Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

Energy Technology Data Exchange (ETDEWEB)

Gilles Youinou; R. Sonat Sen

2013-09-01

The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy – UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

Summary of Preliminary Criticality Analysis for Peach Bottom Fuel in the DOE Standardized Spent Nuclear Fuel Canister

International Nuclear Information System (INIS)

Henrikson, D.J.

1999-01-01

The Department of Energy's (DOE's) National Spent Nuclear Fuel Program is developing a standardized set of canisters for DOE spent nuclear fuel (SNF). These canisters will be used for DOE SNF handling, interim storage, transportation, and disposal in the national repository. Several fuels are being examined in conjunction with the DOE SNF canisters. This report summarizes the preliminary criticality safety analysis that addresses general fissile loading limits for Peach Bottom graphite fuel in the DOE SNF canister. The canister is considered both alone and inside the 5-HLW/DOE Long Spent Fuel Co-disposal Waste Package, and in intact and degraded conditions. Results are appropriate for a single DOE SNF canister. Specific facilities, equipment, canister internal structures, and scenarios for handling, storage, and transportation have not yet been defined and are not evaluated in this analysis. The analysis assumes that the DOE SNF canister is designed so that it maintains reasonable geometric integrity. Parameters important to the results are the canister outer diameter, inner diameter, and wall thickness. These parameters are assumed to have nominal dimensions of 45.7-cm (18.0-in.), 43.815-cm (17.25-in), and 0.953-cm (0.375-in.), respectively. Based on the analysis results, the recommended fissile loading for the DOE SNF canister is 13 Peach Bottom fuel elements if no internal steel is present, and 15 Peach Bottom fuel elements if credit is taken for internal steel

Preliminary design report for the prototypical fuel rod consolidation system

International Nuclear Information System (INIS)

Rosa, J.M.

1986-01-01

This report documents NUTECH's preliminary design of a dry, spent fuel rod consolidation system. This preliminary design is the result of Phase I of a planned four phase project. The present report on this project provides a considerable amount of detail for a preliminary design effort. The design and all of its details are described in this Preliminary Design Report (PDR). The NUTECH dry rod consolidation system described herein is remotely operated. It provides for automatic operation, but with operator hold points between key steps in the process. The operator has the ability to switch to a manual operation mode at any point in the process. The system is directed by the operator using an executive computer which controls and coordinates the operation of the in-cell equipment. The operator monitors the process using an in-cell closed circuit television (CCTV) system with audio output and equipment status displays on the computer monitor. The in-cell mechanical equipment consists of the following: (1) two overhead cranes with manipulators; (2) a multi-degree of freedom fuel handling table and its clamping equipment; (3) a fuel assembly end fitting removal station and its tools; (4) a consolidator (which pulls rods, assembles the consolidated bundle and loads the canister); (5) a canister end cap welder and weld inspection system; (6) decontamination systems; and (7) the CCTV and microphone systems

Canadian CANDU fuel development program and recent fuel operating experience

International Nuclear Information System (INIS)

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

1999-01-01

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

Fuel Cell and Hydrogen Technologies Program | Hydrogen and Fuel Cells |

Science.gov (United States)

NREL Fuel Cell and Hydrogen Technologies Program Fuel Cell and Hydrogen Technologies Program Through its Fuel Cell and Hydrogen Technologies Program, NREL researches, develops, analyzes, and validates fuel cell and hydrogen production, delivery, and storage technologies for transportation

DEMONSTRATION OF FUEL CELLS TO RECOVER ENERGY FROM ANAEROBIC DIGESTER GAS - PHASE I. CONCEPTUAL DESIGN, PRELIMINARY COST, AND EVALUATION STUDY

Science.gov (United States)

The report discusses Phase I (a conceptual design, preliminary cost, and evaluation study) of a program to demonstrate the recovery of energy from waste methane produced by anaerobic digestion of waste water treatment sludge. The fuel cell is being used for this application becau...

PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

International Nuclear Information System (INIS)

Kirk, W.L.

1976-08-01

The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters

PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

Energy Technology Data Exchange (ETDEWEB)

Kirk, W.L. (comp.)

1976-08-01

The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters.

The Canadian CANDU fuel development program and recent fuel operating experience

International Nuclear Information System (INIS)

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

1999-01-01

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

In Vitro Studies and Preliminary Mathematical Model for Jet Fuel and Noise Induced Auditory Impairment

Science.gov (United States)

2015-06-01

of JP-8 and a Fischer- Tropsch synthetic jet fuel following subacute inhalation exposure in rats. Toxicol Sci 116(1): 239-248. Gallinat, J...AFRL-RH-WP-TR-2015-0084 IN VITRO STUDIES AND PRELIMINARY MATHEMATICAL MODEL FOR JET FUEL AND NOISE INDUCED AUDITORY IMPAIRMENT...April 2014 – September 2014 4. TITLE AND SUBTITLE In Vitro Studies and Preliminary Mathematical Model for Jet Fuel and Noise Induced Auditory

The preliminary design and feasibility study of the spent fuel and high level waste repository in the Czech Republic

International Nuclear Information System (INIS)

Valvoda, Z.; Holub, J.; Kucerka, M.

1996-01-01

In the year 1993, began the Program of Development of the Spent Fuel and High Level Waste Repository in the Conditions of the Czech Republic. During the first phase, the basic concept and structure of the Program has been developed, and the basic design criteria and requirements were prepared. In the conditions of the Czech Republic, only an underground repository in deep geological formation is acceptable. Expected depth is between 500 to 1000 meters and as host rock will be granites. A preliminary variant design study was realized in 1994, that analyzed the radioactive waste and spent fuel flow from NPPs to the repository, various possibilities of transportation in accordance to the various concepts of spent fuel conditioning and transportation to the underground structures. Conditioning and encapsulation of spent fuel and/or radioactive waste is proposed on the repository site. Underground disposal structures are proposed at one underground floor. The repository will have reserve capacity for radioactive waste from NPPs decommissioning and for waste non acceptable to other repositories. Vertical disposal of unshielded canisters in boreholes and/or horizontal disposal of shielded canisters is studied. As the base term of the start up of the repository operation, the year 2035 has been established. From this date, a preliminary time schedule of the Project has been developed. A method of calculating leveled and discounted costs within the repository lifetime, for each of selected 5 variants, was used for economic calculations. Preliminary expected parametric costs of the repository are about 0,1 Kc ($0.004) per MWh, produced in the Czech NPPs. In 1995, the design and feasibility study has gone in more details to the technical concept of repository construction and proposed technologies, as well as to the operational phase of the repository. Paper will describe results of the 1995 design work and will present the program of the repository development in next period

Core fuel management using TVS-2M fuel assembly and economic analysis

International Nuclear Information System (INIS)

Xu Min; Wang Hongxia; Li Youyi

2014-01-01

To improve the economic efficiency, TVS-2M fuel assembly was considered to apply in Tianwan Nuclear Power Plant units 3, 4. Using KASKAD program package, a preliminary research and design was carried out for the Tianwan Nuclear Power Plant loading TVS-2M fuel assembly from the first cycle to equilibrium cycle. An improved fuel management program was obtained, and the economic analysis of the two fuel management programs with or without TVS-2M assembly was studied. The analysis results show that TVS-2M fuel assembly can improve the economic efficiency of the plant remarkably. (authors)

Romanian nuclear fuel program

International Nuclear Information System (INIS)

Budan, O.

1999-01-01

The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle

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...

Preliminary study on direct recycling of spent PWR fuel in PWR system

International Nuclear Information System (INIS)

Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

2012-01-01

Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

Spent-fuel disassembly and canning programs at the Barnwell Nuclear Fuel Plant (BNFP)

International Nuclear Information System (INIS)

Townes, G.A.

1980-02-01

Doubling pool storage and tripling truck cask shipping capability appear feasible based on preliminary development work performed at the BNFP. This would be accomplished by disassembly of the fuel assembly and canning of the fuel pins. The dry, remote disassembly and canning in the headend cells offer adequate rates, easily controlled operator exposure, and visual inspection of the fuel pins and operations through shielded windows

Oxygenates in automotive fuels. Consequence analysis - preliminary study

International Nuclear Information System (INIS)

Brandberg, Aa.; Saevbark, B.

1994-01-01

Oxygenates is used in gasoline due to several reasons. They are added as high-octane components in unleaded gasoline and as agents to reduce the emission of harmful substances. Oxygenates produced from biomass might constitute a coming market for alternative fuels. This preliminary study describes the prerequisites and consequences of such an oxygenate utilization. 39 refs, 9 figs, 5 tabs

A preliminary assessment of thorium as a fuel for thermal reactors

International Nuclear Information System (INIS)

Duret, M.F.; Halsall, M.J.

1965-08-01

This report presents a preliminary survey of the contributing factors to total fuel cost, for a variety of potentially economic fuel cycles. U 238 and Th 232 are considered as fertile materials with enrichments of U 235 , U 233 and plutonium. The total fuel cost comprises: (a) a fuel supply cost which is derived from a manufacturing cost and the fuel burnup, (b) as spent fuel credit, and (c) inventory charges. The interrelation of these three factors and their dependence on reactor flux, fuel burnup, the heat rating of the fuel and the interest rate are considered for each combination of fertile and fissile material. Results show inventory charges, spent fuel credit (hence processing costs) and fabrication costs have an important effect on the fuelling cost. In addition, using highly rated thorium-based fuels will affect reactor capital costs. Further, more detailed studies of reactor design, fuel design and fuel management schemes are now required to verify the attraction of using thorium as a reactor fuel. (author)

Spent fuels program

International Nuclear Information System (INIS)

Shappert, L.B.

1983-01-01

The goal of this task is to support the Domestic Spent Fuel Storage Program through studies involving the transport of spent fuel. A catalog was developed to provide authoritative, timely, and accessible transportation information for persons involved in the transport of irradiated reactor fuel. The catalog, drafted and submitted to the Transportation Technology Center, Sandia National Laboratories, for their review and approval, covers such topics as federal, state, and local regulations, spent fuel characteristics, cask characteristics, transportation costs, and emergency response information

Treatment of spent fuels from research reactors and reactor development programs in Germany

International Nuclear Information System (INIS)

Closs, K.D.

1999-01-01

Quite a great number of different types of spent fuel from research reactors and development programs exists in Germany. The general policy is to send back to the USA as long as possible fuel from MTRs and TRIGAs of USA origin. An option is reprocessing in Great Britain or France. This option is pursued as long as reprocessing and reuse of the recovered material is economically justifiable. For those fuels which cannot be returned to the USA or which will not be reprocessed, a domestic back-up solution of spent fuel management has been developed in Germany, compatible with the management of spent fuel from power reactors. It consists in dry storage in special casks and, later on, direct disposal. Preliminary results from experimental R and D investigations with research reactor fuel and experience from LWR fuel lead to the conclusion that the direct disposal option even for research reactor fuel or exotic fuel does not impose major technical difficulties for the German waste management and disposal concept. (author)

Preliminary study on metallic inclusion in nuclear fuel

International Nuclear Information System (INIS)

Fukuzawa, Takashi; Tanaka, Masahiro; Tanabe, Tetsuo; Imoto, Shosuke

1984-01-01

In recent postirradiation tests, metallic fission products such as Mo, Ru, Rh and Pd are known to precipitate as metallic inclusions in the fuel. These inclusions remain as insoluble residues and provide various problems in different fields of the reprocessing. In this report are presented preliminary results of the study on the ternary phase diagram of Mo-Ru-Pd system and on their properties in nitric acid or various oxidative environments. It is concluded that (1) most of metallic inclusions which are insoluble in nitric acid show epsilon phase, Ru base hcp alloy, in which a large amount of Mo and Pd are soluble, (2) Pd, however, seems to deposit separately in the fuel pin because of its high vapor pressure, (3) Mo fraction in the inclusion would be highly dependent on oxygen potential in the fuel pin. (author)

Development of a Computer Program (CASK) for the Analysis of Logistics and Transportation Cost of the Spent Fuels

International Nuclear Information System (INIS)

Cha, Jeong-Hun; Choi, Heui-Joo; Cho, Dong-Keun; Kim, Seong-Ki; Lee, Jong-Youl; Choi, Jong-Won

2008-07-01

The cost for the spent fuel management includes the costs for the interim storage, the transportation, and the permanent disposal of the spent fuels. The CASK(Cost and logistics Analysis program for Spent fuel transportation in Korea) program is developed to analyze logistics and transportation cost of the spent fuels. And the total amount of PWR spent fuels stored in four nuclear plant sites, a centralized interim storage facility near coast and a permanent disposal facility near the interim storage facility are considered in this program. The CASK program is developed by using Visual Basic language and coupled with an excel sheet. The excel sheet shows a change of logistics and transportation cost. Also transportation unit cost is easily changed in the excel sheet. The scopes of the report are explanation of parameters in the CASK program and a preliminary calculation. We have developed the CASK version 1.0 so far, and will update its parameters in transportation cost and transportation scenario. Also, we will incorporate it into the program which is used for the projection of spent fuels from the nuclear power plants. Finally, it is expected that the CASK program could be a part of the cost estimation tools which are under development at KAERI. And this program will be a very useful tool for the establishment of transportation scenario and transportation cost in Korean situations

Canadian fuel development program

International Nuclear Information System (INIS)

Gacesa, M.; Young, E.G.

1992-11-01

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

Preliminary cost analysis of a universal package concept in the spent fuel management system

International Nuclear Information System (INIS)

1984-09-01

The purpose of this study is to provide a preliminary cost assessment of a universal spent fuel package concept as it applies to the backend of the once through nuclear fuel cycle; i.e., a package that would be qualified for spent fuel storage, transportation, and disposal. To provide this preliminary cost assessment, costs for each element of the spent fuel management system have been compiled for system scenarios employing the universal package, and these costs are compared against system costs for scenarios employing the universal package, and these costs are compared against system costs for scenarios employing other types of storage, transportation, and disposal packages. The system elements considered in this study are storage at the nuclear power plant, spent fuel transportation, a Monitored Retrievable Storage (MRS) facility, and a geologic repository. In accordance with the Nuclear Waste Policy Act, most of these system elements and associated functions will be the responsibility of the Department of Energy. 10 refs., 25 figs., 22 tabs

Consolidated fuel reprocessing program

Science.gov (United States)

1985-04-01

A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

Fuel isolation research for the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

1982-06-01

This document is intended to give a broad outline of the Fuel Isolatikn program and to indicate how this program fits into the overall framework of the Canadian Nuclear Fuel Waste Management Program. Similar activities in other countries are described, and the differences in philosophy behind these and the Canadian program are highlighted. A program plan is presented that outlines the development of research programs that contribute to the safety assessment of the disposal concept and the development of technology required for selection and optimization of a feasible fuel isolation system. Some indication of the work that might take place beyond concept assessment, at the end of the decade, is also given. The current program is described in some detail, with emphasis on what the prkgram has achieved to date and hopes to achieve in the future for the concept assessment phase of the waste management program. Finally, some major capital facilities associated with the fuel isolation program are described

Canadian fuel development program in 1997/98

International Nuclear Information System (INIS)

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

1997-01-01

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

Preliminary concepts: coordinated safeguards for materials management in a thorium--uranium fuel reprocessing plant

International Nuclear Information System (INIS)

Hakkila, E.A.; Barnes, J.W.; Dayem, H.A.; Dietz, R.J.; Shipley, J.P.

1978-10-01

This report addresses preliminary concepts for coordinated safeguards materials management in a typical generic thorium--uranium-fueled light-water reactor (LWR) fuels reprocessing plant. The reference facility is designed to recover thorium and uranium from first-generation (denatured 235 U) startup fuels, first-recycle and equilibrium (denatured 233 U) thorium--uranium LWR fuels, and to recover the plutonium generated in the 238 U denaturant as well. 12 figures, 3 tables

The Advanced Petroleum-Based Fuels Program Evaluation of EC-Diesel and Diesel Particulate Filters in Southern California Vehicle Fleets

International Nuclear Information System (INIS)

None

2000-01-01

The EC-Diesel and particulate filter combination greatly reduced the particulate matter, hydrocarbon, and carbon monoxide emissions of all vehicles tested in the program to date. Particulate matter reductions greater than 98% were achieved. For several vehicles tested, the PM and HC emissions were less than background levels. Based on preliminary statistical analysis, there is 95%+ confidence that EC-D and particulate filters reduced emissions from three different types of vehicles. A fuel consumption penalty was not detectable using the current test procedures and chassis dynamometer laboratory. Test vehicles equipped with the CRT and DPX particulate filters and fueled with EC-Diesel fuel have operated reliably during the program start-up period

Preliminary analysis of the creep behaviour of nuclear fuel-waste container materials

International Nuclear Information System (INIS)

Dutton, R.; Leitch, B.W.; Crosthwaite, J.L.; Kasprick, G.R.

1996-12-01

In the Canadian Nuclear Fuel Waste Management Program, it is proposed that nuclear fuel waste be placed in a durable container and disposed of in a deep underground vault. Consideration of various disposal-container designs has identified either titanium or copper as the material suitable for constructing the container shell. As part of the R and D program to examine the structural integrity of the container, creep tests are being conducted on commercially pure titanium and oxygen-free copper. This report presents the preliminary data obtained. It also describes the evaluation of various constitutive equations to represent the creep curves, thus providing the basis for extrapolation of the creep behaviour over the design lifetime of the container. In this regard, a specific focus is placed on equations derived from the 0-Projection Concept. Recognizing that the container lifetime will be determined by the onset of tertiary creep leading to creep rupture, we present the results of the metallographic examination of creep damage. This shows that the tertiary stage in titanium is associated with the formation of transgranular cavities within the region of localized necking of the creep specimens. In contrast, creep damage in copper is in the form of intergranular cavities uniformly distributed throughout the gauge length. These results are analyzed within the context of the extant literature, and their implications for future container design are discussed. (author)

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

International Nuclear Information System (INIS)

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

1978-12-01

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

The DUPIC fuel development program in KAERI

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

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

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

Preliminary assessment on compatibility of DUPIC fuel with CANDU-6

International Nuclear Information System (INIS)

Choi, Hang-Bok; Roh, G.H.; Jeong, C.J.; Rhee, B.W.; Choi, J.W.; Boss, C.R.

1997-01-01

The compatibility of DUPIC fuel with the existing CANDU-6 reactor was assessed. The technical issues of DUPIC fuel compatibility were chosen based on the CANDU physics design requirements and inherent characteristics of DUPIC fuel. The compatibility was assessed for the reference DUPIC fuel composition which was determined to reduce the composition heterogeneity and improve the spent PWR fuel utilization. Preliminary studies on a CANDU core loaded with DUPIC fuel have shown that the nominal power distribution is flatter than that of a natural uranium core when a 2-bundle shift refueling scheme is used, which reduces the reactivity worths of devices in the core and, therefore, the performance of reactivity devices was assessed. The safety of the core was assessed by a LOCA simulation and it was found that the power pulse upon LOCA can be maintained below that in the natural uranium core when a poison material is used in the DUPIC fuel. For the feasibility of handling DUPIC fuel in the plant, it will be necessary to introduce new equipment to load the DUPIC fuel in the refueling magazine. The radiation effect of DUPIC fuel on both the reactor hardware and the environment will be qualitatively analyzed later. (author)

Preliminary Analysis of the Bundle-Duct Interaction for the fuel of SFR

Energy Technology Data Exchange (ETDEWEB)

Lee, Byoung Oon; Cheon, Jin Sik; Hahn, Do Hee; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-10-15

BDI (Bundle-Duct Interaction) occurs in the fuel of SFR (Sodium-cooled Fast Reactor) due to the radial expansion and bowing of a fuel pin bundle. Under the BDI condition, excess cladding strain and hot spots would occur. Therefore, BDI, which is the dominant deformation mechanisms in a fuel pin bundle, should be considered to evaluate the FBR fuel integrity. The analysis codes such as ETOILE and BMBOO, have been developed to evaluate the BDI behavior. The bundle duct interaction model is also being developed for SFR in Korea. This model is based on ANSYS. In this paper, the fuel pin configuration model for the BDI calculation was established. The preliminary analysis of the bundle-duct interaction was performed to evaluate the fuel design concept.

Preliminary Economics for Hydrocarbon Fuel Production from Cellulosic Sugars

Energy Technology Data Exchange (ETDEWEB)

Collett, James R.; Meyer, Pimphan A.; Jones, Susanne B.

2014-05-18

Biorefinery process and economic models built in CHEMCAD and a preliminary, genome-scale metabolic model for the oleaginous yeast Lipomyces starkeyi were used to simulate the bioconversion of corn stover to lipids, and the upgrading of these hydrocarbon precursors to diesel and jet fuel. The metabolic model was based on the recently released genome sequence for L. starkeyi and on metabolic pathway information from the literature. The process model was based on bioconversion, lipid extraction, and lipid oil upgrading data found in literature, on new laboratory experimental data, and on yield predictions from the preliminary L. starkeyi metabolic model. The current plant gate production cost for a distillate-range hydrocarbon fuel was estimated by the process model Base Case to be $9.5/gallon ($9.0 /gallon of gasoline equivalent) with assumptions of 2011$, 10% internal return on investment, and 2205 ton/day dry feed rate. Opportunities for reducing the cost to below $5.0/gallon, such as improving bioconversion lipid yield and hydrogenation catalyst selectivity, are presented in a Target Case. The process and economic models developed for this work will be updated in 2014 with new experimental data and predictions from a refined metabolic network model for L. starkeyi. Attaining a production cost of $3.0/gallon will require finding higher value uses for lignin other than power generation, such as conversion to additional fuel or to a co-product.

Alcohol fuels program technical review

Energy Technology Data Exchange (ETDEWEB)

None

1981-07-01

The last issue of the Alcohol Fuels Process R/D Newsletter contained a work breakdown structure (WBS) of the SERI Alcohol Fuels Program that stressed the subcontracted portion of the program and discussed the SERI biotechnology in-house program. This issue shows the WBS for the in-house programs and contains highlights for the remaining in-house tasks, that is, methanol production research, alcohol utilization research, and membrane research. The methanol production research activity consists of two elements: development of a pressurized oxygen gasifier and synthesis of catalytic materials to more efficiently convert synthesis gas to methanol and higher alcohols. A report is included (Finegold et al. 1981) that details the experimental apparatus and recent results obtained from the gasifier. The catalysis research is principally directed toward producing novel organometallic compounds for use as a homogeneous catalyst. The utilization research is directed toward the development of novel engine systems that use pure alcohol for fuel. Reforming methanol and ethanol catalytically to produce H/sub 2/ and CO gas for use as a fuel offers performance and efficiency advantages over burning alcohol directly as fuel in an engine. An application of this approach is also detailed at the end of this section. Another area of utilization is the use of fuel cells in transportation. In-house researchers investigating alternate electrolyte systems are exploring the direct and indirect use of alcohols in fuel cells. A workshop is being organized to explore potential applications of fuel cells in the transportation sector. The membrane research group is equipping to evaluate alcohol/water separation membranes and is also establishing cost estimation and energy utilization figures for use in alcohol plant design.

Russian research reactor fuel return program starts shipping fuel to Russia

International Nuclear Information System (INIS)

Dedik, T.; Bolshinsky, I.; Krass, A.

2003-01-01

For almost four years the United States (U.S), the Russian Federation (R.F.), and the International Atomic Energy Agency (IAEA) have been discussing an initiative to return Soviet/Russian-origin research reactor fuel to the Russian Federation. In a series of bilateral and trilateral meetings in Vienna and Moscow, considerable progress has been made toward defining the Russian Research Reactor Fuel Return Program as well as obtaining the necessary technical data to facilitate the return. More than 20 research reactors in 17 countries that have Soviet- or Russian-supplied fuel have identified. Most of these reactors have stocks of both fresh and irradiated HEU fuel that must be carefully stored and managed for many years to come. On September 21, 2003 the Russian Research Reactor Fuel Return program shipped 14 kg of fresh Russian-origin HEU fuel from Romania to the nuclear fuel fabrication facility in Russia, which represented the beginning of the practical implementation of the program. (author)

Preliminary Analysis of the Fuel Bundle Stiffness by ANSYS for SFR

Energy Technology Data Exchange (ETDEWEB)

Lee, Byoung Oon; Cheon, Jin Sik; Hahn, Do Hee; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-05-15

In SFR (Sodium-cooled Fast Reactor) the temperature of the fuel pin is higher than that of the hexagonal duct, so the thermal expansion rate of the fuel bundle is higher than that of the duct. The neutron fluence and the fuel pin pressure are also increased according to the burnup. So the radial expansion and bowing of a fuel pin bundle would occur, and then fuel bundle would interact with a duct. This phenomenon is called bundle-to-duct interaction (BDI). Under the BDI condition, excess cladding strain and hot spots would occur. Therefore BDI as well as the core mechanics should be considered to evaluate the FBR fuel integrity. The analysis codes such as ETOILE, SHADOW, and MARSE, have been developed to evaluate the BDI behavior. The ANSYS based model is also being developed to analysis the bundle duct interaction for SFR in Korea. In this paper, the fuel pin/bundle model for analyzing the bending deflection and oval deformation was described. The preliminary analysis of the fuel bundle stiffness was performed by the developed model.

Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

Energy Technology Data Exchange (ETDEWEB)

Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [Unidad de Actividad Combustibles Nucleares Comision Nacional de Energia Atomica (CNE4), Avda. del Libertador, 8250 C1429BNO Buenos Aires (Argentina)

2002-07-01

One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm{sup 3}. PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

International Nuclear Information System (INIS)

Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H.

2002-01-01

One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm 3 . PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

The Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Dixon, R.S.

1984-12-01

The Canadian Nuclear Fuel Waste Management Program involves research into the storage and transportation of used nuclear fuel, immobilization of fuel waste, and deep geological disposal of the immobilized waste. The program is now in the fourth year of a ten-year generic research and development phase. The objective of this phase of the program is to assess the safety and environmental aspects of the deep underground disposal of immobilized fuel waste in plutonic rock. The objectives of the research for each component of the program and the progress made to the end of 1983 are described in this report

Current status of VEGA program and a preliminary test with cesium iodide

International Nuclear Information System (INIS)

Hidaka, A.; Nakamura, T.; Kudo, T.; Hayashida, R.; Nakamura, J.; Otomo, T.; Uetsuka, H.

2000-01-01

The VEGA program has been performed at JAERI to clarify the mechanism of FP release from irradiated PWR/BWR fuels including MOX fuel and to improve predictability of the source term. The principal purposes are to investigate the release of actinides and FPs including non-volatile radionuclides from irradiated fuel at 3000degC under high pressure condition up to 1.0 MPa. The short-life radionuclides will be accumulated by re-irradiation of test fuel just before the experiment using the JAERI's research reactor such as JRR-3 or NSRR. The test facility was installed into the beta/gamma concrete No.5 cell at RFEF and completed in February, 1999. Before the first VEGA-1 test in September, 1999, a preliminary test using a cold simulant, cesium iodide (CsI) was performed to confirm the fundamental capabilities of the test facility. The test results showed that the trapping efficiency of the aerosol filters is about 98%. The amount of CsI which arrived at the downstream pipe of the filters was quite small while a small amount of I 2 gas which can pass through the filters was condensed just before the cold condenser as expected in the design. (author)

LEU fuel development at CERCA. Status as of October 1997. Preliminary developments of MTR plates with UMo fuel

International Nuclear Information System (INIS)

Durand, J.P.; Lavastre, Y.; Grasse, M.

1997-01-01

UMo fuels are considered by the RERTR programme because of their higher density as compared to U 3 Si 2 . This paper is focused on the preliminary results about the manufacture feasibility of Uranium/Molybdenum fuel plates carried out by CERCA. A special procedure of casting and heat treatment has been developed in order to get an homogeneous gamma phase of UMo alloy Although U-5%Mo allows to reach densities up to 9.9 U/cm3 with the advanced process developed by CERCA for the high loaded plates, it is not a good candidate on the thermal stability point of view. U-9%Mo alloy seems to gather all the criteria for a good fuel alloy but it is a little less effective on the Uranium density point of view as compared to U-5%Mo alloy. In any case, the preliminary feasibility results are very much encouraging because UMo alloys seem to be compatible with the Aluminium matrix when taking special care while manufacturing. A good compromise could be an intermediate percentage of Molybdenum or the addition of metal traces in order to thermally stabilise 5%Mo. (author)

Nuclear-fuel-cycle costs. Consolidated Fuel-Reprocessing Program

International Nuclear Information System (INIS)

Burch, W.D.; Haire, M.J.; Rainey, R.H.

1981-01-01

The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel-cycle costs are given for the pressurized-water reactor once-through and fuel-recycle systems, and for the liquid-metal fast-breeder-reactor system. These calculations show that fuel-cycle costs are a small part of the total power costs. For breeder reactors, fuel-cycle costs are about half that of the present once-through system. The total power cost of the breeder-reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment

Preliminary analysis of aircraft fuel systems for use with broadened specification jet fuels

Science.gov (United States)

Pasion, A. J.; Thomas, I.

1977-01-01

An analytical study was conducted on the use of broadened specification hydrocarbon fuels in present day aircraft. A short range Boeing 727 mission and three long range Boeing 747 missions were used as basis of calculation for one-day-per-year extreme values of fuel loading, airport ambient and altitude ambient temperatures with various seasonal and climatic conditions. Four hypothetical fuels were selected; two high-vapor-pressure fuels with 35 kPa and 70 kPa RVP and two high-freezing-point fuels with -29 C and -18 C freezing points. In-flight fuel temperatures were predicted by Boeing's aircraft fuel tank thermal analyzer computer program. Boil-off rates were calculated for the high vapor pressure fuels and heating/insulation requirements for the high freezing point fuels were established. Possible minor and major heating system modifications were investigated with respect to heat output, performance and economic penalties for the high freezing point fuels.

Preliminary nuclear design for test MOX Fuel rods

Energy Technology Data Exchange (ETDEWEB)

Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

1997-10-01

As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

Preliminary plan for decommissioning - repository for spent nuclear fuel

International Nuclear Information System (INIS)

Hallberg, Bengt; Tiberg, Liselotte

2010-06-01

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

National spent fuel program preliminary report RCRA characteristics of DOE-owned spent nuclear fuel DOE-SNF-REP-002. Revision 3

International Nuclear Information System (INIS)

1995-07-01

This report presents information on the preliminary process knowledge to be used in characterizing all Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) types that potentially exhibit a Resource Conservation and Recovery Act (RCRA) characteristic. This report also includes the process knowledge, analyses, and rationale used to preliminarily exclude certain SNF types from RCRA regulation under 40 CFR section 261.4(a)(4), ''Identification and Listing of Hazardous Waste,'' as special nuclear and byproduct material. The evaluations and analyses detailed herein have been undertaken as a proactive approach. In the event that DOE-owned SNF is determined to be a RCRA solid waste, this report provides general direction for each site regarding further characterization efforts. The intent of this report is also to define the path forward to be taken for further evaluation of specific SNF types and a recommended position to be negotiated and established with regional and state regulators throughout the DOE Complex regarding the RCRA-related policy issues

National spent fuel program preliminary report RCRA characteristics of DOE-owned spent nuclear fuel DOE-SNF-REP-002. Revision 3

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-01

This report presents information on the preliminary process knowledge to be used in characterizing all Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) types that potentially exhibit a Resource Conservation and Recovery Act (RCRA) characteristic. This report also includes the process knowledge, analyses, and rationale used to preliminarily exclude certain SNF types from RCRA regulation under 40 CFR {section}261.4(a)(4), ``Identification and Listing of Hazardous Waste,`` as special nuclear and byproduct material. The evaluations and analyses detailed herein have been undertaken as a proactive approach. In the event that DOE-owned SNF is determined to be a RCRA solid waste, this report provides general direction for each site regarding further characterization efforts. The intent of this report is also to define the path forward to be taken for further evaluation of specific SNF types and a recommended position to be negotiated and established with regional and state regulators throughout the DOE Complex regarding the RCRA-related policy issues.

