Site selection - siting of the final repository for spent nuclear fuel

International Nuclear Information System (INIS)

2011-03-01

SKB has selected Forsmark as the site for the final repository for spent nuclear fuel. The site selection is the end result of an extensive siting process that began in the early 1990s. The strategy and plan for the work was based on experience from investigations and development work over a period of more than ten years prior to then. This document describes the siting work and SKB's choice of site for the final repository. It also presents the information on which the choice was based and the reasons for the decisions made along the way. The document comprises Appendix PV to applications under the Nuclear Activities Act and the Environmental Code for licences to build and operate an encapsulation plant adjacent to the central interim storage facility for spent nuclear fuel in Oskarshamn, and to build and operate a final repository for spent nuclear fuel in Forsmark in Oesthammar Municipality

Site selection - siting of the final repository for spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

2011-03-15

SKB has selected Forsmark as the site for the final repository for spent nuclear fuel. The site selection is the end result of an extensive siting process that began in the early 1990s. The strategy and plan for the work was based on experience from investigations and development work over a period of more than ten years prior to then. This document describes the siting work and SKB's choice of site for the final repository. It also presents the information on which the choice was based and the reasons for the decisions made along the way. The document comprises Appendix PV to applications under the Nuclear Activities Act and the Environmental Code for licences to build and operate an encapsulation plant adjacent to the central interim storage facility for spent nuclear fuel in Oskarshamn, and to build and operate a final repository for spent nuclear fuel in Forsmark in Oesthammar Municipality

HEU to LEU fuel conversion. Final report

International Nuclear Information System (INIS)

Mulder, R.U.

1994-10-01

The Nuclear Regulatory Commission issued a ruling, effective March 27, 1986, that all U.S. non-power reactors convert from HEU fuel to LEU fuel. A Reduced Enrichment for Research and Test Reactors Program was conducted by the Department of Energy at Argonne National Laboratory to coordinate the development of the high density LEU fuel and assist in the development of Safety Analysis Reports for the smaller non-power reactors. Several meetings were held at Argonne in 1987 with the non-power reactor community to discuss the conversion and to set up a conversion schedule for university reactors. EG ampersand G at Idaho was assigned the coordination of the fuel element redesigns. The fuel elements were manufactured by the Babcock ampersand Wilcox Company in Lynchburg, Virginia. The University of Virginia was awarded a grant by the DOE Idaho Operations Office in 1988 to perform safety analysis studies for the LEU conversion for its 2 MW UVAR and 100 Watt CAVALIER reactors. The University subsequently decided to shut down the CAVALIER reactor. A preliminary SAR on the UVAR, along with Technical Specification changes, was submitted to the NRC in November, 1990. An updated SAR was approved by the NRC in January, 1991. In September, 1992, representatives from the fuel manufacturer (B ampersand W) and the fuel designer (EG ampersand G, Idaho) came to the UVAR facility to observe trial fittings of new 22 plate LEU mock fuel elements. B ampersand W fabricated two non-fuel bearing elements, a regular 22 plate element and a control rod element. The elements were checked against the drawings and test fitted in the UVAR grid plate. The dimensions were acceptable and the elements fit in the grid plate with no problems. The staff made several suggestions for minor construction changes to the end pieces on the elements, which were incorporated into the final design of the actual fuel elements. Selected papers are indexed separately for inclusion in the Energy Science and Technology

HEU to LEU fuel conversion. Final report

Energy Technology Data Exchange (ETDEWEB)

Mulder, R.U.

1994-10-01

The Nuclear Regulatory Commission issued a ruling, effective March 27, 1986, that all U.S. non-power reactors convert from HEU fuel to LEU fuel. A Reduced Enrichment for Research and Test Reactors Program was conducted by the Department of Energy at Argonne National Laboratory to coordinate the development of the high density LEU fuel and assist in the development of Safety Analysis Reports for the smaller non-power reactors. Several meetings were held at Argonne in 1987 with the non-power reactor community to discuss the conversion and to set up a conversion schedule for university reactors. EG&G at Idaho was assigned the coordination of the fuel element redesigns. The fuel elements were manufactured by the Babcock & Wilcox Company in Lynchburg, Virginia. The University of Virginia was awarded a grant by the DOE Idaho Operations Office in 1988 to perform safety analysis studies for the LEU conversion for its 2 MW UVAR and 100 Watt CAVALIER reactors. The University subsequently decided to shut down the CAVALIER reactor. A preliminary SAR on the UVAR, along with Technical Specification changes, was submitted to the NRC in November, 1990. An updated SAR was approved by the NRC in January, 1991. In September, 1992, representatives from the fuel manufacturer (B&W) and the fuel designer (EG&G, Idaho) came to the UVAR facility to observe trial fittings of new 22 plate LEU mock fuel elements. B&W fabricated two non-fuel bearing elements, a regular 22 plate element and a control rod element. The elements were checked against the drawings and test fitted in the UVAR grid plate. The dimensions were acceptable and the elements fit in the grid plate with no problems. The staff made several suggestions for minor construction changes to the end pieces on the elements, which were incorporated into the final design of the actual fuel elements. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Final Report on the Fuel Saving Effectiveness of Various Driver Feedback Approaches

Energy Technology Data Exchange (ETDEWEB)

Gonder, J.; Earleywine, M.; Sparks, W.

2011-03-01

This final report quantifies the fuel-savings opportunities from specific driving behavior changes, identifies factors that influence drivers' receptiveness to adopting fuel-saving behaviors, and assesses various driver feedback approaches.

Fuel Cooling in Absence of Forced Flow at Shutdown Condition with PHTS Partially Drained

Energy Technology Data Exchange (ETDEWEB)

Parasca, L.; Pecheanu, D.L., E-mail: laurentiu.parasca@cne.ro, E-mail: doru.pecheanu@cne.ro [Cernavoda Nuclear Power Plant, Cernavoda (Romania)

2014-09-15

During the plant outage for maintenance on primary side (e.g. for the main Heat Transport System pumps maintenance, the Steam Generators inspection), there are situations which require the primary heat transport system (HTS) drainage to a certain level for opening the circuit. The primary fuel heat sink for this configuration is provided by the shutdown cooling system (SDCS). In case of losing the forced cooling (e.g. due to the loss of SDCS, design basis earthquake-DBE), flow conditions in the reactor core may become stagnant. Inside the fuel channels, natural circulation phenomena known as Intermittent Buoyancy Induced Flow (IBIF) will initiate, providing an alternate heat sink mechanism for the fuel. However, this heat sink is effective only for a limited period of time (recall time). The recall time is defined as the elapsed time until the water temperature in the HTS headers exceeds a certain limit. Until then, compensatory measures need to be taken (e.g. by re-establishing the forced flow or initiate Emergency Core Cooling system injection) to preclude fuel failures. The present paper briefly presents the results of an analysis performed to demonstrate that fuel temperature remains within acceptable limits during IBIF transient. One of the objectives of this analysis was to determine the earliest moment since the reactor shut down when maintenance activities on the HTS can be started such that IBIF is effective in case of losing the forced circulation. The resulting peak fuel sheath and pressure tube temperatures due to fuel heat up shall be within the acceptable limits to preclude fuel defect or fuel channel defects.Thermalhydraulic circuit conditions were obtained using a CATHENA model for the primary side of HTS (drained to a certain level), an ECC system model and a system model for SDCS. A single channel model was developed in GOTHIC code for the fuel assessment analysis. (author)

Fuel safety criteria technical review - Results of OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria

International Nuclear Information System (INIS)

Hollasky, N.; Valtonen, K.; Hache, G.; Gross, H.; Bakker, K.; Recio, M.; Bart, G.; Zimmermann, M.; Van Doesburg, W.; Killeen, J.; Meyer, R.O.; Speis, T.

2000-01-01

With the advent of advanced fuel and core designs, the adoption of more aggressive operational modes and the implementation of more accurate (best estimate or statistical) design and analysis methods, there is a concern if safety margins have remained adequate. Most - if not all - of the currently existing safety criteria were established during the 60's and early 70's, and verified against experiments with fuel that was available at that time, mostly with unirradiated specimens. Verification was of course performed as designs progressed in later years, however mostly with the aim to be able to prove that these designs adequately complied with existing criteria, and not to establish new limits. The OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria (TFFSC) was therefore given the mandate to technically review the existing fuel safety criteria, focusing on the 'new design' elements (new fuel and core design, cladding materials, manufacturing processes, high burnup, MOX, etc.) introduced by the industry. It should also identify if additional efforts may be required (experimental, analytical) to ensure that the basis for fuel safety criteria is adequate to address the relevant safety issues. In this report, fuel-related criteria are discussed without attempting to categorize them according to event type or risk significance. For each of these 20 criteria, we present a brief description of the criterion as it is used in several applications along with the rationale for having such a criterion. New design elements, such as different cladding materials, higher burnup, and the use of MOX fuels, can affect fuel-related margins and, in some cases, the criteria themselves. Some of the more important effects are mentioned in order to indicate whether the criteria need to be re-evaluated. The discussion may not cover all possible effects, but should be sufficient to identify those criteria that need to be addressed. A summary of these discussions is given in Section 7. As part

Handling final storage of unreprocessed spent nuclear fuel

International Nuclear Information System (INIS)

1978-01-01

The present second report from KBS describes how the safe final storage of spent unreprocessed nuclear fuel can be implemented. According to the Swedish Stipulation Law, the owner must specify in which form the waste is to be stored, how final storage is to be effected, how the waste is to be transported and all other aspects of fuel handling and storage which must be taken into consideration in judging whether the proposed final storage method can be considered to be absolutely safe and feasible. Thus, the description must go beyond general plans and sketches. The description is therefore relatively detailed, even concerning those parts which are less essential for evaluating the safety of the waste storage method. For those parts of the handling chain which are the same for both alternatives of the Stipulation Law, the reader is referred in some cases to the first report. Both of the alternatives of the Stipulation Law may be used in the future. Handling equipment and facilities for the two storage methods are so designed that a combination in the desired proportions is practically feasible. In this first part of the report are presented: premises and data, a description of the various steps of the handling procedure, a summary of dispersal processes and a safety analysis. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-30

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

Physics-based modeling of live wildland fuel ignition experiments in the Forced Ignition and Flame Spread Test apparatus

Science.gov (United States)

C. Anand; B. Shotorban; S. Mahalingam; S. McAllister; D. R. Weise

2017-01-01

A computational study was performed to improve our understanding of the ignition of live fuel in the forced ignition and flame spread test apparatus, a setup where the impact of the heating mode is investigated by subjecting the fuel to forced convection and radiation. An improvement was first made in the physics-based model WFDS where the fuel is treated as fixed...

Light-duty vehicle greenhouse gas emission standards and corporate average fuel economy standards : final rule

Science.gov (United States)

2010-05-07

Final Rule to establish a National Program consisting of new standards for light-duty vehicles that will reduce greenhouse gas emissions and improve fuel economy. This joint : Final Rule is consistent with the National Fuel Efficiency Policy announce...

MaxiFuels. Final report; Maxi-Fuels. Afslutningsrapport

Energy Technology Data Exchange (ETDEWEB)

2008-07-01

The MaxiFuel pilot plant opened September 2006 with a view to testing and developing a competitive integrated concept for 2nd generation bioethanol production from lignocellulosic biomass such as straw. This makes the raw material cheaper than crops cultivated for energy production purposes only, and the bioethanol production will not compete with production of food. The MaxiFuels concept is a patented technology which has proven in lab-scale to have the potential of producing bioethanol from residual biomass for a lower price than other existing 2nd generation bioethanol concepts. Lately a suitable pre-treatment process, Wet Explosion, has been developed, and enzymes and C6 fermentation based on yeast have been tested and further developed on pre-treated straw. Furthermore a special micro organism has been isolated, genetically engineered and tested on the pre-treated biomass. The micro organism converts C5 sugars to ethanol. Finally an attempt has been made to produce biogas from the remaining organic material, which can't be fermented to ethanol. All the abovementioned processes have been developed in lab-scale. The operation of these processes at the same time in one concept in pilot scale makes it possible to evaluate the concept's productivity in full-scale. The data of the pilot plant operation have been used for the set-up of an economical model of the whole concept. Based on this model a full-scale plant has been designed and a business plan for the implementation of this step has been elaborated. (BA)

Inhalation exposure to jet fuel (JP8) among U.S. Air Force personnel.

Science.gov (United States)

Smith, Kristen W; Proctor, Susan P; Ozonoff, Al; McClean, Michael D

2010-10-01

As jet fuel is a common occupational exposure among military and civilian populations, this study was conducted to characterize jet fuel (JP8) exposure among active duty U.S. Air Force personnel. Personnel (n = 24) were divided a priori into high, moderate, and low exposure groups. Questionnaires and personal air samples (breathing zone) were collected from each worker over 3 consecutive days (72 worker-days) and analyzed for total hydrocarbons (THC), benzene, toluene, ethylbenzene, xylenes, and naphthalene. Air samples were collected from inside the fuel tank and analyzed for the same analytes. Linear mixed-effects models were used to evaluate the exposure data. Our results show that the correlation of THC (a measure of overall JP8 inhalation exposure) with all other analytes was moderate to strong in the a priori high and moderate exposure groups combined. Inhalation exposure to all analytes varied significantly by self-reported JP8 exposure (THC levels higher among workers reporting JP8 exposure), a priori exposure group (THC levels in high group > moderate group > low group), and more specific job task groupings (THC levels among workers in fuel systems hangar group > refueling maintenance group > fuel systems office group > fuel handling group > clinic group), with task groupings explaining the most between-worker variability. Among highly exposed workers, statistically significant job task-related predictors of inhalation exposure to THC indicated that increased time in the hangar, working close to the fuel tank (inside > less than 25 ft > greater than 25 ft), primary job (entrant > attendant/runner/fireguard > outside hangar), and performing various tasks near the fuel tank, such as searching for a leak, resulted in higher JP8 exposure. This study shows that while a priori exposure groups were useful in distinguishing JP8 exposure levels, job task-based categories should be considered in epidemiologic study designs to improve exposure classification. Finally

The final management of nuclear fuel. Legal and economic aspects; La gestion final del combustible nuclear. Aspectos legales y economicos

Energy Technology Data Exchange (ETDEWEB)

Villota, C. de

2009-07-01

This article gives a brief summary of the characteristics of spent fuel and the lines of action considered for its management. It describes the legal framework that supports the Radioactive Waste Management Plan (PGRR), which contains the lines applicable to Final Spent Fuel Management, as well as the evolution of this legal framework. The article contains the 2008 updated costs of the various items of the PGRR, with a more detailed description of those related to this type of fuel, as well the source and amount of the financial contributions to the fund for meeting these costs, including how they have evolved over time. finally, it provides some personal reflections on this issue. (Author)

Fluid elastic vibration of nuclear fuel assemblies

International Nuclear Information System (INIS)

Kim, S. N.; Jung, S. Y.

1998-01-01

Since utilities and fuel venders have adopted the fuel design of high burn-up and improved thermal margin flow mixing vane, several PWR nuclear power plants have in recent years experienced fretting wear fuel rod failure due to flow induced vibration. Flow induced vibration can be resulted from fluidelastic instability, periodic shedding, turbulence-induced excitation, and acoustic resonance (1). Among these mechanisms found in the core of nuclear power plant, the governing mechanism that is fluidelastic instability, could be inferred from the analysis of fuel failure patterns. Therefore, to simulate the fuel failure in nuclear power plants, Tanaka's model (2) was chosen as most suitable one, which is well explaining the damage pattern, in particular it's second row damage characteristics. In the model, unsteady fluid dynamic forces acting on the vibrating cyclinders were included which consists of the inertia forces due to the added mass of fluid, damping forces of fluid in phase to the cylinder vibrating velocity, and stiffness forces proportional to cylinder displacements. However, the model did not account for radiation effect-spring forces deflection. So, the model was modified to account for the spring force relaxation due to radiation exposure. The stiffness of spring was fitted with experimental data. Finally the critical velocities were calculated with the modified spring force at beginning and end of cycle

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

International Nuclear Information System (INIS)

1983-03-01

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

Demonstration of fuel resistant to pellet-cladding interaction. Phase I. Final report

International Nuclear Information System (INIS)

Rosenbaum, H.S.

1979-03-01

This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel, and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress, and reactive fission products during reactor service. This is the final report for PHASE 1 of this program. Support tests have shown that the barrier fuel resists PCI far better than does the conventional Zircaloy-clad fuel. Power ramp tests thus far have shown good PCI resistance for Cu-barrier fuel at burnup > 12 MWd/kg-U and for Zr-liner fuel > 16 MWd/kg-U. The program calls for continued testing to still higher burnup levels in PHASE 2

Achilles tests finally nail PWR fuel clad ballooning fears

International Nuclear Information System (INIS)

Dore, P.; McMinn, K.

1992-01-01

A conclusive series of experiments carried out by AEA Reactor Services at its Achilles rig in the UK has finally allayed fears that fuel clad ballooning is a major safety problem for Sizewell B, Britain's first Pressurized Water Reactor. The experiments are described in this article. (author)

Exposures to jet fuel and benzene during aircraft fuel tank repair in the U.S. Air Force.

Science.gov (United States)

Carlton, G N; Smith, L B

2000-06-01

Jet fuel and benzene vapor exposures were measured during aircraft fuel tank entry and repair at twelve U.S. Air Force bases. Breathing zone samples were collected on the fuel workers who performed the repair. In addition, instantaneous samples were taken at various points during the procedures with SUMMA canisters and subsequent analysis by mass spectrometry. The highest eight-hour time-weighted average (TWA) fuel exposure found was 1304 mg/m3; the highest 15-minute short-term exposure was 10,295 mg/m3. The results indicate workers who repair fuel tanks containing explosion suppression foam have a significantly higher exposure to jet fuel as compared to workers who repair tanks without foam (p fuel, absorbed by the foam, to volatilize during the foam removal process. Fuel tanks that allow flow-through ventilation during repair resulted in lower exposures compared to those tanks that have only one access port and, as a result, cannot be ventilated efficiently. The instantaneous sampling results confirm that benzene exposures occur during fuel tank repair; levels up to 49.1 mg/m3 were found inside the tanks during the repairs. As with jet fuel, these elevated benzene concentrations were more likely to occur in foamed tanks. The high temperatures associated with fuel tank repair, along with the requirement to wear vapor-permeable cotton coveralls for fire reasons, could result in an increase in the benzene body burden of tank entrants.

Final safety analysis report for the irradiated fuels storage facility

International Nuclear Information System (INIS)

Bingham, G.E.; Evans, T.K.

1976-01-01

A fuel storage facility has been constructed at the Idaho Chemical Processing Plant to provide safe storage for spent fuel from two commercial HTGR's, Fort St. Vrain and Peach Bottom, and from the Rover nuclear rocket program. The new facility was built as an addition to the existing fuel storage basin building to make maximum use of existing facilities and equipment. The completed facility provides dry storage for one core of Peach Bottom fuel (804 elements), 1 1 / 2 cores of Fort St. Vrain fuel (2200 elements), and the irradiated fuel from the 20 reactors in the Rover program. The facility is designed to permit future expansion at a minimum cost should additional storage space for graphite-type fuels be required. A thorough study of the potential hazards associated with the Irradiated Fuels Storage Facility has been completed, indicating that the facility is capable of withstanding all credible combinations of internal accidents and pertinent natural forces, including design basis natural phenomena of a 10,000 year flood, a 175-mph tornado, or an earthquake having a bedrock acceleration of 0.33 g and an amplification factor of 1.3, without a loss of integrity or a significant release of radioactive materials. The design basis accident (DBA) postulated for the facility is a complete loss of cooling air, even though the occurrence of this situation is extremely remote, considering the availability of backup and spare fans and emergency power. The occurrence of the DBA presents neither a radiation nor an activity release hazard. A loss of coolant has no effect upon the fuel or the facility other than resulting in a gradual and constant temperature increase of the stored fuel. The temperature increase is gradual enough that ample time (28 hours minimum) is available for corrective action before an arbitrarily imposed maximum fuel centerline temperature of 1100 0 F is reached

TMI-2 Lessons Learned Task Force. Final report

International Nuclear Information System (INIS)

1979-10-01

In its final report reviewing the Three Mile Island accident, the TMI-2 Lessons Learned Task Force has suggested change in several fundamental aspects of basic safety policy for nuclear power plants. Changes in nuclear power plant design and operations and in the regulatory process are discussed in terms of general goals. The appendix sets forth specific recommendations for reaching these goals

Ontario Select Committee on Alternative Fuel Sources : Final Report

International Nuclear Information System (INIS)

Galt, D.

2002-06-01

On June 28, 2001, the Ontario Legislative Assembly appointed the Select Committee an Alternative Fuel Sources, comprised of representatives of all parties, with a broad mandate to investigate, report and offer recommendations with regard to the various options to support the development and application of environmentally sustainable alternatives to the fossil fuel sources already existing. The members of the Committee elected to conduct extensive public hearings, conduct site visits, attend relevant conferences, do some background research to examine a vast number of alternative fuel and energy sources that could be of relevance to the province of Ontario. A discussion paper (interim report) was issued by the Committee in November 2001, and the present document represents the final report, containing 141 recommendations touching 20 topics. The information contained in the report is expected to assist in the development and outline of policy and programs designed to specifically support alternative fuels and energy sources and applicable technologies. Policy issues were discussed in Part A of the report, along with the appropriate recommendations. The recommendations on specific alternative fuels and energy sources were included in Part B of the report. It is believed that the dependence of Ontario on traditional petroleum-based fuels and energy sources can be reduced through aggressive action on alternative fuels and energy. The benefits of such action would be felt in the area of air quality, with social, and economic benefits as well. 3 tabs

Safeguards for final disposal of spent nuclear fuel. Methods and technologies for the Olkiluoto site

International Nuclear Information System (INIS)

Okko, O.

2003-05-01

The final disposal of the nuclear material shall introduce new safeguards concerns which have not been addressed previously in IAEA safeguards approaches for spent fuel. The encapsulation plant to be built at the site will be the final opportunity for verification of spent fuel assemblies prior to their transfer to the geological repository. Moreover, additional safety and safeguards measures are considered for the underground repository. Integrated safeguards verification systems will also concentrate on environmental monitoring to observe unannounced activities related to possible diversion schemes at the repository site. The final disposal of spent nuclear fuel in geological formation will begin in Finland within 10 years. After the geological site investigations and according to legal decision made in 2001, the final repository of the spent nuclear fuel shall be located at the Olkiluoto site in Eurajoki. The next phase of site investigations contains the construction of an underground facility, called ONKALO, for rock characterisation purposes. The excavation of the ONKALO is scheduled to start in 2004. Later on, the ONKALO may form a part of the final repository. The plans to construct the underground facility for nuclear material signify that the first safeguards measures, e.g. baseline mapping of the site area, need to take prior to the excavation phase. In order to support the development and implementation of the regulatory control of the final disposal programme, STUK established an independent expert group, LOSKA. The group should support the STUK in the development of the technical safeguards requirements, in the implementation of the safeguards and in the evaluation of the plans of the facility operator. This publication includes four background reports produced by this group. The first of these 'NDA verification of spent fuel, monitoring of disposal canisters, interaction of the safeguards and safety issues in the final disposal' describes the new

Impact forces on a core shroud of an excited PWR fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Collard, B.; Vallory, J. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

2001-07-01

Seismic excitation of PWR internals may induce large motions of the fuel assemblies (FA). This could result in impact between assemblies or between assemblies and core shroud. Forces generated during these shocks are often the basis for the maximum design loads of the spacer grids and fuel rods. An experimental program has been conducted at the French Nuclear Reactor Directorate (CEA) to measure the impact forces of a reduced scale FA on the test section under different environmental conditions. Within the framework of the tests presented, the effect of the FA environment (air, stagnant water, water under flow) on the maximum impact forces measured at grid levels and on the energy dissipated during the shock is examined. A 'fluid cushioning' effect (dissipative) between the grids and the wall is sought. Experimental results show that the axial flow has a great influence on the impact forces. The greater the axial flow velocity is, the lower the impact forces are. The tests of impact of an assembly on a wall were analyzed compared to the tests carried out without impact. This analysis related on the measured forces of impact and the variation of the measured/computed total energy of the system. The whole of these tests in air and water shows that the 'fluid cushioning' effect required exists but is not significant. Thus the presence of water does not decrease the forces of impact, and does not amplify the quantity of energy dissipated during the shock. The fact that the 'fluid cushioning' effect is weak compared to more analytical tests probably comes from our 'not perfect' or 'realistic' conditions of tests which involve an angle between the grid and the wall at the shock moment.

Impact forces on a core shroud of an excited PWR fuel assembly

International Nuclear Information System (INIS)

Collard, B.; Vallory, J.

2001-01-01

Seismic excitation of PWR internals may induce large motions of the fuel assemblies (FA). This could result in impact between assemblies or between assemblies and core shroud. Forces generated during these shocks are often the basis for the maximum design loads of the spacer grids and fuel rods. An experimental program has been conducted at the French Nuclear Reactor Directorate (CEA) to measure the impact forces of a reduced scale FA on the test section under different environmental conditions. Within the framework of the tests presented, the effect of the FA environment (air, stagnant water, water under flow) on the maximum impact forces measured at grid levels and on the energy dissipated during the shock is examined. A 'fluid cushioning' effect (dissipative) between the grids and the wall is sought. Experimental results show that the axial flow has a great influence on the impact forces. The greater the axial flow velocity is, the lower the impact forces are. The tests of impact of an assembly on a wall were analyzed compared to the tests carried out without impact. This analysis related on the measured forces of impact and the variation of the measured/computed total energy of the system. The whole of these tests in air and water shows that the 'fluid cushioning' effect required exists but is not significant. Thus the presence of water does not decrease the forces of impact, and does not amplify the quantity of energy dissipated during the shock. The fact that the 'fluid cushioning' effect is weak compared to more analytical tests probably comes from our 'not perfect' or 'realistic' conditions of tests which involve an angle between the grid and the wall at the shock moment

Alternative Low-Sulfur Diesel Fuel Transition Program for Alaska Final Rule

Science.gov (United States)

This final rule will implement the requirements for sulfur, cetane and aromatics for highway, nonroad, locomotive and marine diesel fuel produced in, imported into, and distributed or used in the rural areas of Alaska.

Getting ready for final disposal in Finland - Independent verification of spent fuel

International Nuclear Information System (INIS)

Tarvainen, Matti; Honkamaa, Tapani; Martikka, Elina; Varjoranta, Tero; Hautamaeki, Johanna; Tiitta, Antero

2001-01-01

Full text: Final disposal of spent nuclear fuel has been known to be the solution for the back-end of the fuel cycle in Finland already for a long time. This has allowed the State system for accounting and control (SSAC) to prepare for the safeguards requirements in time. The Finnish SSAC includes the operator, the State authority STUK and the parties above them e.g. the Ministry for Trade and Industry. Undisputed responsibility of the safe disposal of spent fuel is on the operator. The role of the safety authority STUK. is to set up detailed requirements, to inspect the operator plans and by using different tools of a quality audit approach to verity that the requirements will be complied with in practice. Responsibility on the safeguards issues is similar with the addition of the role of the regional and the international verification organizations represented by Euratom and the IAEA, As the competent safeguards authority, STUK has decided to maintain its active role also in the future. This will be reflected in the future in the increasing cooperation between the SSAC and the IAEA in the new safeguards activities related to the Additional Protocol. The role of Euratom will remain the same concerning the implementation of conventional safeguards. Based on its SSAC role, STUK has continued carrying out safeguards inspections including independent verification measurements on spent fuel also after joining the EU and Euratom safeguards in 1995. Verification of the operator declared data is the key verification element of safeguards. This will remain to be the case also under the Integrated Safeguards (IS) in the future. It is believed that the importance of high quality measurements will rather increase than decrease when the frequency of interim inspections will decrease. Maintaining the continuity of knowledge makes sense only when the knowledge is reliable and independently verified. One of the corner stones of the high quality of the Finnish SSAC activities is

Decision nearing on final disposal of spent fuel in Finland

International Nuclear Information System (INIS)

Vira, J.

2000-01-01

The programme for final disposal of spent fuel from Finnish nuclear power plants is entering into important phase: in the year 2000 the Finnish Government is expected to decide whether the proposal made by Posiva Oy on the spent fuel disposal is in line with the overall good of society. Associated with the decision is also Posiva's proposal on siting the disposal facility at Olkiluoto in Eurajoki municipality on the western coast of Finland. An important document underlying Posiva's application for this principle decision is the report of the environmental impact assessment, which was completed in 1999. Safety considerations play an important role in the application. New assessments have, therefore, been made on both the operational and long-term safety as well as on safety of spent fuel transportation. (author)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Handling of spent nuclear fuel and final storage of nitrified high level reprocessing waste

International Nuclear Information System (INIS)

The following stages of handling and transport of the fuel on its way to final storage are dealt with in the report. 1) The spent nuclear fuel is stored at the power station or in the central fuel storage facility awaiting reprocessing. 2) The fuel is reprocessed, i.e. uranium, plutonium and waste are separated from each other. Reprocessing does not take place in Sweden. The highlevel waste is vitrified and can be sent back to Sweden in the 1990s. 3) Vitrified waste is stored for about 30 years awaiting deposition in the final repository. 4) The waste is encapsulated in highly durable materials to prevent groundwater from coming into contact with the waste glass while the radioactivity of the waste is still high. 5) The canisters are emplaced in a final repository which is built at a depth of 500 m in rock of low permeability. 6) All tunnels and shafts are filled with a mixture of clay and sand of low permeability. A detailed analysis of possible harmful effects resulting from normal acitivties and from conceivable accidents is presented in a special section. (author)

Site-selection studies for final disposal of spent fuel in Finland

International Nuclear Information System (INIS)

Vuorela, P.; Aeikaes, T.

1984-02-01

In the management of waste by the Industrial Power Company Ltd. (TVO) preparations are being made for the final disposal of unprocessed spent fuel into the Finnish bedrock. The site selection program will advance in three phases. The final disposal site must be made at the latest by the end of the year 2000, in accordance with a decision laid down by the Finnish Government. In the first phase, 1983-85, the main object is to find homogeneous stable bedrock blocks surrounded by fracture zones located at a safe distance from the planned disposal area. The work usually starts with a regional structural analysis of mosaics of Landsat-1 winter and summer imagery. Next an assortment of different maps, which cover the whole country, is used. Technical methods for geological and hydrogeological site investigations are being developed during the very first phase of the studies, and a borehole 1000 meters deep will be made in southwestern Finland. Studies for the final disposal of spent fuel or high-level reprocessing waste have been made since 1974 in Finland. General suitability studies of the bedrock have been going on since 1977. The present results indicate that suitable investigation areas for the final disposal of highly active waste can be found in Finland

The final disposal facility of spent nuclear fuel

International Nuclear Information System (INIS)

Prvakova, S.; Necas, V.

2001-01-01

Today the most serious problem in the area of nuclear power engineering is the management of spent nuclear fuel. Due to its very high radioactivity the nuclear waste must be isolated from the environment. The perspective solution of nuclear fuel cycle is the final disposal into geological formations. Today there is no disposal facility all over the world. There are only underground research laboratories in the well developed countries like the USA, France, Japan, Germany, Sweden, Switzerland and Belgium. From the economical point of view the most suitable appears to build a few international repositories. According to the political and social aspect each of the country prepare his own project of the deep repository. The status of those programmes in different countries is described. The development of methods for the long-term management of radioactive waste is necessity in all countries that have had nuclear programmes. (authors)

Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

Energy Technology Data Exchange (ETDEWEB)

Bragg-Sitton, Shannon Michelle [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force (TF1) includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, and identification of fuel performance and system codes applicable to ATF evaluation. The Cladding and Core Materials (TF2) and Fuel Concepts (TF3) task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment task force is chaired by Shannon Bragg-Sitton (Idaho National Laboratory [INL], U.S.), the Cladding Task Force is chaired by Marie Moatti (Electricite de France [EdF], France), and the Fuels Task Force is chaired by a Masaki Kurata (Japan Atomic Energy Agency [JAEA], Japan). The original Expert Group mandate was established for June 2014 to June 2016. In April 2016 the Expert Group voted to extend the mandate one additional year to June 2017 in order to complete the task force deliverables; this request was subsequently approved by the Nuclear Science Committee. This

Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

International Nuclear Information System (INIS)

Bragg-Sitton, Shannon Michelle

2016-01-01

The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force (TF1) includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, and identification of fuel performance and system codes applicable to ATF evaluation. The Cladding and Core Materials (TF2) and Fuel Concepts (TF3) task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment task force is chaired by Shannon Bragg-Sitton (Idaho National Laboratory [INL], U.S.), the Cladding Task Force is chaired by Marie Moatti (Electricite de France [EdF], France), and the Fuels Task Force is chaired by a Masaki Kurata (Japan Atomic Energy Agency [JAEA], Japan). The original Expert Group mandate was established for June 2014 to June 2016. In April 2016 the Expert Group voted to extend the mandate one additional year to June 2017 in order to complete the task force deliverables; this request was subsequently approved by the Nuclear Science Committee. This

Forced-convection boiling tests performed in parallel simulated LMR fuel assemblies

International Nuclear Information System (INIS)

Rose, S.D.; Carbajo, J.J.; Levin, A.E.; Lloyd, D.B.; Montgomery, B.H.; Wantland, J.L.

1985-01-01

Forced-convection tests have been carried out using parallel simulated Liquid Metal Reactor fuel assemblies in an engineering-scale sodium loop, the Thermal-Hydraulic Out-of-Reactor Safety facility. The tests, performed under single- and two-phase conditions, have shown that for low forced-convection flow there is significant flow augmentation by thermal convection, an important phenomenon under degraded shutdown heat removal conditions in an LMR. The power and flows required for boiling and dryout to occur are much higher than decay heat levels. The experimental evidence supports analytical results that heat removal from an LMR is possible with a degraded shutdown heat removal system

Methods of characterization of salt formations in view of spent fuel final disposal

International Nuclear Information System (INIS)

Diaconu, Daniela; Balan, Valeriu; Mirion, Ilie

2002-01-01

Deep disposal in geological formations of salt, granite and clay seems to be at present the most proper and commonly adopted solution for final disposal of high-level radioactive wastes and spent fuel. Disposing such wastes represents the top-priority issue of the European research community in the field of nuclear power. Although seemingly premature for Romanian power system, the interest for final disposal of spent fuel is justified by the long duration implied by the studies targeting this objective. At the same time these studies represent the Romanian nuclear research contribution in the frame of the efforts of integration within the European research field. Although Romania has not made so far a decision favoring a given geological formation for the final disposal of spent fuel resulting from Cernavoda NPP, the most generally taken into consideration appears the salt formation. The final decision will be made following the evaluation of its performances to spent fuel disposal based on the values of the specific parameters of the geological formation. In order to supply the data required as input parameters in the codes of evaluation of the geological formation performances, the INR Pitesti initiated a package of modern and complex methodologies for such determinations. The studies developed so far followed up the special phenomenon of salt convergence, a phenomenon characteristic for only this kind of rock, as well as the radionuclide migration. These studies allow a better understanding of these processes of upmost importance for disposal's safety. The methods and the experimental installation designed and realized at INR Pitesti aimed at determination of thermal expansion coefficient, thermal conductivity, specific heat, which are all parameters of high specific interest for high level radioactive waste or spent fuel disposal. The paper presents the results of these studies as well as the methodologies, the experimental installations and the findings

Key convention on safe management of spent fuel and radioactive waste to enter into force

International Nuclear Information System (INIS)

2001-01-01

At a ceremony at IAEA Headquarters today, Ireland deposited its instrument of ratification to an important convention on the safe management of spent fuel and radioactive waste, thereby ensuring its entry into force. The Convention will be the first international instrument to address the safety of management and storage of radioactive wastes and spent fuels in countries with and without nuclear programmes

NUFACTS-nuclear fuel cycle activity simulator: reference manual. Final report

International Nuclear Information System (INIS)

Triplett, M.B.; Waddell, J.D.; Breese, T.A.

1978-01-01

The Nuclear Fuel Cycle Activity Simulator (NUFACTS) is a package of FORTRAN subroutines which facilitate the simulation of a diversity of nuclear power growth scenarios. An approach to modeling the nuclear fuel cycle has been developed that is highly adaptive and capable of addressing a variety of problems. Being a simulation model rather than an optimization model, NUFACTS mimics the events and processes that are characteristic of the nuclear fuel cycle. This approach enables the model user to grasp the modeling approach rather quickly. Within this report descriptions of the model and its components are provided with several emphases. First, a discussion of modeling approach and basic assumptions is provided. Next, instructions are provided for generating data, inputting the data properly, and running the code. Finally, detailed descriptions of individual program element are given as an aid to modifying and extending the present capabilities

Unified force and its relation with global warming crave for hydrogen energy and promote fuel cell technology

International Nuclear Information System (INIS)

Krishnan, K.J.; Kalam, A.

2011-01-01

Global warming is presently a tremendous public interest and has become a threat to every individual. Huge quantities of CO/sub 2/ are emitted to the atmosphere by burning of fossil fuels to produce electricity in power plants and burning of gasoline in aeroplanes and vehicles. Enormous amount of greenhouse gasses are sent into the air when garbage is burnt in landfills. Cutting down of trees and other plants which collect CO/sub 2/ a greenhouse gas which is inhaled and which gives back oxygen which is exhaled makes global warming worse. 'Self-Compressive Surrounding Pressure Force' which is also known as Unified Force is also related with global warming which is proportional to increase of H/sub 2/O level in sea and causes floods, storms, droughts and severe impacts to the environment and society. In order to better understand global warming and its relation with Unified Force, this paper discusses the cause and effect system on the amount of greenhouse gases emitted to the atmosphere from the burning of fossil fuels and also the other green house gases like CH/sub 4/, water vapour, NOx etc. and emphasis its importance to focus on crave for Hydrogen Energy and to promote Fuel Cell technology to keep the earth green and safer from the impacts of global warming. The benefit of switching from fossil fuels to Hydrogen Energy and Fuel Cell technology reduces the impact of global warming, elimination of pollution caused by fossil fuels and greenhouse gases, economic dependence and distributed production. (author)

Final summary report of fuel-dynamics tests H2 and E4

International Nuclear Information System (INIS)

Doerner, R.C.; Rothman, A.B.; De Volpi, A.; Dickerman, C.E.; Deitrich, L.W; Stahl, D.; Murphy, W.F.

1976-02-01

Results of two failure experiments using LMFBR-type fuel during simulated unprotected transient overpower accidents are reported and analyzed. In both experiments, a single fresh fuel pin in a Mark-IIA loop was subjected to a temperature-limited, step-reactivity irradiation in the TREAT reactor. Total energy was 490 MJ in Test H2 and 690 MJ in Test E4. Except for their timing, the sequence of events in the failure scenario was the same for both tests. Local coolant boiling began 25-50 msec before failure. Significant upward fuel flow in the center of the pin started as early as 100 msec before cladding failure. Cladding failure was due to melting after contact with molten fuel and occurred at the top of the fuel column. Formation of an outlet flow-channel blockage began about 10 msec after failure and was complete by 50 msec. Inlet blockage began later and was less extensive. No significant amount of fuel sweepout was observed. Fuel remains separated into a small group of 50-1000-μm fragments and a macroscopic group of chunks and clinkers. The final distribution of fuel remains may have resulted from a delayed fuel/steel interaction in the inlet region

Final disposal of spent nuclear fuel - basis for site selection

International Nuclear Information System (INIS)

Anttila, P.

1995-05-01

International organizations, e.g. IAEA, have published several recommendations and guides for the safe disposal of radioactive waste. There are three major groups of issues affecting the site selection process, i.e. geological, environmental and socioeconomic. The first step of the site selection process is an inventory of potential host rock formations. After that, potential study areas are screened to identify sites for detailed investigations, prior to geological conditions and overall suitability for the safe disposal. This kind of stepwise site selection procedure has been used in Finland and in Sweden. A similar approach has been proposed in Canada, too. In accordance with the amendment to the Nuclear Energy Act, that entered into force in the beginning of 1995, Imatran Voima Oy has to make preparations for the final disposal of spent fuel in the Finnish bedrock. Relating to the possible site selection, the following geological factors, as internationally recommended and used in the Nordic countries, should be taken into account: topography, stability of bedrock, brokenness and fracturing of bedrock, size of bedrock block, rock type, predictability and natural resources. The bedrock of the Loviisa NPP site is a part of the Vyborg rapakivi massif. As a whole the rapakivi granite area forms a potential target area, although other rock types or areas cannot be excluded from possible site selection studies. (25 refs., 7 figs.)

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

International Nuclear Information System (INIS)

1987-01-01

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

Effects of fuel Lewis number on localised forced ignition of turbulent homogeneous mixtures: A numerical investigation

Directory of Open Access Journals (Sweden)

Dipal Patel

2016-09-01

Full Text Available The influences of fuel Lewis number LeF (ranging from 0.8 to 1.2 on localised forced ignition and early stages of combustion of stoichiometric and fuel-lean homogeneous mixtures have been analysed using simple chemistry three-dimensional compressible direct numerical simulations for different values of root-mean-square velocity fluctuation and the energy deposition characteristics (i.e. characteristic width and the duration of energy deposition by the ignitor. The localised forced ignition is modelled using a source term in the energy transport equation, which deposits energy in a Gaussian manner from the centre of the ignitor over a stipulated period of time. The fuel Lewis number LeF has been found to have significant influences on the extent of burning of stoichiometric and fuel-lean homogeneous mixtures. It has been shown that the width of ignition energy deposition and the duration over which the ignition energy is deposited have significant influences on the success of ignition and subsequent flame propagation. An increase in the width of ignition energy deposition and the duration of energy deposition for a given amount of ignition energy have been found to have detrimental effects on the ignition event, which may ultimately lead to misfire. For a given value of u' (LeF, the rate of heat transfer from the hot gas kernel increases with increasing LeF (u', which in turn leads to a reduction in the extent of overall burning for both stoichiometric and fuel-lean homogeneous mixtures but the detrimental effects of high values of u' on localised forced ignition are particularly prevalent for fuel-lean mixtures. Detailed physical explanations have been provided for the observed LeF,u' and energy deposition characteristics effects.

Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

Energy Technology Data Exchange (ETDEWEB)

Bragg-Sitton, Shannon Michelle [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, parametric studies, and selection of system codes. The Cladding and Core Materials and Fuel Concepts task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment Task Force is chaired by Shannon Bragg-Sitton (INL), while the Cladding Task Force will be chaired by a representative from France (Marie Moatti, Electricite de France [EdF]) and the Fuels Task Force will be chaired by a representative from Japan (Masaki Kurata, Japan Atomic Energy Agency [JAEA]). This report provides an overview of the Systems Assessment Task Force charter and status of work accomplishment.

Impact Force Applied on the Spent Nuclear Fuel Disposal Canister that Accidentally Drops and Collides onto the Ground

International Nuclear Information System (INIS)

Kwon, Young Joo

2016-01-01

In this paper, a mathematical methodology was theoretically studied to obtain the impact force caused by the collision between rigid bodies. This theoretical methodology was applied to compute the impact force applied on the spent nuclear fuel disposal canister that accidentally drops and collides onto the ground. From this study, the impact force required to ensure a structurally safe canister design was theoretically formulated. The main content of the theoretical study concerns the rigid body kinematics and equation of motion during collision between two rigid bodies. On the basis of this study, a general impact theory to compute the impact force caused by the collision between two bodies was developed. This general impact theory was applied to theoretically formulate the approximate mathematical solution of the impact force that affects the spent nuclear fuel disposal canister that accidentally falls to the ground. Simultaneously, a numerical analysis was performed using the computer code to compute the numerical solution of the impact force, and the numerical result was compared with the approximate mathematical solution

Final Scientific Report - "Improved Fuel Efficiency from Nanocomposite Tire Tread"

Energy Technology Data Exchange (ETDEWEB)

Dr. Andrew Myers

2005-12-30

Rolling resistance, a measure of the energy lost as a tire rotates while moving, is a significant source of power and fuel loss. Recently, low rolling resistant tires have been formulated by adding silica to tire tread. These "Green Tires" (so named from the environmental advantages of lower emissions and improved fuel economy) have seen some commercial success in Europe, where high fuel prices and performance drive tire selection. Unfortunately, the higher costs of the silica and a more complicated manufacturing process have prevented significant commercialization - and the resulting fuel savings - in the U.S. In this project, TDA Research, Inc. (TDA) prepared an inexpensive alternative to silica that leads to tire components with lower rolling resistance. These new tire composite materials were processed with traditional rubber processing equipment. We prepared specially designed nanoparticle additives, based on a high purity, inorganic mineral whose surface can be easily modified for compatibility with tire tread formulations. Our nanocomposites decreased energy losses to hysteresis, the loss of energy from the compression and relaxation of an elastic material, by nearly 20% compared to a blank SBR sample. We also demonstrated better performance than a leading silica product, with easier production of our final rubber nanocomposite.

