Nuclear fuel assembly incorporating primary and secondary structural support members

International Nuclear Information System (INIS)

Carlson, W.R.; Gjertsen, R.K.; Miller, J.V.

1987-01-01

A nuclear fuel assembly, comprising: (a) an upper end structure; (b) a lower end structure; (c) elongated primary structural members extending longitudinally between and rigidly interconnecting the upper and lower end structures, the upper and lower end structures and primary structural members together forming a rigid structural skeleton of the fuel assembly; (d) transverse grids supported on the primary structural members at axially spaced locations therealong between the upper and lower end structures; (e) fuel rods extending through and supported by the grids between the upper and lower end structures so as to extend in generally side-by-side spaced relation to one another and to the primary structural members; and (f) elongated secondary structural members extending longitudinally between but unconnected with the upper and lower end structures, the secondary structural members extending through and rigidly interconnected with the grids to extend in generally side-by-side spaced relation to one another, to the fuel rods and to the primary structural members so as to bolster the stiffness of the structural skeleton of the fuel assembly

Premixed flame chemistry of a gasoline primary reference fuel surrogate

KAUST Repository

Selim, Hatem; Mohamed, Samah; Hansen, Nils; Sarathy, Mani

2017-01-01

Investigating the combustion chemistry of gasoline surrogate fuels promises to improve detailed reaction mechanisms used for simulating their combustion. In this work, the combustion chemistry of one of the simplest, but most frequently used

Calibration of the TVO spent BWR reference fuel assembly

International Nuclear Information System (INIS)

Tarvainen, M.; Baecklin, A.; Haakanson, A.

1992-02-01

In 1989 the Support Programmes of Finland (FSP) and Sweden (SSP) initiated a joint task to cross calibrate the burnup of the IAEA spent BWR reference fuel assembly at the TVO AFR storage facility (TVO KPA-STORE) in Finland. The reference assembly, kept separately under the IAEA seal, is used for verification measurements of spent fuel by GBUV method (SG-NDA-38). The cross calibration was performed by establishing a calibration curve, 244 Cm neutron rate versus burnup, using passive neutron assay (PNA) measurements. The declared burnup of the reference assembly was compared with the burnup value deduced from the calibration curve. A calibration line was also established by using the GBUV method with the aid of high resolution gamma ray spectrometry (HRGS). Normalization between the two different facilities was performed using sealed neutron and gamma calibration sources. The results of the passive neutron assay show consistency, better than 1 %, between the declared mean burnup of the reference assembly and the burnup deduced from the calibration curve. The corresponding consistency is within +-2 % for the HRGS measurements

Criticality reference benchmark calculations for burnup credit using spent fuel isotopics

International Nuclear Information System (INIS)

Bowman, S.M.

1991-04-01

To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as ''burnup credit,'' the fuel loading capacity of these casks can be increased. One of the difficulties in implementing burnup credit in criticality analyses is that there have been no critical experiments performed with spent fuel which can be used for computer code validation. In lieu of that, a reference problem set of fresh fuel critical experiments which model various conditions typical of light water reactor (LWR) transportation and storage casks has been identified and used in the validation of SCALE-4. This report documents the use of this same problem set to perform spent fuel criticality benchmark calculations by replacing the actual fresh fuel isotopics from the experiments with six different sets of calculated spent fuel isotopics. The SCALE-4 modules SAS2H and CSAS4 were used to perform the analyses. These calculations do not model actual critical experiments. The calculated k-effectives are not supposed to equal unity and will vary depending on the initial enrichment and burnup of the calculated spent fuel isotopics. 12 refs., 11 tabs

Criteria of reference radionuclides for safety analysis of spent fuel waste disposal

International Nuclear Information System (INIS)

Suryanto

1998-01-01

Study on the criteria for reference radionuclides selection for assessment on spent fuel disposal have done. The reference radionuclides in this study means radionuclides are predicted to contribute of the most radiological effect for man if spent fuel waste are discharged on deep geology formation. The research was done by investigate critically of parameters were used on evaluation a kind of radionuclide. Especially, this research study of parameter which relevant disposal case and or spent fuel waste on deep geology formation . The research assumed that spent fuel discharged on deep geology by depth 500-1000 meters from surface of the land. The migration scenario Radionuclides from waste form to man was assumed particularly for normal release in which Radionuclides discharge from waste form in a series thorough container, buffer, geological, rock, to fracture(fault) and move together with ground water go to biosphere and than go into human body. On this scenario, the parameter such as radionuclides inventory, half life, heat generation, hazard index based on maximum permissible concentration (MPC) or annual limit on intake (ALI) was developed as criteria of reference radionuclides selection. The research concluded that radionuclides inventory, half live, heat generated, hazard index base on MPC or ALI can be used as criteria for selection of reference Radionuclide. The research obtained that the main radionuclides are predicted give the most radiological effect to human are as Cs-137, Sr-90, I-129, Am-243, Cm-244, Pu-238, Pu-239, Pu-240. The radionuclides reasonable to be used as reference radionuclides in safety analysis at spent fuel disposal. (author)

A new improvement on a chemical kinetic model of primary reference fuel for multi-dimensional CFD simulation

International Nuclear Information System (INIS)

Zhen, Xudong; Wang, Yang; Liu, Daming

2016-01-01

Highlights: • A new optimized chemical kinetic mechanism for PRF is developed. • New mechanism optimization is performed based on the CHEMKIN simulations. • More reactions of C_0–C_1 oxidation are added in the present mechanism. • Good performance is achieved of mechanism by validating various reactors and operating conditions. - Abstract: In the present study, for the multi-dimensional CFD (computational fluid dynamics) combustion simulations of internal combustion engines, a new optimized chemical kinetic reaction mechanism for the oxidation of PRF (primary reference fuel) instead of gasoline has been developed. In order to carry out the in-depth research for combustion phenomenon of internal combustion engines, an optimized reduced PRF mechanism including more intermediate species and radicals was developed. The developed mechanism contains of iso-octane (C_8H_1_8) and n-heptane (C_7H_1_6) surrogates, which contains of 51-species and 193 reactions. Compared with many other mechanisms of PRF, more reactions of C_0–C_1 oxidation (100 reactions) are added in the present mechanism. In order to improve the performances of the model, the developed mechanism focused on the improvement through the prediction of the ignition delay time. The developed mechanism has been validated against various experimental and simulation data including shock tube data, laminar flame speed data and HCCI (homogeneous charge compression ignition) engine data. The results showed that the developed PRF mechanism was agreements with the experimental data and other approved reduced mechanisms, and it could be applied to the multi-dimensional CFD simulations for internal combustion engines.

A reference regional nuclear fuel centre

International Nuclear Information System (INIS)

1978-01-01

A nuclear fuel centre groups the facilities for spent fuel reprocessing, plutonium fuel fabrication, waste conditioning, and interim storage on a single site. The technical aspects of safety and protection, and the socio-economic consequences of two types of centre have been studied. The reference centre has an initial reprocessing capacity of 1500 tonnes. This capacity is quadrupled by the construction of two new units in 15 years. The other centre considered is a quarter of this size. A description is given of the processes used, the personal and capital requirements for construction and operation of the plant, the transport of radioactive waste and products, and the quantities involved. The local radiological impact is low and could be further reduced to a level well below that of natural radioactivity. The resulting increase in economic activity, employment, income redistribution and the new infrastructure requirements are estimated for a rural or semi-rural region. Measures to prevent tension are proposed. The impact of the host country's balance of payments, finances, employment situation and technological knowhow is evaluated. The original centre is compared with equivalent facilities scattered geographically

Combustor with two stage primary fuel tube with concentric members and flow regulating

Science.gov (United States)

Parker, David Marchant; Whidden, Graydon Lane; Zolyomi, Wendel

1999-01-01

A combustor for a gas turbine having a centrally located fuel nozzle and inner, middle and outer concentric cylindrical liners, the inner liner enclosing a primary combustion zone. The combustor has an air inlet that forms two passages for pre-mixing primary fuel and air to be supplied to the primary combustion zone. Each of the pre-mixing passages has a circumferential array of swirl vanes. A plurality of primary fuel tube assemblies extend through both pre-mixing passages, with each primary fuel tube assembly located between a pair of swirl vanes. Each primary fuel tube assembly is comprised of two tubular members. The first member supplies fuel to the first pre-mixing passage, while the second member, which extends through the first member, supplies fuel to the second pre-mixing passage. An annular fuel manifold is divided into first and second chambers by a circumferentially extending baffle. The proximal end of the first member is attached to the manifold itself while the proximal end of the second member is attached to the baffle. The distal end of the first member is attached directly to the second member at around its mid-point. The inlets of the first and second members are in flow communication with the first and second manifold chambers, respectively. Control valves separately regulate the flow of fuel to the two chambers and, therefore, to the two members of the fuel tube assemblies, thereby allowing the flow of fuel to the first and second pre-mixing passages to be separately controlled.

Soot and chemiluminescence in diesel combustion of bio-derived, oxygenated and reference fuels

NARCIS (Netherlands)

Klein-Douwel, R.J.H.; Donkerbroek, A.J.; Vliet, A.P. van; Boot, M.D.; Somers, L.M.T.; Baert, R.S.G.; Dam, N.J.; Meulen, J.J. ter

2009-01-01

High-speed imaging, spectroscopy and thermodynamical characterization are applied to an optically accessible, heavy-duty diesel engine in order to compare sooting and chemiluminescence behaviour of bio-derived, oxygenated fuels and various reference fuels. The fuels concerned include the bio-derived

Effect of reference parameters and properties of materials for WWER-type fuel elements on their reliability

Energy Technology Data Exchange (ETDEWEB)

Bibilashvili, Yu K; Malachenko, L L; Medvedev, A V; Solyany, V I; Sukhanov, G I; Tonkov, V Yu

1987-05-01

Present approach to requirements for reference parameters and properties of materials for WWER-1000 fuel elements is presented as well as evaluation of their effects on fuel reliability. Some results of investigations with the aim of improving fuel element reliability in operational NPP conditions are discussed. 4 references, 7 figures, 3 tables.

Performance of FFTF reference fuel and control assemblies

International Nuclear Information System (INIS)

Leggett, R.D.; Weber, E.T.

1984-11-01

This paper describes the performance of the reference fuel and control assemblies used in FFTF through the first four cycles of irradiation (446 equivalent full power days, EFPD). These assemblies performed flawlessly through the rigors of the Startup Testing Program, STP, (beginning in late 1979) with its cyclic operation and continued to do so throughout Cycles 1, 2, 3 and 4, the latter ending in April 1984

78 FR 21008 - Publication of Inflation Adjustment Factor, Nonconventional Source Fuel Credit, and Reference...

Science.gov (United States)

2013-04-08

... products) for calendar year 2012. DATES: The 2012 inflation adjustment factor, nonconventional source fuel... DEPARTMENT OF THE TREASURY Internal Revenue Service Publication of Inflation Adjustment Factor, Nonconventional Source Fuel Credit, and Reference Price for Calendar Year 2012 AGENCY: Internal Revenue Service...

76 FR 19524 - Publication of Inflation Adjustment Factor, Nonconventional Source Fuel Credit, and Reference...

Science.gov (United States)

2011-04-07

... products) for calendar year 2010. DATES: The 2010 inflation adjustment factor, nonconventional source fuel... DEPARTMENT OF THE TREASURY Internal Revenue Service Publication of Inflation Adjustment Factor, Nonconventional Source Fuel Credit, and Reference Price for Calendar Year 2010 AGENCY: Internal Revenue Service...

77 FR 22067 - Publication of Inflation Adjustment Factor, Nonconventional Source Fuel Credit, and Reference...

Science.gov (United States)

2012-04-12

... calendar year 2011. DATES: The 2011 inflation adjustment factor and nonconventional source fuel credit... DEPARTMENT OF THE TREASURY Internal Revenue Service Publication of Inflation Adjustment Factor, Nonconventional Source Fuel Credit, and Reference Price for Calendar Year 2011 AGENCY: Internal Revenue Service...

Prices versus policy: An analysis of the drivers of the primary fossil fuel mix

International Nuclear Information System (INIS)

Atalla, Tarek; Blazquez, Jorge; Hunt, Lester C.; Manzano, Baltasar

2017-01-01

Energy policymakers often attempt to shape their countries' energy mix, rather than leave it purely to market forces. By calibrating and simulating a Dynamic Stochastic General Equilibrium (DSGE) model, this paper analyzes the primary fossil fuel mix in the USA and compares it to Germany and the UK, given the different evolution of the mixes and the different roles played by relative prices and policy in North America and Europe. It is found that the model explains well the evolution of the primary fossil fuel mix in the USA for the period 1980–2014, suggesting that relative fossil fuel prices generally dominated in determining the mix during this time. However, this is not the case for Germany and the UK. For both countries, the model performs well only for the period after the market-oriented reforms in the 1990s. Additionally, the volatility of private consumption and output for the pre- and post-reform periods is evaluated for Germany and the UK and it is found that the liberalized energy markets brought about a transition from coal to natural gas, but with increased macroeconomic volatility. - Highlights: • Macroeconomic analysis of the importance of prices vs policy in driving the primary fossil fuel mix. • USA primary fossil fuel mix chiefly driven by relative prices since the early 1980s. • Germany and UK primary fossil fuel mix chiefly driven by policy until 1990s. • Germany and UK primary fossil fuel mix chiefly driven by relative prices since early to mid-1990s. • Transition from coal to natural gas in Germany and UK increased macroeconomic volatility.

The disposal of Canada's nuclear fuel waste: comments on the postclosure assessment of a reference system

International Nuclear Information System (INIS)

Allan, C.J.; Goodwin, B.W.

1996-07-01

Canada, like other countries, is developing technology for disposal of its nuclear fuel waste , based on the concept of geological disposal in stable plutonic rock of the Canadian Shield. The choice of methods, materials, and designs for a disposal system will ultimately be made on the basis of safety, taking into account the characteristics of the specific site on which the facility is to be developed, costs and practicality. As part of its work in developing the technology for the disposal of Canada's nuclear fuel waste, AECL analyzed the performance of a hypothetical disposal facility that incorporates specific design choices for the engineered barriers and that assumes a specific geological setting. This system, comprising the disposal facility and the geological setting, and the results of the performance analysis, is described in an Environmental Impact Statement that AECL submitted in 1994 and in a Primary Reference for the EIS 'The Disposal of Canada's Nuclear Fuel Waste: Postclosure Assessment of a Reference System.' The performance analysis was not intended to be a general proof of the safety of disposal, but rather it presents a safety analysis of one specific system to illustrate the postclosure assessment methodology and to demonstrate that safety could be achieved for the system in question. Although the design of the disposal facility analyzed and the geological setting have specific features, the results obtained from the safety analysis can, however, be used to provide considerable insight into the performance of the various components that comprise the multibarrier geological disposal system. Moreover, the results can show how changes in the performance of specific components can affect the overall performance of the system. This report discusses these aspects of the postclosure analysis. (author)

Selection of LEU/Th reference fuel for the HTGR-SC/C lead plant

International Nuclear Information System (INIS)

Turner, R.F.; Neylan, A.J.; Baxter, A.M.; McEachern, D.W.; Stansfield, O.M.

1983-05-01

This paper describes the reference fuel materials for the high-temperature gas-cooled reactor (HTGR) plant for steam cycle/cogeneration (SC/C). A development and testing program carried out in 1978 through 1982 led to the selection of coated fuel particles of uranium-oxycarbide (UCO) for fissile materials and thorium oxide (ThO 2 ) for fertiel materials. Low-enriched uranium (LEU) is the enrichment basis for the HTGR-SC/C application. While UC 2 and UO 2 would also meet the essential criteria for fissile fuel, the UCO, alternative was selected on the basis of improved performance, economics, and process conditions

The effect of reference parameters and properties of materials for WWER-type fuel elements on their reliability

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.; Malachenko, L.L.; Medvedev, A.V.; Solyany, V.I.; Sukhanov, G.I.; Tonkov, V.Yu.

1987-01-01

Present approach to requirements for reference parameters and properties of materials for WWER-1000 fuel elements is presented as well as evaluation of their effects on fuel reliability. Some results of investigations with the aim of improving fuel element reliability in operational NPP conditions are discussed. (author)

Radionuclide release from PWR fuels in a reference tuff repository groundwater

International Nuclear Information System (INIS)

Wilson, C.N.; Oversby, V.M.

1985-03-01

The Nevada Nuclear Waste Storage Investigations Project (NNWSI) is studying the suitability of the welded devitrified Topopah Spring tuff at Yucca Mountain, Nye County, Nevada, for potential use as a high-level nuclear waste repository. In support of the Waste Package task of NNWSI, tests have been conducted under ambient air environment to measure radionuclide release from two pressurized water reactor (PWR) spent fuels in water obtained from the J-13 well near the Yucca Mountain site. Four specimen types, representing a range of fuel physical conditions that may exist in a failed waste canister containing a limited amount of water were tested. The specimen types were: fuel rod sections split open to expose bare fuel particles; rod sections with water-tight end fittings with a 2.5-cm long by 150-μm wide slit through the cladding; rod sections with water-tight end fittings and two 200-μm-diameter holes through the cladding; and undefected rod segments with water-tight end fittings. Radionuclide release results from the first 223-day test runs on H.B. Robinson spent fuel specimens in J-13 water are reported and compared to results from a previous test series in which similar Turkey Point reactor spent fuel specimens were tested on deionized water. Selected initial results are also given for Turkey Point fuel specimens tested on J-13 water. Results suggest that the actinides Pu, Am, Cm and Np are released congruently with U as the UO 2 spent fuel matrix dissolves. Fractional release of 137 Cs and 99 Tc was greater than that measured for the actinides. Generally, lower radionuclide releases were measured for the H.B. Robinson fuel in J-13 water than for Turkey Point Fuel in deionized water. 8 references, 7 figures, 9 tables

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

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

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

International Nuclear Information System (INIS)

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

1997-09-01

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

Primary care: constipation and encopresis treatment strategies and reasons to refer.

Science.gov (United States)

Philichi, Lisa; Yuwono, Melawati

2010-01-01

The purpose of the study was to assess constipation and encopresis treatment strategies of primary care providers and determine reasons to refer to a pediatric gastroenterology specialist. A closed-ended questionnaire was mailed to a convenience sampling of 237 pediatric primary care providers. Ninety-one questionnaires were returned with a 38% response rate: 74 (81%) pediatricians and 17 (19%) nurse practitioners. The majority of responders recommended pharmacologic treatment and diet changes. Many providers (73%) estimated a 75%-100% success rate when managing constipation, whereas 19% providers estimated a greater than 80% success rate with encopresis patients. The number one reason to refer was unresponsiveness to treatment (71%), followed by parents want a second opinion (15%), rule out organic cause (9%), and management is too time-consuming (5%). Both primary care providers and pediatric gastroenterologists use medication strategies, but diet recommendations are not the same. Unresponsiveness to treatment is the main reason for referral. If better management can occur in the primary care setting, costly specialty services may be avoided and possibly reduce healthcare costs.

Ignition of alkane-rich FACE gasoline fuels and their surrogate mixtures

KAUST Repository

Sarathy, Mani; Kukkadapu, Goutham; Mehl, Marco; Wang, Weijing; Javed, Tamour; Park, Sungwoo; Oehlschlaeger, Matthew A.; Farooq, Aamir; Pitz, William J.; Sung, Chihjen

2015-01-01

Engines) gasoline test fuels and their corresponding PRF (primary reference fuel) blend in fundamental combustion experiments. Shock tube ignition delay times were measured in two separate facilities at pressures of 10, 20, and 40 bar, temperatures from

Calculation for Primary Combustion Characteristics of Boron-Based Fuel-Rich Propellant Based on BP Neural Network

OpenAIRE

Wan'e, Wu; Zuoming, Zhu

2012-01-01

A practical scheme for selecting characterization parameters of boron-based fuel-rich propellant formulation was put forward; a calculation model for primary combustion characteristics of boron-based fuel-rich propellant based on backpropagation neural network was established, validated, and then was used to predict primary combustion characteristics of boron-based fuel-rich propellant. The results show that the calculation error of burning rate is less than ± 7 . 3 %; in the formulation rang...