DOE Hydrogen and Fuel Cells Program Plan (September 2011)

Energy Technology Data Exchange (ETDEWEB)

none,

2011-09-01

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

Canadian fuel development program and recent operational experience

International Nuclear Information System (INIS)

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

1995-01-01

This paper provides an overview of the current Canadian CANDU fuel R and D programs and operational experience. The details of operational experience for fuel in Canadian reactors are summarized for the period 1991-1994; excellent fuel performance has been sustained, with steady-state bundle defect rates currently as low as 0.02%. The status of introducing long 37-element bundles, and bundles with rounded bearing pads is reviewed. These minor changes in fuel design have been selectively introduced in response to operational constraints (end-plate cracking and pressure-tube fretting) at Ontario Hydro's Bruce-B and Darlington stations. The R and D programs are generating a more complete understanding of CANDU fuel behaviour, while the CANDU Owners Group (COG) Fuel Technology Program is being re-aligned to a more exclusive focus on the needs of operating stations. Technical highlights and realized benefits from the COG program are summarized. Re-organization of AECL to provide a one-company focus, with an outward looking view to new CANDU markets, has strengthened R and D in advanced fuel cycles. Progress in AECL's key fuel cycle programs is also summarized. (author)

Prototypical spent fuel rod consolidation equipment preliminary design report: Volume 2, Drawings

International Nuclear Information System (INIS)

1986-01-01

This volume consists of 65 E size drawings and 4 sketches of the NUS spent fuel rod consolidation equipment. The drawings have been grouped into categories; a detailed list of the drawings is included. The sketches prepared during the preliminary design process have been included

Program summary. Nuclear waste management and fuel cycle programs

International Nuclear Information System (INIS)

1982-07-01

This Program Summary Document describes the US Department of Energy (DOE) Nuclear Waste Management and Fuel Cycle Programs. Particular emphasis is given to near-term, specifically Fiscal Year (FY) 1982, activities. The overall objective of these programs will be achieved by the demonstration of: (1) safe radioactive waste management practices for storage and disposal of high-level waste and (2) advanced technologies necessary to close the nuclear fuel cycle on a schedule which would assure a healthy future for the development of nuclear power in this country

National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Quality assurance program for licensing

International Nuclear Information System (INIS)

1978-12-01

A National Waste Terminal Storage Repository, in bedded salt, for spent unreprocessed fuel is the subject of a conceptual design project which began in January 1977. This volume presents a preliminary quality assurance program to guide the license applicant in developing a detailed program that will be compatible with anticipated National Waste Terminal Storage (NWTSR2) contracting arrangements and provide the documentation required by regulatory bodies. This QA program is designed to provide confidence that the quality-related activities pertaining to safety-related structures, systems, and components will be identified and controlled. Specific responsibilities for quality-related activities are documented and assigned to personnel and organizations for the major phases of facility design and construction. These responsibilities encompass a broad range of activities and are addressed in this preliminary program. The quality assurance program elements are organized and discussed herein as follows: (1) quality assurance during design and construction; (2) the applicant (DOE); (3) siting contractor; (4) architect/engineer; (5) project field management; and (6) operations contractor

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

Energy Technology Data Exchange (ETDEWEB)

Renfro, David [ORNL; Chandler, David [ORNL; Cook, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant [ORNL; Valentine, Jennifer [ORNL

2014-10-30

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

Energy Technology Data Exchange (ETDEWEB)

Renfro, David G [ORNL; Chandler, David [ORNL; Cook, David Howard [ORNL; Ilas, Germina [ORNL; Jain, Prashant K [ORNL; Valentine, Jennifer R [ORNL

2014-11-01

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present

76 FR 18066 - Regulation of Fuels and Fuel Additives: Changes to Renewable Fuel Standard Program

Science.gov (United States)

2011-04-01

... ENVIRONMENTAL PROTECTION AGENCY 40 CFR Part 80 Regulation of Fuels and Fuel Additives: Changes to Renewable Fuel Standard Program CFR Correction In Title 40 of the Code of Federal Regulations, Parts 72 to...-generating foreign producers and importers of renewable fuels for which RINs have been generated by the...

Fuel channel in-service inspection programs program design for maximum cost effectiveness

International Nuclear Information System (INIS)

Van den Brekel, N.C.

1995-01-01

Inspection is an integral part of fuel channel life management strategy. Inspection data is used to assess the state of reactor core integrity and provide the information necessary to optimize long term maintenance programs. This paper will provide an overview of the structured approach to developing fuel channel inspection programs within OHN. The inspection programs are designed to balance the resources utilized (cost, outage time, and dose expenditure) with the benefits provided by the inspection data obtained (improved knowledge of component status, degradation mechanisms and rates, etc..). The CANDU community has yet to have a fuel channel operate for a full 30 year design life. Since research programs can not fully simulate reactor operating conditions, inspections become an essential feature of the life management strategy as the components age. Inspection programs often include activities designed to develop predictive capability for long term fuel channel behaviour and provide early warning of changes in behaviour. It should be noted that although this paper addresses the design of fuel channel inspection programs, the basic principles presented can be applied to the design of inspection programs for any major power plant component or system. (author)

French development program on fuel cycle

International Nuclear Information System (INIS)

Viala, M.; Bourgeois, M.

1991-01-01

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

Preliminary CFD analysis methodology for flow in a LFR fuel assembly

International Nuclear Information System (INIS)

Catana, A.; Ioan, M.; Serbanel, M.

2013-01-01

In this paper a preliminary Computational Fluid Dynamics (CFD) analysis was performed in order to setup a methodology to be used for more complex coolant flow analysis inside ALFRED nuclear reactor fuel assembly. The core contains 171 separated fuel assembly, each consisting in a hexagonal array of 127 fuel rods. Three honey comb spacer grids are proposed along fuel rods with the aim to keep flow geometry intact during reactor operation. The main goal of this paper is to compute some hydraulic parameters: pressure, velocity, wall shear stress and turbulence parameters with and without spacer grids. In this analysis we consider an adiabatic case, so far no heat transfer is considered but we pave the road toward more complex thermo hydraulic analysis for ALFRED (LFR in general). The CAELINUX CFD distribution was used with its main components: Salome-Meca (for geometry and mesh) and Code-Saturne as mono-phase CFD solver. Paraview and Visist Postprocessors were used for data extraction and graphical displays. (authors)

Metallic fuel design development

International Nuclear Information System (INIS)

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

1999-04-01

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

U-Mo fuel qualification program in HANARO

International Nuclear Information System (INIS)

Lee, K.H.; Lee, C.S.; Kim, H.R.; Kuk, I.H.; Kim, C.K.

2000-01-01

Atomized U-Mo fuel has shown good performance from the results of previous out-of-pile tests and post-irradiation examinations. A qualification program of rod type U-Mo fuel is in progress and the fuel will be irradiated in HANARO. 6 gU/cm 3 U-7Mo, U-8Mo and U-9Mo are considered in this program. The laboratory test results of porosity, mechanical property, thermal conductivity, and thermal compatibility test are discussed in this paper. In parallel with this qualification program, the feasibility study on the core conversion from the present U 3 Si fuel to U-Mo in HANARO will be initiated to provide technical bases for the policy making. Several options of core conversion for HANARO are proposed and each option will be addressed briefly in terms of the operation policy, fuel management, and licensing of HANARO. (author)

DOE Hydrogen & Fuel Cell Overview

Science.gov (United States)

2011-01-13

AUTHOR(S) 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT NUMBER 7. PERFORMING ORGANIZATION NAME(S) AND ADDRESS(ES) U.S. Department of Energy...Overview of Combined Heat+Power PowerElectricity Natural Gas Heat + Cooling Natural Gas or Biogas ...Fuel Cell Technologies Program eere.energy.gov Source: US DOE 10/2010 Biogas Benefits: Preliminary Analysis Stationary fuel

RERTR program progress in qualifying reduced-enrichment fuels

International Nuclear Information System (INIS)

Snelgrove, James L.

1983-01-01

In order to provide the technical means for reducing the enrichment of uranium used to fuel research and test reactors, the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program has been engaged in the development and testing of higher-uranium-density fuels than had been used previously. This fuel development effort included work to increase the density of fuels which were being used at the time the Program began and work on a fuel with the potential for much higher density. The ultimate goal of the fuel development and testing phase of the Program is to 'qualify' the fuel for use. A fuel is considered qualified when a sufficient data base for the fuel exists that it can be approved by regulating bodies for use in reactors. To convert a core to the use of reduced-enrichment fuel it is necessary to show that the core will behave properly during normal and off-normal operating conditions and to show that the fuel will behave properly to a reasonable margin beyond the conditions expected during normal operation. It is this latter area that this paper will address. The main characteristics to be considered in evaluating the performance of a fuel are its swelling, its blister-threshold temperature, and its metallurgical appearance. Data for the qualification of the reduced-enrichment fuels being developed by the RERTR Program are obtained from examination of miniature fuel plates (miniplates) which successfully pass the irradiation screening tests and from examinations of full-sized fuel elements. This paper will summarize the miniplate data reported in other papers presented during this meeting and will give the status of full-sized element irradiations. Finally, the current status of qualification of the various fuel types will be discussed and some projections of the future will be given

1986 fuel cell seminar: Program and abstracts

Energy Technology Data Exchange (ETDEWEB)

None

1986-10-01

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

Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

International Nuclear Information System (INIS)

1990-02-01

The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs

Preliminary assessment of safeguardability on the concepture design of advanced spent fuel conditioning process

Energy Technology Data Exchange (ETDEWEB)

Lee, Sang Yoon; Ha, Jang Ho; Ko, Won Il; Song, Dae Yong; Kim, Ho Dong

2003-04-01

In this report, a preliminary study on the safeguardability of ACP (Advanced spent fuel Conditioning Process) was conducted with Los Alamos National Laboratory. The proposed ACP concept is an electrometallurgical treatment technique to convert oxide-type spent nuclear fuels into metal forms, which can achieve significant reduction of the volume and heat load of spent fuel to be stored and disposed of. For the safeguardability analysis of the ACP facility, sub-processes and their KMPs (Key Measurement Points) were defined first, and then their material flows were analyzed. Finally, the standard deviation of the Inventory Difference (ID) value of the facility was estimated with assumption by assuming international target values for the uncertainty of measurement methods and their uncertainty. From the preliminary calculation, we concluded that if the assumptions regarding measurement instruments can be achieved in a safeguards system for the ACP facility, the safeguards goals of International Atomic Energy Agency (IAEA) could be met. In the second phase of this study, further study on sensitivity analyses considering various factors such as measurement errors, facility capacities, MBA periods etc. may be needed.

Preliminary assessment of safeguardability on the concepture design of advanced spent fuel conditioning process

International Nuclear Information System (INIS)

Lee, Sang Yoon; Ha, Jang Ho; Ko, Won Il; Song, Dae Yong; Kim, Ho Dong

2003-04-01

In this report, a preliminary study on the safeguardability of ACP (Advanced spent fuel Conditioning Process) was conducted with Los Alamos National Laboratory. The proposed ACP concept is an electrometallurgical treatment technique to convert oxide-type spent nuclear fuels into metal forms, which can achieve significant reduction of the volume and heat load of spent fuel to be stored and disposed of. For the safeguardability analysis of the ACP facility, sub-processes and their KMPs (Key Measurement Points) were defined first, and then their material flows were analyzed. Finally, the standard deviation of the Inventory Difference (ID) value of the facility was estimated with assumption by assuming international target values for the uncertainty of measurement methods and their uncertainty. From the preliminary calculation, we concluded that if the assumptions regarding measurement instruments can be achieved in a safeguards system for the ACP facility, the safeguards goals of International Atomic Energy Agency (IAEA) could be met. In the second phase of this study, further study on sensitivity analyses considering various factors such as measurement errors, facility capacities, MBA periods etc. may be needed

Romanian-Canadian joint program for qualification of FCN as a CANDU fuel supplier

International Nuclear Information System (INIS)

Galeriu, C.A.; Andrei, G.; Bailescu, A.

1995-01-01

RENEL (Romania Power Authority), the co-ordinator of Romanian Nuclear Program, have decided to improve, starting 1990 the existing capability to produce CANDU nuclear fuel at FCN Pitesti. The objective of the program was defined with AAC (AECL - ANSALDO Consortium) for the qualification of FCN fuel plant according to Canadian Z299.2 standard. The Qualification Program was performed under AAC Work Order C-003. The co-ordination was assumed by AECL, as overall Design Authority. ZPI (Zircatec Precision Industries Inc., Canada), were designated to supply technical assistance, equipments and know how where necessary. After a preliminary verification of the FCN fuel plant, including the processes and system investigation, performed under AECL and ZPI assistance, the Qualification Program was defined in all details. The upgrading of documentation on all aspects required by Z299.2 was performed. Few processes needed to be reconsidered and equipment was delivered by ZPI or other suppliers. This includes mainly welding equipments and special inspection equipments. Health Physics was practically fully reconsidered. New equipment and practice were adapted to provide adequate control on health conditions. Every manufacturing and inspection process was checked to determine their performance during a Qualification Run based on acceptance criteria which have been established in the Qualification Plan. Manufacturing Demonstration Run was an important step to prove that all plant functions have been accomplished during the fabrication of 200 fuel bundles. These bundles have been fully accepted and 66 of them have been loaded in the first charge of Unit 1 Cemavoda NPS. The surveillance and audit actions made by AECL and ZPI during this period confirmed the FCN capability to operate an adequate system meeting the to required quality assurance standard. The very open attitude of AECL, Zircatec and FCN staff have stimulated the progress of the project and a successful achievement of the

Experimental-demonstrative system for energy conversion using hydrogen fuel cell - preliminary results

International Nuclear Information System (INIS)

Stoenescu, D.; Stefanescu, I.; Patularu, I.; Culcer, M.; Lazar, R.E.; Carcadea, E.; Mirica, D. . E-mail address of corresponding author: daniela@icsi.ro; Stoenescu, D.)

2005-01-01

It is well known that hydrogen is the most promising solution of future energy, both for long and medium term strategies. Hydrogen can be produced using many primary sources (natural gas, methane, biomass, etc.), it can be burned or chemically react having a high yield of energy conversion, being a non-polluted fuel. This paper presents the preliminary results obtained by ICSI Rm. Valcea in an experimental-demonstrative conversion energy system made by a sequence of hydrogen purification units and a CO removing reactors until a CO level lower than 10ppm, that finally feeds a hydrogen fuel stack. (author)

The Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Rummery, T.E.; Rosinger, E.L.J.

1983-05-01

The Canadian Nuclear Fuel Waste Management Program is now well established. This report outlines the generic research and technological development underway in this program to assess the concept of immobilization and subsequent disposal of nuclear fuel waste deep in a stable plutonic rock in the Canadian Shield. The program participants, funding, schedule and associated external review processes are briefly outlined. The major scientific and engineering components of the program, namely, immobilization studies, geoscience research and environmental and safety assessment, are described in more detail

The Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Rummery, T.E.; Rosinger, E.L.J.

1984-12-01

The Canadian Nuclear Fuel Waste Management Program is in the fourth year of a ten-year generic research and development phase. The objective of this phase of the program is to assess the basic safety and environmental aspects of the concept of isolating immobilized fuel waste by deep underground disposal in plutonic rock. The major scientific and engineering components of the program, namely immobilization studies, geoscience research, and environmental and safety assessment, are described. Program funding, scheduling and associated external review processes are briefly outlined

The Canadian fuel waste management program

International Nuclear Information System (INIS)

McConnell, D.B.

1986-04-01

This report is the sixth in the series of annual reports on the research and development program for the safe management and disposal of Canada's nuclear fuel waste. The report summarizes progress in 1984 for the following activities: storage and transportation of used fuel, immobilization of nuclear fuel waste, geotechnical research, environmental research, and environmental and safety assessment. 186 refs

LOFT fuel modules design, characterization, and fabrication program

International Nuclear Information System (INIS)

Russell, M.L.

1977-06-01

The loss-of-fluid test [LOFT) fuel modules have evolved from a comprehensive five-year design, characterization, and fabrication program which has resulted in the accomplishment of many technical activities of interest in pressurized water reactor fuel design development and safety research. Information is presented concerning: determination of fundamental high-temperature reactor material properties; design invention related to in-core instrumentation attachment; implementation of advanced and/or unique fuel bundle characterization techniques; implementation of improved fuel bundle fabrication techniques; and planning and execution of a multimillion dollar design, characterization, and fabrication program for pressurized water reactor fuel

Overview of the US spent nuclear fuel program

International Nuclear Information System (INIS)

Hurt, W.L.

1999-01-01

This report, Overview of the United States Spent Nuclear Fuel Program, December, 1997, summarizes the U.S. strategy for interim management and ultimate disposition of spent nuclear fuel from research and test reactors. The key elements of this strategy include consolidation of this spent nuclear fuel at three sites, preparation of the fuel for geologic disposal in road-ready packages, and low-cost dry interim storage until the planned geologic repository is opened. The U.S. has a number of research programs in place that are intended to Provide data and technologies to support both characterization and disposition of the fuel. (author)

Fuel cell energy service Enron`s commerical program

Energy Technology Data Exchange (ETDEWEB)

Jacobson, M.W.

1996-04-01

Enron, the premier provider of clean fuels worldwide, has launched a unique energy service based on fuel cell technology. The goal of this program is to bring the benefits of fuel cell power to the broad commercial marketplace. Enron`s Energy Service is currently based on a 200 kilowatt phosphoric acid power plant manufactured by ONSI Corporation. This plant is fueled by natural gas or propane, and exhibits superior performance. Enron offers a `no hassle` package that provides customers with immediate benefits with no upfront capital or technical risks. This paper describes Enron`s fuel cell commercial program.

Segmented fuel irradiation program: investigation on advanced materials

International Nuclear Information System (INIS)

Uchida, H.; Goto, K.; Sabate, R.; Abeta, S.; Baba, T.; Matias, E. de; Alonso, J.

1999-01-01

The Segmented Fuel Irradiation Program, started in 1991, is a collaboration between the Japanese organisations Nuclear Power Engineering Corporation (NUPEC), the Kansai Electric Power Co., Inc. (KEPCO) representing other Japanese utilities, and Mitsubishi Heavy Industries, Ltd. (MHI); and the Spanish Organisations Empresa Nacional de Electricidad, S.A. (ENDESA) representing A.N. Vandellos 2, and Empresa Nacional Uranio, S.A. (ENUSA); with the collaboration of Westinghouse. The objective of the Program is to make substantial contribution to the development of advanced cladding and fuel materials for better performance at high burn-up and under operational power transients. For this Program, segmented fuel rods were selected as the most appropriate vehicle to accomplish the aforementioned objective. Thus, a large number of fuel and cladding combinations are provided while minimising the total amount of new material, at the same time, facilitating an eventual irradiation extension in a test reactor. The Program consists of three major phases: phase I: design, licensing, fabrication and characterisation of the assemblies carrying the segmented rods (1991 - 1994); phase II: base irradiation of the assemblies at Vandellos 2 NPP, and on-site examination at the end of four cycles (1994-1999). Phase III: ramp testing at the Studsvik facilities and hot cell PIE (1996-2001). The main fuel design features whose effects on fuel behaviour are being analysed are: alloy composition (MDA and ZIRLO vs. Zircaloy-4); tubing texture; pellet grain size. The Program is progressing satisfactorily as planned. The base irradiation is completed in the first quarter of 1999, and so far, tests and inspections already carried out are providing useful information on the behaviour of the new materials. Also, the Program is delivering a well characterized fuel material, irradiated in a commercial reactor, which can be further used in other fuel behaviour experiments. The paper presents the main

DOE-owned spent nuclear fuel program plan

International Nuclear Information System (INIS)

1995-11-01

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

DOE-owned spent nuclear fuel program plan

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-11-01

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

Update on the Vancouver Fuel Cell Vehicle Program

International Nuclear Information System (INIS)

Rothwell, B.

2004-01-01

'Full text:' The Vancouver Fuel Cell Vehicle Program (VFCVP) is a $5.8 million initiative designed to test four Ford Focus Fuel Cell Vehicles for three years in the Lower Mainland of British Columbia. The project is the first of its kind in Canada and is led by Fuel Cells Canada (FCC), the Ford Motor Company (Ford), and the Governments of Canada and British Columbia. This presentation will provide program details and an update on activities leading up to currently planned delivery to Vancouver in November 2004. The VFCVP will test the performance, durability and reliability of the Ford fuel cell vehicle cars in real-world conditions and will examine fuelling issues and solutions, the reduction of greenhouse gas emissions and public acceptance of hydrogen fuel cell vehicles. The program will generate data to help evolve the technology and develop international codes and standards E cents Epnd the implementation and adoption of fuel cell technology. (author)

Coordinated irradiation plan for the Fuel Refabrication and Development Program

International Nuclear Information System (INIS)

Barner, J.O.

1979-04-01

The Department of Energy's Fuel Refabrication and Development (FRAD) Program is developing a number of proliferation-resistant fuel systems and forms for alternative use in nuclear reactors. A major portion of the program is the development of irradiation behavioral information for the fuel system/forms with the ultimate objective of qualifying the design for licensing and commercial utilization. The nuclear fuel systems under development include denatured thoria--urania fuels and spiked urania--plutonia or thoria--plutonia fuels. The fuel forms being considered include pellet fuel produced from mechanically mixed or coprecipitated feed materials, pellet fuel fabricated from partially calcined gel-derived or freeze-dried spheres (hybrid fuel) and packed-particle fuel produced from sintered gel-derived spheres (sphere-pac). This document describes the coordinated development program that will be used to test and demonstrate the irradiation performance of alternative fuels

Romanian nuclear fuel program: past, present and future

International Nuclear Information System (INIS)

Budan, O.; Rotaru, I.; Galeriu, C.A.

1997-01-01

The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. In this paper the word 'past' refers to the period before 1990 and 'present' to the 1990-1997 period. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. - ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified

GCRA review and appraisal of fuel material development programs

International Nuclear Information System (INIS)

1980-09-01

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

Fuel Cell Demonstration Program

Energy Technology Data Exchange (ETDEWEB)

Gerald Brun

2006-09-15

In an effort to promote clean energy projects and aid in the commercialization of new fuel cell technologies the Long Island Power Authority (LIPA) initiated a Fuel Cell Demonstration Program in 1999 with six month deployments of Proton Exchange Membrane (PEM) non-commercial Beta model systems at partnering sites throughout Long Island. These projects facilitated significant developments in the technology, providing operating experience that allowed the manufacturer to produce fuel cells that were half the size of the Beta units and suitable for outdoor installations. In 2001, LIPA embarked on a large-scale effort to identify and develop measures that could improve the reliability and performance of future fuel cell technologies for electric utility applications and the concept to establish a fuel cell farm (Farm) of 75 units was developed. By the end of October of 2001, 75 Lorax 2.0 fuel cells had been installed at the West Babylon substation on Long Island, making it the first fuel cell demonstration of its kind and size anywhere in the world at the time. Designed to help LIPA study the feasibility of using fuel cells to operate in parallel with LIPA's electric grid system, the Farm operated 120 fuel cells over its lifetime of over 3 years including 3 generations of Plug Power fuel cells (Lorax 2.0, Lorax 3.0, Lorax 4.5). Of these 120 fuel cells, 20 Lorax 3.0 units operated under this Award from June 2002 to September 2004. In parallel with the operation of the Farm, LIPA recruited government and commercial/industrial customers to demonstrate fuel cells as on-site distributed generation. From December 2002 to February 2005, 17 fuel cells were tested and monitored at various customer sites throughout Long Island. The 37 fuel cells operated under this Award produced a total of 712,635 kWh. As fuel cell technology became more mature, performance improvements included a 1% increase in system efficiency. Including equipment, design, fuel, maintenance

Metallic fuel design development

Energy Technology Data Exchange (ETDEWEB)

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

1999-04-01

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

GCFR Fuels and Materials Program at Argonne National Laboratory

International Nuclear Information System (INIS)

Neimark, L.A.; Greenberg, S.; Johnson, C.E.; Purohit, A.; Liu, Y.Y.; Rest, J.; Reimann, K.J.; McLennan, G.A.

1979-01-01

The F-5 fuel-pin irradiation experiment in EBR-II is a cornerstone of the GCFR program. It is the largest-scale fuel-pin experiment in the present program and will provide data on the performance of pins and a pin-support structure that are prototypic of the GCFR Demonstration Plant. The fuel pins are presently undergoing interim examination after successfully achieving 4.6 at.% burnup. Studies of the thermodynamics and kinetics of the U--Cs--O system, supplemented by analysis of the results of previously irradiated fuel pins, have led to the incorporation of fuel-design modifications in the F-5 experiment to insure adequate performance of the vented fuel. The effect of ribbing, as well as the ribbing process, on the short- and long-term structural performance of fuel-pin cladding is being evaluated via in-reactor and out-of-reactor tests and with the fuel-element modeling code LIFE-GCFR and the finite element program, ADINA

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.

Preliminary Evaluation of Removing Used Nuclear Fuel From Nine Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J.; Best, Ralph; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul

2013-04-30

The Blue Ribbon Commission on America’s Nuclear Future identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses. In this report, a preliminary evaluation of removing used nuclear fuel from nine shutdown sites was conducted. The shutdown sites included Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion. At these sites a total of 7649 used nuclear fuel assemblies and a total of 2813.2 metric tons heavy metal (MTHM) of used nuclear fuel are contained in 248 storage canisters. In addition, 11 canisters containing greater-than-Class C (GTCC) low-level radioactive waste are stored at these sites. The evaluation was divided in four components: • characterization of the used nuclear fuel and GTCC low-level radioactive waste inventory at the shutdown sites • an evaluation of the onsite transportation conditions at the shutdown sites • an evaluation of the near-site transportation infrastructure and experience relevant to the shipping of transportation casks containing used nuclear fuel from the shutdown sites • an evaluation of the actions necessary to prepare for and remove used nuclear fuel and GTCC low-level radioactive waste from the shutdown sites. Using these evaluations the authors developed time sequences of activities and time durations for removing the used nuclear fuel and GTCC low-level radioactive waste from a single shutdown site, from three shutdown sites located close to each other, and from all nine shutdown sites.

A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

International Nuclear Information System (INIS)

Andress, D.; Joy, D.S.; McLeod, N.B.; Peterson, R.W.; Rahimi, M.

1991-01-01

The Department of Energy has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design--the GA-4 and GA-9 truck casks and the BR-100 rail cask. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elements as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. This paper summarizes preliminary results of one part of that evaluation related to the ability of the From-Reactor Initiative I casks to accommodate the physical and radiological characteristics of the Spent Nuclear Fuel projected to be accepted into the Federal Waste Management System. 3 refs., 5 tabs

14 CFR 129.113 - Fuel tank system maintenance program.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Fuel tank system maintenance program. 129... Continued Airworthiness and Safety Improvements § 129.113 Fuel tank system maintenance program. (a) Except... on which an auxiliary fuel tank is installed under a field approval, before June 16, 2008, the...

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.

2013-09-30

This report fulfills the M2 milestone M2FT-13PN0912022, “Stranded Sites De-Inventorying Report.” In January 2013, the U.S. Department of Energy (DOE) issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013). Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of the Blue Ribbon Commission on America’s Nuclear Future, which identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses (BRC 2012). Shutdown sites are defined as those commercial nuclear power reactor sites where the nuclear power reactors have been shut down and the site has been decommissioned or is undergoing decommissioning. In this report, a preliminary evaluation of removing used nuclear fuel from 12 shutdown sites was conducted. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. These sites have no other operating nuclear power reactors at their sites and have also notified the U.S. Nuclear Regulatory Commission that their reactors have permanently ceased power operations and that nuclear fuel has been permanently removed from their reactor vessels. Shutdown reactors at sites having other operating reactors are not included in this evaluation.

14 CFR 125.507 - Fuel tank system inspection program.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Fuel tank system inspection program. 125... Airworthiness and Safety Improvements § 125.507 Fuel tank system inspection program. (a) Except as provided in... fuel tank is installed under a field approval, before June 16, 2008, the certificate holder must submit...

14 CFR 91.1507 - Fuel tank system inspection program.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 2 2010-01-01 2010-01-01 false Fuel tank system inspection program. 91... Airworthiness and Safety Improvements § 91.1507 Fuel tank system inspection program. (a) Except as provided in... fuel tank is installed under a field approval, before June 16, 2008, the operator must submit to the...

75 FR 15893 - Tire Fuel Efficiency Consumer Information Program

Science.gov (United States)

2010-03-30

... how much the proposed consumer information program would affect consumer tire purchasing behavior and... 575 Tire Fuel Efficiency Consumer Information Program; Final Rule #0;#0;Federal Register / Vol. 75, No... 2127-AK45 Tire Fuel Efficiency Consumer Information Program AGENCY: National Highway Traffic Safety...

Preliminary Design Analysis of a HGD for the NHDD Program at Korea

International Nuclear Information System (INIS)

Song, Kee Nam; Lee, H. Y.; Lee, S. B.; Kim, Y. W.

2007-01-01

Korea Atomic Energy Research Institute is in the process of carrying out a Nuclear Hydrogen Development and Demonstration (NHDD) Program by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950 .deg. C. A coaxial double-tube Hot Gas Duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger for the NHDD program. Recently, a preliminary design evaluation for the hot gas duct of the NHDD program was carried out. These preliminary design activities include a decision on the geometric dimensions, a strength evaluation, an appropriate material selection, and identifying the design code for the HGD. In this study, a preliminary strength evaluation for the HGD of the NHDD program has been undertaken based on the HTR-10 design concepts. Also, a preliminary evaluation of the creep-fatigue damage for a high temperature HGD structure has been carried out according to the draft code case for Alloy 617. Preliminary strength evaluation results for the HGD showed that the geometric dimensions of the proposed HGD would be acceptable for the design requirements

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.

Results of 200 KW fuel cell evaluation programs

Energy Technology Data Exchange (ETDEWEB)

Torrey, J.M.; Merten, G.P. [SAIC, San Diego, CA (United States); Binder, M.J. [Army Construction Engineering Research Labs., Champaign, IL (United States)] [and others

1996-12-31

Science Applications International Corporation (SAIC) has installed six monitoring systems on ONSI Corporation 200 kW phosphoric acid fuel cells. Three of the systems were installed for the U.S. Army Construction Engineering Research Laboratories (USACERL) which is coordinating the Department of Defense (DoD) fuel cell Demonstration Program and three were installed under a contract with the New York State Energy Research and Development Authority (NYSERDA). Monitoring of the three NYSERDA sites has been completed. Monitoring systems for the DoD fuel cells were installed in August, 1996 and thus no operating data was available at the time of this writing, but will be presented at the Fuel Cell Seminar. This paper will present the monitoring configuration and research approach for each program. Additionally, summary performance data is presented for the completed NYSERDA program.

Role of the consolidated fuel reprocessing program in the United States Breeder Reactor Program

International Nuclear Information System (INIS)

Ballard, W.W.; Burch, W.D.

1980-01-01

While present US policy precludes the commercial reprocessing of LWR fuels and the recycle of plutonium, the policy does encompass the need to continue a program to develop the technology for reprocessing breeder fuels. Some questions have again risen this year as to the pace of the entire breeder program, including recycle, and the answers are evolving. This paper and the other companion papers which describe several aspects of the Consolidated Fuel Reprocessing Program take a longer-range perspective on the total program. Whether the program is implemented in the general time frame described is dependent on future government actions dedicated to carrying out a systematic program that would permit breeders to be commercialized early in the next century

14 CFR 121.1113 - Fuel tank system maintenance program.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Fuel tank system maintenance program. 121... Improvements § 121.1113 Fuel tank system maintenance program. (a) Except as provided in paragraph (g) of this... capacity of 7500 pounds or more. (b) For each airplane on which an auxiliary fuel tank is installed under a...

The low-enrichment fuel development program

International Nuclear Information System (INIS)

Stahl, D.

1993-01-01

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

Review of the nuclear fuel waste management program

International Nuclear Information System (INIS)

Hatcher, S.R.

1980-06-01

Progress over the previous year in the nuclear fuel waste management program is reviewed. Universities, industry and consultants have become increasingly involved, and the work is being overseen by a Technical Advisory Committee. The program has also been investigated by Ontario's Porter Commission and Select Committe on Ontario Hydro Affairs. A public information program has been extended to cover most of the Canadian Shield region of Ontario. Ontario Hydro is studying spent fuel storage and transportation, while AECL is covering immobilization of spent fuel or processing wastes, geotechnical and geochemical research in the laboratory and in the field, design of disposal facilities, and environmental and safety assessments. (L.L.)

Oak Ridge TNS program: study of fueling techniques in support of TNS development

International Nuclear Information System (INIS)

Simpson, W.

1979-12-01

The objective of this study was to survey fueling techniques and determine which approaches were compatible with the TNS requirements. Specifically, the following tasks were undertaken: survey of existing fueling concepts for use in TNS, determination of available physical properties for D-T fuel pellets, performance of preliminary load analysis of selected pellet acceleration machines, preparation of conceptual designs, and recommendations for follow-up work

Alternative bio-based fuels for aviation: the clean airports program

International Nuclear Information System (INIS)

Shauck, M.E.; Zanin, M.G.

1997-01-01

The Renewable Aviation Fuels Development Center at Baylor University in Waco, Texas, has been designated as the national coordinator of the Clean Airports Program. The U.S. Dept. of Energy (US DOE) conferred this designation in March 1996. This program, a spin-off of the Clean Cities Program, was initiated to increase the use of alternative fuels in aviation. The two major fuels used in aviation are the current piston engine aviation gasoline and the current turbine engine fuel. The environmental impact of each of these fuels is significant. Aviation gasoline (100LL), currently used in the general aviation piston engine fleet, contributes 100% of the emissions containing lead in the U.S. today. Turbine engine fuel (jet fuel) produces two major environmental impacts: a local one, in the vicinity of the airports, and a global impact on climate change. The Clean Airports Program was established to achieve and maintain clean air at and in the vicinity of airports, through the use of alternative fuel-powered air and ground transportation vehicles. (author)

Alternative transportation fuels in the USA: government hydrogen vehicle programs

International Nuclear Information System (INIS)

Cannon, J.S.

1993-01-01

The linkage between natural gas-based transportation and hydrogen-based transportation strategies, two clean burning gaseous fuels, provides a strong policy rationale for increased government sponsorship of hydrogen vehicle research and demonstration programs. Existing federal and state government hydrogen vehicle projects are discussed in this paper: research at the NREL, alternate-fueled buses, Renewable Hydrogen for the State of Hawaii program, New York state alternative transportation fuels program, Colorado program. 9 refs

The Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

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

1984-04-01

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

Fuel Flexible Turbine System (FFTS) Program

Energy Technology Data Exchange (ETDEWEB)

None, None

2012-12-31

In this fuel flexible turbine system (FFTS) program, the Parker gasification system was further optimized, fuel composition of biomass gasification process was characterized and the feasibility of running Capstone MicroTurbine(TM) systems with gasification syngas fuels was evaluated. With high hydrogen content, the gaseous fuel from a gasification process of various feed stocks such as switchgrass and corn stover has high reactivity and high flashback propensity when running in the current lean premixed injectors. The research concluded that the existing C65 microturbine combustion system, which is designed for natural gas, is not able to burn the high hydrogen content syngas due to insufficient resistance to flashback (undesired flame propagation to upstream within the fuel injector). A comprehensive literature review was conducted on high-hydrogen fuel combustion and its main issues. For Capstone's lean premixed injector, the main mechanisms of flashback were identified to be boundary layer flashback and bulk flow flashback. Since the existing microturbine combustion system is not able to operate on high-hydrogen syngas fuels, new hardware needed to be developed. The new hardware developed and tested included (1) a series of injectors with a reduced propensity for boundary layer flashback and (2) two new combustion liner designs (Combustion Liner Design A and B) that lead to desired primary zone air flow split to meet the overall bulk velocity requirement to mitigate the risk of core flashback inside the injectors. The new injector designs were evaluated in both test apparatus and C65/C200 engines. While some of the new injector designs did not provide satisfactory performance in burning target syngas fuels, particularly in improving resistance to flashback. The combustion system configuration of FFTS-4 injector and Combustion Liner Design A was found promising to enable the C65 microturbine system to run on high hydrogen biomass syngas. The FFTS-4 injector

Current status of spent fuel disposal program in Taiwan, Republic of China

International Nuclear Information System (INIS)

Soong, K.L.; Liu, S.J.