Spent fuel management strategies in eight countries and applicability to Sweden. Final report

International Nuclear Information System (INIS)

1986-01-01

The spent fuel management activities described in volume 1 are compared in three areas. The first section summarizes the spent fuel management options being followed in each country and compares those options with regard to cost, environmental impact and public acceptability. Next section reviews and compares national policies on nuclear power, spent fuel management and high-level waste disposal and assesses their impact on the development and licensing of nuclear power plants. The third section compares the regulatory requirements affecting spent fuel managementin terms of their overall spirit and characteristics and in terms of the responsibilities of the utilities and the regulatory authorities. Finally, the last section addresses the applicability to Sweden of the findings from these comparisons, focusing on cost efficiency, health and safety, environmental impact, public acceptance and licensing procedures

SRE fuel decladding. Final report

International Nuclear Information System (INIS)

Dennison, W.F.

1977-01-01

This report summarizes the task of decladding the SRE fuel assemblies, and shipment of the fuel to Savannah River for eventual reprocessing. The disposition of 16 unidentified RMDF storage canisters is also covered

Effects of clamping force on the water transport and performance of a PEM (proton electrolyte membrane) fuel cell with relative humidity and current density

International Nuclear Information System (INIS)

Cha, Dowon; Ahn, Jae Hwan; Kim, Hyung Soon; Kim, Yongchan

2015-01-01

The clamping force should be applied to a proton electrolyte membrane (PEM) fuel cell due to its structural characteristics. The clamping force affects the ohmic and mass transport resistances in the PEM fuel cell. In this study, the effects of the clamping force on the water transport and performance characteristics of a PEM fuel cell are experimentally investigated with variations in the relative humidity and current density. The water transport characteristics were analyzed by calculating the net drag coefficient. The ohmic resistance decreased with the increase in the clamping force due to the reduced contact resistance and more even membrane hydration. However, the mass transport resistance increased with the increase in the clamping force due to the gas diffusion layer compression. The net drag coefficient decreased with the increase in the clamping force due to high water back-diffusion. Additionally, the relationship between the total resistance and the net drag coefficient was investigated. - Highlights: • Effects of clamping force on the performance of a PEM fuel cell are investigated. • Water transport characteristics are analyzed using net drag coefficient. • Ohmic resistance decreased with clamping force, but mass transport resistance increased. • Net drag coefficient decreased with the increase in clamping force. • Total resistance was significantly degraded for a net drag coefficient below 0.2.

Direct fuel cell power plants: the final steps to commercialization

Science.gov (United States)

Glenn, Donald R.

Since the last paper presented at the Second Grove Fuel Cell Symposium, the Energy Research Corporation (ERC) has established two commercial subsidiaries, become a publically-held firm, expanded its facilities and has moved the direct fuel cell (DFC) technology and systems significantly closer to commercial readiness. The subsidiaries, the Fuel Cell Engineering Corporation (FCE) and Fuel Cell Manufacturing Corporation (FCMC) are perfecting their respective roles in the company's strategy to commercialize its DFC technology. FCE is the prime contractor for the Santa Clara Demonstration and is establishing the needed marketing, sales, engineering, and servicing functions. FCMC in addition to producing the stacks and stack modules for the Santa Clara demonstration plant is now upgrading its production capability and product yields, and retooling for the final stack scale-up for the commercial unit. ERC has built and operated the tallest and largest capacities-to-date carbonate fuel cell stacks as well as numerous short stacks. While most of these units were tested at ERC's Danbury, Connecticut (USA) R&D Center, others have been evaluated at other domestic and overseas facilities using a variety of fuels. ERC has supplied stacks to Elkraft and MTU for tests with natural gas, and RWE in Germany where coal-derived gas were used. Additional stack test activities have been performed by MELCO and Sanyo in Japan. Information from some of these activities is protected by ERC's license arrangements with these firms. However, permission for limited data releases will be requested to provide the Grove Conference with up-to-date results. Arguably the most dramatic demonstration of carbonate fuel cells in the utility-scale, 2 MW power plant demonstration unit, located in the City of Santa Clara, California. Construction of the unit's balance-of-plant (BOP) has been completed and the installed equipment has been operationally checked. Two of the four DFC stack sub-modules, each

Life cycle assessment of geological repositories for the final disposal of spent fuel in Finland and Sweden

International Nuclear Information System (INIS)

Puhrer, A.; Bauer, C.

2014-01-01

This paper presents a Life Cycle Assessment (LCA) of the geological repositories for the final disposal of spent nuclear fuel in Finland and Sweden. A separate LCA has been performed for the geological spent fuel repository in each country and the results have been compared. A further benchmark comparison has been made with the LCA of the Swiss geological repository for high-level waste and spent fuel. The life cycle inventory (LCI) product system boundaries include the spent fuel repository and encapsulation facility in each country. All materials, processes, consumed utilities and transport associated with the construction, operation and closure of the repositories for spent fuel are included in the LCI. The life cycle impact assessment (LCIA) is performed using two methods: IPCC 2007 Climate Change and ReCiPe. These assessment methods return results pertaining to global warming potential (GWP) as well as a number of environmental impact categories such as human toxicity and natural land transformation. Results indicate that the use of copper for disposal canister fabrication and bentonite for repository backfilling are the causes for most of the environmental impact of the spent fuel repositories in Finland and Sweden. Alternate, less bentonite-intensive backfilling scenarios may mitigate this impact. While the Swiss bentonite consumption is lower and no copper is used for canister fabrication, the Swiss electricity and fuel consumption associated with final disposal of high-level waste and spent fuel is significantly higher than in Finland or Sweden. Approximately 1 g CO 2 -eq is emitted due to the final disposal of spent fuel and HLW per kWh of nuclear generated electricity. This represents some 10% of the emissions due to the entire nuclear energy chain and is practically negligible in the context of GHG emissions of other energy technologies. (authors)

Fuel rods

International Nuclear Information System (INIS)

Hattori, Shinji; Kajiwara, Koichi.

1980-01-01

Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

Evaluation of the Field Performance of Residential Fuel Cells: Final Report

Energy Technology Data Exchange (ETDEWEB)

Torrero, E.; McClelland, R.

2004-05-01

Distributed generation has attracted significant interest from rural electric cooperatives and their customers. Cooperatives have a particular nexus because of inherently low customer density, growth patterns at the end of long lines, and an influx of customers and high-tech industries seeking to diversify out of urban environments. Fuel cells are considered a particularly interesting DG candidate for these cooperatives because of their power quality, efficiency, and environmental benefits. The National Rural Electric Cooperative Association Cooperative Research Network residential fuel cell program demonstrated RFC power plants and assessed related technical and application issues. This final subcontract report is an assessment of the program's results. This 3-year program leveraged Department of Energy (DOE) and National Renewable Energy Laboratory (NREL) funding.

Conditioning of spent fuel for interim and final storage in the pilot conditioning plant (PKA) at Gorleben

International Nuclear Information System (INIS)

Lahr, H.; Willax, H.O.; Spilker, H.

1999-01-01

In 1994, due to the change of the nuclear law in Germany, the concept of direct final disposal for spent fuel was developed as an equivalent alternative to the waste management with reprocessing. Since 1979, tests for the direct final disposal of spent fuel have been conducted in Germany. In 1985, the State and the utilities came to an agreement to develop this concept of waste management to technical maturity. Gesellschaft fuer Nuklear-Service (GNS) was commissioned by the utilities with the following tasks: to develop and test components with regard to conditioning technology, to construct and operate the pilot conditioning plant (PKA), and to develop casks suitable for final disposal. Since 1990, the construction of the PKA has taken place at the Brennelementlager Gorleben site. The PKA has been designed as a multipurpose facility and can thus fulfil various tasks within the framework of the conditioning and management of spent fuel assemblies and radioactive waste. The pilot character of the plant allows for development and testing in the field of spent fuel assembly conditioning. The objectives of the PKA may be summarized as follows: to condition spent fuel assemblies, to reload spent fuel assemblies and waste packages, to condition radioactive waste, and to do maintenance work on transport and storage casks as well as on waste packages. Currently, the buildings of the PKA are constructed and the technical facilities are installed. The plant will be ready for service in the middle of 1999. It is the first plant of its kind in the world. (author)

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

International Nuclear Information System (INIS)

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

1997-09-01

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

Long term fuel price elasticity: effects on mobility tool ownership and residential location choice - Final report

Energy Technology Data Exchange (ETDEWEB)

Erath, A.; Axhausen, K. W.

2010-04-15

This comprehensive final report for the Swiss Federal Office of Energy (SFOE) examines the long-term effects of fuel price elasticity. The study analyses how mobility tool usage and ownership as well as residence location choice are affected by rising fuel costs. Based on econometric models, long-term fuel price elasticity is derived. The authors quote that the demand reactions to higher fuel prices mainly observed are the reduction of mileage and the consideration of smaller-engined and diesel-driven cars. As cars with natural gas powered engines and electric drives were hardly considered in the survey, the results of the natural gas model can, according to the authors, only serve as a trend. No stable model could be estimated for the demand and usage of electric cars. A literature overview is presented and the design of the survey is discussed, whereby socio-demographical variables and the effects of price and residence changes are discussed. Modelling of mobility tool factors and results obtained are looked at. Finally, residence choice factors are modelled and discussed. Several appendices complete the report.

Fuel safety criteria and review by OECD / CSNI task force

International Nuclear Information System (INIS)

Van Doesburg, W.

1999-01-01

Full text of publication follows: with the advent of advanced fuel and core designs, and the implementation of more accurate (best estimate or statistical) design and analysis methods, there is a general feeling that safety margins have been or are being reduced. Historically, fuel safety margins were defined by adding conservatism to the safety limits, which in turn were also fixed in a conservative manner, here, the expression 'conservatism' expresses the fact that bounding or limiting numbers were chosen for model parameters, plant and fuel design data, and fuel operating history values. Unfortunately, as these conservatisms were not quantified (or quantifiable), the amount of safety available or the reduction thereof is difficult to substantiate. For the regulator, it is important to know the margin available with the utilities' request for approval of new fuel or methods; likewise, for the utility and vendor it is important to know what margins exist and what they are based on, to identify in which direction they can make further progress and optimize fuel and fuel cycle cost. Naturally, each party involved will have to decide on how much margin should be in place, to establish operational criteria and ensure that these can actually be met during operation. To assess the margins issue, safety criteria themselves need to be reviewed first. Most - if not all - of the currently existing safety criteria were established during the 60's and early 70's, and verified against experiments with fuel available at that time - mostly at zero exposure. Of course, verification was performed as designs progressed in later years, primarily with the aim to be able to prove that safety criteria were adequate as long as the said conservatisms would be retained, and not with the aim to reestablish limits. The mandate to the OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria (TFFSC) is to assess the adequacy of existing fuel safety criteria, in view of the 'new design' elements (new

75 FR 37733 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program

Science.gov (United States)

2010-06-30

... Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program AGENCY... direct final rule to amend the Renewable Fuel Standard program requirements on May 10, 2010. Because EPA... Fuel Standard program requirements, published on May 10, 2010. We stated in that direct final rule that...

Final Environmental Assessment for the California Space Center at Vandenberg Air Force Base, California

Science.gov (United States)

2010-06-02

rooted , mesophylic plant species that Chapter 3. Affected Environment Final Environmental Assessment - California Space Center, Vandenberg Air...Chapter 3. Affected Environment 3-12 Final Environmental Assessment - California Space Center, Vandenberg Air Force Base the root and debris zone of the...protruding objects, slippery soils or mud, and biological hazards including vegetation (i.e. poison oak and stinging nettle ), animals (i.e. insects

Final disposal of spent fuels and high activity waste: status and trends in the world. Part 2

International Nuclear Information System (INIS)

Herscovich de Pahissa, Marta

2008-01-01

The proper management of spent fuel arising from nuclear power production is a key issue for the sustainable development of nuclear energy. Some countries have adopted reprocessing of spent fuel and part of them has continued to develop and improve closed fuel cycle technologies; some other countries have adopted a direct final disposal. The objective in this article is to provide an update on the latest development in the world related with the geological disposal of spent nuclear fuel and high level wastes. (author) [es

Spent Nuclear Fuel (SNF) Cold Vacuum Drying (CVD) Facility Operations Manual; FINAL

International Nuclear Information System (INIS)

IRWIN, J.J.

1999-01-01

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

Lessons learned in demonstration projects regarding operational safety during final disposal of vitrified waste and spent fuel

International Nuclear Information System (INIS)

Filbert, Wolfgang; Herold, Philipp

2015-01-01

The paper summarizes the lessons learned in demonstration projects regarding operational safety during the final disposal of vitrified waste and spent fuel. The three demonstration projects for the direct disposal of vitrified waste and spent fuel are described. The first two demonstration projects concern the shaft transport of heavy payloads of up to 85 t and the emplacement operations in the mine. The third demonstration project concerns the borehole emplacement operation. Finally, open issues for the next steps up to licensing of the emplacement and disposal systems are summarized.

Environmental Impact Statement. March 2011. Interim storage, encapsulation and final disposal of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

2011-07-01

An Environmental Impact Statement (EIS) shall be prepared and submitted along with applications for permissibility and a licence under the Environmental Code and a licence under the Nuclear Activities Act for new nuclear facilities. This Environmental Impact Statement has been prepared by Svensk Kaernbraenslehantering AB (the Swedish Nuclear Fuel and Waste Management Co, SKB) to be included in the licence applications for continued operation of Clab (central interim storage facility for spent nuclear fuel) in Simpevarp in Oskarshamn Municipality and construction and operation of facilities for encapsulation (integrated with Clab) and final disposal of spent nuclear fuel in Forsmark in Oesthammar Municipality

Environmental Impact Statement. March 2011. Interim storage, encapsulation and final disposal of spent nuclear fuel

International Nuclear Information System (INIS)

2011-01-01

An Environmental Impact Statement (EIS) shall be prepared and submitted along with applications for permissibility and a licence under the Environmental Code and a licence under the Nuclear Activities Act for new nuclear facilities. This Environmental Impact Statement has been prepared by Svensk Kaernbraenslehantering AB (the Swedish Nuclear Fuel and Waste Management Co, SKB) to be included in the licence applications for continued operation of Clab (central interim storage facility for spent nuclear fuel) in Simpevarp in Oskarshamn Municipality and construction and operation of facilities for encapsulation (integrated with Clab) and final disposal of spent nuclear fuel in Forsmark in Oesthammar Municipality

TWTF project criticality task force final review and assessment

International Nuclear Information System (INIS)

McKinley, K.B.; Cannon, J.W.; Wheeler, F.J.; Worle, H.A.

1980-11-01

The Transuranic Waste Treatment Facility (TWTF) is being developed to process transuranic waste, stored and buried at the Idaho National Engineering Laboratory, into a chemically inert, physically stable basalt-like residue acceptable at a federal repository. A task force was assembled by the TWTF Project Division to review and assess all aspects of criticality safety for the TWTF. This document presents the final review, assessments, and recommendations of this task force. The following conclusions were made: Additional criticality studies are needed for the entire envelope of feed compositions and temperature effects. Safe operating k/sub eff/'s need to be determined for process components. Criticality analyses validation experiments may also be required. SRP neutron interrogation should be replaced by DDT neutron interrogation. Accuracy studies need to be performed for the proposed assaying techniques. Time-correlated neutron monitoring needs to be mocked up for process components to prove feasibility and determine accuracy. The criticality control techniques developed for the TWTF conceptual design are in compliance with ERDAM 0530, including the Double Contingency Rule. Detailed procedures and controls need to be developed

Single Fuel Concept for Croatian Army Ground Vehicles

Directory of Open Access Journals (Sweden)

Robert Spudić

2008-05-01

Full Text Available During the process of approaching the European associationsand NATO the Republic of Croatia has accepted the singlefuel concept for all ground vehicles of the Croatian Army.Croatia has also undertaken to insure that all aircraft, motorvehicles and equipment with turbo-engines or with pressurizedfuel injection, for participation in NATO and PfP led operationscan • operate using the kerosene-based aviation fuel(NATO F-34. The paper gives a brief overview and the resultsof the earned out activities in the Armed Forces of the Republicof Croatia, the expected behaviour of the motor vehicle andpossible delays caused by the use of kerosene fuel (NATOF-34 as fuel for motor vehicles. The paper also gives the advantagesand the drawbacks of the single fuel concept. By acquiringnew data in the Croatian Armed Forces and experienceexchange with other nations about the method of using fuelF-34, the development of the technologies of engine manufacturingand its vital parts or by introducing new standards in theproductjon of fuels and additives new knowledge will certainlybe acquired for providing logistics support in the area of operations,and its final implementation will be a big step forward forthe Republic of Croatia towards Europe and NATO.

Handling of spent nuclear fuel and final storage of vitrified high level reprocessing waste

International Nuclear Information System (INIS)

1978-01-01

The report gives a general summary of the Swedish KBS-project on management and disposal of vitrified reprocessed waste. Its final aim is to demostrate that the means of processing and managing power reactor waste in an absolutely safe way, as stipulated in the Swedish so called Conditions Act, already exist. Chapters on Storage facility for spent fuel, Intermidiate storage of reprocessed waste, Geology, Final repository, Transportation, Protection, and Siting. (L.E.)

Carbon-based Fuel Cell. Final report

International Nuclear Information System (INIS)

Steven S. C. Chuang

2005-01-01

The direct use of coal in the solid oxide fuel cell to generate electricity is an innovative concept for power generation. The C-fuel cell (carbon-based fuel cell) could offer significant advantages: (1) minimization of NOx emissions due to its operating temperature range of 700-1000 C, (2) high overall efficiency because of the direct conversion of coal to CO 2 , and (3) the production of a nearly pure CO 2 exhaust stream for the direct CO 2 sequestration. The objective of this project is to determine the technical feasibility of using a highly active anode catalyst in a solid oxide fuel for the direct electrochemical oxidation of coal to produce electricity. Results of this study showed that the electric power generation from Ohio No 5 coal (Lower Kittanning) Seam, Mahoning County, is higher than those of coal gas and pure methane on a solid oxide fuel cell assembly with a promoted metal anode catalyst at 950 C. Further study is needed to test the long term activity, selectivity, and stability of anode catalysts

75 FR 59622 - Supplemental Determination for Renewable Fuels Produced Under the Final RFS2 Program From Canola Oil

Science.gov (United States)

2010-09-28

..., heating oil or jet fuel). In addition, this rule includes a new regulatory provision establishing a... work would be completed through a supplemental final rulemaking process. This supplemental final rule... the final RFS2 rule, EPA will revisit our lifecycle analyses in the future as new information becomes...

Spent fuel management strategies in eight countries and applicability to Sweden. Final report

International Nuclear Information System (INIS)

1986-01-01

International Energy Associates Limited (IEAL) undertook this study on behalf of Sweden's National Board for Spent Nuclear Fuel (SKN) from June to October 1986. The purpose of the project was to compare the programs and regulations for the management of spent fuel from nuclear power plants and disposal of high-level radioactive waste in eight countries: Belgium, Canada, the Federal Republic of Germany, France, Japan, Switzerland, the United Kingdom and the United States. This final report includes revisions requested by SKN upon review of the draft report dated in September 26, 1986. The study is presented in three volumes. Volume I (Section 2.0 of the report) consists of detailed country-specific reports on the policies, regulations and strategies for spent fuel and high-level waste management in each of the eight countries. The information contained in these country-specific reports was used as the basis for comparing the options in each country in terms of cost, environmental impact, and public acceptability, and for comparing the policies and regulatory requirements affecting these activities in each country. These comparisons are provided in Volume II (Section 3.0 of the report). Section 3.0 also includes a discussion of the applicability to Sweden of the strategies and policies in the eight countries studied. Finally, Volume III of the report (Section 4.0) presents the laws, regulations and other documents pertinent to spent fuel and high-level waste management in these countries. Descriptive summaries of the documents are provided in Section 4.0, a comparison guide to the documents themselves (the great majority of them in English) which are provided in 15 volumes of appendices

Research reactor fuel - an update

International Nuclear Information System (INIS)

Finlay, M.R.; Ripley, M.I.

2003-01-01

In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

Gothic simulation of single-channel fuel heatup following a loss of forced flow

Energy Technology Data Exchange (ETDEWEB)

Chen, X-Q; Tahir, A. [NSS, Dept. of Thermal Hydraulics Analysis, Toronto, Ontario (Canada); Parlatan, Y. [Ontario Power Generation, NSATD, Pickering, Ontario (Canada); Kwee, M. [Bruce Power, NSASD, Toronto, Ontario (Canada)

2011-07-01

GOTHIC v7.2 was used to develop a computer model for the simulation of 28- and 37-element fuel heat-up at a loss of forced flow. The model has accounted for the non-uniformity of both axial and radial power distributions along the fuel channel for a typical CANDU reactor. In addition, the model has also accounted for the fuel rods, end-fittings, feeders and headers. Experimental test conditions for both 28- and 37-element bundles at either low or high powers were used for model validation. GOTHIC predictions of the rod and/or pressure-tube temperatures at a variety of test locations were compared with the corresponding experimental measurements. It is found that the numerical results agree well with the experimental measurements for most of the test locations. Results have also shown that the channel venting time is sensitive to the initial temperature distribution in the feeders and headers. An imposed temperature asymmetry at the beginning will cause the channel flow to vent earlier. (author)

Geophysical borehole logging. Final disposal of spent fuel

International Nuclear Information System (INIS)

Rouhiainen, P.

1984-01-01

Teollisuuden Voima Oy (Industrial Power Company Ltd.) will take precautions for final disposal of spent fuel in the Finnish bedrock. The first stage of the site selection studies includes drilling of a deep borehole down to approximately 1000 meters in the year 1984. The report deals with geophysical borehole logging methods, which could be used for the studies. The aim of geophysical borehole logging methods is to descripe specially hydrogeological and structural features. Only the most essential methods are dealt with in this report. Attention is paid to the information produced with the methods, derscription of the methods, interpretation and limitations. The feasibility and possibilities for the aims are evaluated. The evaluations are based mainly on the results from Sweden, England, Canada and USA as well as experiencies gained in Finland

Final Report - MEA and Stack Durability for PEM Fuel Cells

Energy Technology Data Exchange (ETDEWEB)

Yandrasits, Michael A.

2008-02-15

the same. (6) Through the use of statistical lifetime analysis methods, it is possible to develop new MEAs with predicted durability approaching the DOE 2010 targets. (7) A segmented cell was developed that extend the resolution from ~ 40 to 121 segments for a 50cm2 active area single cell which allowed for more precise investigation of the local phenomena in a operating fuel cell. (8) The single cell concept was extended to a fuel size stack to allow the first of its kind monitoring and mapping of an operational fuel cell stack. An internal check used during this project involved evaluating the manufacturability of any new MEA component. If a more durable MEA component was developed in the lab, but could not be scaled-up to ‘high speed, high volume manufacturing’, then that component was not selected for the final MEA-fuel cell system demonstration. It is the intent of the team to commercialize new products developed under this project, but commercialization can not occur if the manufacture of said new components is difficult or if the price is significantly greater than existing products as to make the new components not cost competitive. Thus, the end result of this project is the creation of MEA and fuel cell system technology that is capable of meeting the DOEs 2010 target of 40,000 hours for stationary fuel cell systems (although this lifetime has not been demonstrated in laboratory or field testing yet) at a cost that is economically viable for the developing fuel cell industry. We have demonstrated over 2,000 hours of run time for the MEA and system developed under this project.

Montana fuel tax refunds : draft final report.

Science.gov (United States)

2011-11-01

"The primary source of funding for transportation infrastructure is the taxes that are imposed on motor fuels. One aspect of fuel tax collections is the process that requires consumers to apply for refunds of taxes paid on fuels used for tax-exempt p...

Final report of fuel dynamics Test E7

International Nuclear Information System (INIS)

Doerner, R.C.; Murphy, W.F.; Stanford, G.S.; Froehle, P.H.

1977-04-01

Test data from an in-pile failure experiment of high-power LMFBR-type fuel pins in a simulated $3/s transient-overpower (TOP) accident are reported and analyzed. Major conclusions are that (1) a series of cladding ruptures during the 100-ms period preceding fuel release injected small bursts of fission gas into the flow stream; (2) gas release influenced subsequent cladding melting and fuel release [there were no measurable FCI's (fuel-coolant interactions), and all fuel motion observed by the hodoscope was very slow]; (3) the predominant postfailure fuel motion appears to be radial swelling that left a spongy fuel crust on the holder wall; (4) less than 4 to 6 percent of the fuel moved axially out of the original fuel zone, and most of this froze within a 10-cm region above the original top of the fuel zone to form the outlet blockage. An inlet blockage approximately 1 cm long was formed and consisted of large interconnected void regions. Both blockages began just beyond the ends of the fuel pellets

Alternative Fuel Vehicle Forecasts : Final report.

Science.gov (United States)

2016-04-01

Federal and state fuel taxes account for the largest share of the Texas State Highway Fund at 48 percent and 29 percent, respectively, in Fiscal Year 2015. These taxes are levied on a per-gallon basis, meaning that as vehicles get more fuel efficient...

Urinary biomarkers of occupational jet fuel exposure among Air Force personnel.

Science.gov (United States)

Smith, Kristen W; Proctor, Susan P; Ozonoff, A L; McClean, Michael D

2012-01-01

There is a potential for widespread occupational exposure to jet fuel among military and civilian personnel. Urinary metabolites of naphthalene have been suggested for use as short-term biomarkers of exposure to jet fuel (jet propulsion fuel 8 (JP8)). In this study, urinary biomarkers of JP8 were evaluated among US Air Force personnel. Personnel (n=24) were divided a priori into high, moderate, and low exposure groups. Pre- and post-shift urine samples were collected from each worker over three workdays and analyzed for metabolites of naphthalene (1- and 2-naphthol). Questionnaires and breathing-zone naphthalene samples were collected from each worker during the same workdays. Linear mixed-effects models were used to evaluate the exposure data. Post-shift levels of 1- and 2-naphthol varied significantly by a priori exposure group (levels in high group>moderate group>low group), and breathing-zone naphthalene was a significant predictor of post-shift levels of 1- and 2-naphthol, indicating that for every unit increase in breathing-zone naphthalene, there was an increase in naphthol levels. These results indicate that post-shift levels of urinary 1- and 2-naphthol reflect JP8 exposure during the work-shift and may be useful surrogates of JP8 exposure. Among the high exposed workers, significant job-related predictors of post-shift levels of 1- and 2-naphthol included entering the fuel tank, repairing leaks, direct skin contact with JP8, and not wearing gloves during the work-shift. The job-related predictors of 1- and 2-naphthol emphasize the importance of reducing inhalation and dermal exposure through the use of personal protective equipment while working in an environment with JP8.

Interagency task force on the health effects of ionizing radiation. final report

International Nuclear Information System (INIS)

1979-06-01

This is the final report of the task force and incorporates the findings and recommendations of six smaller work groups, each with a more specific focus; i.e., science, privacy, care and benefits, exposure reduction, public information, and institutional arrangements. A research agenda that could provide some answers to questions about the effects of low-level radiation is proposed, along with recommendations to facilitate research. A public information program is outlined. Recommendations are advanced to improve systems that deliver care and benefits to those who may have been injured by exposure to radiation, and proposals for steps that might reduce unnecessary radiation exposure in the future are identified. The task force also recommends measures to institutionalize the interagency cooperation that characterized the task force. Three tables and one figure show the collective estimates of the U.S. general population, Federal research financing, cancer linked to radiation in particular populations, and a general dose-response model

Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report

International Nuclear Information System (INIS)

Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier; Klennert, Lindsay A.; Nolte, Oliver; Molecke, Martin Alan; Autrusson, Bruno A.; Koch, Wolfgang; Pretzsch, Gunter Guido; Brucher, Wenzel; Steyskal, Michele D.

2008-01-01

This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO 2 , CeO 2 , plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively supported and

Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report.

Energy Technology Data Exchange (ETDEWEB)

Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Klennert, Lindsay A.; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

2008-03-01

This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO{sub 2}, CeO{sub 2}, plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively

Study and modeling of fluctuating fluid forces exerted on fuel rods in pressurized water reactors

International Nuclear Information System (INIS)

Bhattacharjee, Saptarshi

2016-01-01

Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in the fuel rods. Due to friction, wear occurs at the contact locations between the spacer grid and the fuel rod. This could compromise the first safety barrier of the nuclear reactor by damaging the fuel rod cladding. In order to ensure the integrity of the cladding, it is necessary to know the random fluctuating forces acting on the rods. However, the spectra for these fluid forces are not well known. The goal of this PhD thesis was to use simple geometrical elements to check the reproducibility of realistic pressurized water reactor spacer grids. As a first step, large eddy simulations were performed on a concentric annular pipe for different mesh refinements using the CFD code Trio CFD (previously Trio U) developed by CEA. A mesh sensitivity study was performed to obtain an acceptable mesh for reproducing standard literature results. This information on mesh resolution was used when carrying out simulations using various geometric obstacles inside the pipe, namely, mixing vanes, circular spacer grid and a combination of square spacer grid with mixing vanes. The last of the three configurations is the closest to a realistic PWR fuel assembly. Structured mesh was generated for the annular pipe case and circular grid case. An innovative hybrid mesh was used for the two remaining cases of the mixing vanes and the square grid: keeping unstructured mesh around the obstacles and structured mesh in the rest of the domain. The inner wall of the domain was representative of the fuel rod cladding. Both hydraulic and wall pressure characteristics were analyzed for each case. The results for the square grid case were found to be an approximate combination of the mixing vane case and circular grid case. Simulation results were compared with experiments performed at CEA Cadarache. Some preliminary comparisons were also made with classical semi-empirical models. (author) [fr

Development of an international safeguards approach to the final disposal of spent fuel in geological repositories

International Nuclear Information System (INIS)

Murphey, W.M.; Moran, B.W.; Fattah, A.

1996-01-01

The International Atomic Energy Agency (IAEA) is currently pursuing development of an international safeguards approach for the final disposal of spent fuel in geological repositories through consultants meetings and through the Program for Development of Safeguards for Final Disposal of Spent Fuel in Geological Repositories (SAGOR). The consultants meetings provide policy guidance to IAEA; SAGOR recommends effective approaches that can be efficiently implemented by IAEA. The SAGOR program, which is a collaboration of eight Member State Support Programs (MSSPs), was initiated in July 1994 and has identified 15 activities in each of three areas (i.e. conditioning facilities, active repositories, and closed repositories) that must be performed to ensure an efficient, yet effective safeguards approach. Two consultants meetings have been held: the first in May 1991 and the last in November 1995. For nuclear materials emplaced in a geological repository, the safeguards objectives were defined to be (1) to detect the diversion of spent fuel, whether concealed or unconcealed, from the repository and (2) to detect undeclared activities of safeguards concern (e.g., tunneling, underground reprocessing, or substitution in containers)

Government--utility interaction on spent fuel disposition

International Nuclear Information System (INIS)

Mills, L.E.

1978-01-01

The question of the needs of the electrical power industry for spent fuel storage in light of the moratorium on fuel reprocessing is addressed. The author feels that since the Federal government has assumed the responsibility for spent fuel storage, it is imperative that a firm plan, program, legislation, and funding be forthcoming immediately. Designation of an existing government site with existing nuclear activities in order to expedite the establishment of a storage facility is recommended. It is felt that the timing for such a site should be ''at the earliest possible date.'' Without storage facilities being provided by the government, utilities will be forced to build storage facilities at the reactor sites. This course of action is not considered cost effective but certainly preferable to shutting down the reactors. It is emphasized that spent fuel storage must be an interim solution and certainly not a final solution to the fuel reprocessing and waste disposal aspects of nuclear technology

Treatment and final storage of radioactive wastes from the nuclear fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Krause, H [Kernforschungszentrum Karlsruhe (Germany, F.R.)

1977-05-01

Types, amounts and activity concentrations of the radioactive wastes arising from the different sections of the fuel cycle are described as well as the methods of their treatment and final disposal. By conversion to glass products, highly active fission product solutions can be transferred into a form well suited for final disposal. Low and medium level waste waters are purified so far that safe discharge or reuse is possible. The concentrates thus produced are incorporated into concrete or bitumen. Baling lends itself for treatment of non-combustible solid wastes. Combustible wastes can be incinerated, the residues are incorporated into concrete. For final storage of the conditioned wastes, salt formations in the deep underground are chosen in the Federal Republic of Germany. They offer a series of favourable preconditions for this purpose and guarantee the isolation of the radionuclides from the biocycle over secular periods of time.

Joint Maneuver Test Range on Eglin Air Force Base, Florida Final Environmental Assessment

Science.gov (United States)

2009-12-14

ir Force B ase, Florida Page 3-2 Final E nvironm ental A ssessm ent A ffected E nvironm ent W ater R esources Figure 3-1. Physical and...nvironm ent W ater R esources Figure 3-2. Physical and Biological Resources Within Range B-9 Existing Components 14. Hills!Qoss Slope Legend --Creek

Characterisation and final disposal behaviour of theoria-based fuel kernels in aqueous phases

International Nuclear Information System (INIS)

Titov, M.

2005-08-01

Two high-temperature reactors (AVR and THTR) operated in Germany have produced about 1 million spent fuel elements. The nuclear fuel in these reactors consists mainly of thorium-uranium mixed oxides, but also pure uranium dioxide and carbide fuels were tested. One of the possible solutions of utilising spent HTR fuel is the direct disposal in deep geological formations. Under such circumstances, the properties of fuel kernels, and especially their leaching behaviour in aqueous phases, have to be investigated for safety assessments of the final repository. In the present work, unirradiated ThO 2 , (Th 0.906 ,U 0.094 )O 2 , (Th 0.834 ,U 0.166 )O 2 and UO 2 fuel kernels were investigated. The composition, crystal structure and surface of the kernels were investigated by traditional methods. Furthermore, a new method was developed for testing the mechanical properties of ceramic kernels. The method was successfully used for the examination of mechanical properties of oxide kernels and for monitoring their evolution during contact with aqueous phases. The leaching behaviour of thoria-based oxide kernels and powders was investigated in repository-relevant salt solutions, as well as in artificial leachates. The influence of different experimental parameters on the kernel leaching stability was investigated. It was shown that thoria-based fuel kernels possess high chemical stability and are indifferent to presence of oxidative and radiolytic species in solution. The dissolution rate of thoria-based materials is typically several orders of magnitude lower than of conventional UO 2 fuel kernels. The life time of a single intact (Th,U)O 2 kernel under aggressive conditions of salt repository was estimated as about hundred thousand years. The importance of grain boundary quality on the leaching stability was demonstrated. Numerical Monte Carlo simulations were performed in order to explain the results of leaching experiments. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-19

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

Fan Fuel Casting Final Report

Energy Technology Data Exchange (ETDEWEB)

Imhoff, Seth D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-09-28

LANL was approached to provide material and design guidance for a fan-shaped fuel element. A total of at least three castings were planned. The first casting is a simple billet mold to be made from high carbon DU-10Mo charge material. The second and third castings are for optimization of the actual fuel plate mold. The experimental scope for optimization is only broad enough for a second iteration of the mold design. It is important to note that partway through FY17, this project was cancelled by the sponsor. This report is being written in order to capture the knowledge gained should this project resume at a later date.

OECD-IAEA Paks Fuel Project. Final Report

International Nuclear Information System (INIS)

2010-05-01

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

Spent-Fuel Test - Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Executive summary of final results

International Nuclear Information System (INIS)

Patrick, W.C.

1986-01-01

This summary volume outlines results that are covered in more detail in the final report of the Spent-Fuel Test - Climate project. The project was conducted between 1978 and 1983 in the granitic Climax stock at the Nevada Test Site. Results indicate that spent fuel can be safely stored for periods of years in this host medium and that nuclear waste so emplaced can be safely retrieved. We also evaluated the effects of heat and radiation (alone and in combination) on emplacement canisters and the surrounding rock mass. Storage of the spent-fuel affected the surrounding rock mass in measurable ways, but did not threaten the stability or safety of the facility at any time

High temperature reactor: Driving force to convert CO2 to fuel - HTR2008-58132

International Nuclear Information System (INIS)

McCormick, J. L.

2008-01-01

The rapidly increasing cost of petroleum products and uncertainty of long-term supply have prompted the U.S. military to aggressively pursue production of alternative fuels (synfuels) such as coal-to-liquids (CTL). U.S. Air Force is particularly active in this effort while the entire military is involved in simultaneously developing fuel specifications for alternative fuels that enable a single fuel for the entire battle space; all ground vehicles, aircraft and fuel cells. By limiting its focus on coal, tar sands and oil shale resources, the military risks violating federal law which requires the use of synfuels that have life cycle greenhouse gas emissions less than or equal to emissions from conventional petroleum fuels. A climate-friendly option would use a high temperature nuclear reactor to split water. The hydrogen (H 2 ) would be used in the reverse water gas shift (RWGS) to react with carbon dioxide (CO 2 ) to produce carbon monoxide (CO) and water. The oxygen (O 2 ) would be fed into a supercritical (SC) coal furnace. The flue gas CO 2 emissions would be stripped of impurities before reacting with H 2 in a RWGS process. Resultant carbon monoxide (CO) is fed, with additional H2, (extra H 2 needed to adjust the stoichiometry: 2 moles H 2 to one mole CO) into a conventional Fischer-Tropsch synthesis (FTS) to produce a heavy wax which is cracked and isomerized and refined to Jet Propulsion 8 (JP-8) and Jet Propulsion 5 (JP-5) fuels. The entire process offers valuable carbon-offsets and multiple products that contribute to lower syn-fuel costs and to comply with the federal limitation imposed on syn-fuel purchases. While the entire process is not commercially available, component parts are being researched; their physical and chemical properties understood and some are state-of-the-art technologies. An international consortium should complete physical, chemical and economic flow sheets to determine the feasibility of this concept that, if pursued, has broad

Equipment to take up the axial forces occuring on fuel elements in the operation of a nuclear reactor

International Nuclear Information System (INIS)

Sankovich, M.

1977-01-01

A constructive solution for the spring support of fuel elements between a lower and upper grid is given which prevents vibrations from the influence of axial forces due to thermal expansion and/or coolant flow with the least possible resistance to the coolant flow. As plate or screw springs usually allow certain vibrations or even encourage these, and to compensate for the flow resistance thus caused nominal increase of the total cooling power is necessary, i.e. the total efficiency of the plants is lowered; therefore a combined torsion and spring was constructed. 4 each of these springs surround in an approximately horizontal plane the head of a fuel element containing the usual number of fuel rods. Each spring forms a U seen from above and surrounds the fuel element head on one side completely and about half the length of the two adjacent sides. The three sides of the spring are inbedded in the openings of the fuel element end pieces so as not to cause any nominal resistance for the coolant flow rising from the fuel elements. (HP) [de

Sensitivity of nuclear fuel-cycle cost to uncertainties in nuclear data. Final report

International Nuclear Information System (INIS)

Becker, M.; Harris, D.R.

1980-11-01

An improved capability for assessing the economic implications of uncertainties in nuclear data and methods on the power reactor fuel cycle was developed. This capability is applied to the sensitivity analysis of fuel-cycle cost with respect to changes in nuclear data and related computational methods. Broad group sensitivities for both a typical BWR and a PWR are determined under the assumption of a throwaway fuel cycle as well as for a scenario under which reprocessing is allowed. Particularly large dollar implications are found for the thermal and resonance cross sections of fissile and fertile materials. Sensitivities for the throwaway case are found to be significantly larger than for the recycle case. Constrained sensitivities obtained for cases in which information from critical experiments or other benchmarks is used in the design calculation to adjust a parameter such as anti ν are compared with unconstrained sensitivities. Sensitivities of various alternate fuel cycles were examined. These included the extended-burnup (18-month) LWR cycle, the mixed-oxide (plutonium) cycle, uranium-thorium and denatured uranium-thorium cycles, as well as CANDU-type reactor cycles. The importance of the thermal capture and fission cross sections of 239 Pu is shown to be very large in all cases. Detailed, energy dependent sensitivity profiles are provided for the thermal range (below 1.855 eV). Finally, sensitivity coefficients are combined with data uncertainties to determine the impact of such uncertainties on fuel-cycle cost parameters

Inventory of concepts for mixed diesel fuels containing renewable components. Final report

Energy Technology Data Exchange (ETDEWEB)

Kronberg, B. [Inst. for Surface Chemistry, Stockholm (Sweden); Berg, R. [Befri Konsult, Solna (Sweden); Berg, J. [Svenska Lantmaennen/Agro Oil, Stockholm (Sweden)

2000-08-01

The present report has involved the assembly of two sub-reports, which have been put together to form this final report. Both of the sub-reports deal with the incorporation of ethanol in diesel fuels. The potential advantages are the decreased net emissions of carbon dioxide, due to the renewable nature of ethanol (if obtained from renewable raw materials), and the decrease of NO{sub x} emissions, due to the decreased combustion temperature. The first sub-report is a compilation of scientific articles and patents/patent applications regarding the possibility to blend ethanol into diesel to form a stable solution in the form of a so called microemulsion, with the aid of surfactants and/or co-solvents. The second sub-report briefly describes the test work, both in the laboratory and in field tests, that is being done in various countries, regarding the blending of ethanol into diesel fuel.

Characterization of inhalation exposure to jet fuel among U.S. Air Force personnel.

Science.gov (United States)

Merchant-Borna, Kian; Rodrigues, Ema G; Smith, Kristen W; Proctor, Susan P; McClean, Michael D

2012-07-01

Jet propulsion fuel-8 (JP-8) is the primary jet fuel used by the US military, collectively consuming ~2.5 billion gallons annually. Previous reports suggest that JP-8 is potentially toxic to the immune, respiratory, and nervous systems. The objectives of this study were to evaluate inhalation exposure to JP-8 constituents among active duty United States Air Force (USAF) personnel while performing job-related tasks, identify significant predictors of inhalation exposure to JP-8, and evaluate the extent to which surrogate exposure classifications were predictive of measured JP-8 exposures. Seventy-three full-time USAF personnel from three different air force bases were monitored during four consecutive workdays where personal air samples were collected and analyzed for benzene, ethylbenzene, toluene, xylenes, total hydrocarbons (THC), and naphthalene. The participants were categorized a priori into high- and low-exposure groups, based on their exposure to JP-8 during their typical workday. Additional JP-8 exposure categories included job title groups and self-reported exposure to JP-8. Linear mixed-effects models were used to evaluate predictors of personal air concentrations. The concentrations of THC in air were significantly different between a priori exposure groups (2.6 mg m(-3) in high group versus 0.5 mg m(-3) in low, P fuel distribution/maintenance, though self-reported exposure to JP-8 was an even stronger predictor of measured exposure in models that explained 72% (THC) and 67% (naphthalene) of between-worker variability. In fact, both self-report JP-8 exposure and a priori exposure groups explained more between-worker variability than job categories. Personal exposure to JP-8 varied by job and was positively associated with the relative humidity. However, self-reported exposure to JP-8 was an even stronger predictor of measured exposure than job title categories, suggesting that self-reported JP-8 exposure is a valid surrogate metric of exposure when

Final environmental statement: US Spent Fuel Policy. Storage of foreign spent power reactor fuel

International Nuclear Information System (INIS)

1980-05-01

In October 1977, the Department of Energy (DOE) announced a Spent Fuel Storage Policy for nuclear power reactors. Under this policy, as approved by the President, US utilities will be given the opportunity to deliver spent fuel to US Government custody in exchange for payment of a fee. The US Government will also be prepared to accept a limited amount of spent fuel from foreign sources when such action would contribute to meeting nonproliferation goals. Under the new policy, spent fuel transferred to the US Government will be delivered - at user expense - to a US Government-approved site. Foreign spent fuel would be stored in Interim Spent Fuel Storage (ISFS) facilities with domestic fuel. This volume of the environmental impact statement includes effects associated with implementing or not implementing the Spent Fuel Storage Policy for the foreign fuels. The analyses show that there are no substantial radiological health impacts whether the policy is implemented or not. In no case considered does the population dose commitment exceed 0.000006% of the world population dose commitment from natural radiation sources over the period analyzed. Full implementation of the US offer to accept a limited amount of foreign spent fuel for storage provides the greatest benefits for US nonproliferation policy. Acceptance of lesser quantities of foreign spent fuel in the US or less US support of foreign spent fuel storage abroad provides some nonproliferation benefits, but at a significantly lower level than full implementation of the offer. Not implementing the policy in regard to foreign spent fuel will be least productive in the context of US nonproliferation objectives. The remainder of the summary provides a brief description of the options that are evaluated, the facilities involved in these options, and the environmental impacts, including nonproliferation considerations, associated with each option

Handling of spent nuclear fuel and final storage of vitrified high level reprocessing waste

International Nuclear Information System (INIS)

1978-01-01

A summary of the planning of transportation and plant design in the Swedish KBS project on management and disposal reprocessed radioactive waste. It describes a transportation system, a central storage facility for used fuel elements, a plant for intermediate storage and encapsulation and a final repository for the vitrified waste. Accounts are given for the reprocessing and vitrification. The safety of the entire system is discussed

Repeated soil application of organic waste amendments reduces draught force and fuel consumption for soil tillage

DEFF Research Database (Denmark)

Peltrea, Clément; Nyord, Tavs; Bruun, Sander

2015-01-01

Abstract Soil application of organic waste products (OWP) can maintain or increase soil organic carbon (SOC) content, which in turn could lead to increased porosity and potentially to reduced energy use for soil tillage. Only a few studies have addressed the effect of SOC content on draught force...... for soil tillage, and this still needs to be addressed for fields that receive diverse types of organic waste of urban, agricultural and agro-industrial origin. The objective of this study was to determine the effect of changes in SOC induced by repeated soil application of OWP on draught force for soil...... tillage and tractor fuel consumption. Draught force was measured for tillage with conventional spring tillage tines, as well as bulk density, soil texture and SOC content in the CRUCIAL field experiment, Denmark in which diverse types of OWP had been applied annually for 11 years. The OWP included...

