a new fuel design which will have improved thermal/mechanical performance characteristics greatly superior to current thermal and mechanical design performance. The mechanical/thermal constraints define the lifetime of the fuel, the maximum power at which the fuel can be operated, the probability of fuel failure over core lifetime, and the integrity of a core during a transient excursion. The thermal/mechanical limits act to degrade fuel integrity when they are violated. The purpose of this project is to investigate a novel design for light water reactor fuel which will extend fuel performance limits and improve reactor safety even further than is currently achieved. This project is investigating liquid metal bonding of LWR fuel in order to radically decrease fuel centerline temperatures which has major performance and safety benefits. The project will verify the compatibility of the liquid metal bond with both the fuel pellets and cladding material, verify the performance enhancement features of the new design over the fuel lifetime, and verify the economic fabricability of the concept and will show how this concept will benefit the LWR nuclear industry

Fuel R and D international programmes, a way to demonstrate future fuel performances

International Nuclear Information System (INIS)

Vanderborck, Y.; Mertens, L.; Dekeyser, J.; Sannen, L.

1997-01-01

As a MOX fuel manufacturer, BELGONUCLEAIRE have spent more than 15 years promoting and managing International R and D Programmes, many of them in close cooperation with SCK''centrdot'' CEN. Such programmes dedicated to MOX versus UO 2 fuel behaviour are most of the time based on irradiation in research reactors in which the investigated fuel is submitted to power variations and to ramp testing or are performed in commercial reactors. This paper is focused on recent programmes concerned by high and medium burn-up in BWR and PWR conditions for MOX fuel. It will present also the new opportunities for new programmes. The goals, the programmes descriptions and the expected data being part of these R and D programmes is presented. (author)

Severe fuel-damage scoping test performance

International Nuclear Information System (INIS)

Gruen, G.E.; Buescher, B.J.

1983-01-01

As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. The first test of Phase I of this series has been successfully completed in the Power Burst Facility at the Idaho National Engineering Laboratory. Following the first test, calculations were performed using the TRAC-BD1 computer code with actual experimental boundary conditions. This paper discusses the test conduct and performance and presents the calculated and measured test bundle results. The test resulted in a slow heatup to 2000 K over about 4 h, with an accelerated reaction of the zirconium cladding at temperatures above 1600 K in the lower part or the bundle and 2000 K in the upper portion of the bundle

Performance of a pellet boiler fired with agricultural fuels

International Nuclear Information System (INIS)

Carvalho, Lara; Wopienka, Elisabeth; Pointner, Christian; Lundgren, Joakim; Verma, Vijay Kumar; Haslinger, Walter; Schmidl, Christoph

2013-01-01

Highlights: ► Performance evaluation of a pellet boiler operated with different agricultural fuels. ► Agricultural fuels could be burn in the tested boiler for a certain period of time. ► All the fuels (except straw and Sorghum) satisfied the European legal requirements. ► Boilers for burning agricultural fuels should have a flexible control system. - Abstract: The increasing demand for woody biomass increases the price of this limited resource, motivating the growing interest in using woody materials of lower quality as well as non-woody biomass fuels for heat production in Europe. The challenges in using non-woody biomass as fuels are related to the variability of the chemical composition and in certain fuel properties that may induce problems during combustion. The objective of this work has been to evaluate the technical and environmental performance of a 15 kW pellet boiler when operated with different pelletized biomass fuels, namely straw (Triticum aestivum), Miscanthus (Miscanthus × giganteus), maize (Zea mays), wheat bran, vineyard pruning (from Vitis vinifera), hay, Sorghum (Sorghum bicolor) and wood (from Picea abies) with 5% rye flour. The gaseous and dust emissions as well as the boiler efficiency were investigated and compared with the legal requirements defined in the FprEN 303-5 (final draft of the European standard 303-5). It was found that the boiler control should be improved to better adapt the combustion conditions to the different properties of the agricultural fuels. Additionally, there is a need for a frequent cleaning of the heat exchangers in boilers operated with agricultural fuels to avoid efficiency drops after short term operation. All the agricultural fuels satisfied the legal requirements defined in the FprEN 303-5, with the exception of dust emissions during combustion of straw and Sorghum. Miscanthus and vineyard pruning were the best fuels tested showing comparable emission values to wood combustion

Annual Performance Reports: 2002-2003 State Assessment Data

Science.gov (United States)

Thurlow, Martha L.; Moen, Ross E.; Wiley, Hilda I.

2005-01-01

States and other educational entities receiving Part B funding under the Individuals with Disabilities Education Act (IDEA) submitted their Annual Performance Reports to the U.S. Secretary of Education on or before March 31, 2004. These reports contained information on a variety of indicators, including assessment participation and performance…

The development of the fuel rod transient performance analysis code FTPAC

International Nuclear Information System (INIS)

Han Zhijie; Ji Songtao

2014-01-01

Fuel rod behavior, especially the integrity of cladding, played an important role in fuel safety research during reactor transient and hypothetical accidents conditions. In order to study fuel rod performance under transient accidents, FTPAC (Fuel Transient Performance Analysis Code) has been developed for simulating light water reactor fuel rod transient behavior when power or coolant boundary conditions are rapidly changing. It is composed of temperature, mechanical deformation, cladding oxidation and gas pressure model. The assessment was performed by comparing FTPAC code analysis result to experiments data and FRAPTRAN code calculations. Comparison shows that, the FTPAC gives reasonable agreement in temperature, deformation and gas pressure prediction. And the application of slip coefficient is more suitable for simulating the sliding between pellet and cladding when the gap is closed. (authors)

Performance concerns for high duty fuel cycle

International Nuclear Information System (INIS)

Esposito, V.J.; Gutierrez, J.E.

1999-01-01

One of the goals of the nuclear industry is to achieve economic performance such that nuclear power plants are competitive in a de-regulated market. The manner in which nuclear fuel is designed and operated lies at the heart of economic viability. In this sense reliability, operating flexibility and low costs are the three major requirements of the NPP today. The translation of these three requirements to the design is part of our work. The challenge today is to produce a fuel design which will operate with long operating cycles, high discharge burnup, power up-rating and while still maintaining all design and safety margins. European Fuel Group (EFG) understands that to achieve the required performance high duty/energy fuel designs are needed. The concerns for high duty design includes, among other items, core design methods, advanced Safety Analysis methodologies, performance models, advanced material and operational strategies. The operational aspects require the trade-off and evaluation of various parameters including coolant chemistry control, material corrosion, boiling duty, boron level impacts, etc. In this environment MAEF is the design that EFG is now offering based on ZIRLO alloy and a robust skeleton. This new design is able to achieve 70 GWd/tU and Lead Test Programs are being executed to demonstrate this capability. A number of performance issues which have been a concern with current designs have been resolved such as cladding corrosion and incomplete RCCA insertion (IRI). As the core duty becomes more aggressive other new issues need to be addressed such as Axial Offset Anomaly. These new issues are being addressed by combination of the new design in concert with advanced methodologies to meet the demanding needs of NPP. The ability and strategy to meet high duty core requirements, flexibility of operation and maintain acceptable balance of all technical issues is the discussion in this paper. (authors)

Castor-V/21 PWR spent fuel storage cask performance test

International Nuclear Information System (INIS)

Creer, J.M.; Schoonen, D.H.

1986-01-01

Performance testing of a CASTOR-V/21 PWR spent fuel storage cask manufactured by Gesellschaft fur Nuklear Service (GNS) was performed as part of a cooperative program between Virginia Power and the US Department of Energy. The performance test consisted of obtaining cask handling experience and heat transfer, shielding, and limited fuel integrity data. Five heat transfer test runs were performed with 21 Surry reactor spent fuel assemblies generating approximately 28 kW. Test runs were performed vacuum, nitrogen, and helium backfill environments with the cask in both vertical and horizontal orientations. Cask exterior surface gamma and neutron dose rates were measured with the cask fully loaded. Gas samples were obtained at the beginning and end of each run with nitrogen or helium environments to verify fuel integrity. The heat transfer performance of the CASTOR-V/21 cask was exceptionally good. Peak clad temperatures with helium and nitrogen environments with the cask in a vertical orientation and with helium with the cask in a horizontal orientation were less than 380 0 C. Vertical vacuum and horizontal nitrogen test runs resulted in peak clad temperatures over 380 0 , but the temperatures were not excessively high ( 0 C). The shielding performance of the cask met the design goal of less than 200 mrem/hr. Cask surface dose rates of <75 mrem/hr can easily be established with minor gamma shielding design refinements if desired. Gas samples obtained during testing indicated no leaking fuel rods were present in the cask. It was concluded that the cask performed satisfactorily from heat transfer and shielding perspectives

CASTOR-V/21 PWR spent fuel storage cask performance test

International Nuclear Information System (INIS)

Creer, J.M.; Schoonen, D.H.

1986-01-01

Performance testing of a CASTOR-V/21 PWR spent fuel storage cask manufactured by Gesellschaft fur Nuklear Service (GNS) was performed as part of a cooperative program between Virginia Power and the US Department of Energy. The performance test consisted of obtaining cask handling experience and heat transfer, shielding, and limited fuel integrity data. Five heat transfer test runs were performed with 21 Surry reactor spent fuel assemblies generating approximately 28 kW. Test runs were performed with vacuum, nitrogen, and helium backfills in both vertical and horizontal orientations. Cask exterior surface gamma and neutron dose rates were measured with the cask fully loaded. Gas samples were obtained at the beginning and end of each run with nitrogen or helium backfills to verify fuel integrity. The heat transfer performance of the CASTOR-V/21 cask was exceptionally good. Peak clad temperatures with helium and nitrogen backfills in a vertical orientation and with helium in a horizontal orientation were less than 380 0 C. Vertical vacuum and horizontal nitrogen runs resulted in peak clad temperatures over 380 0 , but the temperatures were not excessively high ( 0 C). The shielding performance of the cask met the design expectation of less than 200 mrem/h. Cask surface dose rates of <75 mrem/h can easily be established with minor gamma shielding design refinements if desired. Gas samples obtained during testing indicated no leaking fuel rods were present in the cask. It was concluded that the cask performed satisfactorily from heat transfer and shielding perspectives

High performance fuel technology development : Development of high performance cladding materials

International Nuclear Information System (INIS)

Park, Jeongyong; Jeong, Y. H.; Park, S. Y.

2012-04-01

The superior in-pile performance of the HANA claddings have been verified by the successful irradiation test and in the Halden research reactor up to the high burn-up of 67GWD/MTU. The in-pile corrosion and creep resistances of HANA claddings were improved by 40% and 50%, respectively, over Zircaloy-4. HANA claddings have been also irradiated in the commercial reactor up to 2 reactor cycles, showing the corrosion resistance 40% better than that of ZIRLO in the same fuel assembly. Long-term out-of-pile performance tests for the candidates of the next generation cladding materials have produced the highly reliable test results. The final candidate alloys were selected and they showed the corrosion resistance 50% better than the foreign advanced claddings, which is beyond the original target. The LOCA-related properties were also improved by 20% over the foreign advanced claddings. In order to establish the optimal manufacturing process for the inner and outer claddings of the dual-cooled fuel, 18 different kinds of specimens were fabricated with various cold working and annealing conditions. Based on the performance tests and various out-of-pile test results obtained from the specimens, the optimal manufacturing process was established for the inner and outer cladding tubes of the dual-cooled fuel

Annual energy performance of R744 and R410A heat pumping systems

International Nuclear Information System (INIS)

Jin, Zhequan; Eikevik, Trygve M.; Nekså, Petter; Hafner, Armin; Wang, Ruzhu

2017-01-01

Highlights: • Annual energy performance of R744 and R410A heat pumping systems is compared. • Several dynamic models of heat pumping systems are developed. • Annual energy efficiency of R744 hybrid ground-coupled heat pumping system. • The seasonal COPc and COPh of R744 hybrid system are 3.55 and 3.32. • The superiority of R744 system on the integration of two heat sinks is discussed. - Abstract: This work compares the annual energy performance of heat pumping systems using R744 and R410A as refrigerant. Focus is the annual energy efficiency of R744 hybrid ground-coupled heat pumping system. The hybrid system uses both ambient air and ground as heat sinks in the cooling mode. This is important to eliminate the underground heat accumulation phenomenon in warm climates. Several quasi-steady state models of heat pumping systems, using R744 and R410A, have been developed. Simulation results show that the annual COP_c and COP_h of an R744 hybrid system reaches 3.55 and 3.32, and its cooling performance is 42% better than for a R744 ASHP and 23% better than for a R744 GCHP system. The annual energy performance factor of a R410A ASHP system is better than for a R744 hybrid system, but the COP_c for the R410A system will be lower when the ambient temperature is higher than 30 °C.