1989-01-01

In the year of 1988, Taiwan has completed a two-year preliminary study and began in late 1988 a second term for the final disposal of spent fuel program. The research conducted in the first phase was mainly concentrated on the reviews of international studies and domestic geological literatures, set-up of siting criteria, and pertinent engineering analysis of a repository. Taiwan is an area of complex and unstable geological structures, abundant ground water and high density of population. Favorable host rocks under consideration for permanent waste disposal are thick shales and mudstones, metamorphosed rocks of mesozoic basement, solidified quartzite and mesozoic granitic gneiss. The analysis of heat transfer for hypothetical case studies indicate that the temperature rises in the repository systems would be well below the proposed maximum admissible temperatures. The repository size would be less than 2 km 2 when appropriate period for surface storage of spent fuel is satisfied. A probabilistic risk analysis also demonstrates that the presence of faulting and other tectonic instabilities characterize Taiwan's disadvantages for geological disposal of nuclear waste

Consolidated fuel reprocessing. Program progress report, April 1-June 30, 1980

Energy Technology Data Exchange (ETDEWEB)

1980-09-01

This progress report is compiled from major contributions from three programs: (1) the Advanced Fuel Recycle Program at ORNL; (2) the Converter Fuel Reprocessing Program at Savannah River Laboratory; and (3) the reprocessing components of the HTGR Fuel Recycle Program, primarily at General Atomic and ORNL. The coverage is generally overview in nature; experimental details and data are limited.

Research and development of FBR fuel reprocessing in PNC

International Nuclear Information System (INIS)

Hoshino, T.

1976-05-01

The research program of the PNC for FBR fuel reprocessing in Japan is discussed. The general characteristics of FBR fuel reprocessing are pointed out and a comparison with LWR fuel is made. The R and D program is based on reprocessing using the aqueous Purex process. So far, some preliminary steps of the research program have been carried out, these include solvent extraction test, off-gas treatment test, voloxidation process study, solidification test of high-level liquid waste, and study of the dissolution behaviour of irradiated mixed oxide fuel. By the end of the 1980s, a pilot plant for FBR fuel reprocessing will be completed. For the design of the pilot plant, further research will be carried out in the following fields: head-end techniques; voloxidation process; dissolution and extraction techniques; waste treatment techniques. A time schedule for the different steps of the program is included

Preliminary neutronic assessment for ATF (Accident Tolerant Fuel) based on iron alloy

International Nuclear Information System (INIS)

Abe, Alfredo; Carluccio, Thiago; Piovezan, Pamela; Giovedi, Claudia; Martins, Marcelo R.

2015-01-01

After Fukushima Daiichi nuclear accident in 2011, the nuclear fuel performance under accident condition became a very important issue and currently different research and development program are in progress toward to reliability and withstand under accident condition. These initiatives are known as ATF (Accident Tolerant Fuel) R and D program, which many countries with different research institutes, fuel vendors and others are nowadays involved. Accident Tolerant Fuel (ATF) can be defined as enhanced fuel which can tolerate loss of active cooling system capability for a considerably longer time period and the fuel/cladding system can be maintained without significant degradation and can also improve the fuel performance during normal operations and transients, as well as design-basis accident (DBA) and beyond design-basis (BDBA) accident. Different materials have being proposed as fuel cladding candidates considering thermo-mechanical properties and lower reaction kinetic with steam and slower hydrogen production. The aim of this work is to perform a neutronic assessment for several cladding candidates based on iron alloy considering a standard PWR fuel rod (fuel pellet and dimension). The purpose of the assessment is to address different parameters that might contribute for possible neutronic reactivity gain in order to overcome the penalty due to increase of neutron absorption in the cladding materials. All the neutronic assessment is performed using MCNP, Monte Carlo code. (author)

Program on MOX fuel utilization in light water reactors

International Nuclear Information System (INIS)

Kenda, Hirofumi

2000-01-01

MOX fuel utilization program by the Japanese electric power companies was released in February, 1997. Principal philosophy for MOX fuel design is that MOX fuel shall be compatible with Uranium fuel and behavior of core loaded with MOX fuel shall be similar to that of conventional core. MOX fuel is designed so that geometry and nuclear capability of MOX fuel are equivalent to Uranium fuel. (author)

Analysis of transient fuel failure mechanisms: selected ANL programs

International Nuclear Information System (INIS)

Deitrich, L.W.

1975-01-01

Analytical programs at Argonne National Laboratory related to fuel pin failure mechanisms in fast-reactor accident transients are described. The studies include transient fuel pin mechanics, mechanics of unclad fuel, and mechanical effects concerning potential fuel failure propagation. (U.S.).

Computer simulation of HTGR fuel microspheres using a Monte-Carlo statistical approach

International Nuclear Information System (INIS)

Hedrick, C.E.

1976-01-01

The concept and computational aspects of a Monte-Carlo statistical approach in relating structure of HTGR fuel microspheres to the uranium content of fuel samples have been verified. Results of the preliminary validation tests and the benefits to be derived from the program are summarized

A cost analysis of Colorado's 1991-92 oxygenated fuels program

International Nuclear Information System (INIS)

Manderino, L.A.; Bowles, S.L.

1993-01-01

This paper discusses the methodology used to conduct a cost analysis of Colorado's 1991-92 Oxygenated Fuels Program. This program requires the use of oxygenated fuels during the winter season in Denver and surrounding areas. The cost analysis was conducted as part of an overall cost-effectiveness study of the 1991-92 program conducted by PRC Environmental Management, Inc. (PRC). The paper, however, focuses on cost analysis and does not consider potential benefits of the program. The study analyzed costs incurred by different segments of society, including government, industry, and consumers. Because the analysis focused on a specific program year, neither past nor future costs were studied. The discussion of government costs includes the agencies interviewed and the types of costs associated with government administration and enforcement of the program. The methodology used to calculate costs to private industry is also present. The study examined the costs to fuel refineries, pipelines, and blenders, as well as fuel retailers and automobile fleet operators. Finally, the paper discusses the potential costs incurred by the consumer purchasing oxygenated fuels. Costs associated with issues such as vehicle driveability, automobile parts durability and performance, and fuel economy are also examined. A summary of all costs by category is presented along with an analysis of the major cost components. These include costs which are sensitive to specific circumstances and which may vary among programs

Development and engineering plan for graphite spent fuels conditioning program

International Nuclear Information System (INIS)

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

1993-09-01

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

The US Army Foreign Comparative Test fuel cell program

Science.gov (United States)

Bostic, Elizabeth; Sifer, Nicholas; Bolton, Christopher; Ritter, Uli; Dubois, Terry

The US Army RDECOM initiated a Foreign Comparative Test (FCT) Program to acquire lightweight, high-energy dense fuel cell systems from across the globe for evaluation as portable power sources in military applications. Five foreign companies, including NovArs, Smart Fuel Cell, Intelligent Energy, Ballard Power Systems, and Hydrogenics, Inc., were awarded competitive contracts under the RDECOM effort. This paper will report on the status of the program as well as the experimental results obtained from one of the units. The US Army has interests in evaluating and deploying a variety of fuel cell systems, where these systems show added value when compared to current power sources in use. For low-power applications, fuel cells utilizing high-energy dense fuels offer significant weight savings over current battery technologies. This helps reduce the load a solider must carry for longer missions. For high-power applications, the low operating signatures (acoustic and thermal) of fuel cell systems make them ideal power generators in stealth operations. Recent testing has been completed on the Smart Fuel Cell A25 system that was procured through the FCT program. The "A-25" is a direct methanol fuel cell hybrid and was evaluated as a potential candidate for soldier and sensor power applications.

Spent fuel disassembly and canning programs at the Barnwell Nuclear Fuel Plant (BNFP)

International Nuclear Information System (INIS)

Townes, G.A.

1979-10-01

Methods of disassembling and canning spent fuel to allow more efficient storage are being investigated at the BNFP. Studies and development programs are aimed at dry disassembly of fuel to allow storage and shipment of fuel pins rather than full fuel assemblies. Results indicate that doubling existing storage capacity or tripling the carrying capacity of existing transportation equipment is achievable. Disassembly could be performed in the BNFP hot cells at rates of about 12 to 15 assemblies per day

Methanol supply issues for alternative fuels demonstration programs

International Nuclear Information System (INIS)

Teague, J.M.; Koyama, K.K.

1995-01-01

This paper surveys issues affecting the supply of fuel-grade methanol for the California Energy Commission's alternative fuels demonstration programs and operations by other public agencies such as transit and school districts. Establishing stable and reasonably priced sources of methanol (in particular) and of alternative fuels generally is essential to their demonstration and commercialization. Development both of vehicle technologies and of fuel supply and distribution are complementary and must proceed in parallel. However, the sequence of scaling up supply and distribution is not necessarily smooth; achievement of volume thresholds in demand and through-put of alternative fuels are marked by different kinds of challenges. Four basic conditions should be met in establishing a fuel supply: (1) it must be price competitive with petroleum-based fuels, at least when accounting for environmental and performance benefits; (2) bulk supply must meet volumes required at each phase; necessitating resilience among suppliers and a means of designating priority for high value users; (3) distribution systems must be reliable, comporting with end users' operational schedules; (4) volatility in prices to the end user for the fuel must be minimal. Current and projected fuel volumes appear to be insufficient to induce necessary economies of scale in production and distribution for fuel use. Despite their benefits, existing programs will suffer absent measures to secure economical fuel supplies. One solution is to develop sources that are dedicated to fuel markets and located within the end-use region

Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

Energy Technology Data Exchange (ETDEWEB)

Ketusky, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-08-09

The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

National fuel cell bus program : proterra fuel cell hybrid bus report, Columbia demonstration.

Science.gov (United States)

2011-10-01

This report summarizes the experience and early results from a fuel cell bus demonstration funded by the Federal Transit Administration (FTA) under the National Fuel Cell Bus Program. A team led by the Center for Transportation and the Environment an...

Renewable Fuel Pathways II Final Rule to Identify Additional Fuel Pathways under Renewable Fuel Standard Program

Science.gov (United States)

This final rule describes EPA’s evaluation of biofuels derived from biogas fuel pathways under the RFS program and other minor amendments related to survey requirements associated with ULSD program and misfueling mitigation regulations for E15.

Preliminary design report for prototypical spent nuclear fuel rod consolidation equipment

International Nuclear Information System (INIS)

Judson, B.F.; Maillet, J.; O'Neill, G.L.; Tsitsichvili, J.; Tucoulat, D.

1986-12-01

The purpose of the Prototypical Consolidation Demonstration Project (PCDP) is to develop and demonstrate the equipment system that will be used to consolidate the bulk of the spent nuclear fuel generated in the United States prior to its placement in a geological repository. The equipment must thus be capable of operating on a routine production basis over a long period of time with stringent requirements for safety, reliability, productivity and cost-effectiveness. Four phases are planned for the PCDP. Phase 1 is the Preliminary Design of generic consolidation equipment that could be installed at a Monitored Retrievable Storage (MRS) facility or in the Receiving ampersand Handling Facility at a geologic repository site. Phase 2 will be the Final Design and preparation of procurement packages for the equipment in a configuration capable of being installed and tested in a special enclosure within the TAN Hot Shop at DOE's Idaho National Engineering Laboratory. In Phase 3 the equipment will be fabricated and then tested with mock fuel elements in a contractor's facility. Finally, in Phase 4 the equipment will be moved to the TAN facility for demonstration operation with irradiated spent fuel elements. 55 figs., 15 tabs

Mathematical simulation of stressed-deformed state in rod cylindrical fuel elemnts KONDOR program

International Nuclear Information System (INIS)

Khmelevskij, M.Ya.; Malakhova, E.I.; Dolmatov, P.S.

1987-01-01

A mathematical model for numerical computation of stressed-deformed stae in a rod cylindrical fuel element is developed. The model is based on preliminary discretization of the design scheme and linearization of radial parameters as radius functions. The formulation generality enables to calculate strength parameter kinetics in any circular cylindrical fuel element (e.g. annular fuel element; solid or tubular core; ceramic, metallic or dispersion fuel) for arbitrary transient operating conditions and taking into account all possible loading factors. The method is realized in the KONDOR programm (FORTRAN, ES-1061 computer). An example illustrating computation of stress kinetics in a fast reactor fuel element during transient operation is given

Preliminary studies on readiness of biojet fuel for commercial aviation: The feasibility and potential in Malaysia

Science.gov (United States)

Noh, H. Mohd; Mahammad Taher, M. N.; Rodrigo, G. A.; Rahman, N. A. Abdul; Othman, J.; Yahaya, N. H. R.

2017-12-01

This paper demonstrates the need for a new alternative energy using biojet fuel in commercial aviation. The demand of air travels leads the authority, airlines and government in seeking for new renewable and sustainable energy for aircraft operation in the future. This study looks into the level of readiness in using biofuel. 40 personnel who are working in the aviation industries have participated and completed the survey questionnaires. The preliminary findings suggest that the impact towards this new fuel will lead to a better environment, less cost, better maintenance and energy sustainability. The usage of biojet fuel seems possible to be pursued in Malaysia.

Career outcomes of nondesignated preliminary general surgery residents at an academic surgical program.

Science.gov (United States)

Ahmad, Rima; Mullen, John T

2013-01-01

There remains a debate as to whether nondesignated preliminary (NDP) positions in surgery ultimately translate into successful surgical careers for those who pursue them. We sought to identify the success with which our NDP residents were able to transition to their desired career and what, if any, factors contributed to their success. The records of all NDP residents accepted into the Massachusetts General Hospital General Surgery Residency Program from 1995 to 2010 were examined and long-term follow-up was completed. Thirty-four NDP residents were identified, including 26.5% US graduates and 73.5% international medical graduates. At the end of the initial preliminary year, 30 (88%) got placed in a postgraduate residency program, whereas 4 (12%) pursued other career paths. Of those who got placed, 25 (83%) attained surgical residency positions, including 17 (57%) who continued as preliminary residents at our institution and 8 (27%) who got placed in categorical surgical positions at other programs. After multiple preliminary years, 15 of 17 achieved a categorical position, of which, 93% were in surgical fields. Overall, 64.7% of all entering NDP residents eventually went on to have careers in general surgery (50%) or surgical subspecialties (14.7%), and 24 of 34 (71%) fulfilled their desired career goals. No factor predicted success. From 1995 to 2012 there have been 15 midlevel (11 postgraduate year 4) vacancies in our program, 4 of which were filled by preliminary residents, 2 from our program and 2 from elsewhere. All have gone on to board certifications and careers in surgery. More than 70% of NDP residents in our program successfully transitioned to their desired career paths, many achieving categorical surgical positions and academic surgical careers, thus demonstrating the benefit of this track to both residency programs and trainees. © 2013 Association of Program Directors in Surgery. Published by Elsevier Inc. All rights reserved.

Nuclear criticality safety program at the Fuel Cycle Facility

International Nuclear Information System (INIS)

Lell, R.M.; Fujita, E.K.; Tracy, D.B.; Klann, R.T.; Imel, G.R.; Benedict, R.W.; Rigg, R.H.

1994-01-01

The Fuel Cycle Facility (FCF) is designed to demonstrate the feasibility of a novel commercial-scale remote pyrometallurgical process for metallic fuels from liquid metal-cooled reactors and to show closure of the Integral Fast Reactor (IFR) fuel cycle. Requirements for nuclear criticality safety impose the most restrictive of the various constraints on the operation of FCF. The upper limits on batch sizes and other important process parameters are determined principally by criticality safety considerations. To maintain an efficient operation within appropriate safety limits, it is necessary to formulate a nuclear criticality safety program that integrates equipment design, process development, process modeling, conduct of operations, a measurement program, adequate material control procedures, and nuclear criticality analysis. The nuclear criticality safety program for FCF reflects this integration, ensuring that the facility can be operated efficiently without compromising safety. The experience gained from the conduct of this program in the Fuel cycle Facility will be used to design and safely operate IFR facilities on a commercial scale. The key features of the nuclear criticality safety program are described. The relationship of these features to normal facility operation is also described

Alternative concepts for spent fuel storage basin expansion at Morris Operation

International Nuclear Information System (INIS)

Graf, W.A. Jr.; King, C.E.; Miller, G.P.; Shadel, F.H.; Sloat, R.J.

1980-08-01

Alternative concepts for increasing basin capabilities for storage of spent fuel at the Morris Operation have been defined in a series of simplified flow diagrams and equipment schematics. Preliminary concepts have been outlined for (1) construction alternatives for an add-on basin, (2) high-density baskets for storage of fuel bundles or possible consolidated fuel rods in the existing or add-on basins, (3) modifications to the existing facility for increasing cask handling and fuel receiving capabilities and (4) accumulation, treatment and disposal of radwastes from storage operations. Preliminary capital and operating costs have been prepared and resource and schedule requirements for implementing the concepts have been estimated. The basin expansion alternatives would readily complement potential dry storage projects at the site in an integrated multi-stage program that could provide a total storage capacity of up to 7000 tonnes of spent fuel

Discussion paper: direction for Canada's alternate fuels program

Energy Technology Data Exchange (ETDEWEB)

1982-09-01

There is a growing need to accelerate the consideration of alternate fuels for use in Canadian vehicle transportation. At the present time various governments and corporations are initiating alternate fuel programs involving ethanol, methanol, CNG, propane, etc. There is a bewildering array of perspectives as to which fuel or fuels will best serve Canada's needs in the future. In response to the 'Discussion Paper on Liquid Fuels Options, 1980', by the Federal Dept. of Energy, Mines and Resources, Ford of Canada has prepared this perspective on each of the alternate fuels from the company's vantage point as a vehicle manufacturer.

Fuel-coolant interactions: preliminary experiments on the effect of gases dissolved in the 'coolant'

International Nuclear Information System (INIS)

Asher, R.C.; Davies, D.; Jones, P.G.

1976-12-01

A simple apparatus has been used to study fuel-coolant interactions under reasonably well controlled conditions. Preliminary experiments have used water as the 'coolant' and molten tin at 800 0 C as the 'fuel' and have investigated how the violence of the interaction is affected by dissolving gases (oxygen, nitrogen, carbon dioxide and nitrous oxide) in the water. It was found that saturating the water with carbon dioxide or nitrous oxide completely suppresses the violent interaction. Experiments in which the concentrations of these gases were varied showed that a certain critical concentration was needed; below this concentration the dissolved gas has no significant effect but above it the suppression is

Spent fuel receipt and lag storage facility for the spent fuel handling and packaging program

International Nuclear Information System (INIS)

Black, J.E.; King, F.D.

1979-01-01

Savannah River Laboratory (SRL) is participating in the Spent Fuel Handling and Packaging Program for retrievable, near-surface storage of spent light water reactor (LWR) fuel. One of SRL's responsibilities is to provide a technical description of the wet fuel receipt and lag storage part of the Spent Fuel Handling and Packaging (SFHP) facility. This document is the required technical description

Characterization of Zircaloy-4 tubing procured for fuel cladding research programs

International Nuclear Information System (INIS)

Chapman, R.H.

1976-01-01

A quantity of Zircaloy-4 tubing [10.92 mm outside diameter by 0.635 mm wall thickness] was purchased specifically for use in a number of related fuel cladding research programs sponsored by the Division of Reactor Safety Research, Nuclear Regulatory Commission (NRC/RSR). Identical tubing (produced simultaneously and from the same ingot) was purchased concurrently by the Electric Power Research Institute (EPRI) for use in similar research programs sponsored by that organization. In this way, source variability and prior fabrication history were eliminated as parameters, thus permitting direct comparison (as far as as-received material properties are concerned) of experimental results from the different programs. The tubing is representative of current reactor technology. Consecutive serial numbers assigned to each tube identify the sequence of the individual tubes through the final tube wall reduction operation. The report presented documents the procurement activities, provides a convenient reference source of manufacturer's data and tubing distribution to the various users, and presents some preliminary characterization data. The latter have been obtained routinely in various research programs and are not complete. Although the number of analyses, tests, and/or examinations performed to date are insufficient to draw statistically valid conclusions with regard to material characterization, the data are expected to be representative of the as-received tubing. It is anticipated that additional characterizations will be performed and reported routinely by the various research programs that use the tubing

The French development program for a UMo fuel

International Nuclear Information System (INIS)

Romano, R.; Nigon, J.L.; Languille, A.; Le Borgne, E.; Freslon, H.

1999-01-01

Until now high density U 3 Si 2 fuels were satisfactory for LEU conversion of certain reactors, but their use is limited because their density is physically limited to 5,8 gU/cm3 and they have very poor reprocessing capacities. After the end of the present US return policy in may 2006, the reactor operators will be indeed in a very difficult position with silicides. The international community is thus interested in a very high density fuel with good reprocessing capacities in order to convert most reactors and to find a back end solution. In France, CEA, CERCA, and COGEMA have thus launched an important program in order to sort potential candidates of uranium alloys. UMo is one of the most interesting candidates. After the selection of UMo alloys, France has pooled different skills to start an important program on UMo fuels: CEA has started an important project for a new reactor (Jules Horowitz); CERCA is the main manufacturer for MTR fuel; TECHNICATOME is the design expert for research reactors and associated cores; FRAMATOME is the parent company of CERCA and is interested in the development of new reactors; COGEMA is interested in reprocessing spent fuels. This new fuel has three aims: to allow reactors to benefit from a high performing fuel; to have a reprocessable fuel to limit the fuel storage period and the associate safety problem, and solve the back end issue; to support the international effort for non proliferation involving the end of the use of HEU. This high density fuel will decrease the number of fuel assemblies needed to run the reactors and decrease the global cost of the fuel cycle as the back end management cost is in proportion with the quantity of fuel. Reactor operators will thus derive an advantage from this new fuel, in terms of economy

Status of the Canadian Nuclear Fuel Waste Management Program

International Nuclear Information System (INIS)

Lyon, R.B.

1985-10-01

The Canadian Nuclear Fuel Waste Management Program is in the fifth year of a ten-year generic research and development phase. The major objective of this phase of the program is to assess the basic safety and environmental aspects of the concept of isolating immobilized fuel waste by deep underground disposal in plutonic rock. The major scientific and engineering components of the program, namely immobilization studies, geoscience research, and environmental and safety assessment, are well established

NEPA implementation: The Department of Energy's program to manage spent nuclear fuel

International Nuclear Information System (INIS)

Shipler, D.B.

1994-05-01

The Department of Energy (DOE) is implementing the National Environmental Protection Act (NEPA) in its management of spent nuclear fuel. The DOE strategy is to address the short-term safety concerns about existing spent nuclear fuel, to study alternatives for interim storage, and to develop a long-range program to manage spent nuclear fuel. This paper discusses the NEPA process, the environmental impact statements for specific sites as well as the overall program, the inventory of DOE spent nuclear fuel, the alternatives for managing the fuel, and the schedule for implementing the program

Program on fuels for transmutation: present status and prospects

Energy Technology Data Exchange (ETDEWEB)

Rouault, J.; Garnier, J.C.; Chauvin, N.; Pillon, S. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Combustibles

2001-07-01

The performance calculations of appropriate fuel cycle facilities and reactor configurations (scenarios) relying on current reactor technologies (Pressurized Water Reactor and Fast neutrons Reactors) or innovative reactors (Accelerator Driven Systems) have proved the scientific feasibility of some P and T strategies. To insure the technological feasibility, a large program on fuels and materials is underway, including advanced concepts for PWRs and the development of specific targets (dispersed fuels) for transmutation in Fast Reactors. Experiments in different reactors including Phenix are being prepared. The program is presented and recent results are given. (author)

Program on fuels for transmutation: present status and prospects

International Nuclear Information System (INIS)

Rouault, J.; Garnier, J.C.; Chauvin, N.; Pillon, S.

2001-01-01

The performance calculations of appropriate fuel cycle facilities and reactor configurations (scenarios) relying on current reactor technologies (Pressurized Water Reactor and Fast neutrons Reactors) or innovative reactors (Accelerator Driven Systems) have proved the scientific feasibility of some P and T strategies. To insure the technological feasibility, a large program on fuels and materials is underway, including advanced concepts for PWRs and the development of specific targets (dispersed fuels) for transmutation in Fast Reactors. Experiments in different reactors including Phenix are being prepared. The program is presented and recent results are given. (author)

A summary of the program and progress to 1984 December of the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Dixon, R.S.

1986-08-01

The Canadian Nuclear Fuel Waste Management Program involves research into the storage and transportation of used nuclear fuel, immobilization of fuel waste, and deep geological disposal of the immobilized waste. The program is now in the fifth year of a ten-year generic research and development phase. The objective of this phase of the program is to assess the safety and environmental aspects of the deep underground disposal of immobilized fuel waste in plutonic rock. The objectives of the research for each component of the program and the progress made to the end of 1984 are described in this report. 74 refs

Operational experience using the OSTR flip fuel self-protection program

International Nuclear Information System (INIS)

Dodd, B.; Ringle, J.C.; Anderson, T.V.; Johnson, A.G.

1982-01-01

Recent changes in NRC Physical Security regulations make it highly desirable for a small number of TRIGA research reactor establishments to maintain each of the fuel elements in their reactor core above the self-protection dose rate criterion. OSTR operations personnel have written a computer program (SPOOF) which calculates the exposure rate (in Rhr -1 ) from an irradiated fuel element at 3 feet in air using the actual operating history of the reactor. The purpose of this current paper is to describe the operational experience gained over the last year and a half while using the SPOOF computer program, and while performing the quarterly dose rate measurements needed to confirm the continuing accuracy of the program, and, most importantly, the self-protection status of the OSTR fuel. The computer program in association with the quarterly dose rate measurements have been accepted by the NRC, and allow the OSTR to take credit for self-protecting FLIP fuel under the current physical security regulations

The EBR-II spent fuel treatment program

International Nuclear Information System (INIS)

Lineberry, M.J.; McFarlane, H.F.

1995-01-01

Argonne National Laboratory has refurbished and equipped an existing hot cell facility for demonstrating a high-temperature electrometallurgical process for treating spent nuclear fuel from the Experimental Breeder Reactor-11. Two waste forms will be produced and qualified for geologic disposal of the fission and activation products. Relatively pure uranium will be separated for storage. Following additional development, transuranium elements will be blended into one of the high-level waste streams. The spent fuel treatment program will help assess the viability of electrometallurgical technology as a spent fuel management option

Fifth annual report to congress. Federal alternative motor fuels programs

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

This report presents the status of the US Department of Energy`s alternative fuel vehicle demonstration and performance tracking programs being conducted in accordance with the Energy Policy and Conservation Act. These programs comprise the most comprehensive data collection effort ever undertaken on alternative transportation fuels and alternative fuel vehicles. The report summarizes tests and results from the fifth year. Electric vehicles are not included in these programs, and the annual report does not include information on them. Since the inception of the programs, great strides have been made in developing commercially viable alternative fuel vehicle technologies. However, as is the case in the commercialization of all new technologies, some performance problems have been experienced on vehicles involved in early demonstration efforts. Substantial improvements have been recorded in vehicle practicality, safety, and performance in real-world demonstrations. An aspect of particular interest is emissions output. Results from light duty alternative fuel vehicles have demonstrated superior inservice emissions performance. Heavy duty alternative fuel vehicles have demonstrated dramatic reductions in particulate emissions. However, emissions results from vehicles converted to run on alternative fuel have not been as promising. Although the technologies available today are commercially viable in some markets, further improvements in infrastructure and economics will result in greater market expansion. Information is included in this report on light and heavy duty vehicles, transit buses, vehicle conversions, safety, infrastructure support, vehicle availability, and information dissemination.

76 FR 67287 - Alternative Fuel Transportation Program; Alternative Fueled Vehicle Credit Program (Subpart F...

Science.gov (United States)

2011-10-31

... additional credits for the use of biodiesel in blends of 20 percent biodiesel or greater and have provided an... discussion in Part II.A), the original program based upon AFV acquisitions and biodiesel use became known as... example, B20 (a 20 percent blend of biodiesel with 80 percent petroleum diesel) is not an alternative fuel...

PBF severe fuel damage program: results and comparison to analysis

International Nuclear Information System (INIS)

McDonald, P.E.; Buescher, B.J.; Gruen, G.E.; Hobbins, R.R.; McCardell, R.K.

1983-01-01

The United States Nuclear Regulatory Commission has initiated a severe fuel damage research program in the Power Burst Facility (PBF) to investigate fuel rod and core response, and fission product and hydrogen release and transport under degraded core cooling accident conditions. This paper presents a description of Phase I of the PBF Severe Fuel Damage Program, discusses the results of the first experiment, and compares those results with analysis performed using the TRAC-BD1 computer code

PBF Severe Fuel-Damage Program: results and comparison to analysis

International Nuclear Information System (INIS)

MacDonald, P.E.; Buescher, B.J.; Hobbins, R.R.; McCardell, R.K.; Gruen, G.E.

1983-01-01

The United States Nuclear Regulatory Commission has initiated a severe fuel-damage research program in the Power Burst Facility (PBF) to investigate fuel-rod and core response, and fission-product and hydrogen release and transport under degraded-core-cooling accident conditions. This paper presents a description of Phase I of the PBF Severe Fuel Damage Program, discusses the results of the first experiment, and compares those results with analysis performed using the TRAC-BD1 computer code

Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program

Energy Technology Data Exchange (ETDEWEB)

Molecke, M.A.; Gregson, M.W.; Sorenson, K.B. [Sandia National Labs. (United States); Billone, M.C.; Tsai, H. [Argonne National Lab. (United States); Koch, W.; Nolte, O. [Fraunhofer Inst. fuer Toxikologie und Experimentelle Medizin (Germany); Pretzsch, G.; Lange, F. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (Germany); Autrusson, B.; Loiseau, O. [Inst. de Radioprotection et de Surete Nucleaire (France); Thompson, N.S.; Hibbs, R.S. [U.S. Dept. of Energy (United States); Young, F.I.; Mo, T. [U.S. Nuclear Regulatory Commission (United States)

2004-07-01

We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this research to better understand potential radiological impacts from sabotage of nuclear material shipments and storage casks, and to support subsequent risk assessments, modeling, and preventative measures. We provide a summary of the overall, multi-phase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on ''surrogate'' spent fuel, unirradiated depleted uranium oxide, and forthcoming actual spent fuel tests. The depleted uranium oxide test rodlets were prepared by the Institut de Radioprotection et de Surete Nucleaire, in France. These surrogate test rodlets closely match the diameter of the test rodlets of actual spent fuel from the H.B. Robinson reactor (high burnup PWR fuel) and the Surry reactor (lower, medium burnup PWR fuel), generated from U.S. reactors. The characterization of the spent fuels and fabrication into short, pressurized rodlets has been performed by Argonne National Laboratory, for testing at SNL. The ratio of the aerosol and respirable particles released from HEDD-impacted spent

Finite element method programs to analyze irradiation behavior of fuel pellets

International Nuclear Information System (INIS)

Yamada, Rayji; Harayama, Yasuo; Ishibashi, Akihiro; Ono, Masao.

1979-09-01

For the safety assessment of reactor fuel, it is important to grasp local changes of fuel pins due to irradiation in a reactor. Such changes of fuel result mostly from irradiation of fuel pellets. Elasto-plastic analysis programs based on the finite element method were developed to analyze these local changes. In the programs, emphasis is placed on the analysis of cracks in pellets; the interaction between cracked-pellets and cladding is not taken into consideration. The two programs developed are FEMF3 based on a two-dimensional axially symmetric model (r-z system) and FREB4 on a two-dimensional plane model (r-theta system). It is discussed in this report how the occurrence and distribution of cracks depend on heat rate of the fuel pin. (author)

2015 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

Energy Technology Data Exchange (ETDEWEB)

None

2015-12-23

The 2015 Annual Progress Report summarizes fiscal year 2015 activities and accomplishments by projects funded by the DOE Hydrogen and Fuel Cells Program. It covers the program areas of hydrogen production; hydrogen delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes and standards; systems analysis; and market transformation.

Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Preliminary design report

International Nuclear Information System (INIS)

Siefken, L.J.

1999-01-01

Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ''Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents.''

Development of fusion fuel cycles: Large deviations from US defense program systems

Energy Technology Data Exchange (ETDEWEB)

Klein, James Edward, E-mail: james.klein@srnl.doe.gov; Poore, Anita Sue; Babineau, David W.

2015-10-15

Highlights: • All tritium fuel cycles start with a “Tritium Process.” All have similar tritium processing steps. • Fusion tritium fuel cycles minimize process tritium inventories for various reasons. • US defense program facility designs did not minimize in-process inventories. • Reduced inventory tritium facilities will lower public risk. - Abstract: Fusion energy research is dominated by plasma physics and materials technology development needs with smaller levels of effort and funding dedicated to tritium fuel cycle development. The fuel cycle is necessary to supply and recycle tritium at the required throughput rate; additionally, tritium confinement throughout the facility is needed to meet regulatory and environmental release limits. Small fuel cycle development efforts are sometimes rationalized by stating that tritium processing technology has already been developed by nuclear weapons programs and these existing processes only need rescaling or engineering design to meet the needs of fusion fuel cycles. This paper compares and contrasts features of tritium fusion fuel cycles to United States Cold War era defense program tritium systems. It is concluded that further tritium fuel cycle development activities are needed to provide technology development beneficial to both fusion and defense programs tritium systems.