77 FR 72746 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard and Diesel...

Science.gov (United States)

2012-12-06

... Fuels and Fuel Additives: Modifications to Renewable Fuel Standard and Diesel Sulfur Programs AGENCY... Fuel Standard (``RFS'') program under section 211(o) of the Clean Air Act. The direct final rule also... marine diesel fuel produced by transmix processors, and the fuel marker requirements for 500 ppm sulfur...

Novel catalysts for hydrogen fuel cell applications:Final report (FY03-FY05).

Energy Technology Data Exchange (ETDEWEB)

Thornberg, Steven Michael; Coker, Eric Nicholas; Jarek, Russell L.; Steen, William Arthur

2005-12-01

qualitatively as well as the ETEK material for the ORR, a non-trivial achievement. A fuel cell test showed that Pt/C outperformed the ETEK material by an average of 50% for a 300 hour test. Increasing surface area decreases the amount of Pt needed in a fuel cell, which translates into cost savings. Furthermore, the increased performance realized in the fuel cell test might ultimately mean less Pt is needed in a fuel cell; this again translates into cost savings. Finally, enhanced long-term stability is a key driver within the fuel cell community as improvements in this area must be realized before fuel cells find their way into the marketplace; these Pt/C materials hold great promise of enhanced stability over time. An external laser desorption ion source was successfully installed on the existing Fourier transform ion-cyclotron resonance (FT-ICR) mass spectrometer. However, operation of this laser ablation source has only generated metal atom ions, no clusters have been found to date. It is believed that this is due to the design of the pulsed-nozzle/laser vaporization chamber. The final experimental configuration and design of the two source housings are described.

Hawaii alternative fuels utilization program. Phase 3, final report

Energy Technology Data Exchange (ETDEWEB)

Kinoshita, C.M.; Staackmann, M.

1996-08-01

The Hawaii Alternative Fuels Utilization Program originated as a five-year grant awarded by the US Department of Energy (USDOE) to the Hawaii Natural Energy Institute (HNEI) of the University of Hawaii at Manoa. The overall program included research and demonstration efforts aimed at encouraging and sustaining the use of alternative (i.e., substitutes for gasoline and diesel) ground transportation fuels in Hawaii. Originally, research aimed at overcoming technical impediments to the widespread adoption of alternative fuels was an important facet of this program. Demonstration activities centered on the use of methanol-based fuels in alternative fuel vehicles (AFVs). In the present phase, operations were expanded to include flexible fuel vehicles (FFVs) which can operate on M85 or regular unleaded gasoline or any combination of these two fuels. Additional demonstration work was accomplished in attempting to involve other elements of Hawaii in the promotion and use of alcohol fuels for ground transportation in Hawaii.

Logistic Fuel Processor Development

National Research Council Canada - National Science Library

Salavani, Reza

2004-01-01

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

Instrumentation of cars for fuel economy. Final report

Energy Technology Data Exchange (ETDEWEB)

Morris, J E

1982-04-01

The development of an electronic system to control the air-fuel ratio (A/F) and ignition timing of an internal combustion engine to optimize fuel economy is described. Dynamometer and drive cycle testing of the system was performed. The results showed that a significant improvement in fuel economy can be achieved by a control system of the type developed. It is clear, however, that considerably more work needs to be done. One area mentioned is the need for more systematic fuel economy testing against speed and load as control parameters are varied for optimization, and a more economic air bypass system must be developed. (LCL)

Greenhouse impact due to the use of combustible fuels: life cycle viewpoint and relative radiative forcing commitment.

Science.gov (United States)

Kirkinen, Johanna; Palosuo, Taru; Holmgren, Kristina; Savolainen, Ilkka

2008-09-01

Extensive information on the greenhouse impacts of various human actions is important in developing effective climate change mitigation strategies. The greenhouse impacts of combustible fuels consist not only of combustion emissions but also of emissions from the fuel production chain and possible effects on the ecosystem carbon storages. It is important to be able to assess the combined, total effect of these different emissions and to express the results in a comprehensive way. In this study, a new concept called relative radiative forcing commitment (RRFC) is presented and applied to depict the greenhouse impact of some combustible fuels currently used in Finland. RRFC is a ratio that accounts for the energy absorbed in the Earth system due to changes in greenhouse gas concentrations (production and combustion of fuel) compared to the energy released in the combustion of fuel. RRFC can also be expressed as a function of time in order to give a dynamic cumulative picture on the caused effect. Varying time horizons can be studied separately, as is the case when studying the effects of different climate policies on varying time scales. The RRFC for coal for 100 years is about 170, which means that in 100 years 170 times more energy is absorbed in the atmosphere due to the emissions of coal combustion activity than is released in combustion itself. RRFC values of the other studied fuel production chains varied from about 30 (forest residues fuel) to 190 (peat fuel) for the 100-year study period. The length of the studied time horizon had an impact on the RRFC values and, to some extent, on the relative positions of various fuels.

Solid oxide fuel cells towards real life applications. Final report

Energy Technology Data Exchange (ETDEWEB)

2010-07-01

Solid Oxide Fuel Cells offer a clean and efficient way of producing electricity and heat from a wide selection of fuels. The project addressed three major challenges to be overcome by the technology to make commercialisation possible. (1) At the cell level, increased efficiency combined with production cost reduction has been achieved through an optimization of the manufacturing processes, b) by using alternative raw materials with a lower purchase price and c) by introducing a new generation of fuel cells with reduced loss and higher efficiency. (2) At the stack level, production cost reduction is reduced and manufacturing capacity is increased through an optimization of the stack production. (3) At the system level, development of integrated hotbox concepts for the market segments distributed generation (DG), micro combined heat and power (mCHP), and auxiliary power units (APU) have been developed. In the mCHP segment, two concepts have been developed and validated with regards to market requirements and scalability. In the APU-segment, different types of reformers have been tested and it has been proven that diesel can be reformed through appropriate reformers. Finally, operation experience and feedback has been gained by deployment of stacks in the test facility at the H.C. OErsted Power Plant (HCV). This demonstration has been carried out in collaboration between TOFC and DONG Energy Power A/S (DONG), who has participated as a subcontractor to TOFC. The demonstration has given valuable knowledge and experience with design, start-up and operation of small power units connected to the grid and future development within especially the mCHP segment will benefit from this. In this report, the project results are described for each of the work packages in the project. (Author)

Metallographic examinations of the wear-marks on fuel pins of the KNK II/2 fuel assembly NY-308

International Nuclear Information System (INIS)

Patzer, G.

1987-12-01

On the fuel pins and pin spacers of the fuel assembly NY-308 of the second core of KNK II pronounced wear marks had been found in the area of the contact points. In order to determine the exact form of the marks, metallographic investigations were performed on two test pieces of fuel pins in the Hot Cells of the KfK Karlsruhe. It was found that the wear marks did show the already observed stratified structure. Next to the unchanged cladding area there is a peripheral zone with modified grain structure, followed by a layer of moved material and finally there is a flake-like zone of accumulated cladding material at the lower end of the wear marks. Longitudinal cuts do not show grain deformations, which could indicate axial friction forces between pin and spacer. The wear marks are rapidly dropping to their maximum depth at the ends and the depth shows a relatively uniform pattern between both. The findings are confirming the picture, that a stirring movement of the fuel pins took place, which caused adhesive wear [de

Fuel rod-to-support contact pressure and stress measurement for CHASNUPP-1(PWR) fuel

International Nuclear Information System (INIS)

Waseem; Elahi, N.; Siddiqui, A.; Murtaza, G.

2011-01-01

Research highlights: → A detailed finite element model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. → The spring hold-down force is calculated using the contact pressure obtained from the FE model. → Experiment has also been conducted in the same environment for the measurement of this force. → The spring hold-down force values obtained from both studies confirm the validation of this analysis. → The stress obtained through this analysis is less than the yield strength of spacer grid material, thus fulfils the structural integrity criteria of grid. - Abstract: This analysis has been made in an attempt to measure the contact pressure of the PWR fuel assembly spacer grid spring and to verify its structural integrity at room temperature in air. A detailed finite element (FE) model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The FE model of a fuel rod-to-support system is produced with shell and contact elements. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies are compared, which show good agreement, and in turn confirm the validation of this analysis. The Stress obtained through this analysis is less than the yield strength of spacer grid material (Inconel-718), thus fulfils the structural integrity criteria of grid.

Fuel rod-to-support contact pressure and stress measurement for CHASNUPP-1(PWR) fuel

Energy Technology Data Exchange (ETDEWEB)

Waseem, E-mail: wazim_me@hotmail.co [Directorate General Nuclear Power Fuel, Pakistan Atomic Energy Commission, P.O. Box No. 1847, Islamabad 44000 (Pakistan); Elahi, N.; Siddiqui, A.; Murtaza, G. [Directorate General Nuclear Power Fuel, Pakistan Atomic Energy Commission, P.O. Box No. 1847, Islamabad 44000 (Pakistan)

2011-01-15

Research highlights: A detailed finite element model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies confirm the validation of this analysis. The stress obtained through this analysis is less than the yield strength of spacer grid material, thus fulfils the structural integrity criteria of grid. - Abstract: This analysis has been made in an attempt to measure the contact pressure of the PWR fuel assembly spacer grid spring and to verify its structural integrity at room temperature in air. A detailed finite element (FE) model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The FE model of a fuel rod-to-support system is produced with shell and contact elements. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies are compared, which show good agreement, and in turn confirm the validation of this analysis. The Stress obtained through this analysis is less than the yield strength of spacer grid material (Inconel-718), thus fulfils the structural integrity criteria of grid.

ENVIRONMENTAL SCENARIOS FOR MANDATORY BIO-FUEL BLENDING TARGETS: AN APPLICATION OF INTUITIVE LOGICS

Directory of Open Access Journals (Sweden)

Marco Antonio Conejero

2010-09-01

Full Text Available Scenarios depicting targets concerning mandatory bio-fuel blending are critical to the strategic planning of food and bio-energy production chains and their design is the purpose of this paper. Each scenario tells a story about how various elements might interact under given conditions. The method herein utilized is primarily based on Schoemaker´s (1995 and Schwartz´s (1991 earlier proposals. A six step framework is followed: i identify the focal issue; ii summarize current mandatory blending targets; iii identify the driving forces as of a macro-environmental analysis; iv validate driving forces with specialists; v rank such key forces by importance before uncertainties, building a correlation matrix; vi design the scenarios. Finally, three alternative scenarios, relative to the adoption on behalf of countries, by the year 2020, of mandatory bio-fuel blending targets, are proposed which might guide these countries’ decision makers when planning production systems.

Fuel dynamics loss-of-flow test L3. Final report

International Nuclear Information System (INIS)

Fischer, A.K.; Lo, R.K.; Barts, E.W.

1976-06-01

The behavior of FTR-type, mixed-oxide, preirradiated, ''intermediate-power-structure'' fuel during a simulation of an FTR loss-of-flow accident was studied in the Mark-IIA integral TREAT loop. Analysis of the data reported here leads to a postulated scenario (sequence and timing) of events in the test. This scenario is presented, together with the calculated timing of events obtained by use of the SAS code. The initial fuel motion, starting during the preheat phase, consisted of coherent motion of the entire intact fuel bundle toward the pump. Incoherence developed as temperature rose. The fuel motion was mostly upward, and the greatest was in the top third of the fuel column. Fuel fragments formed against the pump side of the fluted tube near the original fuel midplane. A penetration of fluted tube occurred. A sudden voiding of the central region of the fuel column occurred at 29.75 s and was largely completed within 150 ms. The lower steel blockage of the fuel elements occurred in the vicinity of the lower insulator pellets. The upper steel blockage just above the tops of the original fuel pins appeared to have channels through it. Cladding and spacer wires melted away in the fuel section. Fuel pellets were only evident at and above the top and at the bottom of the original fuel column, where a large mass of melted fuel was present. Over the length of the fuel column, most of the fluted tube had melted away

Final disposal of high levels waste and spent nuclear fuel

International Nuclear Information System (INIS)

Gelin, R.

1984-05-01

Foreign and international activities on the final disposal of high-level waste and spent nuclear fuel have been reviewed. A considerable research effort is devoted to development of acceptable disposal options. The different technical concepts presently under study are described in the report. Numerous studies have been made in many countries of the potential risks to future generations from radioactive wastes in underground disposal repositories. In the report the safety assessment studies and existing performance criteria for geological disposal are briefly discussed. The studies that are being made in Canada, the United States, France and Switzerland are the most interesting for Sweden as these countries also are considering disposal into crystalline rocks. The overall time-tables in different countries for realisation of the final disposal are rather similar. Normally actual large-scale disposal operations for high-level wastes are not foreseen until after year 2000. In the United States the Congress recently passed the important Nuclear Waste Policy Act. It gives a rather firm timetable for site-selection and construction of nuclear waste disposal facilities. According to this act the first repository for disposal of commercial high-level waste must be in operation not later than in January 1998. (Author)

Modeling and Parameterization of Fuel Economy in Heavy Duty Vehicles (HDVs

Directory of Open Access Journals (Sweden)

Yunjung Oh

2014-08-01

Full Text Available The present paper suggests fuel consumption modeling for HDVs based on the code from the Japanese Ministry of the Environment. Two interpolation models (inversed distance weighted (IDW and Hermite and three types of fuel efficiency maps (coarse, medium, and dense were adopted to determine the most appropriate combination for further studies. Finally, sensitivity analysis studies were conducted to determine which parameters greatly impact the fuel efficiency prediction results for HDVs. While vitiating each parameter at specific percentages (±1%, ±3%, ±5%, ±10%, the change rate of the fuel efficiency results was analyzed, and the main factors affecting fuel efficiency were summarized. As a result, the Japanese transformation algorithm program showed good agreement with slightly increased prediction accuracy for the fuel efficiency test results when applying the Hermite interpolation method compared to IDW interpolation. The prediction accuracy of fuel efficiency remained unchanged regardless of the chosen fuel efficiency map data density. According to the sensitivity analysis study, three parameters (fuel consumption map data, driving force, and gross vehicle weight have the greatest impact on fuel efficiency (±5% to ±10% changes.

Air quality effects of alternative fuels. Final report

Energy Technology Data Exchange (ETDEWEB)

Guthrie, P.; Ligocki, M.; Looker, R.; Cohen, J.

1997-11-01

To support the Alternative Fuels Utilization Program, a comparison of potential air quality effects of alternative transportation fuels is being performed. This report presents the results of Phase 1 of this program, focusing on reformulated gasoline (RFG), methanol blended with 15 percent gasoline (M85), and compressed natural gas (CNG). The fuels are compared in terms of effects on simulated future concentrations of ozone and mobile source air toxics in a photochemical grid model. The fuel comparisons were carried out for the future year 2020 and assumed complete replacement of gasoline in the projected light-duty gasoline fleet by each of the candidate fuels. The model simulations were carried out for the areas surrounding Los Angeles and Baltimore/DC, and other (non-mobile) sources of atmospheric emissions were projected according to published estimates of economic and population growth, and planned emission control measures specific to each modeling domain. The future-year results are compared to a future-year run with all gasoline vehicle emissions removed. The results of the comparison indicate that the use of M85 is likely to produce similar ozone and air toxics levels as those projected from the use of RFG. Substitution of CNG is projected to produce significantly lower levels of ozone and the mobile source air toxics than those projected for RFG or M85. The relative benefits of CNG substitution are consistent in both modeling domains. The projection methodologies used for the comparison are subject to a large uncertainty, and modeled concentration distributions depend on meteorological conditions. The quantitative comparison of fuel effects is thus likely to be sensitive to alternative assumptions. The consistency of the results for two very different modeling domains, using very different base assumptions, lends credibility to the qualitative differentiation among these fuels. 32 refs., 42 figs., 47 tabs.

Light and Heavy Tactical Wheeled Vehicle Fuel Consumption Evaluations Using Fuel Efficient Gear Oils (FEGO)

Science.gov (United States)

2016-05-01

UNCLASSIFIED LIGHT AND HEAVY TACTICAL WHEELED VEHICLE FUEL CONSUMPTION EVALUATIONS USING FUEL EFFICIENT GEAR OILS (FEGO) FINAL... HEAVY TACTICAL WHEELED VEHICLE FUEL CONSUMPTION EVALUATIONS USING FUEL EFFICIENT GEAR OILS (FEGO) FINAL REPORT TFLRF No. 477 by Adam C...August 2014 – March 2016 4. TITLE AND SUBTITLE LIGHT AND HEAVY TACTICAL WHEELED VEHICLE FUEL CONSUMPTION EVALUATIONS USING FEUL EFFICIENT GEAR OILS

Theoretical models to predict the transient heat transfer performance of HIFAR fuel elements under non-forced convective conditions

International Nuclear Information System (INIS)

Green, W.J.

1987-04-01

Simple theoretical models have been developed which are suitable for predicting the thermal responses of irradiated research fuel elements of markedly different geometries when they are subjected to loss-of-coolant accident conditions. These models have been used to calculate temperature responses corresponding to various non-forced convective conditions. Comparisons between experimentally observed temperatures and calculated values have shown that a suitable value for surface thermal emissivity is 0.35; modelling of the fuel element beyond the region of the fuel plate needs to be included since these areas account for approximately 25 per cent of the thermal power dissipated; general agreement between calculated and experimental temperatures for both transient and steady-state conditions is good - the maximum discrepancy between calculated and experimental temperatures for a HIFAR Mark IV/V fuel element is ∼ 70 deg C, and for an Oak Ridge Reactor (ORR) box-type fuel element ∼ 30 deg C; and axial power distribution does not significantly affect thermal responses for the conditions investigated. Overall, the comparisons have shown that the models evolved can reproduce experimental data to a level of accuracy that provides confidence in the modelling technique and the postulated heat dissipation mechanisms, and that these models can be used to predict thermal responses of fuel elements in accident conditions that are not easily investigated experimentally

Material Performance of Fully-Ceramic Micro-Encapsulated Fuel under Selected LWR Design Basis Scenarios: Final Report

International Nuclear Information System (INIS)

Boer, B.; Sen, R.S.; Pope, M.A.; Ougouag, A.M.

2011-01-01

% and a more modest burnup target level of 500 MWd/kg ,the failure probability drops below 2.0 x 10 -5 , the typical performance of TRISO fuel made under the German HTR research program. An optimization study on particle design shows improved performance if the buffer size is increased from 100 to 120 (micro)mm while reducing the OPyC layer. The presence of the latter layer does not provide much benefit at high burnup levels (and fast fluence levels). Normally the shrinkage of the OPyC would result in a beneficial compressive force on the SiC coating. However, at high fluence levels the shrinkage is expected to turn into swelling, resulting in the opposite effect. However, this situation is different when the SiC-matrix, in which the particles are embedded, is also considered: the OPyC swelling can result in a beneficial compressive force on the SiC coating since outward displacement of the OPyC outer surface is inhibited by the presence of the also-swelling SiC matrix. Taking some credit for this effect by adopting a 5 (micro)mm SiC-matrix layer, the optimized particle (100 (micro)mm buffer and 10 (micro)mm OPyC), gives a failure probability of 1.9 x 10 -4 for conservative conditions. During a LOCA transient, assuming core re-flood in 30 seconds, the temperature of the coated particle can be expected to be about 200K higher than nominal temperature (900K). For this event the particle failure fraction for a conservative case is 1.0 x 10 -2 , for the optimized particle design. For a FGR-fraction of 50% this value reduces to 6.4 x 10 -4 .

Interim Storage of Spent Nuclear Fuel before Final Disposal in Germany - Regulator's view

International Nuclear Information System (INIS)

Arens, G.; Goetz, Ch.; Geupel, Sandra; Gmal, B.; Mester, W.

2014-01-01

For spent nuclear fuel management in Germany the concept of dry interim storage in dual purpose casks before direct disposal is applied. The Federal Office for Radiation Protection (BfS) is the competent authority for licensing of interim storage facilities. The competent authority for surveillance of operation is the responsible authority of the respective federal state (Land). Currently operation licenses for storage facilities have been granted for a storage time of 40 years and are based on safety demonstrations for all safety issues as safe enclosure, shielding, sub-criticality and decay heat removal under consideration of operation conditions. In addition, transportability of the casks for the whole storage period has to be provided. Due to current delay in site selection and exploration of a disposal site, an extension of the storage time beyond 40 years could be needed. This will cause appropriate actions by the licensee and the competent authorities as well. A brief description of the regulatory base of licensing and surveillance of interim storage is given from the regulators view. Furthermore the current planning for final disposal of spent nuclear fuel and high level waste and its interconnections between storage and disposal concepts are shortly explained. Finally the relevant aspects for licensing of extended storage time beyond 40 years will be discussed. Current activities on this issue, which have been initiated by the Federal Government, will be addressed. On the regulatory side a review and amendment of the safety guideline for interim storage of spent fuel has been performed and the procedure of periodic safety review is being implemented. A guideline for implementing an ageing management programme is available in a draft version. Regarding safety of long term storage a study focussing on the identification and evaluation of long term effects as well as gaps of knowledge has been finished in 2010. A continuation and update is currently underway

Final environmental impact statement: US Spent Fuel Policy. Executive summary

International Nuclear Information System (INIS)

1980-05-01

The analysis of the environmental impacts for storage of domestic fuel shows that the impacts for the full range of alternatives considered are relatively small compared with available resources or background exposure of the population from natural radiation sources. The differences in impacts of storage of domestic fuel are attributed to the amount of fuel stored in Independent Spent Fuel Storage (ISFS) facilities, the storage time, and, to a lesser degree, the differences in spent fuel transportation. The differences between comparable alternatives of implementing the policy or not implementing the policy are small. The difference in impacts of storage of foreign fuel are attributed to the amount of fuel received under the policy and to the disposition mode analyzed. The impact of storage of foreign fuel (a small fraction of the amount of domestic fuel considered) is also small. As a result of the small differences in environmental impacts of all cases considered for foreign fuel, environmental impacts probably will not strongly influence the selection of the case that best meets US nonproliferation goals. Proliferation benefits of the various cases analyzed vary significantly. The structure and level of fee charged for storage of spent fuel will affect the degree of participation in the spent fuel storage program by utilities. However, the range of participation is within the range of alternatives analyzed in the draft EISs on storage of US and foreign fuels, for which the environmental effects were found to be relatively small. The fee computed on the basis of full recovery of government costs should not significantly affect the cost of generating nuclear power

Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

Energy Technology Data Exchange (ETDEWEB)

Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

2016-08-01

To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

Reproduction of the RA reactor fuel element fabrication

International Nuclear Information System (INIS)

Novakovic, M.

1961-12-01

This document includes the following nine reports: Final report on task 08/12 - testing the Ra reactor fuel element; design concept for fabrication of RA reactor fuel element; investigation of the microstructure of the Ra reactor fuel element; Final report on task 08/13 producing binary alloys with Al, Mo, Zr, Nb and B additions; fabrication of U-Al alloy; final report on tasks 08/14 and 08/16; final report on task 08/32 diffusion bond between the fuel and the cladding of the Ra reactor fuel element; Final report on task 08/33, fabrication of the RA reactor fuel element cladding; and final report on task 08/36, diffusion of solid state metals [sr

Female Reproductive Effects of Exposure to Jet Fuel at U.S. Air Force Bases

Science.gov (United States)

1998-11-01

first breath sample until the Monday morning ofyour final breath sample. Avoid self-service refueling of your vehicle or lawn mower this week (outside of...fuel the lawn mower and mow the lawn for her - Post-pone painting, spraying pesticides/insecticides or using solvents if she might be in the area and...U No U_ 2. In the PAST WEEK, that is, one week ago today, did you... a).. .use the self-service tank when refueling of your vehicle or lawn mower this

Life cycle assessment of biomass-to-liquid fuels - Final report

Energy Technology Data Exchange (ETDEWEB)

Jungbluth, N.; Buesser, S.; Frischknecht, R.; Tuchschmid, M.

2008-02-15

This study elaborates a life cycle assessment of using of BTL-fuels (biomass-to-liquid). This type of fuel is produced in synthesis process from e.g. wood, straw or other biomass. The life cycle inventory data of the fuel provision with different types of conversion concepts are based on the detailed life cycle assessment compiled and published within a European research project. The inventory of the fuel use emissions is based on information published by automobile manufacturers on reductions due to the use of BTL-fuels. Passenger cars fulfilling the EURO3 emission standards are the basis for the comparison. The life cycle inventories of the use of BTL-fuels for driving in passenger cars are investigated from cradle to grave. The full life cycle is investigated with the transportation of one person over one kilometre (pkm) as a functional unit. This includes all stages of the life cycle of a fuel (biomass and fuel production, distribution, combustion) and the necessary infrastructure (e.g. tractors, conversion plant, cars and streets). The use of biofuels is mainly promoted for the reason of reducing the climate change impact and the use of scarce non-renewable resources e.g. crude oil. The possible implementation of BTL-fuel production processes would potentially help to achieve this goal. The emissions of greenhouse gases due to transport services could be reduced by 28% to 69% with the BTL-processes using straw, forest wood or short-rotation wood as a biomass input. The reduction potential concerning non-renewable energy resources varies between 37% und 61%. A previous study showed that many biofuels cause higher environmental impacts than fossil fuels if several types of ecological problems are considered. The study uses two single score impact assessment methods for the evaluation of the overall environmental impacts, namely the Eco-indicator 99 (H,A) and the Swiss ecological scarcity 2006 method. The transportation with the best BTL-fuel from short

Evaluation of friction forces between control bars and guide tubes of CAREM fuel element

International Nuclear Information System (INIS)

Ghiselli, Alberto M.; Fiori, Jose M.; Yedros, Pablo A.

2003-01-01

The design of the MSAC (Mecanismo del Sistema de ajuste y control, Mechanism of the Adjustment and control system, MACS) for CAREM reactor shall consider all loads that the reactor shall support during its operation. Several charges are originated by friction forces produced on contact points of the control Bars/Guide tubes assembly against other components. A test facility was designed and manufactured for the determination of friction components and the damping characteristics of the assembly formed by guide tubes fuel element, the control bars and the guide of the bar of the MACS. In this work a description of the test performed and the results obtained is presented. (author)

Review of the IAEA nuclear fuel cycle and material section activities connected with nuclear fuel including WWER fuel

International Nuclear Information System (INIS)

Sokolov, F.

2001-01-01

Program activities on Nuclear Fuel Cycle and Materials cover the areas of: 1) raw materials (B.1.01); 2) fuel performance and technology (B.1.02); 3) pent fuel (B.1.03); 4) fuel cycle issues and information system (B.1.04); 5) support to technical cooperation activities (B.1.05). The IAEA activities in fuel performance and technology in 2001 include organization of the fuel experts meetings and completion of the Co-ordinate Research Projects (CRP). The special attention is given to the advanced post-irradiation examination techniques for water reactor fuel and fuel behavior under transients and LOCA conditions. An international research program on modeling of activity transfer in primary circuit of NPP is finalized in 2001. A new CRP on fuel modeling at extended burnup (FUMEX II) has planed to be carried out during the period 2002-2006. In the area of spent fuel management the implementation of burnup credit (BUC) in spent fuel management systems has motivated to be used in criticality safety applications, based on economic consideration. An overview of spent fuel storage policy accounting new fuel features as higher enrichment and final burnup, usage of MOX fuel and prolongation of the term of spent fuel storage is also given

Fuel exchanging apparatus

International Nuclear Information System (INIS)

Imada, Takahiko; Sato, Hideo.

1975-01-01

Object: To provide a centripetal device, which has an initial spring force greater than a frictional force in an oscillating direction of a telescope mast, on a mast fixing device mounted on a body of fuel exchanging apparatus so that the telescope mast may be secured quickly returning to a predetermined initial position. Structure: When the body of fuel exchanging apparatus is stopped at a predetermined position, a tension spring, which has an initial spring force greater than a frictional force in an oscillating direction of the telescope mast, causes a lug to be forced by means of a push rod to position a sliding base plate to its original position. At the same time, a device of similar structure causes an operating arm to be positioned to the original position, and a lock pin urged by a cylinder is inserted into a through hole in the sliding base plate and operating arm so that the telescope mast may be fixed and retained. (Hanada, M.)

Full-Length High-Temperature Severe Fuel Damage Test No. 5: Final safety analysis

International Nuclear Information System (INIS)

Lanning, D.D.; Lombardo, N.J.; Panisko, F.E.

1993-09-01

This report presents the final safety analysis for the preparation, conduct, and post-test discharge operation for the Full-Length High Temperature Experiment-5 (FLHT-5) to be conducted in the L-24 position of the National Research Universal (NRU) Reactor at Chalk River Nuclear Laboratories (CRNL), Ontario, Canada. The test is sponsored by an international group organized by the US Nuclear Regulatory Commission. The test is designed and conducted by staff from Pacific Northwest Laboratory with CRNL staff support. The test will study the consequences of loss-of-coolant and the progression of severe fuel damage

Monitoring instrumentation spent fuel management program. Final report

International Nuclear Information System (INIS)

1979-01-01

Preliminary monitoring system methodologies are identified as an input to the risk assessment of spent fuel management. Conceptual approaches to instrumentation for surveillance of canister position and orientation, vault deformation, spent fuel dissolution, temperature, and health physics conditions are presented. In future studies, the resolution, reliability, and uncertainty associated with these monitoring system methodologies will be evaluated

Characteristic of The RSG-Gas Oxide Fuel Element Temperature Under Forced Convection And Natural Convection Mode

International Nuclear Information System (INIS)

Sudarmono

2000-01-01

One of the methods used for fuel element plate temperature measurement in RSG-Gas is a direct measurement. Evaluation on the measurement results were done by using HEATHYDE and NATCON code, which was then compared to the safety margin criteria. Results of thermalhydraulic measurement on transitional core both under forced and natural convection were compared with the results of calculations using the two codes. Measurement result for maximum fuel element plate temperature at typical working core of 30 MW, was 121 o C. The deviation between calculation and measurement result was under 9.75 %. Under normal operation, safety margin on DNB and OFI are 3.56 and 2.60, respectively. Natcon calculation result showed that the typical working core under the natural circulation mode, an onset of nucleate boiling (ONB)occurred at a core power level of 826 kW (2.8% of the nominal power)

Texas Hydrogen Highway Fuel Cell Hybrid Bus and Fueling Infrastructure Technology Showcase - Final Scientific/Technical Report

Energy Technology Data Exchange (ETDEWEB)

Hitchcock, David

2012-06-29

The Texas Hydrogen Highway project has showcased a hydrogen fuel cell transit bus and hydrogen fueling infrastructure that was designed and built through previous support from various public and private sector entities. The aim of this project has been to increase awareness among transit agencies and other public entities on these transportation technologies, and to place such technologies into commercial applications, such as a public transit agency. The initial project concept developed in 2004 was to show that a skid-mounted, fully-integrated, factory-built and tested hydrogen fueling station could be used to simplify the design, and lower the cost of fueling infrastructure for fuel cell vehicles. The approach was to design, engineer, build, and test the integrated fueling station at the factory then install it at a site that offered educational and technical resources and provide an opportunity to showcase both the fueling station and advanced hydrogen vehicles. The two primary technology components include: Hydrogen Fueling Station: The hydrogen fueling infrastructure was designed and built by Gas Technology Institute primarily through a funding grant from the Texas Commission on Environmental Quality. It includes hydrogen production, clean-up, compression, storage, and dispensing. The station consists of a steam methane reformer, gas clean-up system, gas compressor and 48 kilograms of hydrogen storage capacity for dispensing at 5000 psig. The station is skid-mounted for easy installation and can be relocated if needed. It includes a dispenser that is designed to provide temperaturecompensated fills using a control algorithm. The total station daily capacity is approximately 50 kilograms. Fuel Cell Bus: The transit passenger bus built by Ebus, a company located in Downey, CA, was commissioned and acquired by GTI prior to this project. It is a fuel cell plug-in hybrid electric vehicle which is ADA compliant, has air conditioning sufficient for Texas operations

Fuel Reduction for the Mobility Air Forces: Executive Summary

Science.gov (United States)

2015-01-01

to reduce fuel consumption. These measures include technology improvements ( aerodynamics , aircraft weight, propulsion, etc.) and fleet, flight, and...calculate fuel savings from an enterprise perspective. For example, there is significant literature on drag reduction of winglets ; however, most of this...an aircraft. If the weight of the paint can be reduced, then the fuel burn can also be reduced. • Microvanes. Microvanes are small aerodynamic

Comparison of radiative forcing impacts of the use of wood, peat, and fossil fuels

International Nuclear Information System (INIS)

Savolainen, I.; Hillebrand, K.; Nousiainen, I.; Sinisalo, J.

1994-01-01

The present study investigates the greenhouse impacts and the relevant time factors of the use of peat and wood for energy production and compares them with those of fossil fuels. Emissions and sinks of the whole energy production chain and subsequent use of the wood or peat production site are taken into account. The radiative forcing caused by energy production is used as a measure for the greenhouse impact. Economical considerations are not included. Radiative forcing is calculated for carbon dioxide (CO 2 ), methane (CH 4 ) and nitrous oxide (N 2 O) emissions. The real emissions of energy production are calculated by subtracting the emissions of non-use from the emissions of energy production. All the emissions are given as a function of time, i.e. their evolution over time is taken into account. At this point the estimates for some emission developments are quite crude and should be considered exemplary. The studied energy production chains can be divided roughly into three groups, if the greenhouse impact caused by continuous energy production of hundred years is considered. In this case forest residues, planted stands and unused merchantable wood cause the least radiative forcing per unit of primary energy generated. Natural gas and peat from cultivated peatland form the middle group. According to the calculations coal and conventional peat cause the greatest greenhouse impact

Nuclear fuel element

International Nuclear Information System (INIS)

Hirama, H.

1978-01-01

A nuclear fuel element comprises an elongated tube having upper and lower end plugs fixed to both ends thereof and nuclear fuel pellets contained within the tube. The fuel pellets are held against the lower end plug by a spring which is supported by a setting structure. The setting structure is maintained at a proper position at the middle of the tube by a wedge effect caused by spring force exerted by the spring against a set of balls coacting with a tapered member of the setting structure thereby wedging the balls against the inner wall of the tube, and the setting structure is moved free by pushing with a push bar against the spring force so as to release the wedge effect

Final Report - Low Temperature Combustion Chemistry And Fuel Component Interactions

Energy Technology Data Exchange (ETDEWEB)

Wooldridge, Margaret [Univ. of Michigan, Ann Arbor, MI (United States)

2017-02-24

Recent research into combustion chemistry has shown that reactions at “low temperatures” (700 – 1100 K) have a dramatic influence on ignition and combustion of fuels in virtually every practical combustion system. A powerful class of laboratory-scale experimental facilities that can focus on fuel chemistry in this temperature range is the rapid compression facility (RCF), which has proven to be a versatile tool to examine the details of fuel chemistry in this important regime. An RCF was used in this project to advance our understanding of low temperature chemistry of important fuel compounds. We show how factors including fuel molecular structure, the presence of unsaturated C=C bonds, and the presence of alkyl ester groups influence fuel auto-ignition and produce variable amounts of negative temperature coefficient behavior of fuel ignition. We report new discoveries of synergistic ignition interactions between alkane and alcohol fuels, with both experimental and kinetic modeling studies of these complex interactions. The results of this project quantify the effects of molecular structure on combustion chemistry including carbon bond saturation, through low temperature experimental studies of esters, alkanes, alkenes, and alcohols.

Final disposal of spent nuclear fuel-equipment for site characterization

International Nuclear Information System (INIS)

Almen, K.; Hansson, K.; Johansson, B.E.; Nilsson, G.; Andersson, O.; Wikberg, P.; Aahagen, H.

1983-05-01

The suitability of a certain geological formation as a repository for the final disposal of spent nuclear fuel can be determined only after detailed investigation and analysis. The purpose of the investigations is to provide information on the geology and the hydrology and chemistry of the site concerned. The value of these data largely depends on the way in which they have been collected. The report of the findings should enable the investigating party to evaluate the function and the accuracy of the equipment with which field data have been collected for KBS 3. This report describes the geophysical equipment, the hydraulic testing equipment, the water chemistry sample extracting equipment and the core-logging equipment used. The objectives of the instrument development have been: - to obtain a high data quality. - to collect data automatically in logs and tape recorders for direct transfer to a central processing unit. - to provide back-up in order to counteract loss of data. - to make instrument more efficient. (author)

Reference concepts for the final disposal of LWR spent fuel and other high activity wastes in Spain

International Nuclear Information System (INIS)

Huertas, F.; Ulibarri, A.

1993-01-01

Studies over the last three years have been recently concluded with the selection of a reference repository concept for the final disposal of spent fuel and other high activity wastes in deep geological formations. Two non-site specific preliminary designs, at a conceptual level, have been developed; one considers granite as the host rock and the other rock salt formations. The Spanish General Radioactive Waste Program also considers clay as a potential host rock for HLW deep disposal; conceptualization for a deep repository in clay is in the initial phase of development. The salt repository concept contemplates the disposal of the HLW in self-shielding casks emplaced in the drifts of an underground facility, excavated at a depth of 850 m in a bedded salt formation. The Custos Type I(7) cask admits up to seven intact PWR fuel assemblies or 21 of BWR type. The final repository facilities are planned to accept a total of 20,000 fuel assemblies (PWR and BWR) and 50 vitrified waste canisters over a period of 25 years. The total space needed for the surface facilities amounts to 322,000 m 2 , including the rock salt dump. The space required for the underground facilities amounts to 1.2 km 2 , approximately. The granite repository concept contemplates the disposal of the HLW in carbon steel canisters, embedded in a 0.75 m thick buffer of swelling smectite clay, in the drifts of an underground facility, excavated at a depth of 55 m in granite. Each canister can host 3 PWR or 9 BWR fuel assemblies. For this concept the total number of canisters needed amounts to 4,860. The space required for the surface and underground facilities is similar to that of the salt concept. The technical principles and criteria used for the design are discussed, and a description of the repository concept is presented

Proposed Expansion of German Air Force Operations at Holloman AFB, New Mexico. Final Environmental Impact Statement. Volume III: Comment Letters and Responses to Comments

National Research Council Canada - National Science Library

1998-01-01

The Final Environmental Impact Statement analyzed the potential environmental consequences from the proposal to beddown 30 additional German Air Force Tornado aircraft and 640 personnel at Holloman Air Force Base (AFB) New Mexico...

Hydropyrolysis of biomass to produce liquid hydrocarbon fuels. Final report. Biomass Alternative-Fuels Program

Energy Technology Data Exchange (ETDEWEB)

Fujita, R K; Bodle, W W; Yuen, P C

1982-10-01

The ojective of the study is to provide a process design and cost estimates for a biomass hydropyrolysis plant and to establish its economic viability for commercial applications. A plant site, size, product slate, and the most probable feedstock or combination of feedstocks were determined. A base case design was made by adapting IGT's HYFLEX process to Hawaiian biomass feedstocks. The HYFLEX process was developed by IGT to produce liquid and/or gaseous fuels from carbonaceous materials. The essence of the process is the simultaneous extraction of valuable oil and gaseous products from cellulosic biomass feedstocks without forming a heavy hard-to-handle tar. By controlling rection time and temperature, the product slate can be varied according to feedstock and market demand. An optimum design and a final assessment of the applicability of the HYFLEX process to the conversion of Hawaiian biomass was made. In order to determine what feedstocks could be available in Hawaii to meet the demands of the proposed hydropyrolysis plant, various biomass sources were studied. These included sugarcane and pineapple wastes, indigenous and cultivated trees and indigenous and cultivated shrubs and grasses.

The production of fuels and chemicals from food processing wastes & cellulosics. Final research report

Energy Technology Data Exchange (ETDEWEB)

Dale, M.C.; Okos, M.; Burgos, N. [and others

1997-06-15

High strength food wastes of about 15-20 billion pounds solids are produced annually by US food producers. Low strength food wastes of 5-10 billion pounds/yr. are produced. Estimates of the various components of these waste streams are shown in Table 1. Waste paper/lignocellulosic crops could produce 2 to 5 billion gallons of ethanol per year or other valuable chemicals. Current oil imports cost the US about $60 billion dollars/yr. in out-going balance of trade costs. Many organic chemicals that are currently derived from petroleum can be produced through fermentation processes. Petroleum based processes have been preferred over biotechnology processes because they were typically cheaper, easier, and more efficient. The technologies developed during the course of this project are designed to allow fermentation based chemicals and fuels to compete favorably with petroleum based chemicals. Our goals in this project have been to: (1) develop continuous fermentation processes as compared to batch operations; (2) combine separation of the product with the fermentation, thus accomplishing the twin goals of achieving a purified product from a fermentation broth and speeding the conversion of substrate to product in the fermentation broth; (3) utilize food or cellulosic waste streams which pose a current cost or disposal problem as compared to high cost grains or sugar substrates; (4) develop low energy recovery methods for fermentation products; and finally (5) demonstrate successful lab scale technologies on a pilot/production scale and try to commercialize the processes. The scale of the wastes force consideration of {open_quotes}bulk commodity{close_quotes} type products if a high fraction of the wastes are to be utilized.

Final Report. Fumex-III. Improvement of Models Used for Fuel Behaviour Simulation

International Nuclear Information System (INIS)

Kulacsy, Katalin

2013-01-01

The FUMEX-III coordinated research programme organised by the IAEA was the first FUMEX exercise in which AEKI (Hungarian Academy of Sciences KFKI Atomic Energy Research Institute) took part with the partial support of Paks NPP. The aim of the participation was to test the code FUROM developed at AEKI against not only measurements but also other fuel behaviour simulation codes, to share and discuss modelling experience and issues, and to establish acquaintance with fuel modellers in other countries. Among the numerous cases proposed for the programme, AEKI chose to simulate normal operation up to high burn-up and ramp tests, with special interest in VVER rods and PWR rods with annular pellets. The US PWR 16x16, the SPC RE GINNA, the Kola3-MIR, the IFA-519.9 cases and the AREVA idealised rod were thus selected. The present Final Report gives a short description of the FUROM models relevant to the selected cases, presents the results for the 5 cases and summarises the conclusions of the FUMEX-III programme. The input parameters used for the simulations can be found in the Appendix at the end of the Report. Observations concerning the IFPE datasets are collected for each dataset in their respective Sections for possible use in the IFPE database. (author)

High pressure combustion of liquid fuels. [alcohol and n-paraffin fuels

Science.gov (United States)

Canada, G. S.

1974-01-01

Measurements were made of the burning rates and liquid surface temperatures for a number of alcohol and n-paraffin fuels under natural and forced convection conditions. Porous spheres ranging in size from 0.64-1.9 cm O.D. were emloyed to simulate the fuel droplets. The natural convection cold gas tests considered the combustion in air of methanol, ethanol, propanol-1, n-pentane, n-heptane, and n-decane droplets at pressures up to 78 atmospheres. The pressure levels of the natural convection tests were high enough so that near critical combustion was observed for methanol and ethanol vaporization rates and liquid surface temperature measurements were made of droplets burning in a simulated combustion chamber environment. Ambient oxygen molar concentrations included 13%, 9.5% and pure evaporation. Fuels used in the forced convection atmospheric tests included those listed above for the natural convection tests. The ambient gas temperature ranged from 600 to 1500 K and the Reynolds number varied from 30 to 300. The high pressure forced convection tests employed ethanol and n-heptane as fuels over a pressure range of one to 40 atmospheres. The ambient gas temperature was 1145 K for the two combustion cases and 1255 K for the evaporation case.

Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

International Nuclear Information System (INIS)

Yang, Won Sik; Lin, C. S.; Hader, J. S.; Park, T. K.; Deng, P.; Yang, G.; Jung, Y. S.; Kim, T. K.; Stauff, N. E.

2016-01-01

This report presents the performance characteristics of two ''two-stage'' fast spectrum fuel cycle options proposed to enhance uranium resource utilization and to reduce nuclear waste generation. One is a two-stage fast spectrum fuel cycle option of continuous recycle of plutonium (Pu) in a fast reactor (FR) and subsequent burning of minor actinides (MAs) in an accelerator-driven system (ADS). The first stage is a sodium-cooled FR fuel cycle starting with low-enriched uranium (LEU) fuel; at the equilibrium cycle, the FR is operated using the recovered Pu and natural uranium without supporting LEU. Pu and uranium (U) are co-extracted from the discharged fuel and recycled in the first stage, and the recovered MAs are sent to the second stage. The second stage is a sodium-cooled ADS in which MAs are burned in an inert matrix fuel form. The discharged fuel of ADS is reprocessed, and all the recovered heavy metals (HMs) are recycled into the ADS. The other is a two-stage FR/ADS fuel cycle option with MA targets loaded in the FR. The recovered MAs are not directly sent to ADS, but partially incinerated in the FR in order to reduce the amount of MAs to be sent to the ADS. This is a heterogeneous recycling option of transuranic (TRU) elements

Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

Energy Technology Data Exchange (ETDEWEB)

Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Lin, C. S. [Purdue Univ., West Lafayette, IN (United States); Hader, J. S. [Purdue Univ., West Lafayette, IN (United States); Park, T. K. [Purdue Univ., West Lafayette, IN (United States); Deng, P. [Purdue Univ., West Lafayette, IN (United States); Yang, G. [Purdue Univ., West Lafayette, IN (United States); Jung, Y. S. [Purdue Univ., West Lafayette, IN (United States); Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Stauff, N. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-01-30

This report presents the performance characteristics of two “two-stage” fast spectrum fuel cycle options proposed to enhance uranium resource utilization and to reduce nuclear waste generation. One is a two-stage fast spectrum fuel cycle option of continuous recycle of plutonium (Pu) in a fast reactor (FR) and subsequent burning of minor actinides (MAs) in an accelerator-driven system (ADS). The first stage is a sodium-cooled FR fuel cycle starting with low-enriched uranium (LEU) fuel; at the equilibrium cycle, the FR is operated using the recovered Pu and natural uranium without supporting LEU. Pu and uranium (U) are co-extracted from the discharged fuel and recycled in the first stage, and the recovered MAs are sent to the second stage. The second stage is a sodium-cooled ADS in which MAs are burned in an inert matrix fuel form. The discharged fuel of ADS is reprocessed, and all the recovered heavy metals (HMs) are recycled into the ADS. The other is a two-stage FR/ADS fuel cycle option with MA targets loaded in the FR. The recovered MAs are not directly sent to ADS, but partially incinerated in the FR in order to reduce the amount of MAs to be sent to the ADS. This is a heterogeneous recycling option of transuranic (TRU) elements

Convective parameters in fuel elements for research nuclear reactors

International Nuclear Information System (INIS)

Lopez Martinez, C.D.