Oxide thickness measurement for monitoring fuel performance at high burnup

International Nuclear Information System (INIS)

Jaeger, M.A.; Van Swam, L.F.P.; Brueck-Neufeld, K.

1991-01-01

For on-site monitoring of the fuel performance at high burnup, Advanced Nuclear Fuels uses the linear scan eddy current method to determine the oxide thickness of irradiated Zircaloy fuel cans. Direct digital data acquisition methods are employed to collect the data on magnetic storage media. This field-proven methodology allows oxide thickness measurements and rapid interpretation of the data during the reactor outages and makes it possible to immediately reinsert the assemblies for the next operating cycle. The accuracy of the poolside measurements and data acquisition/interpretation techniques have been verified through hot cell metallographic measurements of rods previously measured in the fuel pool. The accumulated data provide a valuable database against which oxide growth models have been benchmarked and allow for effective monitoring of fuel performance. (orig.) [de

Clean Cities 2013 Annual Metrics Report

Energy Technology Data Exchange (ETDEWEB)

Johnson, C.; Singer, M.

2014-10-01

Each year, the U.S. Department of Energy asks its Clean Cities program coordinators to submit annual reports of their activities and accomplishments for the previous calendar year. Data and information are submitted via an online database that is maintained as part of the Alternative Fuels Data Center (AFDC) at the National Renewable Energy Laboratory (NREL). Coordinators submit a range of data that characterize the membership, funding, projects, and activities of their coalitions. They also submit data about sales of alternative fuels, deployment of alternative fuel vehicles (AFVs) and hybrid electric vehicles (HEVs), idle-reduction (IR) initiatives, fuel economy activities, and programs to reduce vehicle miles traveled (VMT). NREL analyzes the data and translates them into petroleum-use reduction impacts, which are summarized in this 2013 Annual Metrics Report.

Clean Cities 2014 Annual Metrics Report

Energy Technology Data Exchange (ETDEWEB)

Johnson, Caley [National Renewable Energy Lab. (NREL), Golden, CO (United States); Singer, Mark [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2015-12-22

Each year, the U.S. Department of Energy asks its Clean Cities program coordinators to submit annual reports of their activities and accomplishments for the previous calendar year. Data and information are submitted via an online database that is maintained as part of the Alternative Fuels Data Center (AFDC) at the National Renewable Energy Laboratory (NREL). Coordinators submit a range of data that characterize the membership, funding, projects, and activities of their coalitions. They also submit data about sales of alternative fuels, deployment of alternative fuel vehicles (AFVs) and hybrid electric vehicles (HEVs), idle-reduction (IR) initiatives, fuel economy activities, and programs to reduce vehicle miles traveled (VMT). NREL analyzes the data and translates them into petroleum-use reduction impacts, which are summarized in this 2014 Annual Metrics Report.

Theoretical and experimental analysis of fast reactor fuel performance

International Nuclear Information System (INIS)

Kummerer, K.R.; Freund, D.; Steiner, H.

1982-09-01

In order to predict behavior, performance, and capability of prototypic fuel pins a standard operational scheme for the SNR-300 fast breeder reactor is established considering besides normal operation unscheduled power changes and shutdowns. The behavior during the whole lifetime is calculated using the updated SATURN codes and - for special conditions as power transients and skewed fuel rod power - the new TRANSIENT and TEXDIF codes. The results of these calculations are compared to experimental findings. It is demonstrated that the level of modeling and the knowledge of material properties under irradiation are sufficient for a quantitative description of the fuel pin performance under the above mentioned conditions. (orig.) [de

Theoretical performance of hydrogen-bromine rechargeable SPE fuel cell

Science.gov (United States)

Savinell, Robert F.; Fritts, S. D.

1987-01-01

A mathematical model was formulated to describe the performance of a hydrogen-bromine fuel cell. Porous electrode theory was applied to the carbon felt flow-by electrode and was coupled to theory describing the solid polymer electrolyte (SPE) system. Parametric studies using the numerical solution to this model were performed to determine the effect of kinetic, mass transfer, and design parameters on the performance of the fuel cell. The results indicate that the cell performance is most sensitive to the transport properties of the SPE membrane. The model was also shown to be a useful tool for scale-up studies.

Spent fuel performance in geologic repository environments

International Nuclear Information System (INIS)

Bradley, D.J.

1985-10-01

The performance assessment of the waste package is a current area of study in the United States program to develop a geologic repository for nuclear waste isolation. The waste package is presently envisioned as the waste form and its surrounding containers and possibly a packing material composed of crushed host rock or mixtures of that rock with clays. This waste package is tied to performance criteria set forth in recent legislation. It is the goal of the Civilian Radioactive Waste Management Program to obtain the necessary information on the waste package, in several geologic environments, to show that the waste package provides reasonable assurance of meeting established performance criteria. This paper discusses the United States program directed toward managing high-level radioactive waste, with emphasis on the current effort to define the behavior of irradiated spent fuel in repository groundwaters. Current studies are directed toward understanding the rate and nature (such as valence state, colloid form if any, solid phase controlling solubility) of radionuclide release from the spent fuel. Due to the strong interactive effect of radiation, thermal fields, and waste package components on this release, current spent fuel studies are being conducted primarily in the presence of waste package components over a wide range of potential environments

Final Report - Effects of Impurities on Fuel Cell Performance and Durability

Energy Technology Data Exchange (ETDEWEB)

Trent Molter

2012-08-18

This program is focused on the experimental determination of the effects of key hydrogen side impurities on the performance of PEM fuel cells. Experimental data has been leveraged to create mathematical models that predict the performance of PEM fuel cells that are exposed to specific impurity streams. These models are validated through laboratory experimentation and utilized to develop novel technologies for mitigating the effects of contamination on fuel cell performance. Results are publicly disseminated through papers, conference presentations, and other means.

WWER fuel performance, development, QA and future prospects at Loviisa NPS

Energy Technology Data Exchange (ETDEWEB)

Loesoenen, P [Imatran Voima Oy, Vantaa (Finland)

1994-12-31

The essential characteristics of Loviisa fuel service are presented. A brief description is given of the steps in developing the performance values including: burnup dependent linear heat rate for fuel rods, comparison of measured and calculated cladding creep down of some rods, development of average discharge burnup for unit 1 and unit 2, results from experimental irradiation of test rods in research reactors. Some noticeable design changes performed and in-reactor behaviour of the lead fuel assemblies are also discussed. The quality assurance of the fuel procurement and operation is explained. Future prospects are connected with power raise of Loviisa reactors, discontinuing of spent fuel transportation to Russia for final disposal and new WWER fuel deliverers. 4 figs., 1 tab., 8 refs.

The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

2012-10-15

Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 Degree-Sign C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated

SKB Annual Report 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-05-01

This is the annual report of the Swedish Nuclear Fuel and Waste Management Co (SKB). Part 1 of the report contains an overview of the SKB activities in different fields, and part 2 gives a description of the research and development work on nuclear waste disposal performed during 1996. Lectures and publications as well as reports issued during 1996 are listed in part 3, and summaries of the reports are listed in part 4. The task of SKB is to transport, store and dispose of the spent nuclear fuel and radioactive wastes from the nuclear power plants and to perform the research and development and other measures necessary for this work. SKB is the owner of CLAB, the Central Interim Storage Facility for spent fuel, located at Oskarshamn. CLAB was taken into operation in July 1985 and by the end of 1996 about 2500 tons of spent fuel have been received. At Forsmark the Final Repository for Radioactive Operational Waste (SFR) was taken into operation in April 1988. The repository is situated in crystalline rock under the Baltic Sea. SFR has currently a capacity of about 60000 m{sup 3} or waste. At the end of 1996 at total of 21000 m{sup 3} of waste has been deposited. Transportation from the reactor sites to CLAB and SFR is made by a specially designed ship, M/S Sigyn. The total cost for R,D and D during 1996 amounted to 124 MSEK (about 15 MUSD).

SKB Annual Report 1996

International Nuclear Information System (INIS)

1997-05-01

This is the annual report of the Swedish Nuclear Fuel and Waste Management Co (SKB). Part 1 of the report contains an overview of the SKB activities in different fields, and part 2 gives a description of the research and development work on nuclear waste disposal performed during 1996. Lectures and publications as well as reports issued during 1996 are listed in part 3, and summaries of the reports are listed in part 4. The task of SKB is to transport, store and dispose of the spent nuclear fuel and radioactive wastes from the nuclear power plants and to perform the research and development and other measures necessary for this work. SKB is the owner of CLAB, the Central Interim Storage Facility for spent fuel, located at Oskarshamn. CLAB was taken into operation in July 1985 and by the end of 1996 about 2500 tons of spent fuel have been received. At Forsmark the Final Repository for Radioactive Operational Waste (SFR) was taken into operation in April 1988. The repository is situated in crystalline rock under the Baltic Sea. SFR has currently a capacity of about 60000 m 3 or waste. At the end of 1996 at total of 21000 m 3 of waste has been deposited. Transportation from the reactor sites to CLAB and SFR is made by a specially designed ship, M/S Sigyn. The total cost for R,D and D during 1996 amounted to 124 MSEK (about 15 MUSD)

Solid Polymer Fuel Cells. Electrode and membrane performance studies

Energy Technology Data Exchange (ETDEWEB)

Moeller-Holst, S.

1996-12-31

This doctoral thesis studies aspects of fuel cell preparation and performance. The emphasis is placed on preparation and analysis of low platinum-loading solid polymer fuel cell (SPEC) electrodes. A test station was built and used to test cells within a wide range of real operating conditions, 40-150{sup o}C and 1-10 bar. Preparation and assembling equipment for single SPFCs was designed and built, and a new technique of spraying the catalyst layer directly onto the membrane was successfully demonstrated. Low Pt-loading electrodes (0.1 mg Pt/cm{sup 2}) prepared by the new technique exhibited high degree of catalyst utilization. The performance of single cells holding these electrodes is comparable to state-of-the-art SPFCs. Potential losses in single cell performance are ascribed to irreversibilities by analysing the efficiency of the Solid Oxide Fuel Cell by means of the second law of thermodynamics. The water management in membranes is discussed for a model system and the results are relevant to fuel cell preparation and performance. The new spray deposition technique should be commercially interesting as it involves few steps as well as techniques that are adequate for larger scale production. 115 refs., 43 figs., 18 tabs.

Solid Polymer Fuel Cells. Electrode and membrane performance studies

Energy Technology Data Exchange (ETDEWEB)

Moeller-Holst, S

1997-12-31

This doctoral thesis studies aspects of fuel cell preparation and performance. The emphasis is placed on preparation and analysis of low platinum-loading solid polymer fuel cell (SPEC) electrodes. A test station was built and used to test cells within a wide range of real operating conditions, 40-150{sup o}C and 1-10 bar. Preparation and assembling equipment for single SPFCs was designed and built, and a new technique of spraying the catalyst layer directly onto the membrane was successfully demonstrated. Low Pt-loading electrodes (0.1 mg Pt/cm{sup 2}) prepared by the new technique exhibited high degree of catalyst utilization. The performance of single cells holding these electrodes is comparable to state-of-the-art SPFCs. Potential losses in single cell performance are ascribed to irreversibilities by analysing the efficiency of the Solid Oxide Fuel Cell by means of the second law of thermodynamics. The water management in membranes is discussed for a model system and the results are relevant to fuel cell preparation and performance. The new spray deposition technique should be commercially interesting as it involves few steps as well as techniques that are adequate for larger scale production. 115 refs., 43 figs., 18 tabs.

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.

Research and development into power reactor fuel performance

International Nuclear Information System (INIS)

Notley, M.J.F.