Forest road and fuel break siting with respect to reference fire intensities

Energy Technology Data Exchange (ETDEWEB)

Eastaugh, C. S.; Molina, D. M.

2012-11-01

Forest roads and permanent fuel breaks are an important part of fire suppression infrastructure, but due to maintenance and environmental costs many forest agencies seek to reduce the extent of these networks. The question of which roads should be retained or where fuel breaks should be established is contentious, and few quantified methods exist to aid management decisions. This study uses GIS procedures and develops a metric for road network vulnerability, which may be used to determine the relative effectiveness of a road network or a particular fuel break as a fire control line. The method constructs reference fire intensities, and compares the fire intensity at roadsides or fuel breaks with the overall forest average. In the case study area in Victoria's Central Highlands (southeast Australia), average fire intensities on the forest road network are found to closely match the forest average, indicating that roads in their current locations are not skewed towards more dangerous parts of the forest. The fuel break network however is likely to face fire intensities substantially greater than those in the average forest area. (Author) 33 refs.

The disposal of Canada's nuclear fuel waste: postclosure assessment of a reference system

International Nuclear Information System (INIS)

Goodwin, B.W.; McConnell, D.B.; Andres, T.H.

1994-01-01

The concept for disposal of Canada's nuclear fuel waste is based on a vault located deep in plutonic rock of the Canadian Shield. We document in this report a method to assess the long-term impacts of a disposal facility for nuclear fuel waste. The assessment integrates relevant information from engineering design studies, site investigations, laboratory studies, expert judgment and detailed mathematical analyses to evaluate system performance in terms of safety criteria, guidelines and standards. The method includes the use of quantitative tools such as the Systems Variability Analysis computer Code (SYVAC) to deal with parameter uncertainty and the use of reasoned arguments based on well-established scientific principles. We also document the utility of the method by describing its application to a hypothetical implementation of the concept called the reference disposal system. The reference disposal system generally conforms to the overall characteristics of the concept, except we have made some specific site and design choices so that the assessment would be more realistic. To make the reference system more representative of a real system, we have used the geological observations of the AECL's Whiteshell Research Area located near Lac du Bonnet, Manitoba, to define the characteristics of the geosphere and the groundwater flow system. This research area has been subject to more than a decade of geological and hydrological studies. The analysis of the reference disposal system provides estimates of radiological and chemical toxicity impacts on members of a critical group and estimates of possible impacts on the environment. The latter impacts include estimates of radiation dose to nonhuman organisms. Other quantitative analyses examine the use of derived constraints to improve the margin of safety, the effectiveness of engineered and natural barriers, and the sensitivity of the results to influential features, events, and processes of the reference disposal

The effect of attitudes on reference-dependent preferences: Estimation and validation for the case of alternative-fuel vehicles

DEFF Research Database (Denmark)

Mabit, Stefan Lindhard; Cherchi, Elisabetta; Jensen, Anders Fjendbo

2015-01-01

reference-dependent preferences and attitudes together may explain individual choices. In a modelling framework based on a hybrid choice model allowing for both concepts, we investigate how attitudes and reference-dependent preferences interact and how they affect willingness-to-pay measures and demand...... elasticities. Using a data set with stated choices among alternative-fuel vehicles, we see that allowing for reference-dependent preferences improves our ability to explain the stated choices in the data and that the attitude (appreciation of car features) explains part of the preference heterogeneity across...... with varying attitudes and reference values will act differently when affected by policy instruments related to the demand for alternative-fuel vehicles, e.g. subsidies....

Developing primary care in Hong Kong: evidence into practice and the development of reference frameworks.

Science.gov (United States)

Griffiths, Sian M; Lee, Jeff P M

2012-10-01

Enhancing primary care is one of the proposals put forward in the Healthcare Reform Consultation Document "Your Health, Your Life" issued in March 2008. In 2009, the Working Group on Primary Care, chaired by the Secretary for Food and Health, recommended the development of age-group and disease-specific primary care conceptual models and reference frameworks. Drawing on international experience and best evidence, the Task Force on Conceptual Model and Preventive Protocols of the Working Group on Primary Care has developed two reference frameworks for the management of two common chronic diseases in Hong Kong, namely diabetes and hypertension, in primary care settings. Adopting a population approach for the prevention and control of diabetes and hypertension across the life course, the reference frameworks aim to provide evidence-based and appropriate recommendations for the provision of continuing and comprehensive care for patients with chronic diseases in the community.

Spent nuclear fuel characterization for a bounding reference assembly for the receiving basin for off-site fuel

International Nuclear Information System (INIS)

Kahook, S.D.; Garrett, R.L.; Canas, L.R.

1995-01-01

A basis for interim operation 1 (BIO) for the receiving basin for off-site fuel (RBOF) facility at the U.S. Department of Energy's (DOE) Savannah River site nuclear materials production complex has been developed in accordance to draft DOE-STD-0019-93 (Ref. 2). The latter document requires a hazard categorization per DOE-STD-1027-92 (Ref. 3) for the safety analysis portion of the BIO. This classification places the facility in one of three categories as defined in DOE 5480.23 (Ref. 4) per the total radioactivity, which can be released during an accident. The diversity of spent nuclear fuels stored in the RBOF made an exacting assessment of the total radioactive inventory virtually impossible. This restriction led to a conservative calculation based on the concept of a hypothetical bounding reference fuel assembly (RFA) integrated over the total capacity of the facility. The RFA is derived from a systematic ranking of the real assemblies (current and expected) according to a maximum burnup criterion. The indicated scheme is not only simple but precluded a potential delay in the completion of the BIO

Behaviour of fission products in PWR primary coolant and defected fuel rods evaluation

International Nuclear Information System (INIS)

Bourgeois, P.; Stora, J.P.

1979-01-01

The activity surveillance of the PWR primary coolant by γ spectometry gives some informations on fuel failures. The activity of different nuclides e.g. Xenons, Kryptons, Iodines, can be correlated with the number of the defected fuel rods. Therefore the precharacterization with eventually a prelocalization of the related fuel assemblies direct the sipping-test and allows a saving of time during refueling. A model is proposed to calculate the number of the defected rods from the activity measurements of the primary coolant. A semi-empirical model of the release of the fission products has been built from the activity measurements of the primary coolant in a 900 MWe PWR. This model allows to calculate the number of the defected rods and also a typical parameter of the mean damage. Fission product release is described by three stages: release from uranium dioxide, transport across the gas gap and behaviour in the primary coolant. The model of release from the oxide considers a diffusion process in the grains with trapping. The release then occurs either directly to free surfaces or with a delay due to a transit into closed porosity of the oxide. The amount released is the same for iodine and rare gas. With the gas gap transit is associated a transport time and a probability of trapping for the iodines. In the primary coolant the purification and the radioactive decay are considered. (orig.)

Primary water chemistry optimization for extended fuel cycle operation. Results of the 'Duo experimentation' after three cycles

International Nuclear Information System (INIS)

Viricel, L.; Andrieu, C.; Segura, J.C.; Rocher, A.; Thomazet, J.; Clinard, M.H.; Dacquait, F.

2002-01-01

The primary coolant conditioning in French nuclear power plants is essentially based on the boron-lithium coordinated chemistry, with a target pH of 7.2 at 300 C and a maximum lithium concentration of 2.2 mg/kg. In 1996, EDF 1300 MWe units began operating 18-month fuel cycles, increasing boron concentrations at the beginning of the cycles. Since today the maximum lithium concentration in normal operation is 2.2 mg/kg, extended cycle operation results in a decrease in the pH at the beginning of the cycles, which may possibly lead to deposits in RCS, and particularly on the fuel cladding, and increased dose rates. It has to be noted that today, the fuel assemblies maximum burnup is set at 52 GWd/tU. One solution is to adjust the pH by increasing the lithium content at the beginning of the cycles, which is easy to implement and does not require any modification on the units. Hence, EDF is testing a ''modified'' chemistry regime in the > during 4 fuel cycles, with a maximum authorized lithium content of 3.5 mg/kg at the beginning of the cycles in the Cattenom 2 pilot unit. The Golfech 1 reference unit implements a standard boron-lithium coordination pH 300 7.2. The major goal of the experimentation is to assess the impact of elevated lithium concentrations at the beginning of the cycles on fuel cladding oxide behavior, mass transport and dose rates. This paper presents the results of the first three cycles of the Duo experimentation. (author)

Characteristics of spray from a GDI fuel injector for naphtha and surrogate fuels

KAUST Repository

Wang, Libing

2016-11-18

Characterization of the spray angle, penetration, and droplet size distribution is important to analyze the spray and atomization quality. In this paper, the spray structure development and atomization characterization of two naphtha fuels, namely light naphtha (LN) and whole naphtha (WN) and two reference fuel surrogates, i.e. toluene primary reference fuel (TPRF) and primary reference fuel (PRF) were investigated using a gasoline direct injection (GDI) fuel injector. The experimental setup included a fuel injection system, a high-speed imaging system, and a droplet size measurement system. Spray images were taken by using a high-speed camera for spray angle and penetration analysis. Sauter mean diameter, Dv(10), Dv(50), Dv(90), and particle size distribution were measured using a laser diffraction technique. Results show that the injection process is very consistent for different runs and the time averaged spray angles during the measuring period are 103.45°, 102.84°, 102.46° and 107.61° for LN, WN, TPRF and PRF, respectively. The spray front remains relatively flat during the early stage of the fuel injection process. The peak penetration velocities are 80 m/s, 75 m/s, 75 m/s and 79 m/s for LN, WN, TPRF and PRF, respectively. Then velocities decrease until the end of the injection and stay relatively stable. The transient particle size and the time-averaged particle size were also analyzed and discussed. The concentration weighted average value generally shows higher values than the arithmetic average results. The average data for WN is usually the second smallest except for Dv90, of which WN is the biggest. Generally the arithmetic average particle sizes of PRF are usually the smallest, and the sizes does not change much with the measuring locations. For droplet size distribution results, LN and WN show bimodal distributions for all the locations while TPRF and PRF shows both bimodal and single peak distribution patterns. The results imply that droplet size

Estimating fuel octane numbers from homogeneous gas-phase ignition delay times

KAUST Repository

Naser, Nimal

2017-11-05

Fuel octane numbers are directly related to the autoignition properties of fuel/air mixtures in spark ignition (SI) engines. This work presents a methodology to estimate the research and the motor octane numbers (RON and MON) from homogeneous gas-phase ignition delay time (IDT) data calculated at various pressures and temperatures. The hypothesis under investigation is that at specific conditions of pressure and temperature (i.e., RON-like and MON-like conditions), fuels with IDT identical to that of a primary reference fuel (PRF) have the same octane rating. To test this hypothesis, IDTs with a detailed gasoline surrogate chemical kinetic model have been calculated at various temperatures and pressures. From this dataset, temperatures that best represent RON and MON have been correlated at a specified pressure. Correlations for pressures in the range of 10–50 bar were obtained. The proposed correlations were validated with toluene reference fuels (TRF), toluene primary reference fuels (TPRF), ethanol reference fuels (ERF), PRFs and TPRFs with ethanol, and multi-component gasoline surrogate mixtures. The predicted RON and MON showed satisfactory accuracy against measurements obtained by the standard ASTM methods and blending rules, demonstrating that the present methodology can be a viable tool for a first approximation. The correlations were also validated against an extensive set of experimental IDT data obtained from literature with a high degree of accuracy in RON/MON prediction. Conditions in homogeneous reactors such as shock tubes and rapid compression machines that are relevant to modern SI engines were also identified. Uncertainty analysis of the proposed correlations with linear error propagation theory is also presented.

Estimating fuel octane numbers from homogeneous gas-phase ignition delay times

KAUST Repository

Naser, Nimal; Sarathy, Mani; Chung, Suk-Ho

2017-01-01

Fuel octane numbers are directly related to the autoignition properties of fuel/air mixtures in spark ignition (SI) engines. This work presents a methodology to estimate the research and the motor octane numbers (RON and MON) from homogeneous gas-phase ignition delay time (IDT) data calculated at various pressures and temperatures. The hypothesis under investigation is that at specific conditions of pressure and temperature (i.e., RON-like and MON-like conditions), fuels with IDT identical to that of a primary reference fuel (PRF) have the same octane rating. To test this hypothesis, IDTs with a detailed gasoline surrogate chemical kinetic model have been calculated at various temperatures and pressures. From this dataset, temperatures that best represent RON and MON have been correlated at a specified pressure. Correlations for pressures in the range of 10–50 bar were obtained. The proposed correlations were validated with toluene reference fuels (TRF), toluene primary reference fuels (TPRF), ethanol reference fuels (ERF), PRFs and TPRFs with ethanol, and multi-component gasoline surrogate mixtures. The predicted RON and MON showed satisfactory accuracy against measurements obtained by the standard ASTM methods and blending rules, demonstrating that the present methodology can be a viable tool for a first approximation. The correlations were also validated against an extensive set of experimental IDT data obtained from literature with a high degree of accuracy in RON/MON prediction. Conditions in homogeneous reactors such as shock tubes and rapid compression machines that are relevant to modern SI engines were also identified. Uncertainty analysis of the proposed correlations with linear error propagation theory is also presented.

Contamination of a PWR primary circuit by fuel pins with failed cladding

International Nuclear Information System (INIS)

Janvier, J.C.; Chagrot, M.

1979-01-01

The safety authorities in the principal nuclear countries appear to be attaching increasing importance to keeping reactor primary circuits as contamination-free as possible. Therefore, the consequences of cladding failures and especially of those resulting from fabrication defects have to be evaluated, for when these failures become systematic in nature they constitute an important source of contamination in pressurized-water reactors. The Grenoble Nuclear Research Centre is implementing a programme on the study of such failures with a view to analysing the behaviour of failed fuel elements. A distinction is made between two types of cladding failure, depending on whether the primary water enters the fuel pin as soon as the circuits are pressurized (fabrication defect) or whether the failure is caused during operation. The emission of gaseous fission products and halogens has been analysed in different operating modes (steady-state or transient), and in spite of the complexity of the phenomena involved, some results have been obtained which already enable one to evaluate fission product contamination of the primary circuit. (author)

Calculation for Primary Combustion Characteristics of Boron-Based Fuel-Rich Propellant Based on BP Neural Network

Directory of Open Access Journals (Sweden)

Wu Wan'e

2012-01-01

Full Text Available A practical scheme for selecting characterization parameters of boron-based fuel-rich propellant formulation was put forward; a calculation model for primary combustion characteristics of boron-based fuel-rich propellant based on backpropagation neural network was established, validated, and then was used to predict primary combustion characteristics of boron-based fuel-rich propellant. The results show that the calculation error of burning rate is less than ±7.3%; in the formulation range (hydroxyl-terminated polybutadiene 28%–32%, ammonium perchlorate 30%–35%, magnalium alloy 4%–8%, catocene 0%–5%, and boron 30%, the variation of the calculation data is consistent with the experimental results.

Pulverulent deposits on fuel assemblies in primary circuit on 2 units at EDF (France)

International Nuclear Information System (INIS)

Zatla, A.; Piana, O.

2015-01-01

In march 2013, during the moving of some fuel assemblies in the fuel deactivation pool of two units nuclear power plants (Blayais 4 and Chinon B2 units), some fine and powdery deposits in suspension have been observed. In the two cases, the fuel has stayed in the deactivation pool during a long time (12 and 3 months), the units have started for the first time to inject Zinc in the primary circuit in the beginning of the precedent fuel cycle, and the units have operated an extended fuel cycle. EDF has performed analysis to evaluate the harmfulness of the particles and to investigate the mechanisms involved in this phenomenon. 3 conclusions can be drawn. First, the deposits are made up of usual corrosion products issued from the primary circuit components. Secondly, the chemical composition and the powdery characteristics of the deposits exclude nuclear safety risks. Thirdly, the atypical behaviour of the deposits could probably be linked with a modification of their structure due to a zinc effect. Because of the lack of new understanding elements, the extension of zinc injection to other NPP units has been broken off currently in France. EDF is studying the international experience feedback to identify the operating parameters of the nuclear power plant which might influent the phenomenon, and to evaluate if the deposits could enhance fuel cladding corrosion

Phenomenological description reference concept (Spent Fuel - Copper - Iron - Bentonite - Granite)

International Nuclear Information System (INIS)

Lehikoinen, J.; Nordman, H.

2008-12-01

This report gives a broad overview of the salient processes and associated time frames and uncertainties pertaining to the long-term normal evolution of the near-field of a deep repository of the KBS-3V type. By normal evolution, reference is made to the description of the evolution of a disposal system subject to the assumption that all the spent fuel canisters remain intact following emplacement and repository closure for at least the design corrosion lifetime of the canister overpack in the expected repository conditions. In this connection, the effects of future glaciations, canister manufacturing defects and small-scale tectonic activity are briefly touched upon. The roles of the different repository near-field barriers in providing safety are briefly noted. The description of the normal evolution of a KBS-3V-type repository near-field conditions external to the unbreached spent fuel canister has been realized in terms of processes rather than timescales. Separate chapters have been devoted to the thermal, hydraulic, mechanical and chemical/biological evolution of the repository system, with due attention to some of the couplings between them. (orig.)

An experimental investigation of the interaction of primary and secondary stresses in fuel plates

International Nuclear Information System (INIS)

Swinson, W.F.; Battiste, R.L.; Yahr, G.T.

1996-01-01

If the load is not relieved as a structure starts to yield, the induced stress is defined as primary stress. If the load relaxes, as a structure begins yield the induced stress is defined as secondary stress. In design it is not uncommon to give more weight to primary stresses than to secondary stresses. However, knowing when this is good design practice and when it is not good design practice represents a problem. In particular, the fuel plates in operating reactors contain both primary stresses and secondary stresses and to properly assess a design there is a need to assign design weights to the stresses. Tests were conducted on reactor fuel plates intended for the Advanced Neutron Source (ANS) to determine the potential of giving different design weights to the primary and secondary stresses. The results of these tests and the conclusion that the stresses should be weighted the same are given in this paper

An experimental investigation on the interaction of primary and secondary stresses in fuel plates

International Nuclear Information System (INIS)

Swinson, W.F.; Battiste, R.L.; Yahr, G.T.

1997-01-01

If the load is not relieved as a structure starts to yield, the induced stress is defined as primary stress. If the load relaxes, as a structure begins to yield the induced stress is defined as secondary stress. In design, it is not uncommon to give more weight to primary stresses than to secondary stresses. However, knowing when this is good design practice and when it is not good design practice represent a problem. In particular, the fuel plates in operating reactors contain both primary stresses and secondary stresses, and to properly assess a design there is a need to assign design weights to the stresses. Tests were conducted on reactor fuel plates intended for the advanced neutron source (ANS) to determine the potential of giving different design weights to the primary and secondary stresses. The results of these tests and the conclusion that the stresses should be weighted the same are given in this paper

DIADEME: A computer code to assess in operation defective fuel characteristics and primary circuit contamination

Energy Technology Data Exchange (ETDEWEB)

Genin, J.B. [DEN/DEC/S3C, CEA Cadarache, 13 - Saint-Paul-lez-Durance (France); Harrer, A. [EdF/SEPTEN, 69 - Villeurbanne (France); Musante, Y. [FRAMATOME-ANP, 69 - Lyon (France)

2002-07-01

DIADEME is a computer code developed within the framework of R and D cooperation between the French Atomic Energy Commission (CEA), Electricite de France (EdF) and FRAMATOME-ANP. Its aim is to assess in operation defective fuel characteristics and primary circuit contamination for actinides and long half-life fission products involved in health physics problems as well as in waste and decommissioning studies. DIADEME has been developed and qualified for the EDF nuclear power plants. For many years, both theoretical and experimental studies have been carried out at the CEA on the release of fission products and actinides out of defective fuel rods in operation, their migration and deposition in PWR primary circuits. These studies have allowed defect characteristic diagnosis methods to be developed, based on radiochemical measurements of the primary coolant. These methods are generally used along with gamma spectrometry measurements on primary water sampling. In order to be completely efficient, these methods can also be used in connection with an on-line primary water gamma spectrometry device. This permits to obtain the most comprehensive data on fission product activity evolutions at steady state and during operation transients, and allows the on-line characterization of the defective fuel assemblies. For long half-life fission products and for actinides, DIADEME is also able to assess the activities of soluble and insoluble forms in the primary water and in the chemical and voluminal control system (CVCS) filters and resins, as well as those activities deposited on primary circuit surfaces. (author)

DIADEME: A computer code to assess in operation defective fuel characteristics and primary circuit contamination

International Nuclear Information System (INIS)

Genin, J.B.; Harrer, A.; Musante, Y.