1992-01-01

The study of a prototype for the simulation of fuel elements for research nuclear reactors by natural convection in water is presented in this paper. This project is carry out in the thermofluids laboratory of National Institute of Nuclear Research. The fuel prototype has already been test for natural convection in air, and the first results in water are presented in this work. In chapter I, a general description of Triga Mark III is made, paying special atention to fuel-moderator components. In chapter II and III an approach to convection subject in its global aspects is made, since the intention is to give a general idea of the events occuring around fuel elements in a nuclear reactor. In chapter II, where an emphasis on forced convection is made, some basic concepts for forced convection as well as for natural convection are included. The subject of flow through cylinders is annotated only as a comparative reference with natural convection in vertical cylinders, noting the difference between used correlations and the involved variables. In chapter III a compilation of correlation found in the bibliography about natural convection in vertical cylinders is presented, since its geometry is the more suitable in the analysis of a fuel rod. Finally, in chapter IV performed experiments in the test bench are detailed, and the results are presented in form of tables and graphs, showing the used equations for the calculations and the restrictions used in each case. For the analysis of the prototypes used in the test bench, a constant and uniform flow of heat in the whole length of the fuel rod is considered. At the end of this chapter, the work conclusions and a brief explanation of the results are presented (Author)

Consolidated fuel reprocessing program

International Nuclear Information System (INIS)

Kuban, D.P.; Noakes, M.W.; Bradley, E.C.

1987-01-01

The Advanced Servomanipulator (ASM) System consists of three major components: the ASM slave, the dual arm master controller or master, and the control system. The ASM is a remotely maintainable force-reflecting servomanipulator developed at the Oak Ridge National Laboratory (ORNL) as part of the Consolidated Fuel Reprocessing Program of (CFRP). This new manipulator addresses requirements of advanced nuclear fuel reprocessing with emphasis on force reflection, remote maintainability, and reliability. It uses an all-gear force transmission system. The master arms were designed as a kinematic replica of ASM and use cable force transmission. Special digital control algorithms were developed to improve the system performance. The system is presently operational and undergoing evaluation. Preliminary testing has been completed and is reported. The system is now undergoing commercialization by transferring the technology to the private sector

Rock quality designation of the hydraulic properties in the near field of a final repository for spent nuclear fuel

International Nuclear Information System (INIS)

Carlsson, Hans; Carlsson, Leif; Pusch, Roland

1989-06-01

Quality assurance of a final repository for spent nuclear fuel requires detailed information on the characteristics of the rock, backfill, canisters and the waste itself. Furthermore, and of fundamental importance, is the knowledge of the behaviour of the integrated system of the waste and the different barriers. The in-situ characteristics of the rock must therefore be assessed and their influence on and interactions with the remaining barriers must be predicted and verified. A rock quality designation process of the hydraulic properties in the near-field is out-lined both for the KBS-3 system as well as for the WP-cave system. The process, once updated and approved, will be included in a Quality Assurance Program for the final repository for spent nuclear fuel. Some of the available methods for the near-field designation process are presented as well as techniques that need further development or are not developed at all. Finally, a presentation is given of a generic designation process of the KBS-3 and WP-cave repository systems in the previously investigated area in Central Sweden, where the final repository for reactor waste, SFR, is located. Geological and hydrogeological data are here at hand and it is therefore possible to carry out a simulation of how the designation process would be accomplished. (authors) (72 figs., 12 tabs., 43 refs.)

RTR spent fuel treatment and final waste storage

International Nuclear Information System (INIS)

Thomasson, J.

2000-01-01

A number of RTR operators have chosen in the past to send their spent fuel to the US in the framework of the US take back program. However, this possibility ends as of May 12th, 2006. 3 different strategies are left for managing RTR spent fuel: extended storage, direct disposal and treatment-conditioning through reprocessing. Whilst former strategies raise a number of uncertainties, the latter already offers a management solution. It features two advantages. It benefits from the long experience of existing flexible industrial facilities from countries like France. Secondly, it offers a dramatic volume reduction of the ultimate waste to be stored under well-characterized, stable and durable forms. RTR spent fuel management through reprocessing-conditioning offers a durable management solution that can be fully integrated in whatever global radioactive waste management policy, including ultimate disposal

TRIGA high wt -% LEU fuel development program. Final report

International Nuclear Information System (INIS)

West, G.B.

1980-07-01

The principal purpose of this work was to investigate the characteristics of TRIGA fuel where the contained U-235 was in a relatively high weight percent (wt %) of LEU (low enriched uranium - enrichment of less than 20%) rather than a relatively low weight percent of HEU (high enriched uranium). Fuel with up to 45 wt % U was fabricated and found to be acceptable after metallurgical examinations, fission product retention tests and physical property examinations. Design and safety analysis studies also indicated acceptable prompt negative temperature coefficient and core lifetime characteristics for these fuels

Final disposal of spent fuels and high activity waste: status and trends in the world

International Nuclear Information System (INIS)

Herscovich de Pahissa, Marta

2007-01-01

Geological disposal of spent nuclear fuel and high level waste from reprocessing, properly conditioned, is described. This issue is a major challenge related to radioactive waste management. Several options are analyzed, such as application of separation and transmutation to high level waste before final disposal; need of multinational repositories; a phased approach to deep geological disposal and long term surface storage. Bearing in mind this information, a future article will report the state of art in the world. (author) [es

Systems study 'Alternative Entsorgung'. Final report. Technical annex 6

International Nuclear Information System (INIS)

1984-08-01

In the conditioning plant, fuel elements which have been stored for ten years are loaded into transport containers, unloaded, identified and welded into a dry storage box. The dry store barrel is introduced into a final storage container, which, after being closed, is packed in lost shielding. This so-called final storage barrel is finally placed in a transport container and leaves the conditioning plant in this form by rail for transport to the final storage mine. The fuel element method of treatment 'packing of three complete fuel elements' was used as the reference process. In addition, the method of treatment 'fuel elements dismantled into fuel rods' was also examined. The handling of fuel elements and secondary waste treatment in the reference process are described in detail. (orig./HP) [de

SKB 91. Final disposal of spent nuclear fuel. Importance of the bedrock for safety

International Nuclear Information System (INIS)

1992-05-01

The safety of a deep repository for spent nuclear fuel has been assessed in this report. The spent fuel is assumed to be encapsulated in a copper canister and deposited at a depth of 600 m in the bedrock. The primary purpose has been to shed light on the importance of the geological features of the site for the safety of a final repository. The assessment shows that the encapsulated fuel will, in all likelihood, be kept isolated from the groundwater for millions of years. This is considerably longer than the more than 100 000 years that are required in order for the toxicity of the waste to have declined to a level equivalent to that of rich uranium ores. However, in order to be able to study the role of the rock as a barrier to the dispersal of radioactive materials, calculations have been carried out under the assumption that waste canisters leak. The results show that the safety of a carefully designed repository is only affected to a small extent by the ability of the rock to retain the escaping radionuclides. The primary role of the rock is to provide stable mechanical and chemical conditions in the repository over a long period of time so that the function of the engineered barriers is not jeopardized. (187 refs.) (au)

Forced convective post CHF heat transfer and quenching

International Nuclear Information System (INIS)

Nelson, R.A.

1980-01-01

This paper discusses mechanisms in the post-CHF region which provide understanding and qualitative prediction capability for several current forced convective heat transfer problems. In the area of nuclear reactor safety, the mechanisms are important in the prediction of fuel rod quenches for the reflood phase, blowdown phase, and possibly some operational transients with dryout. Results using the mechanisms to investigate forced convective quenching are presented. Data reduction of quenching experiments is discussed, and the way in which the quenching transient may affect the results of different types of quenching experiments is investigated. This investigation provides an explanation of how minimum wall superheats greater than the homogeneous nucleation temperature result, as well as how these may appear to be either hydrodynamically or thermodynamically controlled. Finally, the results of a parametric study of the effects of the mechanisms upon the LOFT L2-3 hotpin calculation are presented

Fuel assembly

International Nuclear Information System (INIS)

Bando, Masaru.

1993-01-01

As neutron irradiation progresses on a fuel assembly of an FBR type reactor, a strong force is exerted to cause ruptures if the arrangement of fuel elements is not displaced, whereas the fuel elements may be brought into direct contact with each other not by way of spacers to cause burning damages if the arrangement is displaced. In the present invention, the circumference of fuel elements arranged in a normal triangle lattice is surrounded by a wrapper tube having a hexagonal cross section, wire spacers are wound therearound, and deformable spacers are distributed to optional positions for fuel elements in the wrapper tube. Interaction between the fuel elements caused by irradiation is effectively absorbed, thereby enabling to delay the occurrence of the rupture and burning damages of the elements. (N.H.)

Nuclear fuel elements

International Nuclear Information System (INIS)

Nakai, Keiichi

1983-01-01

Purpose: To decrease the tensile stresses resulted in a fuel can as well as prevent decladding of fuel pellets into the bore holes by decreasing the inner pressure within the nuclear fuel element. Constitution: A fuel can is filled with hollow fuel pellets, inserted with a spring for retaining the hollow fuel pellets with an appropriate force and, thereafter, closely sealed at the both ends with end plugs. A cylindrical body is disposed into the bore holes of the hollow fuel pellets. Since initial sealing gases and/or gaseous nuclear fission products can thus be excluded from the bore holes where the temperature is at the highest level, the inner pressure of the nuclear fuel element can be reduced to decrease the tensile strength resulted to the fuel can. Furthermore, decladding of fuel pellets into the bore holes can be prevented. (Moriyama, K.)

Final Report of the National Black Health Providers Task Force on High Blood Pressure Education and Control.

Science.gov (United States)

Public Health Service (DHHS), Rockville, MD.

This is the final report of National Black Health Providers Task Force (NBHPTF) on High Blood Pressure Education and Control. The first chapter of the report recounts the history of the NBHPTF and its objectives. In the second chapter epidemiological evidence is presented to demonstrate the need for a suggested 20 year plan aimed at controlling…

FAMREC, PWR Lateral Mechanical Fuel Rod Assembly Response

International Nuclear Information System (INIS)

Guenzler, R.C.

1995-01-01

1 - Description of program or function: The Fuel Assembly Mechanical Response Code (FAMREC) calculates the lateral mechanical response of a row of fuel assemblies while allowing for two types of nonlinearities. The first type is a geometric nonlinearity in the form of gaps between individual assemblies and between peripheral assemblies and a boundary wall. Impacting is monitored across the gaps. The second nonlinearity is the permanent deformation of the fuel assembly spacer grid to compressive loading. 2 - Method of solution: The response is calculated in the modal plane. The coupled differential equations are solved in closed form using Laplace transformations. The discrete displacements and velocities are then calculated and the gaps in the system monitored at each axial elevation for impacting. These impact forces are then applied statistically at a given time-step, and equilibrium is found using a Gaussian elimination technique. Three impact force calculation methods are available: 1- a linear impact force and crushing load audit calculation, 2- a more detailed linear impact force and crushing load calculation, and 3- a non-linear grid calculation which allows for plastic deformation of the fuel assembly spacer grids. 3 - Restrictions on the complexity of the problem: Maxima of: 3601 time-steps and forces; 80 modes; 30 applied forces; 15 fuel assemblies; and 5 impact grids per assembly

Effect of fabrication and operating parameters on electrochemical property of anode and cathode for direct methanol fuel cells

International Nuclear Information System (INIS)

Liu, Guicheng; Zhou, Hongwei; Ding, Xianan; Li, Xinping; Zou, Dechun; Li, Xinyang; Wang, Xindong; Lee, Joong Kee

2016-01-01

Highlights: • A quick and simple method for optimizing assembly force of fuel cells. • Effect mechanisms of operating parameters on polarization of each electrode. • Working temperature is main factor to affect the optimal flow rates. • This paper is helpful to simulate the relation between operating parameters. - Abstract: A quick and simple method for optimizing assembly force of the direct methanol fuel cell has been introduced. Meanwhile, the effect mechanism of operating parameters on fuel cell performance and the properties of single anode and cathode have been intuitively investigated by a three-electrode system in this paper. The impedance curves indicate that internal resistance is the suitable intermediate to connect assembly torque and assembly force. The cathode polarization curve and limiting current density of methanol crossover are shown that the increasing methanol concentration markedly exacerbates the polarization in cathode due to serious methanol crossover phenomenon. Also, the higher cathode backpressure mainly improves cathode property, and lowers methanol crossover simultaneously. Finally, the summaries of peak power densities prove that the main factor that affected the optimal flow rates of methanol and oxygen is not the concentration or backpressure, but the working temperature.

Further analysis of extended storage of spent fuel. Final report of a co-ordinated research programme on the behaviour of spent fuel assemblies during extended storage (BEFAST-III) 1991-1996

International Nuclear Information System (INIS)

1997-05-01

Considerable quantities of spent fuel continue to be produced and to accumulate in a number of countries. Although some new reprocessing facilities have been constructed, many countries are investigating the option of extended spent fuel storage prior to reprocessing or fuel disposal. Wet storage continues to predominate as an established technology. However, dry storage is becoming increasingly used with many countries considering dry storage for the longer term. This Technical Document is the final report of the IAEA Co-ordinated Research Programme on the Behaviour of Spent Fuel Assemblies During Extended Storage (BEFAST-III, 1991-1996). It contains analyses of wet and dry spent fuel storage technologies obtained from 16 organizations representing 13 countries (Canada, Finland, France, Germany, Hungary, the Republic of Korea, Japan, the Russian Federation, Slovakia, Spain, Sweden, the United Kingdom and the USA) which participated in the co-ordinated research programme as participants or observers. The report contains information presented during the three Research Co-ordination meetings and also data which were submitted by the participants in response to request by the Scientific Secretary. 48 refs, 4 tabs

Stainless steel clad for light water reactor fuels. Final report

International Nuclear Information System (INIS)

Rivera, J.E.; Meyer, J.E.

1980-07-01

Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

Liquid fuels production from biomass. Final report

Energy Technology Data Exchange (ETDEWEB)

Levy, P. F.; Sanderson, J. E.; Ashare, E.; Wise, D. L.; Molyneaux, M. S.

1980-06-30

The current program to convert biomass into liquid hydrocarbon fuels is an extension of a previous program to ferment marine algae to acetic acid. In that study it was found that marine algae could be converted to higher aliphatic organic acids and that these acids could be readily removed from the fermentation broth by membrane or liquid-liquid extraction. It was then proposed to convert these higher organic acids via Kolbe electrolysis to aliphatic hydrocarbons, which may be used as a diesel fuel. The specific goals for the current porgram are: (1) establish conditions under which substrates other than marine algae may be converted in good yield to organic acids, here the primary task is methane suppression; (2) modify the current 300-liter fixed packed bed batch fermenter to operate in a continuous mode; (3) change from membrane extraction of organic acids to liquid-liquid extraction; (4) optimize the energy balance of the electrolytic oxidation process, the primary task is to reduce the working potential required for the electrolysis while maintaining an adequate current density; (5) scale the entire process up to match the output of the 300 liter fermenter; and (6) design pilot plant and commercial size plant (1000 tons/day) processes for converting biomass to liquid hydrocarbon fuels and perform an economic analysis for the 1000 ton/day design.

Closing the gap between spent fuel storage and final disposal in a multinational management system

International Nuclear Information System (INIS)

Bredell, P.J.

1999-01-01

In this paper, a multinational spent fuel management concept is proposed. The management concept is based on a service agreement between countries, which intend participating in a common spent fuel (SNF) management venture. Accordingly, one of the participants in this venture would act as the hosting country, while the others fulfil the role of customer countries. The hosting country would agree to accept SNF from customer countries under specific conditions, as required by the service agreement. The service agreement should cover a sufficient number of options that customers can use, such as storage, reprocessing or disposal. The service offering should be flexible enough to accommodate diverse customer requirements. Typically, the first step in the multinational management process is the storage of the SNF delivered to the hosting country. The final step being the disposal of the material in a deep geologic repository. This paper explores the ways and means of closing the gap between the first and last steps in the management process. (author)

Determination of consumption biogenic solid fuels in the commercial sector, trade, services (tertiary sector). Final report

International Nuclear Information System (INIS)

Viehmann, Cornelia; Westerkamp, Tanja; Schwenker, Andre; Schenker, Marian; Thraen, Daniela; Lenz, Volker; Ebert, Marcel

2012-01-01

The policy has both national and European level ambitious program aimed at expansion of renewable energy and related to the reduction of greenhouse gas emissions. In the national action plan for renewable energy of the Federal Republic of Germany these goals are defined by 2020. The share of renewable energy in the provision of heat and cold should therefore rise from 6.6% to 15.5% of gross final energy consumption. According to the increasing importance of solar-thermal, near-surface and geothermal heat, the relative share of biomass is decreasing. However biomass makes with those listed in the national action plan with 79% an essential amount in regenerative heat market [BMU 2010]. For the pursuit of goals and reviews, the support measures and packages of measures which are initiated in this context, a regular and timely reporting on the development of the above objectives is mandatory. The diverse and growing reporting requirements such as in the EU directive on the promotion of renewable energy, require, however well-founded knowledge of the sector-specific energy consumption from renewable sources. While the data available for use of biogenic solid fuels in the sectors household and industry has improved significantly in recent years, for the sector commercial sector, trade, services (tertiary sector) reliable figures are still lacking. Against this background, the objective is to present study, in close cooperation with the Federal Environment Agency (UBA), the determination of the final energy consumption biogenic solid fuels in the tertiary sector in Germany for the year 2008. The basis is, in addition to the development of the current knowledge of the energy and heat consumption, the delimitation and characterization of the sector and the development of an extrapolation tools. The demand for this tool is its expandability and update possibility. From the industry-nonspecific and industry-specific input data can be derived, collecting for the extrapolation

Fuel utilization improvements in a once-through PWR fuel cycle. Final report on Task 6

International Nuclear Information System (INIS)

Dabby, D.

1979-06-01

In studying the position of the United States Department of Energy, Non-proliferation Alternative Systems Assessment Program, this report determines the uranium saving associated with various improvement concepts applicable to a once-through fuel cycle of a standard four-loop Westinghouse Pressurized Water Reactor. Increased discharged fuel burnup from 33,000 to 45,000 MWD/MTM could achieve a 12% U 3 O 8 saving by 1990. Improved fuel management schemes combined with coastdown to 60% power, could result in U 3 O 8 savings of 6%

Fuel cell opportunities

Energy Technology Data Exchange (ETDEWEB)

Harris, K. [Hydrogenics Corporation, Mississauga, ON (Canada)

2002-07-01

The opportunities for fuel cell development are discussed. Fuel cells are highly efficient, reliable and require little maintenance. They also produce virtually zero emissions. The author stated that there are some complicated issues to resolve before fuel cells can be widely used. These include hydrogen availability and infrastructure. While the cost of fuel cells is currently very high, these costs are constantly coming down. The industry is still in the early stages of development. The driving forces for the development of fuel cells are: deregulation of energy markets, growing expectations for distributed power generation, discontinuity between energy supply and demand, and environmental concerns. 12 figs.

Case histories of West Valley spent fuel shipments: Final report

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

In 1983, NRC/FC initiated a study on institutional issues related to spent fuel shipments originating at the former spent fuel processing facility in West Valley, New York. FC staff viewed the shipment campaigns as a one-time opportunity to document the institutional issues that may arise with a substantial increase in spent fuel shipping activity. NRC subsequently contracted with the Aerospace Corporation for the West Valley Study. This report contains a detailed description of the events which took place prior to and during the spent fuel shipments. The report also contains a discussion of the shipment issues that arose, and presents general findings. Most of the institutional issues discussed in the report do not fall under NRC's transportation authority. The case histories provide a reference to agencies and other institutions that may be involved in future spent fuel shipping campaigns. 130 refs., 7 figs., 19 tabs.

Case histories of West Valley spent fuel shipments: Final report

International Nuclear Information System (INIS)

1987-01-01

In 1983, NRC/FC initiated a study on institutional issues related to spent fuel shipments originating at the former spent fuel processing facility in West Valley, New York. FC staff viewed the shipment campaigns as a one-time opportunity to document the institutional issues that may arise with a substantial increase in spent fuel shipping activity. NRC subsequently contracted with the Aerospace Corporation for the West Valley Study. This report contains a detailed description of the events which took place prior to and during the spent fuel shipments. The report also contains a discussion of the shipment issues that arose, and presents general findings. Most of the institutional issues discussed in the report do not fall under NRC's transportation authority. The case histories provide a reference to agencies and other institutions that may be involved in future spent fuel shipping campaigns. 130 refs., 7 figs., 19 tabs

Fuel cycle services

International Nuclear Information System (INIS)

Gruber, Gerhard J.

1990-01-01

TRIGA reactor operators are increasingly concerned about the back end of their Fuel Cycle due to a new environmental policy in the USA. The question how to close the Fuel Cycle will have to be answered by all operators sooner or later. Reprocessing of the TRIGA fuel elements is not available. Only long term storage and final disposal can be considered. But for such a storage or disposal a special treatment of the fuel elements and of course a final depository is necessary. NUKEM plans to undertake efforts to assist the TRIGA operators in this area. For that reason we need to know your special needs for today and tomorrow - so that potential processors can consider whether to offer these services on the market. (orig.)

Ohio State University Nuclear Reactor Laboratory HEU fuel shipment summary. Final

International Nuclear Information System (INIS)

1997-01-01

In November 1988, OSURR converted from HEU fuel to LEU fuel. As a result they needed to get rid of their HEU fuel by shipping it to Savannah River. The players in the fuel shipping game are: OSURR as the keeper of the fuel; DOE as the owner of fuel and shipper of record; Tri-State Motor Transit Co. for transporting the cask; Muth Brothers as the rigger responsible for getting the cask on and off the truck and in and out of the building; Hoffman LaRoche/Cintichem as the owner of the cask; Savannah River as the receiver of the fuel; and the NRC for approval of the Security Plan, QA Plan, etc. This report gives a chronological history of the events from February 1989 to June 1, 1995, the actual day of shipment. The cask was received at Savannah River on June 2, 1995

Evaluation of gasification and gas cleanup processes for use in molten carbonate fuel cell power plants. Final report. [Contains lists and evaluations of coal gasification and fuel gas desulfurization processes

Energy Technology Data Exchange (ETDEWEB)

Jablonski, G.; Hamm, J.R.; Alvin, M.A.; Wenglarz, R.A.; Patel, P.

1982-01-01

This report satisfies the requirements for DOE Contract AC21-81MC16220 to: List coal gasifiers and gas cleanup systems suitable for supplying fuel to molten carbonate fuel cells (MCFC) in industrial and utility power plants; extensively characterize those coal gas cleanup systems rejected by DOE's MCFC contractors for their power plant systems by virtue of the resources required for those systems to be commercially developed; develop an analytical model to predict MCFC tolerance for particulates on the anode (fuel gas) side of the MCFC; develop an analytical model to predict MCFC anode side tolerance for chemical species, including sulfides, halogens, and trace heavy metals; choose from the candidate gasifier/cleanup systems those most suitable for MCFC-based power plants; choose a reference wet cleanup system; provide parametric analyses of the coal gasifiers and gas cleanup systems when integrated into a power plant incorporating MCFC units with suitable gas expansion turbines, steam turbines, heat exchangers, and heat recovery steam generators, using the Westinghouse proprietary AHEAD computer model; provide efficiency, investment, cost of electricity, operability, and environmental effect rankings of the system; and provide a final report incorporating the results of all of the above tasks. Section 7 of this final report provides general conclusions.

Final Report - Stationary and Emerging Market Fuel Cell System Cost Assessment

Energy Technology Data Exchange (ETDEWEB)

Contini, Vince [Battelle Memorial Inst., Columbus, OH (United States); Heinrichs, Mike [Battelle Memorial Inst., Columbus, OH (United States); George, Paul [Battelle Memorial Inst., Columbus, OH (United States); Eubanks, Fritz [Battelle Memorial Inst., Columbus, OH (United States); Jansen, Mike [Battelle Memorial Inst., Columbus, OH (United States); Valluri, Manoj [Battelle Memorial Inst., Columbus, OH (United States); Mansouri, Mahan [Battelle Memorial Inst., Columbus, OH (United States); Swickrath, Mike [Battelle Memorial Inst., Columbus, OH (United States)

2017-04-30

The U.S. Department of Energy (DOE) is focused on providing a portfolio of technology solutions to meet energy security challenges of the future. Fuel cells are a part of this portfolio of technology offerings. To help meet these challenges and supplement the understanding of the current research, Battelle has executed a five-year program that evaluated the total system costs and total ownership costs of two technologies: (1) an ~80 °C polymer electrolyte membrane fuel cell (PEMFC) technology and (2) a solid oxide fuel cell (SOFC) technology, operating with hydrogen or reformate for different applications. Previous research conducted by Battelle, and more recently by other research institutes, suggests that fuel cells can offer customers significant fuel and emission savings along with other benefits compared to incumbent alternatives. For this project, Battelle has applied a proven cost assessment approach to assist the DOE Fuel Cell Technologies Program in making decisions regarding research and development, scale-up, and deployment of fuel cell technology. The cost studies and subsequent reports provide accurate projections of current system costs and the cost impact of state-of-the-art technologies in manufacturing, increases in production volume, and changes to system design on system cost and life cycle cost for several near-term and emerging fuel cell markets. The studies also provide information on types of manufacturing processes that must be developed to commercialize fuel cells and also provide insights into the optimization needed for use of off-the-shelf components in fuel cell systems. Battelle’s analysis is intended to help DOE prioritize investments in research and development of components to reduce the costs of fuel cell systems while considering systems optimization.

Advanced remotely maintainable force-reflecting servomanipulator concept

International Nuclear Information System (INIS)

Kuban, D.P.; Martin, H.L.

1984-01-01

A remotely maintainable force-reflecting servomanipulator concept is being developed at the Oak Ridge National Laboratory as part of the Consolidated Fuel Reprocessing Program. This new manipulator addresses requirements of advanced nuclear fuel reprocessing with emphasis on force reflection, remote maintainability, reliability, radiation tolerance, and corrosion resistance. The advanced servomanipulator is uniquely subdivided into remotely replaceable modules which will permit in situ manipulator repair by spare module replacement. Manipulator modularization and increased reliability are accomplished through a force transmission system that uses gears and torque tubes. Digital control algorithms and mechanical precision are used to offset the increased backlash, friction, and inertia resulting from the gear drives. This results in the first remotely maintainable force-reflecting servomanipulator in the world. 10 references, 4 figures, 1 table

The Swedish Radiation Protection Institute's regulations concerning the final management of spent nuclear fuel and nuclear waste - with background and comments

International Nuclear Information System (INIS)

2000-11-01

This report presents and comments on the Swedish Radiation Protection Institute's Regulations concerning the Protection of Human Health and the Environment in connection with the Final Management of Spent Nuclear Fuel or Nuclear Waste, SSI FS 1998: 1

Experimental study of defect power reactor fuel. Final report

International Nuclear Information System (INIS)

Forsyth, R.S.; Jonsson, T.

1982-01-01

Two BWR fuel rods, one intact and one defect, with the same manufacturing and irradiation data have been examined in a comparative study. The defect rod has been irradiated in a defect condition during approximately one reactor cycle and has consequently some secondary defects. The defect rod has two penetrating defects at a distance of about 1.5 meters from each other. Comparison with the intact rod shows a large Cs loss from the defect rod, especially between the cladding defects, where the loss is measured to about 30 %. The leachibility in deionized water is higher for Cs, U and Cm for fuel from the defect rod. The leaching results are more complex for Sr-90, Pu and Am. The fuel in the defect rod has undergone a change of structure with gain growth and formation of oriented fuel structure. The cladding of the defect rod is hydrided locally in some parts of the lower part of the rod and furthermore over a more extended region near the end of the rod. (Authors)

LMFBR fuel analysis. Task A: oxide fuel dynamics. Final report, October 1977--September 1978

International Nuclear Information System (INIS)

Dhir, V.K.; Frank, M.; Kastenberg, W.E.; McKone, T.E.

1979-03-01

Three aspects of LMFBR safety are discussed. The first concerns the potential reactivity effects of whole core fuel motion prior to pin failure in low ramp rate transient overpower accidents. The second concerns the effects of flow blockages following pin failure on the coolability of a core following an unprotected overpower transient. The third aspect concerns the safety related implications of using thorium based fuels in LMFBR's

Safety criteria for spent-fuel transport. Final report

International Nuclear Information System (INIS)

Goldmann, K.; Gekler, W.C.

1986-10-01

The focus of this study is on the question, ''Do current regulations provide reasonable assurance of safety for a transport scenario of spent fuel, as presently anticipated by the Department of Energy, under the Nuclear Waste Policy Act.'' This question has been addressed by developing a methodology for identifying the expected frequency of Accidents Which Exceed Regulatory Conditions in Severity (AWERCS) for spent fuel transport casks and then assessing the health effects resulting from that frequency. By applying the methodology to an illustrative case of road transports, it was found that the accidental release of radioactive material from impact AWERCS would make negligible contributions to health effects associated with spent fuel transports by road. It is also concluded that the current regulatory drop test requirements in 10 CFR 71.51 which form the basis for cask design and were used to establish AWERCS screening criteria for this study are adequate, and that no basis was found to conclude that cask performance under expected road accident conditions represents an undue risk to the public

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

The impact of a final disposal facility for spent nuclear fuel on a municipality's image

International Nuclear Information System (INIS)

Kankaanpaeae, H.; Haapavaara, L.; Lampinen, T.

1999-02-01

The study comprised on one hand a nationwide telephone interview (totally 800 interviews) aimed at mapping out the current image of possible host municipalities to a final disposal facility for spent nuclear fuel, and on the other hand some group interviews of people of another parish but of interest from the municipalities' point of view. The purpose of these group interviews was the same as that of the telephone interview, i.e. to find out what kind of an impact locating a final disposal facility of spent nuclear fuel in a certain municipality would have on the host municipality's image. Because the groups interviewed were selected on different grounds the results of the interviews are not fully comparable. The most important result of the study is that the current attitude towards a final disposal facility for spent nuclear fuel is calm and collected and that the matter is often considered from the standpoint of an outsider. The issue is easily ignored, classified as a matter 'which does not concern me', provided that the facility will not be placed too near one's own home. Among those interviewed the subject seemed not to be of any 'great interest and did not arouse spontaneous feelings for or against'. There are, however, deeply rooted beliefs concerning the facility and quite strong negative and positive attitudes towards it. The facility itself and the associated decision-making procedure arouse many questions, which at present to a large extent are still unexpressed because the subject is considered so remote. It is, however, necessary to give concrete answers to the questions because this makes it possible for people to relate the issue to daily life. It is further important that things arousing fear and doubts also can be discussed because a silence in this respect only emphasizes their importance. The attitude towards the facility is varying. On one hand there are economic and technical factors: the probable economic benefit from it, the obligation to

Axially alignable nuclear fuel pellets

International Nuclear Information System (INIS)

Johansson, E.B.; Klahn, D.H.; Marlowe, M.O.

1978-01-01

An axially alignable nuclear fuel pellet of the type stacked in end-to-end relationship within a tubular cladding is described. Fuel cladding failures can occur at pellet interface locations due to mechanical interaction between misaligned fuel pellets and the cladding. Mechanical interaction between the cladding and the fuel pellets loads the cladding and causes increased cladding stresses. Nuclear fuel pellets are provided with an end structure that increases plastic deformation of the pellets at the interface between pellets so that lower alignment forces are required to straighten axially misaligned pellets. Plastic deformation of the pellet ends results in less interactions beween the cladding and the fuel pellets and significantly lowers cladding stresses. The geometry of pellets constructed according to the invention also reduces alignment forces required to straighten fuel pellets that are tilted within the cladding. Plastic deformation of the pellets at the pellet interfaces is increased by providing pellets with at least one end face having a centrally-disposed raised area of convex shape so that the mean temperature and shear stress of the contact area is higher than that of prior art pellets

Final Technical Report Transport Task Force Activities

International Nuclear Information System (INIS)

P.W. Terry

2006-01-01

The Transport Task Force has functioned as the primary scientific organization in the area of magnetic-fusion confinement and transport since its inception in 1988. It has defined and set research directions, coordinated broad research efforts, advocated new funding initiatives, and created a highly successful and widely admired interactive culture between experiment, theory and modeling. The Transport Task Force carries out its activities under the direction of its chair and the Executive Committee. The Executive Committee is comprised of the leaders and deputy leaders of the scientific working groups. The working groups are structured and organized according to research needs and priorities and have been organized around the areas of Core Transport, H Mode and Pedestal, Fast Particle Transport, Transient Transport Phenomena, and Modeling and Simulation. A steering committee provides advise on TTF activities. Further information on the working groups and the structure and management of the TTF can be found at http://psfcwww2.psfc.mit.edu/ttf/index.html. The TTF holds an annual workshop. A summary of the workshops held during the period of this report is given in Appendix I. During the period of this report the Transport Task Force was involved in several significant activities. Foremost of these was a sweeping review of the status of transport science, the key research tasks for progress during the next 5-10 years, and a proposal for a funding initiative to ensure application of adequate resources to these problems. The conclusions of this study were incorporated into a white paper, which is copied below in Appendix II. Other significant activities have included the introduction of an extended, ongoing discussion on verification and validation as a requisite for defining and codifying the path toward predictive capability, the orchestration of a gradual shift of focus from ion thermal confinement to electron thermal confinement, and a joining of efforts on edge

Alternative Fuels Market and Policy Trends (Presentation)

Energy Technology Data Exchange (ETDEWEB)

Schroeder, A. N.

2013-09-01

Market forces and policies are increasing opportunities for alternative fuels. There is no one-size-fits-all, catch-all, silver-bullet fuel. States play a critical role in the alternative fuel market and are taking a leading role.

The relevance of axial burn-up profiles for the criticality safety analysis of spent nuclear fuel in a final repository

International Nuclear Information System (INIS)

Kilger, R.; Gmal, B.; Moser, E.F.

2008-01-01

Due to inhomogeneous neutron flux and moderator density distributions in the reactor core, the burn-up of a nuclear fuel assembly is not homogeneous but shows an axial distribution, typically with lower partial burn-up and thus higher remaining reactivity at the fuel ends in particular at the assembly top end. Beyond a burn-up of about 15 to 20 GWd/tHM, the multiplication factor K of the whole assembly is dominated by this lower-burnt end regions, and is usually higher than for assuming a homogeneous uniform distribution of the averaged burn-up. This behaviour commonly referred to as positive ''end effect'' is well known in burn-up credit considerations for transportation and storage casks and is being investigated also in the context of criticality analyses for final disposition of spent nuclear fuel. Sign and value of the end effect depend on several parameters. Based on a generic model one may not conclude that criticality in a final repository is a likely or expected event, but nevertheless it draws the attention to the fact that criticality is not excluded per se but has to be considered in the analysis and probably has to be encountered by certain appropriate measures, maybe e.g. by limitation of the amount of fissile material inside one single cask, or a rigorous prove for prevention of water ingress. The authors also conclude that the higher partial reactivity of the fuel ends has to be accounted for carefully in more realistic analyses of post-closure scenarios with respect to criticality safety.

Environmental implications of thorium use in selected nuclear fuel cycles. Final

International Nuclear Information System (INIS)

Buckley, D.W.; Simmons, G.L.; Ziskind, R.A.

1978-01-01

The objective of this study was to assess the environmental implications of the nuclear fuel cycle associated with the highly enriched uranium concept of the High Temperature Gas Cooled Reactor. Model fuel cycles were constructed for the HTGR and a reference light water reactor (LWR) cycle. Mass flows were developed, control technology cases proposed and costed, effluents determined, and population doses calculated. Emphasis was given to the intercomparison of the fuel cycles to delineate areas which show pronounced departure. The dose commitment received by the population both within and outside a radius of 50 miles of each facility was determined. The 100 year population dose commitments due to a single year's plant operation was selected to facilitate intercomparison among fuel cycle components. No account was taken for long term waste sources associated with the fuel cycle such as mill tailing piles or terminal waste storage (study groundrule). The resource utilization and radionuclide activity of various fuel cycle options for using thorium in a Pressurized Water Reactor were studied. These data were contrasted with similar results obtained for a uranium fuel PWR

A Study on the Fuel Assembly Seismic Analysis without Holddown Springs

International Nuclear Information System (INIS)

Kwon, O Cheol; Ha, Dong Geun; Lee, Kyou Seok; Jeon, Sang Yoon; Suh, Jung Min

2013-01-01

In this study, the effect for the fuel assembly removed holddown spring under seismic event has been evaluated through the comparison with the seismic analysis result of fuel assembly with holddown spring. In order to compare each design, the simplified fuel assembly seismic analysis models have been established according to reference. The mid grid impact force, natural frequency, and top nozzle displacement for each fuel assembly model has been analyzed using ANSYS. The fuel assembly seismic analyses without holddown springs are performed and compared to the model with holddown springs. The grid impact forces of CPM 1 and CPM 2 are almost doubled in comparison with CPM 3 and almost tripled in comparison with CPM 4 so the grid impact forces depend on CPM types. The grid impact forces of the fuel assembly model without holddown springs have similar tendencies in comparison with fuel assembly with holddown springs. Moreover, the model without holddown springs analysis time is much longer than the model with holddown springs. Consequently, it is moderate that the fuel assembly analysis model with holddown springs would be used for effective analysis even though the actual model has no holddown springs

A Study on the Fuel Assembly Seismic Analysis without Holddown Springs

Energy Technology Data Exchange (ETDEWEB)

Kwon, O Cheol; Ha, Dong Geun; Lee, Kyou Seok; Jeon, Sang Yoon; Suh, Jung Min [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

2013-10-15

In this study, the effect for the fuel assembly removed holddown spring under seismic event has been evaluated through the comparison with the seismic analysis result of fuel assembly with holddown spring. In order to compare each design, the simplified fuel assembly seismic analysis models have been established according to reference. The mid grid impact force, natural frequency, and top nozzle displacement for each fuel assembly model has been analyzed using ANSYS. The fuel assembly seismic analyses without holddown springs are performed and compared to the model with holddown springs. The grid impact forces of CPM{sub 1} and CPM{sub 2} are almost doubled in comparison with CPM{sub 3} and almost tripled in comparison with CPM{sub 4} so the grid impact forces depend on CPM types. The grid impact forces of the fuel assembly model without holddown springs have similar tendencies in comparison with fuel assembly with holddown springs. Moreover, the model without holddown springs analysis time is much longer than the model with holddown springs. Consequently, it is moderate that the fuel assembly analysis model with holddown springs would be used for effective analysis even though the actual model has no holddown springs.

Fuel pin transfer tool

International Nuclear Information System (INIS)

Patenaude, R. S.

1985-01-01

A fuel pin transfer tool has a latching device of the collet type attached to a first member movable vertically through a long work stroke enabling a fuel pin in an under water assembly to be engaged and withdrawn therefrom or placed therein and released. The latching device has a collet provided with a plurality of resilient fingers having cam portions normally spaced apart to receive the upper end of a fuel pin between them and a second member, movable vertically through a short stroke relative to the first member is provided with cam portions engageable with those of the fingers and is yieldably and resiliently held in a raised position in which its cam portions engage those of the fingers and force the fingers into their pin-gripping positions. When a predetermined force is applied to the second member, it is so moved that its cam portions are disengaged from the cam portions of the fingers permitting the latter to move into their normal relationship in which a gripped pin is released or another pin received but with their pin-gripping relationship positively re-established and maintained once the force on the tubular member is lessened. Movement of the first member in either direction and movement of the second member into its raised position is attended by forces inadequate to affect the integrity of fuel pin cladding. That force is applied in the preferred embodiment, by a power operated actuator which is within the upper portion of a housing and, in the preferred embodiment, carried by the long stroke member but always in the upper housing portion which is of a material sufficiently translucent to enable the actuator to be observed throughout the work stroke and is sufficiently light in weight to prevent the tool from being top heavy

Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P.O. 1236909 Final report

International Nuclear Information System (INIS)

Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

1978-12-01

The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel

Measurement of dynamic interaction between a vibrating fuel element and its support

Energy Technology Data Exchange (ETDEWEB)

Fisher, N.J.; Tromp, J.H.; Smith, B.A.W. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada). Chalk River Labs.

1996-12-01

Flow-induced vibration of CANDU{reg_sign} fuel can result in fretting damage of the fuel and its support. A WOrk-Rate Measuring Station (WORMS) was developed to measure the relative motion and contact forces between a vibrating fuel element and its support. The fixture consists of a small piece of support structure mounted on a micrometer stage. This arrangement permits position of the support relative to the fuel element to be controlled to within {+-} {micro}m. A piezoelectric triaxial load washer is positioned between the support and micrometer stage to measure contact forces, and a pair of miniature eddy-current displacement probes are mounted on the stage to measure fuel element-to-support relative motion. WORMS has been utilized to measure dynamic contact forces, relative displacements and work-rates between a vibrating fuel element and its support. For these tests, the fuel element was excited with broadband random force excitation to simulate flow-induced vibration due to axial flow. The relationship between fuel element-to-support gap or preload (i.e., interference or negative gap) and dynamic interaction (i.e., relative motion, contact forces and work-rates) was derived. These measurements confirmed numerical simulations of in-reactor interaction predicted earlier using the VIBIC code.

Man -made greenhouse gases trigger unified force to start global warming impacts referred to as climate change

International Nuclear Information System (INIS)

Karishnan, K.J.; Kalam, A.

2011-01-01

Global warming problems due to man-made greenhouse gases (GHGs), appear to be a serious concern and threat to the globe. CO/sub 2/, O/sub 3, NOx and HFC's are the main greenhouse gases and CO/sub 2/ is one of the main cause of global warming. CO/sub 2/ is emitted from burning fossil fuels to produce electricity from power plants and burning of gasoline in vehicles and airplanes. Global greenhouse gases and its sources in regions are discussed in this paper. This paper initially discusses the CO/sub 2/ emissions and the recycle of CO/sub 2/ in biodiesel. This paper mainly focuses on 'Unified Force'. The increase of H/sub 2/O in the sea due to warming of the globe triggers the 'Unified Force' or 'Self-Compressive Surrounding Pressure Force' which is proportional to the H/sub 2/O level in the sea to start global warming impacts referred to as climate change. This paper also points out the climate change and the ten surprising results of global warming. Finally, this paper suggests switching from fossil fuel technology to green energy technologies like biodiesel which recycles CO/sub 2/ emissions and also Hydrogen Energy and Fuel Cell Technologies which eradicates global warming impacts. The benefits of switching from fossil fuel to biodiesel and Hydrogen Energy utilization includes reduction of greenhouse gas emissions and pollution, economic independence by having distributed production and burning of biodiesel does not add extra CO/sub 2/ to the air that contributes global warming impacts. (author)

Final repository for spent nuclear fuel - the role of the municipality

International Nuclear Information System (INIS)

Berggren, Marie; Lindfors, Virpi; Andersson Oehrn, Barbro; Alm, Bertil; Soederblom, Anna-Lena; Berggren, Marie; Lindfors, Virpi

2014-01-01

In Sweden there is a long tradition of local self-government which is enshrined within the Swedish constitution, and the municipalities are responsible for matters relating to its inhabitants and their immediate environment. The municipality of Oesthammar has been engaged in the project of final repository for spent nuclear fuel since 1995 and by that time a consultative committee was established with representatives from all the political parties within the municipality and neighbouring municipalities. Future potentials as well as threats must be considered when making decisions on the most favourable site and the method used for the disposal of nuclear waste, and the application from SKB, as well as the review by the authorities, must stand up to a number of public demands. The work has included several stages of decisions for the municipality, due to the site selection process for SKB. The dialogue between the municipality and SKB as well as between the municipality and the authorities has been of great importance for getting the stepwise decision making process that has become practice in this question. The municipality has intensively followed the process concerning establishment of a final repository through consultation meetings, by being observer on meetings between SKB and Swedish Radiation Safety Authority (SSM), seminars, statements, etc. The openness and transparency throughout the process has been essential between all actors. However, if the municipalities have a right of absolute veto, the government still can say yes even if the municipality has said no

Site safety progress review of spent fuel central interim storage facility. Final report

International Nuclear Information System (INIS)

Gurpinar, A.; Serva, L.; Giuliani

1995-01-01

Following the request of the Czech Power Board (CEZ) and within the scope of the Technical Cooperation Project CZR/9/003, a progress review of the site safety of the Spent Fuel Central Interim Storage Facility (SFCISF) was performed. The review involved the first two stages of the works comprising the regional survey and identification of candidate sites for the underground and surface storage options. Five sites have been identified as a result of the previous works. The following two stages will involved the identification of the preferred candidate sites for the two options and the final site qualification. The present review had the purpose of assessing the work already performed and making recommendations for the next two stages of works

Honeywell's Working Parents Task Force. Final Report and Recommendations.