1983-07-01

The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO 2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

Study on thermal performance and margins of BWR fuel elements

International Nuclear Information System (INIS)

Stosic, Zoran

1999-01-01

This paper contributes to developing a methodology of predicting and analyzing thermal performance and margins of Boiling Water Reactor (BWR) fuel assemblies under conditions of reaching high quality Boiling Crisis and subsequent post-dryout thermal hydraulics causing temperature excursion of fuel cladding. Operational margins against dryout and potential for increasing fuel performance with appropriate benefits are discussed. The philosophy of modeling with its special topics are demonstrated on the HECHAN (HEated CHannel ANalyzer) model as the state-of-art for thermal-hydraulics analysis of BWR fuel assemblies in pre- and post-dryout two-phase flow regimes. The scope of further work either being or has to be performed concerning implementation of new physical aspects, including domain extension of HECHAN model applications to the Pressurized Water Reactors (PWRs), is discussed. Finally, a comprehensive overview of the literature dealing with development of the model is given. (author)

Petroleum supply annual 1996: Volume 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-06-01

The Petroleum Supply Annual (PSA) contains information on the supply and disposition of crude oil and petroleum products. The publication reflects data that were collected from the petroleum industry during 1996 through annual and monthly surveys. The PSA is divided into two volumes. This first volume contains three sections: Summary Statistics, Detailed Statistics, and Refinery Capacity; each with final annual data. The summary statistics section show 16 years of data depicting the balance between supply, disposition and ending stocks for various commodities including crude oil, motor gasoline, distillate fuel oil, residual fuel oil, jet fuel propane/propylene, and liquefied petroleum gases. The detailed statistics section provide 1996 detailed statistics on supply and disposition, refinery operations, imports and exports, stocks, and transportation of crude oil and petroleum products. The refinery capacity contain listings of refineries and associated crude oil distillation and downstream capacities by State, as of January 1, 1997, as well as summaries of corporate refinery capacities and refinery storage capacities. In addition, refinery receipts of crude oil by method of transportation for 1996 are provided. Also included are fuels consumed at refineries, and lists of shutdowns, sales, reactivations, and mergers during 1995 and 1996. 16 figs., 59 tabs.

Petroleum supply annual 1996: Volume 1

International Nuclear Information System (INIS)

1997-06-01

The Petroleum Supply Annual (PSA) contains information on the supply and disposition of crude oil and petroleum products. The publication reflects data that were collected from the petroleum industry during 1996 through annual and monthly surveys. The PSA is divided into two volumes. This first volume contains three sections: Summary Statistics, Detailed Statistics, and Refinery Capacity; each with final annual data. The summary statistics section show 16 years of data depicting the balance between supply, disposition and ending stocks for various commodities including crude oil, motor gasoline, distillate fuel oil, residual fuel oil, jet fuel propane/propylene, and liquefied petroleum gases. The detailed statistics section provide 1996 detailed statistics on supply and disposition, refinery operations, imports and exports, stocks, and transportation of crude oil and petroleum products. The refinery capacity contain listings of refineries and associated crude oil distillation and downstream capacities by State, as of January 1, 1997, as well as summaries of corporate refinery capacities and refinery storage capacities. In addition, refinery receipts of crude oil by method of transportation for 1996 are provided. Also included are fuels consumed at refineries, and lists of shutdowns, sales, reactivations, and mergers during 1995 and 1996. 16 figs., 59 tabs

Performance study of sugar-yeast-ethanol bio-hybrid fuel cells

Science.gov (United States)

Jahnke, Justin P.; Mackie, David M.; Benyamin, Marcus; Ganguli, Rahul; Sumner, James J.

2015-05-01

Renewable alternatives to fossil hydrocarbons for energy generation are of general interest for a variety of political, economic, environmental, and practical reasons. In particular, energy from biomass has many advantages, including safety, sustainability, and the ability to be scavenged from native ecosystems or from waste streams. Microbial fuel cells (MFCs) can take advantage of microorganism metabolism to efficiently use sugar and other biomolecules as fuel, but are limited by low power densities. In contrast, direct alcohol fuel cells (DAFCs) take advantage of proton exchange membranes (PEMs) to generate electricity from alcohols at much higher power densities. Here, we investigate a novel bio-hybrid fuel cell design prepared using commercial off-the-shelf DAFCs. In the bio-hybrid fuel cells, biomass such as sugar is fermented by yeast to ethanol, which can be used to fuel a DAFC. A separation membrane between the fermentation and the DAFC is used to purify the fermentate while avoiding any parasitic power losses. However, shifting the DAFCs from pure alcohol-water solutions to filtered fermented media introduces complications related to how the starting materials, fermentation byproducts, and DAFC waste products affect both the fermentation and the long-term DAFC performance. This study examines the impact of separation membrane pore size, fermentation/fuel cell byproducts, alcohol and salt concentrations, and load resistance on fuel cell performance. Under optimized conditions, the performance obtained is comparable to that of a similar DAFC run with a pure alcohol-water mixture. Additionally, the modified DAFC can provide useable amounts of power for weeks.

An introduction to current modeling techniques in nuclear fuel performance analysis

International Nuclear Information System (INIS)

Faya, S.C.S.

1981-10-01

An overview of several important phenomena considered in fuel performance analysis, such as fuel restructuring, fuel swelling, fission gas release, pellet cracking, etc... are provided. All these phenomena are independent. For the purpose of illustration of integral fuel rod analysis the LIFE-1 computer code is discussed. (Author) [pt

Influence of the fuel and dosage on the performance of double-compartment microbial fuel cells.

Science.gov (United States)

Asensio, Y; Fernandez-Marchante, C M; Lobato, J; Cañizares, P; Rodrigo, M A

2016-08-01

This manuscript focuses on the evaluation of the use of different types and dosages of fuels in the performance of double-compartment microbial fuel cell equipped with carbon felt electrodes and cationic membrane. Five types of fuels (ethanol, glycerol, acetate, propionate and fructose) have been tested for the same organic load (5,000 mg L(-1) measured as COD) and for one of them (acetate), the range of dosages between 500 and 20,000 mg L(-1) of COD was also studied. Results demonstrate that production of electricity depends strongly on the fuel used. Carboxylic acids are much more efficient than alcohols or fructose for the same organic load and within the range 500-5,000 mg L(-1) of acetate the production of electricity increases linearly with the amount of acetate fed but over these concentrations a change in the population composition may explain a worse performance. Copyright © 2016 Elsevier Ltd. All rights reserved.

Spent fuel transportation in the United States: commercial spent fuel shipments through December 1984

International Nuclear Information System (INIS)

1986-04-01

This report has been prepared to provide updated transportation information on light water reactor (LWR) spent fuel in the United States. Historical data are presented on the quantities of spent fuel shipped from individual reactors on an annual basis and their shipping destinations. Specifically, a tabulation is provided for each present-fuel shipment that lists utility and plant of origin, destination and number of spent-fuel assemblies shipped. For all annual shipping campaigns between 1980 and 1984, the actual numbers of spent-fuel shipments are defined. The shipments are tabulated by year, and the mode of shipment and the casks utilized in shipment are included. The data consist of the current spent-fuel inventories at each of the operating reactors as of December 31, 1984. This report presents historical data on all commercial spent-fuel transportation shipments have occurred in the United States through December 31, 1984

In-reactor performance of methods to control fuel-cladding chemical interaction

International Nuclear Information System (INIS)

Weber, E.T.; Gibby, R.L.; Wilson, C.N.; Lawrence, L.A.; Adamson, M.G.

1979-01-01

Inner surface corrosion of austenitic stainless steel cladding by oxygen and reactive fission product elements requires a 50 μm wastage allowance in current FBR reference oxide fuel pin design. Elimination or reduction of this wastage allowance could result in better reactor efficiency and economics through improvements in fuel pin performance and reliability. Reduction in cladding thickness and replacement of equivalent volume with fuel result in improved breeding capability. Of the factors affecting fuel-cladding chemical interaction (FCCI), oxygen activity within the fuel pin can be most readily controlled and/or manipulated without degrading fuel pin performance or significantly increasing fuel fabrication costs. There are two major approaches to control oxygen activity within an oxide fuel pin: (1) control of total oxygen inventory and chemical activity (Δ anti GO 2 ) by use of low oxygen-to-metal ratio (O/M) fuel; and (2) incorporation of a material within the fuel pin to provide in-situ control of oxygen activity (Δ anti GO 2 ) and fixation of excess oxygen prior to, or in preference to reaction with the cladding. The paper describes irradiation tests which were conducted in EBR-II and GETR incorporating oxygen buffer/getter materials and very low O/M fuel to control oxygen activity in sealed fuel pins

Development of Out-pile Test Technology for Fuel Assembly Performance Verification

Energy Technology Data Exchange (ETDEWEB)

Chun, Tae Hyun; In, W. K.; Oh, D. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)

2007-03-15

Out-pile tests with full scale fuel assembly are to verify the design and to evaluate the performance of the final products. HTL for the hydraulic tests and FAMeCT for mechanical/structural tests were constructed in this project. The maximum operating conditions of HTL are 30 bar, 320 .deg. C, and 500 m3/hr. This facility can perform the pressure drop test, fuel assembly uplift test, and flow induced vibration test. FAMeCT can perform the bending and vibration tests. The verification of the developed facilities were carried out by comparing the reference data of the fuel assembly which was obtained at the Westinghouse Co. The compared data showed a good coincidence within uncertainties. FRETONUS was developed for high temperature and high pressure fretting wear simulator and performance test. A performance test was conducted for 500 hours to check the integrity, endurance, data acquisition capability of the simulator. The technology of turbulent flow analysis and finite element analysis by computation was developed. From the establishments of out-pile test facilities for full scale fuel assembly, the domestic infrastructure for PWR fuel development has been greatly upgraded.

Performance of PEM Liquid-Feed Direct Methanol-Air Fuel Cells

Science.gov (United States)

Narayanan, S. R.

1995-01-01

A direct methanol-air fuel cell operating at near atmospheric pressure, low-flow rate air, and at temperatures close to 60oC would tremendously enlarge the scope of potential applications. While earlier studies have reported performance with oxygen, the present study focuses on characterizing the performance of a PEM liquid feed direct methanol-air cell consisting of components developed in house. These cells employ Pt-Ru catalyst in the anode, Pt at the cathode and Nafion 117 as the PEM. The effect of pressure, flow rate of air and temperature on cell performance has been studied. With air, the performance level is as high as 0.437 V at 300 mA/cm2 (90oC, 20 psig, and excess air flow) has been attained. Even more significant is the performance level at 60oC, 1 atm and low flow rates of air (3-5 times stoichiometric), which is 0.4 V at 150 mA/cm2. Individual electrode potentials for the methanol and air electrode have been separated and analyzed. Fuel crossover rates and the impact of fuel crossover on the performance of the air electrode have also been measured. The study identifies issues specific to the methanol-air fuel cell and provides a basis for improvement strategies.

Calculation study of the WWER-440 fuel performance for extended burnup

International Nuclear Information System (INIS)

Kujal, J.; Pazdera, F.; Barta, O.

1984-01-01

The results of preliminary calculational study of extended burnup cycling schemes impact on WWER-440 fuel performance are presented. Two high burnup schemes were proposed with three and four cycles, resp. Comparison was made with three cycle reference case. The thermal mechanical analysis was performed with PIN and RELA codes. The values of rod internal pressure, fuel centerline temperatures and fuel-cladding gap are expressed as function of power history. (author)

PWR fuel performance and future trend in Japan

International Nuclear Information System (INIS)

Kondo, Y.