2002-01-01

DIADEME is a computer code developed within the framework of R and D cooperation between the French Atomic Energy Commission (CEA), Electricite de France (EdF) and FRAMATOME-ANP. Its aim is to assess in operation defective fuel characteristics and primary circuit contamination for actinides and long half-life fission products involved in health physics problems as well as in waste and decommissioning studies. DIADEME has been developed and qualified for the EDF nuclear power plants. For many years, both theoretical and experimental studies have been carried out at the CEA on the release of fission products and actinides out of defective fuel rods in operation, their migration and deposition in PWR primary circuits. These studies have allowed defect characteristic diagnosis methods to be developed, based on radiochemical measurements of the primary coolant. These methods are generally used along with gamma spectrometry measurements on primary water sampling. In order to be completely efficient, these methods can also be used in connection with an on-line primary water gamma spectrometry device. This permits to obtain the most comprehensive data on fission product activity evolutions at steady state and during operation transients, and allows the on-line characterization of the defective fuel assemblies. For long half-life fission products and for actinides, DIADEME is also able to assess the activities of soluble and insoluble forms in the primary water and in the chemical and voluminal control system (CVCS) filters and resins, as well as those activities deposited on primary circuit surfaces. (author)

Cost update: Technology, safety, and costs of decommissioning a reference uranium fuel fabrication plant

International Nuclear Information System (INIS)

Miles, T.L.; Liu, Y.

1994-06-01

The cost estimates originally developed in NUREG/CR-1266 for commissioning a reference low-enrichment uranium fuel fabrication plant are updated from 1978 to early 1993 dollars. During this time, the costs for labor and materials increased approximately at the rate of inflation, the cost of energy increased more slowly than the rate of inflation, and the cost of low-level radioactive waste disposal increased much more rapidly than the rate of inflation. The results of the analysis indicate that the estimated costs for the immediate dismantlement and decontamination for unrestricted facility release (DECON) of the reference plant have increased from the mid-1978 value of $3.57 million to $8.08 million in 1993 with in-compact low-level radioactive waste disposal at the US Ecoloay facility near Richland, Washington. The cost estimate rises to $19.62 million with out-of-compact radioactive waste disposal at the Chem-Nuclear facility near Barnwell, South Carolina. A methodology and a formula are presented for estimating the cost of decommissioning the reference uranium fuel fabrication plant at some future time, based on these early 1993 cost estimates. The formula contains essentially the same elements as the formula given in 10 CFR 50.75 for escalating the decommissioning costs for nuclear power reactors to some future time

Simplified Reference Electrode for Electrorefining of Spent Nuclear Fuel in High Temperature Molten Salt

Energy Technology Data Exchange (ETDEWEB)

Kim Davies; Shelly X Li

2007-09-01

Pyrochemical processing plays an important role in development of proliferation- resistant nuclear fuel cycles. At the Idaho National Laboratory (INL), a pyrochemical process has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor II (EBR-II) in the last decade. Electrorefining in a high temperature molten salt is considered a signature or central technology in pyroprocessing fuel cycles. Separation of actinides from fission products is being demonstrated by electrorefining the spent fuel in a molten UCl3-LiCl-KCl electrolyte in two engineering scale electrorefiners (ERs). The electrorefining process is current controlled. The reference electrode provides process information through monitoring of the voltage difference between the reference and the anode and cathode electrodes. This information is essential for monitoring the reactions occurring at the electrodes, investigating separation efficiency, controlling the process rate, and determining the process end-point. The original reference electrode has provided good life expectancy and signal stability, but is not easily replaceable. The reference electrode used a vycor-glass ion-permeable membrane containing a high purity silver wire with one end positioned in ~2 grams of LiCl/KCl salt electrolyte with a low concentration (~1%) AgCl. It was, however, a complex assembly requiring specialized skill and talent to fabricate. The construction involved multiple small pieces, glass joints, ceramic to glass joints, and ceramic to metal joints all assembled in a high purity inert gas environment. As original electrodes reached end-of-life it was uncertain if the skills and knowledge were readily available to successfully fabricate replacements. Experimental work has been conducted to identify a simpler electrode design while retaining the needed long life and signal stability. This improved design, based on an ion-permeable membrane of mullite has been completed. Use of the silver wire

Simplified Reference Electrode for Electrorefining of Spent Nuclear Fuel in High Temperature Molten Salt

International Nuclear Information System (INIS)

Kim Davies; Shelly X Li

2007-01-01

Pyrochemical processing plays an important role in development of proliferation-resistant nuclear fuel cycles. At the Idaho National Laboratory (INL), a pyrochemical process has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor II (EBR-II) in the last decade. Electrorefining in a high temperature molten salt is considered a signature or central technology in pyroprocessing fuel cycles. Separation of actinides from fission products is being demonstrated by electrorefining the spent fuel in a molten UCl3-LiCl-KCl electrolyte in two engineering scale electrorefiners (ERs). The electrorefining process is current controlled. The reference electrode provides process information through monitoring of the voltage difference between the reference and the anode and cathode electrodes. This information is essential for monitoring the reactions occurring at the electrodes, investigating separation efficiency, controlling the process rate, and determining the process end-point. The original reference electrode has provided good life expectancy and signal stability, but is not easily replaceable. The reference electrode used a vycor-glass ion-permeable membrane containing a high purity silver wire with one end positioned in ∼2 grams of LiCl/KCl salt electrolyte with a low concentration (∼1%) AgCl. It was, however, a complex assembly requiring specialized skill and talent to fabricate. The construction involved multiple small pieces, glass joints, ceramic to glass joints, and ceramic to metal joints all assembled in a high purity inert gas environment. As original electrodes reached end-of-life it was uncertain if the skills and knowledge were readily available to successfully fabricate replacements. Experimental work has been conducted to identify a simpler electrode design while retaining the needed long life and signal stability. This improved design, based on an ion-permeable membrane of mullite has been completed. Use of the silver

Primary energy and greenhouse gas implications of increasing biomass production through forest fertilization

Energy Technology Data Exchange (ETDEWEB)

Sathre, Roger [Ecotechnology, Mid Sweden University, Ostersund (Sweden); Gustavsson, Leif [Ecotechnology, Mid Sweden University, Ostersund (Sweden); Bergh, Johan [Ecotechnology, Mid Sweden University, Ostersund (Sweden); Southern Swedish Forest Research Centre, Swedish University of Agricultural Sciences, Alnarp (Sweden)

2010-04-15

In this study we analyze the primary energy and greenhouse gas (GHG) implications of increasing biomass production by fertilizing 10% of Swedish forest land. We estimate the primary energy use and GHG emissions from forest management including production and application of N and NPK fertilizers. Based on modelled growth response, we then estimate the net primary energy and GHG benefits of using biomaterials and biofuels obtained from the increased forest biomass production. The results show an increased annual biomass harvest of 7.4 million t dry matter, of which 41% is large-diameter stemwood. About 6.9 PJ/year of additional primary energy input is needed for fertilizer production and forest management. Using the additional biomass for fuel and material substitution can reduce fossil primary energy use by 150 or 164 PJ/year if the reference fossil fuel is fossil gas or coal, respectively. About 22% of the reduced fossil energy use is due to material substitution and the remainder is due to fuel substitution. The net annual primary energy benefit corresponds to about 7% of Sweden's total primary energy use. The resulting annual net GHG emission reduction is 11.9 million or 18.1 million tCO{sub 2equiv} if the reference fossil fuel is fossil gas or coal, respectively, corresponding to 18% or 28% of the total Swedish GHG emissions in 2007. A significant one-time carbon stock increase also occurs in wood products and forest tree biomass. These results suggest that forest fertilization is an attractive option for increasing energy security and reducing net GHG emission.

Primary energy and greenhouse gas implications of increasing biomass production through forest fertilization

International Nuclear Information System (INIS)

Sathre, Roger; Gustavsson, Leif; Bergh, Johan

2010-01-01

In this study we analyze the primary energy and greenhouse gas (GHG) implications of increasing biomass production by fertilizing 10% of Swedish forest land. We estimate the primary energy use and GHG emissions from forest management including production and application of N and NPK fertilizers. Based on modelled growth response, we then estimate the net primary energy and GHG benefits of using biomaterials and biofuels obtained from the increased forest biomass production. The results show an increased annual biomass harvest of 7.4 million t dry matter, of which 41% is large-diameter stemwood. About 6.9 PJ/year of additional primary energy input is needed for fertilizer production and forest management. Using the additional biomass for fuel and material substitution can reduce fossil primary energy use by 150 or 164 PJ/year if the reference fossil fuel is fossil gas or coal, respectively. About 22% of the reduced fossil energy use is due to material substitution and the remainder is due to fuel substitution. The net annual primary energy benefit corresponds to about 7% of Sweden's total primary energy use. The resulting annual net GHG emission reduction is 11.9 million or 18.1 million tCO 2equiv if the reference fossil fuel is fossil gas or coal, respectively, corresponding to 18% or 28% of the total Swedish GHG emissions in 2007. A significant one-time carbon stock increase also occurs in wood products and forest tree biomass. These results suggest that forest fertilization is an attractive option for increasing energy security and reducing net GHG emission.

Primary energy and greenhouse gas implications of increasing biomass production through forest fertilization

Energy Technology Data Exchange (ETDEWEB)

Sathre, Roger; Gustavsson, Leif [Ecotechnology, Mid Sweden University, Oestersund (Sweden); Bergh, Johan [Ecotechnology, Mid Sweden University, Oestersund (Sweden); Southern Swedish Forest Research Centre, Swedish University of Agricultural Sciences, Alnarp (Sweden)

2010-04-15

In this study we analyze the primary energy and greenhouse gas (GHG) implications of increasing biomass production by fertilizing 10% of Swedish forest land. We estimate the primary energy use and GHG emissions from forest management including production and application of N and NPK fertilizers. Based on modelled growth response, we then estimate the net primary energy and GHG benefits of using biomaterials and biofuels obtained from the increased forest biomass production. The results show an increased annual biomass harvest of 7.4 million t dry matter, of which 41% is large-diameter stemwood. About 6.9 PJ/year of additional primary energy input is needed for fertilizer production and forest management. Using the additional biomass for fuel and material substitution can reduce fossil primary energy use by 150 or 164 PJ/year if the reference fossil fuel is fossil gas or coal, respectively. About 22% of the reduced fossil energy use is due to material substitution and the remainder is due to fuel substitution. The net annual primary energy benefit corresponds to about 7% of Sweden's total primary energy use. The resulting annual net GHG emission reduction is 11.9 million or 18.1 million tCO{sub 2equiv} if the reference fossil fuel is fossil gas or coal, respectively, corresponding to 18% or 28% of the total Swedish GHG emissions in 2007. A significant one-time carbon stock increase also occurs in wood products and forest tree biomass. These results suggest that forest fertilization is an attractive option for increasing energy security and reducing net GHG emission. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Hansen, U

1973-03-16

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

Neutronic analysis of a reference LEU core for Pakistan research reactor using oxide fuel

International Nuclear Information System (INIS)

Akhtar, K.M.; Qazi, M.K.; Bokhari, I.H.; Khan, L.A.; Pervez, S.

1988-07-01

Neutronic analysis of a 10 MW reference core for PARR, having 28 fresh LEU fuel elements arranged in a 6x5 configuration has been carried out using standard computer codes WIMS-D, EXTERMINATOR-II, and CITATION. Total nuclear power peaking of 3.2 has bee found to occur in the fuel plate adjacent to the water filled central flux trap at the depth of 43.8 cm from the top of the active core. Replacement of water in central flux trap with an aluminum block, having a 50 mm diameter water filled irradiation channel changes the flux profiles in fuel, core side flux trap and reflector. The thermal flux in the central flux trap decreases by about 53%. Therefore some of the fuel elements will have to be removed and the new configuration has to be analysed to determine the first operating core. However, after achieving some burn-up and confirmation from thermal hydraulic analysis, the core configuration analysed, will be the final working core. (orig./A.B.)

Dual Tank Fuel System

Science.gov (United States)

Wagner, Richard William; Burkhard, James Frank; Dauer, Kenneth John

1999-11-16

A dual tank fuel system has primary and secondary fuel tanks, with the primary tank including a filler pipe to receive fuel and a discharge line to deliver fuel to an engine, and with a balance pipe interconnecting the primary tank and the secondary tank. The balance pipe opens close to the bottom of each tank to direct fuel from the primary tank to the secondary tank as the primary tank is filled, and to direct fuel from the secondary tank to the primary tank as fuel is discharged from the primary tank through the discharge line. A vent line has branches connected to each tank to direct fuel vapor from the tanks as the tanks are filled, and to admit air to the tanks as fuel is delivered to the engine.

Thermal performance of annular-coated and sphere-pac LWR fuel rod designs

International Nuclear Information System (INIS)

Guenther, R.J.; Hsieh, K.A.; Barner, J.O.; Freshley, M.D.

1980-01-01

Two FCI-resistant UO 2 fuel rod designs are being compared to a reference design in irradiation tests in the Halden Boiling Water Reactor (HBWR) as part of the DOE-sponsored Fuel Performance Improvement Program (FPIP). The primary fuel design (annular-coated-pressurized) incorporates annular pellets, a graphite coating on the inner surface of the Zircaloy cladding, and pressurized helium fill gas. Also being investigated is an 87% smear density sphere-pac design with pressurized helium fill gas. The solid pellet (reference) and annular-coated designs described had helium fill gas at approx. 100 kPa and the sphere-pac rods were pressurized at approx. 455 kPa

Standard Test Method for Calibration of Primary Non-Concentrator Terrestrial Photovoltaic Reference Cells Using a Tabular Spectrum

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This test method is intended to be used for calibration and characterization of primary terrestrial photovoltaic reference cells to a desired reference spectral irradiance distribution, such as Tables G173. The recommended physical requirements for these reference cells are described in Specification E1040. Reference cells are principally used in the determination of the electrical performance of photovoltaic devices. 1.2 Primary photovoltaic reference cells are calibrated in natural sunlight using the relative spectral response of the cell, the relative spectral distribution of the sunlight, and a tabulated reference spectral irradiance distribution. 1.3 This test method requires the use of a pyrheliometer that is calibrated according to Test Method E816, which requires the use of a pyrheliometer that is traceable to the World Radiometric Reference (WRR). Therefore, reference cells calibrated according to this test method are traceable to the WRR. 1.4 This test method is a technique that may be used ...

Pressurised light water reactor failed fuel monitoring. Comparison of international strategies used in the surveillance of the primary barrier

International Nuclear Information System (INIS)

Aldworth, Robin; Dinse, Christelle; Mole, Didier

2014-01-01

As part of its focus on continuous improvement, the nuclear industry has devoted considerable effort to the prevention, monitoring, characterisation and impact mitigation of fuel failures. Failed fuel and primary circuit activity monitoring strategies have thus evolved over time, with utilities changing their strategies as a function of operational experience, regulator encouragement and improvements to scientific understanding. Differences among utility fuel monitoring strategies are therefore to be expected. This paper compares and analyses a selection of failed fuel and primary circuit activity monitoring strategies employed at French, American, South-African and Belgian pressurised water reactors (PWRs). These strategies, which were received in response to a questionnaire sent by Electricité de France, are discussed with a number of potential best practices and areas for future research and development being provided. Finally, and so as to ensure the utility of this paper with regards to the development, or evolution of, failed fuel monitoring strategies, this paper has been split into four sections representing the primary radiochemistry indicator groups: gross, calculated, fission products and activated corrosion products. (author)

Simulation of primary fuel atomization processes at subcritical pressures.

Energy Technology Data Exchange (ETDEWEB)

Arienti, Marco

2013-06-01

This report documents results from an LDRD project for the first-principles simulation of the early stages of spray formation (primary atomization). The first part describes a Cartesian embedded-wall method for the calculation of flow internal to a real injector in a fully coupled primary calculation. The second part describes the extension to an all-velocity formulation by introducing a momentum-conservative semi-Lagrangian advection and by adding a compressible term in the Poissons equation. Accompanying the description of the new algorithms are verification tests for simple two-phase problems in the presence of a solid interface; a validation study for a scaled-up multi-hole Diesel injector; and demonstration calculations for the closing and opening transients of a single-hole injector and for the high-pressure injection of liquid fuel at supersonic velocity.

Thermal Expansion and Density Data of UO2 and Simulated Fuel for Standard Reference

International Nuclear Information System (INIS)

Yang, Jae Hwan; Na, S. H.; Lee, J. W.; Kang, K. H.

2010-01-01

Standard Reference Data (SRD) is the scientific, technical data whose reliability and accuracy are evaluated by scientist group. Since SRD has a great impact on the improvement of national competitiveness by stirring up technological innovation in every sector of industries, many countries are making great efforts on establishing SRD in various areas. Data center for nuclear fuel material in Korea Atomic Energy Research Institute plays a role to providing property data of nuclear fuel material at high temperature, pressure, and radiation which are essential for the safety evaluation of nuclear power. In this study, standardization of data on thermal expansion and density of UO 2 were carried out in the temperature range from 300 K to 3100 K via uncertainty evaluation of indirectly produced data. Besides, standardization of data on thermal expansion and density of simulated fuel were also done in the temperature range from 350 K to 1750 K via uncertainty evaluation of directly produced data

Methodology for comparing the health effects of electricity generation from uranium and coal fuels

International Nuclear Information System (INIS)

Rhyne, W.R.; El-Bassioni, A.A.

1981-01-01

A methodology was developed for comparing the health risks of electricity generation from uranium and coal fuels. The health effects attributable to the construction, operation, and decommissioning of each facility in the two fuel cycle were considered. The methodology is based on defining (1) requirement variables for the materials, energy, etc., (2) effluent variables associated with the requirement variables as well as with the fuel cycle facility operation, and (3) health impact variables for effluents and accidents. The materials, energy, etc., required for construction, operation, and decommissioning of each fuel cycle facility are defined as primary variables. The materials, energy, etc., needed to produce the primary variable are defined as secondary requirement variables. Each requirement variable (primary, secondary, etc.) has associated effluent variables and health impact variables. A diverging chain or tree is formed for each primary variable. Fortunately, most elements reoccur frequently to reduce the level of analysis complexity. 6 references, 11 figures, 6 tables

Primary water chemistry optimization for extended fuel cycle operation. Results of the 'Duo experimentation' after three cycles

Energy Technology Data Exchange (ETDEWEB)

Viricel, L.; Andrieu, C.; Segura, J.C.; Rocher, A. [Electricite de France (France); Thomazet, J.; Clinard, M.H. [Framatome ANP (France); Dacquait, F. [Commissariat a l' Energie Atomique (France)

2002-07-01

The primary coolant conditioning in French nuclear power plants is essentially based on the boron-lithium coordinated chemistry, with a target pH of 7.2 at 300 C and a maximum lithium concentration of 2.2 mg/kg. In 1996, EDF 1300 MWe units began operating 18-month fuel cycles, increasing boron concentrations at the beginning of the cycles. Since today the maximum lithium concentration in normal operation is 2.2 mg/kg, extended cycle operation results in a decrease in the pH at the beginning of the cycles, which may possibly lead to deposits in RCS, and particularly on the fuel cladding, and increased dose rates. It has to be noted that today, the fuel assemblies maximum burnup is set at 52 GWd/tU. One solution is to adjust the pH by increasing the lithium content at the beginning of the cycles, which is easy to implement and does not require any modification on the units. Hence, EDF is testing a ''modified'' chemistry regime in the << Duo experimentation >> during 4 fuel cycles, with a maximum authorized lithium content of 3.5 mg/kg at the beginning of the cycles in the Cattenom 2 pilot unit. The Golfech 1 reference unit implements a standard boron-lithium coordination pH{sub 300} 7.2. The major goal of the experimentation is to assess the impact of elevated lithium concentrations at the beginning of the cycles on fuel cladding oxide behavior, mass transport and dose rates. This paper presents the results of the first three cycles of the Duo experimentation. (author)

An overview of REARA reference design

International Nuclear Information System (INIS)

Menezes, A.