On-site fuel cell field test support program

Science.gov (United States)

Staniunas, J. W.; Merten, G. P.

1982-01-01

In order to assess the impact of grid connection on the potential market for fuel cell service, applications studies were conducted to identify the fuel cell operating modes and corresponding fuel cell sizing criteria which offer the most potential for initial commercial service. The market for grid-connected fuel cell service was quantified using United's market analysis program and computerized building data base. Electric and gas consumption data for 268 buildings was added to our surveyed building data file, bringing the total to 407 buildings. These buildings were analyzed for grid-isolated and grid-connected fuel cell service. The results of the analyses indicated that the nursing home, restaurant and health club building sectors offer significant potential for fuel cell service.

Non destructive examination of UN / U-Si fuel pellets using neutrons (preliminary assessment)

Energy Technology Data Exchange (ETDEWEB)

Bourke, Mark Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Voit, Stewart Lancaster [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Losko, Adrian S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tremsin, Anton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-03-31

Tomographic imaging and diffraction measurements were performed on nine pellets; four UN/ U Si composite formulations (two enrichment levels), three pure U₃Si₅ reference formulations (two enrichment levels) and two reject pellets with visible flaws (to qualify the technique). The U-235 enrichments ranged from 0.2 to 8.8 wt.%. The nitride/silicide composites are candidate compositions for use as Accident Tolerant Fuel (ATF). The monophase U₃Si₅ material was included as a reference. Pellets from the same fabrication batches will be inserted in the Advanced Test Reactor at Idaho during 2016. The goal of the Advanced Non-destructive Fuel Examination work package is the development and application of non-destructive neutron imaging and scattering techniques to ceramic and metallic nuclear fuels. Data reported in this report were collected in the LANSCE run cycle that started in September 2015 and ended in March 2016. Data analysis is ongoing; thus, this report provides a preliminary review of the measurements and provides an overview of the characterized samples.

Optimization programs for reactor core fuel loading exhibiting reduced neutron leakage

International Nuclear Information System (INIS)

Darilek, P.

1991-01-01

The program MAXIM was developed for the optimization of the fuel loading of WWER-440 reactors. It enables the reactor core reactivity to be maximized by modifying the arrangement of the fuel assemblies. The procedure is divided into three steps. The first step includes the passage from the three-dimensional model of the reactor core to the two-dimensional model. In the second step, the solution to the problem is sought assuming that the multiplying properties, or the reactivity in the zones of the core, vary continuously. In the third step, parameters of actual fuel assemblies are inserted in the ''continuous'' solution obtained. Combined with the program PROPAL for a detailed refinement of the loading, the program MAXIM forms a basis for the development of programs for the optimization of fuel loading with burnable poisons. (Z.M.). 16 refs

Preliminary Results of the Louisiana Sex Offender Treatment Program

Directory of Open Access Journals (Sweden)

Lee A. Underwood

2015-12-01

Full Text Available The purpose of this study was to offer preliminary support for the Louisiana Sex Offender Treatment Program (LSOTP in addressing the needs of juvenile sex offenders. Research objectives were (1 to offer statistical evidence for reductions in anxiety, depression, cognitive distortion and negative attitudes towards women comparing a group of 21 adolescents, 12 of whom received services as usual and nine of whom participated in the LSOTP. A controlled experimental evaluation design was utilized. The juvenile sex offenders were randomly assigned to the experimental group for 12 weeks receiving treatment services and a control group receiving care “as usual” in a residential group care program. Participants in the experimental group experienced statistically significant decreases in cognitive distortions related specifically to rape and molestation.The results of this study offer preliminary support of the LSOTP as a best practices alternative to other treatment modalities.

Space reactor preliminary mechanical design

International Nuclear Information System (INIS)

Meier, K.L.

1983-01-01

An analysis was performed on the SABRE reactor space power system to determine the effect of the number and size of heat pipes on the design parameters of the nuclear subsystem. Small numbers of thin walled heat pipes were found to give a lower subsystem mass, but excessive fuel swelling resulted. The SP-100 preliminary design uses 120 heat pipes because of acceptable fuel swelling and a minimum nuclear subsystem mass of 1875 kg. Salient features of the reactor preliminary design are: individual fuel modules, ZrO 2 block core mounts, bolted collar fuel module restraints, and a BeO central plug

Preliminary study on flexible core design of super FBR with multi-axial fuel shuffling

International Nuclear Information System (INIS)

Sukarman; Yamaji, Akifumi; Someya, Takayuki; Noda, Shogo

2017-01-01

Preliminary study has been conducted on developing a new flexible core design concept for the Supercritical water-cooled Fast Breeder Reactor (Super FBR) with multi-axial fuel shuffling. The proposed new concept focuses on the characteristic large axial coolant density change in supercritical water cooled reactors (SCWRs) when the coolant inlet temperature is below the pseudocritical point and large coolant enthalpy rise is taken in the core for achieving high thermal efficiency. The aim of the concept is to attain both the high breeding performance and good thermal-hydraulic performance at the same time. That is, short Compound System Doubling Time (CSDT) for high breeding, large coolant enthalpy rise for high thermal efficiency, and large core power. The proposed core concept consists of horizontal layers of mixed oxide (MOX) fuels and depleted uranium (DU) blanket layers at different elevation levels. Furthermore, the upper core and the lower core are separated and independent fuel shuffling schemes in these two core regions are considered. The number of fuel batches and fuel shuffling scheme of the upper core were changed to investigate influence of multi-axial fuel shuffling on the core characteristics. The core characteristics are evaluated with-three-dimensional diffusion calculations, which are fully-coupled with thermal-hydraulics calculations based on single channel analysis model. The results indicate that the proposed multi-axial fuel shuffling scheme does have a large influence on CSDT. Further investigations are necessary to develop the core concept. (author)

First annual report of the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Boulton, J.; Gibson, A.R.

1979-12-01

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

Quality evaluation of the nuclear fuel manufactured in Romania before 1990

International Nuclear Information System (INIS)

Budan, O.; Galeriu, C.A.; Pascu, A.; Bailescu, A.; Andrei, G.

1997-01-01

Nuclear fuel fabricated in Romania before June 1990 was stored and was not licensed for the use in reactor. A preliminary appraisal performed by AECL and Zircatec experts in 1990, considered this fuel as being of suspect quality. Taking into consideration that Nuclear Fuel Plant was qualified to produce CANDU-6 fuel, only after 1990, RENEL-GEN authorized FCN to evaluate the actual quality of the stock fuel, in 1996. A documented evaluation program was developed and implemented. Stock production of each year was considered as one fuel bundle lot. Quality verification was done on statistical basis. Random sampling was applied, with the condition that each production week was represented in the sample. The results show that the fuel quality is not acceptable for in-reactor use. Recovery solutions for the stock fuel were established and already applied. An independent review of the program, results and recovery solutions, made by AECL confirmed the consistency of the work performed by FCN for stock fuel quality evaluation. (author)

Preliminary estimates of the charge for spent-fuel storage and disposal services

International Nuclear Information System (INIS)

1978-07-01

The purpose of this report is to stimulate discussion among a wide range of interested parties concerning a one-time charge by the U.S. Government for disposal, or interim storage and disposal, of spent unreprocessed nuclear fuel. The report contains a set of estimates of the charge based on current cost figures and a variety of demand, logistical, institutional, and cost overrun assumptions. The services are to be offered to domestic utilities by the U.S. Government in connection with the spent fuel policy approved by the President and announced by the Department of Energy (DOE) on October 18, 1977. This policy is a direct result of the indefinite deferral of all commercial reprocessing of spent fuel announced by President Carter on April 7, 1977. The services will also be offered to foreign governments on a limited basis in cases where this action would contribute to U.S. goals for nonproliferation of nuclear weapons. The report does not establish new policy and it does not commit DOE to any specific program, schedule or charge. No scenario or case is to be considered most important, no methodology is to be considered definitive, and no charge is to be considered most likely or to represent a proposed charge. The report describes basic principles and methodologies for calculating the charge and highlights primary cost centers. Current estimates of program and facility costs are used. Various aspects of the DOE Spent Fuel Storage Program are brought into focus through this analysis. Interested parties should find these assessment criteria helpful for their planning and useful in discussions concerning the program

Preliminary report. Preliminary findings and views concerning the exemption of kerojet fuels from the Mandatory Petroleum Allocation and Price Regulations

Energy Technology Data Exchange (ETDEWEB)

None

1978-01-01

Preliminary findings indicate the following: kerojet fuel is not in short supply; it will not adversely impact on the supply of other petroleum products subject to the Emergency Petroleum Allocation Act; competition and market forces are adequate; and it will not result in inequitable prices for kerojet or other products. Chapter II provides background information on the use, production, and distribution of kerojet. Chapter III analyzes the historical interaction of supply, demand, and price, and explores the market structure for kerojet during 1968 to 1976, prior to and during imposition of allocation and price controls. Chapter IV examines the effect upon kerojet supply, demand, price, and market structure of exempting kerojet from controls and indicates the benefits to be derived from such exemption. In Chapter V, the potential economic impacts of exemption are evaluated. Chapter VI provides a final summary of the DOE's findings and views in support of its preliminary judgment that kerojet should be exempted from allocation and price regulations. (MCW)

Power Burst Facility severe-fuel-damage test program

International Nuclear Information System (INIS)

McCardell, R.K.; MacDonald, P.E.

1982-01-01

As a result of the Three Mile Island Unit 2 (TMI-2) accident, the United States Nuclear Regulatory Commission (USNRC) has initiated a severe fuel damage research program to investigate fuel rod and core response, and fission product and hydrogen release and transport during degraded core cooling accidents. This paper presents a discussion of the expected benefits of the PBF severe fuel damage tests to the nuclear industry, a description of the first five planned experiments, the results of pretest analysis performed to predict the fuel bundle heatup for the first two experiments, and a discussion of Phase II severe fuel damage experiments. Modifications to the fission product detection system envisioned for the later experiments are also described

2016 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

Energy Technology Data Exchange (ETDEWEB)

None, None

2017-03-09

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

Studsvik's implementation of the DOE spent fuel return program: The US perspective

International Nuclear Information System (INIS)

Grover, Stephen; Browser, Rita C.

1996-01-01

The government of Sweden has long been a proponent of nuclear nonproliferation policies. This includes laws governing the conduct of the Studsvik R2 Reactor in Nykoeping, Sweden. Studsvik became a participant in the RERTR program early on and was one of the first research reactors to convert from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel. Since Studsvik operates the reactor on a commercial basis, it is important to fully comply with all Swedish laws and regulations for the operation of a nuclear reactor. This includes development of a proper long term program for the final disposition for the irradiated fuel. For this reason it was very important to Studsvik to be involved in the process for development of the DOE Spent Fuel Return Program as early as possible. This process included input into the Environmental Impact Statement process, early contractual negotiations after the issuance of the Record of Decision, participation in litigation involving the Spent Fuel Return Program, and plans for return of HEU fuel on the first European shipment. Not only was program involvement important, but actual program implementation was necessary for Studsvik to maintain its reactor license. Studsvik needed to have fuel included in the first shipment in the return program not only for licensing considerations, but also to reduce the proliferation risk associated with the storage of non-self protecting fuel. To ensure participation in the first shipment, Studsvik's activities included the early development of Appendix A's, coordinating the shipment, obtaining approvals and authorizations, and other technical aspects. (author)

A preliminary evaluation of a combined tire- and refuse-derived fuel (TDF-RDF)

Energy Technology Data Exchange (ETDEWEB)

Stessel, R.I.; Amari, T.; Themelis, N.J.; Wearnick, I.K.

1999-07-01

In dense urban areas of the US, it is now becoming clear that waste management is far from economically-optimum. Even with the popularity of inexpensive land disposal, hauling and recycling costs are driving up the average waste bill. An historic option has been refuse-derived fuel, or RDF. Difficulties included low energy content and difficulty obtaining uniformity. Today, many resource-recovery technologies used in RDF are finding their way into materials recovery facilities (MRFs), some of which are reviving the automated processing of waste. Any MRF, automated or not, will have residue streams. Currently, one of the most significant problems is waste tires. Local options are difficult to locate in dense urban areas. As fuels, tires typically have energy contents considerably above those for which most solid-fuel combustors are designed, leading to thermal imbalances and various forms of failure. This paper suggests a new fuel that can be either co-fired with coal, or used in its own right in a combustor primarily designed for coal: TDF-RDF. A preliminary examination is undertaken of thermal and emissions characteristics, and possible costs for a few applications of the fuel. Immediately, TDF is already cleaner-burning than many coals, even in sulfur emissions. RDF has been widely-regarded as being similarly cleanly. Posited MRF residue streams should be still cleaner, and more consistent, than RDF. Overall, there is quite a potential for developing a fuel that would allow old coal powerplants in historic urban centers to be better neighbors, while helping with a few problems in municipal waste management.

Preliminary study on the feasibility of ductless fuel assembly for fast reactors

International Nuclear Information System (INIS)

Shibahara, Itaru; Enokido, Yuji

1988-01-01

Preliminary study on the feasibility of ductless fuel assembly for fast reactors has been conducted. The primary concern is with forecasting the thermal hydraulic characteristics and the heat removal efficiency from the core. The thermal hydraulic analysis revealed the coolant mixing in the core at steady state operating condition was not intensive and the coolant temperature increase was almost proportional to the power of each assembly. The hot spot analysis of the ductless core indicated that the hottest temperature in the core could be comparable with the temperature of the conventional ducted core, even in case the radial power flattening was not actively pursued but with adopting ducted radial blanket assemblies. Under off-normal conditions, the ductless core had improved heat removal capability which was caused by inter-assembly coolant flow. The study has indicated the feasibility of the ductless fuel assembly for fast reactors. The experiments to demonstrate the feasibility will be the next key process for the development. (author)

A status report on the integral fast reactor fuels and safety program

International Nuclear Information System (INIS)

Pedersen, D.R.; Seidel, B.R.

1990-01-01

The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor (ALMR) concept being developed at Argonne National Laboratory. The IFR program is specifically responsible for the irradiation performance, advanced core design, safety analysis, and development of the fuel cycle for the US Department of Energy's ALMR program. The basic elements of the IFR concept are (a) metallic fuel, (b) liquid-sodium cooling, (c) modular, pool-type reactor configuration, (d) an integral fuel cycle based upon pyrometallurgical processing. The most significant safety aspects of the IFR program result from its unique fuel design, a ternary alloy of uranium, plutonium, and zirconium. This fuel is based on experience gained through > 25 yr operation of the Experimental Breeder Reactor II (EBR-II) with a uranium alloy metallic fuel. The ultimate criteria for fuel pin design is the overall integrity at the target burnup. The probability of core meltdown is remote; however, a theoretical possibility of core meltdown remains. The next major step in the IFR development program will be a full-scale pyroprocessing demonstration to be carried out in conjunction with EBR-II. The IFR fuel cycle closure based on pyroprocessing will also have a dramatic impact on waste management options and on actinide recycling

The NRU blowdown test facility commissioning program

Energy Technology Data Exchange (ETDEWEB)

Walsworth, J A; Zanatta, R J; Yamazaki, A R; Semeniuk, D D; Wong, W; Dickson, L W; Ferris, C E; Burton, D H [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.

1990-12-31

A major experimental program has been established at the Chalk River Nuclear Laboratories (CRL) that will provide essential data on the thermal and mechanical behaviour of nuclear fuel under abnormal reactor operating conditions and on the transient release, transport and deposition of fission product activity from severely degraded fuel. A number of severe fuel damage (SFD) experiments will be conducted within the Blowdown Test Facility (BTF) at CRL. A series of experiments are being conducted to commission this new facility prior to the SFD program. This paper describes the features and the commissioning program for the BTF. A development and testing program is described for critical components used on the reactor test section. In-reactor commissioning with a fuel assembly simulator commenced in 1989 June and preliminary results are given. The paper also outlines plans for future all-effects, in-reactor tests of CANDU-designed fuel. (author). 11 refs., 3 tabs., 7 figs.

Fire, Fuel, and Smoke Program: 2014 Research Accomplishments

Science.gov (United States)

Faith Ann Heinsch; Robin J. Innes; Colin C. Hardy; Kristine M. Lee

2015-01-01

The Fire, Fuel, and Smoke Science Program (FFS) of the U.S. Forest Service, Rocky Mountain Research Station focuses on fundamental and applied research in wildland fire, from fire physics and fire ecology to fuels management and smoke emissions. Located at the Missoula Fire Sciences Laboratory in Montana, the scientists, engineers, technicians, and support staff in FFS...

Alternative fuel cycles

International Nuclear Information System (INIS)

Penn, W.J.

1979-05-01

Uranium resource utilization and economic considerations provide incentives to study alternative fuel cycles as future options to the PHWR natural uranium cycle. Preliminary studies to define the most favourable alternatives and their possible introduction dates are discussed. The important and uncertain components which influence option selection are reviewed, including nuclear capacity growth, uranium availability and demand, economic potential, and required technological developments. Finally, a summary of Ontario Hydro's program to further assess cycle selection and define development needs is given. (auth)

LEU fuel fabrication program for the RECH-1 reactor. Status report

International Nuclear Information System (INIS)

Chavez, J.C.; Barrera, M.; Jimenez, O.; Lisboa, J.; Marin, J.

2000-01-01

In 1995 a 50 LEU U 3 Si 2 fuel elements fabrication program for the RECH-1 research reactor was established at the Comision Chilena de Energia Nuclear, CCHEN. After a fabrication process qualification stage, in 1998, four elements were early delivered to the reactor in order to start an irradiation qualification stage. The irradiation has reached an estimated 10% burn-up and no fabrication problems have been detected up to this burn-up level. During 1999 and up to the first quarter of 2000, 19 fuel elements were produced and 7 fuel elements are expected for the end of 2000. This report presents an updated summary of the main results obtained in this fuel fabrication program. A summary of other activities generated by this program, such as in core follow-up of the four leader fuel elements, ISO 9001 implementation for the fabrication process and a fabrication and qualification optimization planning, is also presented here. (author)

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

Methods and computer programs for PWR's fuel management: Programs Sothis and Ciclon

International Nuclear Information System (INIS)

Aragones, J.M.; Corella, M.R.; Martinez-Val, J.M.

1976-01-01

Methos and computer programs developed at JEN for fuel management in PWR are discussed, including scope of model, procedures for sistematic selection of alternatives to be evaluated, basis of model for neutronic calculation, methods for fuel costs calculation, procedures for equilibrium and trans[tion cycles calculation with Soth[s and Ciclon codes and validation of methods by comparison of results with others of reference (author) ' [es

Safety assessment to support NUE fuel full core implementation in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Fan, H.Z.; Laurie, T.; Siddiqi, A.; Li, Z.P.; Rouben, D.; Zhu, W.; Lau, V.; Cottrell, C.M. [CANDU Energy Inc., Mississauga, Ontario (Canada)

2013-07-01

The Natural Uranium Equivalent (NUE) fuel contains a combination of recycled uranium and depleted uranium, in such a manner that the resulting mixture is similar to the natural uranium currently used in CANDU® reactors. Based on successful preliminary results of 24 bundles of NUE fuel demonstration irradiation in Qinshan CANDU 6 Unit 1, the NUE full core implementation program has been developed in cooperation with the Third Qinshan Nuclear Power Company and Candu Energy Inc, which has recently received Chinese government policy and funding support from their National-Level Energy Innovation program. This paper presents the safety assessment results to technically support NUE fuel full core implementation in CANDU reactors. (author)

TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

International Nuclear Information System (INIS)

Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

1983-01-01

TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

Romanian concern for advanced fuels development

International Nuclear Information System (INIS)

Ohai, Dumitru

2001-01-01

The Institute for Nuclear Research (ICN), a subsidiary of Romanian Authority for Nuclear Activities, at Pitesti - Romania, has developed a preliminary design of a fuel bundle with 43 elements named SEU 43 for high burnup in CANDU Reactor. A very high experience in nuclear fuels manufacturing and control has also been accumulated. Additionally, on the nuclear site Pitesti there is the Nuclear Fuel Plant (NFP) qualified to manufacturing CANDU 6 type fuel, the main fuel supplier for NPP Cernavoda. A very good collaboration of ICN with NFP can lead to a low cost upgrading the facilities which ensure at present the CANDU standard fuel fabrication to be able of manufacturing also SEU 43 fuel for extended burnup. The financial founds are allocated by Romanian Authority for Nuclear Activities of the Ministry of Industry and Resources to sustain the departmental R and D program 'Nuclear Fuel'. This Program has the main objective to establish a technology for manufacturing a new CANDU fuel type destined for extended burnup. It is studied the possibility to use the Recovered Uranium (RU) resulted from LWR spent fuel reprocessing facility existing in stockpiles. The International Agency for Atomic Energy (IAEA) sustains also this program. By ROM/4/025/ Model Project, IAEA helps ICN to solve the problems regarding materials (RU, Zircaloy 4 tubes) purchasing, devices' upgrading and personnel training. The paper presents the main actions needing to be create the technical base for SEU 43 fuel bundle manufacturing. First step, the technological experiments and experimental fuel element manufacturing, will be accomplished in ICN installations. Second step, the industrial scale, need thorough studies for each installation from NFP to determine tools and technology modification imposed by the new CANDU fuel bundle manufacturing. All modifications must be done such as to the NFP, standard CANDU and SEU fuel bundles to be manufactured alternatively. (author)

Post-irradiation examination and R and D programs using irradiated fuels at KAERI

International Nuclear Information System (INIS)

Chun, Yong Bum; Min, Duck Kee; Kim, Eun Ka and others

2000-12-01

This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

Post-irradiation examination and R and D programs using irradiated fuels at KAERI

International Nuclear Information System (INIS)

Chun, Yong Bum; So, Dong Sup; Lee, Byung Doo; Lee, Song Ho; Min, Duck Kee

2001-09-01

This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

2011 Annual Progress Report: DOE Hydrogen and Fuel Cells Program (Book)

Energy Technology Data Exchange (ETDEWEB)

2011-11-01

In the past year, the DOE Hydrogen and Fuel Cells Program (the Program) made substantial progress toward its goals and objectives. The Program has conducted comprehensive and focused efforts to enable the widespread commercialization of hydrogen and fuel cell technologies in diverse sectors of the economy. With emphasis on applications that will effectively strengthen our nation's energy security and improve our stewardship of the environment, the Program engages in research, development, and demonstration of critical improvements in the technologies. Highlights of the Program's accomplishments can be found in the sub-program chapters of this report.

Program description for the program Fuel program conversion 2011-07-01 - 2015-06-30; Programbeskrivning foer programmet Braensleprogrammet omvandling 2011-07-01 - 2015-06-30

Energy Technology Data Exchange (ETDEWEB)

2011-04-15

The Fuel program conversion is included as one of three programs in a cohesive commitment to increased, sustainable and efficient production and use of indigenous and renewable fuels that are implemented by the Swedish Energy Agency during the period July 1, 2011 - June 30, 2015. The program focuses on increasing knowledge and development of technologies for processing of the fuel feedstock, and conversion to primarily heat and electricity in plants up to about 10 MW. Issues of particular concern is to strengthen competitiveness, efficiency and sustainability in the chain from raw material to end use, and increasing raw material base by developing knowledge and technologies for new materials that are current. The program includes basic and applied research, development and demonstration. Within the sub-area Processing efforts are prioritized which streamline and reduce costs in different types of processing operations. Processing here means not just technology for pelleting and briquetting, but also sub-processes that prepare, dry, handle or process raw materials in other ways. Within the sub-area Conversion to heat and electricity in the range <10 MW, the focus is principally, on one hand increasing the knowledge of the interaction between fuel and combustion plant, and on the other hand to implement results of research in collaboration with users and equipment suppliers. Furthermore, great importance is given to increasing the competitiveness of local heating in the transition from oil to renewable heat supply and the use of new fuel materials and fuels. Questions about the fuel supply from forestry and agriculture, including issues of sustainability, environment and resource use are treated in the two parallel running programs, the Fuel program supply and the Fuel program sustainability

A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

International Nuclear Information System (INIS)

Andress, D.; McLeod, N.B.; Rahimi, M.; Joy, D.S.; Peterson, R.W.

1991-01-01

The DOE has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design - the GA-4 and GA-9 truck casks and the BR-100 rail cask. The GA-4 cask is designed for PWR fuel only; the GA-9 cask is a longer cask with less shielding designed for BWR fuel only; and the BR-100 cask is designed to accommodate both PWR and BWR fuels. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elements as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. Because of the button and spring interference, the basket openings in these casks will not accommodate assemblies in the BWR/2,3 and BWR/4-6 fuel classes with the fuel channels in place

Sustainable Harvest for Food and Fuel Preliminary Food & Fuel Gap Analysis Report

Energy Technology Data Exchange (ETDEWEB)

Ray Grosshans; Kevin M. Kostelnik; Jake Jacobson

2007-04-01

To promote economic growth and energy security, and to protect the environment, the U.S. is pursuing a national strategy of energy independence and climatic protection in which domestic renewable carbon-neutral biofuels displace 30 percent of U.S. oil consumption by the mid-21st century. Such fuels, including ethanol and biodiesel, will be produced from biological feed stocks (biomass). The availability of this billion-ton biomass will hinge on the application of modern scientific and engineering tools to create a highly-integrated biofuel production system. Efforts are underway to identify and develop energy crops, ranging from agricultural residues to genetically engineered perennials; to develop biology-based processing methods; and, to develop large-scale biorefineries to economically convert biomass into fuels. In addition to advancing the biomass-to-biofuel research and development agenda, policy makers are concurrently defining the correct mix of governmental supports and regulations. Given the volumes of biomass and fuels that must flow to successfully enact a national biomass strategy, policies must encourage large-scale markets to form and expand around a tightly integrated system of farmers, fuel producers and transporters, and markets over the course of decades. In formulating such policies, policy makers must address the complex interactions of social, technical, economic, and environmental factors that bound energy production and use. The Idaho National Laboratory (INL) is a science-based, applied engineering national laboratory dedicated to supporting the U.S. Department of Energy (DOE). The INL Bioenergy Program supports the DOE and the U.S. Department of Agriculture. Key multidisciplinary INL capabilities are being leveraged to address major science and technology needs associated with the cost-effective utilization of biomass. INL’s whole crop utilization (WCU) vision is focused on the use of the entire crop, including both the grain and

Safety assessment of spent-fuel transportation in extreme environments

International Nuclear Information System (INIS)

Sandoval, R.P.; Weber, J.P.; Newton, G.J.

1981-01-01

Preliminary estimates of the health effects and/or consequences resulting from a malevolent attack on a spent fuel truck shipment in downtown New York City have been made. This estimate is based upon a measured quantity (0.78 +- 0.05 g) of respirable radioactive material released from a 1/4 scale event. A linear extrapolation from the 1/4 scale event to the generic full scale event has been made and an aerosolized release fraction (0.0023 percent) of the total heavy metal inventory of a three-PWR assembly truck cask has been calculated. Although scaling of the source term parameters is tentative at this point in the program, a full scale experiment is planned in 1981 to verify the scaling methodology used in these calculations. A preliminary correlation between spent fuel and surrogate fuel source terms has been shown to be feasible and that radionuclide size partitioning can be determined experimentally. Finally, it has been shown, based on our preliminary experimental source term data, that a maximum of 25 total latent cancer fatalities could occur, assuming a release in downtown New York City. This is 20 times smaller than the latent cancer fatalities predicted in the Urban Study

Extended burnup demonstration: reactor fuel program. Pre-irradiation characterization and summary of pre-program poolside examinations. Big Rock Point extended burnup fuel

International Nuclear Information System (INIS)

Exarhos, C.A.; Van Swam, L.F.; Wahlquist, F.P.

1981-12-01

This report is a resource document characterizing the 64 fuel rods being irradiated at the Big Rock Point reactor as part of the Extended Burnup Demonstration being sponsored jointly by the US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities. The program entails extending the exposure of standard BWR fuel to a discharge average of 38,000 MWD/MTU to demonstrate the feasibility of operating fuel of standard design to levels significantly above current limits. The fabrication characteristics of the Big Rock Point EBD fuel are presented along with measurement of rod length, rod diameter, pellet stack height, and fuel rod withdrawal force taken at poolside at burnups up to 26,200 MWD/MTU. A review of the fuel examination data indicates no performance characteristics which might restrict the continued irradiation of the fuel

Small-scale irradiated fuel electrorefining

International Nuclear Information System (INIS)

Benedict, R.W.; Krsul, J.R.; Mariani, R.D.; Park, K.; Teske, G.M.

1993-01-01

In support of the metallic fuel cycle development for the Integral Fast Reactor (IFR), a small scale electrorefiner was built and operated in the Hot Fuel Examination Facility (HFEF) at Argonne National Laboratory-West. The initial purpose of this apparatus was to test the single segment dissolution of irradiated metallic fuel via either direct dissolution in cadmium or anodic dissolution. These tests showed that 99.95% of the uranium and 99.99% of the plutonium was dissolved and separated from the fuel cladding material. The fate of various fission products was also measured. After the dissolution experiments, the apparatus was upgraded to stady fission product behavior during uranium electrotransport. Preliminary decontamination factors were estimated for different fission products under different processing conditions. Later modifications have added the following capabilities: Dissolution of multiple fuel segments simultaneously, electrotransport to a solid cathode or liquid cathode and actinide recovery with a chemical reduction crucible. These capabilities have been tested with unirradiated uranium-zirconium fuel and will support the Fuel Cycle Demonstration program

Technical basis for the proposed high efficiency nuclear fuel program

International Nuclear Information System (INIS)

MacDonald, P.E.; Herring, J.S.; Crawford, D.C.; Neimark, L.E.

1999-01-01

Greenhouse gas emissions from fossil fired electricity generating stations will dramatically increase over the next 20 years. Nuclear energy is the only fully developed technology able to supply large amounts of electricity without generation of greenhouse gases. However, the problem of noncompetitive economics and public concerns about radioactive waste disposal, safety, and nuclear weapons proliferation may prevent the reemergence of nuclear power as a preferred option for new electric energy generation in the U.S. This paper discusses a new research program to help address these issues, by developing fuel designs capable of burnup values in excess of 60 MWD/kgU. The objectives of the program are to: improve the reliability and robustness of light water reactor fuel, thereby improving safety margins; Significantly increase the energy generated by each fuel loading, thereby achieving longer operating cycles, higher capacity factors, and lower cost electric power; Significantly reduce the volume of spent nuclear fuel discharged for disposal by allowing more energy to be extracted from each fuel element prior to discharge; Develop fuel that is much more proliferation resistant. (author)

Actinide recycle potential in the integral fast reactor (IFR) fuel cycle

International Nuclear Information System (INIS)

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

1991-01-01

In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The use of metallic fuel in the IFR allows a radically improved fuel cycle technology. Based on the recent IFR process development, a preliminary assessment has been made to investigate the feasibility of further adapting pyrochemical processes to directly extract actinides from LWR spent fuel. The results of this assessment indicate very promising potential and two most promising flowsheet options have been identified for further research and development. This paper also summarizes current thinking on the rationale for actinide recycle, its ramifications on the geologic repository and the current high-level waste management plans, and the necessary development programs

Impacts of the Weatherization Assistance Program in fuel-oil heated houses

Energy Technology Data Exchange (ETDEWEB)

Levins, W.P.; Ternes, M.P.

1994-10-01

In 1990, the US Department of Energy (DOE) initiated a national evaluation of its lowincome Weatherization Assistance Program. This report, which is one of five parts of that evaluation, evaluates the energy savings and cost-effectiveness of the Program as it had been applied to single-family houses heated primarily by fuel-oil. The study was based upon a representative sample (41 local weatherization agencies, 222 weatherized and 115 control houses) from the nine northeastern states during 1991 and 1992 program years. Dwelling-specific and agency-level data on measures installed, costs, and service delivery procedures were collected from the sampled agencies. Space-heating fuel-oil consumption, indoor temperature, and outdoor temperature were monitored at each house. Dwelling characteristics, air-leakage measurements, space-heating system steady-state efficiency measurements, safety inspections, and occupant questionnaires were also collected or performed at each monitored house. We estimate that the Program weatherized a total of 23,400 single-family fuel-oil heated houses in the nine northeastern states during program years 1991 and 1992. Annual fuel-oil savings were calculated using regression techniques to normalize the savings to standard weather conditions. For the northeast region, annual net fuel-oil savings averaged 160 gallons per house, or 17.7% of pre-weatherization consumption. Although indoor temperatures changed in individual houses following weatherization, there was no average change and no significant difference as compared to the control houses; thus, there was no overall indoor temperature takeback effect influencing fuel-oil savings. The weatherization work was performed cost effectively in these houses from the Program perspective, which included both installation costs and overhead and management costs but did not include non-energy benefits (such as employment and environmental).

Automotive Stirling engine development program. [fuel economy assessment

Science.gov (United States)

Kitzner, E. W.

1978-01-01

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

Preliminary investigation study of code of developed country for developing Korean fuel cycle code

International Nuclear Information System (INIS)

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

2012-01-01

In order to develop Korean fuel cycle code, the analyses has been performed with the fuel cycle codes which are used in advanced country. Also, recommendations were proposed for future development. The fuel cycle codes are AS FLOOWS: VISTA which has been developed by IAEA, DANESS code which developed by ANL and LISTO, and VISION developed by INL for the Advanced Fuel Cycle Initiative (AFCI) system analysis. The recommended items were proposed for software, program scheme, material flow model, isotope decay model, environmental impact analysis model, and economics analysis model. The described things will be used for development of Korean nuclear fuel cycle code in future

Fuel cells for transportation program: FY1997 national laboratory annual report

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

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

Future of alcohol fuels programs in Brasil

Science.gov (United States)

Carvalho, A. V., Jr.; Rechtschaffen, E.; Goldstein, L., Jr.