1987-01-01

Since the first PWR power plant Mihama Unit 1 initiated its commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts to improve PWR technology. The results are already seen in significantly improved performance of 16 PWR plants now in operation. Mitsubishi Heavy Industries Ltd. (MHI) has been supplying them with nuclear fuel assemblies, which are over 5700. As the reliability of the current design fuel has been achieved, the direction of R and D on nuclear fuel has changed to make nuclear power more competitive to the other power generation methods. The most important R and D targets are the burnup extension, Gd contained fuel, Pu utilizatoin and the load follow capacility. (author)

Solid oxide fuel cell performance comparison fueled by methane, MeOH, EtOH and gasoline surrogate C_8H_1_8

International Nuclear Information System (INIS)

Liso, Vincenzo; Cinti, Giovanni; Nielsen, Mads P.; Desideri, Umberto

2016-01-01

Carbon deposition is a major cause of degradation in solid oxide fuel cell systems. The ability to predict carbon formation in reforming processes is thus absolutely necessary for stable operation of solid oxide fuel cell systems. In the open literature it is found that the steam input is always considered in large excess compared to what required by the reforming process with the purpose of reducing carbon formation and avoiding rapid degradation of the cell performance. This makes it difficult to consistently compare system performance with different fuels. In this work, the molar compositions at equilibrium are calculated for a minimum steam to carbon ratio for each fuel type. We carry out a thermodynamic analysis of fuel/steam system using Gibbs Free Energy minimization method. A mathematical relationship between Lagrange's multipliers and carbon activity in the gas phase was deduced. Minimum steam required for the reforming process for each fuel was related to the heat required for the reforming process and fuel cell open circuit voltage. Furthermore, in an experimental test, steam reforming product compositions were used to evaluate and compare SOFC performance with different hydrocarbons. Comparing the model to the experimental activity, it is revealed that at temperatures exceeding 800 °C the gas composition is dominated by hydrogen and carbon monoxide for any of the fuels considered leading to similar cell polarization curves performance for different fuels. The main effect on the performance is related to OCV values which are dependent on different steam content for each fuel. It was concluded that the magnitude of the heat requested for the fuel reforming process is the major difference in system performance when comparing different fuels. However, reforming kinetic effects can become predominant rather than thermodynamics, especially at lower temperatures.

The improvement of performances for PWR fuels

International Nuclear Information System (INIS)

Debes

2001-01-01

UO 2 fuels used in French nuclear power plants are authorized for values of burn-ups up to 52 GWj/t. Constant technological progress concerning pellets, cladding, and the design of the assembly has led to better performance and a broader safety margin. EDF is gathering all the elements to qualify and back its demand to increase the limit burn-up to 65 GWj/t in 2004 and to 70 GWj/t in 2008. For the same amount of energy produced, this policy of higher burn-ups will allow: - a reduction of the number of spent fuel assemblies, - a direct economic spare by using less fuel assemblies, - a reduction of personnel dosimetry because of longer irradiation campaigns, and - less quantity of residual plutonium produced. (A.C.)

Transient performance simulation of aircraft engine integrated with fuel and control systems

International Nuclear Information System (INIS)

Wang, C.; Li, Y.G.; Yang, B.Y.

2017-01-01

Highlights: • A new performance simulation method for engine hydraulic fuel systems is introduced. • Time delay of engine performance due to fuel system model is noticeable but small. • The method provides details of fuel system behavior in engine transient processes. • The method could be used to support engine and fuel system designs. - Abstract: A new method for the simulation of gas turbine fuel systems based on an inter-component volume method has been developed. It is able to simulate the performance of each of the hydraulic components of a fuel system using physics-based models, which potentially offers more accurate results compared with those using transfer functions. A transient performance simulation system has been set up for gas turbine engines based on an inter-component volume (ICV) method. A proportional-integral (PI) control strategy is used for the simulation of engine controller. An integrated engine and its control and hydraulic fuel systems has been set up to investigate their coupling effect during engine transient processes. The developed simulation system has been applied to a model aero engine. The results show that the delay of the engine transient response due to the inclusion of the fuel system model is noticeable although relatively small. The developed method is generic and can be applied to any other gas turbines and their control and fuel systems.

11. annual report of the technical advisory committee on the nuclear fuel waste management program

International Nuclear Information System (INIS)

1990-10-01

The Eleventh Annual Report of the Technical Advisory Committee (TAC) assesses the scientific and technical progress made within the Canadian Nuclear Fuel Waste Management Program (NFWMP) during the period July 1989 to June 1990. The Committee notes that the general concept of a multibarrier system involving geologic media and engineered systems is based on known technologies and current scientific knowledge, and has gained strong international scientific and engineering support as currently the most feasible and practical. TAC continues to endorse the full investigation of the concept of nuclear waste disposal deep in plutonic formations, such as those in the Canadian Shield

Tradeoffs in fuel cycle performance for most promising options - 15346

International Nuclear Information System (INIS)

Taiwo, T.; Kim, T.K.; Feng, B.; Stauff, N.; Hoffman, E.; Ganda, F.; Todosow, M.; Brown, N.; Raitses, G.; Gehin, J.; Powers, J.; Youinou, G.; Hiruta, H.; Wigeland, R.

2015-01-01

A recent Evaluation and Screening (E/S) study of nuclear fuel cycle options was conducted by grouping all potential options into 40 Evaluation Groups (EGs) based on similarities in fundamental physics characteristics and fuel cycle performance. Through a rigorous evaluation process considering benefit and challenge metrics, 4 of these EGs were identified by the E/S study as 'most promising'. All 4 involve continuous recycle of U/Pu or U/TRU with natural uranium feed in fast critical reactors. However, these most promising EGs also include fuel cycle groups with variations on feed materials, neutron spectra, and reactor criticality. Therefore, the impacts of the addition of natural thorium fuel feed to a system that originally only used natural uranium fuel feed, using an intermediate spectrum instead of a fast spectrum, and using externally-driven systems versus critical reactors were evaluated. It was found that adding thorium to the natural uranium feed mixture leads to lower burnup, higher mass flows, and degrades fuel cycle benefit metrics (waste management, resource utilization, etc.) for fuel cycles that continuously recycle U/Pu or U/TRU. Adding thorium results in fissions of 233 U instead of just 239 Pu and in turn results in a lower average number of neutrons produced per absorption (η) for the fast reactor system. For continuous recycling systems, the lower η results in lower excess reactivity and subsequently lower achievable fuel burnup. This in turn leads to higher mass flows (fabrication, reprocessing, disposal, etc.) to produce a given amount of energy and subsequent lower metrics performance. The investigated fuel cycle options with intermediate spectrum reactors also exhibited degraded performance in the benefit metrics compared to fast spectrum reactors. Similarly, this is due to lower η values as the spectrum softens. The best externally-driven systems exhibited similar performance as fast critical reactors in terms of mass flows

Relations between Air-Fuel Ratio and Dynamic Performance of Small Race Cars

OpenAIRE

位田, 晴良; Ida, Haruyoshi; 漁, 佑一郎; Sunadori, Yuichiro; 牧田, 俊太郎; Makita, Syuntaro; 宮﨑, 真央; Miyazaki, Manaka; 磯松, 弥司; Isomatsu, Yatsuka

2017-01-01

'It goes without saying that engine output power characteristics greatly affect the dynamic performance of the race car. One of the methods of changing the output power of the engine is to adjust the set amount of fuel supply. This method changes the air-fuel ratio of the air fuel mixture supplied to the engine. In this study, a slalom test run of a small race car was used to examine dynamic performance with attention to the air-fuel ratio changed by adjusting the set amount of fuel supply. T...

Performance Evaluation of a High Bandwidth Liquid Fuel Modulation Valve for Active Combustion Control

Science.gov (United States)

Saus, Joseph R.; DeLaat, John C.; Chang, Clarence T.; Vrnak, Daniel R.

2012-01-01

At the NASA Glenn Research Center, a characterization rig was designed and constructed for the purpose of evaluating high bandwidth liquid fuel modulation devices to determine their suitability for active combustion control research. Incorporated into the rig s design are features that approximate conditions similar to those that would be encountered by a candidate device if it were installed on an actual combustion research rig. The characterized dynamic performance measures obtained through testing in the rig are planned to be accurate indicators of expected performance in an actual combustion testing environment. To evaluate how well the characterization rig predicts fuel modulator dynamic performance, characterization rig data was compared with performance data for a fuel modulator candidate when the candidate was in operation during combustion testing. Specifically, the nominal and off-nominal performance data for a magnetostrictive-actuated proportional fuel modulation valve is described. Valve performance data were collected with the characterization rig configured to emulate two different combustion rig fuel feed systems. Fuel mass flows and pressures, fuel feed line lengths, and fuel injector orifice size was approximated in the characterization rig. Valve performance data were also collected with the valve modulating the fuel into the two combustor rigs. Comparison of the predicted and actual valve performance data show that when the valve is operated near its design condition the characterization rig can appropriately predict the installed performance of the valve. Improvements to the characterization rig and accompanying modeling activities are underway to more accurately predict performance, especially for the devices under development to modulate fuel into the much smaller fuel injectors anticipated in future lean-burning low-emissions aircraft engine combustors.

Performance evaluation of UO2-Zr fuel in power ramp tests

International Nuclear Information System (INIS)

Knudsen, P.; Bagger, C.

1977-01-01

In power reactors using UO 2 -Zr fuel, rapid power increases may lead to failures in fuel pins that have been irradiated at steady or decreasing heat loads. This paper presents results which extend the experience with power ramp performance of high burn-up fuel pins. A test fuel element containing both pellet and vipac UO 2 -Zr fuel pins was irradiated in the HBWR at Halden for effectively 2 1/2 years to an average burn-up of 21,000 MWD/te UO 2 at gradually decreasing power levels. The subsequent non-destructive characterization revealed formation of transverse cracks in the vipac fuel columns. After the HBWR irradiation, five of the fuel pins were power ramp tested individually in the DR 3 Reactor at Riso. The ramp rates in this test series were in the range 3-60 W/cm min. The maximum local heat loads seen in the ramp tests were 20-120% above the highest levels experienced at the same axial positions during the HBWR irradiation. Three pellets and one vipac fuel pin failed, whereas another vipac pin gave no indication of clad penetration. Profilometry after the ramp testing indicated the formation of small ridges for both types of fuel pins. For vipac fuel, the ridges were less regularly distributed along the pin length than for pellet fuel. Neutron radiography revealed the formation of additional transverse and longitudinal fuel cracks during the power ramps for both types of fuel pins. The observed failures seemed to be marginal since little or no indication as to the locations of the clad penetrations could be derived from the non-destructive post-irradiation examinations. The cases have been analyzed by means of the Danish fuel performance codes. The calculations, which are in general agreement with the observations, are discussed. The results of the investigations indicate qualitative similarities in over power performance of the two fuel types

Operating performance and reliability of CANDU PHWR fuel channels in Canada

International Nuclear Information System (INIS)

Strachan, B.; Brown, D.R.

1983-03-01

CANDU nuclear plants use many small-diameter high-pressure fuel channels. Good operating performance from the CANDU fuel channels has made a major contribution to the world-leading operating record of the CANDU nuclear power plants. As of 1982 December 31, there were 7,480 fuel channels installed in 18 CANDU reactors over 500 MW(e) in size. Eight of these reactors have been declared in-service and have accumulated 24,000 fuel channel-years of operation. The only significant operating problems with fuel channels have been the occurrence of leaking cracks in 70 fuel channels and a larger amount of axial creep on the early reactors than was originally provided for in the design. Both of these problems have been corrected on all CANDU reactors built since the Bruce GS 'A' station and the newer reactors should exhibit even better performance

Design characteristics of metallic fuel rod on its in-LMR performance

International Nuclear Information System (INIS)

Hwang, Woan; Kang Hee Young; Nam, Cheol; Kim, Jong Oh

1997-01-01

Fuel design is a key feature to assure LMR safety goals. To date, a large effort had been devoted to develop metallic fuels at ANL's experimental breeder reactor (EBR-II). The major design and performance parameters investigated include; thermal conductivity and temperature profile; smear density; axial plenum; FCMI and cladding deformation including creep, and fission gas release. In order to evaluate the sensitivity of each parameter, in-LMR performances of metallic fuels are not only reviewed by the experiment results in literatures, but also key design characteristics according to the variation of metallic fuel rod design parameters are analyzed by using the MACSIS code which simulates in-reactor behaviors of metal fuel rod. In this study, key design characteristics and the criteria which must be considered to design fuel rod in LMR, are proposed and discussed. (author). 14 refs., 4 figs

Evaluation of fuel performance with different enrichment degrees for an experimental device

International Nuclear Information System (INIS)

Giovedi, Claudia; Pino, Eddy S.; Gomes, Daniel S.; Abe, Alfredo Y.; Silva, Antonio Teixeira e