Science.gov (United States)

Honeywell, Inc., Minneapolis, Minn.

This publication provides a summary of the Honeywell Working Parent Task Force's recommendations on how to solve problems experienced by working parents. The Task Force consisted of three committees: the Employment Practices Committee (EPC); the Parent Education Committee (PEC); and the Child Care Facilities Committee (CCFC). After examining a…

Impact of fuel fabrication and fuel management technologies on uranium management

International Nuclear Information System (INIS)

Arnsberger, P.L.; Stucker, D.L.

1994-01-01

Uranium utilization in commercial pressurized water reactors is a complex function of original NSSS design, utility energy requirements, fuel assembly design, fuel fabrication materials and fuel fabrication materials and fuel management optimization. Fuel design and fabrication technologies have reacted to the resulting market forcing functions with a combination of design and material changes. The technologies employed have included ever-increasing fuel discharge burnup, non-parasitic structural materials, burnable absorbers, and fissile material core zoning schemes (both in the axial and radial direction). The result of these technological advances has improved uranium utilization by roughly sixty percent from the infancy days of nuclear power to present fuel management. Fuel management optimization technologies have also been developed in recent years which provide fuel utilization improvements due to core loading pattern optimization. This paper describes the development and impact of technology advances upon uranium utilization in modern pressurized water reactors. 10 refs., 3 tabs., 10 figs

Fuel element cellular grid structure and procedure to insert and withdraw fuel rods from that structure

International Nuclear Information System (INIS)

1975-01-01

A typical embodiment of the invention provides a means for selectively inserting and withdrawing one or more fuel rods from a fuel element cellular grid structure. The transverse stubs on one side of a long, thin bar are turned through 90deg to extend across the gap between mutually perpendicular grid structure plates. The extreme ends of these stubs engage the adhacent portions of the associated plates that form part of the grid cells. Pressing the stubs against the plate portions through the application of appropriate force in a longitudinal direction relative to the bar deflects the engaged plates through a sufficient distance to enable fuel rods to be inserted into or withdrawn from respective cells. After rod insertion, the force applied to the bar is released to enable the plates to relax and engage the fuel rods. The bars are rotated once more through 90deg and withdrawn from the grid structure. A similar procedure is employed to withdraw fuel rods from the grid structure

Vibration mechanism of fuel rod in axial flow

International Nuclear Information System (INIS)

Kang, Heung Seok; Yoon, Kyung Ho; Kim, Hyung Kyu; Song, Kee Nam

1998-08-01

This is a review on the previous researches for the vibration of fuel rod induced by axial flow. The analysis methods are classified into three categories accordingly as the researchers postulate the vibration to be self-excited, forced and parametric; the self-excited mechanism by Burgreen and Quinn, the forced one by Reavis, Gorman, kanazawa, and S. Chen, and the parametric one by Y. Chen. Quinn supposed that the centrifugal force by flow exaggerated the natural bow in the cylinder, and the flexural force by it diminished the bow by turns; this interactive motion leaded cylinder to vibration. The supporters to the forced mechanism considered the forces arising from pressure perturbation within the boundary layers as vibrating sources. Y. Chen insisted that the cylinder could only be excited to vibration in resonance by the small oscillation of mean flow velocity. The previous studies were based on the simple boundary conditions such as hinged-hinged or fixed-fixed single span. Therefore, for the more accurate prediction of the fuel rod vibration in reactor, the further studies need to reflect the actual boundary conditions of the fuel rod like axial force and continuous supports by grids. (author). 25 refs

Accelerating Acceptance of Fuel Cell Backup Power Systems - Final Report

Energy Technology Data Exchange (ETDEWEB)

Petrecky, James; Ashley, Christopher

2014-07-21

Since 2001, Plug Power has installed more than 800 stationary fuel cell systems worldwide. Plug Power’s prime power systems have produced approximately 6.5 million kilowatt hours of electricity and have accumulated more than 2.5 million operating hours. Intermittent, or backup, power products have been deployed with telecommunications carriers and government and utility customers in North and South America, Europe, the United Kingdom, Japan and South Africa. Some of the largest material handling operations in North America are currently using the company’s motive power units in fuel cell-powered forklifts for their warehouses, distribution centers and manufacturing facilities. The low-temperature GenSys fuel cell system provides remote, off-grid and primary power where grid power is unreliable or nonexistent. Built reliable and designed rugged, low- temperature GenSys delivers continuous or backup power through even the most extreme conditions. Coupled with high-efficiency ratings, low-temperature GenSys reduces operating costs making it an economical solution for prime power requirements. Currently, field trials at telecommunication and industrial sites across the globe are proving the advantages of fuel cells—lower maintenance, fuel costs and emissions, as well as longer life—compared with traditional internal combustion engines.

Low enrichment fuel conversion for Iowa State University. Final report

International Nuclear Information System (INIS)

Bullen, D.B.; Wendt, S.E.

1996-01-01

The UTR-10 research and teaching reactor at Iowa State University (ISU) has been converted from high-enriched fuel (HEU) to low- enriched fuel (LEU) under Grant No. DE-FG702-87ER75360 from the Department of Energy (DOE). The original contract period was August 1, 1987 to July 31, 1989. The contract was extended to February 28, 1991 without additional funding. Because of delays in receiving the LEU fuel and the requirement for disassembly of the HEU assemblies, the contract was renewed first through May 31, 1992, then through May 31, 1993 with additional funding, and then again through July 31, 1994 with no additional funding. In mid-August the BMI cask was delivered to Iowa State. Preparations are underway to ship the HEU fuel when NRC license amendments for the cask are approved

Holddown device for nuclear fuel assembly

International Nuclear Information System (INIS)

Anthony, A.J.

1978-01-01

An apparatus for preventing ''floating'' of nuclear-reactor fuel assemblies due to hydraulic forces is disclosed. The apparatus uses a holddown column made of the same material as the core barrel. The column is positioned in a center guide-tube location in the fuel assembly in such a manner as to enable it either to slide within the center guide tube or, if the center guide tube is replaced by the column, to slide through openings in the spacer grids. The lower end of the holddown column engages the lower end fitting of the fuel assembly, and the upper end of the column engages a flow plate to which holddown force is applied. As a consequence of this arrangement, holddown force is transmitted from the flow plate through the holddown column to the lower end fitting. Movement of the fuel assembly is thereby prevented without a compression load being applied to the fuel-assemb1ly structure. In addition, variations due to thermal expansion in the distance between the lower core plate and the upper core plate are largely made up for by corresponding variations in the holddown column because the holddown column and the core barrel can be made of the same material

Fuel exchanger control device

International Nuclear Information System (INIS)

Kurabayashi, Masaharu.

1985-01-01

Purpose: To improve the stability and the operationability of the fuel exchanging work by checking the validity of the data before the initiation of the work. Constitution: A floppy disc stores the initial charging state data showing the arrangement of fuel assemblies in the reactor core pool, data showing the working procedures for the fuel exchange and a final charged state data upon completion of the work. The initial data and the procedure data are read from the disk and stored once into a memory. Then, the initial data are sequentially performed on the memory in accordance with the procedure data and, thereafter, they were compared with the final data read from the disk. After confirming that there are no errors in the working data, the procedure data are orderly instructed to the fuel exchanger for performing fuel replacement. Accordingly, since the data are checked before the initiation of the work, the fuel exchange can be performed automatically thereby improving the operationability thereof. (Yoshino, Y.)

Final disposal of spent nuclear fuel in the Finnish bedrock

International Nuclear Information System (INIS)

1992-12-01

Teollisuuden Voima Oy (TVO) studies Finnish bedrock for the final disposal of the spent nuclear fuel from the Olkiluoto nuclear power plant. The study is in accordance with the decision in principle by Finnish government in 1983. The report is the summary of the preliminary site investigations carried out during the years 1987-1992. On the basis of these investigations a few areas will be selected for detailed site investigation. The characterization comprises five areas selected from the shortlist of potential candidate areas resulted in the earlier study during 1983-1985. Areas are located in different parts of Finland and they represent the main formations of the Finnish bedrock. Romuvaara area in Kuhmo and Veitsivaara area in Hyrynsalmi represent the Archean basement. Kivetty area in Konginkangas consists of mainly younger granitic rocks. Syyry in Sievi is located in transition area of Svecofennidic rocks and granitic rocks. Olkiluoto in Eurajoki represents migmatites in southern Finland. For the field investigations area-specific programs were planned and executed. The field investigations have comprised airborne survey by helicopter, geophysical surveys, geological mappings and samplings, deep and shallow core drillings, geophysical and hydrological borehole measurements and groundwater samplings

Prototypical spent nuclear nuclear fuel rod consolidation equipment, Phase 2: Final design report: Volume 2, Appendices: Part 1

International Nuclear Information System (INIS)

Ciez, A.P.

1987-01-01

The purpose of this specification is to establish functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation System. The Department of Energy-Idaho Operations Office (DOE-ID) is responsible for the implementation of the Prototypic Dry Rod Consolidation Demonstration Project. This program is to develop and demonstrate a fully qualified, licensable, cost-effective, dry spent fuel rod consolidation system by July 1989. The work is divided into four phases as follows: Phase I--Preliminary Design, Phase II--Final Design Option, Phase III--Fabrication and System Checkout Option, and Phase IV--Installation and Hot Demonstration Option. This specification is part of the Phase II effort. The objectives of this specification are to provide functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation equipment; establish specific tool and subsystem requirements such that the integrated and overall system requirements are satisfied; and establish positioning, envelope and size interface control requirements for each tool or subsystem such that the individual components will interface properly with the overall system design

Phase 1A Final Report for the AREVA Team Enhanced Accident Tolerant Fuels Concepts

Energy Technology Data Exchange (ETDEWEB)

Morrell, Mike E. [AREVA Federal Services LLC, Charlotte, NC (United States)

2015-03-19

In response to the Department of Energy (DOE) funded initiative to develop and deploy lead fuel assemblies (LFAs) of Enhanced Accident Tolerant Fuel (EATF) into a US reactor within 10 years, AREVA put together a team to develop promising technologies for improved fuel performance during off normal operations. This team consisted of the University of Florida (UF) and the University of Wisconsin (UW), Savannah River National Laboratory (SRNL), Duke Energy and Tennessee Valley Authority (TVA). This team brought broad experience and expertise to bear on EATF development. AREVA has been designing; manufacturing and testing nuclear fuel for over 50 years and is one of the 3 large international companies supplying fuel to the nuclear industry. The university and National Laboratory team members brought expertise in nuclear fuel concepts and materials development. Duke and TVA brought practical utility operating experience. This report documents the results from the initial “discovery phase” where the team explored options for EATF concepts that provide enhanced accident tolerance for both Design Basis (DB) and Beyond Design Basis Events (BDB). The main driver for the concepts under development were that they could be implemented in a 10 year time frame and be economically viable and acceptable to the nuclear fuel marketplace. The economics of fuel design make this DOE funded project very important to the nuclear industry. Even incremental changes to an existing fuel design can cost in the range of $100M to implement through to LFAs. If this money is invested evenly over 10 years then it can take the fuel vendor several decades after the start of the project to recover their initial investment and reach a breakeven point on the initial investment. Step or radical changes to a fuel assembly design can cost upwards of $500M and will take even longer for the fuel vendor to recover their investment. With the projected lifetimes of the current generation of nuclear power

Developments in fuel manufacturing

International Nuclear Information System (INIS)

Williams, T.

1997-01-01

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

A seismic analysis of Korean standard PWR fuels under transition core conditions

International Nuclear Information System (INIS)

Kim, Hyeong Koo; Park, Nam Kyu; Jang, Young Ki; Kim, Jae Ik; Kim, Kyu Tae

2005-01-01

The PLUS7 fuel is developed to achieve higher thermal performance, burnup and more safety margin than the conventional fuel used in the Korean Standard Nuclear Plants (KSNPs) and to sustain structural integrity under increased seismic requirement in Korea. In this study, a series of seismic analysis have been performed in order to evaluate the structural integrity of fuel assemblies associated with seismic loads in the KSNPs under transition core conditions replacing the Guardian fuel, which is a resident fuel in the KSNP reactors, with the PLUS7 fuel. For the analysis, transition core seismic models have been developed, based on the possible fuel loading patterns. And the maximum impact forces on the spacer grid and various stresses acting on the fuel components have been evaluated and compared with the through-grid strength of spacer grids and the stress criteria specified in the ASME code for each fuel component, respectively. Then three noticeable parameters regarding as important parameters governing fuel assembly dynamic behavior are evaluated to clarify their effects on the fuel impact and stress response. As a result of the study, it has been confirmed that both the PLUS7 and the Guardian fuel sustain their structural integrity under the transition core condition. And when the damping ratio is constant, increasing the natural frequency of fuel assembly results in a decrease in impact force. The fuel assembly flexural stiffness has an effect increasing the stress of fuel assembly, but not the impact force. And the spacer grid stiffness is directly related with the impact force response. (author)

NEW SOLID FUELS FROM COAL AND BIOMASS WASTE; FINAL

International Nuclear Information System (INIS)

Hamid Farzan

2001-01-01

Under DOE sponsorship, McDermott Technology, Inc. (MTI), Babcock and Wilcox Company (B and W), and Minergy Corporation developed and evaluated a sludge derived fuel (SDF) made from sewage sludge. Our approach is to dry and agglomerate the sludge, combine it with a fluxing agent, if necessary, and co-fire the resulting fuel with coal in a cyclone boiler to recover the energy and to vitrify mineral matter into a non-leachable product. This product can then be used in the construction industry. A literature search showed that there is significant variability of the sludge fuel properties from a given wastewater plant (seasonal and/or day-to-day changes) or from different wastewater plants. A large sewage sludge sample (30 tons) from a municipal wastewater treatment facility was collected, dried, pelletized and successfully co-fired with coal in a cyclone-equipped pilot. Several sludge particle size distributions were tested. Finer sludge particle size distributions, similar to the standard B and W size distribution for sub-bituminous coal, showed the best combustion and slagging performance. Up to 74.6% and 78.9% sludge was successfully co-fired with pulverized coal and with natural gas, respectively. An economic evaluation on a 25-MW power plant showed the viability of co-firing the optimum SDF in a power generation application. The return on equity was 22 to 31%, adequate to attract investors and allow a full-scale project to proceed. Additional market research and engineering will be required to verify the economic assumptions. Areas to focus on are: plant detail design and detail capital cost estimates, market research into possible project locations, sludge availability at the proposed project locations, market research into electric energy sales and renewable energy sales opportunities at the proposed project location. As a result of this program, wastes that are currently not being used and considered an environmental problem will be processed into a renewable

Simulated coal-gas fueled carbonate fuel cell power plant system verification. Final report, September 1990--June 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-03-01

This report summarizes work performed under U.S. Department of Energy, Morgantown Energy Technology Center (DOE/METC) Contract DE-AC-90MC27168 for September 1990 through March 1995. Energy Research Corporation (ERC), with support from DOE, EPRI, and utilities, has been developing a carbonate fuel cell technology. ERC`s design is a unique direct fuel cell (DFC) which does not need an external fuel reformer. An alliance was formed with a representative group of utilities and, with their input, a commercial entry product was chosen. The first 2 MW demonstration unit was planned and construction begun at Santa Clara, CA. A conceptual design of a 10OMW-Class dual fuel power plant was developed; economics of natural gas versus coal gas use were analyzed. A facility was set up to manufacture 2 MW/yr of carbonate fuel cell stacks. A 100kW-Class subscale power plant was built and several stacks were tested. This power plant has achieved an efficiency of {approximately}50% (LHV) from pipeline natural gas to direct current electricity conversion. Over 6,000 hours of operation including 5,000 cumulative hours of stack operation were demonstrated. One stack was operated on natural gas at 130 kW, which is the highest carbonate fuel cell power produced to date, at 74% fuel utilization, with excellent performance distribution across the stack. In parallel, carbonate fuel cell performance has been improved, component materials have been proven stable with lifetimes projected to 40,000 hours. Matrix strength, electrolyte distribution, and cell decay rate have been improved. Major progress has been achieved in lowering stack cost.

Fuel Cells

DEFF Research Database (Denmark)

Smith, Anders; Pedersen, Allan Schrøder

2014-01-01

Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications...... of the different types of fuel cells. Finally, their role in a future energy supply with a large share of fluctuating sustainable power sources, e.g., solar or wind, is surveyed....

Standard model for the safety analysis report of nuclear fuel reprocessing plants

International Nuclear Information System (INIS)

1980-02-01

This norm establishes the Standard Model for the Safety Analysis Report of Nuclear Fuel Reprocessing Plants, comprehending the presentation format, the detailing level of the minimum information required by the CNEN for evaluation the requests of Construction License or Operation Authorization, in accordance with the legislation in force. This regulation applies to the following basic reports: Preliminary Safety Analysis Report - PSAR, integrating part of the requirement of Construction License; and Final Safety Analysis Report (FSAR) which is the integrating part of the requirement for Operation Authorization

[Research and workshop on alternative fuels for aviation. Final report

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-09-01

The Renewable Aviation Fuels Development Center (RAFDC) at Baylor University was granted U. S. Department of Energy (US DOE) and Federal Aviation Administration (FAA) funds for research and development to improve the efficiency in ethanol powered aircraft, measure performance and compare emissions of ethanol, Ethyl Tertiary Butyl Ether (ETBE) and 100 LL aviation gasoline. The premise of the initial proposal was to use a test stand owned by Engine Components Inc. (ECI) based in San Antonio, Texas. After the grant was awarded, ECI decided to close down its test stand facility. Since there were no other test stands available at that time, RAFDC was forced to find additional support to build its own test stand. Baylor University provided initial funds for the test stand building. Other obstacles had to be overcome in order to initiate the program. The price of the emission testing equipment had increased substantially beyond the initial quote. Rosemount Analytical Inc. gave RAFDC an estimate of $120,000.00 for a basic emission testing package. RAFDC had to find additional funding to purchase this equipment. The electronic ignition unit also presented a series of time consuming problems. Since at that time there were no off-the-shelf units of this type available, one had to be specially ordered and developed. FAA funds were used to purchase a Super Flow dynamometer. Due to the many unforeseen obstacles, much more time and effort than originally anticipated had to be dedicated to the project, with much of the work done on a volunteer basis. Many people contributed their time to the program. One person, mainly responsible for the initial design of the test stand, was a retired engineer from Allison with extensive aircraft engine test stand experience. Also, many Baylor students volunteered to assemble the. test stand and continue to be involved in the current test program. Although the program presented many challenges, which resulted in delays, the RAFDC's test

From waste to traffic fuel -projects. Final report

Energy Technology Data Exchange (ETDEWEB)

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

2012-11-01

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

Effect of aviation fuel type and fuel injection conditions on the spray characteristics of pressure swirl and hybrid air blast fuel injectors

Science.gov (United States)

Feddema, Rick

fuels. Optical patternation data and line of sight laser diffraction data show that there is significant difference between jet fuels. Particularly at low fuel injection pressures (0.345 MPa) and cold temperatures (-40 C), the patternation data shows that the total surface area in the spray at 38.1 mm from the pressure swirl injector for the JP-10 fuel type is one-sixth the amount of the JP-8. Finally, this study compares the atomizer performance of a pressure swirl nozzle to a hybrid air blast nozzle. The total surface area for both the hybrid air blast nozzle and the pressure swirl nozzle show a similar decline in atomization performance at low fuel injection pressures and cold temperatures. However, the optical patternator radial profile data and the line of sight laser diffraction data show that the droplet size and spray distribution data are less affected by injection conditions and fuel type in the hybrid air blast nozzle, than they are in the pressure swirl nozzle. One explanation is that the aerodynamic forces associated with the swirler on the hybrid air blast nozzle control the distribution droplets in the spray. This is in contrast to the pressure swirl nozzle droplet distribution that is controlled by internal geometry and droplet ballistics.

Hoisting appliances and fuel handling equipment at nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-12-31

The guide is followed by the Finnish Centre for Radiation and Nuclear Safety (STUK) in regulating hoisting and handling equipment Class 3 at nuclear facilities. The guide is applied e.g. to the following equipment: reactor building overhead cranes, hoisting appliances at nuclear fuel storages, fuel handling machines, other hoisting appliances, which because of nuclear safety aspects are classified in Safety Class 3, and load-bearing devices connected with the above equipment, such as replaceable hoisting tools and auxiliary lifting devices. The regulating of hoisting and handling equipment comprises the following stages: handling of preliminary and final safety analysis reports, inspection of the construction plan, supervision of fabrication and construction inspection, and supervision of initial start-up and commissioning inspection. 36 refs. Translation. The original text is published under the same guide number. The guide is valid from 5 January 1987 and will be in force until further notice.

Procedure and apparatus for measuring the radial gap between fuel and surrounding cladding in a fuel rod for a nuclear reactor

International Nuclear Information System (INIS)

Olshausen, K.D.

1976-01-01

A device is described for measuring non-destructively the annular fuel-cladding gap in an irradiated or fresh fuel rod. The principle applied is that a force is applied to an arm which presses the cladding diametrically, thus deforming it until it touches the fuel pellet. By presenting the values of the force applied and the deformation produced on an XY recorder, the width of the gap is obtained. Alternatively the gap width may be obtained digitally. Since the gap is so small that the deformation is within the elastic range, the fuel rod may be reloaded in the reactor for further irradiation. (JIW)

Nuclear fuel

International Nuclear Information System (INIS)

Azevedo, J.B.L. de.

1980-01-01

All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.) [pt

Fuel assembly supporting structure

International Nuclear Information System (INIS)

Aisch, F.W.; Fuchs, H.P.; Knoedler, D.; Steinke, A.; Steven, J.

1976-01-01

For use in forming the core of a pressurized-water reactor, a fuel assembly supporting structure for holding a bundle of interspaced fuel rods, is formed by interspaced end pieces having holes in which the end portions of control rod guide tubes are inserted, fuel rod spacer grids being positioned by these guide tubes between the end pieces. The end pieces are fastened to the end portions of the guide tubes, to integrate the supporting structure, and in the case of at least one of the end pieces, this is done by means which releases that end piece from the guide tubes when the end pieces receive an abnormal thrust force directed towards each other and which would otherwise place the guide tubes under a compressive stress that would cause them to buckle. The spacer grids normally hold the fuel rods interspaced by distances determined by nuclear physics, and buckling of the control rod guide tubes can distort the fuel rod spacer grids with consequent dearrangement of the fuel rod interspacing. A sudden loss of pressure in a pressurized-water reactor pressure vessel can result in the pressurized coolant in the vessel discharging from the vessel at such high velocity as to result in the abnormal thrust force on the end pieces of each fuel assembly, which could cause buckling of the control rod guide tubes when the end pieces are fixed to them in the normal rigid and unyielding manner

Evolution of fuel rod support under irradiation impact on the mechanical behaviour of fuel assemblies

International Nuclear Information System (INIS)

Billerey, Antoine; Waeckel, Nicolas

2005-01-01

New fuel management targets imply to increase fuel assembly discharge burnup. Therefore, the prediction of the mechanical behaviour of the irradiated fuel assembly is essential such as excessive fuel assembly distortion induce incomplete Rod Cluster Control Assembly insertion problems (safety issue) or fuel rod vibration induced wear leading to leaking rods (plant operation problems). Within this framework, one of the most important parameter is the knowledge of the fuel rod support in the grid cell because it directly governs the mechanical behaviour of the fuel assembly and consequently allows to predict the behaviour of irradiated structures in terms of (1) axial and lateral deformation (global behaviour of the assembly) and (2) rod vibration induced wear (local behaviour of the rod). Generally, fuel rod support is provided by a spring-dimple system fixed to the grid. During irradiation, the spring force decreases and a gap between the rod and the spring may occur. This phenomenon is due to (1) stress relieving in the spring and in the dimples, (2) grid growth and (3) reduction of the rod diameter. Two models have been developed to predict the behaviour of the rod in the cell. The first model is dedicated to the evaluation of the spring force relaxation during irradiation. The second one can assess the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (1) the creep laws of the grid materials, (2) the growth law of the grid, (3) the evolution of rod diameter and (4) the design of the fuel rod support. The aim of this paper is to: (1) evaluate the consequences of grid support design modifications on the rod vibration sensitivity in terms of predicted rod to grid maximum gap during irradiation and time in operation with an open rod to grid gap, (2) evaluate, using a linear or non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the overall mechanical

Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

International Nuclear Information System (INIS)

D.R. Jackson; G.R. Kiebel

1999-01-01

Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training

Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

Energy Technology Data Exchange (ETDEWEB)

D.R. Jackson; G.R. Kiebel

1999-08-24

Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

Department of the Navy final environmental impact statement for a container system for the management of naval spent nuclear fuel

International Nuclear Information System (INIS)

1996-11-01

This Final Environmental Impact Statement (EIS) addresses six general alternative systems for the loading, storage, transport, and possible disposal of naval spent nuclear fuel following examination. This EIS describes environmental impacts of (1) producing and implementing the container systems (including those impacts resulting from the addition of the capability to load the containers covered in this EIS in dry fuel handling facilities at Idaho National Engineering Laboratory (INEL)); (2) loading of naval spent nuclear fuel at the Expended Core Facility or at the Idaho Chemical Processing Plant with subsequent storage at INEL; (3) construction of a storage facility (such as a paved area) at alternative locations at INEL; and (4) loading of containers and their shipment to a geologic repository or to a centralized interim storage site outside the State of Idaho once one becomes available. As indicated in the EIS, the systems and facilities might also be used for handling low-level radiological waste categorized as special case waste. The Navy's preferred alternative for a container system for the management of naval spent fuel is a dual-purpose canister system. The primary benefits of a dual-purpose canister system are efficiencies in container manufacturing and fuel reloading operations, and potential reductions in radiation exposure

Final Report - Durable Catalysts for Fuel Cell Protection during Transient Conditions

Energy Technology Data Exchange (ETDEWEB)

Atanasoski, Radoslav [3M Company, St. Paul, MN (United States); van der Vliet, Dennis [3M Company, St. Paul, MN (United States); Cullen, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Atanasoska, Ljiljana [3M Company, St. Paul, MN (United States)

2015-01-26

The objective of this project was to develop catalysts that will enable proton exchange membranes (PEM) fuel cell systems to weather the damaging conditions in the fuel cell at voltages beyond the thermodynamic stability of water during the transient periods of start-up/shut-down and fuel starvation. Such catalysts are required to make it possible for the fuel cell to satisfy the 2015 DOE targets for performance and durability. The project addressed a key issue of importance for successful transition of PEM fuel cell technology from development to pre-commercial phase. This issue is the failure of the catalyst and the other thermodynamically unstable membrane electrode assembly (MEA) components during start-up/shut-down and local fuel starvation at the anode, commonly referred to as transient conditions. During these periods the electrodes can reach potentials higher than the usual 1.23V upper limit during normal operation. The most logical way to minimize the damage from such transient events is to minimize the potential seen by the electrodes. At lower positive potentials, increased stability of the catalysts themselves and reduced degradation of the other MEA components is expected.

An analysis of heating fuel market behavior, 1989--1990

Energy Technology Data Exchange (ETDEWEB)

1990-06-01

The purpose of this report is to fully assess the heating fuel crisis from a broader and longer-term perspective. Using EIA final, monthly data, in conjunction with credible information from non-government sources, the pricing phenomena exhibited by heating fuels in late December 1989 and early January 1990 are described and evaluated in more detail and more accurately than in the interim report. Additionally, data through February 1990 (and, in some cases, preliminary figures for March) make it possible to assess the market impact of movements in prices and supplies over the heating season as a whole. Finally, the longer time frame and the availability of quarterly reports filed with the Securities and Exchange Commission make it possible to weigh the impact of revenue gains in December and January on overall profits over the two winter quarters. Some of the major, related issues raised during the House and Senate hearings in January concerned the structure of heating fuel markets and the degree to which changes in this structure over the last decade may have influenced the behavior and financial performance of market participants. Have these markets become more concentrated Was collusion or market manipulation behind December's rising prices Did these, or other, factors permit suppliers to realize excessive profits What additional costs were incurred by consumers as a result of such forces These questions, and others, are addressed in the course of this report.

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

Energy Technology Data Exchange (ETDEWEB)

McDeavitt, Sean M

2011-04-29

Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºC to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

International Nuclear Information System (INIS)

McDeavitt, Sean M.

2011-01-01

Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500 C to 600 C) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: (1) Hot working fabrication using mechanical alloying and extrusion - Design, fabricate, and assemble extrusion equipment - Extrusion database on DU metal - Extrusion database on U-10Zr alloys - Extrusion database on U-20xx-10Zr alloys - Evaluation and testing of tube sheath metals (2) Low-temperature sintering of U alloys - Design, fabricate, and assemble equipment - Sintering database on DU metal - Sintering database on U-10Zr alloys - Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research and Development (FCR and D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the

Fuel assembly

International Nuclear Information System (INIS)

Watanabe, Shoichi; Hirano, Yasushi.

1998-01-01

A one-half or more of entire fuel rods in a fuel assembly comprises MOX fuel rods containing less than 1wt% of burnable poisons, and at least a portion of the burnable poisons comprises gadolinium. Then, surplus reactivity at an initial stage of operation cycle is controlled to eliminate burnable poisons remained unburnt at a final stage, as well as increase thermal reactivity. In addition, the content of fission plutonium is determined to greater than the content of uranium 235, and fuel rods at corner portions are made not to incorporate burnable poisons. Fuel rods not containing burnable poisons are disposed at positions in adjacent with fuel rods facing to a water rod at one or two directions. Local power at radial center of the fuel assembly is increased to flatten the distortion of radial power distribution. (N.H.)

A porous medium approach for the fluid structure interaction modelling of a water pressurized nuclear reactor core fuel assemblies: simulation and experimentation

International Nuclear Information System (INIS)

Ricciardi, G.

2008-10-01

The designing of a pressurized water reactor core subjected to seismic loading, is a major concern of the nuclear industry. We propose, in this PhD report, to establish the global behaviour equations of the core, in term of a porous medium. Local equations of fluid and structure are space averaged on a control volume, thus we define an equivalent fluid and an equivalent structure, of which unknowns are defined on the whole space. The non-linear fuel assemblies behaviour is modelled by a visco-elastic constitutive law. The fluid-structure coupling is accounted for by a body force, the expression of that force is based on empirical formula of fluid forces acting on a tube subject to an axial flow. The resulting equations are solved using a finite element method. A validation of the model, on three experimental device, is proposed. The first one presents two fuel assemblies subjected to axial flow. One of the two fuel assemblies is deviated from its position of equilibrium and released, while the other is at rest. The second one presents a six assemblies row, immersed in water, placed on a shaking table that can simulate seismic loading. Finally, the last one presents nine fuel assemblies network, arranged in a three by three, subject to an axial flow. The displacement of the central fuel assembly is imposed. The simulations are in agreement with the experiments, the model reproduces the influence of the flow of fluid on the dynamics and coupling of the fuel assemblies. (author)

Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

Energy Technology Data Exchange (ETDEWEB)

Wegst, Ulrike G.K. [Dartmouth College, Hanover, NH (United States). Thayer School of Engineering; Allen, Todd [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States); Sridharan, Kumar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States)

2014-04-07

Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

International Nuclear Information System (INIS)

Wegst, Ulrike G.K.; Sridharan, Kumar

2014-01-01

Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

Suitability of Haestholmen Loviisa for final disposal of spent fuel. Preliminary study; Loviisan Haestholmenin soveltuvuus kaeytetyn polttoaineen loppusijoitukseen. Esiselvitys

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-12-01

Based on the amendment of the Nuclear Energy Act the spent nuclear fuel of Imatran Voima Oy (IVO) will be disposed of in Finland instead of returning it to Russia. After Teollisuuden Voima Oy (TVO) and IVO had founded a joint company Posiva Oy the work IVO started in 1995 was brought together with the ongoing research programme for final disposal of spent fuel and extended to a feasibility study. The feasibility study was launched in the beginning of 1996. The geological evaluation was mainly based on the previous investigations at the island. For this study the complementary geological mapping has been carried out at the Haestholmen and on the surrounding area with a radius of 20 km. (49 refs.).

Supply Chain-Based Solution to Prevent Fuel Tax Evasion: Final Report

Energy Technology Data Exchange (ETDEWEB)

Capps, Gary J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Engineering and Transportation Sciences Division; Franzese, Oscar [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Engineering and Transportation Sciences Division; Lascurain, Mary Beth [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Engineering and Transportation Sciences Division; Siekmann, Adam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Engineering and Transportation Sciences Division; Barker, Alan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Electrical and Electronics Systems Research Division

2016-07-28

The primary source of funding for the United States transportation system is derived from motor fuel and other highway use taxes. Loss of revenue attributed to fuel-tax evasion has been assessed to be somewhere between 1 billion and 3 billion per year. Any solution that addresses this problem needs to include not only the tax-collection agencies and auditors, but also the carriers transporting oil products and the carriers customers. This report presents a system developed by the Oak Ridge National Laboratory (ORNL) for the Federal Highway Administration which has the potential to reduce or eliminate many fuel-tax evasion schemes. The solution balances the needs of tax-auditors and those of the fuel-hauling companies and their customers. The system has three main components. The on-board subsystem combined sensors, tracking and communication devices, and software (the on-board Evidential Reasoning System, or obERS) to detect, monitor, and geo-locate the transfer of fuel among different locations. The back office sub-system (boERS) used self-learning algorithms to determine the legitimacy of the fuel loading and offloading (important for tax auditors) and detect potential illicit operations such as fuel theft (important for carriers and their customers, and may justify the deployment costs). The third sub-system, the Fuel Distribution and Auditing System or FDAS, is a centralized database, which together with a user interface allows tax auditors to query the data submitted by the fuel-hauling companies and correlate different parameters to quickly identify any anomalies. Industry partners included Barger Transport of Weber City, Virginia (fleet); Air-Weigh, of Eugene, Oregon (and their wires and harnesses); Liquid Bulk Tank (LBT) of Omaha, Nebraska (three five-compartment trailers); and Innovative Software Engineering (ISE) of Coralville, Iowa(on-board telematics device and back-office system). ORNL conducted a pilot test with the three instrumented vehicles

Method and jig for dismantling nuclear fuel assembly

International Nuclear Information System (INIS)

Urata, Megumi; Watahiki, Minoru.

1989-01-01

The object of the present inention is to extract a fuel element from a lower tie plate safely and at high efficiency by a remote control operation. That is, a forked top end of a lever of a dismantling jig is inserted between the tapered portion of a lower end plug and a lower tie plate. Then, a load is applied to the counter-lower end side of the lever by a motor. This exerts an elevating force to the fuel elements to easily release fixture between the lower end plug and the lower tie plate. Since the fuel can of fuel elements is not applied with a force by this mehtod, operation safety can be improved. (I.J.)

U.S. Air Force Hydroprocessed Renewable Jet (HRJ) Fuel Research

Science.gov (United States)

2012-07-01

general finding is that the neat HRJ fuels show high BOCLE unadditized and that most fuels respond immediately to low dosages of additive but quickly...Lo Power Cruise 3 Combat 4 Descent 5 Ground Idle 6 Start Time (el. min) 0 25 46 88 91 97 End Time (el. min

Long term integrity of spent fuel and construction materials for dry storage facilities

Energy Technology Data Exchange (ETDEWEB)

Saegusa, T [CRIEPI (Japan)

2012-07-01

In Japan, two dry storage facilities at reactor sites have already been operating since 1995 and 2002, respectively. Additionally, a large scale dry storage facility away from reactor sites is under safety examination for license near the coast and desired to start its operation in 2010. Its final storage capacity is 5,000tU. It is therefore necessary to obtain and evaluate the related data on integrity of spent fuels loaded into and construction materials of casks during long term dry storage. The objectives are: - Spent fuel rod: To evaluate hydrogen migration along axial fuel direction on irradiated claddings stored for twenty years in air; To evaluate pellet oxidation behaviour for high burn-up UO{sub 2} fuels; - Construction materials for dry storage facilities: To evaluate long term reliability of welded stainless steel canister under stress corrosion cracking (SCC) environment; To evaluate long term integrity of concrete cask under carbonation and salt attack environment; To evaluate integrity of sealability of metal gasket under long term storage and short term accidental impact force.

Gas-cooled fast reactor fuel-cost assessment. Final report, October 1978-September 1979

Energy Technology Data Exchange (ETDEWEB)

Thompson, M.L.

1979-01-01

This program, contracted to provide a Gas Cooled Fast Reactor (GCFR) fuel assembly fabrication cost assessment, comprised the following basic activities: establish agreement on the ground rules for cost assessment, prepare a fuel factory flow sheet, and prepare a cost assessment for fuel assembly fabrication. Two factory sizes, 250 and 25 MTHM/year, were considered for fuel assembly fabrication cost assessment. The work on this program involved utilizing GE LMFBR cost assessment and fuel factory studies experience to provide a cost assessment of GCFR fuel assembly fabrication. The recent impact of highly sensitive safety and safeguards environment policies on fuel factory containment, safety, quality assurance and safeguards costs are significantly higher than might have been expected just a few years ago. Fuel assembly fabrication costs are significant because they represent an estimated 30 to 60% of the total fuel cycle costs. In light of the relative high cost of fabrication, changes in the core and assembly design may be necessary in order to enhance the overall fuel cycle economics. Fabrication costs are based on similar operations and experience used in other fuel cycle studies. Because of extrapolation of present technology (e.g., remote fuel fabrication versus present contact fabrication) and regulatory requirements, conservative cost estimates were made.

Fuel temperature characteristics of the 37-element and CANFLEX fuel bundle

International Nuclear Information System (INIS)

Bae, Jun Ho; Rho, Gyu Hong; Park, Joo Hwan

2009-10-01

This report describes the fuel temperature characteristics of CANFLEX fuel bundles and 37-element fuel bundles for a different burnup of fuel. The program was consisted for seeking the fuel temperature of fuel bundles of CANFLEX fuel bundles and 37-element fuel bundles by using the method in NUCIRC. Fuel temperature has an increasing pattern with the burnup of fuel for CANFLEX fuel bundles and 37-element fuel bundles. For all the case of burnup, the fuel temperature of CANFLEX fuel bundles has a lower value than that of 37-element fuel bundles. Especially, for the high power channel, the CANFLEX fuel bundles show a lower fuel temperature as much as about 75 degree, and the core averaged fuel temperature has a lower fuel temperature of about 50 degree than that of 37-element fuel bundles. The lower fuel temperature of CANFLEX fuel bundles is expected to enhance the safety by reducing the fuel temperature coefficient. Finally, for each burnup of CANFLEX fuel bundles and 37-element fuel bundles, the equation was present for predicting the fuel temperature of a bundle in terms of a coolant temperature and bundle power

Development of Techniques for Spent Fuel Assay - Differential Dieaway Final Report

International Nuclear Information System (INIS)

Swinhoe, Martyn Thomas; Goodsell, Alison; Ianakiev, Kiril Dimitrov; Iliev, Metodi; Desimone, David J.; Rael, Carlos D.; Henzl, Vladimir; Polk, Paul John

2016-01-01

This report summarizes the work done under a DNDO R&D funded project on the development of the differential dieaway method to measure plutonium in spent fuel. There are large amounts of plutonium that are contained in spent fuel assemblies, and currently there is no way to make quantitative non-destructive assay. This has led NA24 under the Next Generation Safeguards Initiative (NGSI) to establish a multi-year program to investigate, develop and implement measurement techniques for spent fuel. The techniques which are being experimentally tested by the existing NGSI project do not include any pulsed neutron active techniques. The present work covers the active neutron differential dieaway technique and has advanced the state of knowledge of this technique as well as produced a design for a practical active neutron interrogation instrument for spent fuel. Monte Carlo results from the NGSI effort show that much higher accuracy (1-2%) for the Pu content in spent fuel assemblies can be obtained with active neutron interrogation techniques than passive techniques, and this would allow their use for nuclear material accountancy independently of any information from the operator. The main purpose of this work was to develop an active neutron interrogation technique for spent nuclear fuel.

Spent fuel verification options for final repository safeguards in Finland. A study on verification methods, their feasibility and safety aspects

International Nuclear Information System (INIS)

Hautamaeki, J.; Tiitta, A.

2000-12-01

The verification possibilities of the spent fuel assemblies from the Olkiluoto and Loviisa NPPs and the fuel rods from the research reactor of VTT are contemplated in this report. The spent fuel assemblies have to be verified at the partial defect level before the final disposal into the geologic repository. The rods from the research reactor may be verified at the gross defect level. Developing a measurement system for partial defect verification is a complicated and time-consuming task. The Passive High Energy Gamma Emission Tomography and the Fork Detector combined with Gamma Spectrometry are the most potential measurement principles to be developed for this purpose. The whole verification process has to be planned to be as slick as possible. An early start in the planning of the verification and developing the measurement devices is important in order to enable a smooth integration of the verification measurements into the conditioning and disposal process. The IAEA and Euratom have not yet concluded the safeguards criteria for the final disposal. E.g. criteria connected to the selection of the best place to perform the verification. Measurements have not yet been concluded. Options for the verification places have been considered in this report. One option for a verification measurement place is the intermediate storage. The other option is the encapsulation plant. Crucial viewpoints are such as which one offers the best practical possibilities to perform the measurements effectively and which would be the better place in the safeguards point of view. Verification measurements may be needed both in the intermediate storages and in the encapsulation plant. In this report also the integrity of the fuel assemblies after wet intermediate storage period is assessed, because the assemblies have to stand the handling operations of the verification measurements. (orig.)

Final environmental impact statement for the construction and operation of an independent spent fuel storage installation to store the Three Mile Island Unit 2 spent fuel at the Idaho National Engineering and Environmental Laboratory. Docket Number 72-20

International Nuclear Information System (INIS)

1998-03-01

This Final Environmental Impact Statement (FEIS) contains an assessment of the potential environmental impacts of the construction and operation of an Independent Spent Fuel Storage Installation (ISFSI) for the Three Mile Island Unit 2 (TMI-2) fuel debris at the Idaho National Engineering and Environmental laboratory (INEEL). US Department of Energy-Idaho Operations Office (DOE-ID) is proposing to design, construct, and operate at the Idaho Chemical Processing Plant (ICPP). The TMI-2 fuel debris would be removed from wet storage, transported to the ISFSI, and placed in storage modules on a concrete basemat. As part of its overall spent nuclear fuel (SNF) management program, the US DOE has prepared a final programmatic environmental impact statement (EIS) that provides an overview of the spent fuel management proposed for INEEL, including the construction and operation of the TMI-2 ISFSI. In addition, DOE-ID has prepared an environmental assessment (EA) to describe the environmental impacts associated with the stabilization of the storage pool and the construction/operation of the ISFSI at the ICPP. As provided in NRC's NEPA procedures, a FEIS of another Federal agency may be adopted in whole or in part in accordance with the procedures outlined in 40 CFR 1506.3 of the regulations of the Council on Environmental Quality (CEQ). Under 40 CFR 1506.3(b), if the actions covered by the original EIS and the proposed action are substantially the same, the agency adopting another agency's statement is not required to recirculate it except as a final statement. The NRC has determined that its proposed action is substantially the same as actions considered in DOE's environmental documents referenced above and, therefore, has elected to adopt the DOE documents as the NRC FEIS

Fuel Rod Vibration Measurement Method using a Flap and its Verification

Energy Technology Data Exchange (ETDEWEB)

Yoo, Joo Young; Park, Nam Gyu; Suh, Jung Min; Jeon, Kyeong Lak [KEPCO NF Co., Daejeon (Korea, Republic of)

2011-10-15

Flow-induced vibration is a critical factor for the mechanical integrity of a fuel rod. This vibration can cause leaked fuel through the mechanism, such as grid to rod fretting. To minimize the failures caused by flow-induced vibration, a robust design is needed which takes into account vibrational characteristics. That is, the spacer grid design should be developed to avoid any excessive vibration. On the one hand, if fuel rod vibration can be measured, an estimation of the excitation forces, which are a critical cause of rod failure, should be possible. Therefore, by applying an external force, flow-induced vibration can be roughly estimated when the fuel rod vibration model is used. KEPCO Nuclear Fuel developed the test loop to research flow-induced vibration as shown in Fig.1. The investigation flow-induced vibration (INFINIT) - the test facility - can measure the grid strap vibration and pressure drop of a 5x5 small scale fuel bundle. Basically, using a Laser Doppler Vibrometer (LDV), the vibration of a structure immersed in high speed fluid can be measured. Grid strap vibration is easily measured using an LDV. However, it is quite difficult to measure fuel rod vibration because of the round surface shape of the rods. In addition, measuring current method using the LDV, it was only possible to directly measure fuel rod vibration at the first row of the bundle as the rods behind the first row are obscured. To solve this problem, a thin flap, as shown in Fig. 2(a) can be used as a reflecting target, gaining access to rods within the bundle. The flap is attached to the fuel rod, as in Fig. 2(b). As a result, most of the inner rod vibration can be measured. Before using a flap to measure fuel rod vibration, a verification process was needed to show whether the LDV signal from the flap vibration provided equivalent and reliable signals. Therefore, impact testing was carried out on the fuel rod using a flap. The LDV signals were then compared with accelerometer

Alternative fuels for vehicles fleet demonstration program. Final report, volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-06-01

The Alternative Fuels for Vehicles Fleet Demonstration Program (AFV-FDP) was a multiyear effort to collect technical data for use in determining the costs and benefits of alternative-fuel vehicles (AFVs) in typical applications in New York State. This report, Volume 2, includes 13 appendices to Volume 1 that expand upon issues raised therein. Volume 1 provides: (1) Information about the purpose and scope of the AFV-FDP; (2) A summary of AFV-FDP findings organized on the basis of vehicle type and fuel type; (3) A short review of the status of AFV technology development, including examples of companies in the State that are active in developing AFVs and AFV components; and (4) A brief overview of the status of AFV deployment in the State. Volume 3 provides expanded reporting of AFV-FDP technical details, including the complete texts of the brochure Garage Guidelines for Alternative Fuels and the technical report Fleet Experience Survey Report, plus an extensive glossary of AFV terminology. The appendices cover a wide range of issues including: emissions regulations in New York State; production and health effects of ozone; vehicle emissions and control systems; emissions from heavy-duty engines; reformulated gasoline; greenhouse gases; production and characteristics of alternative fuels; the Energy Policy Act of 1992; the Clean Fuel Fleet Program; garage design guidelines for alternative fuels; surveys of fleet managers using alternative fuels; taxes on conventional and alternative fuels; and zero-emission vehicle technology.