1996-01-01

In an effort to keep fast reactors as an option for an eventual future utilization in Brazil, the Advanced Studies Institute (IEAv) is coordinating a research project (called REARA, from the Portuguese REAtores RApidos (Fast Reactors)) which aims to establish a basic know-how in important aspects of fast reactor technology. A reference design for the primary circuit of a 60 MWt experimental fast reactor has been completed. Fuel pin dimensions and other data were taken mostly from PRISM and EBR-II and were used for calculations which lead to a general core configuration. The reference design is presently being used for testing our calculational tools, for checking different methodologies for core calculations, etc. The REARA reference design is presented, with emphasis in the main results and also in the contribution given by the project team in areas such as neutronics, thermohydraulics, etc. (author). 6 refs, 1 fig., 7 tabs

Relating the octane numbers of fuels to ignition delay times measured in an ignition quality tester (IQT)

KAUST Repository

Naser, Nimal

2016-09-21

A methodology for estimating the octane index (OI), the research octane number (RON) and the motor octane number (MON) using ignition delay times from a constant volume combustion chamber with liquid fuel injection is proposed by adopting an ignition quality tester. A baseline data of ignition delay times were determined using an ignition quality tester at a charge pressure of 21.3 bar between 770 and 850 K and an equivalence ratio of 0.7 for various primary reference fuels (PRFs, mixtures of isooctane and n-heptane). Our methodology was developed using ignition delay times for toluene reference fuels (mixtures of toluene and n-heptane). A correlation between the OI and the ignition delay time at the initial charge temperature enabled the OI of non-PRFs to be predicted at specified temperatures. The methodology was validated using ignition delay times for toluene primary reference fuels (ternary mixtures of toluene, iso-octane, and n-heptane), fuels for advanced combustion engines (FACE) gasolines, and certification gasolines. Using this methodology, the RON, the MON, and the octane sensitivity were estimated in agreement with values obtained from standard test methods. A correlation between derived cetane number and RON is also provided. (C) 2016 Elsevier Ltd. All rights reserved.

Opportunities for PEM fuel cell commercialization : fuel cell electric vehicle demonstration in Shanghai

Energy Technology Data Exchange (ETDEWEB)

Ma, Z.F. [Shanghai Jiao Tong Univ., Shanghai (China). Dept. of Chemical Engineering

2006-07-01

The research and development activities devoted to the development of the proton exchange membrane fuel cell (PEMFC) were discussed with reference to its application in the fuel cell electric vehicle (FCEV). In the past decade, PEMFC technology has been successfully applied in both the automobile and residential sector worldwide. In China, more than one billion RMB yuan has been granted by the Chinese government to develop PEM fuel cell technology over the past 5 years, particularly for commercialization of the fuel cell electric vehicle (FCEV). The City of Shanghai has played a significant role in the FCEV demonstration with involvement by Shanghai Auto Industrial Company (SAIC), Tongji University, Shanghai Jiaotong University, and Shanghai Shenli High Tech Co. Ltd. These participants were involved in the development and integration of the following components into the FCEV: fuel cell engines, batteries, FCEV electric control systems, and primary materials for the fuel cell stack. During the course of the next five year-plan (2006-2010), Shanghai will promote the commercialization of FCEV. More than one thousand FCEVs will be manufactured and an FCEV fleet will be in operation throughout Shanghai City by 2010.

A comparative study of the oxidation characteristics of two gasoline fuels and an n-heptane/iso-octane surrogate mixture

KAUST Repository

Javed, Tamour; Nasir, Ehson F.; Es-sebbar, Et-touhami; Farooq, Aamir

2015-01-01

Ignition delay times and CO, H2O, OH and CO2 time-histories were measured behind reflected shock waves for two FACE (Fuels for Advanced Combustion Engines) gasolines and one PRF (Primary Reference Fuel) blend. The FACE gasolines chosen for this work

Reported characteristics of children referred from primary care to pediatric allergy specialist care

NARCIS (Netherlands)

Brakel, T.M.; Flokstra-De Blok, B.M.J.; Oude Elberink, J.N.G.; Schuttelaar, M.L.A.; Christoffers, W.A.; Roerdink, E.M.; Molen, van der Thys; Dubois, A.E.J.

2013-01-01

Background: For the majority of patients suspected of allergy, management is mostly the responsibility of general practitioners (GPs), but may include specialist referral in selected cases. The purpose of this study is to obtain insight in characteristics of children that are referred from primary

Ethanol Fuels Reference Guide: A Decision-Makers Guide to Ethanol Fuels

Energy Technology Data Exchange (ETDEWEB)

1982-10-01

This guide is a compendium of information on alcohol fuel production and use. Chapter titles are: facts about ethanol; gasohol-answers to the basic questions; feedstocks and their coproducts; ethanol production processes; and vehicle fuel use and performance. In addition, there are 8 appendices which include fermentation guides for common grains and potatoes, component and enzyme manufacturers, and information on regulations and permits. (DMC)

Costs, CO{sub 2}- and primary energy balances of forest-fuel recovery systems at different forest productivity

Energy Technology Data Exchange (ETDEWEB)

Eriksson, Lisa; Gustavsson, Leif [Ecotechnology, Department of Engineering and Sustainable Development, Mid Sweden University, SE-831 25 Oestersund (Sweden)

2010-05-15

Here we examine the cost, primary energy use, and net carbon emissions associated with removal and use of forest residues for energy, considering different recovery systems, terrain, forwarding distance and forest productivity. We show the potential recovery of forest fuel for Sweden, its costs and net carbon emissions from primary energy use and avoided fossil carbon emissions. The potential annual net recovery of forest fuel is about 66 TWh, which would cost one billion EUR{sub 2005} to recover and would reduce fossil emissions by 6.9 Mt carbon if coal were replaced. Of the forest fuel, 56% is situated in normal terrain with productivity of >30 t dry-matter ha{sup -1} and of this, 65% has a forwarding distance of <400 m. In normal terrain with >30 t dry-matter ha{sup -1} the cost increase for the recovery of forest fuel, excluding stumps, is around 4-6% and 8-11% for medium and longer forwarding distances, respectively. The stump and small roundwood systems are less cost-effective at lower forest fuel intensity per area. For systems where loose material is forwarded, less dry-matter per hectare increases costs by 6-7%, while a difficult terrain increases costs by 3-4%. Still, these systems are quite cost-effective. The cost of spreading ash is around 40 EUR{sub 2005} ha{sup -1}, while primary energy use for spreading ash in areas where logging residues, stumps, and small roundwood are recovered is about 0.025% of the recovered bioenergy. (author)

An application of indirect model reference adaptive control to a low-power proton exchange membrane fuel cell

Science.gov (United States)

Yang, Yee-Pien; Liu, Zhao-Wei; Wang, Fu-Cheng

2008-05-01

Nonlinearity and the time-varying dynamics of fuel cell systems make it complex to design a controller for improving output performance. This paper introduces an application of a model reference adaptive control to a low-power proton exchange membrane (PEM) fuel cell system, which consists of three main components: a fuel cell stack, an air pump to supply air, and a solenoid valve to adjust hydrogen flow. From the system perspective, the dynamic model of the PEM fuel cell stack can be expressed as a multivariable configuration of two inputs, hydrogen and air-flow rates, and two outputs, cell voltage and current. The corresponding transfer functions can be identified off-line to describe the linearized dynamics with a finite order at a certain operating point, and are written in a discrete-time auto-regressive moving-average model for on-line estimation of parameters. This provides a strategy of regulating the voltage and current of the fuel cell by adaptively adjusting the flow rates of air and hydrogen. Experiments show that the proposed adaptive controller is robust to the variation of fuel cell system dynamics and power request. Additionally, it helps decrease fuel consumption and relieves the DC/DC converter in regulating the fluctuating cell voltage.

ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

International Nuclear Information System (INIS)

Wright, A.L.; Kress, T.S.; Smith, A.M.

1980-01-01

This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

Fuel Handbook[Wood and other renewable fuels

Energy Technology Data Exchange (ETDEWEB)

Stroemberg, Birgitta [TPS Termiska Processer AB, Nykoeping (SE)] (ed.)

2006-03-15

This handbook on renewable fuels is intended for power and heat producers in Sweden. This fuel handbook provides, from a plant owner's perspective, a method to evaluate different fuels on the market. The fuel handbook concerns renewable fuels (but does not include household waste) that are available on the Swedish market today or fuels that have potential to be available within the next ten years. The handbook covers 26 different fuels. Analysis data, special properties, operating experiences and literature references are outlined for each fuel. [Special properties, operating experiences and literature references are not included in this English version] The handbook also contains: A proposed methodology for introduction of new fuels. A recommendation of analyses and tests to perform in order to reduce the risk of problems is presented. [The recommendation of analyses and tests is not included in the English version] A summary of relevant laws and taxes for energy production, with references to relevant documentation. [Only laws and taxes regarding EU are included] Theory and background to evaluate a fuel with respect to combustion, ash and corrosion properties and methods that can be used for such evaluations. Summary of standards, databases and handbooks on biomass fuels and other solid fuels, and links to web sites where further information about the fuels can be found. The appendices includes: A methodology for trial firing of fuels. Calculations procedures for, amongst others, heating value, flue gas composition, key number and free fall velocity [Free fall velocity is not included in the English version]. In addition, conversion routines between different units for a number of different applications are provided. Fuel analyses are presented in the appendix. (The report is a translation of parts of the report VARMEFORSK--911 published in 2005)

Calculating analysis of firing different composition artificial coal liquid fuels (ACLF) in the cyclone primary furnace

Energy Technology Data Exchange (ETDEWEB)

Tsepenok, A. [Novosibirsk State Technological Univ. (Russian Federation); Joint Stock company ' ' ZiO-COTES' ' , Novosibirsk (Russian Federation); Ovchinnikov, Yu. [Novosibirsk State Technological Univ. (Russian Federation); Serant, F. [Joint Stock company ' ' ZiO-COTES' ' , Novosibirsk (Russian Federation)

2013-07-01

This chapter describes the preparation technologies, results of computer simulation of combustion processes in a cyclone primary furnace during firing of artificial coal liquid fuels prepared from different coal grades and results of live testing. As a result the values of unburned carbon, NO{sub x} emissions and other concentrations in the outlet section primary furnace were estimated.

Revised Analyses of Decommissioning Reference Non-Fuel-Cycle Facilities

International Nuclear Information System (INIS)

Bierschbach, M.C.; Haffner, D.R.; Schneider, K.J.; Short, S.M.

2002-01-01

Cost information is developed for the conceptual decommissioning of non-fuel-cycle nuclear facilities that represent a significant decommissioning task in terms of decontamination and disposal activities. This study is a re-evaluation of the original study (NUREG/CR-1754 and NUREG/CR-1754, Addendum 1). The reference facilities examined in this study are the same as in the original study and include: a laboratory for the manufacture of 3 H-labeled compounds; a laboratory for the manufacture of 14 C-labeled compounds; a laboratory for the manufacture of 123 I-labeled compounds; a laboratory for the manufacture of 137 Cs sealed sources; a laboratory for the manufacture of 241 Am sealed sources; and an institutional user laboratory. In addition to the laboratories, three reference sites that require some decommissioning effort were also examined. These sites are: (1) a site with a contaminated drain line and hold-up tank; (2) a site with a contaminated ground surface; and (3) a tailings pile containing uranium and thorium residues. Decommissioning of these reference facilities and sites can be accomplished using techniques and equipment that are in common industrial use. Essentially the same technology assumed in the original study is used in this study. For the reference laboratory-type facilities, the study approach is to first evaluate the decommissioning of individual components (e.g., fume hoods, glove boxes, and building surfaces) that are common to many laboratory facilities. The information obtained from analyzing the individual components of each facility are then used to determine the cost, manpower requirements and dose information for the decommissioning of the entire facility. DECON, the objective of the 1988 Rulemaking for materials facilities, is the decommissioning alternative evaluated for the reference laboratories because it results in the release of the facility for restricted or unrestricted use as soon as possible. For a facility, DECON requires

Revised Analyses of Decommissioning Reference Non-Fuel-Cycle Facilities

Energy Technology Data Exchange (ETDEWEB)

MC Bierschbach; DR Haffner; KJ Schneider; SM Short

2002-12-01

Cost information is developed for the conceptual decommissioning of non-fuel-cycle nuclear facilities that represent a significant decommissioning task in terms of decontamination and disposal activities. This study is a re-evaluation of the original study (NUREG/CR-1754 and NUREG/CR-1754, Addendum 1). The reference facilities examined in this study are the same as in the original study and include: a laboratory for the manufacture of {sup 3}H-labeled compounds; a laboratory for the manufacture of {sup 14}C-labeled compounds; a laboratory for the manufacture of {sup 123}I-labeled compounds; a laboratory for the manufacture of {sup 137}Cs sealed sources; a laboratory for the manufacture of {sup 241}Am sealed sources; and an institutional user laboratory. In addition to the laboratories, three reference sites that require some decommissioning effort were also examined. These sites are: (1) a site with a contaminated drain line and hold-up tank; (2) a site with a contaminated ground surface; and (3) a tailings pile containing uranium and thorium residues. Decommissioning of these reference facilities and sites can be accomplished using techniques and equipment that are in common industrial use. Essentially the same technology assumed in the original study is used in this study. For the reference laboratory-type facilities, the study approach is to first evaluate the decommissioning of individual components (e.g., fume hoods, glove boxes, and building surfaces) that are common to many laboratory facilities. The information obtained from analyzing the individual components of each facility are then used to determine the cost, manpower requirements and dose information for the decommissioning of the entire facility. DECON, the objective of the 1988 Rulemaking for materials facilities, is the decommissioning alternative evaluated for the reference laboratories because it results in the release of the facility for restricted or unrestricted use as soon as possible. For a

Electricity generation from fermented primary sludge using single-chamber air-cathode microbial fuel cells

KAUST Repository

Yang, Fei; Ren, Lijiao; Pu, Yuepu; Logan, Bruce E.

2013-01-01

Single-chamber air-cathode microbial fuel cells (MFCs) were used to generate electricity from fermented primary sludge. Fermentation (30°C, 9days) decreased total suspended solids (26.1-16.5g/L), volatile suspended solids (24.1-15.3g/L) and pH (5

LPG diesel dual fuel engine – A critical review

Directory of Open Access Journals (Sweden)

B. Ashok

2015-06-01

Full Text Available The engine, which uses both conventional diesel fuel and LPG fuel, is referred to as ‘LPG–diesel dual fuel engines’. LPG dual fuel engines are modified diesel engines which use primary fuel as LPG and secondary fuel as diesel. LPG dual fuel engines have a good thermal efficiency at high output but the performance is less during part load conditions due to the poor utilization of charges. This problem can be overcome by varying factors such as pilot fuel quantity, injection timing, composition of the gaseous fuel and intake charge conditions, for improving the performance, combustion and emissions of dual fuel engines. This article reviews about the research work done by the researchers in order to improve the performance, combustion and emission parameters of a LPG–diesel dual fuel engines. From the studies it is shown that the use of LPG in diesel engine is one of the capable methods to reduce the PM and NOx emissions but at same time at part load condition there is a drop in efficiency and power output with respect to diesel operation.

Genetics Home Reference: primary spontaneous pneumothorax

Science.gov (United States)

... Home Health Conditions Primary spontaneous pneumothorax Primary spontaneous pneumothorax Printable PDF Open All Close All Enable Javascript ... view the expand/collapse boxes. Description Primary spontaneous pneumothorax is an abnormal accumulation of air in the ...

Pilot-benchmarking of the WENRA safety reference levels for the spent fuel intermediate storage facility Ahaus

International Nuclear Information System (INIS)

Lorenz, Bernd; Roeder, Markus; Brandt, Klaus-Dieter

2008-01-01

Full text: The Western European Nuclear Regulator's Association (WENRA) has 2007 issued the draft of the 'Waste and Spent Fuel Storage Safety Reference Levels'. The objective of WENRA is to strive for a harmonized safety level of nuclear facilities within the European Community and these Reference Levels are a benchmark method to demonstrate the achieved level for the regulatory system and the implementation as well. Safety Reference Levels exist at the moment for Reactor Safety, Waste Storage and Decommissioning in different stages of development. ENISS, the European Nuclear Installations Safety Standards Initiative, a FORATOM based special organisation of nuclear operators, has discussed these Safety Reference Levels very intensively with WENRA and the agreement was to make a implementation benchmark-exercise for the storage facilities before the authorities finally agree on the Reference Levels. This benchmark was scheduled for the year 2008. Because of the special situation in Germany where a large number of storage facilities is in operation the German authorities felt that it would be useful to initiate a Pilot-Benchmark to get first results on the feasibility of the Reference Levels and the burden imposed to authorities and operators by these benchmark-exercises. GNS, a subsidiary company of the utilities, agreed to step into this process on a voluntary basis with its storage facility for spent fuel in Ahaus. The exercise was done in a very efficient way and in good co-operation between the authorities, local and federal, and the operator. The results in terms of safety assessments have been very satisfactory showing the high degree of safety. Although the facility was for the first time licensed already in 1987 the compliance with nearly all Reference Levels from 2007 could be demonstrated. It became also clear that newer facilities would fulfil the desired safety standard too. Nevertheless, in spite of the good results the exercise revealed some weak

Development of a reference spent fuel library of 17x17 PWR fuel assemblies

International Nuclear Information System (INIS)

Rossa, Riccardo; Borella, Alessandro; Van der Meer, Klaas

2013-01-01

One of the most common ways to investigate new Non-Destructive Assays (NDA) for the spent fuel assemblies are Monte Carlo simulations. In order to build realistic models the user must define in an accurate way the material compositions and the source terms in the system. This information can be obtained using burnup codes such as ORIGEN-ARP and ALEPH2.2, developed at SCK-CEN. These software applications allow the user to select the irradiation history of the fuel assembly and to calculate the corresponding isotopic composition and neutron/gamma emissions as a function of time. In the framework of the development of an innovative NDA for spent fuel verifications, SCK•CEN built an extensive fuel library for 17x17 PWR assemblies, using both ORIGEN-ARP and ALEPH2.2. The parameters considered in the calculations were initial enrichment, discharge burnup, and cooling time. The combination of these variables allows to obtain more than 1500 test cases. Considering the broad range of the parameters, the fuel library can be used for other purposes apart from spent fuel verifications, for instance for the direct disposal in geological repositories. In addition to the isotopic composition of the spent fuel, the neutron and photon emissions were also calculated and compared between the two codes. The comparison of the isotopic composition showed a good agreement between the codes for most of the relevant isotopes in the spent fuel. However, specific isotopes as well as neutron and gamma spectra still need to be investigated in detail.

Is drug treatment for dementia followed up in primary care? A Swedish study of dementia clinics and referring primary care centres.