2013-01-01

Evaluation of fuel performance is conventionally carried out using specific codes developed to this aim. The obtained data are confirmed by experimental measurements performed using devices, which are located inside research reactors, projected to simulate reactor conditions under normal operation. Due to the limitations of the available reactor core length for irradiation in research reactors core, fuel rods used to obtain experimental data must present the same characteristics of the real fuel rod, but with a shorter length. Then, in order to compare the obtained results to the expected behavior of the real fuel rod, the experimental fuel rod should be designed with a free volume to fuel volume ratio very closed to the one of the full scale fuel rod. The aim of this paper is to evaluate some parameters and aspects related to the fuel rod behavior in a rod applied to the experimental irradiation device called Nuclear Fuel Irradiation Circuit (CAFE-Mod1) considering two fuel enrichment degrees: a typical commercial PWR enrichment and a value about 4 times higher. This evaluation is carried out by means of an adapted fuel performance code. Some of the parameter evaluated were fuel temperature and fission gas release as function of the fuel enrichment level. The results obtained in this paper were very similar to the ones previously obtained without consider similar free volume between the experimental and the full length fuel rod, regardless of low increases observed for the internal rod pressure and the amount of fission gas released. (author)

AREVA's fuel assemblies addressing high performance requirements of the worldwide PWR fleet

International Nuclear Information System (INIS)

Anniel, Marc; Bordy, Michel-Aristide

2009-01-01

Taking advantage of its presence in the fuel activities since the start of commercial nuclear worldwide operation, AREVA is continuing to support the customers with the priority on reliability, to: >participate in plant operational performance for the in core fuel reliability, the Zero Tolerance for Failure ZTF as a continuous improvement target and the minimisation of manufacturing/quality troubles, >guarantee the supply chain a proven product stability and continuous availability, >support performance improvements with proven design and technology for fuel management updating and cycle cost optimization, >support licensing assessments for fuel assembly and reloads, data/methodologies/services, >meet regulatory challenges regarding new phenomena, addressing emergent performance issues and emerging industry challenges for changing operating regimes. This capacity is based on supplies by AREVA accumulating very large experience both in manufacturing and in plant operation, which is demonstrated by: >manufacturing location in 4 countries including 9 fuel factories in USA, Germany, Belgium and France. Up to now about 120,000 fuel assemblies and 8,000 RCCA have been released to PWR nuclear countries, from AREVA European factories, >irradiation performed or in progress in about half of PWR world wide nuclear plants. Our optimum performances cover rod burn ups of to 82GWD/tU and fuel assemblies successfully operated under various world wide fuel management types. AREVA's experience, which is the largest in the world, has the extensive support of the well known fuel components such as the M5'TM'cladding, the MONOBLOC'TM'guide tube, the HTP'TM' and HMP'TM' structure components and the comprehensive services brought in engineering, irradiation and post irradiation fields. All of AREVA's fuel knowledge is devoted to extend the definition of fuel reliability to cover the whole scope of fuel vendor support. Our Top Reliability and Quality provide customers with continuous

Performance of the Westinghouse WWER-1000 fuel design

International Nuclear Information System (INIS)

Höglund, J.; Jansson, A.; Latorre, R.; Davis, D.

2015-01-01

In 2005, six (6) Westinghouse WWER-1000 Lead Test Assemblies (LTAs) were loaded in South Ukraine Unit 3 (SU3). The LTAs completed the planned four cycles of operation and reached an average assembly burnup in excess of 43 MWd/ kgU. Post Irradiation Examination (PIE) inspections were performed after completion of each cycle and it was concluded that the 6 Westinghouse LTAs performed as expected during their operational regimes. In 2010, a full region of 42 assemblies of an enhanced WWER-1000 fuel design for Ukrainian reactors, designated WFA, was loaded in SU3. The WFA includes features that further mitigate assembly bow while at the same time improving the fuel cycle economy. In 2015, 26 WFAs completed their planned four cycles of operation reaching an average assembly burnup in excess of 42 MWd/ kgU. Currently 36 WFAs continue operating their fourth cycle in SU3. In addition, South Ukraine Unit 2 (SU2) has been loaded with WFAs and 27 assemblies have completed two cycles of operation reaching an average assembly burnup above 24 MWd/kgU. PIE for the WFAs has been completed after each cycle of operation. All assemblies have been examined for visible damage or non-standard position of fuel assembly components during unloading and reloading. All WFAs have also been subject to the standard leak testing process, with all fuel rods found to be hermetically sealed and non-leaking. Each outage, six WFAs have been subject to a more extensive inspection program. In 2012, 2013, and 2015, the Westinghouse Fuel Inspection and Repair Equipment (FIRE) workstation were used for the SU3 inspections. Excellent irradiation fuel performance has been observed and measured on all WFAs. The fuel assembly growth, rod cluster control assembly (RCCA) drag forces, oxide thickness, total fuel rod-to-nozzle gap channel closure, and fuel assembly bow data were within the bounds of the Westinghouse experience database. Results and concluding remarks from the PIEs are provided in this paper. In

Δ14C level of annual plants and fossil fuel derived CO2 distribution across different regions of China

International Nuclear Information System (INIS)

Xi, X.T.; Ding, X.F.; Fu, D.P.; Zhou, L.P.; Liu, K.X.

2013-01-01

The 14 C level in annual plants is a sensitive tracer for monitoring fossil fuel derived CO 2 in the atmosphere. Corn leave samples were selected from different regions of China, including high mountains in the Tibetan Plateau, grassland in Inner Mongolia, and inland and coastal cities during the summer of 2010. The 14 C/ 12 C ratio of the samples was measured with the NEC compact AMS system at the Institute of Heavy Ion Physics, Peking University. The fossil fuel derived CO 2 was estimated by comparing the measured Δ 14 C values of corn leave samples to background atmospheric Δ 14 C level. The influences of topography, meteorological conditions and carbon cycling processes on the fossil fuel derived CO 2 concentration are considered when interpreting the data. Our results show a clear association of the low Δ 14 C values with regions where human activities are intensive.

Metal fuel safety performance

International Nuclear Information System (INIS)

Miles, K.J. Jr.; Tentner, A.M.

1988-01-01

The current development of breeder reactor systems has lead to the renewed interest in metal fuels as the driver material. Modeling efforts were begun to provide a mechanistic description of the metal fuel during anticipated and hypothetical transients within the context of the SAS4A accident analysis code system. Through validation exercises using experimental results of metal fuel TREAT tests, confidence is being developed on the nature and accuracy of the modeling and implementation. Prefailure characterization, transient pin response, margins to failure, axial in-pin fuel relocation prior to cladding breach, and molten fuel relocation after cladding breach are considered. Transient time scales ranging from milliseconds to many hours can be studied with all the reactivity feedbacks evaluated

Hydrocarbon Fuel Thermal Performance Modeling based on Systematic Measurement and Comprehensive Chromatographic Analysis

Science.gov (United States)

2016-07-31

distribution unlimited Hydrocarbon Fuel Thermal Performance Modeling based on Systematic Measurement and Comprehensive Chromatographic Analysis Matthew...vital importance for hydrocarbon -fueled propulsion systems: fuel thermal performance as indicated by physical and chemical effects of cooling passage... analysis . The selection and acquisition of a set of chemically diverse fuels is pivotal for a successful outcome since test method validation and

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2013

Energy Technology Data Exchange (ETDEWEB)

Eudy, Leslie [National Renewable Energy Lab. (NREL), Golden, CO (United States); Gikakis, Christina [Federal Transit Administration, Washington, DC (United States)

2013-12-01

This report is the seventh in an annual series of reports that summarize the progress of fuel cell electric bus (FCEB) development in the United States and discuss the achievements and challenges of introducing fuel cell propulsion in transit. The report also provides a snapshot of current FCEB performance results from August 2012 through July 2013 for five FCEB demonstrations at four transit agencies.

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2012

Energy Technology Data Exchange (ETDEWEB)

Eudy, L.; Chander, K.; Gikakis, C.

2012-11-01

This report is the sixth in an annual series of reports that summarize the progress of fuel cell electric bus (FCEB) development in the United States and discuss the achievements and challenges of introducing fuel cell propulsion in transit. The report also provides a snapshot of current FCEB performance results over the last year. There are 25 active FCEBs in demonstrations this year at eight locations.

Highlights from the IAEA coordinated research programme on fuel performance and fission product data

International Nuclear Information System (INIS)

Nabielek, H.; Schenk, W.; Verfondern, K.

1996-01-01

Seven countries are cooperating with the objectives (i) to document the status of the experimental data base and of the predictive methods for Gas-Cooled Reactor fuel performance and fission product behaviour; (ii) to verify and validate methods in fuel performance and fission product retention prediction. These countries are China, France, Germany, Japan, Russia, USA and the UK. Duration of the programme is 1993-96. The technology areas addressed in this IAEA Coordinated Research Programme are: Fuel design and manufacture, Normal operation fuel performance and fission product behaviour, Accident condition fuel performance and fission product behaviour, -core heatup, -fast transients, -oxidising conditions (water and air ingress), Plateout, re-entrainment of plateout, fission product behaviour in the reactor building, and Performance of advanced fuels. Work performed so far has generated a 300-page draft document with important information for normal operations (Germany, Japan, China, Russia) and accident conditions (USA, Japan, Germany, Russia) and, additionally, a special chapter on advanced fuels (Japan). (author)

11. International conference on WWER fuel performance, modelling and experimental support. Proceedings

International Nuclear Information System (INIS)

Manolova, M.; Boneva, S.; Mitev, M.

2015-01-01

This publication is a compilation of the papers presented in 11th International Conference on WWER Fuel Performance, Modeling and Experimental Support, organized by the Institute for Nuclear Research and Nuclear Energy (INRNE) of the Bulgarian Academy of Sciences in co-operation with the International Atomic Energy Agency (IAEA), Vienna, Austria, supported by the Kozloduy Nuclear Power Plant (KNPP), the Bulgarian Nuclear Regulatory Agency, and TVEL Fuel Company, Russia. The Conference took place in hotel Bolero, Golden Sands Resort, Bulgaria, from 26 September 2015 to 3 October 2015. It was attended by 117 participants, among them more than 100 experts and specialists from 22 countries, including representatives of 3 international organizations, 16 Russian organizations and other 36 foreign institutes, nuclear fuel plants, nuclear power plants and organizations responsible for WWER and PWR fuel design, manufacturing and research, and 3 Bulgarian organizations, working for the Bulgarian nuclear industry. 70 papers have been presented in the Conference in 6 oral and 1 poster session, covering: (1) general overview lectures; (2) fuel performance and operational experience; (3) fuel modeling and experimental support; (4) fuel safety and QA; (5) spent fuel performance and management; (6) specific issues of WWER-1000 fuel reliability. The proceedings provide Summary, Conclusions and Recommendations of the Conference, together with the full text of the presentations. IAEA Technical Meeting (TM) “Achieving zero fuel failure rates: challenges and perspectives”, 1 – 2 October 2015 was organized in conjunction with the 11th International Conference on WWER Fuel Performance, Modelling and Experimental Support. The reports presented on TM sessions are included in the Conference Proceedings too

Study on the micro direct ethanol fuel cell (Micro-DEFC) performance

Science.gov (United States)

Saisirirat, Penyarat; Joommanee, Bordindech

2018-01-01

The direct ethanol fuel cell (DEFC) is selected for this research. DEFC uses ethanol in the fuel cell instead of the more toxic methanol. Ethanol is more attractive than methanol by many reasons. Ethanol is a hydrogen-rich liquid and it has a higher specific energy (8.0 kWh/kg) compared to that of methanol (6.1 kWh/kg). Ethanol can be obtained in great quantity from biomass through a fermentation process from renewable resources such as sugar cane, wheat, corn, and even straw. The use of ethanol would also overcome both the storage and infrastructure challenge of hydrogen for fuel cell applications. The experimental apparatus on the micro direct ethanol fuel cell for measuring the cell performance has been set for this research. The objective is to study the micro direct ethanol fuel cell performance for applying with the portable electronic devices. The cell performance is specified in the terms of cell voltage, cell current and power of the cell at room operating temperature and 1 atm for the pressure and also includes the ethanol fuel consumption. The effect of operating temperature change on the electrical production performance is also studied. The steady-state time for collecting each data value is about 5-10 minutes. The results show that with the increase of concentrations of ethanol by volume, the reactant concentration at the reaction sites increases so the electrochemical rate also increases but when it reaches the saturated point the performance gradually drops.

Hybrid Direct Carbon Fuel Cell Performance with Anode Current Collector Material

DEFF Research Database (Denmark)

Deleebeeck, Lisa; Kammer Hansen, Kent

2015-01-01

collectors were studied: Au, Ni, Ag, and Pt. It was shown that the performance of the direct carbon fuel cell (DCFC) is dependent on the current collector materials, Ni and Pt giving the best performance, due to their catalytic activity. Gold is suggested to be the best material as an inert current collector......The influence of the current collector on the performance of a hybrid direct carbon fuel cell (HDCFC), consisting of solid oxide fuel cell (SOFC) with a molten carbonate-carbon slurry in contact with the anode, has been investigated using current-voltage curves. Four different anode current...