Method and device for taking out spent fuel

International Nuclear Information System (INIS)

Kono, Takayuki; Shimamoto, Takijiro; Otsuka, Shiomi.

1994-01-01

Both ends of a fuel assembly are cut and removed, and spacers are removed into a single body of a fuel cladding tube. A plurality of the fuel cladding tubes as single bodies are sent into a ring-like electrode. An arc forming electric power source is actuated, to form arcs between the fuel cladding tube and the ring-like electric power source. A magnetic field forming power source is actuated, to form magnetic fields around the ring-like electrode by solenoid coils. The direction of the magnetic fluxes in this magnetic fields intersect the arcs. Then, the arcs are applied with a rotational force by the magnetic fields formed by the solenoid coils and rotate on the surface of the fuel cladding tube and the inner surface of the ring-like electrode. The fuel cladding tube is exhausted circumferentially uniformly by the rotational arcs. Since spent fuels are taken out by melting and scattering the fuel cladding tube by the heat and force of the arcs, the operation is facilitated. (I.N.)

New fuel review and assessment

International Nuclear Information System (INIS)

Miasnikov, A.

2001-01-01

This paper elaborated the complexity of the licensing process, concerning fuel core design and fuel management. All activities of the Czech Regulatory Body including activities, which shall require the State Office for Nuclear Safety (SUJB) license or authorization are based on legal framework. The licensing process consists of three major stages during which demonstration of safety is to be submitted by the applicants and reviewed and assessed by the regulatory body. The regulatory review and assessment leads to a series of regulatory decisions, which result in granting of an authorization (or the refusal) for site, construction and operation permits. To receive each of these permissions, the licensee has to submit to the SUJB Safety Analysis Reports (preliminary or final); technical specifications for safe operation; programs for non-active and active test (fuel loading, physical start-up, power start-up, trial operation); programs for quality assurance and quality control. Special attention is given to the design reliability and safety related influences of any design changes and usage of new fuel system design. The design compatibility should be reviewed especially from the standpoint of: 1) hermal hydraulic properties (vibration, hydraulic resistance, CHF correlation, fuel rod bowing, effect of spacing grids, pressure losses); 2) technical properties (rigidity, cycling fatigue, wear, cladding abrasion, deformation by external forces, kinetics of control assemblies drop); 3) chemical properties (corrosion, hydriding); 4) neutronic-physical properties (peaking factors, influence of different enrichments, water-uranium ration, shutdown reactivity margin, stability, maximum speed of the reactivity insertion, both calculated and experimental. The licensee is encouraged not only to show the fulfillment of the regulatory requirements but also to provide sufficient arguments with evidence that the safety of the total NPP is maintained

Responses to comments received on the draft final report of the Secretary of Energy Advisory Board Task Force on Radioactive Waste Management

Energy Technology Data Exchange (ETDEWEB)

1994-10-01

The Task Force solicited comments on its Draft Final Report from a variety of sources. Letters were sent to over 400 individuals who had expressed interest in the interest in the Department`s radioactive waste, management programs, a notice was placed in the Federal Register, the morning session of the January 1993 meeting of the full Secretary of Energy Advisory Board was given over to discussion of the draft, and Task Force members and staff presented the effort at several professional meetings. Altogether 32 written comments were received. They are reproduced here, followed in each case by the Task Force`s response to specific suggestions made to improve the draft. (The panel did not respond to comments that simply reflected policy preferences or that praised the group`s effort.) With one exception, those specific suggestions are highlighted and given a letter designation from {open_quotes}A{close_quotes} to {open_quotes}Z{close_quotes}. The Task Force`s responses, written in the Fall 1993, are labeled in a like manner. For the one exception, a comments submitted by Judy Treichel, the Task Force`s response is printed on copies of her annotated pages.

Water confinement effects on fuel assembly motion and damping

International Nuclear Information System (INIS)

Brenneman, B.; Shah, S.J.; Williams, G.T.; Strumpell, J.H.

2003-01-01

It has been established by other authors that the accelerations of the water confined by the reactor core baffle plates has a significant effect on the responses of all the fuel assemblies during LOCA or seismic transients. This particular effect is a consequence of the water being essentially incompressible, and thus experiencing the same horizontal accelerations as the imposed baffle plate motions. These horizontal accelerations of the fluid induce lateral pressure gradients that cause horizontal buoyancy forces on any submerged structures. These forces are in the same direction as the baffle accelerations and, for certain frequencies at least, tend to reduce the relative displacements between the fuel and baffle plates. But there is another confinement effect - the imposed baffle plate velocities must also be transmitted to the water. If the fuel assembly grid strips are treated as simple hydro-foils, these horizontal velocity components change the fluid angle of attack on each strip, and thus may induce large horizontal lift forces on each grid in the same direction as the baffle plate velocity. There is a similar horizontal lift due to inclined flow over the rods when axial flow is present. These combined forces appear to always reduce the relative displacements between the fuel and baffle plates for any significant axial flow velocity. Modeling this effect is very simple. It was shown in previous papers that the mechanism for the large fuel assembly damping due to axial flow may be the hydrodynamic forces on the grid strips, and that this is very well represented by discrete viscous dampers at each grid elevation. To include the imposed horizontal water velocity effects, on both the grids and rods, these dampers are simply attached to the baffle plate rather than 'ground'. The large flow-induced damping really acts in a relative reference frame rather than an absolute or inertial reference frame, and thus it becomes a flow-induced coupling between the fuel

Final Environmental Assessment Travis Air Force Base Burke Property Housing

National Research Council Canada - National Science Library

1999-01-01

... (40 CFR Parts 1500-1508) and Air Force Instruction 32-7061, the U.S. Air Force conducted an assessment of the potential environmental consequences of the construction of up to 281 military family housing units at Travis AFB...

Evolution of fuel rod support under irradiation consequences on the mechanical behavior of fuel assembly

International Nuclear Information System (INIS)

Billerey, A.; Bouffioux, P.

2002-01-01

The complete paper follows. According to the fuel management policy in French PWR with respect to high burn-up, the prediction of the mechanical behavior of the irradiated fuel assembly is required as far as excessive deformations of fuel assembly might lead to incomplete Rod Cluster Control Assembly insertion (safety problems) and fretting wear lead to leaking rods (plant operation problems). One of the most important parameter is the evolution of the fuel rod support in the grid cell as it directly governs the mechanical behavior of the fuel assembly and consequently allows to predict the behavior of irradiated structure in terms of (i) axial and lateral deformation (global behavior of the assembly) and (ii) fretting wear (local behavior of the rod). Fuel rod support is provided by a spring-dimple system fixed on the grid. During irradiation, the spring force decreases and a gap between the rod and the spring might open. This phenomenon is due to (i) irradiation-induced stress relaxation for the spring and for the dimples, (ii) grid growth and (iii) reduction of rod diameter. Two models have been developed to predict the behavior of the rod in the grid cell. The first model is able to evaluate the spring force relaxation during irradiation. The second one is able to evaluate the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (i) the creep laws of the grid materials, (ii) the growth law of the grid, (iii) the evolution of rod diameter and (iv) the design of the fuel rod support. The objectives of this paper are to: (i) evaluate the consequences of grid support design modifications on the fretting sensitivity in terms of predicted maximum gap during irradiation and operational time to gap appearance; (ii) evaluate, using a non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the mechanical behavior of the full assembly in terms of axial and

Final environmental impact statement: US Spent Fuel Policy. Comment letters on draft statements and major comments with DOE responses

International Nuclear Information System (INIS)

1980-05-01

Notice of availability of the draft environmental impact statement for storage of US spent power reactor fuel was published by DOE in the Federal Register on September 6, 1978. A subsequent notice was published in the Federal Register on December 14, 1978, on the availability of the draft EISs on storage of foreign fuel, on charge for spent fuel storage, and a supplement to the draft EIS on storage of US fuel. Copies of the draft EISs were distributed for review and comment by appropriate Federal agencies, state governments, and other organizations and individuals who were known to have an interest in spent fuel storage activities and those who requested them. Comments and views concerning the draft EISs were requested from other interested agencies, organizations, and individuals by means of the Federal Register notices. Approximately 1600 copies each of the draft Domestic, Foreign, and Charge EISs were distributed for comment. The closing date for comments to be received on the draft EISs was February 15, 1979. Copies of the EISs (upon publication) and comment letters received were placed for public inspection in DOE public document rooms at 10 locations throughout the country. To the extent practicable, comments received after the closing date were also considered in the preparation of the final EIS. A total of 78 comment letters (some with supplements) were received. These comment letters are reproduced in their entirety in Section I of this volume. Approximately 600 specific comments were identified in these letters

Novel materials for fuel cells operating on liquid fuels

Directory of Open Access Journals (Sweden)

César A. C. Sequeira

2017-05-01

Full Text Available Towards commercialization of fuel cell products in the coming years, the fuel cell systems are being redefined by means of lowering costs of basic elements, such as electrolytes and membranes, electrode and catalyst materials, as well as of increasing power density and long-term stability. Among different kinds of fuel cells, low-temperature polymer electrolyte membrane fuel cells (PEMFCs are of major importance, but their problems related to hydrogen storage and distribution are forcing the development of liquid fuels such as methanol, ethanol, sodium borohydride and ammonia. In respect to hydrogen, methanol is cheaper, easier to handle, transport and store, and has a high theoretical energy density. The second most studied liquid fuel is ethanol, but it is necessary to note that the highest theoretically energy conversion efficiency should be reached in a cell operating on sodium borohydride alkaline solution. It is clear that proper solutions need to be developed, by using novel catalysts, namely nanostructured single phase and composite materials, oxidant enrichment technologies and catalytic activity increasing. In this paper these main directions will be considered.

The Swedish Radiation Protection Institute's regulations concerning the final management of spent nuclear fuel and nuclear waste - with background and comments

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-11-01

This report presents and comments on the Swedish Radiation Protection Institute's Regulations concerning the Protection of Human Health and the Environment in connection with the Final Management of Spent Nuclear Fuel or Nuclear Waste, SSI FS 1998: 1.

Development of TRU transmuters for optimization of the global fuel cycle. Final Report for the NERI Project

International Nuclear Information System (INIS)

Lee, John C.

2009-01-01

This final report summarizes the research activities during the entire performance period of the NERI grant, including the extra 9 months granted under a no-cost time extension. Building up on the 14 quarterly reports submitted through October 2008, we present here an overview of the research accomplishments under the five tasks originally proposed in July 2004, together with citations for publications resulting from the project. The AFCI-NERI project provided excellent support for two undergraduate and 10 graduates students at the University of Michigan during a period of three years and nine months. Significant developments were achieved in three areas: (1) Efficient deterministic fuel cycle optimization algorithms both for PWR and SFR configurations, (2) Efficient search algorithm for PWR equilibrium cycles, and (3) Simplified Excel-based script for dynamic fuel cycle analysis of diverse cycles. The project resulted in a total of 8 conference papers and three journal papers, including two that will be submitted shortly. Three pending publications are attached to the report

Analysis of burnable poison in Ford Nuclear Reactor fuel to extend fuel lifetime. Final report, August 1, 1994--September 29, 1996

Energy Technology Data Exchange (ETDEWEB)

Burn, R.R.; Lee, J.C.

1996-12-01

The objective of the project was to establish the feasibility of extending the lifetime of fuel elements for the Ford Nuclear Reactor (FNR) by replacing current aluminide fuel with silicide fuel comprising a heavier uranium loading but with the same fissile enrichment of 19.5 wt% {sup 235}U. The project has focused on fuel designs where burnable absorbers, in the form of B{sub 4}C, are admixed with uranium silicide in fuel plates so that increases in the control reactivity requirements and peak power density, due to the heavier fuel loading, may be minimized. The authors have developed equilibrium cycle models simulating current full-size aluminide core configurations with 43 {approximately} 45 fuel elements. Adequacy of the overall equilibrium cycle approach has been verified through comparison with recent FNR experience in spent fuel discharge rates and simulation of reactor physics characteristics for two representative cycles. Fuel cycle studies have been performed to compare equilibrium cycle characteristics of silicide fuel designs, including burnable absorbers, with current aluminide fuel. These equilibrium cycle studies have established the feasibility of doubling the fuel element lifetime, with minimal perturbations to the control reactivity requirements and peak power density, by judicious additions of burnable absorbers to silicide fuel. Further study will be required to investigate a more practical silicide fuel design, which incorporates burnable absorbers in side plates of each fuel element rather than uniformly mixes them in fuel plates.

Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

Energy Technology Data Exchange (ETDEWEB)

N/A

2000-04-14

The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign and domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.

Messiah College Biodiesel Fuel Generation Project Final Technical Report

Energy Technology Data Exchange (ETDEWEB)

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

2012-03-30

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

Transitioning aluminum clad spent fuels from wet to interim dry storage

International Nuclear Information System (INIS)

Louthan, M.R. Jr.; Iyer, N.C.; Sindelar, R.L.; Peacock, H.B. Jr.

1994-01-01

The United States Department of Energy (DOE) currently owns several hundred metric tons of aluminum clad, spent nuclear fuel and target assemblies. The vast majority of these irradiated assemblies are currently stored in water basins that were designed and operated for short term fuel cooling prior to fuel reprocessing. Recent DOE decisions to severely limit the reprocessing option have significantly lengthened the time of storage, thus increasing the tendency for corrosion induced degradation of the fuel cladding and the underlying core material. The portent of continued corrosion, coupled with the age of existing wet storage facilities and the cost of continuing basin operations, including necessary upgrades to meet current facility standards, may force the DOE to transition these wet stored, aluminum clad spent fuels to interim dry storage. The facilities for interim dry storage have not been developed, partially because fuel storage requirements and specifications for acceptable fuel forms are lacking. In spite of the lack of both facilities and specifications, current plans are to dry store fuels for approximately 40 to 60 years or until firm decisions are developed for final fuel disposition. The transition of the aluminum clad fuels from wet to interim dry storage will require a sequence of drying and canning operations which will include selected fuel preparations such as vacuum drying and conditioning of the storage atmosphere. Laboratory experiments and review of the available literature have demonstrated that successful interim dry storage may also require the use of fuel and canister cleaning or rinsing techniques that preclude, or at least minimize, the potential for the accumulation of chloride and other potentially deleterious ions in the dry storage environment. This paper summarizes an evaluation of the impact of fuel transitioning techniques on the potential for corrosion induced degradation of fuel forms during interim dry storage

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

Energy Technology Data Exchange (ETDEWEB)

Klingler, James J [GENCO Infrastructure Solutions, Inc.

2014-05-06

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

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.

Evaluation of mechanical integrity for helical coil hold-down spring of PLUS7TM fuel

International Nuclear Information System (INIS)

Choi, Ki Sung; Kim, Yong Hwan; Kwon, Jung Tack; Kim, Kyu Tae

2004-01-01

Nuclear fuel assembly is subject to hydraulic forces generated by primary coolant flow during reactor operation. These forces may be sufficient to overcome the fuel assembly weight thereby allowing the fuel assembly to lift off of its support. To provide a positive hold-down margin against the upward coolant flow forces, helical coil springs or leaf springs are installed in the fuel assemblies. An advanced fuel for Korean Standard Nuclear Power Plants (KSNP), PLUS7 fuel has developed to get the thermal margin increase, failure free and high seismic resistance, etc. And the new designed helical coil hold-down spring was introduced into PLUS7 fuel assembly. The purpose of this paper is to evaluate the mechanical integrity for the helical coil hold-down spring of PLUS7 fuel assembly

Reliability prediction of large fuel cell stack based on structure stress analysis

Science.gov (United States)

Liu, L. F.; Liu, B.; Wu, C. W.

2017-09-01

The aim of this paper is to improve the reliability of Proton Electrolyte Membrane Fuel Cell (PEMFC) stack by designing the clamping force and the thickness difference between the membrane electrode assembly (MEA) and the gasket. The stack reliability is directly determined by the component reliability, which is affected by the material property and contact stress. The component contact stress is a random variable because it is usually affected by many uncertain factors in the production and clamping process. We have investigated the influences of parameter variation coefficient on the probability distribution of contact stress using the equivalent stiffness model and the first-order second moment method. The optimal contact stress to make the component stay in the highest level reliability is obtained by the stress-strength interference model. To obtain the optimal contact stress between the contact components, the optimal thickness of the component and the stack clamping force are optimally designed. Finally, a detailed description is given how to design the MEA and gasket dimensions to obtain the highest stack reliability. This work can provide a valuable guidance in the design of stack structure for a high reliability of fuel cell stack.

INNOVATIVE FOSSIL FUEL FIRED VITRIFICATION TECHNOLOGY FOR SOIL REMEDIATION. FINAL REPORT

International Nuclear Information System (INIS)

J. Hnat; L.M. Bartone; M. Pineda

2001-01-01

This Final Report summarizes the progress of Phases 3,3A and 4 of a waste technology Demonstration Project sponsored under a DOE Environmental Management Research and Development Program and administered by the U.S. Department of Energy National Energy Technology Laboratory-Morgantown (DOE-NETL) for an ''Innovative Fossil Fuel Fired Vitrification Technology for Soil Remediation''. The Summary Reports for Phases 1 and 2 of the Program were previously submitted to DOE. The total scope of Phase 3 was to have included the design, construction and demonstration of Vortec's integrated waste pretreatment and vitrification process for the treatment of low level waste (LLW), TSCA/LLW and mixed low-level waste (MLLW). Due to funding limitations and delays in the project resulting from a law suit filed by an environmental activist and the extended time for DOE to complete an Environmental Assessment for the project, the scope of the project was reduced to completing the design, construction and testing of the front end of the process which consists of the Material Handling and Waste Conditioning (MH/C) Subsystem of the vitrification plant. Activities completed under Phases 3A and 4 addressed completion of the engineering, design and documentation of the MH/C System such that final procurement of the remaining process assemblies can be completed and construction of a Limited Demonstration Project be initiated in the event DOE elects to proceed with the construction and demonstration testing of the MH/C Subsystem. Because of USEPA policies and regulations that do not require treatment of low level or low-level/PCB contaminated wastes, DOE terminated the project because there is no purported need for this technology

Modular dry storage of spent fuel

International Nuclear Information System (INIS)

Baxter, J.W.

1982-01-01

Long term uncertainties in US spent fuel reprocessing and storage policies and programs are forcing the electric utilities to consider means of storing spent fuel at the reactor site in increasing quantitities and for protracted periods. Utilities have taken initial steps in increasing storage capacity. Existing wet storage pools have in many cases been reracked to optimize their capacity for storing spent fuel assemblies

Ongoing evaluation of alternatively fueled buses : final report.

Science.gov (United States)

2016-05-01

The goal of this project is to continue collecting and reporting data on the performance and costs of alternatively fueled public transit vehicles in Florida in a consistent manner. Over the course of this project, researchers sent repeated data requ...

Can Sweden be forced to manage nuclear waste from other countries?

International Nuclear Information System (INIS)

Bergstroem, Ulf; Lindgren, Jonas; Nordfors, Lennart; Hallerby, Christer

2005-06-01

This report tries to answer two questions: Can Sweden, through supranational decisions, be forced to dispose of other nations nuclear waste within its own borders? and How can a Swedish stakeholder follow and influence EU-agendas, in particular concerning nuclear waste? Sometimes it has been argued that, through the EU-membership, Sweden could be forced to accept foreign radioactive waste for disposal in Sweden. However, the Swedish Nuclear Technology Act clearly states that license for final disposal spent nuclear fuels can not be given for fuel from another country than Sweden. At the Swedish EU entry it was found that the national legislation were compatible with the EU regulations, and as a consequence waste from foreign nuclear plants can not be disposed of in Sweden, even if the plant is owned by a Swedish company. Other conclusions in the report are: The system for decision making in EU is complex and slow and must be supported by supranational and international organs. The process is complex, but offers many opportunities for following-up and influencing. The risk for anything happening 'over-night' in the radioactive waste question is minimal. The issue of regulating radioactive waste management in a supranational way, i.e. by forcing a country to accept wastes from another country is not on the agenda. The Euratom treaty is interpreted in the way that such decision can not be made. The international discussions about regional repositories build upon voluntary agreements. Scenarios leading to coersive legislation are judged to be politically unrealistic - and needing important changes to be realised

Prediction of Combustion Stability and Flashback in Turbines with High-Hydrogen Fuel

Energy Technology Data Exchange (ETDEWEB)

Lieuwen, Tim [Georgia Inst. of Technology, Atlanta, GA (United States); Santavicca, Dom [Georgia Inst. of Technology, Atlanta, GA (United States); Yang, Vigor [Georgia Inst. of Technology, Atlanta, GA (United States)

2012-03-31

During the duration of this sponsorship, we broadened our understanding of combustion instabilities through both analytical and experimental work. Predictive models were developed for flame response to transverse acoustic instabilities and for quantifying how a turbulent flame responds to velocity and fuel/air ratio forcing. Analysis was performed on the key instability mechanisms controlling heat release response for flames over a wide range of instability frequencies. Importantly, work was done closely with industrial partners to transition existing models into internal instability prediction codes. Experimentally, the forced response of hydrogen-enriched natural gas/air premixed and partially premixed flames were measured. The response of a lean premixed flame was investigated, subjected to velocity, equivalence ratio, and both forcing mechanisms simultaneously. In addition, important physical mechanisms controlling the response of partially premixed flames to inlet velocity and equivalence ratio oscillations were analyzed. This final technical report summarizes our findings and major publications stemming from this program.

On the problems of the fuel cycles of nuclear fuels

International Nuclear Information System (INIS)

Schmidt-Kuester, W.J.; Wagner, H.F.

1976-01-01

A secured procurement with nuclear energy can be only achieved if a completely closed fuel cycle will be established. In the Federal Republic of Germany efforts are concentrated on the front end as well as on the back end of the fuel cycle. At the front end the main tasks are to secure uranium supply and to establish the necessary enrichment capacity. The German concept for the back end of the fuel cycle will provide for an integrated and co-located system for all necessary facilities including reprocessing, plutonium fuel fabrication, treatment, interim storage and final disposal of the radioactive wastes to be operational in the mid-80's. Responsibilities for establishing this system are shared between government and private industry. Government will provide for final waste disposal, industry will built and operate the other facilities. Another important point for the introduction of nuclear energy is to solve reliably the problems of protection of fissionable material, radioactive waste and nuclear facilities. German government has initiated respective activities and has started appropriate R+D-work. (orig.) [de

LEU fuel element produced by the Egyptian fuel manufacturing pilot plant

International Nuclear Information System (INIS)

Zidan, W.I.

2000-01-01

The Egyptian Fuel Manufacturing Pilot Plant, FMPP, is a Material Testing Reactor type (MTR) fuel element facility, for producing the specified fuel elements required for the Egyptian Second Research Reactor, ETRR-2. The plant uses uranium hexafluoride (UF 6 , 19.75% U 235 by wt) as a raw material which is processed through a series of the manufacturing, inspection and test plan to produce the final specified fuel elements. Radiological safety aspects during design, construction, operation, and all reasonably accepted steps should be taken to prevent or reduce the chance of accidents occurrence. (author)

Forced Air-Breathing PEMFC Stacks

Directory of Open Access Journals (Sweden)

K. S. Dhathathreyan

2012-01-01

Full Text Available Air-breathing fuel cells have a great potential as power sources for various electronic devices. They differ from conventional fuel cells in which the cells take up oxygen from ambient air by active or passive methods. The air flow occurs through the channels due to concentration and temperature gradient between the cell and the ambient conditions. However developing a stack is very difficult as the individual cell performance may not be uniform. In order to make such a system more realistic, an open-cathode forced air-breathing stacks were developed by making appropriate channel dimensions for the air flow for uniform performance in a stack. At CFCT-ARCI (Centre for Fuel Cell Technology-ARC International we have developed forced air-breathing fuel cell stacks with varying capacity ranging from 50 watts to 1500 watts. The performance of the stack was analysed based on the air flow, humidity, stability, and so forth, The major advantage of the system is the reduced number of bipolar plates and thereby reduction in volume and weight. However, the thermal management is a challenge due to the non-availability of sufficient air flow to remove the heat from the system during continuous operation. These results will be discussed in this paper.

Hydrogen in CANDU fuel elements

International Nuclear Information System (INIS)

Sejnoha, R.; Manzer, A.M.; Surette, B.A.

1995-01-01

Unirradiated and irradiated CANDU fuel cladding was tested to compare the role of stress-corrosion cracking and of hydrogen in the development of fuel defects. The results of the tests are compared with information on fuel performance in-reactor. The role of hydriding (deuteriding) from the coolant and from the fuel element inside is discussed, and the control of 'hydrogen gas' content in the element is confirmed as essential for defect-free fuel performance. Finally, implications for fuel element design are discussed. (author)

Fuel cycle management in Finland

International Nuclear Information System (INIS)

Vaeyrynen, H.; Mikkola, I.

1987-01-01

Both Finnish utilities producing nuclear power - Imatran Voima Oy (IVO) and Teollisuuden Voima Oy (Industrial Power Co. Ltd, TVO) - have created efficient fuel cycle management systems. The systems however differ in almost all respects. The reason is that the principal supplier for IVO is the Soviet Union and for TVO is Sweden. A common feature of both systems at the front end of the cycle is the building of stockpiles in order to provide for interruptions in fuel deliveries. Quality assurance supervision at the fuel factory for IVO is regulated by the Soviet Chamber of Commerce and Industry and a final control is made in Finland. The in-core fuel management is done by IVO using codes developed in Finland. The whole IVO fuel cycle is basically a leasing arrangement. The spent fuel is returned to the USSR after five years cooling. TVO carries out the in-core fuel management using a computer code system supplied by Asea-Atom. TVO is responsable for the back end of the cycle and makes preparations for the final disposal of the spent fuel in Finland. 6 refs., 2 figs

TAO2000 V2 computer-assisted force feedback tele-manipulators used as maintenance and production tools at the AREVA NC-La Hague fuel recycling plant

International Nuclear Information System (INIS)

Geffard, Franck; Garrec, Philippe; Piolain, Gerard; Brudieu, Marie-Anne; Thro, Jean-Francois; Coudray, Alain; Lelann, Eric

2012-01-01

During a 15-year joint research program, French Atomic Energy Agency Interactive Robotics Laboratory (CEA LIST) and AREVA have developed several remote operation devices, also called tele-robots. Some of them are now commonly used for maintenance operations at the AREVA NC (Nuclear Cycle) La Hague reprocessing plant. Since the first maintenance operation in 2005, several other successful interventions have been realized using the industrial MA23/RX170 tele-manipulation system. Moreover, since 2010, the through-the-wall tele-robot named MT200 TAO based on the slave arm of the MSM MT200 (La Calhene TM ), has been evaluated in an active production cell at the AREVA NC La Hague fuel recycling plant. Although these evaluations are ongoing, the positive results obtained have led to an update and industrialization program. All these developments are based on the same generic control platform, called TAO2000 V2. TAO2000 V2 is the second release of the CEA LIST core software platform dedicated to computer aided force-feedback tele-operation (TAO is the French acronym for computer aided tele-operation). This paper presents all these developments resulting from the joint research program CEA LIST/AREVA. The TAO2000 V2 controller is first detailed, and then two maintenance operations using the industrial robot RX170 are presented: the removal of the nuclear fuel dissolver wheel rollers and the cleanup of the dissolver wheel inter-bucket spaces. Finally, the new MT200 TAO system and its evaluations at the AREVA NC La Hague facilities are discussed. (authors)

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

Science.gov (United States)

Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

2017-05-01

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

Energy Technology Data Exchange (ETDEWEB)

Keiser, Dennis D., E-mail: dennis.keiser@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States); Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States); Ross Finlay, M. [Australian Nuclear Science and Technology Organization, PMB 1, Menai, NSW 2234 (Australia); Moore, Glenn; Medvedev, Pavel; Meyer, Mitch [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States)

2017-05-15

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

Safety research needs for carbide and nitride fueled LMFBR's. Final report

International Nuclear Information System (INIS)

Kastenberg, W.E.

1975-01-01

The results of a study initiated at UCLA during the academic year 1974--1975 to evaluate and review the potential safety related research needs for carbide and nitride fueled LMFBR's are presented. The tasks included the following: (1) Review Core and primary system designs for any significant differences from oxide fueled reactors, (2) Review carbide (and nitride) fuel element irradiation behavior, (3) Review reactor behavior in postulated accidents, (4) Examine analytical methods of accident analysis to identify major gaps in models and data, and (5) Examine post accident heat removal. (TSS)

Technical considerations associated with spent fuel acceptance. Final report

International Nuclear Information System (INIS)

Supko, E.M.

1996-06-01

This study was initiated by the Electric Power Research Institute (EPRI) to identify technical considerations associated with spent fuel acceptance and implementation of a waste management system that includes the use of transportable storage systems, and to serve as an opening dialogue among Standard Contract Holders and the department of Energy's Office of Civilian Radioactive Waste management (OCRWM) prior to the development of waste acceptance criteria or issuance of a Notice of Proposed Rulemaking by OCRWM to amend the Standard Contract. The original purpose of the Notice of Proposed Rulemaking was to address changes to the Standard Contract to implement a multi-purpose canister based system and to address other issues that were not adequately addressed in the standard contract. Even if DOE does not develop a multi-purpose canister based system for waste acceptance, it will still be necessary to develop waste acceptance criteria in order to accept spent fuel in transportable storage systems that are being deployed for at-reactor storage. In this study, technical issues associated with spent fuel acceptance will be defined and potential options and alternatives for resolution of technical considerations will be explored

Disposal of spent fuel

International Nuclear Information System (INIS)

Blomeke, J.O.; Ferguson, D.E.; Croff, A.G.

1978-01-01

Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed between the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom

Fuel cracking in relation to fuel oxidation in support of an out-reactor instrumented defected fuel experiment

Energy Technology Data Exchange (ETDEWEB)

Quastel, A.; Thiriet, C. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Lewis, B., E-mail: brent.lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada); Corcoran, E., E-mail: emily.corcoran@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

2014-07-01

An experimental program funded by the CANDU Owners Group (COG) is studying an out-reactor instrumented defected fuel experiment in Stern Laboratories (Hamilton, Ontario) with guidance from Atomic Energy of Canada Limited (AECL). The objective of this test is to provide experimental data for validation of a mechanistic fuel oxidation model. In this experiment a defected fuel element with UO{sub 2} pellets will be internally heated with an electrical heater element, causing the fuel to crack. By defecting the sheath in-situ the fuel will be exposed to light water coolant near normal reactor operating conditions (pressure 10 MPa and temperature 265-310{sup o}C) causing fuel oxidation, especially near the hotter regions of the fuel in the cracks. The fuel thermal conductivity will change, resulting in a change in the temperature distribution of the fuel element. This paper provides 2D r-θ plane strain solid mechanics models to simulate fuel thermal expansion, where conditions for fuel crack propagation are investigated with the thermal J integral to predict fuel crack stress intensity factors. Finally since fuel crack geometry can affect fuel oxidation this paper shows that the solid mechanics model with pre-set radial cracks can be coupled to a 2D r-θ fuel oxidation model. (author)

Statistical methods in the mechanical design of fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Radsak, C.; Streit, D.; Muench, C.J. [AREVA NP GmbH, Erlangen (Germany)

2013-07-01

The mechanical design of a fuel assembly is still being mainly performed in a de terministic way. This conservative approach is however not suitable to provide a realistic quantification of the design margins with respect to licensing criter ia for more and more demanding operating conditions (power upgrades, burnup increase,..). This quantification can be provided by statistical methods utilizing all available information (e.g. from manufacturing, experience feedback etc.) of the topic under consideration. During optimization e.g. of the holddown system certain objectives in the mechanical design of a fuel assembly (FA) can contradict each other, such as sufficient holddown forces enough to prevent fuel assembly lift-off and reducing the holddown forces to minimize axial loads on the fuel assembly structure to ensure no negative effect on the control rod movement.By u sing a statistical method the fuel assembly design can be optimized much better with respect to these objectives than it would be possible based on a deterministic approach. This leads to a more realistic assessment and safer way of operating fuel assemblies. Statistical models are defined on the one hand by the quanti le that has to be maintained concerning the design limit requirements (e.g. one FA quantile) and on the other hand by the confidence level which has to be met. Using the above example of the holddown force, a feasible quantile can be define d based on the requirement that less than one fuel assembly (quantile > 192/19 3 [%] = 99.5 %) in the core violates the holddown force limit w ith a confidence of 95%. (orig.)

Preliminary evaluation of fuel oil produced from pyrolysis of waste ...

African Journals Online (AJOL)

It could be refined further to produce domestic kerosene and gasoline. The physical and structural properties of the fuel oil produced compared favorably with that of Aviation fuel JP-4 (a wide-cut US Air force fuel). Presently African countries are importing aviation fuels. The fuel oil produced from the pyrolysis of waste water ...

Signatures of Extended Storage of Used Nuclear Fuel Comprehensive Final Report

Energy Technology Data Exchange (ETDEWEB)

Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-09-21

This report serves as a comprehensive overview of the Extended Storage of Used Nuclear Fuel work performed for the Material Protection, Accounting and Control Technologies campaign under the Department of Energy Office of Nuclear Energy. This paper describes a signature based on the source and fissile material distribution found within a population of used fuel assemblies combined with the neutron absorbers found within cask design that is unique to a specific cask with its specific arrangement of fuel. The paper describes all the steps used in producing and analyzing this signature from the beginning to the project end.

Jet Fuel Based High Pressure Solid Oxide Fuel Cell System

Science.gov (United States)

Gummalla, Mallika (Inventor); Yamanis, Jean (Inventor); Olsommer, Benoit (Inventor); Dardas, Zissis (Inventor); Bayt, Robert (Inventor); Srinivasan, Hari (Inventor); Dasgupta, Arindam (Inventor); Hardin, Larry (Inventor)

2015-01-01

A power system for an aircraft includes a solid oxide fuel cell system which generates electric power for the aircraft and an exhaust stream; and a heat exchanger for transferring heat from the exhaust stream of the solid oxide fuel cell to a heat requiring system or component of the aircraft. The heat can be transferred to fuel for the primary engine of the aircraft. Further, the same fuel can be used to power both the primary engine and the SOFC. A heat exchanger is positioned to cool reformate before feeding to the fuel cell. SOFC exhaust is treated and used as inerting gas. Finally, oxidant to the SOFC can be obtained from the aircraft cabin, or exterior, or both.

Biomass-derived Syngas Utilization for Fuels and Chemicals - Final Report

Energy Technology Data Exchange (ETDEWEB)

Dayton, David C

2010-03-24

Executive Summary The growing gap between petroleum production and demand, mounting environmental concerns, and increasing fuel prices have stimulated intense interest in research and development (R&D) of alternative fuels, both synthetic and bio-derived. Currently, the most technically defined thermochemical route for producing alternative fuels from lignocellulosic biomass involves gasification/reforming of biomass to produce syngas (carbon monoxide [CO] + hydrogen [H2]), followed by syngas cleaning, Fischer-Tropsch synthesis (FTS) or mixed alcohol synthesis, and some product upgrading via hydroprocessing or separation. A detailed techno-economic analysis of this type of process has recently been published [1] and it highlights the need for technical breakthroughs and technology demonstration for gas cleanup and fuel synthesis. The latter two technical barrier areas contribute 40% of the total thermochemical ethanol cost and 70% of the production cost, if feedstock costs are factored out. Developing and validating technologies that reduce the capital and operating costs of these unit operations will greatly reduce the risk for commercializing integrated biomass gasification/fuel synthesis processes for biofuel production. The objective of this project is to develop and demonstrate new catalysts and catalytic processes that can efficiently convert biomass-derived syngas into diesel fuel and C2-C4 alcohols. The goal is to improve the economics of the processes by improving the catalytic activity and product selectivity, which could lead to commercialization. The project was divided into 4 tasks: Task 1: Reactor Systems: Construction of three reactor systems was a project milestone. Construction of a fixed-bed microreactor (FBR), a continuous stirred tank reactor (CSTR), and a slurry bubble column reactor (SBCR) were completed to meet this milestone. Task 2: Iron Fischer-Tropsch (FT) Catalyst: An attrition resistant iron FT catalyst will be developed and tested

Fuel charging machine

International Nuclear Information System (INIS)

Uchikawa, Sadao.

1978-01-01

Purpose: To enable continuous fuel discharging and charging steps in a bwr type reactor by effecting positioning only for once by providing a plurality of fuel assembly grippers and their drives co-axially on a rotatable surface. Constitution: A plurality of fuel assembly grippers and their drives are provided co-axially on a rotatable surface. For example, a gripper A, a drive B, a gripper C and a drive D are arranged co-axially in symmetric positions on a disk rotated on rails by wheels and rotational drives. A new fuel in a fuel pool is gripped by the gripper A and transported above the reactor core. Then, the disk is positioned so that the gripper C can grip the spent fuel in the core, and the fuel to be discharged is gripped and raised by the gripper C. Then the disk is rotated by 180 0 and the new fuel in the gripper A is charged into the position from which the old fuel has been discharged and, finally, the discharged fuel is sent to the fuel pool for storage. (Seki, T.)

Technology development goals for automotive fuel cell power systems. Final report

Energy Technology Data Exchange (ETDEWEB)

James, B.D.; Baum, G.N.; Kuhn, I.F. Jr. [Directed Technologies, Inc., Arlington, VA (United States)

1994-08-01

This report determines cost and performance requirements for Proton Exchange Membrane (PEM) fuel cell vehicles carrying pure H{sub 2} fuel, to achieve parity with internal combustion engine (ICE) vehicles. A conceptual design of a near term FCEV (fuel cell electric vehicle) is presented. Complete power system weight and cost breakdowns are presented for baseline design. Near term FCEV power system weight is 6% higher than ICE system, mid-term FCEV projected weights are 29% lower than ICE`s. There are no inherently high-cost components in FCE, and at automotive production volumes, near term FCEV cost viability is closer at hand than at first thought. PEM current vs voltage performance is presented for leading PEM manufacturers and researchers. 5 current and proposed onboard hydrogen storage techniques are critically compared: pressurized gas, cryogenic liquid, combined pressurized/cryogenic, rechargeable hydride, adsorption. Battery, capacitor, and motor/controller performance is summarized. Fuel cell power system component weight and cost densities (threshold and goal) are tabulated.

Development of oxygen sensing technology in an irradiated fuel rod. Characteristic test of oxygen sensor

International Nuclear Information System (INIS)

Saito, Junichi; Hoshiya, Taiji; Sakurai, Fumio; Sakai, Haruyuki

1996-03-01

At the Department of JMTR (Japan Materials Test Reactor), the re-instrumentation technologies to a high burnup fuel rod irradiated in an LWR have been developed to study irradiation behavior of the fuel during power transient. It has been progressed developing a chemical sensor as one of the re-instrumentation technologies. This report summarizes the results of characteristic tests of an oxygen sensor made of Yttria Stabilized Zirconia (YSZ) as a solid electrolyte. Several kinds of experiments were carried out to evaluate the electromotive force (emf) performance, stability and lifetime of the oxygen sensor with Ni/NiO, Cr/Cr 2 O 3 and Fe/FeO, respectively as a reference electrode. From the experimental data, it is suggested that the reference electrode of Ni/NiO reveals the most appropriate characteristic of the sensor to measure the partial oxygen pressure in a fuel rod. It is the final goal of this development to clarify the change of oxygen chemical potential in a fuel rod during power transient. (author)

Alternative Fuel News: Official Publication of the Clean Cities Network and the Alternative Fuels Data Center, Vol. 5, No. 3

Energy Technology Data Exchange (ETDEWEB)

2001-11-01

A quarterly magazine with articles on alternative fuel school buses, the market growth of biodiesel fuel, National AFV Day 2002, model year 2002 alternative fuel passenger cars and light trucks, the Michelin Challenge Bibendum road rally, and advanced technology vehicles at Robins Air Force Base, the Top Ten Clean Cities coalitions for 2000, and AFVs on college campuses.

Conversion of hydrocarbons and alcohols for fuel cells

Science.gov (United States)

Joensen, Finn; Rostrup-Nielsen, Jens R.

The growing demand for clean and efficient energy systems is the driving force in the development of fuel processing technology for providing hydrogen or hydrogen-containing gaseous fuels for power generation in fuel cells. Successful development of low cost, efficient fuel processing systems will be critical to the commercialisation of this technology. This article reviews various reforming technologies available for the generation of such fuels from hydrocarbons and alcohols. It also briefly addresses the issue of carbon monoxide clean-up and the question of selecting the appropriate fuel(s) for small/medium scale fuel processors for stationary and automotive applications.

Nuclear-powered pacemaker fuel cladding study

International Nuclear Information System (INIS)

Shoup, R.L.

1976-01-01

The composite of metals and alloys used in the fabrication of 238 Pu cardiac pacemaker fuel capsules resists the effects of high temperatures, high mechanical forces, and chemical corrosives and provides more than adequate protection to the fuel pellet even from deliberate attempts to dissolve the cladding in inorganic acids. This does not imply that opening a pacemaker fuel capsule by inorganic acids is impossible but that it would not be a wise choice

BWR fuel assembly having fuel rod spacers axially positioned by exterior springs

International Nuclear Information System (INIS)

Taleyarkhan, R.P.

1988-01-01

In a fuel assembly having spaced fuel rods, an outer hollow tubular flow channel surrounding the fuel rods so as to direct flow of coolant/moderator fluid there-along, and at least one spacer being disposed along the channel and about the fuel rods so as to maintain them in side-by-side spaced relationship, an arrangement for disposing the spacer in a desired axial position along the fuel rods is described comprising: yieldably resilient springs disposed between an interior side of the outer channel and an exterior side of the spacer. The springs have an inherent spring bias directed away from the exterior sides of the spacers and toward the interior side of the channel such that by contact with the channel and spacer the springs assume states in which they are deflected away from the channel interior side so as to exert sufficient compressive contacting force thereon to maintain the spacer substantially stationary in the desired axial position along the fuel rods

Thin film battery/fuel cell power generating system. Final report, Task E-4, April 1976-April 1978

Energy Technology Data Exchange (ETDEWEB)

Feduska, W.

1978-03-31

A two-year researth program to design and demonstrate the technical feasibility of a high-temperature solid-electrolyte fuel cell is described in detail. A rare-earth chromite, in particular, La /sub 95/Mg /sub 05/Cr /sub 75/Al /sub 25/0/sub 3/ was identified, synthesized by RF-sputtering tested for resistivity, thermal expansion and inertness in contact with yttria-stabilized zirconia, and was found promising as a candidate interconnection material. Films of these interconnection materials have been successfully deposited onto stabilized zirconia tubes by electrochemical vapor deposition (EVD) and the technique has been used to fabricate such films in building fuel cell stacks. Tin-doped indium oxide and antimony-doped tin oxide air electrode current collector materials have been successfully (CVD) chemically vapor deposited, as thin films, onto zirconia tubes. Fabrication procedures for the preparation of thin films of the nickel-cermet fuel electrode and yttria-stabilized zirconia solid electrolyte have been re-verified and improved for use in preparing unit cells and cell stacks on the program. An in-house extrusion technology for porous calcia-stabilized zirconia tubes has been developed and has been used to provide suitable support tubes for component combination samples, unit cell and cell stack sample preparation. Test concepts for component combinations and for unit cells and cell stacks have been evolved, particularly, the crossed electrode technique, and test equipment has been designed, built and used to evaluate fuel cell components and their interfaces. A five-cell fuel cell stack has been fabricated and operated for 700 hours at 200 mA/cm/sup 2/ at 950 to 980/sup 0/C and was subjected to three temperature cycles during the testing. Three series connected cells of this five cell stack met the 80% voltage efficiency final target objective of the program (less than 10% voltage degradation in 700 hours - with only 300 hours required.)