Directory of Open Access Journals (Sweden)

Lars Sonde

Full Text Available PURPOSE: It is largely unknown how the medical treatment of patients diagnosed with dementia is followed up in primary care. Therefore, we studied patient medical records from two dementia clinics and from the referring primary care centres. METHODS: A retrospective study of 241 patients was conducted from April to October 2011 in north west Stockholm, Sweden. Over half (51.5% of the patients had Alzheimer's disease (AD, the remainder had mixed AD/vascular dementia (VaD. Eighty-four medical reports from primary care (35% of the study group were analysed at follow-up 18 months after diagnosis. RESULTS: All four dementia drugs available on the Swedish market (three cholinesterase inhibitors [donepezil, rivastigmine and galantamine] and memantine were prescribed at the two dementia clinics. The most commonly used dementia drug was galantamine. There were differences between the two dementia clinics in preference and combination of drugs and of treatment given to male and female patients. At follow-up, 84% were still on dementia medication. Drug use was followed up by the general practitioners (GPs in two-thirds of the cases. Eighteen per cent of the GPs' medical records made no reference to the patient's dementia or treatment even though dementia drugs were included in the list of medications prescribed. CONCLUSIONS: The results indicate that the Swedish guidelines for treatment of cognitive symptoms in AD are being followed in primary care. However, documentation of follow-up of drug treatment was sometimes insufficient, which calls for development of guidelines for complete medical records and medication lists.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-02-01

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

Development of the Korean Reference Vertical Disposal System Concept for Spent Fuels

International Nuclear Information System (INIS)

Lee, J.Y.; Cho, D.K.; Kim, S.G.; Choi, H.J.; Choi, J.W.; Hahn, P.S.

2006-01-01

The development of a deep geologic disposal system for the spent fuel from nuclear power plants has been carried out since this program was launched at 1997 in Korea. In ' this paper, a pre-conceptual design of the Korean Reference HLW Vertical disposal System (KRS-V1) is presented. Though no site for the underground repository has yet been specified in Korea, a generic site with granitic rock is considered for reference HLW repository design. Depth of the repository is assumed to be 500 meters. The repository consists of the disposal area, technical rooms with four shafts to connect them to the ground level in the controlled area and technical rooms with an access tunnel and three shafts to connect them to the ground level in the uncontrolled area. Disposal area consists of disposal tunnels, panel tunnels and a central tunnel. The repository will be excavated, operated and backfilled in several phases including an Underground Research Laboratory (URL) phase. The result of this preliminary conceptual design will be used for an evaluation of the feasibility, analyses of the long term safety, information for public communication and a cost estimation etc. (authors)

Long-term exposure to jet fuel. An investigation on occupationally exposed workers with special reference to the nervous system

Energy Technology Data Exchange (ETDEWEB)

Knave, B; Persson, H E; Goldberg, M; Westerholm, P

1976-09-01

Chronic solvent intoxication due to jet fuel vapor was diagnosed after tests on 29 aircraft factory workers. The exposed subjects were classified into (1) heavily exposed and (2) less heavily exposed groups. Tests included a clinical neurological examination, measurements of the peipheral nerve conduction velocities and threshold determination of vibratory sensations in the extremities. All 13 examined in the first group and 7 of 16 in the second group stated that, on repeated occasions, they had experienced acute effects (dizziness, dyspnoea, heart palpitations, chest striction, nausea, headaches) of the jet fuel vapors in the inhaled air. Air overrepresentation of symptoms of neurasthenia and physical disturbances, polyneuropathic symptoms and plyneuropathic findings were evident from the neurological status both in the heavily exposed group and in the two groups taken as a whole when comparison was made with reference groups. Moreover, relatively lower nerve conduction velocities and higher vibration thresholds were witnessed in both groups compared with a reference group from the heavy metal industry.

Integral benchmarks with reference to thorium fuel cycle

International Nuclear Information System (INIS)

Ganesan, S.

2003-01-01

This is a power point presentation about the Indian participation in the CRP 'Evaluated Data for the Thorium-Uranium fuel cycle'. The plans and scope of the Indian participation are to provide selected integral experimental benchmarks for nuclear data validation, including Indian Thorium burn up benchmarks, post-irradiation examination studies, comparison of basic evaluated data files and analysis of selected benchmarks for Th-U fuel cycle

Genetics Home Reference: primary ciliary dyskinesia

Science.gov (United States)

... primary ciliary dyskinesia also have year-round nasal congestion and a chronic cough. Chronic respiratory tract infections ... likely due to abnormal cilia in the fallopian tubes. Another feature of primary ciliary dyskinesia is recurrent ...

Technology, safety and costs of decommissioning a reference small mixed oxide fuel fabrication plant. Volume 1. Main report

Energy Technology Data Exchange (ETDEWEB)

Jenkins, C. E.; Murphy, E. S.; Schneider, K J

1979-01-01

Detailed technology, safety and cost information are presented for the conceptual decommissioning of a reference small mixed oxide fuel fabrication plant. Alternate methods of decommissioning are described including immediate dismantlement, safe storage for a period of time followed by dismantlement and entombment. Safety analyses, both occupational and public, and cost evaluations were conducted for each mode.

The Nuclear Fuel Cycle

International Nuclear Information System (INIS)

2011-08-01

This brochure describes the nuclear fuel cycle, which is an industrial process involving various activities to produce electricity from uranium in nuclear power reactors. The cycle starts with the mining of uranium and ends with the disposal of nuclear waste. The raw material for today's nuclear fuel is uranium. It must be processed through a series of steps to produce an efficient fuel for generating electricity. Used fuel also needs to be taken care of for reuse and disposal. The nuclear fuel cycle includes the 'front end', i.e. preparation of the fuel, the 'service period' in which fuel is used during reactor operation to generate electricity, and the 'back end', i.e. the safe management of spent nuclear fuel including reprocessing and reuse and disposal. If spent fuel is not reprocessed, the fuel cycle is referred to as an 'open' or 'once-through' fuel cycle; if spent fuel is reprocessed, and partly reused, it is referred to as a 'closed' nuclear fuel cycle.

Benchmark reference data on post irradiation analysis of light water reactor fuel samples

International Nuclear Information System (INIS)

Guardini, S.; Guzzi, G.

1983-01-01

The structure of the present report is as follows: in section I the benchmark activity (BM) is described in detail; characteristics of the reactors and fuel assemblies examinated are given, and the technical aspects of the chemical and analytical processes are discussed. In section II all the techniques used to certify the analytical data are presented, together with a discussion of evaluated random and systematic uncertainties. A comparison with the calculated values and the interpretation with ICT (Isotopic Correlation Techniques) is also presented in this section. Section III presents the results. In practice the complete sets of results referring to all JRC measurements are given here for the sake of the completeness and consistency of this final report

Economic analysis of grid and wire wrap supported hydride and oxide fueled pressurized water reactors

International Nuclear Information System (INIS)

Shuffler, C.; Diller, P.; Malen, J.; Todreas, N.; Greenspan, E.; Petrovic, B.

2009-01-01

An economic analysis is performed to calculate the levelized unit cost of electricity (COE) for a pressurized water reactor (PWR) retrofitted with a range of potential U (45 wt.%)-ZrH 1.6 hydride and UO 2 oxide fueled geometries (i.e., combinations of rod diameter and pitch) supported by traditional grid spacers (square array) and wire wrap spacers (hexagonal array). The time frame considered in computing the COE is the remaining plant life, beginning at the time of retrofit. The goals of the analysis are twofold: (1) comparing the economic performance of UO 2 and U-ZrH 1.6 fuels for a range of retrofitted geometries supported by grid and wire wrap spacers; and (2) investigating the potential economic benefits for nuclear utilities considering retrofitting new fuels and/or geometries into existing PWR pressure vessels. Fuel cycle, operations and maintenance (O and M), and capital costs are considered. The economic performance of U-ZrH 1.6 and UO 2 fuels is found to be similar, with UO 2 fueled designs providing a slight advantage when supported by grid spacers, and U-ZrH 1.6 providing a slight advantage when supported by wire wrap spacers. These small differences in cost, however, are within the bounds of uncertainty of this study and are not believed to provide a strong economic argument for the use of one fuel type over the other. To demonstrate the potential economic benefits of retrofitted designs to nuclear utilities, two different comparisons are made. The first compares the COE for retrofitted designs with the COE for a reference PWR, assumed to have operated long enough to recuperate its initial capital investment. The costs for this reference PWR reflect the 'do-nothing' case for current plant owners whose primary expenditures are fuel cycle and O and M costs. The second comparison introduces a different reference PWR that includes the costs to operate an existing unit and the cost to purchase power from a newly constructed PWR, for comparison with

Energy in France. References

International Nuclear Information System (INIS)

2000-01-01

This small booklet is a compilation of key data about the energy in France: energy and economy (energy industries and gross internal product, employment, investments), overall energies (primary energy production and consumption, sectoral consumption, energy bill, price of imported crude oil), petroleum (primary production, sectoral consumption of refined petroleum products, automotive fuels demand, import and export of petroleum products), natural gas (production, sectoral consumption, imports per country of origin), coal (production, sectoral consumption, imports), electric power (production per origin, classical thermal production per type of fuel, sectoral consumption), renewable energies (overall production, heat networks supplied with non-conventional energy sources, wood consumption, wind power production, solar thermal and photovoltaic production), rational use of energy (primary energy intensity, cumulated energy saving), energy prices (in industries and households, automotive fuel prices, energy consumptions in households), energy and environment (CO 2 emissions). A synthesis of the main energy tariffs and prices is given in a separate folder. (J.S.)

WWER expert system for fuel failure analysis using the RTOP-CA code

International Nuclear Information System (INIS)

Likhanskii, V.; Evdokimov, I.; Sorokin, A.; Khromov, A.; Kanukova, V.; Apollonova, O.; Ugryumov, A.

2008-01-01

The computer expert system for fuel failure analysis of WWER during operation is presented. The diagnostics is based on the measurement of specific activity of reference nuclides in reactor primary coolant and application of a computer code for the data interpretation. The data analysis includes an evaluation of tramp uranium mass in reactor core, detection of failures by iodine and caesium spikes, evaluation of burnup of defective fuel. Evaluation of defective fuel burnup was carried out by applying the relation of caesium nuclides activity in spikes and relations of activities of gaseous fission products for steady state operational conditions. The method of burnup evaluation of defective fuel by use of fission gas activity is presented in details. The neural-network analysis is performed for determination of failed fuel rod number and defect size. Results of the expert system application are illustrated for several fuel campaigns on operating WWER NPPs. (authors)

Global absenteeism and presenteeism in mental health patients referred through primary care.

Science.gov (United States)

Bailey, S Kathleen; Haggarty, John; Kelly, Sara

2015-01-01

Disability from mental health (MH) symptoms impairs workers' functioning. Most of what is known about the MH of workers relates to their experiences after intervention or work absence. To profile the clinical symptoms, self-reported absenteeism and presenteeism and treatment response of workers with MH symptoms at the point of accessing MH care and compare the characteristics of patients referred with or without problems related to work. Analysis of 11 years of patient data collected in a Shared Mental Health Care (SMHC) clinic referred within a primary care setting in Ontario, Canada. Multiple regression with MH disorders was used to predict absenteeism and presenteeism. Absenteeism and presenteeism were assessed using the 12-item self-administered version of the WHO-DAS 2. Symptom profiles were assessed with the Patient Health Questionnaire (PHQ). Some psychiatric disorders (depression, somatization, anxiety) contributed more to predicting absenteeism and presenteeism than others. Patients referred with work-related problems differed from the general SMHC population in terms of sex and type and number of symptoms. Treatment response was good in both groups after a mean of three treatment visits. Patients with work-related mental health complaints formed a distinct clinical group that benefitted equally from the intervention(s) provided by SMHC.

High accuracy Primary Reference gas Mixtures for high-impact greenhouse gases

Science.gov (United States)

Nieuwenkamp, Gerard; Zalewska, Ewelina; Pearce-Hill, Ruth; Brewer, Paul; Resner, Kate; Mace, Tatiana; Tarhan, Tanil; Zellweger, Christophe; Mohn, Joachim

2017-04-01

Climate change, due to increased man-made emissions of greenhouse gases, poses one of the greatest risks to society worldwide. High-impact greenhouse gases (CO2, CH4 and N2O) and indirect drivers for global warming (e.g. CO) are measured by the global monitoring stations for greenhouse gases, operated and organized by the World Meteorological Organization (WMO). Reference gases for the calibration of analyzers have to meet very challenging low level of measurement uncertainty to comply with the Data Quality Objectives (DQOs) set by the WMO. Within the framework of the European Metrology Research Programme (EMRP), a project to improve the metrology for high-impact greenhouse gases was granted (HIGHGAS, June 2014-May 2017). As a result of the HIGHGAS project, primary reference gas mixtures in cylinders for ambient levels of CO2, CH4, N2O and CO in air have been prepared with unprecedented low uncertainties, typically 3-10 times lower than usually previously achieved by the NMIs. To accomplish these low uncertainties in the reference standards, a number of preparation and analysis steps have been studied and improved. The purity analysis of the parent gases had to be performed with lower detection limits than previously achievable. E.g., to achieve an uncertainty of 2•10-9 mol/mol (absolute) on the amount fraction for N2O, the detection limit for the N2O analysis in the parent gases has to be in the sub nmol/mol domain. Results of an OPO-CRDS analyzer set-up in the 5µm wavelength domain, with a 200•10-12 mol/mol detection limit for N2O, will be presented. The adsorption effects of greenhouse gas components at cylinder surfaces are critical, and have been studied for different cylinder passivation techniques. Results of a two-year stability study will be presented. The fit-for-purpose of the reference materials was studied for possible variation on isotopic composition between the reference material and the sample. Measurement results for a suit of CO2 in air

Jet Fuel Based High Pressure Solid Oxide Fuel Cell System

Science.gov (United States)

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

2015-01-01

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

Long-term implications of alternative light-duty vehicle technologies for global greenhouse gas emissions and primary energy demands

International Nuclear Information System (INIS)

Kyle, Page; Kim, Son H.

2011-01-01

This study assesses global light-duty vehicle (LDV) transport in the upcoming century, and the implications of vehicle technology advancement and fuel-switching on greenhouse gas emissions and primary energy demands. Five different vehicle technology scenarios are analyzed with and without a CO 2 emissions mitigation policy using the GCAM integrated assessment model: a reference internal combustion engine vehicle scenario, an advanced internal combustion engine vehicle scenario, and three alternative fuel vehicle scenarios in which all LDVs are switched to natural gas, electricity, or hydrogen by 2050. The emissions mitigation policy is a global CO 2 emissions price pathway that achieves 450 ppmv CO 2 at the end of the century with reference vehicle technologies. The scenarios demonstrate considerable emissions mitigation potential from LDV technology; with and without emissions pricing, global CO 2 concentrations in 2095 are reduced about 10 ppmv by advanced ICEV technologies and natural gas vehicles, and 25 ppmv by electric or hydrogen vehicles. All technological advances in vehicles are important for reducing the oil demands of LDV transport and their corresponding CO 2 emissions. Among advanced and alternative vehicle technologies, electricity- and hydrogen-powered vehicles are especially valuable for reducing whole-system emissions and total primary energy. - Highlights: → Alternative-fuel LDVs reduce whole-system CO 2 emissions, even without carbon pricing. → Alternative-fuel LDVs enhance the CO 2 mitigation capacity of the transportation sector. → Electric and hydrogen vehicles reduce whole-system primary energy supporting LDV transport.

Increased power generation from primary sludge by a submersible microbial fuel cell and optimum operational conditions

DEFF Research Database (Denmark)

Vologni, Valentina; Kakarla, Ramesh; Angelidaki, Irini

2013-01-01

Microbial fuel cells (MFCs) have received attention as a promising renewable energy technology for waste treatment and energy recovery. We tested a submersible MFC with an innovative design capable of generating a stable voltage of 0.250 ± 0.008 V (with a fixed 470 Ω resistor) directly from prima...... prolonged the current generation and increased the power density by 7 and 1.5 times, respectively, in comparison with raw primary sludge. These findings suggest that energy recovery from primary sludge can be maximized using an advanced MFC system with optimum conditions....

Relating the octane numbers of fuels to ignition delay times measured in an ignition quality tester (IQT)

KAUST Repository

Naser, Nimal; Yang, Seung Yeon; Kalghatgi, Gautam; Chung, Suk-Ho

2016-01-01

an ignition quality tester. A baseline data of ignition delay times were determined using an ignition quality tester at a charge pressure of 21.3 bar between 770 and 850 K and an equivalence ratio of 0.7 for various primary reference fuels (PRFs, mixtures

Recent advances in ratio primary reference measurement procedures (definitive methods) and their use in certification of reference materials and controlling assigned values in proficiency testing

International Nuclear Information System (INIS)

Dybczynski, R.S.; Polkowska-Motrenko, H.; Chajduk, E.; Danko, B.; Pyszynska, M.

2014-01-01

Three very accurate (definitive) methods by RNAA for the determination of Se, As and Fe respectively, which were recently elaborated in our laboratory, are reviewed and their use in certification of reference materials and in checking the assigned values in proficiency tests is demonstrated on several examples. According to VIM 3 nomenclature these methods may be called: ratio primary reference measurement procedures (RPRMPs). RPRMPs with their expanded uncertainties of 2.7-3.6 % are comparable to ID-MS methods and are the only methods of such high metrological quality which can be used for the determination of trace amounts of monoisotopic elements. (author)

Experimental investigation of the auto-ignition characteristics of oxygenated reference fuel compounds

Science.gov (United States)

Walton, Stephen Michael

The increased use of biofuels presents an opportunity to improve combustion performance while simultaneously reducing greenhouse gases and pollutant emissions. This work focused on improving the fundamental understanding of the auto-ignition chemistry of oxygenated reference fuel compounds. A systematic study of the effects of ester structure on ignition chemistry was performed using the University of Michigan Rapid Compression Facility. The ignition properties of the ester compounds were investigated over a broad range of pressures (P=5-20 atm) and temperatures (T=850-1150 K) which are directly relevant to advanced combustion engine strategies. Ignition delay times for five esters were determined using the RCF. The esters were selected to systematically consider the chemical structure of the compounds. Three esters were saturated: methyl butanoate, butyl methanoate, and ethyl propanoate; and two were unsaturated: methyl crotonate and methyl trans-3-hexenoate. The unsaturated esters were more reactive than their saturated counterparts, with the largest unsaturated ester, methyl trans-3-hexenoate having the highest reactivity. Two isomers of the saturated esters, butyl methanoate and ethyl propanoate, were more reactive than the isomer methyl butanoate. The results are explained if we assume that butyl methanoate and ethyl propanoate form intermediate ring structures which decompose more rapidly than esters such as methyl butanoate, which do not form ring structures. Modeling studies of the reaction chemistry were conducted for methyl butanoate and ethyl propanoate, for which detailed mechanisms were available in the literature. The new experimental data indicated that literature rate coefficients for some of the methyl butanoate/HO2 reactions were too fast. Modifying these within the theoretical uncertainties for the reaction rates, led to excellent agreement between the model predictions and the experimental data. Comparison of the modeling results with the

Fuel characteristics pertinent to the design of aircraft fuel systems

Science.gov (United States)

Barnett, Henry C; Hibbard, R R

1953-01-01

Because of the importance of fuel properties in design of aircraft fuel systems the present report has been prepared to provide information on the characteristics of current jet fuels. In addition to information on fuel properties, discussions are presented on fuel specifications, the variations among fuels supplied under a given specification, fuel composition, and the pertinence of fuel composition and physical properties to fuel system design. In some instances the influence of variables such as pressure and temperature on physical properties is indicated. References are cited to provide fuel system designers with sources of information containing more detail than is practicable in the present report.

Survey of reference materials. V. 2: Environmentally related reference materials for trace elements, nuclides and microcontaminants

International Nuclear Information System (INIS)

1996-05-01

The present report presently contains over 250 reference materials with trace element and organic contaminant information on fuel, geological and mineral, anthropogenic disposal, soil reference and miscellaneous reference materials. Not included in the current report is information on most biological and environmental reference materials with trace element, stable isotope, radioisotope and organic contaminant information. 8 refs, tabs

Fuel gases generation in the primary contention during a coolant loss accident in a nuclear power plant with reactor type BWR

International Nuclear Information System (INIS)

Salaices, M.; Salaices, E.; Ovando, R.; Esquivias, J.