An updating is given of the Brazilian National Alcohol Program's production and utilization achievements to date in the substitution of ethanol and methanol for imported oil products. A series of Eucalyptus forestry and processing-industry projections are made for fuel output and jobs creation that may be expected by the year 2000. With few exceptions, methanol produced from wood grown on poorer soils than can now be used for sugarcane substitute for oil products and result in jobs creation several orders of magnitude higher than petroleum fuels.

Quality assurance program for surveillance of fast reactor mixed oxide fuel analytical chemistry

International Nuclear Information System (INIS)

Rein, J.E.; Zeigler, R.K.; Waterbury, G.R.; McClung, W.E.; Praetorius, P.R.; Delvin, W.L.

1976-01-01

An effective quality assurance program for the chemical analysis of nuclear fuel is essential to assure that the fuel will meet the strict chemical specifications required for optimum reactor performance. Such a program has been in operation since 1972 for the fuels manufactured for the Fast Flux Test Facility. This program, through the use of common quality control and calibration standards, has consistently provided high levels of agreement among laboratories in all areas of analysis. The paper presented gives a summary of the chemical specifications for the fuel and source material, an outline of the requirements for laboratory qualifications and the preparation of calibration and quality control materials, general administration details of the plan, and examples where the program has been useful in solving laboratory problems

Ageing management program for the Spanish low and intermediate level waste disposal and spent fuel and high-level waste centralised storage facilities

Science.gov (United States)

Zuloaga, P.; Ordoñez, M.; Andrade, C.; Castellote, M.

2011-04-01

The generic design of the centralised spent fuel storage facility was approved by the Spanish Safety Authority in 2006. The planned operational life is 60 years, while the design service life is 100 years. Durability studies and surveillance of the behaviour have been considered from the initial design steps, taking into account the accessibility limitations and temperatures involved. The paper presents an overview of the ageing management program set in support of the Performance Assessment and Safety Review of El Cabril low and intermediate level waste (LILW) disposal facility. Based on the experience gained for LILW, ENRESA has developed a preliminary definition of the Ageing Management Plan for the Centralised Interim Storage Facility of spent Fuel and High Level Waste (HLW), which addresses the behaviour of spent fuel, its retrievability, the confinement system and the reinforced concrete structure. It includes tests plans and surveillance design considerations, based on the El Cabril LILW disposal facility.

Ageing management program for the Spanish low and intermediate level waste disposal and spent fuel and high-level waste centralised storage facilities

Directory of Open Access Journals (Sweden)

Andrade C.

2011-04-01

Full Text Available The generic design of the centralised spent fuel storage facility was approved by the Spanish Safety Authority in 2006. The planned operational life is 60 years, while the design service life is 100 years. Durability studies and surveillance of the behaviour have been considered from the initial design steps, taking into account the accessibility limitations and temperatures involved. The paper presents an overview of the ageing management program set in support of the Performance Assessment and Safety Review of El Cabril low and intermediate level waste (LILW disposal facility. Based on the experience gained for LILW, ENRESA has developed a preliminary definition of the Ageing Management Plan for the Centralised Interim Storage Facility of spent Fuel and High Level Waste (HLW, which addresses the behaviour of spent fuel, its retrievability, the confinement system and the reinforced concrete structure. It includes tests plans and surveillance design considerations, based on the El Cabril LILW disposal facility.

A fast-running fuel management program for a CANDU reactor

International Nuclear Information System (INIS)

Choi, Hangbok

2000-01-01

A fast-running fuel management program for a CANDU reactor has been developed. The basic principle of this program is to select refueling channels such that the reference reactor conditions are maintained by applying several constraints and criteria when selecting refueling channels. The constraints used in this program are the channel and bundle power and the fuel burnup. The final selection of the refueling channel is determined based on the priority of candidate channels, which enhances the reactor power distribution close to the time-average model. The refueling simulation was performed for a natural uranium CANDU reactor and the results were satisfactory

Development of a Computer Program for the Analysis Logistics of PWR Spent Fuels

International Nuclear Information System (INIS)

Choi, Heui Joo; Choi, Jong Won; Cha, Jeong Hun

2008-01-01

It is expected that the temporary storage facilities at the nuclear power plants will be full of the spent fuels within 10 years. Provided that a centralized interim storage facility is constructed along the coast of the Korean peninsula to solve this problem, a substantial amount of spent fuels should be transported by sea or by land every year. In this paper we developed a computer program for the analysis of transportation logistics of the spent fuels from 4 different nuclear power plant sites to the hypothetical centralized interim storage facility and the final repository. Mass balance equations were used to analyze the logistics between the nuclear power plants and the interim storage facility. To this end a computer program, CASK, was developed by using the VISUAL BASIC language. The annual transportation rates of spent fuels from the four nuclear power plant sites were determined by using the CASK program. The parameter study with the program illustrated the easiness of logistics analysis. The program could be used for the cost analysis of the spent fuel transportation as well.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

FUEL-FLEXIBLE GASIFICATION-COMBUSTION TECHNOLOGY FOR PRODUCTION OF H2 AND SEQUESTRATION-READY CO2

International Nuclear Information System (INIS)

George Rizeq; Ravi Kumar; Janice West; Vitali Lissianski; Neil Widmer; Vladimir Zamansky

2001-01-01

It is expected that in the 21st century the Nation will continue to rely on fossil fuels for electricity, transportation, and chemicals. It will be necessary to improve both the thermodynamic efficiency and environmental impact performance of fossil fuel utilization. General Electric Energy and Environmental Research Corporation (GE-EER) has developed an innovative fuel-flexible Advanced Gasification-Combustion (AGC) concept to produce H(sub 2) and sequestration-ready CO(sub 2) from solid fuels. The AGC module offers potential for reduced cost and increased energy efficiency relative to conventional gasification and combustion systems. GE-EER was awarded a Vision-21 program from U.S. DOE NETL to develop the AGC technology. Work on this three-year program started on October 1, 2000. The project team includes GE-EER, California Energy Commission, Southern Illinois University at Carbondale, and T. R. Miles, Technical Consultants, Inc. In the AGC technology, coal/opportunity fuels and air are simultaneously converted into separate streams of (1) pure hydrogen that can be utilized in fuel cells, (2) sequestration-ready CO(sub 2), and (3) high temperature/pressure oxygen depleted air to produce electricity in a gas turbine. The process produces near-zero emissions and, based on preliminary modeling work in the first quarter of this program, has an estimated process efficiency of approximately 67% based on electrical and H(sub 2) energy outputs relative to the higher heating value of coal. The three-year R and D program will determine the operating conditions that maximize separation of CO(sub 2) and pollutants from the vent gas, while simultaneously maximizing coal conversion efficiency and hydrogen production. The program integrates lab-, bench- and pilot-scale studies to demonstrate the AGC concept. This is the 1st quarterly progress report for the Vision-21 AGC program supported by U.S. DOE NETL (Contract: DE-FC26-00FT40974). This report summarizes program

The Australian Hydrogen and Fuel Cells Education Program

International Nuclear Information System (INIS)

Luigi Bonadio

2006-01-01

The next generation of engineers and scientists will face great technical, economic and political challenges to satisfy increasing demands for a secure, reliable and affordable global energy system that maintains and enhances current standards of living. The Australian Hydrogen and Fuel Cells Education Program aims to bolster the quality and relevance of primary and secondary school teaching in emerging areas of science, technology and environmental/sustainability studies using hydrogen, in its capacity as a versatile energy carrier, as the educational basis for teacher and student learning. Critical advances in specific areas of hydrogen production, distribution, storage and end-use technologies arise when students are engaged to develop and apply a broad range of disciplinary and interdisciplinary knowledge and practical skills. A comprehensive hydrogen and fuel cell technology teaching module will be developed to complement existing fuels and energy curricula across Australian schools. The pilot program will be delivered via the collaboration of nine trial schools, a broad range of technical and pedagogy experts and representatives of professional bodies and industry. The program features essential and extensive teacher consultation, a professional learning and development course, industry site visits and a dedicated research and evaluation study. This initiative aims to bolster teacher literacy and student participation in the design, construction and operation of various hydrogen and fuel cell devices and extended activities. Students will reflect on and formally present their learning experiences via several dedicated fora including an awards ceremony where outstanding performance of leading schools, teachers and student groups within the cluster will be acknowledged. (authors)

The Australian Hydrogen and Fuel Cells Education Program

Energy Technology Data Exchange (ETDEWEB)

Luigi Bonadio [Senior Consultant Luigi Bonadio and Associates (Australia)

2006-07-01

The next generation of engineers and scientists will face great technical, economic and political challenges to satisfy increasing demands for a secure, reliable and affordable global energy system that maintains and enhances current standards of living. The Australian Hydrogen and Fuel Cells Education Program aims to bolster the quality and relevance of primary and secondary school teaching in emerging areas of science, technology and environmental/sustainability studies using hydrogen, in its capacity as a versatile energy carrier, as the educational basis for teacher and student learning. Critical advances in specific areas of hydrogen production, distribution, storage and end-use technologies arise when students are engaged to develop and apply a broad range of disciplinary and interdisciplinary knowledge and practical skills. A comprehensive hydrogen and fuel cell technology teaching module will be developed to complement existing fuels and energy curricula across Australian schools. The pilot program will be delivered via the collaboration of nine trial schools, a broad range of technical and pedagogy experts and representatives of professional bodies and industry. The program features essential and extensive teacher consultation, a professional learning and development course, industry site visits and a dedicated research and evaluation study. This initiative aims to bolster teacher literacy and student participation in the design, construction and operation of various hydrogen and fuel cell devices and extended activities. Students will reflect on and formally present their learning experiences via several dedicated fora including an awards ceremony where outstanding performance of leading schools, teachers and student groups within the cluster will be acknowledged. (authors)

The American 'severe fuel damage program'

International Nuclear Information System (INIS)

Sdouz, G.

1982-03-01

The TMI-2 accident has initiated a new phase of safety research. It is necessary to consider severe accidents with degraded or molten core. For NRC there was a need for an improved understanding of this reactor behaviour and the 'Severe Fuel Dage Program' was initiated. Planned are in-pile experiments in PBF, NRU and ESSOR and in addition separate effects tests and results from TMI-2. The analytical component of the program is the development of different versions of the code SCDAP for the detailed analysis during severe accident transients. (Author) [de

OVERVIEW OF ADVANCED PETROLEUM-BASED FUELS-DIESEL EMISSIONS CONTROL PROGRAM (APBF-DEC)

Energy Technology Data Exchange (ETDEWEB)

Sverdrup, George M.

2000-08-20

The Advanced Petroleum-Based Fuels-Diesel Emissions Control Program (APBF-DEC) began in February 2000 and is supported by government agencies and industry. The purpose of the APBF-DEC program is to identify and evaluate the optimal combinations of fuels, lubricants, diesel engines, and emission control systems to meet the projected emission standards for the 2000 to 2010 time period. APBF-DEC is an outgrowth of the earlier Diesel Emission Control-Sulfur Effects Program (DECSE), whose objective is to determine the impact of the sulfur levels in fuel on emission control systems that could lower the emissions of NOx and particulate matter (PM) from diesel powered vehicles in the 2002 to 2004 period. Results from the DECSE studies of two emission control technologies-diesel particle filter (DPF) and NOx adsorber-will be used in the APBF-DEC program. These data are expected to provide initial information on emission control technology options and the effects of fuel properties (including additives) on the performance of emission control systems.

Fire, Fuel, and Smoke Science Program 2015 Research Accomplishments

Science.gov (United States)

Faith Ann Heinsch; Charles W. McHugh; Colin C. Hardy

2016-01-01

The Fire, Fuel, and Smoke Science Program (FFS) of the U.S. Forest Service, Rocky Mountain Research Station focuses on fundamental and applied research in wildland fire, from fire physics and fire ecology to fuels management and smoke emissions. Located at the Missoula Fire Sciences Laboratory in Montana, the scientists, engineers, technicians, and support...

Spent fuel sabotage aerosol ratio program : FY 2004 test and data summary.

Energy Technology Data Exchange (ETDEWEB)

Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Mo, Tin (U.S. Nuclear Regulatory Commission, Washington, DC); Billone, Michael C. (Argonne National Laboratory, Argonne, IL); Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Young, F. I. (U.S. Nuclear Regulatory Commission, Washington, DC); Coats, Richard Lee; Burtseva, Tatiana (Argonne National Laboratory, Argonne, IL); Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Thompson, Nancy Slater (U.S. Department of Energy, Washington, DC); Hibbs, Russell S. (U.S. Department of Energy, Washington, DC); Gregson, Michael Warren; Lange, Florentin (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Molecke, Martin Alan; Tsai, Han-Chung (Argonne National Laboratory, Argonne, IL)

2005-07-01

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 has been underway for several years. This program provides data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of

Spent fuel sabotage aerosol ratio program : FY 2004 test and data summary

International Nuclear Information System (INIS)

Brucher, Wenzel; Koch, Wolfgang; Pretzsch, Gunter Guido; Loiseau, Olivier; Mo, Tin; Billone, Michael C.; Autrusson, Bruno A.; Young, F. I.; Coats, Richard Lee; Burtseva, Tatiana; Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver; Thompson, Nancy Slater; Hibbs, Russell S.; Gregson, Michael Warren; Lange, Florentin; Molecke, Martin Alan; Tsai, Han-Chung

2005-01-01

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 has been underway for several years. This program provides data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of

Ciclon: A neutronic fuel management program for PWR's consecutive cycles

International Nuclear Information System (INIS)

Aragones, J.M.

1977-01-01

The program description and user's manual of a new computer code is given. Ciclon performs the neutronic calculation of consecutive reload cycles for PWR's fuel management optimization. Fuel characteristics and burnup data, region or batch sizes, loading schemes and state of previously irradiated fuel are input to the code. Cycle lengths or feed enrichments and burnup sharing for each region or batch are calculate using different core neutronic models and printed or punched in standard fuel management format. (author) [es

Preliminary modelling study of geochemical interactions between a used-fuel disposal vault and the surrounding geosphere

International Nuclear Information System (INIS)

McMurry, J.

1995-10-01

In the Environmental Impact Statement (EIS) and the related documents that describe the Canadian nuclear fuel waste disposal concept (AECL 1994), it has been assumed that a disposal vault would have no significant geochemical impact on the geosphere, and so no such effects were included explicitly in the postclosure assessment model. The purpose of this study was to estimate the general magnitude and significance of vault-induced geochemical changes over an expected range of temperatures. The results of the preliminary modelling are used broadly to evaluate the implications of these changes for the migration of radionuclides through the geosphere. The geochemical modelling program PHREEQE was used to calculate the changes in mineral solubilities that would result from the transfer of aqueous species from the vault to the geosphere or that would result from groundwater-granite interactions enhanced by vault-derived elevated temperatures. Twelve representative vault water compositions, derived from predicted interactions with buffer material and backfill over a range of temperatures up to 95 deg C, were used in the modelling. For the conditions modelled it was determined that the interactions of the geosphere with dissolved vault constituents, and the relatively modest maximum increase in groundwater temperature produced by a vault, would have a limited impact on the geosphere. The conclusions of this preliminary study are qualified by some of the simplifying assumptions used in the modelling. More realistic modelling of natural systems requires a more detailed representation of water-solid interactions with a variety of vault materials at elevated temperatures. (author) 48 refs., 13 tabs, 4 figs

The Canadian program for management of spent fuel and high level wastes

International Nuclear Information System (INIS)

Barnes, R.W.; Mayman, S.A.

A brief history and description of the nuclear power program in Canada is given. Schedules and programs are described for storing spent fuel in station fuel bays, centralized water pool storage facilities, concrete canisters, convection vaults, and rock or salt formations. High-level wastes will be retrievable initially, therefore the focus is on storage in mined cavities. The methods developed for high-level waste storage/disposal will ideally be flexible enough to accommodate spent fuel. (E.C.B.)

Fire, Fuel, and Smoke Science Program: 2013 Research accomplishments

Science.gov (United States)

Faith Ann Heinsch; Robin J. Innes; Colin C. Hardy; Kristine M. Lee

2014-01-01

The Fire, Fuel, and Smoke Science Program (FFS) of the U.S. Forest Service, Rocky Mountain Research Station, focuses on fundamental and applied research in wildland fire, from fire physics and fire ecology to fuels management and smoke emissions. Located at the Missoula Fire Sciences Laboratory in Montana, the scientists, engineers, technicians, and support staff in...

Trade-off results and preliminary designs of Near-Term Hybrid Vehicles

Science.gov (United States)

Sandberg, J. J.

1980-01-01

Phase I of the Near-Term Hybrid Vehicle Program involved the development of preliminary designs of electric/heat engine hybrid passenger vehicles. The preliminary designs were developed on the basis of mission analysis, performance specification, and design trade-off studies conducted independently by four contractors. THe resulting designs involve parallel hybrid (heat engine/electric) propulsion systems with significant variation in component selection, power train layout, and control strategy. Each of the four designs is projected by its developer as having the potential to substitute electrical energy for 40% to 70% of the petroleum fuel consumed annually by its conventional counterpart.

The KFB Program on Biobased Fuels for Vehicles

International Nuclear Information System (INIS)

1996-12-01

KFB supports research and demonstration projects for bio-based transport fuels, alcohols and biogas. The program started in 1991 and will continue through 1997. The program focuses on heavy vehicles, e.g. buses for public transportation. Projects and intermediate results are described in the brochure. Information is also available at the KFB homepage. //www.kfb.se

IAEA activities related to research reactor fuel conversion and spent fuel return programs

International Nuclear Information System (INIS)

Goldman, Ira N.; Adelfang, Pablo; Ritchie, Iain G.

2005-01-01

The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programs have supported research reactor fuel conversion from HEU to low enriched uranium (LEU), and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. (author)

CERCA'S experience in UMO fuel manufacturing

International Nuclear Information System (INIS)

Jarousse, Ch.; Lavastre, Y.; Grasse, M.

2003-01-01

Considered as a suitable solution for non-proliferation and reprocessing purposes, UMo fuel has been chosen and studied by the RERTR program since 1996. Involved in the RERTR fuel developments since 1978, with more than 20 years of U 3 SI 2 fuel production, and closely linked to the French Commissariat a l'Energie Atomique, CERCA was able to define properly, from the beginning, the right R and D actions plan for UMo fuel development. CERCA has already demonstrated during the last 4 years its ability to manufacture plates and fuel elements with high density UMo fuel. UMo full size plates produced for 4 irradiation experiments in 3 European reactors afforded us a unique experience. In addition, as a main part of our R and D effort, we have always studied in depth a key part of the CERCA process outline which is the plate rolling stage. After some preliminary investigation in order to define the phenomenological model describing the behavior of the fuel core when rolling, we have developed a rolling digital simulator. (author)

Computerised programming of the Dragon reactor fuel handling operations

International Nuclear Information System (INIS)

Butcher, P.

1976-11-01

Two suites of FORTRAN IV computer programs have been written to produce check lists for the operation of the two remote control fuel handling machines of the Dragon Reactor. This document describes the advantages of these programs over the previous manual system of writing check lists, and provides a detailed guide to the programs themselves. (author)

Super-Efficient Refrigerator Program (SERP) evaluation volume 2: Preliminary impact and market transformation assessment

Energy Technology Data Exchange (ETDEWEB)

Lee, A.D.; Conger, R.L.

1996-08-01

The Super Efficient Refrigerator Program (SERP) is a collaborative utility program intended to transform the market for energy-efficient and environmentally friendly refrigerators. It is one of the first examples of a large-scale {open_quotes}market transformation{close_quotes} energy efficiency program. This report documents the preliminary impact and market transformation evaluation of SERP ({open_quotes}the Program{close_quotes}). Pacific Northwest National Laboratory (PNNL) conducted this evaluation for the U.S. Department of Energy. This study focuses on the preliminary impact evaluation and market transformation assessment, but also presents limited process evaluation information. It is based on interviews with refrigerator dealers and manufacturers, interviews with utility participants, industry data, and information from the Program administrators. Results from this study complement those from prior process evaluation also conducted by PNNL. 42 refs., 5 figs., 4 tabs.

Status of Away From Reactor spent fuel storage program

International Nuclear Information System (INIS)

King, F.D.

1979-07-01

The Away From Reactor (AFR) Spent Fuel Program that the US Department of Energy established in 1977 is intended to preclude the shutting down of commercial nuclear power reactors because of lack of storage space for spent fuel. Legislation now being considered by Congress includes plans to provide storage space for commercial spent fuel beginning in 1983. Utilities are being encouraged to provide as much storage space as possible in their existing storage facilities, but projections indicate that a significant amount of AFR storage will be required. The government is evaluating the use of both existing and new storage facilities to solve this forecasted storage problem for commercial spent fuel

ROBOT3: a computer program to calculate the in-pile three-dimensional bowing of cylindrical fuel rods (AWBA Development Program)

International Nuclear Information System (INIS)

Kovscek, S.E.; Martin, S.E.

1982-10-01

ROBOT3 is a FORTRAN computer program which is used in conjunction with the CYGRO5 computer program to calculate the time-dependent inelastic bowing of a fuel rod using an incremental finite element method. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the CYGRO5 axisymmetric model. Fuel rod supports are modeled as displacement, force, or spring-type nodal boundary conditions. The program input is described and a sample problem is given

75 FR 79964 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program

Science.gov (United States)

2010-12-21

... renewable fuel facility to be considered biogas for purposes of Table 1 to Sec. 80.1426; --Sec. 80.1452(b... explained that many developers of ethanol facilities, including their own, sought to obtain construction permits without going through EPA's New Source Review (NSR) program, and were able to do so by obtaining...

Methodology and a preliminary data base for examining the health risks of electricity generation from uranium and coal fuels

Energy Technology Data Exchange (ETDEWEB)

El-Bassioni, A.A.

1980-08-01

An analytical model was developed to assess and examine the health effects associated with the production of electricity from uranium and coal fuels. The model is based on a systematic methodology that is both simple and easy to check, and provides details about the various components of health risk. A preliminary set of data that is needed to calculate the health risks was gathered, normalized to the model facilities, and presented in a concise manner. Additional data will become available as a result of other evaluations of both fuel cycles, and they should be included in the data base. An iterative approach involving only a few steps is recommended for validating the model. After each validation step, the model is improved in the areas where new information or increased interest justifies such upgrading. Sensitivity analysis is proposed as the best method of using the model to its full potential. Detailed quantification of the risks associated with the two fuel cycles is not presented in this report. The evaluation of risks from producing electricity by these two methods can be completed only after several steps that address difficult social and technical questions. Preliminary quantitative assessment showed that several factors not considered in detail in previous studies are potentially important. 255 refs., 21 figs., 179 tabs.

Summary of NRC LWR safety research programs on fuel behavior, metallurgy/materials and operational safety

International Nuclear Information System (INIS)

Bennett, G.L.

1979-09-01

The NRC light-water reactor safety-research program is part of the NRC regulatory program for ensuring the safety of nuclear power plants. This paper summarizes the results of NRC-sponsored research into fuel behavior, metallurgy and materials, and operational safety. The fuel behavior research program provides a detailed understanding of the response of nuclear fuel assemblies to postulated off-normal or accident conditions. Fuel behavior research includes studies of basic fuel rod properties, in-reactor tests, computer code development, fission product release and fuel meltdown. The metallurgy and materials research program provides independent confirmation of the safe design of reactor vessels and piping. This program includes studies on fracture mechanics, irradiation embrittlement, stress corrosion, crack growth, and nondestructive examination. The operational safety research provides direct assistance to NRC officials concerned with the operational and operational-safety aspects of nuclear power plants. The topics currently being addressed include qualification testing evaluation, fire protection, human factors, and noise diagnostics

Coated Particle Fuel and Deep Burn Program Monthly Highlights May 2011

International Nuclear Information System (INIS)

Snead, Lance Lewis; Bell, Gary L.; Besmann, Theodore M.

2011-01-01

During FY 2011 the CP and DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for April 2011, ORNL/TM-2011/125, was distributed to program participants on May 10, 2011. As reported previously, the final Quarterly for FY 2010, Deep Burn Program Quarterly Report for July - September 2010, ORNL/TM-2010/301, was announced to program participants and posted to the website on December 28, 2010. This report discusses the following: (1) Fuel Performance Modeling - Fuel Performance Analysis; (2) Thermochemical Data and Model Development - (a) Thermochemical Modeling, (b) Thermomechanical Modeling, (c) Actinide and Fission Product Transport; (3) TRU (transuranic elements) TRISO (tri-structural isotropic) Development - (a) TRU Kernel Development, (b) Coating Development; and (4) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing.

Coated Particle Fuel and Deep Burn Program Monthly Highlights June 2011

International Nuclear Information System (INIS)

Snead, Lance Lewis; Bell, Gary L.; Besmann, Theodore M.

2011-01-01

During FY 2011 the CP and DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for May 2011, ORNL/TM-2011/126, was distributed to program participants on June 9, 2011. As reported previously, the final Quarterly for FY 2010, Deep Burn Program Quarterly Report for July - September 2010, ORNL/TM-2010/301, was announced to program participants and posted to the website on December 28, 2010. This report discusses the following: (1) Fuel Performance Modeling - Fuel Performance Analysis; (2) Thermochemical Data and Model Development - (a) Thermochemical Behavior, (b) Thermomechanical Modeling, (c) Actinide and Fission Product Transport; (3) TRU (transuranic elements) TRISO (tri-structural isotropic) Development - (a) TRU Kernel Development, (b) Coating Development; and (4) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing.

Industry perspectives on the U.S. used nuclear fuel management program

International Nuclear Information System (INIS)

Matzie, R.

2001-01-01

The U.S. used nuclear fuel management program is at a critical juncture. It has been 17 years since the U.S. Congress authorized geologic disposal of used nuclear fuel and 12 years since Congress elected to study Yucca Mountain, Nevada, as the nation's lone long-term disposal option. During this period, Yucca Mountain's ability to isolate long-lived radionuclides was the subject of considerable scientific study. Concurrently, specialists expended significant design effort on the engineered barriers that work in conjunction with the proposed repository's natural features to ensure that health protection standards are met. Confidence as to the soundness of the geologic disposal option is high. While technical progress on geologic disposal is substantial, policy level decision-making needed to move the program forward remains lacking. A number of important policy questions concerning used fuel management remain unanswered as this fuel accumulates at 103 operating U.S. commercial nuclear power reactors at 65 sites. Wet fuel storage space at those locations is becoming increasingly scarce. This paper will review both the scientific successes and policy challenges on the road to geologic disposal. It will outline the nation's used nuclear fuel storage needs, highlight what the nation's scientists have learned about geologic disposal, discuss what is being done to meet near-term interim storage requirements, describe the emerging regulatory framework associated with the program and summarize the ongoing policy debate. (author)

The new revision of NPP Krsko decommissioning, radioactive waste and spent fuel management program: analyses and results

International Nuclear Information System (INIS)

Zeleznik, Nadja; Kralj, Metka; Lokner, Vladimir; Levanat, Ivica; Rapic, Andrea; Mele, Irena

2010-01-01

The preparation of the new revision of the Decommissioning and Spent Fuel (SF) and Low and Intermediate level Waste (LILW) Disposal Program for the NPP Krsko (Program) started in September 2008 after the acceptance of the Term of Reference for the work by Intergovernmental Committee responsible for implementation of the Agreement between the governments of Slovenia and Croatia on the status and other legal issues related to investment, exploitation, and decommissioning of the Nuclear power plant Krsko. The responsible organizations, APO and ARAO together with NEK prepared all new technical and financial data and relevant inputs for the new revision in which several scenarios based on the accepted boundary conditions were investigated. The strategy of immediate dismantling was analyzed for planned and extended NPP life time together with linked radioactive waste and spent fuel management to calculate yearly annuity to be paid by the owners into the decommissioning funds in Slovenia and Croatia. The new Program incorporated among others new data on the LILW repository including the costs for siting, construction and operation of silos at the location Vrbina in Krsko municipality, the site specific Preliminary Decommissioning Plan for NPP Krsko which included besides dismantling and decontamination approaches also site specific activated and contaminated radioactive waste, and results from the referenced scenario for spent fuel disposal but at very early stage. Important inputs for calculations presented also new amounts of compensations to the local communities for different nuclear facilities which were taken from the supplemented Slovenian regulation and updated fiscal parameters (inflation, interest, discount factors) used in the financial model based on the current development in economical environment. From the obtained data the nominal and discounted costs for the whole nuclear program related to NPP Krsko which is jointly owned by Slovenia and Croatia have

Nonproliferation and safeguards aspects of fuel cycle programs in reduction of excess separated plutonium and high-enriched uranium

International Nuclear Information System (INIS)

Persiani, P.J.

1995-01-01

The purpose of this preliminary investigation is to explore alternatives and strategies aimed at the gradual reduction of the excess inventories of separated plutonium and high-enriched uranium (HEU) in the civilian nuclear power industry. The study attempts to establish a technical and economic basis to assist in the formation of alternative approaches consistent with nonproliferation and safeguards concerns. Reference annual mass flows and inventories for a representative 1,400 Mwe Pressurized Water Reactor (PWR) fuel cycle have been investigated for three cases: the 100 percent uranium oxide UO 2 fuel loading once through cycle, and the 33 percent mixed oxide MOX loading configuration for a first and second plutonium recycle. The analysis addresses fuel cycle developments; plutonium and uranium inventory and flow balances; nuclear fuel processing operations; UO 2 once-through and MOX first and second recycles; and the economic incentives to draw-down the excess separated plutonium stores. The preliminary analysis explores several options in reducing the excess separated plutonium arisings and HEU, and the consequences of the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials on nonproliferation and safeguards policy assessments

Fuel performance improvement program. Quarterly/annual progress report, October 1977--September 1978

International Nuclear Information System (INIS)

Crouthamel, C.E.

1978-10-01

This quarterly/annual report reviews and summarizes the activities performed in support of the Fuel Performance Improvement Program (FPIP) during Fiscal Year 1978 with emphasis on those activities that transpired during the quarter ending September 30, 1978. Significant progress has been made in achieving the primary objectives of the program, i.e., to demonstrate commercially viable fuel concepts with improved fuel - cladding interaction (FCI) behavior. This includes out-of-reactor experiments to support the fuel concepts being evaluated, initiation of instrumented test rod experiments in the Halden Boiling Water Reactor (HBWR), and fabrication of the first series of demonstration rods for irradiation in the Big Rock Point Reactor

Program description for the program Fuel program sustainability July 1, 2011 through June 30, 2015; Programbeskrivning foer programmet Braensleprogrammet haallbarhet 1 juli 2011 till och med 30 juni 2015

Energy Technology Data Exchange (ETDEWEB)

2011-04-15

The Fuel program sustainability is included as one of three programs in a cohesive commitment to increased, sustainable and efficient production and use of indigenous and renewable fuels that are implemented by the Swedish Energy Agency from July 1, 2011 to June 30, 2015. The program focuses on issues of environment and ecological sustainability of the production of biofuels, and systems and resource issues concerning the use. It does not include purely technical issues related to biofuels production or processes in which fuel is converted to heat, electricity or fuel. Questions about fuel supply and fuel processing / scale combustion are treated in the parallel running programs; the Fuel program supply and the Fuel program conversion. The four-year program will commence on July 1, 2011 and will run until June 30, 2015 and has an annual budget of total SEK 16 millions. Ambiguities in the environmental area may through various regulations mean various forms of barriers on the biofuel market. The Fuel program sustainability aims to sort out such ambiguities and, if possible, eliminate such obstacles, identify solutions and develop opportunities. The availability of biofuels and croplands is limited relative to needs. Thus the program also aims to describe the resource efficiency and climate benefits of current biofuel chains, and the possibility of using instruments to stimulate good practice. The program consists of sub-areas that partly overlap. - The sub-area Environment and sustainability focuses on how biofuel production will be designed to meet national and international environmental objectives and sustainability criteria. - The sub-area Biofuels and greenhouse gases describes different climate aspects related to production and exploitation of biofuels. - The sub-area System and market focuses on resource- and climate-efficient solutions in a system perspective, and how the bio-energy system can be affected by policy instruments

Fuelling study of CANDU reactors using neutron absorber poisoned fuel

Energy Technology Data Exchange (ETDEWEB)

Song, J.J.; Chan, P.K.; Bonin, H.W., E-mail: s25815@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

2014-07-01

A comparative fuelling study is conducted to determine the potential gain in operating margin for CANDU reactors incurred by implementing a change to the design of the conventional 37-element natural uranium (NU) fuel. The change involves insertion of minute quantities of neutron absorbers, Gd{sub 2}O{sub 3} and Eu{sub 2}O{sub 3}, into the fuel pellets. The Reactor Fuelling Simulation Program (RFSP) is used to conduct core-following simulations, for the regular 37-element NU fuel, which is to be used as control for comparison. Preliminary results are presented for fuelling with the regular 37-element NU fuel, which indicate constraints on fuelling that may be relaxed with addition of neutron absorbers. (author)

Preliminary results of the BTF-104 experiment: an in-reactor test of fuel behaviour and fission-product release and transport under LOCA/LOECC conditions

Energy Technology Data Exchange (ETDEWEB)

Dickson, L W; Elder, P H; Devaal, J W; Irish, J D; Yamazaki, A R [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

1996-12-31

The BTF-104 experiment is one of a series of in-reactor tests being performed to measure fuel behaviour and fission-product release from nuclear fuel subjected to accident conditions. The primary objective of the BTF-104 experiment was to measure fission-product releases from a CANDU-sized fuel element under combined Loss-of-Coolant Accident (LOCA) and Loss-of-Emergency-Core-Cooling (LOECC) conditions at an average fuel temperature of about 1550 deg C. The preliminary results of the BTF-104 experiment are presented in this paper. (author). 6 refs., 12 figs.