Fueling method in LMFBR type reactors

International Nuclear Information System (INIS)

Kawashima, Katsuyuki; Inoue, Kotaro.

1985-01-01

Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

Development of Techniques for Spent Fuel Assay – Differential Dieaway Final Report

Energy Technology Data Exchange (ETDEWEB)

Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Goodsell, Alison [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Ianakiev, Kiril Dimitrov [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Iliev, Metodi [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Desimone, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rael, Carlos D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzl, Vladimir [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Polk, Paul John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-07-28

This report summarizes the work done under a DNDO R&D funded project on the development of the differential dieaway method to measure plutonium in spent fuel. There are large amounts of plutonium that are contained in spent fuel assemblies, and currently there is no way to make quantitative non-destructive assay. This has led NA24 under the Next Generation Safeguards Initiative (NGSI) to establish a multi-year program to investigate, develop and implement measurement techniques for spent fuel. The techniques which are being experimentally tested by the existing NGSI project do not include any pulsed neutron active techniques. The present work covers the active neutron differential dieaway technique and has advanced the state of knowledge of this technique as well as produced a design for a practical active neutron interrogation instrument for spent fuel. Monte Carlo results from the NGSI effort show that much higher accuracy (1-2%) for the Pu content in spent fuel assemblies can be obtained with active neutron interrogation techniques than passive techniques, and this would allow their use for nuclear material accountancy independently of any information from the operator. The main purpose of this work was to develop an active neutron interrogation technique for spent nuclear fuel.

Storage device for a long nuclear reactor fuel element and/or a long nuclear reactor fuel element part

International Nuclear Information System (INIS)

Vogt, M.; Schoenwitz, H.P.; Dassbach, W.

1986-01-01

The storage device can be erected in a dry storage room for new fuel elements and also in a storage pond for irradiated fuel elements. It consists of shells, which are arranged vertically and which have a lid. A suspension for the fuel element is provided on the underside of the lid, which acts as a support against squashing or bending in case of vertical forces acting (earthquake). (DG) [de

Safeguards aspects for future fuel management alternatives

International Nuclear Information System (INIS)

Richter, B.; Stein, G.; Gerstler, R.

1987-01-01

In the future, more flexible fuel management strategies will be realized in light-water reactor power stations. The incentives for this development are based on considerations related to safe and economic plant operation, e.g. improved fuel strategies can save fuel resources and waste management efforts. A further important aspect of the nuclear fuel cycle deals with recycling strategies. At the back-end of the fuel cycle, the direct final disposal of spent fuel will have to be assessed as an alternative to recycling strategies. These major development fields will also have consequences for international safeguards. In particular, reactor fuel strategies may involve higher burn-up, conditioning of spent fuel directly in the power plant, gadolinium-poisoned fuel and different levels of enrichment. These strategies will have an impact on inspection activities, especially on the applicability of NDA techniques. The inspection frequency could also be affected in recycling strategies using MOX fuel. There may be problems with NDA methods if reprocessed feed is used in enrichment plants. On the other hand, the direct final disposal of spent fuel will raise safeguards problems regarding design verification, long-term safeguarding and the very feasibility of inaccessible nuclear material

Seismic behaviour of fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Song, Heuy Gap; Jhung, Myung Jo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-11-01

A general approach for the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced from earthquake. The dynamic responses such as fuel assembly shear force, bending moment and displacement, and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed. (Author) 9 refs., 24 figs., 1 tab.

Nuclear fuel element

International Nuclear Information System (INIS)

Yamamoto, Seigoro.

1994-01-01

Ultrafine particles of a thermal neutron absorber showing ultraplasticity is dispersed in oxide ceramic fuels by more than 1% to 10% or lower. The ultrafine particles of the thermal neutron absorber showing ultrafine plasticity is selected from any one of ZrGd, HfEu, HfY, HfGd, ZrEu, and ZrY. The thermal neutron absorber is converted into ultrafine particles and solid-solubilized in a nuclear fuel pellet, so that the dispersion thereof into nuclear fuels is made uniform and an absorbing performance of the thermal neutrons is also made uniform. Moreover, the characteristics thereof, for example, physical properties such as expansion coefficient and thermal conductivity of the nuclear fuels are also improved. The neutron absorber, such as ZrGd or the like, can provide plasticity of nuclear fuels, if it is mixed into the nuclear fuels for showing the plasticity. The nuclear fuel pellets are deformed like an hour glass as burning, but, since the end portion thereof is deformed plastically within a range of a repulsive force of the cladding tube, there is no worry of damaging a portion of the cladding tube. (N.H.)

GAT 3 - fuel cells and their management (PACoGES). Progress report; GAT 3 - piles a combustible et leur gestion (PACoGES). Rapport final (juillet 2002 a juin 2004)

Energy Technology Data Exchange (ETDEWEB)

Lamy, C.

2005-07-01

The Topic Analysis Group PACoGES ('Piles a Combustible et leur Gestion') has conducted thoughts on fuel cells and their management with all the searchers concern with researches and developments on fuel cells and in particular on solid oxide fuel cells (SOFC, ITSOFC) running at high temperature (600 to 1000 C). This has concerned about 200 searchers working in about fifty laboratories (CNRS, CEA, EDF, GDF, INRETS, CNAM, Armines, and several industrial teams). Here is given the final report 2002-2004 concerning all the researches carried out by this Group. (O.M.)

Performance of a diesel engine operating on raw coal-diesel fuel and solvent refined coal-diesel fuel slurries. Final report

Energy Technology Data Exchange (ETDEWEB)

Marshall, H.P.

1980-03-01

Performance tests using an 11 kW single cylinder diesel engine were made to determine the effects of three different micronized coal-fuel oil slurries being considered as alternative fuels. Slurries containing 20, 32, and 40%-wt micronized raw coal in No. 2 fuel oil were used. Results are presented indicating the changes in the concentrations of SO/sub X/ and NO/sub X/ in the exhaust, exhaust opacity, power and efficiency, and in wear rates relative to operation on fuel oil No. 2. The engine was operated for 10 h at full load and 1400 rpm on al fuels except the 40%-wt slurry. This test was discontinued because of extremely poor performance.

Regional environmental impacts of methanol-fueled vehicles. Final report

International Nuclear Information System (INIS)

Belian, T.; Morris, R.E.; Ligocki, M.P.; Whitten, G.Z.

1991-01-01

The objectives of the study were to obtain, through simulation modeling, preliminary estimates of the regional environmental impacts methanol-fueled vehicles and to estimate the sensitivity of the model to important parameters and assumptions that affect the calculation of the impacts. The regional environmental effects of the use of M85 fuel (85 percent methanol and 15 percent gasoline) and M100 (neat methanol) relative to gasoline (an indoline blend) were estimated using a Lagrangian (trajectory) acid deposition model. The Comprehensive Chemistry Acid Deposition Model (CCADM), contains a detailed treatment of gas-phase and aqueous-phase chemistry and associated mass transfer, but provides for a less comprehensive representation of advection and diffusion. Two different meteorological regimes were analyzed: clear sky conditions and cloudy skies with a rain event. The study also included a review of gas- and aqueous-phase chemistry, with particular emphasis on methanol. The CCADM chemical mechanism was updated to include state-of-the-science (as of 1990) gas- and aqueous-phase chemistry including methanol chemistry. The CCADM was then used to analyze the regional environmental impacts from the use of methanol fuels. In performing such an analysis it was necessary to make several assumptions. The sensitivity of the analysis was examined through a series of simulations that varied key input parameters within their ranges of uncertainty

FFTF fuel failure detection and characterization by cover gas monitoring. Final report

International Nuclear Information System (INIS)

Miller, W.C.; Holt, F.E.

1977-01-01

The Fast Flux Test Facility (FFTF) will include a Fuel Failure Monitoring (FFM) System designed to detect, characterize, and locate fuel and absorber pin failures (i.e., cladding breaches) using a combination of delayed neutron detection, cover gas radioisotope monitoring, and gas tagging. During the past several years the Hanford Engineering Development Laboratory has been involved in the development, design, procurement, and installation of this integrated system. The paper describes one portion of the FFM System, the Cover Gas Monitoring System (CGMS), which has the primary function of fuel failure detection and characterization in the FFTF. By monitoring the various radioisotopes in the cover gas, the CGMS will both detect fuel and absorber pin failures and characterize those failures as to magnitude and severity

Annex 34 : task 1 : analysis of biodiesel options : biomass-derived diesel fuels : final report

Energy Technology Data Exchange (ETDEWEB)

McGill, R [Oak Ridge National Laboratory, TN (United States); Aakko-Saksa, P; Nylund, N O [TransEnergy Consulting Ltd., Helsinki (Finland)

2009-06-15

Biofuels are derived from woody biomass, non-woody biomass, and organic wastes. The properties of vegetable oil feedstocks can have profound effects on the properties of the finished biodiesel product. However, all biodiesel fuels have beneficial effects on engine emissions. This report discussed the use of biodiesel fuels as replacements for part of the diesel fuel consumed throughout the world. Biodiesel fuels currently being produced from fatty acid esters today were reviewed, as well as some of the more advanced diesel replacement fuels. The report was produced as part of the International Energy Agency (IEA) Advanced Motor Fuels (AMF) Implementing Agreement Annex 34, and was divided into 14 sections: (1) an introduction, (2) biodiesel and biomass, (3) an explanation of biodiesel, (4) properties of finished biodiesel fuels, (5) exhaust emissions of finished biodiesel fuels and blends, (6) life-cycle emissions and energy, (7) international biodiesel (FAME) technical standards and specifications, (8) growth in production and use of biodiesel fuels, (9) biofuel refineries, (10) process technology, (11) development and status of biorefineries, (12) comparison of options to produce biobased diesel fuels, (13) barriers and gaps in knowledge, and (14) references. 113 refs., 37 tabs., 74 figs.

Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

Energy Technology Data Exchange (ETDEWEB)

Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

International Nuclear Information System (INIS)

Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung

2016-01-01

In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

Fuel Thermo-physical Characterization Project. Fiscal Year 2014 Final Report

Energy Technology Data Exchange (ETDEWEB)

Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Edwards, Matthew K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); MacFarlan, Paul J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pool, Karl N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Slonecker, Bruce D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Frances N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Steen, Franciska H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-03-15

The Office of Material Management and Minimization (M3) Reactor Conversion Fuel Thermo-Physical Characterization Project at Pacific Northwest National Laboratory (PNNL) was tasked with using PNNL facilities and processes to receive irradiated low enriched uranium–molybdenum (LEU-Mo) fuel plate samples and perform analysis in support of the M3 Reactor Conversion Program. This work is in support of the M3 Reactor Conversion Fuel Development Pillar that is managed by Idaho National Laboratory. The primary research scope was to determine the thermo-physical properties as a function of temperature and burnup. Work conducted in Fiscal Year (FY) 2014 complemented measurements performed in FY 2013 on four additional irradiated LEU-Mo fuel plate samples. Specifically, the work in FY 2014 investigated the influence of different processing methods on thermal property behavior, the absence of aluminum alloy cladding on thermal property behavior for additional model validation, and the influence of higher operating surface heat flux / more aggressive irradiation conditions on thermal property behavior. The model developed in FY 2013 and refined in FY 2014 to extract thermal properties of the U-Mo alloy from the measurements conducted on an integral fuel plate sample (i.e., U-Mo alloy with a thin Zr coating and clad in AA6061) continues to perform very well. Measurements conducted in FY 2014 on samples irradiated under similar conditions compare well to measurements performed in FY 2013. In general, there is no gross influence of fabrication method on thermal property behavior, although the difference in LEU-Mo foil microstructure does have a noticeable influence on recrystallization of grains during irradiation. Samples irradiated under more aggressive irradiation conditions, e.g., higher surface heat flux, revealed lower thermal conductivity when compared to samples irradiated at moderate surface heat fluxes, with the exception of one sample. This report documents thermal

Dresden 1 plutonium recycle program. Final report

International Nuclear Information System (INIS)

Bresnick, S.D.

1980-01-01

This is the final report on the Dresden 1 Plutonium Recycle Demonstration Program. It covers the work performed from July 1, 1978 to completion, which includes in-pool inspection of two fuel assemblies, removal of two fuel rods, and post-irradiation examination (PIE) of six fuel rods. Appendix A describes the inspection and rod removal operations, and Appendix B describes the PIE work

French programme for HTR fuel

International Nuclear Information System (INIS)

Gillet, R.M.

1991-01-01

It is reported that in the frameworks of the French HTR research program, stopped in 1979 the HTR coated particle fuel, fuel rod and prismatic fuel element design have been successfully developed and irradiation tested in France and specific examination methods for irradiated fuel particles, rods and graphite blocks have been developed. Currently CEA is involved in fission product transport experiments sponsored by the US Department of Energy and performed in the COMEDIE loop. Finally the CEA follows progress and developments in HTR fuel research and development throughout the world. 1 tab

Practice and prospect of advanced fuel management and fuel technology application in PWR in China

International Nuclear Information System (INIS)

Xiao Min; Zhang Hong; Ma Cang; Bai Chengfei; Zhou Zhou; Wang Lei; Xiao Xiaojun

2015-01-01

Since Daya Bay nuclear power plant implemented 18-month refueling strategy in 2001, China has completed a series of innovative fuel management and fuel technology projects, including the Ling Ao Advanced Fuel Management (AFM) project (high-burnup quarter core refueling) and the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. First, this paper gives brief introduction to China's advanced fuel management and fuel technology experience. Second, it introduces practices of the advanced fuel management in China in detail, which mainly focuses on the implementation and progress of the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. Finally, the paper introduces the practices of advanced fuel technology in China and gives the outlook of the future advanced fuel management and fuel technology in this field. (author)

Solid oxide fuel cell systems development. Final report

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-12-15

The main objective in this project has been to develop a generic and dynamic tool for SOFC systems simulation and development. Developing integrated fuel cell systems is very expensive and therefore having the right tools to reduce the development cost and time to market for products becomes an important feature. The tools developed in this project cover a wide range of needs in Dantherm Power, R and D, and can be divided into 3 categories: 1. Component selection modeling; to define component specification requirements and selection of suppliers. 2. Application simulation model built from scratch, which can simulate the interface between customer demand and system output and show operation behavior for different control settings. 3. System operation strategy optimization with respect to operation cost and customer benefits. a. Allows to see how system size, in terms of electricity and heat output, and operation strategy influences a specific business case. b. Gives a clear overview of how a different property, in the system, affects the economics (e.g. lifetime, electrical and thermal efficiency, fuel cost sensitivity, country of deployment etc.). The main idea behind the structure of the tool being separated into 3 layers is to be able to service different requirements, from changing stakeholders. One of the major findings in this project has been related to thermal integration between the existing installation in a private household and the fuel cell system. For a normal family requiring 4500 kWh of electricity a year, along with the possibility of only running the system during the heating season (winter), the heat storage demand is only 210kWh of heat with an approximate value of Dkr 160,- in extra gas consumption. In this case, it would be much more cost effective to dump the heat, in the house, and save the expense of adding heat storage to the system. This operation strategy is only valid in Denmark for the time being, since the feed-In-Tariff allows for a

Optimum Performance of Direct Hydrogen Hybrid Fuel Cell Vehicles

OpenAIRE

Zhao, Hengbing; Burke, Andy

2009-01-01

Proton Exchange Membrane fuel cell (PEMFC) technology is one of the most attractive candidates for transportation applications due to its inherently high efficiency and high power density. However, the fuel cell system efficiency can suffer because of the need for forced air supply and water-cooling systems. Hence the operating strategy of the fuel cell system can have a significant impact on the fuel cell system efficiency and thus vehicle fuel economy. The key issues are how the fuel cell b...

Scratch direction and threshold force in nanoscale scratching using atomic force microscopes

Energy Technology Data Exchange (ETDEWEB)

Tseng, Ampere A., E-mail: ampere.tseng@asu.edu [Department of Materials Science and Engineering, National Taiwan University of Science and Technology, Taipei 106, Taiwan (China); Kuo, Chung-Feng Jeffrey; Jou, Shyankay [Department of Materials Science and Engineering, National Taiwan University of Science and Technology, Taipei 106, Taiwan (China); Nishimura, Shinya; Shirakashi, Jun-ichi [Department of Electrical and Electronic Engineering, Tokyo University of Agriculture and Technology, Tokyo 184-8588 (Japan)

2011-09-01

The nanoscaled tip in an AFM (atomic force microscope) has become an effective scratching tool for material removing in nanofabrication. In this article, the characteristics of using a diamond-coated pyramidal tip to scratch Ni-Fe thin film surfaces was experimentally investigated with the focus on the evaluation of the influence of the scratch or scan direction on the final shape of the scratched geometry as well as the applied scratch force. Results indicated that both the scratched profile and the scratch force were greatly affected by the scratch direction. It has been found that, to minimize the formation of protuberances along the groove sides and to have a better control of the scratched geometry, the tip face should be perpendicular to the scratching direction, which is also known as orthogonal cutting condition. To demonstrate the present findings, three groove patterns have been scratched with the tip face perpendicular to the scratching direction and very little amount of protuberances was observed. The threshold scratch force was also predicted based on the Hertz contact theory. Without considering the surface friction and adhesive forces between the tip and substrate, the threshold force predicted was twice smaller than the measurement value. Finally, recommendations for technical improvement and research focuses are provided.

Scratch direction and threshold force in nanoscale scratching using atomic force microscopes

International Nuclear Information System (INIS)

Tseng, Ampere A.; Kuo, Chung-Feng Jeffrey; Jou, Shyankay; Nishimura, Shinya; Shirakashi, Jun-ichi

2011-01-01

The nanoscaled tip in an AFM (atomic force microscope) has become an effective scratching tool for material removing in nanofabrication. In this article, the characteristics of using a diamond-coated pyramidal tip to scratch Ni-Fe thin film surfaces was experimentally investigated with the focus on the evaluation of the influence of the scratch or scan direction on the final shape of the scratched geometry as well as the applied scratch force. Results indicated that both the scratched profile and the scratch force were greatly affected by the scratch direction. It has been found that, to minimize the formation of protuberances along the groove sides and to have a better control of the scratched geometry, the tip face should be perpendicular to the scratching direction, which is also known as orthogonal cutting condition. To demonstrate the present findings, three groove patterns have been scratched with the tip face perpendicular to the scratching direction and very little amount of protuberances was observed. The threshold scratch force was also predicted based on the Hertz contact theory. Without considering the surface friction and adhesive forces between the tip and substrate, the threshold force predicted was twice smaller than the measurement value. Finally, recommendations for technical improvement and research focuses are provided.

Ansaldo programs on fuel cell vehicles

Energy Technology Data Exchange (ETDEWEB)

Marcenaro, B.G.; Federici, F. [Ansaldo Ricerche Srl, Genova (Italy)

1996-12-31

The growth in traffic and the importance of maintaining a stable ecology at the global scale, particularly with regard to atmospheric pollution, raises the necessity to realize a new generation of vehicles which are more efficient, more economical and compatible with the environment. At European level, the Car of Tomorrow task force has identified fuel cells as a promising alternative propulsion system. Ansaldo Ricerche has been involved in the development of fuel cell vehicles since the early nineties. Current ongoing programs relates to: (1) Fuel cell bus demonstrator (EQHEPP BUS) Test in 1996 (2) Fuel cell boat demonstrator (EQHHPP BOAT) Test in 1997 (3) Fuel cell passenger car prototype (FEVER) Test in 1997 (4) 2nd generation Fuel cell bus (FCBUS) 1996-1999 (5) 2nd generation Fuel cell passenger car (HYDRO-GEN) 1996-1999.

78 FR 50313 - Physical Protection of Irradiated Reactor Fuel in Transit

Science.gov (United States)

2013-08-19

... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Orders; rescission. SUMMARY... the NRC published a final rule, ``Physical Protection of Irradiated Fuel in Transit,'' on May 20, 2013... of Irradiated Reactor Fuel in Transit'' (RIN 3150-AI64; NRC-2009-0163). The final rule incorporates...

Public comments and Task Force responses regarding the environmental survey of the reprocessing and waste management portions of the LWR fuel cycle

International Nuclear Information System (INIS)

1977-03-01

This document contains responses by the NRC Task Force to comments received on the report ''Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle'' (NUREG-0116). These responses are directed at all comments, inclding those received after the close of the comment period. Additional information on the environmental impacts of reprocessing and waste management which has either become available since the publication of NUREG-0116 or which adds requested clarification to the information in that document

Final Technical Report for the MIT Annular Fuel Research Project

International Nuclear Information System (INIS)

Mujid S. Kazimi; Pavel Hejzlar

2008-01-01

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

Judgement of properties and function of concrete in connection with final disposal of nuclear fuel wastes in rock

International Nuclear Information System (INIS)

Bergstroem, S.G.; Fagerlund, G.; Romben, L.

1977-06-01

This report deals with the possibility of using concrete in conjuction with the permanent storage of nuclear fuel waste in rock storage facilities. The emphasis has been placed on properties such as strength and tightness and how these may be affected by internal and external causes of destruction during a filling stage of approximately 100 years and during the final storage stage of 1 000 - 100 000 years. It is established that spontaneous structural changes, which lead to a certain increase in porosity, cannot be precluded during the filling stage and uring the final storage stage. It is deemed possible to avoid cracking during the manufacture and during the filling stage if the concrete is kept moist. The risk for cracking during the final storage stage is difficult to assess. Attempts are made to estimate the tightness of aged concrete during the various stages. The tightness during the final storage stage is difficult to assess due to the fact that the scope of the cracking cannot be estimated. Chemical attacks during the filling stage are deemed to be small and can be repaired. The risk for destruction due to radioactive radiation is extremely small. Reinforcement, if any, can be protected during the filling stage on condition that the concrete is kep saturated but all reinforcement will be destroyed during the final storage stage. By way of conclusion, a number of general views on the choice of concrete and work methods are provided. (author)

Mixing Characteristics of Strongly-Forced Jet Flames in Crossflow

Science.gov (United States)

Marr, Kevin; Clemens, Noel; Ezekoye, Ofodike

2008-11-01

The effects of high frequency, large-amplitude forcing on the characteristics of a non-premixed jet flame in crossflow (JFICF) at mean Reynolds numbers of 3,200 and 4,850 are studied experimentally. Harmonic forcing of the jet fuel results in a drastic decrease in flame length and complete suppression of soot luminosity. Visualization by planar laser Mie scattering shows that forced JFICF, similar to forced free or coflow jet flames, are characterized by ejection of high-momentum, deeply penetrating vortical structures. These structures rapidly breakdown and promote intense turbulent mixing in the near region of the jet. The rapid mixing resembles a ``one-step'' process going from a fuel rich state far in the nozzle to a well-mixed, but significantly diluted, state just a few diameters from the jet exit plane. Exhaust gas emissions measurements indicate a decrease in NOx, but increases in CO and unburned hydrocarbons with increasing forcing amplitude. Acetone PLIF measurements are used to investigate the effect of partial-premixing on these emissions findings.

Development of oxygen scavenger additives for jet fuels

Energy Technology Data Exchange (ETDEWEB)

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

1995-05-01

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

Characteristics of strongly-forced turbulent jets and non-premixed jet flames

Energy Technology Data Exchange (ETDEWEB)

Lakshminarasimhan, K.; Ezekoye, O.A. [University of Texas at Austin, Department of Mechanical Engineering, Austin, TX (United States); Clemens, N.T. [University of Texas at Austin, Department of Aerospace Engineering and Engineering Mechanics, Austin, TX (United States)

2006-10-15

Previous researchers have demonstrated that strong pulsations of the fuel flow rate can significantly reduce the flame length and luminosity of laminar/transitional non-premixed jet flames. The physical mechanisms responsible for these changes are investigated experimentally in acoustically-forced jet flows where the peak velocity fluctuations are up to eight times the mean flow velocity. Both reacting and non-reacting flows were studied and Reynolds numbers, based on the mean flow properties, ranged from 800 to 10,000 (corresponding to peak Reynolds numbers of 1,450-23,000), and forcing frequencies ranged from 290 to 1,140 Hz. Both the first and second organ-pipe resonance modes of the fuel delivery tube were excited to obtain these frequencies. An analysis of the acoustic forcing characteristics within the resonance tube is provided in order to understand the source of the high amplitude forcing. Flow visualization of jets with first resonant forcing confirms the presence of large-scale coherent vortices and strong reverse flow near the exit of the fuel tube. With second-resonant forcing, however, vortices are not emitted from the tube as they are drawn back into the fuel tube before they can fully form. Increased fine-scale turbulence is associated with both resonant cases, but particularly at second resonance. The power spectra of the velocity fluctuations for a resonantly pulsed jet show the presence of an inertial subrange indicating that the flow becomes fully turbulent even for mean-Reynolds-number jets that are nominally laminar. It is shown that these pulsed jet flows exhibit strong similarities to synthetic jets and that the Strouhal number, based on the maximum velocity at the fuel tube exit, is the dominant parameter for scaling these flows. The Strouhal number determines the downstream location where the coherent vortices breakdown, and is found to provide better collapse of flame length data (both current and previous) than other parameters that have

Dermal Exposure to Jet Fuel JP-8 Significantly Contributes to the Production of Urinary Naphthols in Fuel-Cell Maintenance Workers

OpenAIRE

Chao, Yi-Chun E.; Kupper, Lawrence L.; Serdar, Berrin; Egeghy, Peter P.; Rappaport, Stephen M.; Nylander-French, Leena A.

2005-01-01

Jet propulsion fuel 8 (JP-8) is the major jet fuel used worldwide and has been recognized as a major source of chemical exposure, both inhalation and dermal, for fuel-cell maintenance workers. We investigated the contributions of dermal and inhalation exposure to JP-8 to the total body dose of U.S. Air Force fuel-cell maintenance workers using naphthalene as a surrogate for JP-8 exposure. Dermal, breathing zone, and exhaled breath measurements of naphthalene were obtained using tape-strip sam...

Responses to comments received on the draft final report of the Secretary of Energy Advisory Board Task Force on Radioactive Waste Management

International Nuclear Information System (INIS)

1994-01-01

The Task Force solicited comments on its Draft Final Report from a variety of sources. Letters were sent to over 400 individuals who had expressed interest in the interest in the Department's radioactive waste, management programs, a notice was placed in the Federal Register, the morning session of the January 1993 meeting of the full Secretary of Energy Advisory Board was given over to discussion of the draft, and Task Force members and staff presented the effort at several professional meetings. Altogether 32 written comments were received. They are reproduced here, followed in each case by the Task Force's response to specific suggestions made to improve the draft. (The panel did not respond to comments that simply reflected policy preferences or that praised the group's effort.) With one exception, those specific suggestions are highlighted and given a letter designation from open-quotes Aclose quotes to open-quotes Zclose quotes. The Task Force's responses, written in the Fall 1993, are labeled in a like manner. For the one exception, a comments submitted by Judy Treichel, the Task Force's response is printed on copies of her annotated pages

A Hold-down Margin Assessment using Statistical Method for the PWR Fuel Assembly

International Nuclear Information System (INIS)

Jeon, S. Y.; Park, N. K.; Lee, K. S.; Kim, H. K.

2007-01-01

The hold-down springs provide an acceptable hold down force against hydraulic uplift force absorbing the length change of the fuel assembly relative to the space between the upper and lower core plates in PWR. These length changes are mainly due to the thermal expansion, irradiation growth and creep down of the fuel assemblies. There are two kinds of hold-down springs depending on the different design concept of the reactor internals of the PWR in Korea, one is a leaf-type hold down spring for Westinghouse type plants and the other is a coil-type hold-down spring for OPR1000 (Optimized Power Reactor 1000). There are four sets of hold-down springs in each fuel assembly for leaf type hold-down spring and each set of the hold-down springs consists of multiple tapered leaves to form a cantilever leaf spring set. The length, width and thickness of the spring leaves are selected to provide the desired spring constant, deflection range, and hold down force. There are four coil springs in each fuel assembly for coil-type hold-down spring. In this study, the hold-down forces and margins were calculated for the leaf-type and coil-type hold-down springs considering geometrical data of the fuel assembly and its components, length changes of the fuel assembly due to thermal expansion, irradiation growth, creep, and irradiation relaxation. The hold-down spring forces were calculated deterministically and statistically to investigate the benefit of the statistical calculation method in view of hold-down margin. The Monte-Carlo simulation method was used for the statistical hold down force calculation

Prototypical spent nuclear fuel rod consolidation equipment: Phase 2, Final design report: Volume 1, Detailed design

International Nuclear Information System (INIS)

Blissell, W.H.; Ciez, A.P.; Goedicke, F.E.; Bessko, C.

1987-01-01

This document describes the Westinghouse Final Design for the Prototypical Spent Fuel Consolidation Equipment Demonstration Project. This design represents a fully qualified, licensable, cost effective spent fuel rod consolidation system. As a result of significant concerns raised by DOE and its Technical Review Committee during the 30% Design Review, significant changes were made to the original Preliminary Design resulting from Phase I activities. These changes focused on increased automation, end fitting removal, the rod pulling process and the need to maintain the consolidation canisters as clean as possible. As a result of these changes, the new system is greatly enhanced with a much greater probability of meeting or exceeding the project functional requirements. As a result of delays in resolving cost and contractual differences, additional bench testing was not conducted during Phase II. It is however our belief that the current design exceeds the 90% confidence level required by DOE because of the confidence gained from the Phase I tests, the additional engineering detail completed and the fact that our rod pulling tool has been demonstrated in a similar application at Oconee while our ID tube cutter is a modified (mounting method only) off-the-shelf design. 7 refs., 49 figs., 36 tabs

Final disposal of radioactive waste

Directory of Open Access Journals (Sweden)

Freiesleben H.

2013-06-01

Full Text Available In this paper the origin and properties of radioactive waste as well as its classification scheme (low-level waste – LLW, intermediate-level waste – ILW, high-level waste – HLW are presented. The various options for conditioning of waste of different levels of radioactivity are reviewed. The composition, radiotoxicity and reprocessing of spent fuel and their effect on storage and options for final disposal are discussed. The current situation of final waste disposal in a selected number of countries is mentioned. Also, the role of the International Atomic Energy Agency with regard to the development and monitoring of international safety standards for both spent nuclear fuel and radioactive waste management is described.

Cermet fuel reactors

International Nuclear Information System (INIS)

Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

1987-09-01

Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

Using LMDI method to analyze the change of China's industrial CO2 emissions from final fuel use: An empirical analysis

International Nuclear Information System (INIS)

Liu Lancui; Fan Ying; Wu Gang; Wei Yiming

2007-01-01

Based on time series decomposition of the Log-Mean Divisia Index (LMDI), this paper analyzes the change of industrial carbon emissions from 36 industrial sectors in China over the period 1998-2005. The changes of industrial CO 2 emission are decomposed into carbon emissions coefficients of heat and electricity, energy intensity, industrial structural shift, industrial activity and final fuel shift. Our results clearly show that raw chemical materials and chemical products, nonmetal mineral products and smelting and pressing of ferrous metals account for 59.31% of total increased industrial CO 2 emissions. The overwhelming contributors to the change of China's industrial sectors' carbon emissions in the period 1998-2005 were the industrial activity and energy intensity; the impact of emission coefficients of heat and electricity, fuel shift and structural shift was relatively small. Over the year 1998-2002, the energy intensity change in some energy-intensive sectors decreased industrial emissions, but increased emissions over the period 2002-2005. The impact of structural shift on emissions have varied considerably over the years without showing any clear trend, and the final fuel shift increased industrial emissions because of the increase of electricity share and higher emissions coefficient. Therefore, raw chemical materials and chemical products, nonmetal mineral products and smelting and pressing of ferrous metals should be among the top priorities for enhancing energy efficiency and driving their energy intensity close to the international advanced level. To some degree, we should reduce the products waste of these sectors, mitigate the growth of demand for their products through avoiding the excessive investment highly related to these sectors, increasing imports or decreasing the export in order to avoid expanding their share in total industrial value added. However, all these should integrate economic growth to harmonize industrial development and CO 2

Alternatives to traditional transportation fuels 1993

Energy Technology Data Exchange (ETDEWEB)

1995-01-01

In recent years, gasoline and diesel fuel have accounted for about 80 percent of total transportation fuel and nearly all of the fuel used in on-road vehicles. Growing concerns about the environmental effects of fossil fuel use and the Nation`s high level of dependence on foreign oil are providing impetus for the development of replacements or alternatives for these traditional transportation fuels. (The Energy Policy Act of 1992 definitions of {open_quotes}replacement{close_quotes} and {open_quotes}alternative{close_quotes} fuels are presented in the following box.) The Alternative Motor Fuels Act of 1988, the Clean Air Act Amendments of 1990 (CAAA90) and the Energy Policy Act of 1992 (EPACT) are significant legislative forces behind the growth of replacement fuel use. Alternatives to Traditional Transportation Fuels 1993 provides the number of on-road alternative fueled vehicles in use in the United States, alternative and replacement fuel consumption, and information on greenhouse gas emissions resulting from the production, delivery, and use of replacement fuels for 1992, 1993, and 1995.

Development of spent fuel remote handling technology

International Nuclear Information System (INIS)

Yoon, Ji Sup; Park, B. S.; Park, Y. S.; Oh, S. C.; Kim, S. H.; Cho, M. W.; Hong, D. H.

1997-12-01

Since the nation's policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

Comprehensive Fuel Cycle - Community Perspective - 13093

Energy Technology Data Exchange (ETDEWEB)

McLeod, Richard V. [Savannah River Community Reuse Organization, P.O. Box 696, Aiken, SC 29802 (United States); Frazier, Timothy A. [Dickstein Shapiro LLP, 1825 Eye Street NW, Washington, DC, 20006-5403 (United States)

2013-07-01

Should a five-county region surrounding the Department of Energy's Savannah River Site ('SRS') use its assets to help provide solutions to closing the nation's nuclear fuel cycle? That question has been the focus of a local ad hoc multi-disciplinary community task force (Tier I) that has been at work in recent months outlining issues and identifying unanswered questions to determine if assuming a leadership role in closing the nuclear fuel cycle is in the community's interest. If so, what are the terms and conditions under which we the community would agree to participate? Our starting point was the President's Blue Ribbon Commission on America's Nuclear Future ('Commission') which made a total of eight (8) recommendations in its final report. There are several recommendations that are directly relevant to the Tier I group and potential efforts of the Region. These are the 'consent-based approach', the creation of an independent nuclear waste management entity funded from the existing nuclear waste fee; the 'prompt efforts to develop one or more consolidated storage facilities', and 'continued U.S. innovation in nuclear energy technology and for workforce development'. (authors)

Final Technical Report for Award DESC0011912, "Trimodal Tapping Mode Atomic Force Microscopy: Simultaneous 4D Mapping of Conservative and Dissipative Probe-Sample Interactions of Energy-Relevant Materials”

Energy Technology Data Exchange (ETDEWEB)

Solares, Santiago D. [George Washington Univ., Washington, DC (United States)

2017-09-22

The final project report covering the period 7/1/14-6/30/17 provides an overview of the technical accomplishments in the areas of (i) fundamental viscoelasticity, (ii) multifrequency atomic force microscopy, and (iii) characterization of energy-relevant materials with atomic force microscopy. A list of publications supported by the project is also provided.

Design principle of TVO's final repository and preliminary adaptation to site specific conditions

International Nuclear Information System (INIS)

Salo, J-P.; Reikkola, R.

1995-01-01

Teollisuuden Voima Oy (TVO) is responsible for the management of spent fuel produced by the Olkiluoto power plant. TVO's current programme of spent fuel management is based on the guidelines and time schedule set by the Finnish Government. TVO has studied a final disposal concept in which the spent fuel bundles are encapsulated in copper canisters and emplaced in Finnish bedrock. According to the plan the final repository for spent fuel will be in operation by 2020. TVO's updated technical plans for the disposal of spent fuel together with a performance analysis (TVO-92) were submitted to the authorities in 1992. The paper describes the design principle of TVO's final repository and preliminary adaptation of the repository to site specific conditions. (author). 10 refs., 5 figs

Final disposal of spent nuclear fuel in Sweden. Some unresolved issues and challenges in the design and implementation of the forthcoming planning and EIA processes

International Nuclear Information System (INIS)

Bjarnadottir, H.; Hilding-Rydevik, T.

2001-06-01

The aim of the study is to highlight some unresolved and challenging issues in the forthcoming approximately six year long Environmental Impact Assessment (EIA) and planning process of the long-term disposal of spent nuclear fuel in Sweden. Different international and Nordic experiences of the processes for final disposal as well as from other development of similar scope, where experiences assumed to be of importance for final disposal of nuclear waste, have been described. Furthermore, issues relating to 'good EIA practice' as well as certain aspects of planning theory have also been presented. The current Swedish situation for the planning and EIA process of the final disposal of spent nuclear fuel was also been summarized. These different 'knowledge areas' have been compared and measured against our perception of the expectations towards the forthcoming process, put forward by different Swedish actors in the field. The result is a presentation of a number of questions and identification issues that the authors consider need special attention in the design and conduction of the planning and EIA process. The study has been realized through a literature survey and followed by reading and analysis of the written material. The main focus of the literature search was on material describing planning processes, actor perspectives and EIA. Material and literature on the technical and scientific aspects of spent nuclear fuel disposal was however deliberately avoided. There is a wealth of international and Swedish literature concerning final disposal of spent nuclear fuel - concerning both technical issues and issues concerning for example public participation and risk perception. But material of a more systematic and comparative nature (relating to both empirical and theoretical issues, and to practical experiences) in relation to EIA processes and communicative planning for final disposal of spent nuclear fuel seems to be more sparsely represented. Our perception of

Interim report of the task force on energy policy

International Nuclear Information System (INIS)

2001-01-01

A Task Force was established by the Premier of British Columbia in August 2001 to draft an energy policy framework for the province. Based on best practices worldwide, and keeping in mind the specific energy needs of British Columbia, this framework aims at fostering energy development in British Columbia, in accordance with exemplary environmental practices. This interim report comprises the preliminary findings of the Task Force, and public input is sought before the final report is finalized and presented to government. The energy sector of British Columbia comprises hydroelectric power, oil, gas and coal resources. In addition, green energy and alternative energy technologies are being developed, such as wind, solar, and wave power, and hydrogen fuel cells. Industry and individual consumers are well served by the highly developed transmission and distribution systems for energy. Several strategic directions were identified by the Task Force for inclusion in the energy policy of British Columbia, to meet its full potential. They are: growth to ensure safe, reliable energy and take advantage of economic opportunities; diversification; competitiveness; industry restructuring and expansion; environmental imperative; government leadership; and community and First Nations' involvement. Some changes are also required for the continuing success of the energy sector in British Columbia: a move to fully competitive markets in the electricity system, the development of natural gas storage capacity in the Lower Mainland, additional considerations for coal use, and the development of alternative energy sources. It is expected that private capital and more energy supply will result from a fully competitive energy market, which in turn would lower energy costs. Jobs and income would increase as a result of the growth in the sector. Diversification makes good economic and environmental sense. tabs., figs

Fuel consumption organization at the Kola NPP

International Nuclear Information System (INIS)

Matveev, A.A.; Ignatenko, E.I.; Volkov, A.P.; Trofimov, B.A.

1981-01-01

Problems of using NPPs in the power systems including hydroelectric power plants and NPPs are considered on the example of the Kola power system. The methods of the WWER-440 reactor fuel loading formation, reactor power forcing, optimization of volumes and time of the NPP main equipment planned maintenance are discussed. It is concluded that the optimal methods for the WWER-440 reactor fuel loading formation are the following: reactor make-up with the lesser number of fuel assemblies with maximum designed enrichment; for the case of decreased loading energy capacity displacement of make-up fuel with 2.4% enrichment by the fuel with 3.6% enrichment when preserving the designed number of make-up fuel assemblies [ru

Water confinement effects in response of fuel assembly to faulted condition loads

International Nuclear Information System (INIS)

Shah, S.J.; Brenneman, B.; Williams, G.T.; Strumpel, J.H.

2004-01-01

It has been established by other authors that the accelerations of the water confined by the reactor core baffle plates has a significant effect on the responses of all the fuel assemblies during LOCA (loss of coolant accident) or seismic transients. This particular effect is a consequence of the water being essentially incompressible, and thus experiencing the same horizontal accelerations as the imposed baffle plate motions. These horizontal accelerations of the fluid induce lateral pressure gradients that cause horizontal buoyancy forces on any submerged structures. These forces are in the same direction as the baffle accelerations and, for certain frequencies at least, tend to reduce the relative displacements between the fuel and baffle plates. But there is another confinement effect: the imposed baffle plate velocities must also be transmitted to the water. If the fuel assembly grid strips are treated as simple hydro-foils, these horizontal velocity components change the fluid angle of attack on each strip, and thus may induce large horizontal lift forces on each grid in the same direction as the baffle plate velocity. There is a similar horizontal lift due to inclined flow over the rods when axial flow is present. These combined forces appear to reduce the relative displacements between the fuel and baffle plates for any significant axial flow velocity. Modeling this effect is very simple. It was shown in previous papers that the mechanism for the large fuel assembly damping due to axial flow may be the hydrodynamic forces on the grid strips, and that this is very well represented by discrete viscous dampers at each grid elevation. To include the imposed horizontal water velocity effects, on both the grids and rods, these dampers are simply attached to the baffle plate rather than 'ground'. The large flow-induced damping really acts in a relative reference frame rather than an inertial reference frame, and thus it becomes a flow-induced coupling between the

Study of fuel systems for LH2-fueled subsonic transport aircraft, volume 1

Science.gov (United States)

Brewer, G. D.; Morris, R. E.; Davis, G. W.; Versaw, E. F.; Cunnington, G. R., Jr.; Riple, J. C.; Baerst, C. F.; Garmong, G.

1978-01-01

Several engine concepts examined to determine a preferred design which most effectively exploits the characteristics of hydrogen fuel in aircraft tanks received major emphasis. Many candidate designs of tank structure and cryogenic insulation systems were evaluated. Designs of all major elements of the aircraft fuel system including pumps, lines, valves, regulators, and heat exchangers received attention. Selected designs of boost pumps to be mounted in the LH2 tanks, and of a high pressure pump to be mounted on the engine were defined. A final design of LH2-fueled transport aircraft was established which incorporates a preferred design of fuel system. That aircraft was then compared with a conventionally fueled counterpart designed to equivalent technology standards.

Characteristics of axial splits in failed BWR fuel rods

International Nuclear Information System (INIS)

Lysell, G.; Grigoriev, V.

2000-01-01

Secondary cladding defects in BWR fuel sometimes have the shape of long axial cracks or ''splits''. Due to the large open UO 2 surfaces exposed to the water, fission product and UO 2 release to the coolant can reach excessive levels leading to forced shut downs to remove the failed fuel rods. A number of such fuel rods have been examined in Studsvik over the last 10 years. The paper describes observations from the PIE of long cracks and discusses the driving force of the cracks. Details such as starting cracks, macroscopic and microscopic fracture surface appearance, cross sections of cracks, hydride precipitates, location and degree of plastic deformation are given. (author)

Uncertainties in forces extracted from non-contact atomic force microscopy measurements by fitting of long-range background forces

Directory of Open Access Journals (Sweden)

Adam Sweetman

2014-04-01

Full Text Available In principle, non-contact atomic force microscopy (NC-AFM now readily allows for the measurement of forces with sub-nanonewton precision on the atomic scale. In practice, however, the extraction of the often desired ‘short-range’ force from the experimental observable (frequency shift is often far from trivial. In most cases there is a significant contribution to the total tip–sample force due to non-site-specific van der Waals and electrostatic forces. Typically, the contribution from these forces must be removed before the results of the experiment can be successfully interpreted, often by comparison to density functional theory calculations. In this paper we compare the ‘on-minus-off’ method for extracting site-specific forces to a commonly used extrapolation method modelling the long-range forces using a simple power law. By examining the behaviour of the fitting method in the case of two radically different interaction potentials we show that significant uncertainties in the final extracted forces may result from use of the extrapolation method.

Nuclear reactor seismic fuel assembly grid

International Nuclear Information System (INIS)

Anthony, A.J.

1977-01-01

The strength of a nuclear reactor fuel assembly is enhanced by increasing the crush strength of the zircaloy spacer grids which locate and support the fuel elements in the fuel assembly. Increased resistance to deformation as a result of laterally directed forces is achieved by increasing the section modulus of the perimeter strip through bending the upper and lower edges thereof inwardly. The perimeter strip is further rigidized by forming, in the central portion thereof, dimples which extend inwardly with respect to the fuel assembly. The integrity of the spacer grid may also be enhanced by providing back-up arches for some or all of the integral fuel element locating springs and the strength of the fuel assembly may be further enhanced by providing, intermediate its ends, a steel seismic grid. 13 claims, 6 figures

Public comments and Task Force responses regarding the environmental survey of the reprocessing and waste management portions of the LWR fuel cycle

Energy Technology Data Exchange (ETDEWEB)

1977-03-01

This document contains responses by the NRC Task Force to comments received on the report ''Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle'' (NUREG-0116). These responses are directed at all comments, inclding those received after the close of the comment period. Additional information on the environmental impacts of reprocessing and waste management which has either become available since the publication of NUREG-0116 or which adds requested clarification to the information in that document.