2011-11-01

During an accident design base of coolant loos, the hydrogen gas can accumulate inside the primary contention as a result of several generation mechanisms among those that are: 1) the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant, 2) the metals corrosion for the solutions used in the emergency cooling and dew of the contention, and 3) the radio-decomposition of the cooling solutions of post-accident emergency. In this work the contribution of each generation mechanism to the hydrogen total in the primary contention is analyzed, considering typical inventories of zirconium, zinc, aluminum and fission products in balance cycle of a reactor type BWR. In the analysis the distribution model of fission products and hydrogen production proposed in the regulator guide 1.7, Rev. 2 of the US NRC was used. The results indicate that the mechanism that more contributes to the hydrogen generation at the end of a period of 24 hours of initiate the accident is the radio-decomposition of the cooling solutions of post-accident emergency continued by the reaction metal-water involving the zirconium of the fuel cladding with the reactor coolant, and lastly the aluminum and zinc oxidation present in the primary contention. However, the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant is the mechanism that more contributes to the hydrogen generation in the first moments after the accident. This study constitutes the first part of the general analysis of the generation, transport and control of fuel gases in the primary contention during a coolant loss accident in BWRs. (Author)

The Kjeldahl method as a primary reference procedure for total protein in certified reference materials used in clinical chemistry. II. Selection of direct Kjeldahl analysis and its preliminary performance parameters.

Science.gov (United States)

Vinklárková, Bára; Chromý, Vratislav; Šprongl, Luděk; Bittová, Miroslava; Rikanová, Milena; Ohnútková, Ivana; Žaludová, Lenka

2015-01-01

To select a Kjeldahl procedure suitable for the determination of total protein in reference materials used in laboratory medicine, we reviewed in our previous article Kjeldahl methods adopted by clinical chemistry and found an indirect two-step analysis by total Kjeldahl nitrogen corrected for its nonprotein nitrogen and a direct analysis made on isolated protein precipitates. In this article, we compare both procedures on various reference materials. An indirect Kjeldahl method gave falsely lower results than a direct analysis. Preliminary performance parameters qualify the direct Kjeldahl analysis as a suitable primary reference procedure for the certification of total protein in reference laboratories.

Reference costs of electricity

International Nuclear Information System (INIS)

Terraz, N.

1997-01-01

The calculation of electric power production reference costs is used in France, even in the present case of over-capacity, for comparing the relative interest of the various means of power generation (nuclear plants, coal plants, hydroelectricity, gas combined cycles, etc.) and as an aid for future investment decisions. Reference costs show a sharp decrease between 1993 and 1997 due to advancements in nuclear plant operating ability and fossil fuel price decrease. Actuarial rates, plant service life, fuel costs and exchange rates are important parameters. The various costs from the research stage to the waste processing stages are discussed and the reference costs of the various power generation systems are presented and compared together with their competitiveness; the future of wind energy and cogeneration and the prospective of the renewal of nuclear plants at the 2010 horizon are also addressed

Primary emissions and secondary organic aerosol formation from the exhaust of a flex-fuel (ethanol) vehicle

Science.gov (United States)

Suarez-Bertoa, R.; Zardini, A. A.; Platt, S. M.; Hellebust, S.; Pieber, S. M.; El Haddad, I.; Temime-Roussel, B.; Baltensperger, U.; Marchand, N.; Prévôt, A. S. H.; Astorga, C.

2015-09-01

Incentives to use biofuels may result in increasing vehicular emissions of compounds detrimental to air quality. Therefore, regulated and unregulated emissions from a Euro 5a flex-fuel vehicle, tested using E85 and E75 blends (gasoline containing 85% and 75% of ethanol (vol/vol), respectively), were investigated at 22 and -7 °C over the New European Driving Cycle, at the Vehicle Emission Laboratory at the European Commission Joint Research Centre Ispra, Italy. Vehicle exhaust was comprehensively analyzed at the tailpipe and in a dilution tunnel. A fraction of the exhaust was injected into a mobile smog chamber to study the photochemical aging of the mixture. We found that emissions from a flex-fuel vehicle, fueled by E85 and E75, led to secondary organic aerosol (SOA) formation, despite the low aromatic content of these fuel blends. Emissions of regulated and unregulated compounds, as well as emissions of black carbon (BC) and primary organic aerosol (POA) and SOA formation were higher at -7 °C. The flex-fuel unregulated emissions, mainly composed of ethanol and acetaldehyde, resulted in very high ozone formation potential and SOA, especially at low temperature (860 mg O3 km-1 and up to 38 mg C kg-1). After an OH exposure of 10 × 106 cm-3 h, SOA mass was, on average, 3 times larger than total primary particle mass emissions (BC + POA) with a high O:C ratio (up to 0.7 and 0.5 at 22 and -7 °C, respectively) typical of highly oxidized mixtures. Furthermore, high resolution organic mass spectra showed high 44/43 ratios (ratio of the ions m/z 44 and m/z 43) characteristic of low-volatility oxygenated organic aerosol. We also hypothesize that SOA formation from vehicular emissions could be due to oxidation products of ethanol and acetaldehyde, both short-chain oxygenated VOCs, e.g. methylglyoxal and acetic acid, and not only from aromatic compounds.

Studies on the primary and secondary residues from the dissolution of high-burnup nuclear fuels

International Nuclear Information System (INIS)

Schmid, M.

1986-01-01

To clarify the composition of residues from the dissolution of high-burnup nuclear fuels a sample with a burnup of 4.5 GWd and a two year cooling period was studied with the help of REM-EDX. In a parallel experiment an inactive simulator of a solution was subjected to a similar chemical treatment. The residues which resulted from this were analysed analogously. As a result of the results the chemistry of the following compounds in HNO 3 were studied: MoO 3 , ZrMo 2 O 5 (OH) 2 x2H 2 O, the oxide of antimony as well as Sb 4 O 4 (OH) 2 (NO 3 ) 2 , PdO.xH 2 O, Ag 2 Se, Ag 2 Te, and CsTcO 4 . Of special interest here were the solubility and precipitation formation of these compounds as well as the influence of a high (ca. 1 mol/l) concentration of uranium on these characteristics. With high radiation doses to the simulated solution a radiolytical reduction of Pd 2+ was established and was studied more closely with pure Pd(NO 3 ) 2 solutions. In primary dissolution residues the presence of the radionuclides Ru-106, Ag-110m, Sb-125, Cs-134, and Cs-137 was γ-spectrometrically proven. The residue was made up primarily of an element combination of Mo and Ru. As other components Rh, Pd and Tc appear in an alloy as the so-called ε phase, which already has to be present in the fuel, because this phase was not exhibited in the similarly handled simulator. Zirconium molybdate was not identified in the real feed slurries, but was definitely present in the precipitation of the simulated feed solution. The analysis of the primary residues also showed pure zirconium particles, presumably from the zirconium alloy of the fuel cans, as well as undissolved fuel particles. The precipitation from the fuel solution was made up of agglomerates of the smallest particles of the ε phase, upon which silver halogenides were crystallized. Radiochemically reduced Pd was also found. (orig./RB) [de

Numerical Simulations of Hollow Cone Injection and Gasoline Compression Ignition Combustion With Naphtha Fuels

KAUST Repository

Badra, Jihad A.

2016-01-11

Gasoline compression ignition (GCI), also known as partially premixed compression ignition (PPCI) and gasoline direct injection compression ignition (GDICI), engines have been considered an attractive alternative to traditional spark ignition engines. Lean burn combustion with the direct injection of fuel eliminates throttle losses for higher thermodynamic efficiencies, and the precise control of the mixture compositions allows better emission performance such as NOx and particulate matter (PM). Recently, low octane gasoline fuel has been identified as a viable option for the GCI engine applications due to its longer ignition delay characteristics compared to diesel and lighter evaporation compared to gasoline fuel [1]. The feasibility of such a concept has been demonstrated by experimental investigations at Saudi Aramco [1, 2]. The present study aims to develop predictive capabilities for low octane gasoline fuel compression ignition engines with accurate characterization of the spray dynamics and combustion processes. Full three-dimensional simulations were conducted using CONVERGE as a basic modeling framework, using Reynolds-averaged Navier-Stokes (RANS) turbulent mixing models. An outwardly opening hollow-cone spray injector was characterized and validated against existing and new experimental data. An emphasis was made on the spray penetration characteristics. Various spray breakup and collision models have been tested and compared with the experimental data. An optimum combination has been identified and applied in the combusting GCI simulations. Linear instability sheet atomization (LISA) breakup model and modified Kelvin-Helmholtz and Rayleigh-Taylor (KH-RT) break models proved to work the best for the investigated injector. Comparisons between various existing spray models and a parametric study have been carried out to study the effects of various spray parameters. The fuel effects have been tested by using three different primary reference fuel (PRF

Real-time monitoring of methanol concentration using a shear horizontal surface acoustic wave sensor for direct methanol fuel cell without reference liquid measurement

Science.gov (United States)

Tada, Kyosuke; Nozawa, Takuya; Kondoh, Jun

2017-07-01

In recent years, there has been an increasing demand for sensors that continuously measure liquid concentrations and detect abnormalities in liquid environments. In this study, a shear horizontal surface acoustic wave (SH-SAW) sensor is applied for the continuous monitoring of liquid concentrations. As the SH-SAW sensor functions using the relative measurement method, it normally needs a reference at each measurement. However, if the sensor is installed in a liquid flow cell, it is difficult to measure a reference liquid. Therefore, it is important to establish an estimation method for liquid concentrations using the SH-SAW sensor without requiring a reference measurement. In this study, the SH-SAW sensor is installed in a direct methanol fuel cell to monitor the methanol concentration. The estimated concentration is compared with a conventional density meter. Moreover, the effect of formic acid is examined. When the fuel temperature is higher than 70 °C, it is necessary to consider the influence of liquid conductivity. Here, an estimation method for these cases is also proposed.

Reference News Release: U.S. Files Complaint, Announces Settlement to Address Alleged Renewable Fuel Standard Violations by NGL Crude Logistics and Western Dubuque Biodiesel

Science.gov (United States)

Reference news release on the complaint against NGL Crude Logistics, LLC and Western Dubuque Biodiesel, LLC and a settlement with Western Dubuque to address alleged violations of the Renewable Fuel Standard.

Greenhouse gas reduction and primary energy savings via adoption of a fuel cell hybrid plant in a hospital

International Nuclear Information System (INIS)

Bizzarri, Giacomo; Morini, Gian Luca

2004-01-01

The Kyoto agreement, expressing great concern about global climate change, has stated emissions of greenhouse gases, especially CO 2 from fossil fuel use, need to be reduced. According to this, existing plants have been required to cut emissions; moreover, there has been a greater emphasis on adopting efficient systems in order to reduce the energy losses. Among high efficiency technologies, fuel cells appear to be the most promising for their high efficiency and their very low environmental impact. This paper first reviews the state-of-the-art of fuel cells systems, then simulates the operation of a hybrid fuel cells plant in a 'typical hospital' analysing how it could optimize the hospitals energetic requirements. Hospitals and sanitary structures are normally characterized by considerable energy demands not often suitable with resolute energy retrofit strategies. Both the considerable primary energy savings and the pollutant emissions reduction, achieved upgrading conventional systems to a fuel cell hybrid plant, have the potential to prompt national boards to support their business development, as long as they achieve a consolidated market penetration

Characterization of spent fuel disassembly hardware and nonfuel bearing components and their relationship to 10 CFR 61

International Nuclear Information System (INIS)

Luksic, A.T.

1987-02-01

There are a variety of wastes that will be disposed of by the federal waste management system under the Nuclear Waste Policy Act of 1982. The primary waste form is spent nuclear fuel. Currently, this is in the form of fuel assemblies. If the fuel pins are removed from the fuel assembly, as in consolidation, then the fuel pins and the structural portion of the fuel assembly must be considered as separate waste streams. The structural hardware consists of end fittings, grid spacers, water rods (BWR 8 x 8 only), control rod guide tubes (PWR only) and various nuts, washers, springs, etc. These are referred to as spent fuel disassembly (SFD) hardware. There will also be a number of other components which are defined in Appendix E of 10 CFR 961, the standard utility contract. These are referred to as nonfuel-bearing (NFB) components, and include fuel channels (BWR), control rods, fission chambers, neutron sources, thimble plugs, and other components. This paper characterizes spent fuel disassembly (SFD) hardware, and nonfuel-bearing (NFB) components for the most abundant fuel types. The descriptions and figures given are representative for the items described. Many subvariants exist due to design evaluation, which are not covered. This paper also discusses the relationship of these wastes to 10 CFR 61 waste classification

Fuel property effects on Navy aircraft fuel systems

Science.gov (United States)

Moses, C. A.

1984-01-01

Problems of ensuring compatibility of Navy aircraft with fuels that may be different than the fuels for which the equipment was designed and qualified are discussed. To avoid expensive requalification of all the engines and airframe fuel systems, methodologies to qualify future fuels by using bench-scale and component testing are being sought. Fuel blends with increasing JP5-type aromatic concentration were seen to produce less volume swell than an equivalent aromatic concentration in the reference fuel. Futhermore, blends with naphthenes, decalin, tetralin, and naphthalenes do not deviate significantly from the correlation line of aromatic blends, Similar results are found with tensile strenth and elongation. Other elastomers, sealants, and adhesives are also being tested.

Electricity generation from fermented primary sludge using single-chamber air-cathode microbial fuel cells

KAUST Repository

Yang, Fei

2013-01-01

Single-chamber air-cathode microbial fuel cells (MFCs) were used to generate electricity from fermented primary sludge. Fermentation (30°C, 9days) decreased total suspended solids (26.1-16.5g/L), volatile suspended solids (24.1-15.3g/L) and pH (5.7-4.5), and increased conductivity (2.4-4.7mS/cm), soluble COD (2.66-15.5g/L), and volatile fatty acids (1.9-10.1g/L). To lower the COD and increase pH, fermentation supernatant was diluted with primary effluent before being used in the MFCs. The maximum power density was 0.32±0.01W/m2, compared to 0.24±0.03W/m2 with only primary effluent. Power densities were higher with phosphate buffer added to the supernatant (1.03±0.06W/m2) or the solution (0.87±0.05W/m2). Coulombic efficiencies ranged from 18% to 57%, and sCOD removals from 84% to 94%. These results demonstrated that sludge can effectively be used for power generation when fermented and then diluted with only primary effluent. © 2012 Elsevier Ltd.

Gravimetric preparation and characterization of primary reference solutions of molybdenum and rhodium.

Science.gov (United States)

Kaltenbach, Angela; Noordmann, Janine; Görlitz, Volker; Pape, Carola; Richter, Silke; Kipphardt, Heinrich; Kopp, Gernot; Jährling, Reinhard; Rienitz, Olaf; Güttler, Bernd

2015-04-01

Gravimetrically prepared mono-elemental reference solutions having a well-known mass fraction of approximately 1 g/kg (or a mass concentration of 1 g/L) define the very basis of virtually all measurements in inorganic analysis. Serving as the starting materials of all standard/calibration solutions, they link virtually all measurements of inorganic analytes (regardless of the method applied) to the purity of the solid materials (high-purity metals or salts) they were prepared from. In case these solid materials are characterized comprehensively with respect to their purity, this link also establishes direct metrological traceability to The International System of Units (SI). This, in turn, ensures the comparability of all results on the highest level achievable. Several national metrology institutes (NMIs) and designated institutes (DIs) have been working for nearly two decades in close cooperation with commercial producers on making an increasing number of traceable reference solutions available. Besides the comprehensive characterization of the solid starting materials, dissolving them both loss-free and completely under strict gravimetric control is a challenging problem in the case of several elements like molybdenum and rhodium. Within the framework of the European Metrology Research Programme (EMRP), in the Joint Research Project (JRP) called SIB09 Primary standards for challenging elements, reference solutions of molybdenum and rhodium were prepared directly from the respective metals with a relative expanded uncertainty associated with the mass fraction of U rel(w) methods required to assist with the preparation and as dissemination tools.

Preparation and certification of reference materials for the nuclear industry

International Nuclear Information System (INIS)

Wagner, J.F.

1994-01-01

CETAMA (Commission for the Establishment of Analytical Methods) of the Cea group is producing and certifying uranium and plutonium reference materials to meet the requirements of analytical laboratories in the nuclear industry. Reference materials are required at all stages in the fuel cycle: from extraction of uranium, purification, conversion, uranium enrichment, fuel fabrication until reprocessing of spent fuel, management and storage of waste. 3 tabs

Translating evidence into practice: Hong Kong Reference Framework for Preventive Care for Children in Primary Care Settings.

Science.gov (United States)

Siu, Natalie P Y; Too, L C; Tsang, Caroline S H; Young, Betty W Y

2015-06-01

There is increasing evidence that supports the close relationship between childhood and adult health. Fostering healthy growth and development of children deserves attention and effort. The Reference Framework for Preventive Care for Children in Primary Care Settings has been published by the Task Force on Conceptual Model and Preventive Protocols under the direction of the Working Group on Primary Care. It aims to promote health and prevent disease in children and is based on the latest research, and contributions of the Clinical Advisory Group that comprises primary care physicians, paediatricians, allied health professionals, and patient groups. This article highlights the comprehensive, continuing, and patient-centred preventive care for children and discusses how primary care physicians can incorporate the evidence-based recommendations into clinical practice. It is anticipated that the adoption of this framework will contribute to improved health and wellbeing of children.

FCRD Transmutation Fuels Handbook 2015

Energy Technology Data Exchange (ETDEWEB)

Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Papesch, Cynthia Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. It is, therefore, important to understand the properties of U-Pu-Zr alloys, both with and without minor actinide additions. In addition to requiring extensive safety precautions, alloys containing U and Pu are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phase-transformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, and that general acceptance of results sometimes indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, and alloys of two or three of these elements. It contains information about phase diagrams and related information (including phases and phase transformations); heat capacity, entropy, and enthalpy; thermal expansion; and thermal conductivity and diffusivity. In addition to presenting information about materials properties, it attempts to provide information about how well the property is known and how much variation exists between measurements. Although the handbook includes some references to publications about modeling

HTGR fuel reprocessing technology

International Nuclear Information System (INIS)

Brooks, L.H.; Heath, C.A.; Shefcik, J.J.

1976-01-01

The following aspects of HTGR reprocessing technology are discussed: characteristics of HTGR fuels, criteria for a fuel reprocessing flowsheet; selection of a reference reprocessing flowsheet, and waste treatment

Crud deposition on fuel in WWER reactors

International Nuclear Information System (INIS)

Kysela, J.; Svarc, V.; Androva, K.; Ruzickova, M.