Proceedings of American Nuclear Society topical meeting on light water reactor fuel performance

International Nuclear Information System (INIS)

1979-01-01

Separate abstracts are included for 32 of the 39 papers presented concerning fuel performance; fuel performance under transient and off-normal operations; fuel-cladding interactions; and fission product release. Seven papers have previously been input into the data base

1981 Annual Status Report. Plutonium fuels and actinide programme

International Nuclear Information System (INIS)

1981-01-01

In this 1981 report the work carried out by the European Institute for Transuranium elements is reviewed. Main topics are: operation limits of plutonium fuels: swelling of advanced fuels, oxide fuel transients, equation of state of nuclear materials; actinide cycle safety: formation of actinides (FACT), safe handling of plutonium fuel (SHAPE), aspects of the head-end processing of carbide fuel (RECARB); actinide research: crystal chemistry, solid state studies, applied actinide research

U.S. Department of Energy Hydrogen and Fuel Cells Program 2016 Annual Merit Review and Peer Evaluation Report: June 6-10, 2016, Washington, DC

Energy Technology Data Exchange (ETDEWEB)

Popovich, Neil

2016-10-01

The fiscal year 2016 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting (AMR), in conjunction with DOE's Vehicle Technologies Office AMR, was held from June 6-10, 2015, in Washington, D.C.. This report is a summary of comments by AMR peer reviewers about the hydrogen and fuel cell projects funded by DOE's Office of Energy Efficiency and Renewable Energy.

U.S. Department of Energy Hydrogen and Fuel Cells Program 2015 Annual Merit Review and Peer Evaluation Report: June 8-12, 2015, Arlington, Virginia

Energy Technology Data Exchange (ETDEWEB)

Popovich, Neil

2015-10-01

The fiscal year 2015 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting (AMR), in conjunction with DOE's Vehicle Technologies Office AMR, was held from June 8-12, 2015, in Arlington, Virginia. This report is a summary of comments by AMR peer reviewers about the hydrogen and fuel cell projects funded by DOE's Office of Energy Efficiency and Renewable Energy.

Shielding Performance Measurements of Spent Fuel Transportation Container

Directory of Open Access Journals (Sweden)

SUN Hong-chao

2015-11-01

Full Text Available The safety supervision of radioactive material transportation package has been further stressed and implemented. The shielding performance measurements of spent fuel transport container is the important content of supervision. However, some of the problems and difficulties reflected in practice need to be solved, such as the neutron dose rate on the surface of package is too difficult to measure exactly, the monitoring results are not always reliable, etc. The monitoring results using different spectrometers were compared and the simulation results of MCNP runs were considered. An improvement was provided to the shielding performance measurements technique and management of spent fuel transport.

Fuel performance in water storage

International Nuclear Information System (INIS)

Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

1993-11-01

Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory (INEL) for the Department of Energy (DOE). A variety of different types of fuels have been stored there since the 1950's prior to reprocessing for uranium recovery. In April of 1992, the DOE decided to end fuel reprocessing, changing the mission at ICPP. Fuel integrity in storage is now viewed as long term until final disposition is defined and implemented. Thus, the condition of fuel and storage equipment is being closely monitored and evaluated to ensure continued safe storage. There are four main areas of fuel storage at ICPP: an original underwater storage facility (CPP-603), a modern underwater storage facility (CPP-666), and two dry fuel storage facilities. The fuels in storage are from the US Navy, DOE (and its predecessors the Energy Research and Development Administration and the Atomic Energy Commission), and other research programs. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels. In the underwater storage basins, fuels are clad with stainless steel, zirconium, and aluminum. Also included in the basin inventory is canned scrap material. The dry fuel storage contains primarily graphite and aluminum type fuels. A total of 55 different fuel types are currently stored at the Idaho Chemical Processing Plant. The corrosion resistance of the barrier material is of primary concern in evaluating the integrity of the fuel in long term water storage. The barrier material is either the fuel cladding (if not canned) or the can material

Performance and emissions of a supercharged dual-fuel engine fueled by hydrogen-rich coke oven gas

Energy Technology Data Exchange (ETDEWEB)

Roy, M.M.; Tomita, E.; Kawahara, N.; Harada, Y.; Sakane, A. [Okayama University, Okayama (Japan). Dept. of Mechanical Engineering

2009-12-15

This study investigated the engine performance and emissions of a supercharged dual-fuel engine fueled by hydrogen-rich coke oven gas and ignited by a pilot amount of diesel fuel. The engine was tested for use as a cogeneration engine, so power output while maintaining a reasonable thermal efficiency was important. Experiments were carried out at a constant pilot injection pressure and pilot quantity for different fuel-air equivalence ratios and at various injection timings without and with exhaust gas recirculation (EGR). The experimental strategy was to optimize the injection timing to maximize engine power at different fuel-air equivalence ratios without knocking and within the limit of the maximum cylinder pressure. The engine was tested first without EGR condition up to the maximum possible fuel-air equivalence ratio of 0.65. A maximum indicated mean effective pressure (IMEP) of 1425 kPa and a thermal efficiency of 39% were obtained. However, the nitrogen oxides (NOx) emissions were high. A simulated EGR up to 50% was then performed to obtain lower NOx emissions. The maximum reduction of NOx was 60% or more maintaining the similar levels of IMEP and thermal efficiency. Two-stage combustion was obtained; this is an indicator of maximum power output conditions and a precursor of knocking combustion.

1982 Annual Status Report Plutonium Fuels and Actinide Programme

International Nuclear Information System (INIS)

Lindner, R.

1983-01-01

The programme of the Transuranium Institute has long included work on advanced fuels for fast breeder reactors. Study of the swelling of carbide and nitride fuels is now nearing completion, the retention of fission gases in bubbles of different sizes in the fuel having been quantified as function of burn-up and temperature. An important step forward has been achieved in the studies of the Equation of State of Nuclear Fuels up to 5000 K. Formation of some of the less abundant isotopes in PWR fuel has been determined experimentally. Aerosol formation during the fabrication of plutonium containing fuels, part of the activity Safe Handling of Plutonium Fuel has been studied. Head-End Processing of carbide fuels has continued experiments with high burn up mixed carbides. In the field of actinide research the preparation and characterisation of pure specimens is carried out. Effect of actinides on the properties of waste glasses is investigated

A fuel performance code TRUST VIc and its validation

Energy Technology Data Exchange (ETDEWEB)

Ishida, M; Kogai, T [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan)

1997-08-01

This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs.

A fuel performance code TRUST VIc and its validation

International Nuclear Information System (INIS)

Ishida, M.; Kogai, T.

1997-01-01

This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs

Performance of U-Pu-Zr fuel cast into zirconium molds

International Nuclear Information System (INIS)

Crawford, D.C.; Lahm, C.E.; Tsai, H.

1992-01-01

Current fabrication techniques for the integral fast reactor (IFR) fuel utilize injection casting into quartz molds after reprocessing in the IFR fuel cycle facility. The quartz molds are destroyed during the fuel demolding process, and the quartz residue must therefore be treated as contaminated waste. Alternatively, if the fuel can be cast into molds that remain as part of the fuel slugs (i.e., if the fuel can be left inside the molds for irradiation), then the quartz mold contribution to the waste stream can be eliminated. This possibility is being addresssed in an ongoing effort to evaluate the irradiation performance of fuel cast into zirconium sheaths rather than quartz molds. Zirconium was chosen as the sheath material because it is the component of the U-Pu-Zr fuel alloy that raises the alloy solidus temperatures and provides resistance to fuel-cladding chemical interaction (FCCI)

Fuel cycle related parametric study considering long lived actinide production, decay heat and fuel cycle performances

International Nuclear Information System (INIS)

Raepsaet, X.; Damian, F.; Lenain, R.; Lecomte, M.

2001-01-01

One of the very attractive HTGR reactor characteristics is its highly versatile and flexible core that can fulfil a wide range of diverse fuel cycles. Based on a GTMHR-600 MWth reactor, analyses of several fuel cycles were carried out without taking into account common fuel particle performance limits (burnup, fast fluence, temperature). These values are, however, indicated in each case. Fuel derived from uranium, thorium and a wide variety of plutonium grades has been considered. Long-lived actinide production and total residual decay heat were evaluated for the various types of fuel. The results presented in this papers provide a comparison of the potential and limits of each fuel cycle and allow to define specific cycles offering lowest actinide production and residual heat associated with a long life cycle. (author)

An innovative fuel design concept for improved light water reactor performance and safety. Final technical report

International Nuclear Information System (INIS)

Tulenko, J.S.; Connell, R.G.

1995-07-01

Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. The purpose of this research was to explore a technique for extending fuel performance by thermally bonding LWR fuel with a non-alkaline liquid metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide (UO 2 ) fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Due to the thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high conductivity liquid metal thermally bonds the fuel to the cladding, and eliminates the large temperature change across the gap, while preserving the expansion and pellet loading capabilities. The resultant lower fuel temperature directly impacts fuel performance limit margins and also core transient performance. The application of liquid bonding techniques to LWR fuel was explored for the purposes of increasing LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) has been developed under the program to analyze the in-reactor performance of the liquid metal bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of Liquid Metal Bonded LWR fuel

Performance analysis of hybrid solid oxide fuel cell and gas turbine cycle: Application of alternative fuels

International Nuclear Information System (INIS)

Zabihian, Farshid; Fung, Alan S.

2013-01-01

Highlights: • Variation of the stream properties in the syngas-fueled hybrid SOFC–GT cycle. • Detailed analysis of the operation of the methane-fueled SOFC–GT cycle. • Investigate effects of inlet fuel type and composition on performance of cycle. • Comparison of system operation when operated with and without anode recirculation. - Abstract: In this paper, the hybrid solid oxide fuel cell (SOFC) and gas turbine (GT) model was applied to investigate the effects of the inlet fuel type and composition on the performance of the cycle. This type of analysis is vital for the real world utilization of manufactured fuels in the hybrid SOFC–GT system due to the fact that these fuel compositions depends on the type of material that is processed, the fuel production process, and process control parameters. In the first part of this paper, it is shown that the results of a limited number of studies on the utilization of non-conventional fuels have been published in the open literature. However, further studies are required in this area to investigate all aspects of the issue for different configurations and assumptions. Then, the results of the simulation of the syngas-fueled hybrid SOFC–GT cycle are employed to explain the variation of the stream properties throughout the cycle. This analysis can be very helpful in understanding cycle internal working and can provide some interesting insights to the system operation. Then, the detailed information of the operation of the methane-fueled SOFC–GT cycle is presented. For both syngas- and methane-fueled cycles, the operating conditions of the equipment are presented and compared. Moreover, the comparison of the characteristics of the system when it is operated with two different schemes to provide the required steam for the cycle, with anode recirculation and with an external source of water, provides some interesting insights to the system operation. For instance, it was shown that although the physical

Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

Energy Technology Data Exchange (ETDEWEB)

Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

2012-02-01

The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code 'FEMAXI-FBR'

International Nuclear Information System (INIS)

Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

2012-01-01

FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code 'ASTERIA-FBR' in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code 'FEMAXI-6', FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P 0 /s (P 0 : steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions. (author)

Development of high performance hybrid rocket fuels

Science.gov (United States)

Zaseck, Christopher R.