Using distributed fuel cells to compete with established utilities under rules permitting retail wheeling

International Nuclear Information System (INIS)

Miller, D.A.; Mathur, A.

1996-01-01

Regulatory reform in the electricity industry clearly has many implications for the corporate structure of electric utilities, the pricing of energy products and services, and the quality of service received by customers. But it also has implications for the selection of energy generating technologies. The rise of the cogenerator and independent power producer has been a major force in the expanded use of combined-cycle power plants and their technological advancement. The next stage of increased deregulation, retail wheeling, must also lead to further technological change, because the economic climate determines the operational characteristics required of new generating resources. This paper discusses the use of distributed fuel cells to compete with established utilities in areas where retail wheeling has been instituted. It will cover in detail the unique advantages of the technology under this industry configuration. The paper will pay particular attention to the operational and design characteristics of fuel cells that will provide companies the flexibility they require to compete successfully. Finally, the paper will discuss the implications for the use of distributed fuel cells of alternative retail wheeling implementation schemes

Crude Glycerol as Cost-Effective Fuel for Combined Heat and Power to Replace Fossil Fuels, Final Technical Report

Energy Technology Data Exchange (ETDEWEB)

Roberts, William L

2012-10-31

The primary objectives of this work can be summed into two major categories. Firstly, the fundamentals of the combustion of glycerol (in both a refined and unrefined form) were to be investigated, with emphasis of the development of a system capable of reliably and repeatedly combusting glycerol as well as an analysis of the emissions produced during glycerol combustion. Focus was placed on quantifying common emissions in comparison to more traditional fuels and this work showed that the burner developed was able to completely combust glycerol within a relatively wide range of operating conditions. Additionally, focus was placed on examining specific emissions in more detail, namely interesting NOx emissions observed in initial trials, acrolein and other volatile organic emissions, and particulate and ash emissions. This work showed that the combustion of crude glycerol could result in significantly reduced NOx emissions as a function of the high fuel bound oxygen content within the glycerol fuel. It also showed that when burned properly, the combustion of crude glycerol did not result in excessive emissions of acrolein or any other VOC compared to the combustion from more traditional fuels. Lastly however, this work has shown that in any practical application in which glycerol is being burned, it will be necessary to explore ash mitigation techniques due to the very high particulate matter concentrations produced during glycerol combustion. These emissions are comparable to unfiltered coal combustion and are directly tied to the biodiesel production method. The second focus of this work was directed to developing a commercialization strategy for the use of glycerol as a fuel replacement. This strategy has identified a 30 month plan for the scaling up of the laboratory scale burner into a pre-pilot scale system. Additionally, financing options were explored and an assessment was made of the economics of replacing a traditional fuel (namely natural gas) with crude

Special equipment for the fabrication and quality control of nuclear fuel elements

International Nuclear Information System (INIS)

Guse, K.; Herbert, W.; Jaeger, K.

1989-01-01

For the fabrication of LWR fuel elements, columns are packed of up to 4 m in length, consisting of fuel pellets with different uranium enrichment, their weight and total length to be measured prior to further processing to fuel rods. An automated column packing device has been developed for this purpose. The packing jobs and other tasks are computer-controlled, measured data are stored and are printed out for quality documentation. The forces in the springs of fuel spacers of LWR fuel elements are to be measured and compared with the standard requirements, deviations to be documented. For this task, another computer-controlled, automated device has been developed for measuring the spring forces at all required positions after positioning and fixation of the spacers. (orig./DG) [de

Nuclear reactor using fuel sphere for combustion and fuel spheres for breeding

International Nuclear Information System (INIS)

Yamashita, Kiyonobu.

1995-01-01

The present invention concerns a pebble bed-type reactor which can efficiently convert parent nuclides to fission nuclides. Fuel spheres for combustion having fission nuclides as main fuels, and fuel spheres for breeding having parent nuclides as main fuels are used separately, in the pebble bed-type reactor. According to the present invention, fuel spheres for breeding can be stayed in a reactor core for a long period of time, so that parent nuclides can be sufficiently converted into fission nuclides. In addition, since fuel spheres for breeding are loaded repeatedly, the amount thereof to be used is reduced. Therefore, the amount of the fuel spheres for breeding is small even when they are re-processed. On the other hand, since the content of the fission nuclides in the fuel spheres for breeding is not great, they can be put to final storage. This is attributable that although the fuel spheres for breeding contain fission nuclides generated by conversion, the fission nuclides are annihilated by nuclear fission reactions at the same time with the generation thereof. (I.S.)

Development of spent fuel remote handling technology

Energy Technology Data Exchange (ETDEWEB)

Yoon, Ji Sup; Park, B S; Park, Y S; Oh, S C; Kim, S H; Cho, M W; Hong, D H

1997-12-01

Since the nation`s policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

Review and Response to the Final Report of the National Black Health Providers Task Force on High Blood Pressure Education and Control.

Science.gov (United States)

Public Health Service (DHHS), Rockville, MD.

This report presents the National Heart, Lung, and Blood Institute's (NHLBI) review of and response to the final report of the National Black Health Providers Task Force on High Blood Pressure Education and Control. The response includes a statement of NHLBI's involvement in health research, and descriptions of what steps can be taken to solve the…

Development of CANFLEX fuel bundle

International Nuclear Information System (INIS)

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

1991-12-01

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

LMFBR fuel analysis. Task A: oxide fuel dynamics. Final report, July 1, 1975--September 30, 1976

International Nuclear Information System (INIS)

Dhir, V.K.; Hauss, B.; Kastenberg, W.E.; Saqui, R.; Sun, Y.H.; Wong, K.

1976-11-01

The report summarizes the results of studies conducted in support of the U.S. Nuclear Regulatory Commission's review of the Preliminary Safety Analysis Report for the Clinch River Breeder Reactor. In particular it deals with three aspects of the unprotected transient overpower accident. The first aspect is the response of the Clinch River Breeder Reactor to low reactivity insertion rates. Second, the investigation of a new method for computing the time, place and mode of fuel pin failure is studied. Lastly, the question of post-failure, fuel freezing, and plate-out is addressed. Several areas of uncertainty in the analysis of these accidents is also discussed

A method for allocating renewable energy source subsidies among final energy consumers

International Nuclear Information System (INIS)

Batlle, Carlos

2011-01-01

In a new context of growing need for renewable energy sources (RES), tariff design has become a critical component of energy system regulation. A methodology for allocating the cost of RES subsidies that ensures an optimal balance between compliance with the main regulatory principles of tariff design and each state's specific policy is of cardinal importance in the current context. This paper presents and discusses a novel methodology to allocate the RES subsidy costs, which consists of distributing them among final energy consumers, in proportion to their consumption, regardless of the type of final energy consumed (liquid fuels, gas, electricity or coal). First, the different designs of RES subsidies are categorized and a review of a good number of the RES burden sharing mechanisms implemented in the EU is presented. Then, the proposed methodology is developed on the basis of the basic regulatory principles underlying the tariff design and the current regulatory context in force in the EU. Finally, to illustrate its actual impact in a real case example, the proposed methodology is applied to the Spanish system, in which the burden of extra costs incurred for RES amounts to a very large proportion of the overall energy system costs. - Research highlights: → This paper develops a novel methodology to allocate RES subsidy costs among final energy consumers. → The methodology is built upon the basic tariff design principles and the 3rd package regulations. → It consists of distributing them in proportion to liquid fuels, gas, electricity or coal consumption. → The 20% RES target affects all energy sectors and justifies allocating the extra costs accordingly. → The methodology is applied to the Spanish energy system, where the RES burden is a very significant.

Fuel compliance model for pellet-cladding mechanical interaction

International Nuclear Information System (INIS)

Shah, V.N.; Carlson, E.R.

1985-01-01

This paper describes two aspects of fuel pellet deformation that play significant roles in determining maximum cladding hoop strains during pellet-cladding mechanical interaction: compliance of fragmented fuel pellets and influence of the pellet end-face design on the transmission of axial compressive force in the fuel stack. The latter aspect affects cladding ridge formation and explains several related observations that cannot be explained by the hourglassing model. An empirical model, called the fuel compliance model and representing the above aspects of fuel deformation, has been developed using the results from two Halden experiments and incorporated into the FRAP-T6 fuel performance code

A fuel in blue

International Nuclear Information System (INIS)

Kelley, T.

1992-01-01

This paper reports that U.s. police departments are increasingly turning to natural gas for the same reasons other vehicle fleet operators do: It's much cheaper than gasoline and better for the environment - and it saves on engine wear-and-tear, too. Of course, the police have a lot of company in seeing the merits of natural gas vehicles (NGVs). Nineteen ninety-two proved itself a good year for NGVs before it was more than a few weeks old. President George Bush personally drove one outside the White House, the General Services Administration ordered 625 NGVs from Chrysler and General Motors, and both Ford and Chevrolet planned to begin this spring to manufacture a small number of vehicles equipped to run solely on natural gas. School districts are happy with the savings they reap by running school buses on natural gas, and the fuel has been embraced by other fleets ranging from garbage trucks in Brooklyn to jet-towing vehicles at Denver's Stapleton Airport. But police departments are different. A patrol care is not just transportation for an officer, but his or her workplace on wheels. And perhaps because they deal with life-and-death situations, police officers aren't eager to change anything - even a vehicle fuel - that they see as doing the job. If marketing natural gas for fleet use is a popular curriculum for gas companies these days, police forces are the demanding final exam

Fuel element

International Nuclear Information System (INIS)

Hirose, Yasuo.

1982-01-01

Purpose: To increase the plenum space in a fuel element used for a liquid metal cooled reactor. Constitution: A fuel pellet is secured at one end with an end plug and at the other with a coil spring in a tubular container. A mechanism for fixing the coil spring composed of a tubular unit is mounted by friction with the inner surface of the tubular container. Accordingly, the recoiling force of the coil spring can be retained by fixing mechanism with a small volume, and since a large amount of plenum space can be obtained, the internal pressure rise in the cladding tube can be suppressed even if large quantities of fission products are discharged. (Kamimura, M.)

Final Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel

International Nuclear Information System (INIS)

2000-01-01

DOE is responsible for the safe and efficient management of its sodium-bonded spent nuclear fuel. This fuel contains metallic sodium, a highly reactive material; metallic uranium, which is also reactive; and in some cases, highly enriched uranium. The presence of reactive materials could complicate the process of qualifying and licensing DOE's sodium-bonded spent nuclear fuel inventory for disposal in a geologic repository. Currently, more than 98 percent of this inventory is located at the Idaho National Engineering and Environmental Laboratory (INEEL), near Idaho Falls, Idaho. In addition, in a 1995 agreement with the State of Idaho, DOE committed to remove all spent nuclear fuel from Idaho by 2035. This EIS evaluates the potential environmental impacts associated with the treatment and management of sodium-bonded spent nuclear fuel in one or more facilities located at Argonne National Laboratory-West (ANL-W) at INEEL and either the F-Canyon or Building 105-L at the Savannah River Site (SRS) near Aiken, South Carolina. DOE has identified and assessed six proposed action alternatives in this EIS. These are: (1) electrometallurgical treatment of all fuel at ANL-W, (2) direct disposal of blanket fuel in high-integrity cans with the sodium removed at ANL-W, (3) plutonium-uranium extraction (PUREX) processing of blanket fuel at SRS, (4) melt and dilute processing of blanket fuel at ANL-W, (5) melt and dilute processing of blanket fuel at SRS, and (6) melt and dilute processing of all fuel at ANL-W. In addition, Alternatives 2 through 5 include the electrometallurgical treatment of driver fuel at ANL-W. Under the No Action Alternative, the EIS evaluates both the continued storage of sodium-bonded spent nuclear fuel until the development of a new treatment technology or direct disposal without treatment. Under all of the alternatives, the affected environment is primarily within 80 kilometers (50 miles) of spent nuclear fuel treatment facilities. Analyses indicate

LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

International Nuclear Information System (INIS)

Warinner, D.K.

1980-01-01

The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

Integrated scheme of long-term for spent fuel management of power nuclear reactors

International Nuclear Information System (INIS)

Ramirez S, J. R.; Palacios H, J. C.; Martinez C, E.

2015-09-01

After of irradiation of the nuclear fuel in the reactor core, is necessary to store it for their cooling in the fuel pools of the reactor. This is the first step in a processes series before the fuel can reach its final destination. Until now there are two options that are most commonly accepted for the end of the nuclear fuel cycle, one is the open nuclear fuel cycle, requiring a deep geological repository for the fuel final disposal. The other option is the fuel reprocessing to extract the plutonium and uranium as valuable materials that remaining in the spent fuel. In this study the alternatives for the final part of the fuel cycle, which involves the recycling of plutonium and the minor actinides in the same reactor that generated them are shown. The results shown that this is possible in a thermal reactor and that there are significant reductions in actinides if they are recycled into reactor fuel. (Author)

Nuclear fuel assembly seismic amplitude limiter

International Nuclear Information System (INIS)

Anthony, A.J.

1977-01-01

The ability of a nuclear reactor to withstand high seismic loading is enhanced by including, on each fuel assembly, at least one seismic grid which reduces the magnitude of the possible lateral deflection of the individual fuel elements and the entire fuel assembly. The reduction in possible deflection minimizes the possibility of impact of the spacer grids of one fuel assembly on those of an adjacent fuel assembly and reduces the magnitude of forces associated with any such impact thereby minimizing the possibility of fuel assembly damage as a result of high seismic loading. The seismic grid is mounted from the fuel assembly guide tubes, has greater external dimensions when compared to the fuel assembly spacer grids and normally does not support or otherwise contact the fuel elements. The reduction in possible deflection is achieved through reduction of the clearance between adjacent fuel assemblies made possible by the use in the seismic grid of a high strength material characterized by favorable thermal expansion characteristics and minimal irradiation induced expansion

Fuel followed control rod installation at AFRRI

International Nuclear Information System (INIS)

Moore, Mark; Owens, Chris; Forsbacka, Matt

1992-01-01

Fuel Followed Control Rods (FFCRs) were installed at the Armed Forces Radiobiology Research Institute's 1 MW TRIGA Reactor. The procedures for obtaining, shipping, and installing the FFCRs is described. As part of the FFCR installation, the transient rod drive was relocated. Core performance due to the addition of the fuel followed control rods is discussed. (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

76 FR 65382 - Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline

Science.gov (United States)

2011-10-21

... Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline AGENCY: Environmental... gasoline. This final rule will provide flexibility to the regulated community by allowing an additional... Method for Olefins in Gasoline III. Statutory and Executive Order Reviews A. Executive Order 12866...

Alcohol fuels for developing countries

International Nuclear Information System (INIS)

Bhattacharya, Partha

1993-01-01

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

LOFT instrumented fuel design and operating experience

International Nuclear Information System (INIS)

Russell, M.L.

1979-01-01

A summary description of the Loss-of-Fluid Test (LOFT) system instrumented core construction details and operating experience through reactor startup and loss-of-coolant experiment (LOCE) operations performed to date are discussed. The discussion includes details of the test instrumentation attachment to the fuel assembly, the structural response of the fuel modules to the forces generated by a double-ended break of a pressurized water reactor (PWR) coolant pipe at the inlet to the reactor vessel, the durability of the LOFT fuel and test instrumentation, and the plans for incorporation of improved fuel assembly test instrumentation features in the LOFT core

Study of the potential uses of the Barnwell Nuclear Fuel Plant (BNFP). Final report

Energy Technology Data Exchange (ETDEWEB)

1980-03-25

The purpose of this study is to provide an evaluation of possible international and domestic uses for the Barnwell Nuclear Fuel Plant, located in South Carolina, at the conclusion of the International Nuclear Fuel Cycle Evaluation. Four generic categories of use options for the Barnwell plant have been considered: storage of spent LWR fuel; reprocessing of LWR spent fuel; safeguards development and training; and non-use. Chapters are devoted to institutional options and integrated institutional-use options.

Study of the potential uses of the Barnwell Nuclear Fuel Plant (BNFP). Final report

International Nuclear Information System (INIS)

1980-01-01

The purpose of this study is to provide an evaluation of possible international and domestic uses for the Barnwell Nuclear Fuel Plant, located in South Carolina, at the conclusion of the International Nuclear Fuel Cycle Evaluation. Four generic categories of use options for the Barnwell plant have been considered: storage of spent LWR fuel; reprocessing of LWR spent fuel; safeguards development and training; and non-use. Chapters are devoted to institutional options and integrated institutional-use options

Post operation: The changing characteristics of nuclear fuel cycle costs

International Nuclear Information System (INIS)

Frank, F.J.

1986-01-01

Fundamental changes have occurred in the nuclear fuel cycle. These changes forged by market forces, legislative action, and regulatory climate appear to be a long term characteristic of the nuclear fuel cycle. The nature of these changes and the resulting emerging importance of post-operation and its impact on fuel cycle costs are examined

Growth rates of breeder reactor fuel. Final report

International Nuclear Information System (INIS)

Ott, K.O.

1979-01-01

During the contract period, a consistent formalism for the definition of the growth rates (and thus the doubling time) of breeder reactor fuel has been developed. This formalism was then extended to symbiotic operation of breeder and converter reactors. Further, an estimation prescription for the growth rate has been developed which is based upon the breeding worth factors. The characteristics of this definition have been investigated, which led to an additional integral concept, the breeding bonus

Potential health and environmental impacts attributable to the nuclear and coal fuel cycles: Final report

International Nuclear Information System (INIS)

Gotchy, R.L.

1987-06-01

Estimates of mortality and morbidity are presented based on present-day knowledge of health effects resulting from current component designs and operations of the nuclear and coal fuel cycles, and anticipated emission rates and occupational exposure for the various fuel cycle facilities expected to go into operation during the next decade. The author concluded that, although there are large uncertainties in the estimates of potential health effects, the coal fuel cycle alternative has a greater health impact on man than the uranium fuel fycle. However, the increased risk of health effects for either fuel cycle represents a very small incremental risk to the average individual in the public for the balance of this century. The potential for large impacts exists in both fuel cycles, but the potential impacts associated with a runaway Greenhouse Effect from combustion of fossil fuels, such as coal, cannot yet be reasonably quantified. Some of the potential environmental impacts of the coal fuel cycle cannot currently be realistically estimated, but those that can appear greater than those from the nuclear fuel cycle. 103 refs., 1 fig., 18 tabs

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.

High Performance Fuel Laboratory, Hanford Reservation, Richland, Washington. Final environmental impact statement

International Nuclear Information System (INIS)

1977-09-01

The High Performance Fuel Laboratory (HPFL) will provide pilot scale tests of manufacturing processes, equipment, and handling systems and of accountability and safeguards, methods, and equipment while keeping radiological and chemical exposures of the workers, public, and environment at the lowest practicable levels. The experience gained from designing, constructing and operating the HPFL can be used in future commitments to commercial fuel fabrication plants in the late 1980s and beyond for processing of nuclear fuel. The HPFL site is located in the 400 Area of the 559-square mile, federally owned Hanford Reservation. This environmental impact statement considers effects of the HPFL under normal conditions and in the event of an accident

Use of Solid Hydride Fuel for Improved long-Life LWR Core Designs. Final summary report

International Nuclear Information System (INIS)

Greenspan, E

2006-01-01

The primary objective of this project was to assess the feasibility of improving the performance of PWR and BWR cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the bus-bar cost of electricity (COE). Additional performance measures considered are safety, fuel bundle design simplicity in particular for BWR's, and plutonium incineration capability. It was found that hydride fuel can safely operate in PWR's and BWR's without restricting the linear heat generation rate of these reactors relative to that attainable with oxide fuel. A couple of promising applications of hydride fuel in PWR's and BWR's were identified: (1) Eliminating dedicated water moderator volumes in BWR cores thus enabling to significantly increase the cooled fuel rods surface area as well as the coolant flow cross section area in a given volume fuel bundle while significantly reducing the heterogeneity of BWR fuel bundles thus achieving flatter pin-by-pin power distribution. The net result is a possibility to significantly increase the core power density ? on the order of 30% and, possibly, more, while greatly simplifying the fuel bundle design. Implementation of the above modifications is, though, not straightforward; it requires a design of completely different control system that could probably be implemented only in newly designed plants. It also requires increasing the coolant pressure drop across the core. (2) Recycling plutonium in PWR's more effectively than is possible with oxide fuel by virtue of a couple of unique features of hydride fuel reduced inventory of U-238 and increased inventory of hydrogen. As a result, the hydride fueled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it incinerates in one pass is double that of the MOX fuel. The fissile fraction of the Pu in the discharged hydride fuel is only ∼2/3 that of the MOX fuel and the discharged hydride fuel is

Fuel channel closure and adapter

International Nuclear Information System (INIS)

Cashen, W.S.

1985-01-01

This invention provides a mechanical closure/actuating ram combination particularly suited for use in sealing the ends of the pressure tubes when a CANDU-type reactor is refueled. It provides a cluster that may be inserted into a fuel channel end fitting to provide at least partial closing off of a pressure tube while permitting the disengagement of the fueling machine and its withdrawal from the closure for other purposes. The invention also provides a ram/closure combination wherein the application of loading force to a deformable sealing disk is regulated by a massive load bar component forming part of the fueling machine and being therefore accessible for maintenance or replacement

Evaluation of spectral shift controlled reactors operating on the uranium fuel cycle. Final report

International Nuclear Information System (INIS)

Matzie, R.A.; Sider, F.M.

1979-08-01

The performance of the spectral shift controlled reactor (SSCR) operating on uranium fuel cycles was evaluated and compared with the conventional pressurized water reactor (PWR). In order to analyze the SSCR, the PSR design methodology was extended to include systems moderated by mixtures of light water and heavy water and these methods were validated by comparison with experimental results. Once the design methods had been formulated, the resouce requirements and power costs were determined for the uranium-fueled SSCR. The ore requirements of the UO 2 once-through fuel cycle and the UO 2 fuel cycle with self-generated recycle (SGR) of plutonium were found to be 10% and 19% less than those of similarly fueled PWRs, respectively. A fuel cycle optimization study was performed for the UO 2 once-through SSCR and the SGR SSCR. By individually altering lattice parameters, discharge exposure or number of in-core batches, savings of less than 8% in resource requirements and less than 1% in power costs were obtained

Fuel Cell Demonstration Project - 200 kW - Phosphoric Acid Fuel Cell Power Plant Located at the National Transportation Research Center: FINAL REPORT

Energy Technology Data Exchange (ETDEWEB)

Berry, JB

2005-05-06

Oak Ridge National Laboratory (ORNL) researches and develops distributed generation technology for the Department of Energy, Energy Efficiency and Renewable Energy Distributed Energy Program. This report describes installation and operation of one such distributed generation system, a United Technology Corporation fuel cell located at the National Transportation Research Center in Knoxville, Tennessee. Data collected from June 2003 to June of 2004, provides valuable insight regarding fuel cell-grid compatibility and the cost-benefit of the fuel cell operation. The NTRC fuel cell included a high-heat recovery option so that use of thermal energy improves project economics and improves system efficiency to 59% year round. During the year the fuel cell supplied a total of 834MWh to the NTRC and provided 300MBtu of hot water. Installation of the NTRC fuel cell was funded by the Distributed Energy Program with partial funding from the Department of Defense's Climate Change Fuel Cell Buy Down Program, administered by the National Energy Technology Laboratory. On-going operational expenses are funded by ORNL's utility budget and are paid from operational cost savings. Technical information and the benefit-cost of the fuel cell are both evaluated in this report and sister reports.

Fundamental and semi-global kinetic mechanisms for hydrocarbon combustion. Final report, March 1977-October 1980

Energy Technology Data Exchange (ETDEWEB)

Dryer, F L; Glassman, I; Brezinsky, K

1981-03-01

Over the past three and one half years, substantial research efforts of the Princeton Fuels Research Group have been directed towards the development of simplified mechanisms which would accurately describe the oxidation of hydrocarbons fuels. The objectives of this combustion research included the study of semi-empirical modeling (that is an overall description) of the chemical kinetic mechanisms of simple hydrocarbon fuels. Such fuels include the alkanes: ethane, propane, butane, hexane and octane as well as the critically important alkenes: ethene, propene and butene. As an extension to this work, the study of the detailed radical species characteristics of combustion systems was initiated as another major aspect of the program, with emphasis on the role of the OH and HO/sub 2/ radicals. Finally, the studies of important alternative fuel problems linked the program to longer range approaches to the energy supply question. Studies of alternative fuels composed the major elements of this area of the program. The efforts on methanol research were completed, and while the aromatics aspects of the DOE work have been a direct extension of efforts supported by the Air Force Office of Scientific Research, they represented a significant part of the overall research effort. The emphasis in the proposed program is to provide further fundamental understanding of the oxidation of hydrocarbon fuels which will be useful in guiding engineering approaches. Although the scope of program ranges from the fundamentals of chemical kinetics to that of alternative fuel combustion, the objective in mind is to provide insight and guidance to the understanding of practical combustion environments. The key to our approach has been our understanding of the fundamental combustion chemistry and its relation to the important practical combustion problems which exist in implementing energy efficient, alternate fuels technologies.

Supply Chain Based Solution to Prevent Fuel Tax Evasion: Proof of Concept Final Report

Energy Technology Data Exchange (ETDEWEB)

Capps, Gary J [ORNL; Lascurain, Mary Beth [ORNL; Franzese, Oscar [ORNL; Earl, Dennis Duncan [ORNL; West, David L [ORNL; McIntyre, Timothy J [ORNL; Chin, Shih-Miao [ORNL; Hwang, Ho-Ling [ORNL; Connatser, Raynella M [ORNL; Lewis Sr, Samuel Arthur [ORNL; Moore, Sheila A [ORNL

2011-12-01

The goal of this research was to provide a proof-of-concept (POC) system for preventing non-taxable (non-highway diesel use) or low-taxable (jet fuel) petrochemical products from being blended with taxable fuel products and preventing taxable fuel products from cross-jurisdiction evasion. The research worked to fill the need to validate the legitimacy of individual loads, offloads, and movements by integrating and validating, on a near-real-time basis, information from global positioning system (GPS), valve sensors, level sensors, and fuel-marker sensors.

Measuring deformation of Fuel pin in a Nuclear Fuel Test Rig

Energy Technology Data Exchange (ETDEWEB)

Heo, S. H.; Yang, T. H.; Hong, J. T.; Joung, C. Y.; Ahn, S. H.; Jang, S. Y.; Kim, J. H. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this study, an LVDT core for measuring the longitudinal displacement of fuel pellets and clad was designed and produced. A signal processing method for the prepared core was investigated. The Nuclear Fuel Test Rig is used to observe changes in the characteristics of the fuel according to the neutron irradiation at HANARO (High-flux Advanced Neutron Application Reactor), which is a research reactor. Which are the strain and internal temperature of the irradiated nuclear fuel and the internal pressure of fuel due to fission gas, the characteristics of the fuel are measured using various sensors such as a thermocouple, SPND and LVDT. In this study, two shaped LVDT (Linear Variable Differential Transformer) cores for displacement measurements were designed and manufactured in order to measure the displacement of a fuel pellet and cladding tube using LVDT sensors for measuring electrical signals by converting the physical variation such as the force and displacement into a linear motion. In addition, signals from the manufactured LVDT sensor were collected and calibrated. Moreover, a method for obtaining the displacement in the core according to the sensing signal was planned. A derived equation can used to predict the change in the position of core. A following study should be conducted to test the output signal and real variation of out-pile system. For further work, a performance verification is required for an in-pile irradiation test.

Cyclone reburn using coal-water fuel: Pilot-scale development and testing. Final report

Energy Technology Data Exchange (ETDEWEB)

Eckhart, C.F.; DeVault, R.F.

1991-10-01

There is an ongoing effort to develop retrofit technologies capable of converting oil- and/or gas-fired boilers to coal combustion. The objective of this project is to demonstrate the technical feasibility of an improved portion of a previously developed retrofit system designed for the purpose of converting oil/gas boilers. This improvement would almost entirely eliminate the use of premium fuels, thereby significantly increasing the economical attractiveness of the system. Specifically, the goals in this program were to replace natural gas as a reburning fuel with coal-water fuel (CWF). The advantages of such a system include: (1) increased return on investment (ROI) for conversions; (2) nearly complete elimination of premium oil or gas fuel; (3) a more integrated approach to the conversion of oil- or gas-designed boilers to CWF.

75 FR 25323 - Light-Duty Vehicle Greenhouse Gas Emission Standards and Corporate Average Fuel Economy Standards...

Science.gov (United States)

2010-05-07

... Greenhouse Gas Emission Standards and Corporate Average Fuel Economy Standards; Final Rule #0;#0;Federal... Fuel Economy Standards; Final Rule AGENCY: Environmental Protection Agency (EPA) and National Highway... reduce greenhouse gas emissions and improve fuel economy. This joint Final Rule is consistent with the...

Roof Box Shape Streamline Adaptation and the Impact towards Fuel Consumption

Directory of Open Access Journals (Sweden)

Abdul Latif M.F.

2017-01-01

Full Text Available The fuel price hike is currently a sensational national issue in Malaysia. Since the rationalization of fuel subsidies many were affected especially the middle income family. Vehicle aerodynamic were directly related to the fuel consumption, were extra frontal area result a higher drag force hence higher fuel consumption. Roof box were among the largest contributor to the extra drag, thus the roof box shape rationalization were prominent to reduce the extra drag. The idea of adopting water drop shape to the roof box design shows prominent result. The roof box has been simulated using MIRA virtual wind tunnel modelling via commercial computational fluid dynamic (CFD package. This streamline shape drastically reduce the drag force by 34% resulting to a 1.7% fuel saving compare to the conventional boxy roof box. This is an effort to reduce the carbon foot print for a sustainable green world.

Investigation of a Shock Absorber for Safeguard of Fuel Assemblies Failure

International Nuclear Information System (INIS)

Karalevicius, Renatas; Dundulis, Gintautas; Rimkevicius, Sigitas; Uspuras, Eugenijus

2006-01-01

The Ignalina NPP has two reactors. The Unit 1 was shut down, therefore the special equipment was designed for transportation of the fuel from Unit 1 to Unit 2. The fuel-loaded basket can drop during transportation. The special shock absorber was designed in order to avoid failure of fuel assemblies during transportation. In case of drop of fuel loaded basket, the failure of fuel assemblies can occur. This shock absorber was studied by scaled experiments at Lithuanian Energy Institute. Static and dynamic investigations of shock absorber are presented in this paper, including dependency of axial force versus axial compression. The finite element codes BRIGADE/Plus and ABAQUS/Explicit were used for analysis. Static simulation was used to optimize the dimensions of shock absorber. Dynamic analysis shows that shock absorber is capable to withstand the dynamic load for successful force suppression function in case of an accident. (authors)

Unsteady Aerodynamic Force Sensing from Measured Strain

Science.gov (United States)

Pak, Chan-Gi

2016-01-01

A simple approach for computing unsteady aerodynamic forces from simulated measured strain data is proposed in this study. First, the deflection and slope of the structure are computed from the unsteady strain using the two-step approach. Velocities and accelerations of the structure are computed using the autoregressive moving average model, on-line parameter estimator, low-pass filter, and a least-squares curve fitting method together with analytical derivatives with respect to time. Finally, aerodynamic forces over the wing are computed using modal aerodynamic influence coefficient matrices, a rational function approximation, and a time-marching algorithm. A cantilevered rectangular wing built and tested at the NASA Langley Research Center (Hampton, Virginia, USA) in 1959 is used to validate the simple approach. Unsteady aerodynamic forces as well as wing deflections, velocities, accelerations, and strains are computed using the CFL3D computational fluid dynamics (CFD) code and an MSC/NASTRAN code (MSC Software Corporation, Newport Beach, California, USA), and these CFL3D-based results are assumed as measured quantities. Based on the measured strains, wing deflections, velocities, accelerations, and aerodynamic forces are computed using the proposed approach. These computed deflections, velocities, accelerations, and unsteady aerodynamic forces are compared with the CFL3D/NASTRAN-based results. In general, computed aerodynamic forces based on the lifting surface theory in subsonic speeds are in good agreement with the target aerodynamic forces generated using CFL3D code with the Euler equation. Excellent aeroelastic responses are obtained even with unsteady strain data under the signal to noise ratio of -9.8dB. The deflections, velocities, and accelerations at each sensor location are independent of structural and aerodynamic models. Therefore, the distributed strain data together with the current proposed approaches can be used as distributed deflection

Bipolar plate materials in molten carbonate fuel cells. Final CRADA report.

Energy Technology Data Exchange (ETDEWEB)

Krumpelt, M.

2004-06-01

Advantages of implementation of power plants based on electrochemical reactions are successfully demonstrated in the USA and Japan. One of the msot promising types of fuel cells (FC) is a type of high temperature fuel cells. At present, thanks to the efforts of the leading countries that develop fuel cell technologies power plants on the basis of molten carbonate fuel cells (MCFC) and solid oxide fuel cells (SOFC) are really close to commercialization. One of the problems that are to be solved for practical implementation of MCFC and SOFC is a problem of corrosion of metal components of stacks that are assembled of a number of fuel cells. One of the major components of MCFC and SOFC stacks is a bipolar separator plate (BSP) that performs several functions - it is separation of reactant gas flows sealing of the joints between fuel cells, and current collection from the surface of electrodes. The goal of Task 1 of the project is to develop new cost-effective nickel coatings for the Russian 20X23H18 steel for an MCFC bipolar separator plate using technological processes usually implemented to apply corrosion stable coatings onto the metal parts for products in the defense. There was planned the research on production of nickel coatings using different methods, first of all the galvanic one and the explosion cladding one. As a result of the works, 0.4 x 712 x 1296 mm plates coated with nickel on one side were to be made and passed to ANL. A line of 4 galvanic baths 600 liters was to be built for the galvanic coating applications. The goal of Task 2 of the project is the development of a new material of an MCFC bipolar separator plate with an upgraded corrosion stability, and development of a technology to produce cold roll sheets of this material the sizes of which will be 0.8 x 712x 1296 mm. As a result of these works, a pilot batch of the rolled material in sheets 0.8 x 712 x 1296 mm in size is to be made (in accordance with the norms and standards of the Russian

Fuel assembly

International Nuclear Information System (INIS)

Ishibashi, Yoko; Aoyama, Motoo; Oyama, Jun-ichi.

1995-01-01

Burnable poison-incorporating fuel rods of a first group are disposed in a region in adjacent with a water rod having a large diameter (neutron moderator rod) disposed to the central portion of a fuel assembly. Burnable poison-incorporating fuel rods of a second group are disposed to a region other than peripheral zone in adjacent with a channel box and corners positioned at an inner zone, in adjacent with the channel box. The average concentration of burnable poisons of the burnable poison-incorporating fuel rods of the first group is made greater than that of the second group. With such a constitution, when the burnable poisons of the first group are burnt out, the burnable poisons of the second group are also burnt out at the same time. Accordingly, an amount of burnable poisons left unburnt at the final stage of the operation cycle is reduced, to improve the reactivity. This can improve the economical property. (I.N.)

Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project (SPAR-II)

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-07-01

As storage of spent fuel has become a key technology in spent fuel management, wet and dry storage have become mature technologies and continue to demonstrate good performance. Increased spent fuel storage capacity in combination with longer storage durations will be needed over the foreseeable future as many countries have delayed their decision on spent fuel disposal or reprocessing. Extended spent fuel storage is, and will remain, an important activity for all countries with nuclear power programmes. A number of countries are planning or have already initiated research programmes on spent fuel storage performance, and there is a continuing benefit in exchanging spent fuel storage experience of the Member States in order to build a comprehensive technology knowledge base. Potential degradation mechanisms that may affect cladding integrity during wet storage are uniform corrosion, pitting, galvanic, and microbiologically-influenced corrosion. Potential degradation mechanisms that may affect cladding integrity during dry storage and subsequent handling and transportation operations are air oxidation, thermal creep, stress corrosion cracking (SCC), delayed hydride cracking (DHC), hydride re-orientation, hydrogen migration and re-distribution. Investigations carried out so far indicate that from the degradation mechanisms that may affect the integrity of spent fuel assembly/bundle structure during interim storage, hydride re-orientation has the potential to impair the ability of the cladding to effectively withstand potentially adverse mechanical challenges resulting from handling or transportation accidents. Fuel integrity issues are related to the definition and criteria of fuel integrity, failure classification, packaging and retrieval of damaged fuel and transport of damaged fuel assemblies. Various monitoring technologies have been developed and used to confirm the continued spent fuel integrity during storage or to provide an early indication of developing

Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project (SPAR-II)

International Nuclear Information System (INIS)

2012-01-01

As storage of spent fuel has become a key technology in spent fuel management, wet and dry storage have become mature technologies and continue to demonstrate good performance. Increased spent fuel storage capacity in combination with longer storage durations will be needed over the foreseeable future as many countries have delayed their decision on spent fuel disposal or reprocessing. Extended spent fuel storage is, and will remain, an important activity for all countries with nuclear power programmes. A number of countries are planning or have already initiated research programmes on spent fuel storage performance, and there is a continuing benefit in exchanging spent fuel storage experience of the Member States in order to build a comprehensive technology knowledge base. Potential degradation mechanisms that may affect cladding integrity during wet storage are uniform corrosion, pitting, galvanic, and microbiologically-influenced corrosion. Potential degradation mechanisms that may affect cladding integrity during dry storage and subsequent handling and transportation operations are air oxidation, thermal creep, stress corrosion cracking (SCC), delayed hydride cracking (DHC), hydride re-orientation, hydrogen migration and re-distribution. Investigations carried out so far indicate that from the degradation mechanisms that may affect the integrity of spent fuel assembly/bundle structure during interim storage, hydride re-orientation has the potential to impair the ability of the cladding to effectively withstand potentially adverse mechanical challenges resulting from handling or transportation accidents. Fuel integrity issues are related to the definition and criteria of fuel integrity, failure classification, packaging and retrieval of damaged fuel and transport of damaged fuel assemblies. Various monitoring technologies have been developed and used to confirm the continued spent fuel integrity during storage or to provide an early indication of developing

A central spent fuel storage in Sweden

International Nuclear Information System (INIS)

Gustafsson, B.; Hagberth, R.

1978-01-01

A planned central spent fuel storage facility in Sweden is described. The nuclear power program and quantities of spent fuel generated in Sweden is discussed. A general description of the facility is given with emphasis on the lay-out of the buildings, transport casks and fuel handling. Finally a possible design of a Swedish transportation system is discussed. (author)

Nuclear forces

International Nuclear Information System (INIS)

Holinde, K.

1990-01-01

In this paper the present status of the meson theory of nuclear forces is reviewed. After some introductory remarks about the relevance of the meson exchange concept in the era of QCD and the empirical features of the NN interaction, the exciting history of nuclear forces is briefly outlined. In the main part, the author gives the basic physical ideas and sketch the derivation of the one-boson-exchange model of the nuclear force, in the Feynman approach. Secondly we describe, in a qualitative way, various necessary extensions, leading to the Bonn model of the N interaction. Finally, points to some interesting pen questions connected with the extended quark structure of the hadrons, which are topics of current research activity

A Methodology to Assess the Benefit of Operational or Tactic Adjustments to Reduce Marine Corps Fuel Consumption

Science.gov (United States)

2015-12-01

force threat level , and the ship-to-firing distance. Cloud cover did not change artillery effectiveness or fuel consumption , so it was not a factor...probability of neutralization • fuel consumption rates of assets • red force threat level structure • asset to target priorities • weapon to target...while consuming the same amount of fuel as the CH-53E, while also advertised to have higher cruise speed at this fuel burn rate. (Sikorsky 2015).Given

Modeling of fuel bundle vibration and the associated fretting wear in a CANDU fuel channel

International Nuclear Information System (INIS)

Mohany, A.; Hassan, M.

2011-01-01

In this paper a numerical model is developed to predict the vibration response of a CANDU® fuel bundle and the associated fretting wear in the surrounding pressure tube. One excitation mechanism is considered in this model; turbulence-induced excitation caused by coolant flow inside the fuel channel. The numerical model can be easily adapted to include the effects of seismic events, fuel bundle impact during refuelling and start-up of the reactor, and the acoustic pressure pulsations caused by the primary heat transport (PHT) pumps. The simulation is performed for a typical CANDU fuel bundle with 37 fuel elements. The clearances between the buttons of the inner fuel elements, and between the bearing pads of the outer fuel elements and the pressure tube were measured from an actual fuel bundle. Some variability among the measured clearance values was observed. Therefore, probability density functions of the measured clearance values were established and the simulation was performed for the probabilistic distribution of the clearance values. The contact between the fuel bundle and the pressure tube is modeled using pseudo-force contact method. The proposed modelling technique can be used in future CANDU reactors to avoid fuel and pressure tube fretting damage due to the aforementioned excitation mechanisms. (author)

Optimised fuel cell operation management. Final report; Optimiertes Betriebsmanagement von Brennstoffzellen. Schlussbericht

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-11-05

Investigations were carried out on two fuel cell systems of the type ONSI PC 25C (large fuel cells) and three fuel cell heaters of the type HXS 1000 PREMIERE by Sulzer HEXIS (small fuel cells) for long periods of time and in different modes of operation. Information was obtained on efficiencies and performance, pollutant emissions, degradation and availability. The status of the project is reported and compared with the goals stated initially. (orig.) [German] Das Hauptziel des Projektes bestand darin, Kriterien fuer den optimalen Betrieb (hohe Verfuegbarkeit, geringe Anzahl von Abschaltungen, hohe Energienutzung etc.) von Brennstoffzellen im Zusammenspiel mit den peripheren Komponenten der Gebaeudetechnik zu ermitteln. Hierzu wurden Untersuchungen an zwei Brennstoffzellen-Anlagen vom Typ ONSI PC 25C (grosse BZ) und an drei Brennstoffzellen-Heizgeraeten HXS 1000 PREMIERE von Sulzer HEXIS (kleine BZ) ueber laengere Zeitraeume und bei weitestgehend unterschiedlichen Betriebsweisen durchgefuehrt. In Auswertung dieser Messkomplexe wurden Aussagen zum Betriebsverhalten der Brennstoffzellen, wie z.B. zu Wirk- und Nutzungsgraden, Schadstoffemissionen, Verlauf der Degradation sowie zur Verfuegbarkeit getroffen. Im vorliegenden Bericht wird der erreichte Sachstand detailliert dargestellt und mit den Zielsetzungen des FuE-Antrages verglichen. (orig.)

Final Scientific Report, New Proton Conductive Composite Materials for PEM Fuel Cells

Energy Technology Data Exchange (ETDEWEB)

Lvov, Serguei

2010-11-08

This project covered one of the main challenges in present-day PEM fuel cell technology: to design a membrane capable of maintaining high conductivity and mechanical integrity when temperature is elevated and water vapor pressure is severely reduced. The DOE conductivity milestone of 0.1 S cm-1 at 120 degrees C and 50 % relative humidity (RH) for designed membranes addressed the target for the project. Our approach presumed to develop a composite membrane with hydrophilic proton-conductive inorganic material and the proton conductive polymeric matrix that is able to “bridge” the conduction paths in the membrane. The unique aspect of our approach was the use of highly functionalized inorganic additives to benefit from their water retention properties and high conductivity as well. A promising result turns out that highly hydrophilic phosphorsilicate gels added in Nafion matrix improved PEM fuel cell performance by over 50% compared with bare Nafion membrane at 120 degrees C and 50 % RH. This achievement realizes that the fuel cell operating pressure can be kept low, which would make the PEM fuel cell much more cost efficient and adaptable to practical operating conditions and facilitate its faster commercialization particularly in automotive and stationary applications.

Review of the Effects of Normal Conditions of Transport on Spent Fuel Integrity in Transportation Casks

Energy Technology Data Exchange (ETDEWEB)

Kim, Junggoo; Yoo, Youngik; Lee, Seongki; Lim, Chaejoon [Korea Nuclear Fuel Co., Daejeon (Korea, Republic of)

2014-10-15

Spent fuel(SF) storage capacity of each domestic nuclear power plant will reach a saturated state in the near future. Although there are several methods of SF disposal, interim storage is suggested as the most realistic and promising alternative. SF integrity evaluation is a regulatory requirement that is described in Part 71 of Code of Federal Regulations, Title 10 of the U..S. NRC licensing requirement. In this paper, the report is reviewed written by EPRI in US and it is helpful to a development of domestic SF integrity evaluation technology. EPRI report about integrity evaluation method on normal conditions of high burn-up spent fuel transport is reviewed. First, dynamic forces occurred in one-foot side drop are calculated. And deformation patterns and fuel rods responses by dynamic forces calculated from spent fuel and cask model are analyzed. It is shown that the damage of fuel rods is not occurred by the dynamic forces on normal conditions. Assembly distortion is not predicted, by virtue of the facts that the spacer grids do not experience significant permanent deformation. Axial forces, bending moments and pinch forces of fuel rods are calculated and compared with the results under the hypothetical accident conditions. No occurrence of transverse tearing mode that is the most serious damage mode in side drop case is predicted. Till now, in Korea, regulatory requirements related with structural integrity of spent fuel are not specified such as 10CFR71. To establish own regulation standards, producing and analyzing sufficient experimental data must be performed preferentially. Based on this, failure analysis and criteria establishment are necessary through modeling and analyzing of spent fuel.

Crud deposition modeling on BWR fuel rods

International Nuclear Information System (INIS)

Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

2014-01-01

Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

Two dimensional point of use fuel cell : a final LDRD project report.

Energy Technology Data Exchange (ETDEWEB)