2008-01-01

Reliability of nuclear fuel and radiation fields surrounding primary systems are important aspects of overall nuclear reactor safety. Corrosion product (crud) deposition on fuel surfaces has implications for fuel performance through heat transfer and local chemistry modifications. Crud is currently one of the key industry issues and has been implicated in several recent cases of crud-related fuel failures and core plugging. Activated crud is deposited on out-of-core surfaces, mainly steam generators, resulting in high radiation fields and high doses of plant staff. Due to radiation build-up in primary circuit systems, decontamination of primary systems components and steam generators is used. Several issues involving decontamination were observed in some cases. After decontamination higher corrosion product release occurs followed by subsequent crud deposition on fuel surfaces. The paper summarizes experience with water chemistry and decontamination that can influence crud deposition on fuel surfaces. The following areas are discussed: 1) Experience with crud deposition, primary water chemistry and decontamination under operating conditions; 2) The behaviour of organic compounds in primary coolant and on fuel surfaces; 3) A proposed experimental programme to study crud deposition. (authors)

Stationary Liquid Fuel Fast Reactor

International Nuclear Information System (INIS)

Yang, Won Sik; Grandy, Andrew; Boroski, Andrew; Krajtl, Lubomir; Johnson, Terry

2015-01-01

For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

Stationary Liquid Fuel Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Grandy, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Boroski, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, Terry [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-09-30

For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

On the effects of fuel properties and injection timing in partially premixed compression ignition of low octane fuels

KAUST Repository

Naser, Nimal

2017-06-29

A better understanding on the effects of fuel properties and injection timing is required to improve the performance of advanced engines based on low temperature combustion concepts. In this work, an experimental and computational study was conducted to investigate the effects of physical and chemical kinetic properties of low octane fuels and their surrogates in partially premixed compression ignition (PPCI) engines. The main objective was to identify the relative importance of physical versus chemical kinetic properties in predicting practical fuel combustion behavior across a range of injection timings. Two fuel/surrogate pairs were chosen for comparison: light naphtha (LN) versus the primary reference fuel (PRF) with research octane number of 65 (PRF 65), and FACE (fuels for advanced combustion engines) I gasoline versus PRF 70. Two sets of parametric studies were conducted: the first varied the amount of injected fuel mass at different injection timings to match a fixed combustion phasing, and the second maintained the same injected fuel mass at each injection timing to assess resulting combustion phasing changes. Full-cycle computational fluid dynamic engine simulations were conducted by accounting for differences in the physical properties of the original and surrogate fuels, while employing identical chemical kinetics. The simulations were found to capture trends observed in the experiments, while providing details on spatial mixing and chemical reactivity for different fuels and injection timings. It was found that differences in physical properties become increasingly important as injection timing was progressively delayed from premixed conditions, and this was rationalized by analysis of mixture stratification patterns resulting from injection of fuels with different physical properties. The results suggest that accurate descriptions of both physical and chemical behavior of fuels are critical in predictive simulations of PPCI engines for a wide range of

On the effects of fuel properties and injection timing in partially premixed compression ignition of low octane fuels

KAUST Repository

Naser, Nimal; Jaasim, Mohammed; Atef, Nour; Chung, Suk-Ho; Im, Hong G.; Sarathy, Mani

2017-01-01

A better understanding on the effects of fuel properties and injection timing is required to improve the performance of advanced engines based on low temperature combustion concepts. In this work, an experimental and computational study was conducted to investigate the effects of physical and chemical kinetic properties of low octane fuels and their surrogates in partially premixed compression ignition (PPCI) engines. The main objective was to identify the relative importance of physical versus chemical kinetic properties in predicting practical fuel combustion behavior across a range of injection timings. Two fuel/surrogate pairs were chosen for comparison: light naphtha (LN) versus the primary reference fuel (PRF) with research octane number of 65 (PRF 65), and FACE (fuels for advanced combustion engines) I gasoline versus PRF 70. Two sets of parametric studies were conducted: the first varied the amount of injected fuel mass at different injection timings to match a fixed combustion phasing, and the second maintained the same injected fuel mass at each injection timing to assess resulting combustion phasing changes. Full-cycle computational fluid dynamic engine simulations were conducted by accounting for differences in the physical properties of the original and surrogate fuels, while employing identical chemical kinetics. The simulations were found to capture trends observed in the experiments, while providing details on spatial mixing and chemical reactivity for different fuels and injection timings. It was found that differences in physical properties become increasingly important as injection timing was progressively delayed from premixed conditions, and this was rationalized by analysis of mixture stratification patterns resulting from injection of fuels with different physical properties. The results suggest that accurate descriptions of both physical and chemical behavior of fuels are critical in predictive simulations of PPCI engines for a wide range of

Fuel performance annual report, period through December 1978

International Nuclear Information System (INIS)

Houston, M.D.

1979-12-01

This annual report, intended to be the first in a series, provides a brief description of fuel performance in commercial nuclear power plants. Brief summaries are given for the reporting period of on-site fuel surveillance programs, fuel performance problems, and changes in commercial fuel designs. The report provides many references to more detailed information and to related NRC evaluations. 57 references

Impact of thermodynamic properties and heat loss on ignition of transportation fuels in rapid compression machines

KAUST Repository

Ahmed, Ahfaz

2018-01-30

Rapid compression machines (RCM) are extensively used to study autoignition of a wide variety of fuels at engine relevant conditions. Fuels ranging from pure species to full boiling range gasoline and diesel can be studied in an RCM to develop a better understanding of autoignition kinetics in low to intermediate temperature ranges. In an RCM, autoignition is achieved by compressing a fuel/oxidizer mixture to higher pressure and temperature, thereby initiating chemical reactions promoting ignition. During these experiments, the pressure is continuously monitored and is used to deduce significant events such as the end of compression and the onset of ignition. The pressure profile is also used to assess the temperature evolution of the gas mixture with time using the adiabatic core hypothesis and the heat capacity ratio of the gas mixture. In such RCM studies, real transportation fuels containing many components are often represented by simpler surrogate fuels. While simpler surrogates such as primary reference fuels (PRFs) and ternary primary reference fuel (TPRFs) can match research and motor octane number of transportation fuels, they may not accurately replicate thermodynamic properties (including heat capacity ratio). This non-conformity could exhibit significant discrepancies in the end of compression temperature, thereby affecting ignition delay (τign) measurements. Another aspect of RCMs that can affect τign measurement is post compression heat loss, which depends on various RCM parameters including geometry, extent of insulation, pre-heating temperature etc. To, better understand the effects of these non-chemical kinetic parameters on τign, thermodynamic properties of a number of FACE G gasoline surrogates were calculated and simulated in a multi-zone RCM model. The problem was further investigated using a variance based analysis and individual sensitivities were calculated. This study highlights the effects on τign due to thermodynamic properties of

Impact of thermodynamic properties and heat loss on ignition of transportation fuels in rapid compression machines

KAUST Repository

Ahmed, Ahfaz; Hantouche, Mireille; Khurshid, Muneeb; Mohamed, Samah; Nasir, Ehson Fawad; Farooq, Aamir; Roberts, William L.; Knio, Omar; Sarathy, Mani

2018-01-01

Rapid compression machines (RCM) are extensively used to study autoignition of a wide variety of fuels at engine relevant conditions. Fuels ranging from pure species to full boiling range gasoline and diesel can be studied in an RCM to develop a better understanding of autoignition kinetics in low to intermediate temperature ranges. In an RCM, autoignition is achieved by compressing a fuel/oxidizer mixture to higher pressure and temperature, thereby initiating chemical reactions promoting ignition. During these experiments, the pressure is continuously monitored and is used to deduce significant events such as the end of compression and the onset of ignition. The pressure profile is also used to assess the temperature evolution of the gas mixture with time using the adiabatic core hypothesis and the heat capacity ratio of the gas mixture. In such RCM studies, real transportation fuels containing many components are often represented by simpler surrogate fuels. While simpler surrogates such as primary reference fuels (PRFs) and ternary primary reference fuel (TPRFs) can match research and motor octane number of transportation fuels, they may not accurately replicate thermodynamic properties (including heat capacity ratio). This non-conformity could exhibit significant discrepancies in the end of compression temperature, thereby affecting ignition delay (τign) measurements. Another aspect of RCMs that can affect τign measurement is post compression heat loss, which depends on various RCM parameters including geometry, extent of insulation, pre-heating temperature etc. To, better understand the effects of these non-chemical kinetic parameters on τign, thermodynamic properties of a number of FACE G gasoline surrogates were calculated and simulated in a multi-zone RCM model. The problem was further investigated using a variance based analysis and individual sensitivities were calculated. This study highlights the effects on τign due to thermodynamic properties of

Description and characterization of BRPR series S-0, S-1, S-2, S-3, and S-4 demonstration fuel rods

International Nuclear Information System (INIS)

Welty, R.K.

1981-07-01

This report describes the process development, fabrication, and pre-irradiation characterization of the demonstration fuel rods for irradiation in the Big Rock Point Reactor as part of the DOE-sponsored Fuel Performance Improvement Program (FPIP). The fuel rods represent advanced designs that are expected to exhibit improved performance with respect to pellet-cladding-interaction and the attainment of extended burnup. Whereas other design variations are described, the primary fuel concepts being evaluated as part of the FPIP are an annular-coated-pressurized design and, at a more modest level, a sphere-pac design. A solid-pellet reference design provides the basis for comparing irradiation behavior

Analysis of Advanced Fuel Kernel Technology

International Nuclear Information System (INIS)

Oh, Seung Chul; Jeong, Kyung Chai; Kim, Yeon Ku; Kim, Young Min; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung

2010-03-01

The reference fuel for prismatic reactor concepts is based on use of an LEU UCO TRISO fissile particle. This fuel form was selected in the early 1980s for large high-temperature gas-cooled reactor (HTGR) concepts using LEU, and the selection was reconfirmed for modular designs in the mid-1980s. Limited existing irradiation data on LEU UCO TRISO fuel indicate the need for a substantial improvement in performance with regard to in-pile gaseous fission product release. Existing accident testing data on LEU UCO TRISO fuel are extremely limited, but it is generally expected that performance would be similar to that of LEU UO 2 TRISO fuel if performance under irradiation were successfully improved. Initial HTGR fuel technology was based on carbide fuel forms. In the early 1980s, as HTGR technology was transitioning from high-enriched uranium (HEU) fuel to LEU fuel. An initial effort focused on LEU prismatic design for large HTGRs resulted in the selection of UCO kernels for the fissile particles and thorium oxide (ThO 2 ) for the fertile particles. The primary reason for selection of the UCO kernel over UO 2 was reduced CO pressure, allowing higher burnup for equivalent coating thicknesses and reduced potential for kernel migration, an important failure mechanism in earlier fuels. A subsequent assessment in the mid-1980s considering modular HTGR concepts again reached agreement on UCO for the fissile particle for a prismatic design. In the early 1990s, plant cost-reduction studies led to a decision to change the fertile material from thorium to natural uranium, primarily because of a lower long-term decay heat level for the natural uranium fissile particles. Ongoing economic optimization in combination with anticipated capabilities of the UCO particles resulted in peak fissile particle burnup projection of 26% FIMA in steam cycle and gas turbine concepts

Nuclear fuels

International Nuclear Information System (INIS)

Gangwani, Saloni; Chakrabortty, Sumita

2011-01-01

Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

Summary of the Environmental Impact Statement on the concept for disposal of Canada's nuclear fuel waste

International Nuclear Information System (INIS)

1994-01-01

implementation - siting. These topics are discussed in more detail in the EIS. In addition, many supporting references have been issued, including nine 'primary references' that deal with major aspects of the disposal concept. The primary references are listed and described on the following two pages.

FAILED FUEL DISPOSITION STUDY

International Nuclear Information System (INIS)

THIELGES, J.R.

2004-01-01

In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

FAILED FUEL DISPOSITION STUDY

Energy Technology Data Exchange (ETDEWEB)

THIELGES, J.R.

2004-12-20

In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

Numerical Simulations of Hollow-Cone Injection and Gasoline Compression Ignition Combustion With Naphtha Fuels

KAUST Repository

Badra, Jihad A.

2016-01-29

tested and compared with the experimental data. An optimum combination has been identified and applied in the combusting GCI simulations. Linear instability sheet atomization (LISA) breakup model and modified Kelvin-Helmholtz and Rayleigh-Taylor (KH-RT) break models proved to work the best for the investigated injector. Comparisons between various existing spray models and a parametric study have been carried out to study the effects of various spray parameters. The fuel effects have been tested by using three different primary reference fuel (PRF) and toluene primary reference fuel (TPRF) surrogates. The effects of fuel temperature and chemical kinetic mechanisms have also been studied. The heating and evaporative characteristics of the low octane gasoline fuel and its PRF and TPRF surrogates were examined. Copyright © 2016 by ASME.

Nuclear Power and Resource Efficiency—A Proposal for a Revised Primary Energy Factor

Directory of Open Access Journals (Sweden)

Ola Eriksson

2017-06-01

Full Text Available Measuring resource efficiency can be achieved using different methods, of which primary energy demand is commonly used. The primary energy factor (PEF is a figure describing how much energy from primary resources is being used per unit of energy delivered. The PEF for nuclear power is typically 3, which refers to thermal energy released from fission in relation to electricity generated. Fuel losses are not accounted for. However; nuclear waste represents an energy loss, as current plans for nuclear waste management mostly include final disposal. Based on a literature review and mathematical calculations of the power-to-fuel ratio for nuclear power, PEF values for the open nuclear fuel cycle (NFC option of nuclear power and different power mixes are calculated. These calculations indicate that a more correct PEF for nuclear power would be 60 (range 32–88; for electricity in Sweden (41% nuclear power PEF would change from 1.8 to 25.5, and the average PEF for electricity in the European Union (EU would change from 2.5 to 18. The results illustrate the poor resource efficiency of nuclear power, which paves the way for the fourth generation of nuclear power and illustrates the policy implication of using PEFs which are inconsistent with current waste management plans.

Shuttle Primary Reaction Control Subsystem Thruster Fuel Valve Pilot Seal Extrusion: A Failure Correlation

Science.gov (United States)

Waller, Jess; Saulsberry, Regor L.

2003-01-01

Pilot operated valves (POVs) are used to control the flow of hypergolic propellants monomethylhydrazine (fuel) and nitrogen tetroxide (oxidizer) to the Shuttle orbiter Primary Reaction Control Subsystem (PRCS) thrusters. The POV incorporates a two-stage design: a solenoid-actuated pilot stage, which in turn controls a pressure-actuated main stage. Isolation of propellant supply from the thruster chamber is accomplished in part by a captive polytetrafluoroethylene (PTFE) pilot seal retained inside a Custom 455.1 stainless steel cavity. Extrusion of the pilot seal restricts the flow of fuel around the pilot poppet, thus impeding or preventing the main valve stage from opening. It can also prevent the main stage from staying open with adequate force margin, particularly if there is gas in the main stage actuation cavity. During thruster operation on-orbit, fuel valve pilot seal extrusion is commonly indicated by low or erratic chamber pressure or failure of the thruster to fire upon command (Fail-Off). During ground turnaround, pilot seal extrusion is commonly indicated by slow gaseous nitrogen (GN2) main valve opening times (greater than 38 ms) or slow water main valve opening response times (greater than 33 ms). Poppet lift tests and visual inspection can also detect pilot seal extrusion during ground servicing; however, direct metrology on the pilot seat assembly provides the most quantitative and accurate means of identifying extrusion. Minimizing PRCS fuel valve pilot seal extrusion has become an important issue in the effort to improve PRCS reliability and reduce associated life cycle costs.

Wood fuel and the environment

International Nuclear Information System (INIS)

Foster, C.A.

1992-01-01

The purpose of this paper is to try and demonstrate the role that the use of Wood as a Fuel can play in our environment. The term ''Wood Fuel'', for the purposes of these proceedings, refers to the use of wood obtained from the forest or the farm. It does not refer to waste wood from for example buildings. The role of wood fuel in the environment can be assessed at many different levels. In this paper three different scales of ''Environment'' and the role of wood fuel in each, will be considered. These three scales are namely the global environment, the local environment, and the National (community) environment. (Author)

Analysis of fuel operational reliability and fuel failures

International Nuclear Information System (INIS)

Smiesko, I.

1999-01-01

In this lecture the fuel failure (loss of fuel rod (cladding) integrity, corruption of second barrier for fission product release from duel and their consequences (increase of primary coolant activity; increase of fission product releases to environment; increase of rad-waste activities and potential increase of personnel exposure) are discussed

Fuel cells 101

Energy Technology Data Exchange (ETDEWEB)

Taylor, B.

2003-06-01

A capsule history of fuel cells is given, beginning with the first discovery in 1839 by William Grove, a Welsh judge who, when experimenting with electrolysis discovered that by re-combining the two components of electrolysis (water and oxygen) an electric charge was produced. A century later, in 1958, Francis Thomas Bacon, a British scientist demonstrated the first working fuel cell stack, a technology which was licensed and used in the Apollo spacecraft. In Canada, early research on the development of fuel cells was carried out at the University of Toronto, the Defence Research Establishment and the National Research Council. Most of the early work concentrated on alkaline and phosphoric acid fuel cells. In 1983, Ballard Research began the development of the electrolyte membrane fuel cell, which marked the beginning of Canada becoming a world leader in fuel cell technology development. The paper provides a brief account of how fuel cells work, describes the distinguishing characteristics of the various types of fuel cells (alkaline, phosphoric acid, molten-carbonate, solid oxide, and proton exchange membrane types) and their principal benefits. The emphasis is on proton exchange membrane fuel cells because they are the only fuel cell technology that is appropriate for providing primary propulsion power onboard a vehicle. Since vehicles are by far the greatest consumers of fossil fuels, it follows that proton exchange membrane fuel cells will have the greatest potential impact on both environmental matters and on our reliance on oil as our primary fuel. Various on-going and planned fuel cell demonstration projects are also described. 1 fig.

Comprehensive Assessment of Composition and Thermochemical Variability by High Resolution GC/QToF-MS and the Advanced Distillation-Curve Method as a Basis of Comparison for Reference Fuel Development.

Science.gov (United States)

Lovestead, Tara M; Burger, Jessica L; Schneider, Nico; Bruno, Thomas J

2016-12-15

Commercial and military aviation is faced with challenges that include high fuel costs, undesirable emissions, and supply chain insecurity that result from the reliance on petroleum-based feedstocks. The development of alternative gas turbine fuels from renewable resources will likely be part of addressing these issues. The United States has established a target for one billion gallons of renewable fuels to enter the supply chain by 2018. These alternative fuels will have to be very similar in properties, chemistry, and composition to existing fuels. To further this goal, the National Jet Fuel Combustion Program (a collaboration of multiple U.S. agencies under the auspices of the Federal Aviation Administration, FAA) is coordinating measurements on three reference gas turbine fuels to be used as a basis of comparison. These fuels are reference fuels with certain properties that are at the limits of experience. These fuels include a low viscosity, low flash point, high hydrogen content "best case" JP-8 (POSF 10264) fuel, a relatively high viscosity, high flash point, low hydrogen content "worst case" JP-5 (POSF 10259) fuel, and a Jet-A (POSF 10325) fuel with relatively average properties. A comprehensive speciation of these fuels is provided in this paper by use of high resolution gas chromatography/quadrupole time-of-flight - mass spectrometry (GC/QToF-MS), which affords unprecedented resolution and exact molecular formula capabilities. The volatility information as derived from the measurement of the advanced distillation curve temperatures, T k and T h , provides an approximation of the vapor liquid equilibrium and examination of the composition channels provides detailed insight into thermochemical data. A comprehensive understanding of the compositional and thermophysical data of gas turbine fuels is required not only for comparison but also for modeling of such complex mixtures, which will, in turn, aid in the development of new fuels with the goals of

Comprehensive Assessment of Composition and Thermochemical Variability by High Resolution GC/QToF-MS and the Advanced Distillation-Curve Method as a Basis of Comparison for Reference Fuel Development*

Science.gov (United States)

Lovestead, Tara M.; Burger, Jessica L.; Schneider, Nico; Bruno, Thomas J.

2018-01-01

Commercial and military aviation is faced with challenges that include high fuel costs, undesirable emissions, and supply chain insecurity that result from the reliance on petroleum-based feedstocks. The development of alternative gas turbine fuels from renewable resources will likely be part of addressing these issues. The United States has established a target for one billion gallons of renewable fuels to enter the supply chain by 2018. These alternative fuels will have to be very similar in properties, chemistry, and composition to existing fuels. To further this goal, the National Jet Fuel Combustion Program (a collaboration of multiple U.S. agencies under the auspices of the Federal Aviation Administration, FAA) is coordinating measurements on three reference gas turbine fuels to be used as a basis of comparison. These fuels are reference fuels with certain properties that are at the limits of experience. These fuels include a low viscosity, low flash point, high hydrogen content “best case” JP-8 (POSF 10264) fuel, a relatively high viscosity, high flash point, low hydrogen content “worst case” JP-5 (POSF 10259) fuel, and a Jet-A (POSF 10325) fuel with relatively average properties. A comprehensive speciation of these fuels is provided in this paper by use of high resolution gas chromatography/quadrupole time-of-flight – mass spectrometry (GC/QToF-MS), which affords unprecedented resolution and exact molecular formula capabilities. The volatility information as derived from the measurement of the advanced distillation curve temperatures, Tk and Th, provides an approximation of the vapor liquid equilibrium and examination of the composition channels provides detailed insight into thermochemical data. A comprehensive understanding of the compositional and thermophysical data of gas turbine fuels is required not only for comparison but also for modeling of such complex mixtures, which will, in turn, aid in the development of new fuels with the goals of

World nuclear fuel cycle requirements 1989

International Nuclear Information System (INIS)

1989-01-01

This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under two nuclear supply scenarios. These two scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries in the World Outside Centrally Planned Economic Areas (WOCA). A No New Orders scenarios is also presented for the Unites States. This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the WOCA projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel; discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2020 for the Lower and Upper Reference cases and through 2036, the last year in which spent fuel is discharged, for the No New Orders case

Advanced Fuel Cycle Cost Basis

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

2009-12-01

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

Advanced Fuel Cycle Cost Basis

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

2007-04-01

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

Advanced Fuel Cycle Cost Basis

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

2008-03-01

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

Estimating the cost of disposal for Canada's nuclear fuel waste

International Nuclear Information System (INIS)

Ates, Y.