. In order to examine paraffin/additive combustion in a motor environment, I conducted experiments on well characterized aluminum based additives. In particular, I investigate the influence of aluminum, unpassivated aluminum, milled aluminum/polytetrafluoroethylene (PTFE), and aluminum hydride on the performance of paraffin fuels for hybrid rocket propulsion. I use an optically accessible combustor to examine the performance of the fuel mixtures in terms of characteristic velocity efficiency and regression rate. Each combustor test consumes a 12.7 cm long, 1.9 cm diameter fuel strand under 160 kg/m 2s of oxygen at up to 1.4 MPa. The experimental results indicate that the addition of 5 wt.% 30 mum or 80 nm aluminum to paraffin increases the regression rate by approximately 15% compared to neat paraffin grains. At higher aluminum concentrations and nano-scale particles sizes, the increased melt layer viscosity causes slower regression. Alane and Al/PTFE at 12.5 wt.% increase the regression of paraffin by 21% and 32% respectively. Finally, an aging study indicates that paraffin can protect air and moisture sensitive particles from oxidation. The opposed burner and aluminum/paraffin hybrid rocket experiments show that additives can alter bulk fuel properties, such as viscosity, that regulate entrainment. The general effect of melt layer properties on the entrainment and regression rate of paraffin is not well understood. Improved understanding of how solid additives affect the properties and regression of paraffin is essential to maximize performance. In this document I investigate the effect of melt layer properties on paraffin regression using inert additives. Tests are performed in the optical cylindrical combustor at ˜1 MPa under a gaseous oxygen mass flux of ˜160 kg/m2s. The experiments indicate that the regression rate is proportional to mu0.08rho 0.38kappa0.82. In addition, I explore how to predict fuel viscosity, thermal conductivity, and density prior to testing

Effect of fuel fabrication parameters on performance- designer's point of view

International Nuclear Information System (INIS)

Prasad, P.N.; Ravi, M.; Soni, R.; Bajaj, S.S.; Bhardwaj, S.A.

2004-01-01

The fuel bundle performance in reactor depends upon the material properties, dimensions of the different components and their inter-compatibility. This paper brings out the fuel parameters required to be optimised to achieve better fuel reliability, operational flexibility, safety and economics from the designer point of view

Probability of spent fuel transportation accidents

International Nuclear Information System (INIS)

McClure, J.D.

1981-07-01

The transported volume of spent fuel, incident/accident experience and accident environment probabilities were reviewed in order to provide an estimate of spent fuel accident probabilities. In particular, the accident review assessed the accident experience for large casks of the type that could transport spent (irradiated) nuclear fuel. This review determined that since 1971, the beginning of official US Department of Transportation record keeping for accidents/incidents, there has been one spent fuel transportation accident. This information, coupled with estimated annual shipping volumes for spent fuel, indicated an estimated annual probability of a spent fuel transport accident of 5 x 10 -7 spent fuel accidents per mile. This is consistent with ordinary truck accident rates. A comparison of accident environments and regulatory test environments suggests that the probability of truck accidents exceeding regulatory test for impact is approximately 10 -9 /mile

Effect analysis of air introduced by pressurization on fuel rod performances

International Nuclear Information System (INIS)

Ren Qisen; Liu Tong; Sheng Guofu

2012-01-01

In the process of pressurization and seal welding, it is common practice to vacuumize before gas filling for the sake of preventing introducing air and other impurities, which would affect the gas composition inside of the fuel rod. However, vacuumization during pressurization is likely not being required sometimes in order to simplify the fabrication procedure. In the present work, based on the AFA3G fuel rod design with 2 MPa of filling gas, analyses on fuel rod performances were carried out under the condition of pressurization with and without vacuumization, respectively. Furthermore, the effect on hydrogen content in fuel rod was preliminarily discussed. Results indicate that the impacts of air composition introduced by pressurization on fuel rod thermal-mechanical performances, such as internal pressure and fuel center temperature, were extremely slight. The gap conductance varies to some extent as a result of the change of gas composition due to air introduced in fuel rod. The impact of humidity on water content in fuel rod is negligible at a low temperature of around 25℃. However, at higher temperature, it is essential to pay attention on the control of fabrication process, and prevent much moisture entering into the fuel rod and increasing the probability of hydriding failure. (authors)

Evaluation and demonstration of methods for improved fuel utilization. Second semi-annual progress report, April 1, 1980-September 30, 1980

International Nuclear Information System (INIS)

1981-01-01

Demonstrations are being performed in the Fort Calhoun reactor. The current program consists of two parts, one to demonstrate low leakage fuel management (SAVFUEL - Shimmed And Very Flexible Uranium Element Loading) and the other to demonstrate high burnup. The first part will demonstrate that the power duty cycle which is characteristic of SAVFUEL does not have a deleterious effect on fuel performance, while the second part will demonstrate that the peak rod average burnup of the current 14 x 14 fuel design can be increased to 45 GWD/T. A visual examination conducted at poolside was completed on four fuel assemblies which are scheduled to demonstrate the SAVFUEL power cycle and seventeen fuel assemblies which are scheduled to provide high burnup fuel performance data. Results of visual examinations, shoulder gap closure, fuel assembly growth, and fuel rod channel width measurements are reported which show excellent fuel performance for the high burnup; demonstration assemblies after four exposure cycles. These results support an additional exposure cycle for the high burnup demonstration assemblies which currently have an assembly average burnup up to 37 GWD/T

Nuclear Safety Project. Annual report 1986

International Nuclear Information System (INIS)

1987-09-01

The annual report 1986 is a detailed description of work within the Nuclear Safety Project performed in 1986 in the nuclear safety field by KfK institutes and departments and by external institutes on behalf of KfK. It includes individual research activities on dynamic loads and strains of reactor components under accident conditions, fuel behaviour under accident conditions, investigation and control of LWR core-meltdown accidents, improvement of fission product retention and reduction of radiation exposure, and on behaviour, impact and removal of released pollutants. (DG)

Performance Evaluation and Suggestion of Upgraded Fuel Handling Equipment for Operating OPR1000

International Nuclear Information System (INIS)

Chang, Sang Gyoon; Hwang, Jeung Ki; Choi, Taek Sang; Na, Eun Seok; Lee, Myung Lyul; Baek, Seung Jin; Kim, Man Su; Kunik, Jack

2011-01-01

The purpose of this study is to evaluate the performance of upgraded FHE (Fuel Handling Equipment) for operating OPR 1000 (Optimized Power Reactor) by using data measured during the fuel reloading, and make some suggestions on enhancing the performance of the FHE. The fuel handling equipment, which serves critical processes in the refueling outage, has been upgraded to increase and improve the operational availability of the plant. The evaluation and suggestion of this study can be a beneficial tool related to the performance of the fuel handling equipment

Annual Status Report (FY2016) Performance Assessment for the Environmental Restoration Disposal Facility

Energy Technology Data Exchange (ETDEWEB)

Casbon, M. A. [CH2M HILL Plateau Remediation Company, Richland, WA (United States); Nichols, W. E. [CH2M HILL Plateau Remediation Company, Richland, WA (United States)

2017-03-15

DOE O 435.1, Radioactive Waste Management, and DOE M 435.1-1, Radioactive Waste Management Manual, require that a determination of continued adequacy of the performance assessment (PA), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and it must consider the results of data collection and analysis from research, field studies, and monitoring. Annual summaries of low-level waste (LLW) disposal operations must be prepared with respect to the conclusions and recommendations of the PA and CA, and a determination of the need to revise the PA or CA must be made. The annual summary requirement provides a structured approach for demonstrating the continued adequacy of the PA and CA in demonstrating a reasonable expectation that the performance objectives will be met. This annual summary addresses only the status of the Environmental Restoration Disposal Facility (ERDF) PA (CP-60089, Performance Assessment for the Environmental Restoration Disposal Facility, Hanford Site, Washington, formerly WCH-520 Rev. 1)1. The CA for ERDF is supported by DOE/RL-2016-62, Annual Status Report (FY 2016): Composite Analysis of Low Level Waste Disposal in the Central Plateau at the Hanford Site. The ERDF PA portion of the CA document is found in Section 3.1.4, and the ERDF operations portion is found in Section 3.3.3.2 of that document.

Effects of coal-derived trace species on performance of molten carbonate fuel cells

Energy Technology Data Exchange (ETDEWEB)

1992-05-01

The Carbonate Fuel Cell is a very promising option for highly efficient generation of electricity from many fuels. If coal-gas is to be used, the interactions of coal-derived impurities on various fuel cell components need to be understood. Thus the effects on Carbonate Fuel Cell performance due to ten different coal-derived contaminants viz., NH{sub 3}, H{sub 2}S, HC{ell}, H{sub 2}Se, AsH{sub 3}, Zn, Pb, Cd, Sn, and Hg, have been studied at Energy Research Corporation. Both experimental and theoretical evaluations were performed, which have led to mechanistic insights and initial estimation of qualitative tolerance levels for each species individually and in combination with other species. The focus of this study was to investigate possible coal-gas contaminant effects on the anode side of the Carbonate Fuel Cell, using both out-of-cell thermogravimetric analysis by isothermal TGA, and fuel cell testing in bench-scale cells. Separate experiments detailing performance decay in these cells with high levels of ammonia contamination (1 vol %) and with trace levels of Cd, Hg, and Sn, have indicated that, on the whole, these elements do not affect carbonate fuel cell performance. However, some performance decay may result when a number of the other six species are present, singly or simultaneously, as contaminants in fuel gas. In all cases, tolerance levels have been estimated for each of the 10 species and preliminary models have been developed for six of them. At this stage the models are limited to isothermal, benchscale (300 cm{sup 2} size) single cells. The information obtained is expected to assist in the development of coal-gas cleanup systems, while the contaminant performance effects data will provide useful basic information for modeling fuel cell endurance in conjunction with integrated gasifier/fuel-cell systems (IGFC).

Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

International Nuclear Information System (INIS)

Dobrevski, I.; Zaharieva, N.

2008-01-01

Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

Alkali slagging problems with biomass fuels

Energy Technology Data Exchange (ETDEWEB)

Miles, T.R.; Miles, T.R. Jr.; Baxter, L.L.; Jenkins, B.M.; Oden, L.L.

1993-12-31

Biomass fueled power boilers are unable to burn more than minor percentages of annually generated agricultural fuels. Determining the mechanisms of deposit formation, and developing means of increasing the proportion of these annual biofuels to be fired are the aims of the ongoing Alkali Deposit Investigation sponsored by DOE/NREL with matching funds from industry sponsors, combining Science, Engineering and Industry.

Transient fuel and target performance testing for the HWR-NPR

International Nuclear Information System (INIS)

Jicha, J.J. Jr.

1990-01-01

This paper describes a five year program of fuel target transient performance testing and model development required for the safety assessment of the HWR new production reactor. Technical issues are described, focusing on fuel and target behavior during extremely low probability transients which can lead to fuel melting. Early work on these issues is reviewed. The program to meet remaining needs is described. Three major transient-testing activities are included: in-cell experiments on small samples of irradiated fuel and target, small-scale phenomenological experiments in the ACRR reactor, and limited-integral experiments in the TREAT reactor. A coordinated development of detailed fuel and target behavior models is also described

Dual-fuel natural gas/diesel engines: Technology, performance, and emissions

Science.gov (United States)

Turner, S. H.; Weaver, C. S.

1994-11-01

An investigation of current dual-fuel natural gas/diesel engine design, performance, and emissions was conducted. The most pressing technological problems associated with dual-fuel engine use were identified along with potential solutions. It was concluded that dual-fuel engines can achieve low NO(sub x) and particulate emissions while retaining fuel-efficiency and BMEP levels comparable to those of diesel engines. The investigation also examined the potential economic impact of dual-fuel engines in diesel-electric locomotives, marine vessels, farm equipment, construction, mining, and industrial equipment, and stand-alone electricity generation systems. Recommendations for further additional funding to support research, development, and demonstration in these applications were then presented.

Conference summaries of the Canadian Nuclear Association 30. annual conference, and the Canadian Nuclear Society 11. annual conference

International Nuclear Information System (INIS)

1990-01-01

This volume contains conference summaries for the 30. annual conference of the Canadian Nuclear Association, and the 11. annual conference of the Canadian Nuclear Society. Topics of discussion include: energy needs and challenges facing the Canadian nuclear industry; the environment and nuclear power; the problems of maintaining and developing industrial capacity; the challenges of the 1990's; programmes and issues for the 1990's; thermalhydraulics; reactor physics and fuel management; nuclear safety; small reactors; fuel behaviour; energy production and the environment; computer applications; nuclear systems; fusion; materials handling; and, reactor components

Fuel performance and operation experience of WWER-440 fuel in improved fuel cycle

International Nuclear Information System (INIS)

Gagarinski, A.; Proselkov, V.; Semchenkov, Yu.