Voluntary Truck and Bus Fuel-Economy-Program marketing plan. Final technical report, September 29, 1980-January 29, 1982

Energy Technology Data Exchange (ETDEWEB)

None

1982-01-01

The aim of the program is to improve the utilization of fuel by commercial trucks and buses by updating and implementing specific approaches for educating and monitoring the trucking industry on methods and means of conserving fuels. The following outlines the marketing plan projects: increase use of program logo by voluntary program members and others; solicit trade publication membership and support; brief Congressional delegations on fuel conservation efforts; increase voluntary program presence before trade groups; increase voluntary program presence at truck and trade shows; create a voluntary program display for use at trade shows and in other areas; review voluntary program graphics; increase voluntary program membership; and produce placemats carrying fuel conservation messages; produce a special edition of Fuel Economy News, emphasizing the driver's involvement in fuel conservation; produce posters carrying voluntary program fuel conservation message. Project objectives, activities, and results for each project are summarized.

HTGR Fuel Technology Program. Semiannual report for the period ending March 31, 1981

International Nuclear Information System (INIS)

1981-05-01

This document reports the technical accomplishments on the HTGR Fuel Technology Program at General Atomic during the first half of FY-81. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, and core component verification testing tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR with a follow-on reformer (R) version. An important effort which was initiated during this period was the preparation of input data for a long-range technology program plan

HTGR Fuel-Technology Program. Semiannual report for the period ending September 30, 1982

International Nuclear Information System (INIS)

1982-11-01

This document reports the technical accomplishments on the HTGR Fuel Technology Program at GA Technologies Inc. during the second half of FY-1982. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, and core component verification testing tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR with a follow-on reformer (R) version. An important effort which was completed during this period was the preparation of input data for a long-range technology program plan

Increased fuel economy in transportation systems by use of energy management. Third year's program. Final report, May 1, 1976--July 1, 1976

Energy Technology Data Exchange (ETDEWEB)

Beachley, N.H.; Frank, A.A.

1976-07-01

A report is given of the results accomplished during the third year of a three-year research program, the overall goal of which has been to conceive and evaluate practical ways to increase automobile fuel economy by energy management within the engine-transmission-vehicle system. The third year was devoted primarily to the detailed design, construction, and preliminary evaluation of a Flywheel Energy Management Powerplant (FEMP) installed in a Pinto. The vehicle has been built to experimentally verify performance simulations and to allow the practical aspects of a real flywheel vehicle to be studied. The FEMP consists basically of an internal combustion engine, a high-speed energy-storage flywheel, and a hydrostatic power-split continuously-variable transmission (CVT) system. The flywheel drives the car, and the engine comes on to ''recharge'' it (with efficient wide-open throttle operation) only when the flywheel speed drops below a predetermined value. The concept also permits effective and efficient regenerative braking. Computer simulations have indicated an improvement in city fuel mileage of about 50%, with improvements of 100% appearing feasible with further research. Preliminary testing of the car shows favorable performance.

Vehicle Technologies and Fuel Cell Technologies Program: Prospective Benefits Assessment Report for Fiscal Year 2016

Energy Technology Data Exchange (ETDEWEB)

Stephens, T. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Taylor, C. H. [TA Engineering, Inc., Catonsville, MD (United States); Moore, J. S. [TA Engineering, Inc., Catonsville, MD (United States); Ward, J. [United States Department of Energy, Washington, DC (United States). Office of Energy Efficiency and Renewable Energy

2016-02-23

Under a diverse set of programs, the Vehicle Technologies and Fuel Cell Technologies offices of DOE’s Office of Energy Efficiency and Renewable Energy invest in research, development, demonstration, and deployment of advanced vehicle, hydrogen production, delivery and storage, and fuel cell technologies. This report estimates the benefits of successfully developing and deploying these technologies (a “Program Success” case) relative to a base case (the “No Program” case). The Program Success case represents the future with completely successful deployment of Vehicle Technologies Office (VTO) and Fuel Cell Technologies Office (FCTO) technologies. The No Program case represents a future in which there is no contribution after FY 2016 by the VTO or FCTO to these technologies. The benefits of advanced vehicle, hydrogen production, delivery and storage, and fuel cell technologies were estimated on the basis of differences in fuel use, primary energy use, and greenhouse gas (GHG) emissions from light-, medium- and heavy-duty vehicles, including energy and emissions from fuel production, between the base case and the Program Success case. Improvements in fuel economy of various vehicle types, growth in the stock of fuel cell vehicles and other advanced technology vehicles, and decreased GHG intensity of hydrogen production and delivery in the Program Success case over the No Program case were projected to result in savings in petroleum use and GHG emissions. Benefits were disaggregated by individual program technology areas, which included the FCTO program and the VTO subprograms of batteries and electric drives; advanced combustion engines; fuels and lubricants; materials (for reduction in vehicle mass, or “lightweighting”); and, for medium- and heavy-duty vehicles, reduction in rolling and aerodynamic resistance. Projections for the Program Success case indicate that by 2035, the average fuel economy of on-road, light-duty vehicle stock could be 47% to 76

Third annual report of the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

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

1981-12-01

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

Fourth annual report of the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

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

1982-12-01

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

Preliminary report on the experiment performed in MARIUS reactor loaded with teledial fuel

Energy Technology Data Exchange (ETDEWEB)

Estiot, J C; Morier, F

1972-06-15

The experimental work described in this paper is part of a collaborative programme agreed between CEA and the Dragon Project. The aim of the programme is the measurement of the relative conversion ratio in a reactor loaded with Teledial fuel elements. The results will allow us to check our calculational methods and assumptions upon which the calculations are based, in the case of a teledial core, which represents a very complicated geometry, specially, due to the presence of the U238 with its resonance. The programme of experiments described in the paper have been completed. Some preliminary results are presented in the second part of this report (Part 2).

National Jet Fuels Combustion Program – Area #3 : Advanced Combustion Tests

Science.gov (United States)

2017-12-31

The goal of this study is to develop, conduct, and analyze advanced laser and optical measurements in the experimental combustors developed under ASCENT National Fuel Combustion Program to measure sensitivity to fuel properties. We conducted advanced...

Second annual report of the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Boulton, J.

1980-12-01

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

Skidder load capacity and fuel consumption HP-41C program

Science.gov (United States)

Ross A. Phillips

1983-01-01

This program gives the log weight that the skidder can move and gives fuel consumption either in liters or gallons per turn. Slope of the skid trail, skidder weight, and skid distance must be entered into the program.

40 CFR 80.610 - What acts are prohibited under the diesel fuel sulfur program?

Science.gov (United States)

2010-07-01

... AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel... supply, store or transport motor vehicle diesel fuel, NRLM diesel fuel, ECA marine fuel or heating oil... transport any diesel fuel for use in motor vehicle or nonroad engines that contains greater than 0.10...

A computer program for calculation of the fuel cycle in pressurized water reactors

International Nuclear Information System (INIS)

Solanilla, R.

1976-01-01

The purpose of the FUCEFURE program is two-fold: first, it is designed to solve the problem of nuclear fuel cycle cost in one pressurized light water reactor calculation. The code was developed primarily for comparative and sensitivity studies. The program contains simple correlations between exposure and available depletion data used to predict the uranium and plutonium content of the fuel as a function of the fuel initial enrichment. Second, it has been devised to evaluate the nuclear fuel demand associated with an expanding nuclear power system. Evaluation can be carried out at any time and stage in the fuel cycle. The program can calculate the natural uranium and separate work requirements of any final and tails enrichment. It also can determine the nuclear power share of each reactor in the system when a decision has been made about the long-term nuclear power installations to be used and the types of PWR and fast breeder reactor characteristics to be involved in them. (author)

CANDU advanced fuel R and D programs for 1997 - 2006 in Korea

International Nuclear Information System (INIS)

Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J.

1997-01-01

KAERI has a comprehensive product development program of CANFLEX and DUPIC fuels to introduce them into CANDU reactors in Korea and a clear vision of how the product will evolve over the next 10 years. CANDU reactors are not the majority of nuclear power plants in Korea, but they produce significant electricity to contribute Korea's economic growth as well as to satisfy the need for energy. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology The CANFLEX and DUPIC R and D programs are conducted currently under the second stage of Korea's Nuclear Energy R and D Project as a national mid- and long-term program over the next 10 years from 1997 to 2006. The specific activities of the programs have taken account of the domestic and international environment concerning on non-proliferation in the Peninsula of Korea. As the first of the development products in the short-term, the CANFLEX-NU fuel will be completely developed jointly by KAERI/AECL and will be useful for the older CANDU-6 Wolsong unit 1. As the second product, the CANFLEX-0.9 % equivalent SEU fuel is expected to be completely developed within the next decade. It will be used in CANDU-6 reactors in Korea immediately after the development, if the existing RU in the world is price competitive with natural uranium. The DUPIC R and D program, as a long term program, is expected to demonstrate the possibility of use of used PWR fuel in CANDU reactors in Korea during the next 10 years. The pilot scale fabrication facility would be completed around 2010. (author)

CANDU advanced fuel R and D programs for 1997 - 2006 in Korea

Energy Technology Data Exchange (ETDEWEB)

Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J. [Korea Atomic Energy Research Inst., Yusong, Taejon (Korea, Republic of)

1997-07-01

KAERI has a comprehensive product development program of CANFLEX and DUPIC fuels to introduce them into CANDU reactors in Korea and a clear vision of how the product will evolve over the next 10 years. CANDU reactors are not the majority of nuclear power plants in Korea, but they produce significant electricity to contribute Korea's economic growth as well as to satisfy the need for energy. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology The CANFLEX and DUPIC R and D programs are conducted currently under the second stage of Korea's Nuclear Energy R and D Project as a national mid- and long-term program over the next 10 years from 1997 to 2006. The specific activities of the programs have taken account of the domestic and international environment concerning on non-proliferation in the Peninsula of Korea. As the first of the development products in the short-term, the CANFLEX-NU fuel will be completely developed jointly by KAERI/AECL and will be useful for the older CANDU-6 Wolsong unit 1. As the second product, the CANFLEX-0.9 % equivalent SEU fuel is expected to be completely developed within the next decade. It will be used in CANDU-6 reactors in Korea immediately after the development, if the existing RU in the world is price competitive with natural uranium. The DUPIC R and D program, as a long term program, is expected to demonstrate the possibility of use of used PWR fuel in CANDU reactors in Korea during the next 10 years. The pilot scale fabrication facility would be completed around 2010. (author)

United States Department of Energy commercial reactor spent fuel programs being conducted at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Piscitella, R.R.; Rasmussen, T.L.; Uhl, D.L.

1987-01-01

The Idaho National Engineering Laboratory participation in OCRWM programs includes the Spent Fuel Storage Cask Testing Program, Dry Rod Consolidation Technology Program, Prototypical Consolidation Demonstration Program, the Nuclear Fuel Services Project, and the Cask Systems Acquisition Program. The DOE has entered into a cooperative agreement with Virginia Power and the Electric Power Research Institute to demonstrate storage of commercial spent fuel in steel storage casks. The Program conducted heat transfer and shielding tests with three storage casks with intact spent fuel assemblies and two casks with consolidated spent fuel rods, one of which was previously tested with intact fuel, and provides test information in support of Virginia Power's at-reactor dry storage licensing effort. 3 figs., 1 tab

Establishing and sustaining a technical program to achieve zero fuel failures

International Nuclear Information System (INIS)

Deshon, J.; Whiteside, K.; Burnham, R.

2015-01-01

In 2006, Chief Nuclear Officers (CNOs) of electric utilities operating 103 commercial reactors in the United States (U.S.) formally endorsed an initiative to achieve failure-free fuel performance by the end of 2010. This became known as the Zero-by- Ten Fuel Failure Initiative. The endorsement manifested during a meeting at the Institute of Nuclear Power Operations (INPO) while nuclear fuel performance trends were being reviewed. Declining fuel performance generated backing for the Initiative. This paper provides a brief review of some of the drivers that caused those trends, the Initiative elements designed to address them, and a review of the technical program that helped achieve, and thus far sustain, improved fuel performance. (author) Key Words: Nuclear Fuel, Fuel, Fuel Failures, Fuel Failure Initiative, Zero by Ten Initiative

Improved self-exclusion program: preliminary results.

Science.gov (United States)

Tremblay, Nicole; Boutin, Claude; Ladouceur, Robert

2008-12-01

The gambling industry has offered self-exclusion programs for quite a long time. Such measures are designed to limit access to gaming opportunities and provide problem gamblers with the help they need to cease or limit their gambling behaviour. However, few studies have empirically evaluated these programs. This study has three objectives: (1) to observe the participation in an improved self-exclusion program that includes an initial voluntary evaluation, phone support, and a mandatory meeting, (2) to evaluate satisfaction and usefulness of this service as perceived by self-excluders, (3) to measure the preliminary impact of this improved program. One hundred sixteen self-excluders completed a questionnaire about their satisfaction and their perception of the usefulness during the mandatory meeting. Among those participants, 39 attended an initial meeting. Comparisons between data collected at the initial meeting and data taken at the final meeting were made for those 39 participants. Data showed that gamblers chose the improved self-exclusion program 75% of the time; 25% preferred to sign a regular self-exclusion contract. Among those who chose the improved service, 40% wanted an initial voluntary evaluation and 37% of these individuals actually attended that meeting. Seventy percent of gamblers came to the mandatory meeting, which was a required condition to end their self-exclusion. The majority of participants were satisfied with the improved self-exclusion service and perceived it as useful. Major improvements were observed between the final and the initial evaluation on time and money spent, consequences of gambling, DSM-IV score, and psychological distress. The applicability of an improved self-exclusion program is discussed and, as shown in our study, the inclusion of a final mandatory meeting might not be so repulsive for self-excluders. Future research directives are also proposed.

Statement on the Consolidated Fuel Reprocessing Program

International Nuclear Information System (INIS)

Trauger, D.B.

1984-01-01

Oak Ridge National Laboratory has chosen the following objectives for future reprocessing plant design: reduced radiation exposure to workers; minimal environmental impact; improved plant operation and maintenance; improved accountability; no plutonium diversion; and reduced overall capital and operating cost. These objectives lead to a plant with totally remote operation. The Breeder Reactor Engineering Test (BRET) has been designed to perform a key role in demonstrating advanced reprocessing technology. It has been scheduled to be available to reprocess spent fuel from the Fast Flux Test Facility. The principal features of the Consolidated Fuel Reprocessing Program and of the BRET facility are appropriate for all reactor types

Guide to the Canadian nuclear fuel waste management program. 2.ed

International Nuclear Information System (INIS)

Rosinger, E.L.J.; Lyon, R.B.; Gillespie, P.; Tamm, J.

1983-02-01

This document describes the administrative structure and major research and development components of the Canadian Nuclear Fuel Waste Management Program. It outlines the participating organizations, summarizes the program statistics, and describes the international cooperation and external review aspects of the program

Experimental advances and preliminary mathematical modeling of the Swiss-roll mixed-reactant direct borohydride fuel cell

Science.gov (United States)

Aziznia, Amin; Oloman, Colin W.; Gyenge, Előd L.

2014-11-01

The Swiss-roll single-cell mixed reactant (SR-MRFC) borohydride - oxygen fuel cell equipped with Pt/carbon cloth 3D anode and either MnO2 or Ag gas-diffusion cathodes is investigated by a combination of experimental studies and preliminary mathematical modeling of the polarization curve. We investigate the effects of four variables: cathode side metallic mesh fluid distributor, separator type (Nafion 112® vs. Viledon®), cathode catalyst (MnO2 vs. Ag), and the hydrophilic pore volume fraction of the gas-diffusion cathode. Using a two-phase feed of alkaline borohydride solution (1 M NaBH4 - 2 M NaOH) and O2 gas in an SR-MRFC equipped with Pt/C 3D anode, MnO2 gas diffusion cathode, Viledon® porous diaphragm, expanded mesh cathode-side fluid distributor, the maximum superficial power density is 2230 W m-2 at 323 K and 105 kPa(abs). The latter superficial power density is almost 3.5 times higher than our previously reported superficial power density for the same catalyst combinations. Furthermore, with a Pt anode and Ag cathode catalyst combination, a superficial power density of 2500 W m-2 is achieved with superior performance durability compared to the MnO2 cathode. The fuel cell results are substantiated by impedance spectroscopy analysis and preliminary mathematical model predictions based on mixed potential theory.

FUEL CONSUMPTION EFFECT OF COMMERCIAL TURBOFANS ON GLOBAL WARMING

Energy Technology Data Exchange (ETDEWEB)

Onder Turan; T. Hikmet Karakoc [School of Civil Aviation, Anadolu University, Eskisehir (Turkey)

2008-09-30

The main objective pursued in this study is to parametrically investigate the fuel consumption effect of commercial turbofans on global warming. In this regard, Of the important parameters, specific fuel consumption of a commercial turbofans is taken into consideration. In order to minimize the effect of fuel consumption on global warming, the values of engine design parameters are optimized for maintaining minimum specific fuel consumption of high bypass turbofan engine under different flight conditions and design criteria. The backbones of optimization approach consisted of elitism-based genetic algorithm coupled with real parametric cycle analysis of a turbofan engine. For solving optimization problem a new software program is developed in MATLAB, while objective function is determined for minimizing the specific fuel consumption by considering the following parameters such as the fan pressure ratio ({pi}{sub f}), bypass ratio ({alpha}) and the fuel heating value [h{sub PR}-(kJ/kg)]. Accordingly, it may be concluded that the software program developed can successfully solve optimization problems at 1.2{le}{pi}{sub f}{le}2, 2{le}{alpha}{le}10 and 23000{le}h{sub PR}{le}120000 with aircraft flight Mach number {le}0.8. Fuel types used in preliminary engine cycle analysis were JP-4, JP-5, JP-8 and hydrogen in this paper.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Characterization program management plan for Hanford K Basin Spent Nuclear Fuel

International Nuclear Information System (INIS)

Lawrence, L.A.

1995-01-01

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

Status of Westinghouse coal-fueled combustion turbine programs

International Nuclear Information System (INIS)

Scalzo, A.J.; Amos, D.J.; Bannister, R.L.; Garland, R.V.

1992-01-01

Developing clean, efficient, cost effective coal utilization technologies for future power generation is an essential part of our National Energy Strategy. Westinghouse is actively developing power plants utilizing advanced gasification, atmospheric fluidized beds (AFB), pressurized fluidized beds (PFB), and direct firing technology through programs sponsored by the U.S. Dept. of Energy (DOE). The DOE Office of Fossil Energy is sponsoring the Direct Coal-Fired Turbine program. This paper presents the status of current and potential Westinghouse Power Generation Business Unit advanced coal-fueled power generation programs as well as commercial plans

Preliminary Report: Bases for Containment Analysis for Transportation of Aluminum-Based Spent Nuclear Fuel

International Nuclear Information System (INIS)

Vinson, D.W.

1998-01-01

Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to SRS under the site FRR/DRR Receipts Program. Shipment of the FRR/DRR assemblies required that the cask with loaded fuel be certified by the US Nuclear Regulatory Commission (for US-owned casks) or the US Department of Transportation (for foreign-owned casks) to comply with the requirements in 10CFR71

Nuclear fuel cycle system analysis

International Nuclear Information System (INIS)

Ko, W. I.; Kwon, E. H.; Kim, S. G.; Park, B. H.; Song, K. C.; Song, D. Y.; Lee, H. H.; Chang, H. L.; Jeong, C. J.

2012-04-01

The nuclear fuel cycle system analysis method has been designed and established for an integrated nuclear fuel cycle system assessment by analyzing various methodologies. The economics, PR(Proliferation Resistance) and environmental impact evaluation of the fuel cycle system were performed using improved DB, and finally the best fuel cycle option which is applicable in Korea was derived. In addition, this research is helped to increase the national credibility and transparency for PR with developing and fulfilling PR enhancement program. The detailed contents of the work are as follows: 1)Establish and improve the DB for nuclear fuel cycle system analysis 2)Development of the analysis model for nuclear fuel cycle 3)Preliminary study for nuclear fuel cycle analysis 4)Development of overall evaluation model of nuclear fuel cycle system 5)Overall evaluation of nuclear fuel cycle system 6)Evaluate the PR for nuclear fuel cycle system and derive the enhancement method 7)Derive and fulfill of nuclear transparency enhancement method The optimum fuel cycle option which is economical and applicable to domestic situation was derived in this research. It would be a basis for establishment of the long-term strategy for nuclear fuel cycle. This work contributes for guaranteeing the technical, economical validity of the optimal fuel cycle option. Deriving and fulfillment of the method for enhancing nuclear transparency will also contribute to renewing the ROK-U.S Atomic Energy Agreement in 2014

The Canadian long-term experimental used fuel storage program

International Nuclear Information System (INIS)

Wasywich, K.M.; Taylor, P.

1993-01-01

The Canadian experimental fuel storage program consists of four components: (1) storage of used CANDU (CANadian Deuterium Uranium, registered trademark of AECL) fuel under water, with periodic examination; (2) storage of used CANDU fuel in dry air at seasonally varying temperatures, and in both dry and moisture-saturated air at 150 C, also with periodic examination; (3) underlying research on the oxidation of unused and used UO 2 in dry and moist air at temperatures up to 300 C; and (4) modeling of UO 2 oxidation in dry air. The primary objective of the fuel-storage experiments is to investigate the stability of used CANDU fuel during long-term storage. Burnup of the fuel in these experiments ranges from ∼43 to 582 MW h/kg U, while the outer-element linear power ratings range from 22 to 79 kW/m. The storage behavior of intact and intentionally defected fuel, and fuel that defected in-reactor, is being investigated in the above experiments. Since differences in UO 2 oxidation behavior were observed between dry-air, moisture-saturated air and wet storage of intentionally defected used CANDU fuel, underlying research was initiated on oxidation of unused and used fuel to develop a better understanding of the different mechanisms. Modeling of UO 2 oxidation based on the results of the dry-storage experiments is also under way

PLUS 7TM advanced fuel assembly development program for KSNPs and APR1400

International Nuclear Information System (INIS)

Kim, Kyutae; Stucker, David L.

2002-01-01

KNFC and Westinghouse have recently completed the development of the PLUS 7 TM advanced 16 X 16 fuel assembly for the Korean Standard Nuclear Plants (KSNPs) and the Advanced Power Reactor 1400 (APR 1400). This fuel design utilized the proven advanced design features including mixing vane spacer grids to increase critical heat flux performance, ZIRLO TM advanced materials to enable high-duty, high burnup fuel management and an optimized fuel rod diameter which improves fuel cycle cost while resulting in significant standardization of Korean fuel manufacture. PLUS 7 TM , also includes a patented spacer grid design with conformal fuel rod support designed to provide superior fuel rod wear/fretting resistance while minimizing pressure drop. This paper will present an overview of the PLUS 7 TM fuel assembly development process including a summary of the three-year design and testing program from a mechanical, neutronic, and thermal/hydraulic perspective. The PLUS 7 TM fuel for the KSNPs and the APR1400 reactors results in multi-million dollar per cycle savings in imported enriched uranium product for the Korean nuclear power program with technology specifically developed for Korea by experienced Korean engineers

Preliminary Results of the Louisiana Sex Offender Treatment Program

OpenAIRE

Lee A. Underwood; Frances L.L. Dailey; Carrie Merino; Yolanda Crump

2015-01-01

The purpose of this study was to offer preliminary support for the Louisiana Sex Offender Treatment Program (LSOTP) in addressing the needs of juvenile sex offenders. Research objectives were (1) to offer statistical evidence for reductions in anxiety, depression, cognitive distortion and negative attitudes towards women comparing a group of 21 adolescents, 12 of whom received services as usual and nine of whom participated in the LSOTP. A controlled experimental evaluation design was utilize...

Zero-gravity cloud physics laboratory: Experiment program definition and preliminary laboratory concept studies

Science.gov (United States)

Eaton, L. R.; Greco, E. V.

1973-01-01

The experiment program definition and preliminary laboratory concept studies on the zero G cloud physics laboratory are reported. This program involves the definition and development of an atmospheric cloud physics laboratory and the selection and delineations of a set of candidate experiments that must utilize the unique environment of zero gravity or near zero gravity.

Extending the foreign spent fuel acceptance program: Policy and implementation issues

International Nuclear Information System (INIS)

Lyman, Edwin S.

2005-01-01

The May 2006 expiration date of the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program is fast approaching. In April 2004, Energy Secretary Spencer Abraham instructed the Energy Department to 'initiate actions necessary to extend .... the fuel acceptance deadline'. However, extending the deadline may not be a simple task. The limits on the original program resulted from a delicate negotiation among many stakeholders. Any proposal to increase the duration and scope of the program will have to be considered in the context of DOE's failure since 1996 to develop viable treatment, packaging and long-term disposal options for FRR SNF. It is also unclear whether accepting additional low-enriched uranium FRR SNF can be justified on security grounds. This paper will propose criteria for acceptance of spent fuel under an extension that are intended to minimize controversy and ensure consistency with a threat-based prioritization of homeland security expenditures. (author)

Impacts of the Weatherization Assistance Program in Fuel-Oil Heated Houses

Energy Technology Data Exchange (ETDEWEB)

Levins, W.P.

1994-01-01

In 1990, the U.S. Department of Energy (DOE) initiated a national evaluation of its low-income Weatherization Assistance Program. This report, which is one of five parts of that evaluation, evaluates the energy savings and cost-effectiveness of the Program as it had been applied to single-family houses heated primarily by fuel-oil. The study was based upon a representative sample (41 local weatherization agencies, 222 weatherized and 115 control houses) from the nine northeastern states during 1991 and 1992 program years. Dwelling-specific and agency-level data on measures installed, costs, and service delivery procedures were collected from the sampled agencies. Space-heating fuel-oil consumption, indoor temperature, and outdoor temperature were monitored at each house. Dwelling characteristics, air-leakage measurements, space-heating system steady-state efficiency measurements, safety inspections, and occupant questionnaires were also collected or performed at each monitored house. We estimate that the Program weatherized a total of 23,400 single-family fuel-oil heated houses in the nine northeastern states during program years 1991 and 1992. Annual fuel-oil savings were calculated using regression techniques to normalize the savings to standard weather conditions. For the northeast region, annual net fuel-oil savings averaged 160 gallons per house, or 17.7% of pre-weatherization consumption. Although indoor temperatures changed in individual houses following weatherization, there was no average change and no significant difference as compared to the control houses; thus, there was no overall indoor temperature takeback effect influencing fuel-oil savings. The weatherization work was performed cost effectively in these houses from the Program perspective, which included both installation costs and overhead and management costs but did not include non-energy benefits (such as employment and environmental). Total average costs were $1819 per house ($1192 for

HEXBU-3D, a three-dimensional PWR-simulator program for hexagonal fuel assemblies

International Nuclear Information System (INIS)

Karvinen, E.

1981-06-01

HEXBU-3D is a three-dimensional nodal simulator program for PWR reactors. It is designed for a reactor core that consists of hexagonal fuel assemblies and of big follower-type control assemblies. The program solves two-group diffusion equations in homogenized fuel assembly geometry by a sophisticated nodal method. The treatment of feedback effects from xenon-poisoning, fuel temperature, moderator temperature and density and soluble boron concentration are included in the program. The nodal equations are solved by a fast two-level iteration technique and the eigenvalue can be either the effective multiplication factor or the boron concentration of the moderator. Burnup calculations are performed by tabulated sets of burnup-dependent cross sections evaluated by a cell burnup program. HEXBY-3D has been originally programmed in FORTRAN V for the UNIVAC 1108 computer, but there is also another version which is operable on the CDC CYBER 170 computer. (author)

Optimization of axial enrichment and gadolinia distributions for BWR fuel under control rod programming, (2)

International Nuclear Information System (INIS)

Hida, Kazuki; Yoshioka, Ritsuo

1992-01-01

A method has been developed for optimizing the axial enrichment and gadolinia distributions for the reload BWR fuel under control rod programming. The problem was to minimize the enrichment requirement subject to the criticality and axial power peaking constraints. The optimization technique was based on the successive linear programming method, each linear programming problem being solved by a goal programming algorithm. A rapid and practically accurate core neutronics model, named the modified one-dimensional core model, was developed to describe the batch-averaged burnup behavior of the reload fuel. A core burnup simulation algorithm, employing a burnup-power-void iteration, was also developed to calculate the rigorous equilibrium cycle performance. This method was applied to the optimization of axial two- and 24-region fuels for demonstrative purposes. The optimal solutions for both fuels have proved the optimality of what is called burnup shape optimization spectral shift. For the two-region fuel with a practical power peaking of 1.4, the enrichment distribution was nearly uniform, because a bottom-peaked burnup shape flattens the axial power shape. Optimization of the 24-region fuel has shown a potential improvement in BWR fuel cycle economics, which will guide future advancement in BWR fuel designs. (author)

A Preliminary Study on the Reuse of the Recovered Uranium from the Spent CANDU Fuel Using Pyroprocessing

International Nuclear Information System (INIS)

Park, C. J.; Na, S. H.; Yang, J. H.; Kang, K. H.; Lee, J. W.

2009-01-01

During the pyroprocessing, most of the uranium is gathered in metallic form around a solid cathode during an electro-refining process, which is composed of about 94 weight percent of the spent fuel. In the previous study, a feasibility study has been done to reuse the recovered uranium for the CANDU reactor fuel following the traditional DUPIC (direct use of spent pressurized water reactor fuel into CANDU reactor) fuel fabrication process. However, the weight percent of U-235 in the recovered uranium is about 1 wt% and it is sufficiently re-utilized in a heavy water reactor which uses a natural uranium fuel. The reuse of recovered uranium will bring not only a huge economic profit and saving of uranium resources but also an alleviation of the burden on the management and the disposal of the spent fuel. The research on recycling of recovered uranium was carried out 10 years ago and most of the recovered uranium was assumed to be imported from abroad at that time. The preliminary results showed there is the sufficient possibility to recycle recovered uranium in terms of a reactor's characteristics as well as the fuel performance. However, the spent CANDU fuel is another issue in the storage and disposal problem. At present, most countries are considering that the spent CANDU fuel is disposed directly due to the low enrichment (∼0.5 wt%) of the discharge fissile content and lots of fission products. If mixing the spent CANDU fuel and the spent PWR fuel, the estimated uranium fissile enrichment will be about 0.6 wt% ∼ 1.0 wt% depending on the mixing ratio, which is sufficiently reusable in a CANDU reactor. Therefore, this paper deals with a feasibility study on the recovered uranium of the mixed spent fuel from the pyroprocessing. With the various mixing ratios between the PWR spent fuel and the CANDU spent fuel, a reactor characteristics including the safety parameters of the CANDU reactor was evaluated

Correlations between fuel pins irradiated in fast and thermal fluxes using the frump fuel pin modelling program

International Nuclear Information System (INIS)

Hayns, M.R.; Adam, J.

1975-08-01

There is no experimental facilities in which a fuel pin can be irradiated in a fast environment under well defined conditions of over power or flow run down. Consequently most of the infor mation which is being accumulated on the behaviour of fuel pins under severe conditions is obtained from either capsule or loop rigs in thermal reactors. It is the purpose of this paper to highlight the differences between the behaviour of fuel pins irradiated in a thermal flux and a fast flux. A typical set of conditions is taken from an overpower experiment in a thermal flux and the behaviour of the system is analysed using the fuel modelling program FRUMP. A second numerical experiment is then performed in which the same conditions prevail, except that a fast flux is assumed, the criterion for comparison being that the total power input to the system is the same in both cases. From the many possible correlations which result from such an exercise the fuel tempreature has been selected to highlight various important features of the two irradiations. It is demonstrated that the flux depression can cause differences in the pin behaviour, even to altering the order of events in a transient. For example fuel melting will occur at different times and at different positions in the fuel in the two cases. It is concluded that the techniques of fuel modelling, as typified in the program FRUMP can provide a very useful tool indeed for the analysis of such experiments and for guiding the establishment of the appropriate correlations for the extrapolation to the fast flux case. (author)

Preliminary topical report on comparison reactor disassembly calculations

International Nuclear Information System (INIS)

McLaughlin, T.P.

1975-11-01

Preliminary results of comparison disassembly calculations for a representative LMFBR model (2100-l voided core) and arbitrary accident conditions are described. The analytical methods employed were the computer programs: FX2-POOL, PAD, and VENUS-II. The calculated fission energy depositions are in good agreement, as are measures of the destructive potential of the excursions, kinetic energy, and work. However, in some cases the resulting fuel temperatures are substantially divergent. Differences in the fission energy deposition appear to be attributable to residual inconsistencies in specifying the comparison cases. In contrast, temperature discrepancies probably stem from basic differences in the energy partition models inherent in the codes. Although explanations of the discrepancies are being pursued, the preliminary results indicate that all three computational methods provide a consistent, global characterization of the contrived disassembly accident

Plutonium fuel program

International Nuclear Information System (INIS)

1979-09-01

A review is presented of the development of the (UPu)C sphere-pac fuel project during 1978. In particular, the problems encountered in obtaining good fuel quality in the fabrication process and their solution is discussed. The development of a fabrication pilot plant is considered, and the post-irradiation examination of fuel pins is presented. (Auth.)

Spent fuel cladding containment credit test

International Nuclear Information System (INIS)

Wilson, C.N.

1983-01-01

As an initial step in addressing the effectiveness of breached cladding as a barrier to radionuclide release from the repository during the post-containment period, preliminary scoping tests have been initiated which compare radionuclide releases from spent fuel specimens with artificially induced cladding defects of various severities. The artificially induced defects are all more severe than the typical in-reactor type breaches which are expected to be the principal type of breach entering the repository for terminal storage. These preliminary scoping tests being conducted by Westinghouse Hanford Company for the Lawrence Livermore National Laboratory Waste Package Development Program in support of the Tuff repository project at the Nevada Test Site are described. Also included in this presentation are selected initial results from these tests. 22 figures

In-core fuel management programs for nuclear power reactors

International Nuclear Information System (INIS)

1984-10-01

In response to the interest shown by Member States, the IAEA organized a co-ordinated research programme to develop and make available in the open domain a set of programs to perform in-core fuel management calculations. This report summarizes the work performed in the context of the CRP. As a result of this programme, complete in-core fuel management packages for three types of reactors, namely PWR's, BWR's and PHWR are now available from the NEA Data Bank. For some reactor types, these program packages are available with three levels of sophistication ranging from simple methods for educational purposes to more comprehensive methods that can be used for reactor design and operation. In addition some operating data have been compiled to allow code validation. (author)

Contextualizing avian mortality: A preliminary appraisal of bird and bat fatalities from wind, fossil-fuel, and nuclear electricity

International Nuclear Information System (INIS)

Sovacool, Benjamin K.