2007-01-01

The paper summarizes WWER-440 second-generation fuel operation experience in improved fuel cycles using the example of Kola NPP units 3 and 4. Basic parameters of fuel assemblies, fuel rods and uranium-gadolinium fuel rods, as well as the principal neutronic parameters and burn-up achieved in fuel assemblies are presented. The paper also contains some data concerning the activity of coolant during operation (Authors)

Performance and specific emissions contours throughout the operating range of hydrogen-fueled compression ignition engine with diesel and RME pilot fuels

Directory of Open Access Journals (Sweden)

Shahid Imran

2015-09-01

Full Text Available This paper presents the performance and emissions contours of a hydrogen dual fueled compression ignition (CI engine with two pilot fuels (diesel and rapeseed methyl ester, and compares the performance and emissions iso-contours of diesel and rapeseed methyl ester (RME single fueling with diesel and RME piloted hydrogen dual fueling throughout the engines operating speed and power range. The collected data have been used to produce iso-contours of thermal efficiency, volumetric efficiency, specific oxides of nitrogen (NOX, specific hydrocarbons (HC and specific carbon dioxide (CO2 on a power-speed plane. The performance and emission maps are experimentally investigated, compared, and critically discussed. Apart from medium loads at lower and medium speeds with diesel piloted hydrogen combustion, dual fueling produced lower thermal efficiency everywhere across the map. For diesel and RME single fueling the maximum specific NOX emissions are centered at the mid speed, mid power region. Hydrogen dual fueling produced higher specific NOX with both pilot fuels as compared to their respective single fueling operations. The range, location and trends of specific NOX varied significantly when compared to single fueling cases. The volumetric efficiency is discussed in detail with the implications of manifold injection of hydrogen analyzed with the conclusions drawn.

Performance and management of IPR-R1 fuel elements

International Nuclear Information System (INIS)

Stasiulevicius, R.; Maretti Junior, F.

1983-01-01

The performance of fuel elements during the 23 years of the reactor operation, is presented aiming to introduce improvements in the fuel load distribution and consequent increase of the reactivity. A computer code CORE was developed aiming to calculate the individual burnup of the fuel elements and the value of the reactivity for several core configurations, establishing a routine to control the nuclear material in the IPR-R1. The values calculated were compared with the experimental results. Some alternatives to augment the reactivity of the present core are presented foreseeing the fuel load availability for operation with 100Km and, for angmenting the power reaction in a next stage. (E.G.) [pt

Evaluation of corrosion on the fuel performance of stainless steel cladding

Directory of Open Access Journals (Sweden)

de Souza Gomes Daniel

2016-01-01

Full Text Available In nuclear reactors, the use of stainless steel (SS as the cladding material offers some advantages such as good mechanical and corrosion resistance. However, its main advantage is the reduction in the amount of the hydrogen released during loss-of-coolant accident, as observed in the Fukushima Daiichi accident. Hence, research aimed at developing accident tolerant fuels should consider SS as an important alternative to existing materials. However, the available computational tools used to analyze fuel rod performance under irradiation are not capable of assessing the effectiveness of SS as the cladding material. This paper addresses the SS corrosion behavior in a modified fuel performance code in order to evaluate its effect on the global fuel performance. Then, data from the literature concerning to SS corrosion are implemented in the specific code subroutines, and the results obtained are compared to those for Zircaloy-4 (Zy-4 under the same power history. The results show that the effects of corrosion on SS are considerably different from those on Zy-4. The thickness of the oxide layer formed on the SS surface is considerably lower than that formed on Zy-4. As a consequence of this, the global fuel performance of SS under irradiation should be less affected by the corrosion.

Evaluation of fuel performance for fresh and aged CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Jung, Jong Yeob; Bae, Jun Ho; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

Like all other industrial plants, nuclear power plants also undergo degradations, so called ageing, with their operation time. Accordingly, in the recent safety analysis for a refurbished Wolsong 1 NPP, various ageing effects were incorporated into the hydraulic models of a number of the components in the primary heat transport system for conservatism. The ageing data of thermal-hydraulic components for 11 EFPY of Wolsong 1 were derived by using NUCIRC code based on the site operation data and they were modified to the appropriate input data for CATHENA code which is a thermal hydraulic code for a postulated accident analysis. This paper deals with the ageing effect of the PHTS (primary heat transport system) of CANDU reactor on the fuel performance during the normal operation. Initial conditions for fuel performance analysis were derived from the thermal-hydraulic analysis for both fresh and aged core models. Here, fresh core means a core state just right after the refurbishment and the aged core is 11 EFPY state after the refurbishment of Wolsong 1. The fuel performance was analyzed by using ELESTRES code for both fresh and aged core state and the results were compared in order to verify the ageing effect of CANDU HTS on the fuel performance.

Modeling constituent redistribution in U–Pu–Zr metallic fuel using the advanced fuel performance code BISON

International Nuclear Information System (INIS)

Galloway, J.; Unal, C.; Carlson, N.; Porter, D.; Hayes, S.

2015-01-01

Highlights: • An improved constituent distribution formulation in metallic nuclear fuels. • The new algorithm is implemented into the advanced fuel performance framework BISON. • Experimental Breeder Reactor-II data, T179, DP16, T459 are reanalyzed. • Phase dependent diffusion coefficients are improved. • Most influential phase is gamma, followed by alpha and thirdly the beta phase. - Abstract: An improved robust formulation for constituent distribution in metallic nuclear fuels is developed and implemented into the advanced fuel performance framework BISON. The coupled thermal diffusion equations are solved simultaneously to reanalyze the constituent redistribution in post irradiation data from fuel tests performed in Experimental Breeder Reactor-II (EBR-II). Deficiencies observed in previously published formulation and numerical implementations are also improved. The present model corrects an inconsistency between the enthalpies of solution and the solubility limit curves of the phase diagram while also adding an artificial diffusion term when in the 2-phase regime that stabilizes the standard Galerkin finite element (FE) method used by BISON. An additional improvement is in the formulation of zirconium flux as it relates to the Soret term. With these new modifications, phase dependent diffusion coefficients are revaluated and compared with the previously recommended values. The model validation included testing against experimental data from fuel pins T179, DP16 and T459, irradiated in EBR-II. A series of viable material properties for U–Pu–Zr based materials was determined through a sensitivity study, which resulted in three cases with differing parameters that showed strong agreement with one set of experimental data, rod T179. Subsequently a full-scale simulation of T179 was performed to reduce uncertainties, particularly relating to the temperature boundary condition for the fuel. In addition a new thermal conductivity model combining all

Performance and emissions of a dual-fuel pilot diesel ignition engine operating on various premixed fuels

International Nuclear Information System (INIS)

Yousefi, Amin; Birouk, Madjid; Lawler, Benjamin; Gharehghani, Ayatallah

2015-01-01

Highlights: • Natural gas/diesel, methanol/diesel, and hydrogen/diesel cases were investigated. • For leaner mixtures, the hydrogen/diesel case has the highest IMEP and ITE. • The methanol/diesel case has the maximum IMEP and ITE for richer mixtures. • Hydrogen/diesel case experiences soot and CO free combustion at rich regions. - Abstract: A multi-dimensional computational fluid dynamics (CFD) model coupled with chemical kinetics mechanisms was applied to investigate the effect of various premixed fuels and equivalence ratios on the combustion, performance, and emissions characteristics of a dual-fuel indirect injection (IDI) pilot diesel ignition engine. The diesel fuel is supplied via indirect injection into the cylinder prior to the end of the compression stroke. Various premixed fuels were inducted into the engine through the intake manifold. The results showed that the dual-fuel case using hydrogen/diesel has a steeper pressure rise rate, higher peak heat release rate (PHRR), more advanced ignition timing, and shorter ignition delay compared to the natural gas/diesel and methanol/diesel dual-fuel cases. For leaner mixtures (Φ_P 0.32). For instance, with an equivalence ratio of 0.35, the ITE is 56.24% and 60.85% for hydrogen/diesel and methanol/diesel dual-fuel cases, respectively. For an equivalence ratio of 0.15, the natural gas/diesel simulation exhibits partial burn combustion and thus results in a negative IMEP. At equivalence ratios of 0.15, 0.2, and 0.25, the methanol/diesel case experiences misfiring phenomenon which consequently deteriorates the engine performance considerably. As for the engine-out emissions, the hydrogen/diesel results display carbon monoxide (CO) free combustion relative to natural gas/diesel and methanol/diesel engines; however, considerable amount of nitrogen oxides (NO_x) emissions are produced at an equivalence ratio of 0.35 which exceeds the Euro 6 NO_x limit. Due to the larger area exposed to high temperature regions

Fuel performance of licensed nuclear power plants through 1974

International Nuclear Information System (INIS)

Bobe, P.E.

1976-01-01

General aspects of fuel element design and specific design data for typical Pressurized and Boiling Water Reactors are presented. Based on a literature search, failure modes and specific failures incurred through December 31, 1974 are described, together with a discussion of those problems which have had a significant impact upon plant operation. The relationship between fuel element failures and the resultant coolant activity/radioactive gaseous effluents upon radiation exposure, plant availability and capacity factors, economic impact, and waste management, are discussed. An assessment was made regarding the generation, availability, and use of fuel performance data

Fuel Chemistry Division annual progress report for 1990

International Nuclear Information System (INIS)

Vaidyanathan, R.

1993-01-01

The progress report gives brief descriptions of the various activities of the Fuel Chemistry Division of Bhabha Atomic Research Centre, Bombay for the year 1990. The descriptions of activities are arranged under the headings: Fuel Development Chemistry, Chemistry of Actinides, Quality Control of Nuclear Fuels, and studies related to Nuclear Materials Accounting. At the end of the report, a list of papers published in journals and presented at various conferences/symposia is also given. (author). 7 figs., 52 tabs

Fuel Chemistry Division: annual progress report for 1988

International Nuclear Information System (INIS)

Vaidyanathan, S.

1991-01-01

The progress report gives the brief descriptions of various activites of the Fuel Chemistry Division of Bhabha Atomic Research Centre, Bombay for the year 1988. The descriptions of activities are arranged under the headings: Fuel Development Chemistry of Actinides, Quality Control of Fuel, and Studies related to Nuclear Material Accounting. At the end of report, a list of publications published in journals and papers presented at various conferences/symposia during 1988 is given. (author). 13 figs., 61 tabs

Electric power annual, 1991

International Nuclear Information System (INIS)

1993-01-01

The Electric Power Annual is prepared by the Survey Management Division; Office of Coal, Nuclear, Electric and Alternate Fuels; Energy Information Administration (EIA); US Department of Energy. The 1991 edition has been enhanced to include statistics on electric utility demand-side management and nonutility supply. ''The US Electric Power Industry at a Glance'' section presents a profile of the electric power industry ownership and performance, and a review of key statistics for the year. Subsequent sections present data on generating capability, including proposed capability additions; net generation; fossil-fuel statistics; electricity sales, revenue, and average revenue per kilowatthour sold; financial statistics; environmental statistics; electric power transactions; demand-side management; and nonutility power producers. In addition, the appendices provide supplemental data on major disturbances and unusual occurrences in US electricity power systems. Each section contains related text and tables and refers the reader to the appropriate publication that contains more detailed data on the subject matter. Monetary values in this publication are expressed in nominal terms

Fuel leak testing performance at NPP Jaslovske Bohunice

International Nuclear Information System (INIS)

Slugen, V.; Krnac, S.; Smiesko, I.

1995-01-01

The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR's. 1 tab., 2 figs., 3 refs

Fuel leak testing performance at NPP Jaslovske Bohunice

Energy Technology Data Exchange (ETDEWEB)

Slugen, V; Krnac, S [Slovak Technical Univ., Bratislava (Slovakia); Smiesko, I [Nuclear Powr Plant EBO, Jaslovske Bohuce (Slovakia)

1996-12-31

The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR`s. 1 tab., 2 figs., 3 refs.

Performance assessment analyses unique to Department of Energy spent nuclear fuel

International Nuclear Information System (INIS)

Loo, H.H.; Duguid, J.J.

2000-01-01

This paper describes the iterative process of grouping and performance assessment that has led to the current grouping of the U.S. Department of Energy (DOE) spent nuclear fuel (SNF). The unique sensitivity analyses that form the basis for incorporating DOE fuel into the total system performance assessment (TSPA) base case model are described. In addition, the chemistry that results from dissolution of DOE fuel and high level waste (HLW) glass in a failed co-disposal package, and the effects of disposal of selected DOE SNF in high integrity cans are presented

Improving the performance of dual fuel engines running on natural gas/LPG by using pilot fuel derived from jojoba seeds