2009-01-01

This article explores the threats that wind farms pose to birds and bats before briefly surveying the recent literature on avian mortality and summarizing some of the problems with it. Based on operating performance in the United States and Europe, this study offers an approximate calculation for the number of birds killed per kWh generated for wind electricity, fossil-fuel, and nuclear power systems. The study estimates that wind farms and nuclear power stations are responsible each for between 0.3 and 0.4 fatalities per gigawatt-hour (GWh) of electricity while fossil-fueled power stations are responsible for about 5.2 fatalities per GWh. While this paper should be respected as a preliminary assessment, the estimate means that wind farms killed approximately seven thousand birds in the United States in 2006 but nuclear plants killed about 327,000 and fossil-fueled power plants 14.5 million. The paper concludes that further study is needed, but also that fossil-fueled power stations appear to pose a much greater threat to avian wildlife than wind and nuclear power technologies.

Contextualizing avian mortality: A preliminary appraisal of bird and bat fatalities from wind, fossil-fuel, and nuclear electricity

Energy Technology Data Exchange (ETDEWEB)

Sovacool, Benjamin K. [Energy Governance Program, Centre on Asia and Globalisation, Lee Kuan Yew School of Public Policy, National University of Singapore, Singapore 259772 (Singapore)], E-mail: bsovacool@nus.edu.sg

2009-06-15

This article explores the threats that wind farms pose to birds and bats before briefly surveying the recent literature on avian mortality and summarizing some of the problems with it. Based on operating performance in the United States and Europe, this study offers an approximate calculation for the number of birds killed per kWh generated for wind electricity, fossil-fuel, and nuclear power systems. The study estimates that wind farms and nuclear power stations are responsible each for between 0.3 and 0.4 fatalities per gigawatt-hour (GWh) of electricity while fossil-fueled power stations are responsible for about 5.2 fatalities per GWh. While this paper should be respected as a preliminary assessment, the estimate means that wind farms killed approximately seven thousand birds in the United States in 2006 but nuclear plants killed about 327,000 and fossil-fueled power plants 14.5 million. The paper concludes that further study is needed, but also that fossil-fueled power stations appear to pose a much greater threat to avian wildlife than wind and nuclear power technologies.

Contextualizing avian mortality. A preliminary appraisal of bird and bat fatalities from wind, fossil-fuel, and nuclear electricity

Energy Technology Data Exchange (ETDEWEB)

Sovacool, Benjamin K. [Energy Governance Program, Centre on Asia and Globalisation, Lee Kuan Yew School of Public Policy, National University of Singapore, Singapore 259772 (Singapore)

2009-06-15

This article explores the threats that wind farms pose to birds and bats before briefly surveying the recent literature on avian mortality and summarizing some of the problems with it. Based on operating performance in the United States and Europe, this study offers an approximate calculation for the number of birds killed per kWh generated for wind electricity, fossil-fuel, and nuclear power systems. The study estimates that wind farms and nuclear power stations are responsible each for between 0.3 and 0.4 fatalities per gigawatt-hour (GWh) of electricity while fossil-fueled power stations are responsible for about 5.2 fatalities per GWh. While this paper should be respected as a preliminary assessment, the estimate means that wind farms killed approximately seven thousand birds in the United States in 2006 but nuclear plants killed about 327,000 and fossil-fueled power plants 14.5 million. The paper concludes that further study is needed, but also that fossil-fueled power stations appear to pose a much greater threat to avian wildlife than wind and nuclear power technologies. (author)

Spent fuel sabotage aerosol test program :FY 2005-06 testing and aerosol data summary

International Nuclear Information System (INIS)

Gregson, Michael Warren; Brockmann, John E.; Nolte, O.; Loiseau, O.; Koch, W.; Molecke, Martin Alan; Autrusson, Bruno; Pretzsch, Gunter Guido; Billone, M. C.; Lucero, Daniel A.; Burtseva, T.; Brucher, W; Steyskal, Michele D.

2006-01-01

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 has been underway for several years. This program provides source-term data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This document focuses on an updated description of the test program and test components for all work and plans made, or revised, primarily during FY 2005 and about the first two-thirds of FY 2006. It also serves as a program status report as of the end of May 2006. We provide details on the significant findings on aerosol results and observations from the recently completed Phase 2 surrogate material tests using cerium oxide ceramic pellets in test rodlets plus non-radioactive fission product dopants. Results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; status on determination of the spent fuel ratio, SFR (the ratio of respirable particles from real spent fuel/respirables from surrogate spent fuel, measured under closely matched test conditions, in a contained test chamber); and, measurements of enhanced volatile fission product species sorption onto respirable particles. We discuss progress and results for the first three, recently performed Phase 3 tests using depleted uranium oxide, DUO 2 , test rodlets. We will also review the status of preparations and the final Phase 4 tests in this program, using short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. These data plus testing results and design are tailored to support and guide, follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments

Preliminary design study of pebble bed reactor HTR-PM base using once-through-then-out fuel recirculation

International Nuclear Information System (INIS)

Topan Setiadipura; Jupiter S Pane; Zuhair

2016-01-01

Pebble Bed Reactor (PBR) is one of the advanced reactor type implementing strong passive safety feature. In this type of design has the potential to do a cogeneration useful for the treatment of various minerals in various islands in Indonesia. The operation of the PBR can be simplified by implementing once-through-then-out (OTTO) fuel recirculation scheme in which pebble fuel only pass the core once time. The purpose of this research is to understand quantitative influence of the changing of fuel element recirculation on the PBR core performance and to find preliminary optimization design of PBR type reactor with OTTO recirculation scheme. PEBBED software was used to find PBR equilibrium core. The calculation result gives quantitative data on the impact of implementing a different fuel recirculation, especially using OTTO scheme. Furthermore, an early optimized PBR design based on HTR-PM using OTTO scheme was obtained where the power must be downgraded into 115 MWt in order to preserve the safety feature. The simplicity of the reactor operation and the reduction of reactor component with OTTO scheme still make this early optimized design an interesting alternative design, despite its power reduction from the reference design. (author)

Building on success. The foreign research reactor spent nuclear fuel acceptance program

International Nuclear Information System (INIS)

Huizenga, David G.; Mustin, Tracy P.; Saris, Elizabeth C.; Massey, Charles D.

1998-01-01

The second year of implementation of the research reactor spent nuclear fuel acceptance program was marked by significant challenges and achievements. In July 1998, the Department of Energy completed by significant challenges and achievements. In July 1998, the Department of Energy completed its first shipment of spent fuel from Asia via the Concord Naval Weapons Station in California to the Idaho National Engineering and Environmental (INEEL). This shipment, which consisted of three casks of spent nuclear fuel from two research reactors in the Republic of Korea, presented significant technical, legal, and political challenges in the United States and abroad. Lessons learned will be used in the planning and execution of our next significant milestone, a shipment of TRIGA spent fuel from research reactors in Europe to INEEL, scheduled for the summer of 1999. This shipment will include transit across the United States for over 2,000 miles. Other challenges and advances include: clarification of the fee policy to address changes in the economic status of countries during the life of the program; resolution of issues associated with cask certification and the specific types and conditions of spent fuel proposed for transport; revisions to standard contract language in order to more clearly address unique shipping situations; and priorization and scheduling of shipments to most effectively implement the program. As of this meeting, eight shipments, consisting of nearly 2,000 spent fuel assemblies from fifteen countries, have been successfully completed. With the continued cooperation of the international research reactor community, we are committed to building on this success in the remaining years of the program. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-06-15

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

Estimation of Shielding Thickness for a Prototype Department of Energy National Spent Nuclear Fuel Program Transport Cask

Energy Technology Data Exchange (ETDEWEB)

SANCHEZ,LAWRENCE C.; MCCONNELL,PAUL E.

2000-07-01

Preliminary shielding calculations were performed for a prototype National Spent Nuclear Fuel Program (NSNFP) transport cask. This analysis is intended for use in the selection of cask shield material type and preliminary estimate of shielding thickness. The radiation source term was modeled as cobalt-60 with radiation exposure strength of 100,000 R/hr. Cobalt-60 was chosen as a surrogate source because it simultaneous emits two high-energy gammas, 1.17 MeV and 1.33 MeV. This gamma spectrum is considered to be large enough that it will upper bound the spectra of all the various spent nuclear fuels types currently expected to be shipped within the prototype cask. Point-kernel shielding calculations were performed for a wide range of shielding thickness of lead and depleted uranium material. The computational results were compared to three shielding limits: 200 mrem/hr dose rate limit at the cask surface, 50 mR/hr exposure rate limit at one meter from the cask surface, and 10 mrem/hr limit dose rate at two meters from the cask surface. The results obtained in this study indicated that a shielding thickness of 13 cm is required for depleted uranium and 21 cm for lead in order to satisfy all three shielding requirements without taking credit for stainless steel liners. The system analysis also indicated that required shielding thicknesses are strongly dependent upon the gamma energy spectrum from the radiation source term. This later finding means that shielding material thickness, and hence cask weight, can be significantly reduced if the radiation source term can be shown to have a softer, lower energy, gamma energy spectrum than that due to cobalt-60.

Emergency fuels utilization guidebook. Alternative Fuels Utilization Program

Energy Technology Data Exchange (ETDEWEB)

1980-08-01

The basic concept of an emergency fuel is to safely and effectively use blends of specification fuels and hydrocarbon liquids which are free in the sense that they have been commandeered or volunteered from lower priority uses to provide critical transportation services for short-duration emergencies on the order of weeks, or perhaps months. A wide variety of liquid hydrocarbons not normally used as fuels for internal combustion engines have been categorized generically, including limited information on physical characteristics and chemical composition which might prove useful and instructive to fleet operators. Fuels covered are: gasoline and diesel fuel; alcohols; solvents; jet fuels; kerosene; heating oils; residual fuels; crude oils; vegetable oils; gaseous fuels.

Solid State Energy Conversion Alliance (SECA) Solid Oxide Fuel Cell Program

Energy Technology Data Exchange (ETDEWEB)

Nguyen Minh

2006-07-31

This report summarizes the work performed for Phase I (October 2001 - August 2006) under Cooperative Agreement DE-FC26-01NT41245 for the U. S. Department of Energy, National Energy Technology Laboratory (DOE/NETL) entitled 'Solid State Energy Conversion Alliance (SECA) Solid Oxide Fuel Cell Program'. The program focuses on the development of a low-cost, high-performance 3-to-10-kW solid oxide fuel cell (SOFC) system suitable for a broad spectrum of power-generation applications. During Phase I of the program significant progress has been made in the area of SOFC technology. A high-efficiency low-cost system was designed and supporting technology developed such as fuel processing, controls, thermal management, and power electronics. Phase I culminated in the successful demonstration of a prototype system that achieved a peak efficiency of 41%, a high-volume cost of $724/kW, a peak power of 5.4 kW, and a degradation rate of 1.8% per 500 hours. . An improved prototype system was designed, assembled, and delivered to DOE/NETL at the end of the program. This prototype achieved an extraordinary peak efficiency of 49.6%.

2013 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

Energy Technology Data Exchange (ETDEWEB)

none,

2013-12-01

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

2014 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

Energy Technology Data Exchange (ETDEWEB)

none,

2014-11-01

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

2011 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

Energy Technology Data Exchange (ETDEWEB)

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

2011-11-01

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

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

International Nuclear Information System (INIS)

Travelli, A.

1983-01-01

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

Fabrication and testing of uranium nitride fuel for space power reactors

Science.gov (United States)

Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.

1988-02-01

Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.

Vinca nuclear decommissioning program

International Nuclear Information System (INIS)

Pesic, M.; Subotic, K.; Sotic, O.; Plecas, I.; Ljubenov, V.; Peric, A.

2002-01-01

In this paper a preliminary program for the nuclear decommissioning in The Vinca Institute of Nuclear Sciences is presented. Proposed Projects and Activities, planned to be done in the next 10 years within the frames of the Program, should improve nuclear and radiation safety and should solve the main problems that have arisen in the previous period. Project of removal of irradiated spent nuclear fuel from the RA reactor, as a first step in all possible decommissioning strategies and the main activity in the first two-three years of the Program realization, is considered in more details. (author)

Preliminary project concerning the straw-fueled combined power-heat plant to be constructed at Glamsbjerg

International Nuclear Information System (INIS)

Gabriel, S.; Koch, T.

1994-06-01

Power and heat generation based on biomass gasification is of great importance due to its beneficial environmental effects and good economy. This report concerns a preliminary project on feasibility and problems of implementing a dual-purpose power plant, supplying both power and district heating to several schools, swimming pools and other public facilities at Glamsbjerg (Funen). The plant is to be based on thermal gasification (pyrolysis) of straw and use of the gas in a diesel engine. The diesels operate the power generator, and their waste heat should be utilized in the local district heating network. In order to establish a stable and flexible straw supply to the plant an evaluation of resources in the area has been carried out. Apart from straw-derived fuel the plant is planned to use natural gas for start and maintenance of the process. The prices of the combined plant and of the fuel processing are estimated in the report. (EG)

Conversion of research and test reactors to low enriched uranium fuel: technical overview and program status

International Nuclear Information System (INIS)

Roglans-Ribas, J.

2008-01-01

Many of the nuclear research and test reactors worldwide operate with high enriched uranium fuel. In response to worries over the potential use of HEU from research reactors in nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel by converting research reactors to low enriched uranium (LEU) fuel. The Reactor Conversion program is currently under the DOE's National Nuclear Security Administration's Global Threat Reduction Initiative (GTRI). 55 of the 129 reactors included in the scope have been already converted to LEU fuel or have shutdown prior to conversion. The major technical activities of the Conversion Program include: (1) the development of advanced LEU fuels; (2) conversion analysis and conversion support; and (3) technology development for the production of Molybdenum-99 (Mo 99 ) with LEU targets. The paper provides an overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives. Nuclear research and test reactors worldwide have been in operation for over 60 years. Many of these facilities operate with high enriched uranium fuel. In response to increased worries over the potential use of HEU from research reactors in the manufacturing of nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. The reactor conversion program was initially focused on U.S.-supplied reactors, but in the early 1990s it expanded and began to collaborate with Russian institutes with the objective of converting Russian supplied reactors to the use of LEU fuel.

Quality assurance program plan fuel supply shutdown project

International Nuclear Information System (INIS)

Metcalf, I.L.

1998-01-01

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

Spent Fuel Handling and Packaging Program: a survey of hot cell facilities

International Nuclear Information System (INIS)

Menon, M.N.

1978-07-01

Hot cell facilities in the United States were surveyed to determine their capabilities for conducting integral fuel assembly and individual fuel rod examinations that are required in support of the Spent Fuel Handling and Packaging Program. The ability to receive, handle, disassemble and reconstitute full-length light water reactor spent fuel assemblies, and the ability to conduct nondestructive and destructive examinations on full-length fuel rods were of particular interest. Three DOE-supported facilities and three commercial facilities were included in the survey. This report provides a summary of the findings

Spent Fuel and Waste Management Technology Development Program. Annual progress report

Energy Technology Data Exchange (ETDEWEB)

Bryant, J.W.

1994-01-01

This report provides information on the progress of activities during fiscal year 1993 in the Spent Fuel and Waste Management Technology Development Program (SF&WMTDP) at the Idaho Chemical Processing Plant (ICPP). As a new program, efforts are just getting underway toward addressing major issues related to the fuel and waste stored at the ICPP. The SF&WMTDP has the following principal objectives: Investigate direct dispositioning of spent fuel, striving for one acceptable waste form; determine the best treatment process(es) for liquid and calcine wastes to minimize the volume of high level radioactive waste (HLW) and low level waste (LLW); demonstrate the integrated operability and maintainability of selected treatment and immobilization processes; and assure that implementation of the selected waste treatment process is environmentally acceptable, ensures public and worker safety, and is economically feasible.

New high density MTR fuel. The CEA-CERCA-COGEMA development program

International Nuclear Information System (INIS)

Languille, A.; Durand, J.P.; Gay, A.

1999-01-01

The development of a new generation of LEU, high in density and with reprocessing capacities MTR fuel, is a key issue to provide reactor operators with a smooth operation which is necessary for a long term development of Nuclear Energy. In the RRFM'98 meeting, a joint contribution of CEA, CERCA and COGEMA presented a technical classification of the potential candidates uranium alloys. In this paper this MTR working group presents the development program of a new high density fuel. This program is composed of three main steps: Basic Data analysis and collection, Plate Tests (Irradiation and Post Irradiation Examinations) and Lead Test Assemblies (Irradiation and Post Irradiation Examinations). The goal to be reached is to make this new fuel available before the end of the present US return policy. (author)

A Deformation Model of TRU Metal Dispersion Fuel Rod for HYPER

International Nuclear Information System (INIS)

Lee, Byoung Oon; Hwang, Woan; Park, Won S.

2002-01-01

Deformation analysis in fuel rod design is essential to assure adequate fuel performance and integrity under irradiation conditions. An in-reactor performance computer code for a dispersion fuel rod is being developed in the conceptual design stage of blanket fuel for HYPER. In this paper, a mechanistic deformation model was developed and the model was installed into the DIMAC program. The model was based on the elasto-plasticity theory and power-law creep theory. The preliminary deformation calculation results for (TRU-Zr)-Zr dispersion fuel predicted by DIMAC were compared with those of silicide dispersion fuel predicted by DIFAIR. It appeared that the deformation levels for (TRU-Zr)-Zr dispersion fuel were relatively higher than those of silicide fuel. Some experimental tests including in-pile and out-pile experiments are needed for verifying the predictive capability of the DIMAC code. An in-reactor performance analysis computer code for blanket fuel is being developed at the conceptual design stage of blanket fuel for HYPER. In this paper, a mechanistic deformation model was developed and the model was installed into the DIMAC program. The model was based on the elasto-plasticity theory and power-law creep theory. The preliminary deformation calculation results for (TRUZr)- Zr dispersion fuel predicted by DIMAC were compared with those of silicide dispersion fuel predicted by DIFAIR. It appears that the deformation by swelling within fuel meat is very large for both fuels, and the major deformation mechanism at cladding is creep. The swelling strain is almost constant within the fuel meat, and is assumed to be zero in the cladding made of HT9. It is estimated that the deformation levels for (TRU-Zr)-Zr dispersion fuel were relatively higher than those of silicide fuel, and the dispersion fuel performance may be limited by swelling. But the predicted volume change of the (TRU-Zr)-Zr dispersion fuel models is about 6.1% at 30 at.% burnup. The value of cladding

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

International Nuclear Information System (INIS)

Martin, M.M.

1993-01-01

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

Optimization of MOX fuel cycles in pebble bed HTGR

International Nuclear Information System (INIS)

Wei Jinfeng; Li Fu; Sun Yuliang

2013-01-01

Compared with light water reactor (LWR), the pebble bed high temperature gas-cooled reactor (HTGR) is able to operate in a full mixed oxide (MOX) fuelled core without significant change to core structure design. Based on a reference design of 250 MW pebble bed HTGR, four MOX fuel cycles were designed and evaluated by VSOP program package, including the mixed Pu-U fuel pebbles and mixed loading of separate Pu-pebbles and U-pebbles. Some important physics features were investigated and compared for these four cycles, such as the effective multiplication factor of initial core, the pebble residence time, discharge burnup, and temperature coefficients. Preliminary results show that the overall performance of one case is superior to other equivalent MOX fuel cycles on condition that uranium fuel elements and plutonium fuel elements are separated as the different fuel pebbles and that the uranium fuel elements are irradiated longer in the core than the plutonium fuel elements, and the average discharge burnup of this case is also higher than others. (authors)

Management of radioactive fuel wastes: the Canadian disposal program

International Nuclear Information System (INIS)

Boulton, J.

1978-10-01

This report describes the research and development program to verify and demonstrate the concepts for the safe, permanent disposal of radioactive fuel wastes from Canadian nuclear reactors. The program is concentrating on deep underground disposal in hard-rock formations. The nature of the radioactive wastes is described, and the options for storing, processing, packaging and disposing of them are outlined. The program to verify the proposed concept, select a suitable site and to build and operate a demonstration facility is described. (author)

Preliminary Accident Analyses for Conversion of the Massachusetts Institute of Technology Reactor (MITR) from Highly Enriched to Low Enriched Uranium

Energy Technology Data Exchange (ETDEWEB)

Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, Erik H. [Argonne National Lab. (ANL), Argonne, IL (United States); Sun, Kaichao S. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Newton, Jr., Thomas H. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

2013-09-30

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. This report presents the preliminary accident analyses for MITR cores fueled with LEU monolithic U-Mo alloy fuel with 10 wt% Mo. Preliminary results demonstrate adequate performance, including thermal margin to expected safety limits, for the LEU accident scenarios analyzed.

Determination of the burn-up of TRIGA fuel elements by calculation with new TRIGLAV program

International Nuclear Information System (INIS)

Zagar, T.; Ravnik, M.

1996-01-01

The results of fuel element burn-up calculations with new TRIGLAV program are presented. TRIGLAV program uses two dimensional model. Results of calculation are compared to results calculated with program, which uses one dimensional model. The results of fuel element burn-up measurements with reactivity method are presented and compared with the calculated results. (author)

Preliminary investigation of the use of monolithic U-Mo fuel in the MIT reactor

International Nuclear Information System (INIS)

Newton, Thomas H. Jr.; Kazimi, Mujid S.; Pilat, Edward E.; Xu Zhiwen

2003-01-01

Studies have begun on the use of monolithic LEU U-Mo fuel in the MIT Nuclear Research Reactor (MITR-II) using the Monte Carlo Transport code MCNP. These studies have included model benchmarking, LEU fuel optimization, burnup evaluation, in-core facility design, and determination of safety attributes. Benchmarking studies on the initial core have shown favorable agreement between the calculated and measured reactivity worths of the six control blades. In addition, optimization studies on LEU U7Mo MITR-II fuel have shown that an arrangement of ten to twelve plates per fuel element would have initial reactivity values and thermal neutron fluxes comparable to the current HEU core. Burnup studies which have been made using the MCODE depletion program will be presented. Safety attributes such as temperature coefficients, shutdown margins, and coolant subcooled margin are under evaluation. (author)

A review of low carbon fuel policies: Principles, program status and future directions

International Nuclear Information System (INIS)

Yeh, Sonia; Witcover, Julie; Lade, Gabriel E.; Sperling, Daniel

2016-01-01

A low carbon fuel standard (LCFS) is a market-based policy that specifies declining standards for the average lifecycle fuel carbon intensity (AFCI) of transportation fuels sold in a region. This paper: (i) compares transportation fuel carbon policies in terms of their economic efficiency, fuel price impacts, greenhouse gas emission reductions, and incentives for innovation; (ii) discusses key regulatory design features of LCFS policies; and (iii) provides an update on the implementation status of LCFS policies in California, the European Union, British Columbia, and Oregon. The economics literature finds that an intensity standard implicitly taxes emissions and subsidizes output. The output subsidy results in an intensity standard being inferior to a carbon tax in a first-best world, although the inefficiency can be corrected with a properly designed consumption tax (or mitigated by a properly designed carbon tax or cap-and-trade program). In California, from 2011 to 2015 the share of alternative fuels in the regulated transportation fuels pool increased by 30%, and the reported AFCI of all alternative fuels declined 21%. LCFS credit prices have varied considerably, rising to above $100/credit in the first half of 2016. LCFS programs in other jurisdictions share many features with California's, but have distinct provisions as well. - Highlights: • LCFS is a market-based policy that sets standards for carbon intensity of fuels. • We compare efficiency, price impacts, GHG emissions, and innovation of C policies. • In California, reported carbon intensity of alternative fuels declined 21% 2011–2015. • LCFS credit prices have varied considerably, rising to above $100/credit in the first half of 2016. • Other LCFS programs share many features with CA's and have distinct provisions.

Facilitating programming comprehension for novice learners with multimedia approach: A preliminary investigation

Science.gov (United States)

Annamalai, Subashini; Salam, Sobihatun Nur Abdul

2017-10-01

This research paper presents the preliminary investigation on the use of an interactive multimedia courseware named MAFPro, to facilitate C Programming lessons for novice learners. The courseware utilizes the elements of multimedia that focus on enhancing learners' programming comprehension. Among the aspects that were examined were the students' programming comprehension and their perceived motivation of MAFPro. This study was carried out in a survey design method with the participation of 30 undergraduates who are novice learners. The data analysis indicates that the multimedia courseware, MAFPro that has been used in the C programming classroom has a significant difference on the undergraduates' programming comprehension. The students also perceived MAFPro as motivating and engaging.

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.

Preliminary assessment of the Louisiana Home Energy Rebate Offer program using IPMVP guidelines

Energy Technology Data Exchange (ETDEWEB)

Kaiser, Mark J.; Pulsipher, Allan G. [Center for Energy Studies, Louisiana State University, Energy Coast and Environment Building, Nicholson Extension Drive, Baton Rouge, LA 70803 (United States)

2010-02-15

The Louisiana Home Energy Rebate Offer (HERO) is a residential energy conservation program established in 1999 to provide rebates for qualified applicants to build new homes that are more energy efficient or improve the energy efficiency of existing homes. Energy conservation programs require careful evaluation because of the high cost to implement the measures and the expectation that they will reduce energy use. The purpose of this paper is to demonstrate that residential energy conservation measures in a hot and humid climate can be evaluated using the International Performance Measurement and Verification Protocol (IPMVP), a best practice methodology commonly used in industrial and commercial performance-based contracts, but rarely, if ever, applied to residential programs. Using a random sample of 60 HERO participants, we were able to construct statistically significant electricity consumption baseline models for 90% of households. We determined that more than half of the sample participants consumed more electricity after their efficiency improvement, with an average net household savings of 172 kWh/yr, about 1% pre-retrofit consumption. A description of the baseline model construction, preliminary program evaluation, and recommendations are provided. All program conclusions are considered preliminary until a larger and more comprehensive study is conducted. (author)

Preliminary assessment of the Louisiana Home Energy Rebate Offer program using IPMVP guidelines

International Nuclear Information System (INIS)

Kaiser, Mark J.; Pulsipher, Allan G.

2010-01-01

The Louisiana Home Energy Rebate Offer (HERO) is a residential energy conservation program established in 1999 to provide rebates for qualified applicants to build new homes that are more energy efficient or improve the energy efficiency of existing homes. Energy conservation programs require careful evaluation because of the high cost to implement the measures and the expectation that they will reduce energy use. The purpose of this paper is to demonstrate that residential energy conservation measures in a hot and humid climate can be evaluated using the International Performance Measurement and Verification Protocol (IPMVP), a best practice methodology commonly used in industrial and commercial performance-based contracts, but rarely, if ever, applied to residential programs. Using a random sample of 60 HERO participants, we were able to construct statistically significant electricity consumption baseline models for 90% of households. We determined that more than half of the sample participants consumed more electricity after their efficiency improvement, with an average net household savings of 172 kWh/yr, about 1% pre-retrofit consumption. A description of the baseline model construction, preliminary program evaluation, and recommendations are provided. All program conclusions are considered preliminary until a larger and more comprehensive study is conducted.

Alternative-fueled truck demonstration natural gas program: Caterpillar G3406LE development and demonstration

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-06-01

In 1990, the California Energy Commission, the South Coast Air Quality Management District, and the Southern California Gas Company joined together to sponsor the development and demonstration of compressed natural gas engines for Class 8 heavy-duty line-haul trucking applications. This program became part of an overall Alternative-Fueled Truck Demonstration Program, with the goal of advancing the technological development of alternative-fueled engines. The demonstration showed natural gas to be a technically viable fuel for Class 8 truck engines.

High-uranium-loaded U3O8--Al fuel element development program

International Nuclear Information System (INIS)

Martin, M.M.

1978-01-01

The High-Uranium-Loaded U 3 O 8 --Al Fuel Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages

Notification: Preliminary Research on EPA's Design for the Environment Product Labeling Program OIG

Science.gov (United States)

Project #OPE-FY14-4012, November 06, 2013. The Office of Inspector General (OIG) is starting preliminary research on the U.S. Environmental Protection Agency’s (EPA’s) Design for the Environment (DfE) Product Labeling Program.

VANTAGE 5 PWR fuel assembly demonstration program at Virgil C. Summer nuclear station

International Nuclear Information System (INIS)

Warner, D.C.; Orr, W.L.

1985-01-01

VANTAGE 5 is an improved PWR fuel product designed and manufactured by Westinghouse Electric Corporation. The VANTAGE 5 fuel design features integral fuel burnable absorbers, intermediate flow mixer grids, axial blankets, high burnup capability, and a reconstitutable top nozzle. A demonstration program for this fuel design commenced in late 1984 in cycle 2 of the Virgil C. Summer Nuclear Station. Objectives for VANTAGE 5 fuel are reduced fuel cycle costs, better core operating margins, and increased design and operating flexibility. Inspections of the VANTAGE 5 demonstration assemblies are planned at each refueling outage

Characterization program management plan for Hanford K Basin spent nuclear fuel

International Nuclear Information System (INIS)

Lawrence, L.A.

1998-01-01

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

Development of a mobility and fuel strategy for Germany. Preliminary investigation; Entwicklung einer Mobilitaets- und Kraftstoffstrategie fuer Deutschland. Voruntersuchung

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-08-15

Mobility is a key element in our society and often synonymous with traffic. It is desirable to secure a clean and diversified energy supply, to design a more sustainable transport structure and to minimize the burden on people and environment. This requires an updating of the fuel strategy 2004 and its progress toward a comprehensive mobility strategy and fuel strategy. Under this aspect, the contribution under consideration presents a preliminary investigation of the development of a mobility strategy and fuel strategy in Germany. Over the past ten years, the transport sector has reduced its energy consumption only slightly. The targets for the energy reduction now are -10 % up to 2020 and -40 % up to 2050. The transport sector challenges a significant contribution to the reduction of greenhouse gas emissions. In order to advance the necessary energy political as well as climate political changes the Federal Government wants to develop a comprehensive mobility strategy and fuel strategy. For this purpose, the contribution under consideration provides nine concrete recommendations for action for this.

A Preliminary Investigation into Parents' Concerns about Programming Education in Japanese Primary Schools

Science.gov (United States)

Maruyama, Yukiko; Kanoh, Hiroko; Adachi, Kinya

2017-01-01

To investigate parents' concerns about programming education in primary school, a preliminary online survey was carried out as a first step of the study. The result of the survey shows that parents seem to think that aim of programming education in primary school is not only learning coding. [For the complete proceedings, see ED579395.

Simulation of a 250 kW diesel fuel processor/PEM fuel cell system

Science.gov (United States)

Amphlett, J. C.; Mann, R. F.; Peppley, B. A.; Roberge, P. R.; Rodrigues, A.; Salvador, J. P.

Polymer-electrolyte membrane (PEM) fuel cell systems offer a potential power source for utility and mobile applications. Practical fuel cell systems use fuel processors for the production of hydrogen-rich gas. Liquid fuels, such as diesel or other related fuels, are attractive options as feeds to a fuel processor. The generation of hydrogen gas for fuel cells, in most cases, becomes the crucial design issue with respect to weight and volume in these applications. Furthermore, these systems will require a gas clean-up system to insure that the fuel quality meets the demands of the cell anode. The endothermic nature of the reformer will have a significant affect on the overall system efficiency. The gas clean-up system may also significantly effect the overall heat balance. To optimize the performance of this integrated system, therefore, waste heat must be used effectively. Previously, we have concentrated on catalytic methanol-steam reforming. A model of a methanol steam reformer has been previously developed and has been used as the basis for a new, higher temperature model for liquid hydrocarbon fuels. Similarly, our fuel cell evaluation program previously led to the development of a steady-state electrochemical fuel cell model (SSEM). The hydrocarbon fuel processor model and the SSEM have now been incorporated in the development of a process simulation of a 250 kW diesel-fueled reformer/fuel cell system using a process simulator. The performance of this system has been investigated for a variety of operating conditions and a preliminary assessment of thermal integration issues has been carried out. This study demonstrates the application of a process simulation model as a design analysis tool for the development of a 250 kW fuel cell system.

Progress in qualifying low-enriched U-Mo dispersion fuels

International Nuclear Information System (INIS)

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

2001-01-01

The U.S. Reduced Enrichment for Research and Test Reactors program is working to qualify dispersions of U-Mo alloys in aluminum with fuel-meat densities of 8 to 9 gU cm -3 . Post irradiation examinations of the small fuel plates irradiated in the Advanced Test Reactor during the high-temperature RERTR-3 tests are virtually complete, and analysis of the large quantity of data obtained is underway. We have observed that the swelling of the fuel plates is stable and modest and that the swelling is dominated by the temperature-dependent interaction of the U-Mo fuel and the aluminum matrix. In order to extract detailed information about the behavior of these fuels from the data, a complex fuel-plate thermal model is being developed to account for the effects of the changing fission rate and thermal conductivity of the fuel meat during irradiation. This paper summarizes the empirical results of the post irradiation examinations and the preliminary results of the model development. In addition, the schedule for irradiation of full-sized elements in the HFR-Petten is briefly discussed. (author)

Results from the DOE Advanced Gas Reactor Fuel Development and Qualification Program

Energy Technology Data Exchange (ETDEWEB)

David Petti

2014-06-01

Modular HTGR designs were developed to provide natural safety, which prevents core damage under all design basis accidents and presently envisioned severe accidents. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude and allows potential elimination of the need for evacuation and sheltering beyond a small exclusion area. This level, however, is predicated on exceptionally high fuel fabrication quality and performance under normal operation and accident conditions. Germany produced and demonstrated high quality fuel for their pebble bed HTGRs in the 1980s, but no U.S. manufactured fuel had exhibited equivalent performance prior to the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The design goal of the modular HTGRs is to allow elimination of an exclusion zone and an emergency planning zone outside the plant boundary fence, typically interpreted as being about 400 meters from the reactor. To achieve this, the reactor design concepts require a level of fuel integrity that is better than that claimed for all prior US manufactured TRISO fuel, by a few orders of magnitude. The improved performance level is about a factor of three better than qualified for German TRISO fuel in the 1980’s. At the start of the AGR program, without a reactor design concept selected, the AGR fuel program selected to qualify fuel to an operating envelope that would bound both pebble bed and prismatic options. This resulted in needing a fuel form that could survive at peak fuel temperatures of 1250°C on a time-averaged basis and high burnups in the range of 150 to 200 GWd/MTHM (metric tons of heavy metal) or 16.4 to 21.8% fissions per initial metal atom (FIMA). Although Germany has demonstrated excellent performance of TRISO-coated UO

Pathways to Commercial Success: Technologies and Products Supported by the Fuel Cell Technologies Program

Energy Technology Data Exchange (ETDEWEB)

Weakley, Steven A.

2012-09-28

The purpose of the project described in this report is to identify and document the commercial and emerging (projected to be commercialized within the next 3 years) hydrogen and fuel cell technologies and products that resulted from Department of Energy support through the Fuel Cell Technologies (FCT) Program in the Office of Energy Efficiency and Renewable Energy (EERE). Pacific Northwest National Laboratory (PNNL) undertook two efforts simultaneously to accomplish this project. The first effort was a patent search and analysis to identify patents related to hydrogen and fuel cells that are associated with FCT-funded projects (or projects conducted by DOE-EERE predecessor programs) and to ascertain the patents’ current status, as well as any commercial products that may have used the technology documented in the patent. The second effort was a series of interviews with current and past FCT personnel, a review of relevant program annual reports, and an examination of grants made under the Small Business Innovation Research and Small Business Technology Transfer Programs that are related to hydrogen and fuel cells.

Pathways to Commercial Success: Technologies and Products Supported by the Fuel Cell Technologies Program

Energy Technology Data Exchange (ETDEWEB)

Weakley, Steven A.; Brown, Scott A.

2011-09-29

The purpose of the project described in this report is to identify and document the commercial and emerging (projected to be commercialized within the next 3 years) hydrogen and fuel cell technologies and products that resulted from Department of Energy support through the Fuel Cell Technologies (FCT) Program in the Office of Energy Efficiency and Renewable Energy (EERE). To do this, Pacific Northwest National Laboratory (PNNL) undertook two efforts simultaneously to accomplish this project. The first effort was a patent search and analysis to identify hydrogen- and fuel-cell-related patents that are associated with FCT-funded projects (or projects conducted by DOE-EERE predecessor programs) and to ascertain the patents current status, as well as any commercial products that may have used the technology documented in the patent. The second effort was a series of interviews with current and past FCT personnel, a review of relevant program annual reports, and an examination of hydrogen- and fuel-cell-related grants made under the Small Business Innovation Research and Small Business Technology Transfer Programs, and within the FCT portfolio.

2016 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

Energy Technology Data Exchange (ETDEWEB)

Satyapal, Sunita [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2017-02-01

In the past year, the DOE Hydrogen Program (the Program) made substantial progress toward its goals and objectives. The Program has conducted comprehensive and focused efforts to enable the widespread commercialization of hydrogen and fuel cell technologies in diverse sectors of the economy. With emphasis on applications that will effectively strengthen our nation's energy security and improve our stewardship of the environment, the Program engages in research, development, and demonstration of critical improvements in the technologies. Highlights of the Program's accomplishments can be found in the sub-program chapters of this report.

2015 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

Energy Technology Data Exchange (ETDEWEB)

Popovich, Neil

2015-12-01

In the past year, the DOE Hydrogen Program (the Program) made substantial progress toward its goals and objectives. The Program has conducted comprehensive and focused efforts to enable the widespread commercialization of hydrogen and fuel cell technologies in diverse sectors of the economy. With emphasis on applications that will effectively strengthen our nation's energy security and improve our stewardship of the environment, the Program engages in research, development, and demonstration of critical improvements in the technologies. Highlights of the Program's accomplishments can be found in the sub-program chapters of this report.

2012 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

Energy Technology Data Exchange (ETDEWEB)

2012-12-01

In the past year, the DOE Hydrogen Program (the Program) made substantial progress toward its goals and objectives. The Program has conducted comprehensive and focused efforts to enable the widespread commercialization of hydrogen and fuel cell technologies in diverse sectors of the economy. With emphasis on applications that will effectively strengthen our nation's energy security and improve our stewardship of the environment, the Program engages in research, development, and demonstration of critical improvements in the technologies. Highlights of the Program's accomplishments can be found in the sub-program chapters of this report.

Fuel System: Automotive Mechanics Instructional Program. Block 4.

Science.gov (United States)

O'Brien, Ralph D.

The fourth of six instructional blocks in automotive mechanics, the lessons and supportive information in the document provide a guide for teachers in planning an instructional program in automotive fuel systems at the secondary and post secondary level. The material, as organized, is a suggested sequence of instruction within each block. Each…

Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

Energy Technology Data Exchange (ETDEWEB)

Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

2015-07-01

In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

none,

2015-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

none,

2011-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

none,

2013-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

none,

2014-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

none,

2012-09-01

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

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

Energy Technology Data Exchange (ETDEWEB)

Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

International Nuclear Information System (INIS)

Armstrong, J.; Hamilton, H.; Hyland, B.

2013-01-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

SOLID STATE ENERGY CONVERSION ALLIANCE (SECA) SOLID OXIDE FUEL CELL PROGRAM

Energy Technology Data Exchange (ETDEWEB)

Unknown

2003-06-01

This report summarizes the progress made during the September 2001-March 2002 reporting period under Cooperative Agreement DE-FC26-01NT41245 for the U. S. Department of Energy, National Energy Technology Laboratory (DOE/NETL) entitled ''Solid State Energy Conversion Alliance (SECA) Solid Oxide Fuel Cell Program''. The program focuses on the development of a low-cost, high-performance 3-to-10-kW solid oxide fuel cell (SOFC) system suitable for a broad spectrum of power-generation applications. The overall objective of the program is to demonstrate a modular SOFC system that can be configured to create highly efficient, cost-competitive, and environmentally benign power plants tailored to specific markets. When fully developed, the system will meet the efficiency, performance, life, and cost goals for future commercial power plants.

Yugoslav spent nuclear fuel management program and international perspectives

International Nuclear Information System (INIS)

Pesic, M.; Subotic, K.; Sotic, O.; Plecas, I.; Ljubenov, V.; Peric, A.; Milosevic, M.

2002-01-01

Spent nuclear fuel stored in the Vinca Institute of Nuclear Sciences, Yugoslavia, consists of about 2.5 tons of metal uranium (initial enrichment 2%) and about 20 kg uranium dioxide (dispersed in aluminum matrix, initial fuel uranium enrichment 80%). This spent nuclear fuel is generated in operation of the RA heavy water research reactor during 1959-1984 period. Both types of fuel are of ex-USSR origin, have the same shape and dimensions and approximately the same initial mass of 235 nuclide. They are known as the TVR-S type of fuel elements. The total of 8030 spent fuel elements are stored at the RA research reactor premises, almost all in the spent fuel pool filled by ordinary water. The last used 480 high-enriched uranium spent fuel elements are kept in the drained RA reactor core since 1984. Fuel layer of both enrichments is covered with thin aluminium cladding. Due to non-suitable chemical parameters of water in the spent fuel storage pool, the corrosion processes penetrated aluminium cladding and aluminium walls od storage containers during storage period long from 20 to 40 years. Activity of fission products ( 137 Cs) is detected in water samples during water inspection in 1996 and experts of the lAEA Russia and USA were invited to help. By end of 2001, some remediation of the water transparency of the storage pool and inspections of water samples taken from the storage containers with the spent fuel elements were carried out by the Vinca Institute staff and with the help of experts from the Russia and the IAEA. Following new initiatives on international perspective on spent fuel management, a proposal was set by the IAEA, and was supported by the governments of the USA and the Russian Federation to ship the spent fuel elements of the RA research reactor to Mayak spent fuel processing plant in Russia. This paper describes current status of the reactor RA spent fuel elements, initiative for new Yugoslav spent fuel management program speculates on some of the

Development of a program for evaluating the temperature of SMART-P fuel rod

Energy Technology Data Exchange (ETDEWEB)

Cheon, Jin Sik; Lee, Byung Ho; Koo, Yang Hyun; Oh, Je Yong; Yim, Jeong Sik; Sohn, Dong Seong

2003-11-01

A code for evaluating the temperature of SMART-P fuel rod has been developed. Finite Element (FE) method is adopted for the developed code sharing the user subroutines which has been prepared for the ABAQUS commercial FE code. The developed program for SMART-P fuel rod corresponds to a nonlinear transient heat transfer problem, and uses a sparse matrix solver for FE equations during iterations at every time step. The verifications of the developed program were conducted using the ABAQUS code. Steady state and transient problems were analyzed for 1/8 rod model due to the symmetry of the fuel rod and full model. From the evaluation of temperature for the 1/8 rod model at steady state, maximal error of 0.18 % was present relative to the ABAQUS result. Analysis for the transient problem using the fuel rod model resulted in the same as the variation of centerline temperature from the ABAQUS code during a hypothetical power transient. Also, given a power depression in fuel meat as a function of burnup, its effect on the centerline temperature was more precisely evaluated by the developed program compared to the ABAQUS code. The distribution of heat flux for the entire cross section and surface was almost identical for the two codes.

Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

Energy Technology Data Exchange (ETDEWEB)

Wiencek, T.C. [Argonne National Lab., IL (United States). Energy Technology Div.

1995-08-01

This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched ({approx}93% {sup 235}U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm{sup 3}) HEU fuel elements to highly loaded (up to 7 g U/cm{sup 3}) low-enrichment (<20% {sup 235}U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing.

Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

International Nuclear Information System (INIS)

Wiencek, T.C.

1995-08-01

This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched (∼93% 235 U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm 3 ) HEU fuel elements to highly loaded (up to 7 g U/cm 3 ) low-enrichment ( 235 U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing

Characteristics of fuel CRUD from Ringhals Unit 4. A comparison of CRUD samples from ultrasonic fuel cleaning and fuel scrape

International Nuclear Information System (INIS)

Chen, Jiaxin; Eskhult, Jonas; Marks, Chuck; Dingee, John; Bengtsson, Bernt; Wells, Daniel

2014-01-01

The characteristics and behaviour of PWR fuel CRUD are closely related to plant radiation field build-up and the risks of CRUD-Induced Power Shifts (CIPS, previously AOA) and CRUD-Induced Localized Corrosion (CILC). At Ringhals NPP fuel scrape technology has frequently been used to collect fuel CRUD samples as a part of plant water chemistry monitoring programs. In 2012, high efficiency ultrasonic fuel cleaning (HE-UFC) technology was first used at Ringhals Unit 4 to reduce the risk of CIPS, which was expected to increase due to steam generator replacement and a subsequent power uprate. In the HE-UFC system a “first-of-a-kind” CRUD sampling system was installed for collecting fuel CRUD. It is of interest to compare the fuel CRUD samples collected by the two different CRUD sampling methods and to understand if HE-UFC crud sampling could be used to replace or supplement the fuel scrape method. This paper presents some preliminary results on isotopic compositions, elemental compositions, and phase compositions of fuel CRUD samples collected from similar fuel rods and assemblies by both methods during the 2012 refueling outage, one cycle after steam generator replacement at Ringhals Unit 4. The results show that the characteristics of fuel CRUD sampled by HE-UFC and fuel scrape, although not always identical, were similar or correlated to some extent in terms of weight ratios of Ni to Fe and Cr to Fe as well as specific activities of Co-58, Co-60 and Cr-51. However, due to the limited experience with the HE-UFC sampling method, more consideration is required if the statistical significance of the obtained data and indications are to be verified. (author)

Development status and potential program for development of proliferation-resistant molten-salt reactors

International Nuclear Information System (INIS)

Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E. Jr.

1979-03-01

Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review

Development status and potential program for development of proliferation-resistant molten-salt reactors

Energy Technology Data Exchange (ETDEWEB)

Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E. Jr.

1979-03-01

Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review.

Site investigations for the disposal of spent fuel - investigation program

International Nuclear Information System (INIS)

Aeikaes, Timo

1985-11-01

The Industrial Power Company Ltd (TVO) is making preparations for the final disposal of spent nuclear fuel into the Finnish bedrock. The revised site investigation program for the years 1986-2010 is presented in this report. The objectives and activities in the near future are described in more detail. The main objectives and frame programs for the investigations in the more distant future are described. The program planning of these investigations are being developed in the preceding site investigations. The investigations for the site selection are divided into four phases: 1983-1985 selection of the investigation areas, preparations for the field investigations, drilling and investigations in a deep test borehole; 1986-1992 preliminary site investigations in 5-10 investigation areas; 1993-2000 detailed site investigations in 2-3 investigation areas. Site selection in the year 2000; 2001-2010 complementary investigations on the selected site. The first investigation phase will be carried out as planned. In this phase a 1001 m deep test borehole was drilled at Lavia in western Finland. With the investigations in the borehole and related development work, preparations were made for the future field investigations. The equipment and investigation methods are being developed during the site investigations. The equipment for taking groundwater samples and the unit for hydraulic testing have been developed. In the future the emphasis in the work will be in developing equipment for monitoring of the hydraulic head and measuring the volumetric flow. In groundwater sampling the present procedure can be improved by adding the test for the in-situ measurements. The results of the field investigations will be stored and processed in a centralized data base. The data base will transmit the results for the interpretation and then the interpreted results transmitted for model calculations and reporting. The cost estimate for the investigations in 1986-2010 is 110-125 million

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs draft environmental impact statement. Volume 1, Appendix B: Idaho National Engineering Laboratory Spent Nuclear Fuel Management Program

Energy Technology Data Exchange (ETDEWEB)

1994-06-01

The US Department of Energy (DOE) has prepared this report to assist its management in making two decisions. The first decision, which is programmatic, is to determine the management program for DOE spent nuclear fuel. The second decision is on the future direction of environmental restoration, waste management, and spent nuclear fuel management activities at the Idaho National Engineering Laboratory. Volume 1 of the EIS, which supports the programmatic decision, considers the effects of spent nuclear fuel management on the quality of the human and natural environment for planning years 1995 through 2035. DOE has derived the information and analysis results in Volume 1 from several site-specific appendixes. Volume 2 of the EIS, which supports the INEL-specific decision, describes environmental impacts for various environmental restoration, waste management, and spent nuclear fuel management alternatives for planning years 1995 through 2005. This Appendix B to Volume 1 considers the impacts on the INEL environment of the implementation of various DOE-wide spent nuclear fuel management alternatives. The Naval Nuclear Propulsion Program, which is a joint Navy/DOE program, is responsible for spent naval nuclear fuel examination at the INEL. For this appendix, naval fuel that has been examined at the Naval Reactors Facility and turned over to DOE for storage is termed naval-type fuel. This appendix evaluates the management of DOE spent nuclear fuel including naval-type fuel.

A nuclear fuel cycle system dynamic model for spent fuel storage options

International Nuclear Information System (INIS)

Brinton, Samuel; Kazimi, Mujid

2013-01-01

Highlights: • Used nuclear fuel management requires a dynamic system analysis study due to its socio-technical complexity. • Economic comparison of local, regional, and national storage options is limited due to the public financial information. • Local and regional options of used nuclear fuel management are found to be the most economic means of storage. - Abstract: The options for used nuclear fuel storage location and affected parameters such as economic liabilities are currently a focus of several high level studies. A variety of nuclear fuel cycle system analysis models are available for such a task. The application of nuclear fuel cycle system dynamics models for waste management options is important to life-cycle impact assessment. The recommendations of the Blue Ribbon Committee on America’s Nuclear Future led to increased focus on long periods of spent fuel storage [1]. This motivated further investigation of the location dependency of used nuclear fuel in the parameters of economics, environmental impact, and proliferation risk. Through a review of available literature and interactions with each of the programs available, comparisons of post-reactor fuel storage and handling options will be evaluated based on the aforementioned parameters and a consensus of preferred system metrics and boundary conditions will be provided. Specifically, three options of local, regional, and national storage were studied. The preliminary product of this research is the creation of a system dynamics tool known as the Waste Management Module (WMM) which provides an easy to use interface for education on fuel cycle waste management economic impacts. Initial results of baseline cases point to positive benefits of regional storage locations with local regional storage options continuing to offer the lowest cost

Preliminary study or RSG-GAS reactor fuel element integrity

International Nuclear Information System (INIS)

Soejoedi, A.; Tarigan, A.; Sujalmo; Prayoga, S.; Suhadi

1996-01-01

After 8 years of operation, RSG-GAS was able to reach 15 cycles of reactor operation with 116 irradiated fuels, whereas 49 fuels were produced by NUKEM; and the other 67 were produced by PEBN-BATAN. At the 15 T h cycles, it have been used 40 standard fuels and 8 control fuels (Forty standard fuels and eight control fuels have been used in the 15 t h core cycles). Several activities have been performed in the reactor, to investigate the fuel integrity, among of them are: .fuel visual test with under water camera, which the results were recorder in the video cassette, primary water quality test during, reactor operation, fuel failure detector system examination and compared the PIE results in the Radiometallurgy Installation (RMI). The results showed that the fuel integrity, before and after irradiation, have still good performance and the fission products have not been released yet

Preliminary performance analysis of exponential experimental system for the determination of neutron effective multiplication factor of PWR spent fuel

International Nuclear Information System (INIS)

Shin, Heesung; Lee, Sang-Yun; Ro, Seung-Gy; Seo, Gi-Seok; Kim, Ho-Dong

2002-01-01

An exponential experiment system which is composed of neutron detector, signal analysis system and neutron source, 10 mCi Cf-252 has been installed in the storage pool of PIEF at KAERI in order to experimentally determining neutron effective multiplication factors of PWR spent fuel assemblies. Preliminary functional characteristic tests of the experimental system are performed for C15, J14 and J44 assemblies loaded in the pool. As a result of preliminary tests, the average neutron counts obtained for 3 minutes in the plateau of the C15, J14 and J44 assemblies are about 1900, 3800 and 3200, respectively. A dip of the neutron flux density distribution is noticed in the spacer grid position. Neutron counts at those positions appear to be reduced to about 70 % in comparison to the fuel position. The measured axial neutron distribution shapes are compared with the result for the P14 assembly and Cs-137 gamma scanning data performed in KAERI. It is revealed that the spacer grid position measured is consistent with the design specifications within a 2.3 % error. The exponential decay constants for the C15 assembly were determined to be 0.152 and 0.165 for detector and source scanning, respectively. (author)

Application of linear programming and perturbation theory in optimization of fuel utilization in a nuclear reactor

International Nuclear Information System (INIS)

Zavaljevski, N.

1985-01-01

Proposed optimization procedure is fast due to application of linear programming. Non-linear constraints which demand iterative application of linear programming are slowing down the calculation. Linearization can be done by different procedures starting from simple empirical rules for fuel in-core management to complicated general perturbation theory with higher order of corrections. A mathematical model was formulated for optimization of improved fuel cycle. A detailed algorithm for determining minimum of fresh fuel at the beginning of each fuel cycle is shown and the problem is linearized by first order perturbation theory and it is optimized by linear programming. Numerical illustration of the proposed method was done for the experimental reactor mostly for saving computer time

Preliminary Analysis of High-Flux RSG-GAS to Transmute Am-241 of PWR’s Spent Fuel in Asian Region

Science.gov (United States)

Budi Setiawan, M.; Kuntjoro, S.

2018-02-01

A preliminary study of minor actinides (MA) transmutation in the high flux profile RSG-GAS research reactor was performed, aiming at an optimal transmutation loading for present nuclear energy development. The MA selected in the analysis includes Am-241 discharged from pressurized water reactors (PWRs) in Asian region. Until recently, studies have been undertaken in various methods to reduce radiotoxicity from actinides in high-level waste. From the cell calculation using computer code SRAC2006, it is obtained that the target Am-241 which has a cross section of the thermal energy absorption in the region (group 8) is relatively large; it will be easily burned in the RSG-GAS reactor. Minor actinides of Am-241 which can be inserted in the fuel (B/T fuel) is 2.5 kg which is equivalent to Am-241 resulted from the partition of spent fuel from 2 units power reactors PWR with power 1000MW(th) operated for one year.

Conservation education for Fire, Fuel and Smoke Program

Science.gov (United States)

Wayne Cook

2009-01-01

The mission of Conservation Education for the Fire, Fuel and Smoke (FFS) Program is to develop and deliver high-quality, science-based education about wildland fire to students, educators, the general public, and agency staff. Goals: 1) Increase awareness of the scope and content of FFS research. 2) Improve understanding of fundamental concepts in wildland fire science...

77 FR 13009 - Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel...

Science.gov (United States)

2012-03-05

... Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel Pathways Under the Renewable Fuel Standard Program AGENCY: Environmental Protection Agency (EPA). ACTION: Withdrawal... Renewable Fuel Standard program regulations. Because EPA received adverse comment, we are withdrawing the...

Enhancing instruction in Fuels and Combustion Laboratory via a developed computer-assisted program for establishing efficient coal-diesel oil mixture (CDOM) fuel proportions

Energy Technology Data Exchange (ETDEWEB)

Maglaya, A.B. [La Salle University, Manila (Philippines). Dept. of Mechanical Engineering

2004-07-01

This paper discusses the relevance of digital computation in Fuels and Combustion Laboratory experiments used by the senior students of the Department of Mechanical Engineering, De La Salle University-Manila, Philippines. One of the students' experiments involved the determination of the most efficient CDOM fuel proportion as alternative fuel to diesel oil for steam generators and other industrial applications. Theoretical calculations show that it requires tedious and repetitive computations. A computer-assisted program was developed to lessen the time-consuming activities. The formulation of algorithms were based on the system of equations of the heat interaction between the CDOM fuel, combustion air and products of combustion and by applying the principles of mass and energy equations (or the First Law of Thermodynamics) for reacting systems were utilized. The developed computer-assisted program output verified alternative fuel selected through actual experimentation.

Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

Energy Technology Data Exchange (ETDEWEB)

Paul A Demkowicz; Paul Demkowicz; David V Laug

2010-10-01

Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

Application of room temperature ionic liquids in advanced fuel cycles RIAR research concept program users

International Nuclear Information System (INIS)

Bychkov, Alexander V.; Kormilitsyn, Michael V.; Savochkin, Yuri P.; Osipenko, Alexander G.; Smolensky, Valeri V.; Shadrin, Alexander Yu.; Babain, Vladimir A.

2005-01-01

The paper reviews briefly the research program on application of Room Temperature Ionic Liquids (RTILs) in some processes of the nuclear fuel reprocessing, particularly in the uranium-aluminum fuel reprocessing and separation of TPEs and REEs from the PUREX process liquid waste, and approaches to implementation of this program. (author)

Preliminary design and analysis on nuclear fuel cycle for fission-fusion hybrid spent fuel burner

International Nuclear Information System (INIS)

Chen Yan; Wang Minghuang; Jiang Jieqiong

2012-01-01

A wet-processing-based fuel cycle and a dry-processing were designed for a fission-fusion hybrid spent fuel burner (FDS-SFB). Mass flow of SFB was preliminarily analyzed. The feasibility analysis of initial loaded fuel inventory, recycle fuel fabrication and spent fuel reprocessing were preliminarily evaluated. The results of mass flow of FDS-SFB demonstrated that the initial loaded fuel inventory, recycle fuel fabrication and spent fuel reprocessing of nuclear fuel cycle of FDS-SFB is preliminarily feasible. (authors)

Preliminary experience and near future utilization programmes of the MPR-30 fueled by LEU [low enriched uranium

International Nuclear Information System (INIS)

Arbie, B.; Soentono, S.

1987-01-01

The MTR type reactor MPR-30 G.A. Siwabessy, located at PUSPIPTEK Serpong has recently reached its first criticality. This multipurpose reactor is supposed to be the first MTR type reactor in the world that is designed and constructed to be fueled by low enriched uranium. Preliminary experience covering the approach to the first criticality and the excess reactivity loading as well as some thermal hydraulics and power ascension tests are briefly presented and discussed. The near future utilization programmes during and after commissioning are also presented. (Author)

Program plan for research and development in support of LWR fuel recycle

International Nuclear Information System (INIS)

1975-01-01

The ERDA program that is being planned to assist industry in the commercialization of the LWR fuel cycle will involve a range of activities, including joint programs with industry, R and D to provide technology, conceptual design of fuel recycle facilities, and environmental and economic assessments. A two-part program to begin in 1976 that is a portion of the overall ERDA plan is described. Responsibility for coordination and management of the tasks described in this document has been assigned to Du Pont as prime contractor to the ERDA Savannah River Operations Office. The first part of the program consists of the conceptual design of complete recycle facilities. The second part of the program, which will proceed concurrently, consists of supporting R and D activities, economic and environmental studies, and other studies to assist in the regulatory process. The R and D program will include both near-term activities in support of the conceptual design effort, and other activities aimed at general improvements in fuel cycle technology. The conceptual design will be used to develop current cost information for a complete reprocessing complex. The design will be based initially on current technology with provision for improvements as confirmatory information and advanced technology become available from the R and D program. The conceptual design and cost estimate will be developed by the Du Pont Atomic Energy Division. The R and D program and supporting studies will be directed at uncertainties in current technology as well as toward development of improved technology. It will include such R and D as might be appropriate for ERDA to undertake in support of joint programs with industry. The Savannah River Laboratory will have responsibility for coordinating the program

Burn-Up Calculation of the Fuel Element in RSG-GAS Reactor using Program Package BATAN-FUEL

International Nuclear Information System (INIS)

Mochamad Imron; Ariyawan Sunardi

2012-01-01

Calculation of burn lip distribution of 2.96 gr U/cc Silicide fuel element at the 78 th reactor cycle using computer code program of BATAN-FUEL has been done. This calculation uses inputs such as generated power, operation time and a core assumption model of 5/1. Using this calculation model burn up for the entire fuel elements at the reactor core are able to be calculated. From the calculation it is obtained that the minimum burn up of 6.82% is RI-50 at the position of A-9, while the maximum burn up of 57.57% is RI 467 at the position of 8-7. Based on the safety criteria as specified in the Safety Analysis Report (SAR) RSG-GAS reactor, the maximum fuel burn up allowed is 59.59%. It then can be concluded that pattern that elements placement at the reactor core are properly and optimally done. (author)

Preliminary safety design analysis of KALIMER

Energy Technology Data Exchange (ETDEWEB)

Suk, Soo Dong; Kwon, Y. M.; Kim, K. D. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-03-01

The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute(KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This document first introduces a set of safety design requirements and accident evaluation criteria established for the conceptual design of KALIMER and then summarizes some of the preliminary results of engineering and design analyses performed for the safety of KALIMER. 19 refs., 19 figs., 6 tabs. (Author)

Effects of an oxidizing atmosphere in a spent fuel packaging facility

International Nuclear Information System (INIS)

Einziger, R.E.

1991-09-01

Sufficient oxidation of spent fuel can cause a cladding breach to propagate, resulting in dispersion of fuel particulates and gaseous radionuclides. The literature for spent fuel oxidation in storage and disposal programs was reviewed to evaluate the effect of an oxidizing atmosphere in a preclosure packaging facility on (1) physical condition of the fuel and (2) operations in the facility. Effects such as cladding breach propagation, cladding oxidation, rod dilation, fuel dispersal, 14 C and 85 Kr release, and crud release were evaluated. The impact of these effects, due to oxidation, upon a spent fuel handling facility is generally predicted to be less than the impact of similar effects due to fuel rod breached during handling in an inert-atmosphere facility. Preliminary temperature limits of 240 degree C and 227 degree C for a 2-week or 4-week handling period and 175 degree C for 2-year lag storage would prevent breach propagation and fuel dispersal. Additional data that are needed to support the assumptions in this analysis or complete the database were identified

Preliminary evaluation of pin power distribution for fuel assemblies of SMART by MCNP

International Nuclear Information System (INIS)

Kim, Kyo Youn

1998-08-01

Monte Carlo transport code MCNP can describe an object sophisticately by use of three-dimensional modelling and can adopt a continuous energy cross-section library. Therefore MCNP has been widely utilized in the field of radiation physics to estimate fluxes and dose rates for nuclear facilities and to review results from conventional methods such a as discrete ordinates method and point kernel method. The Monte Carlo method has recently been introduced to estimated the neutron multiplication factor and pin power distribution in the fuel assembly of a reactor core. The operating thermal power of SMART core is 330 MWt and there are 57 fuel assemblies in the core. In this study it was assumed that the core has 4 types of fuel assemblies. In this study, MCNP4a was used to perform to estimate criticality and normalized pin power distribution in a fuel assembly of SMART core. The results from MCNP4a calculations are able to be used review those from nuclear design/analysis code. It is very complicated to pick up interested data from MCNP output list and to normalize pin power distribution in a fuel assembly because MCNP is not only a nuclear design/analysis code. In this study a program FAPIN was developed to generated a generate a normalized pin power distribution from the MCNP output list. (author). 11 refs

European passive plant program preliminary safety analyses to support system design

International Nuclear Information System (INIS)

Saiu, Gianfranco; Barucca, Luciana; King, K.J.

1999-01-01

In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In the Phase 1 of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) was completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. Incorporation of the EUR has been a key design requirement for the EP1000 form the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. The present paper describes the EP1000 approach to safety analysis and, in particular, to the Design Extension Conditions that, according to the EUR, represent the preferred method for giving consideration to the Complex Sequences and Severe Accidents at the design stage without including them in the design bases conditions. Preliminary results of some DEC analyses and an overview of the probabilistic safety assessment (PSA) are also presented. (author)

Direct electrical heating of irradiated metal fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Emerson, J.E.; Savoie, F.E.; Johanson, E.W.

1985-01-01

The Integral Fast Reactor (IFR) concept proposed by Argonne National Laboratory utilizes a metal fuel core. Reactor safety analysis requires information on the potential for fuel axial expansion during severe thermal transients. In addition to a comparatively large thermal expansion coefficient, metallic fuel has a unique potential for enhanced pre-failure expansion driven by retained fission gas and ingested bond sodium. In this paper, the authors present preliminary results from three direct electrical heating (DEH) experiments performed on irradiated metal fuel to investigate axial expansion behavior. The test samples were from Experimental Breeder Reactor II (EBR-II) driver fuel ML-11 irradiated to 8 at.% burnup. Preliminary analysis of the results suggest that enhanced expansion driven by trapped fission gas can occur

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...

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Whole-core damage analysis of EBR-II driver fuel elements following SHRT program

International Nuclear Information System (INIS)

Chang, L.K.; Koenig, J.F.; Porter, D.L.

1987-01-01

In the Shutdown Heat Removal Testing (SHRT) program in EBR-II, fuel element cladding temperatures of some driver subassemblies were predicted to exceed temperatures at which cladding breach may occur. A whole-core thermal analysis of driver subassemblies was performed to determine the cladding temperatures of fuel elemnts, and these temperatures were used for fuel element damage calculation. The accumulated cladding damage of fuel element was found to be very small and fuel element failure resulting from SHRT transients is unlikely. No element breach was noted during the SHRT transients. The reactor was immediately restarted after the most severe SHRT transient had been completed and no driver fuel breach has been noted to date. (orig.)

Program SYVAC, for stochastic assessment of nuclear fuel waste disposal

International Nuclear Information System (INIS)

Sherman, G.R.; Hoffman, K.J.; Donahue, D.C.

1985-01-01

In this paper, the computer program SYVAC, used to assess concepts for the disposal of nuclear fuel waste, is described with regard to the development approach, the basic program structure, and quality assurance. The interrelationships of these aspects are illustrated by detailed descriptions of two concepts of fundamental importance to the program: the method of selecting parameter values from input probability density functions, and the numerical evaluation of the convolution integral. Quality assurance procedures, including different types of comparisons and peer review, are presented

Irradiation program of slightly enriched fuel elements at the Atucha I nuclear power plant

International Nuclear Information System (INIS)

Casario, J.A.; Cesario, R.H.; Perez, R.A.; Sidelnik, J.I.

1987-01-01

An irradiation program of fuel elements with slightly enriched uranium is implemented, tending to the homogenization of core at Atucha I nuclear power plant. The main benefits of the enrichment program are: a) to extend the average discharge burnup of fuel elements, reducing the number of elements used to generate the same amount of energy. This implies a smaller annual consumption of elements and consequently the reduction of transport and replacement operations and of the storage pool systems as well as that of radioactive wastes; b) the saving of uranium and structural materials (Zircaloy and others). In the initial stage of program an homogeneous core enrichment of 0.85% by weight of U-235 is anticipated. The average discharge burnup of fuel elements, as estimated by previous studies, is approximately 11.6 MW d/kg U. The annual consumption of fuel elements is reduced from 396 of natural uranium to 205, with a load factor of 0.85. It is intended to reach the next equilibrium steps with an enrichment of 1.00 and 1.20% in U-235. (Author)

Developments in the Canadian program for geological disposal of nuclear fuel waste

International Nuclear Information System (INIS)

Allan, C.J.; Nuttall, K.

1996-01-01

The Canadian Nuclear Fuel Waste Management Program is at the end of disposal concept and technology development and is now undergoing a comprehensive environmental review. This paper will review: the history of the Canadian program; the disposal concept and the associated technologies; the program achievements and the lessons learned; and the status of the environmental review. (author)

Examining the Preliminary Efficacy of a Dating Violence Prevention Program for Hispanic Adolescents