1996-07-01

Atomic Energy of Canada Ltd (AECL) prepared an Environmental Impact Statement and nine supporting Primary Reference Documents on the concept for disposal of Canada's nuclear fuel waste. This report summarizes the basis of the cost estimate which is provided in the primary reference document on engineering for a disposal facility. The scope of the cost estimate is explained by describing the key features of the disposal facility design, by noting the major assumptions made in preparing the estimates, and by listing the included and excluded cost components. An activity-based project planning and control method is explained whereby the project schedule, costs, and personnel requirements are interlinked; forming an integrated perspective on the total project life cycle. The summary and distribution of costs in each project stage by major facility or activity are presented. The results of studies which reviewed the overall cost estimate are also described. These studies indicate that, within the scope, the estimate is reasonable and compares well with similar international studies. (author)

Using Checklists to Assess Your Transition to Alternative Fuels: A Technical Reference

Energy Technology Data Exchange (ETDEWEB)

Risch, C. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Santini, D. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, L. R. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-01

The Checklist for Transition to New Alternative Fuel(s) was published in September 2011 by Chuck Risch and Dan Santini. Many improvements, described below, have been incorporated into this current document, Checklists for Assessing the Transitions to New Highway Fuels.2 Further, the original authors and Larry Johnson, co-author of the current report, identified a need for a succinct version of the full report and prepared a brochure based on it to aid busy decisionmakers: Check It Out: Using Checklists to Assess Your Transition to Alternative Fuels.2 These checklists are tools for those stakeholders charged with determining a feasible alternative fuel or fuels for highway transportation systems of the future. The original had four major players whose needs had to be satisfied for a successful transition. The term “activist,” intended to encompass environmental and other special interests, was included in the “customers” category. Activists are customers of the government in the sense that they organize citizens to exert political pressure to regulate the design of vehicles, fuel infrastructure, and roadway networks. Many who evaluate alternative fuels view activists, particularly environmental activists, as a separate category. Further, “activist” has become a pejorative term to many people. Therefore, we have used the word “advocate” or “activist/advocate” instead. Thus, in this update we recognize that environmental and other activists/advocates have been--and will continue to be--a powerful force promoting change in the nature of the fuels that are used in transportation.

Nuclear fuel manufacture

International Nuclear Information System (INIS)

Costello, J.M.

1980-09-01

The technologies used to manufacture nuclear fuel from uranium ore are outlined, with particular reference to the light water reactor fuel cycle. Capital and operating cost estimates for the processing stages are given, and the relevance to a developing uranium industry in Australia is discussed

Fuel performance annual report for 1981

International Nuclear Information System (INIS)

Bailey, W.J.; Tokar, M.

1982-12-01

This annual report, the fourth in a series, provides a brief description of fuel performance during 1981 in commercial nuclear power plants. Brief summaries of fuel operating experience, fuel problems, fuel design changes and fuel surveillance programs, and high-burnup fuel experience are provided. References to additional, more detailed information and related NRC evaluations are included

Evaluation Of Radioactivity Concentration In The Primary Cooling Water System Of The RSG-GAS During Operation With 30% Silicide Fuels

International Nuclear Information System (INIS)

Hartoyo, Unggul; Udiyani, P.M.; Setiawanto, Anto

2001-01-01

The evaluating radioactivity concentration in the primary cooling water of the RSG-GAS during operation with 30% silicide fuels has been performed. The method of the research is sampling of primary cooling water during operation of the reactor and calculation of its radioactivity concentration. Based on the data obtained from calculation, the identified nuclides in the water are, Mn-56, Sb-124, Sb-122 and Na-24, under the limit of safety value

GDI fuel sprays of light naphtha, PRF95 and gasoline using a piezoelectric injector under different ambient pressures

KAUST Repository

Wu, Zengyang; Wang, Libing; Badra, Jihad A.; Roberts, William L.; Fang, Tiegang

2018-01-01

This study investigates fuel sprays of light naphtha (LN), primary reference fuel (PRF) and gasoline under different ambient pressures with an outwardly opening piezo gasoline direct injection (GDI) fuel injector. The tested gasoline fuel (regular grade with up to 10% ethanol, E10) was obtained by mixing fuels with AKI (the average of the research octane number (RON) and the motor octane number (MON)) of 87 from three local gas stations. Primary reference fuel (PRF) is commonly used as gasoline surrogate fuel and is blended by iso-octane and n-heptane. PRF95 is the blend of 95% iso-octane and 5% n-heptane by volume. LN fuel was provided by Saudi Aramco Oil Company. Five different ambient pressure conditions varied from 1 bar to 10 bar were tested. The spray was visualized by applying a Mie-scattering technique and a high-speed camera was employed to capture the spray images. The spray structure, spray angle, spray penetration length and spray front fluctuation were analyzed and compared among three fuels. Spray images show that a clear filamentary hollow-cone spray structure is formed for all three fuels at atmospheric conditions, and toroidal recirculation vortexes are observed at the downstream spray edges. A higher ambient pressure leads to a stronger vortex located closer to the injector outlet. Generally speaking, larger spray angles are found under higher ambient pressure conditions for all three fuels. Gasoline fuel always has the largest spray angle for each ambient pressure, while PRF95 has the smallest at most time. For each fuel, the spray front penetration length and spray front penetration velocity decrease with increasing ambient pressure. LN, PRF95 and gasoline show similar penetration length and velocity under the tested conditions. A two-stage spray front fluctuation pattern is observed for all three fuels. Stage one begins from the start of the injection and ends at 450–500 μs after the start of the injection trigger (ASOIT) with a slow

GDI fuel sprays of light naphtha, PRF95 and gasoline using a piezoelectric injector under different ambient pressures

KAUST Repository

Wu, Zengyang

2018-03-20

This study investigates fuel sprays of light naphtha (LN), primary reference fuel (PRF) and gasoline under different ambient pressures with an outwardly opening piezo gasoline direct injection (GDI) fuel injector. The tested gasoline fuel (regular grade with up to 10% ethanol, E10) was obtained by mixing fuels with AKI (the average of the research octane number (RON) and the motor octane number (MON)) of 87 from three local gas stations. Primary reference fuel (PRF) is commonly used as gasoline surrogate fuel and is blended by iso-octane and n-heptane. PRF95 is the blend of 95% iso-octane and 5% n-heptane by volume. LN fuel was provided by Saudi Aramco Oil Company. Five different ambient pressure conditions varied from 1 bar to 10 bar were tested. The spray was visualized by applying a Mie-scattering technique and a high-speed camera was employed to capture the spray images. The spray structure, spray angle, spray penetration length and spray front fluctuation were analyzed and compared among three fuels. Spray images show that a clear filamentary hollow-cone spray structure is formed for all three fuels at atmospheric conditions, and toroidal recirculation vortexes are observed at the downstream spray edges. A higher ambient pressure leads to a stronger vortex located closer to the injector outlet. Generally speaking, larger spray angles are found under higher ambient pressure conditions for all three fuels. Gasoline fuel always has the largest spray angle for each ambient pressure, while PRF95 has the smallest at most time. For each fuel, the spray front penetration length and spray front penetration velocity decrease with increasing ambient pressure. LN, PRF95 and gasoline show similar penetration length and velocity under the tested conditions. A two-stage spray front fluctuation pattern is observed for all three fuels. Stage one begins from the start of the injection and ends at 450–500 μs after the start of the injection trigger (ASOIT) with a slow

Simulation of fuel dispersion in the MYRRHA-FASTEF primary coolant with CFD and SIMMER-IV

Energy Technology Data Exchange (ETDEWEB)

Buckingham, Sophia, E-mail: sophia.buckingham@vki.ac.be [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Eboli, Marica [University of Pisa, Largo Lucio Lazzarino 2, 56122 Pisa (Italy); Moreau, Vincent [CRS4, Science and Technology Park Polaris – Piscina Manna, 09010 Pula (Italy); Van Tichelen, Katrien [SCK-CEN, Boeretang 200, 2400 Mol (Belgium)

2015-12-15

Highlights: • A comparison between CFD and system codes applied to long-term dispersion of fuel particles inside the MYRRHA reactor is proposed. • Important accumulations at the free-surface level are to be expected. • The risk of core blockage should not be neglected. • Numerical approach and modeling assumptions have a strong influence on the simulation results and accuracy. - Abstract: The objective of this work is to assess the behavior of fuel redistribution in heavy liquid metal nuclear systems under fuel pin failure conditions. Two different modeling approaches are considered using Computational Fluid Dynamics (CFD) codes and a system code, applied to the MYRRHA facility primary coolant loop version 1.4. Two different CFD models are constructed: the first is a single-phase steady model prepared in ANSYS Fluent, while the second is a two-phase model based on the volume of fluid (VOF) method in STARCCM+ to capture the upper free-surface dynamics. Both use a Lagrangian tracking approach with oneway coupling to follow the particles throughout the reactor. The system code SIMMER-IV is used for the third model, without neutronic coupling. Although limited regarding the fluid dynamic aspects compared to the CFD codes, comparisons of particle distributions highlight strong similarities despite quantitative discrepancies in the size of fuel accumulations. These disparities should be taken into account while performing the safety analysis of nuclear systems and developing strategies for accident mitigation.

Emissions from Diesel and Gasoline Vehicles Fuelled by Fischer-Tropsch Fuels and Similar Fuels

DEFF Research Database (Denmark)

Larsen, Ulrik; Lundorff, Peter; Ivarsson, Anders

2007-01-01

The described investigation was carried out under the umbrella of IEA Advanced Motor Fuels Agreement. The purpose was to evaluate the emissions of carbon monoxide (CO), unburned hydrocarbons (HC), nitrogen oxides (NOx), particulate matter (PM) and polycyclic aromatic hydrocarbons (PAH) from...... vehicles fuelled by Fischer Tropsch (FT) based diesel and gasoline fuel, compared to the emissions from ordinary diesel and gasoline. The comparison for diesel fuels was based on a literature review, whereas the gasoline comparison had to be based on our own experiments, since almost no references were...... found in this field. In this context measurement according to the Federal Test Procedure (FTP) and the New European Driving Cycle (NEDC) were carried out on a chassis dynamometer with a directly injected gasoline vehicle. Experiments were carried out with a reference fuel, a fuel based 70% on FT...

Investigation into fuel pin reshuffling options in PWR in-core fuel management for enhancement of efficient use of nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Daing, Aung Tharn, E-mail: atdaing@khu.ac.kr; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

2014-07-01

Highlights: • This paper discusses an alternative option, fuel pin reshuffling for maximization of cycle energy production. • The prediction results of isotopic compositions of each burnt pin are verified. • The operating performance is analyzed at equilibrium core with fuel pin reshuffling. • The possibility of reuse of spent fuel pins for reduction of fresh fuel assemblies is investigated. - Abstract: An alternative way to enhance efficient use of nuclear fuel is investigated through fuel pin reshuffling options within PWR fuel assembly (FA). In modeling FA with reshuffled pins, as prerequisite, the single pin calculation method is proposed to estimate the isotopic compositions of each pin of burnt FA in the core-wide environment. Subsequently, such estimation has been verified by comparing with the neutronic performance of the reference design. Two scenarios are concerned, i.e., first scenario was targeted on the improvement of the uniform flux spatial distribution and on the enhancement of neutron economy by simply reshuffling the existing fuel pins in once-burnt fuel assemblies, and second one was focused on reduction of fresh fuel loading and discharged fuel assemblies with more economic incentives by reusing some available spent fuel pins still carrying enough reactivity that are mechanically sound ascertained. In scenario-1, the operating time was merely somewhat increased for few minutes when treating eight FAs by keeping enough safety margins. The scenario-2 was proved to reduce four fresh FAs loading without largely losing any targeted parameters from the safety aspect despite loss of 14 effective full power days for operation at reference plant full rated power.

Closed Fuel Cycle Waste Treatment Strategy

Energy Technology Data Exchange (ETDEWEB)

Vienna, J. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Collins, E. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Crum, J. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, S. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Garn, T. G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gombert, D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maio, V. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Matyas, J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Nenoff, T. M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Riley, B. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sevigny, G. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Thallapally, P. K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, J. H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-02-01

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form

Nuclear primary energy carriers. Short version

Energy Technology Data Exchange (ETDEWEB)

Jaeck, W

1978-04-01

Basing on our present knowledge the following energy sources for energy supply must be taken into consideration in the long term: regenerative energy sources, fission energy gained by breeder reactors, nuclear fusion. While regenerative energy sources were treated at full length in the study 'Energy Sources for Tomorrow' the present study specifies the other two energy options. The availability and the reliability of nuclear primary energy carrier supply is described in detail and the conversion systems available or still being developed are investigated with regard to their specific consumption of primary energy. Topical questions concerning the proliferation stability of the fuel cycles and techniques are subject to the INFCE program. With reference to the nuclear energy documentation activities of the Federal Government this study is supposed to supply further fundamental material on nuclear primary energy carriers, consumption and readiness for application. Thus it will contribute to the question: 'Is nuclear energy an option which guarantees energy supply in the long term for the Federal Republic of Germany'. (orig.) 891 UA 892 ARA.

Nuclear primary energy carriers. Pt. 1

International Nuclear Information System (INIS)

1978-04-01

Basing on our present knowledge the following energy sources for energy supply must be taken into consideration in the long term: regenerative energy sources, fission energy gained by breeder reactors, nuclear fusion. While regenerative energy sources were treated at full length in the study 'Energy Sources for tomorrow' the present study specifies the other two energy options. The availability and the reliability of nuclear primary energy carrier supply is described in detail and the conversion systems available or still being developed are investigated with regard to their specific consumption of primary energy. Topical questions concerning the proliferation stability of the fuel cycles and techniques are subject to the INFCE programme. With reference to the nuclear energy documentation activities of the Federal Govenment this study is supposed to supply further fundamental material on nuclear primary energy carriers, consumption and readiness for application. Thus it will contribute to the question: 'Is nuclear energy an option which guarantees energy supply in the long term for the Federal Republic of Germany'. (orig.) [de

Fuel performance annual report for 1990

International Nuclear Information System (INIS)

Preble, E.A.; Painter, C.L.; Alvis, J.A.; Berting, F.M.; Beyer, C.E.; Payne, G.A.; Wu, S.L.

1993-11-01

This annual report, the thirteenth in a series, provides a brief description of fuel performance during 1990 in commercial nuclear power plants. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience and trends, fuel problems high-burnup fuel experience, and items of general significance are provided . References to additional, more detailed information, and related NRC evaluations are included where appropriate

An investigation into fuel pulverization with specific reference to high burn-up LOCA

International Nuclear Information System (INIS)

Yagnik, Suresh; Turnbull, James; Noirot, Jean; Walker, Clive; Hallstadius, Lars; Waeckel, N.; Blanpain, P.

2014-01-01

To investigate the phenomenon of high burn-up fuel pellet material potentially disintegrating into powder under a rapid temperature transient, such as in a LOCA-type accident scenario, two independent scoping studies were commissioned. The first was to investigate the effect of hydrostatic restraint pressure on Fission Gas Release (FGR) from small samples of highly irradiated fuel (71 MWd/kgU) during a series of rapid temperature ramps. Experimentally, when the FGR increased rapidly during the temperature transients, the fuel was assumed to be 'pulverized', i.e., fragmented into powder. In the second series of experiments, laser heating of small samples was used to investigate the temperature at which fuel pulverization was initiated. Subsequent to fuel disintegration, there was always a spectrum of particle sizes present. The significance of this observation was recognized in the context of extended burn-up operation in commercial reactors. Based on the observation from these investigations, a fuel fragmentation threshold has been discussed and developed. We conclude that fuel disintegration could be of potential importance in limiting the performance and productive lifetime of nuclear fuel. However, since only fuel closely adjacent to ballooned or ruptured cladding would be released in a LOCA-type transient, expulsion of pulverized fuel from the ruptured fuel rod is not considered a safety issue; cooling of the defected assembly remains possible and there is no issue with respect to local criticality. (author)

Fuel cells: Project Volta

Energy Technology Data Exchange (ETDEWEB)

Vellone, R.; Di Mario, F.

1987-09-01

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

Evaluating primary forest fuel rail terminals with discrete event simulation: A case study from Austria

OpenAIRE

Ulrich J. Wolfsmayr; Rossana Merenda; Peter Rauch; Francesco Longo; Manfred Gronalt

2016-01-01

Biomass rail transport is a useful alternative to unimodal truck transport for medium or long transport distances, if only a short road pre-haulage is required. Up to now primary forest fuels (PFF) are rarely transported on the rail network in Austria and rail terminals able to tranship notable volumes are not established yet. The objective of this study is to investigate the potentials of existing transhipment infrastructure for introducing and operating PFF terminals. Such PFF terminals ena...

The Kjeldahl method as a primary reference procedure for total protein in certified reference materials used in clinical chemistry. I. A review of Kjeldahl methods adopted by laboratory medicine.

Science.gov (United States)

Chromý, Vratislav; Vinklárková, Bára; Šprongl, Luděk; Bittová, Miroslava

2015-01-01

We found previously that albumin-calibrated total protein in certified reference materials causes unacceptable positive bias in analysis of human sera. The simplest way to cure this defect is the use of human-based serum/plasma standards calibrated by the Kjeldahl method. Such standards, commutative with serum samples, will compensate for bias caused by lipids and bilirubin in most human sera. To find a suitable primary reference procedure for total protein in reference materials, we reviewed Kjeldahl methods adopted by laboratory medicine. We found two methods recommended for total protein in human samples: an indirect analysis based on total Kjeldahl nitrogen corrected for its nonprotein nitrogen and a direct analysis made on isolated protein precipitates. The methods found will be assessed in a subsequent article.

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Fuel performance annual report for 1989

International Nuclear Information System (INIS)

Bailey, W.J.; Berting, F.M.; Wu, S.

1992-06-01

This annual report, the twelfth in a series, provides a brief description of fuel performance during 1989 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related US Nuclear Regulatory Commission evaluations are included

Nuclear Fuel Reprocessing

International Nuclear Information System (INIS)

Simpson, Michael F.; Law, Jack D.

2010-01-01

This is a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. Nuclear reprocessing is the chemical treatment of spent fuel involving separation of its various constituents. Principally, it is used to recover useful actinides from the spent fuel. Radioactive waste that cannot be re-used is separated into streams for consolidation into waste forms. The first known application of nuclear reprocessing was within the Manhattan Project to recover material for nuclear weapons. Currently, reprocessing has a peaceful application in the nuclear fuel cycle. A variety of chemical methods have been proposed and demonstrated for reprocessing of nuclear fuel. The two most widely investigated and implemented methods are generally referred to as aqueous reprocessing and pyroprocessing. Each of these technologies is described in detail in Section 3 with numerous references to published articles. Reprocessing of nuclear fuel as part of a fuel cycle can be used both to recover fissionable actinides and to stabilize radioactive fission products into durable waste forms. It can also be used as part of a breeder reactor fuel cycle that could result in a 14-fold or higher increase in energy utilization per unit of natural uranium. Reprocessing can also impact the need for geologic repositories for spent fuel. The volume of waste that needs to be sent to such a repository can be reduced by first subjecting the spent fuel to reprocessing. The extent to which volume reduction can occur is currently under study by the United States Department of Energy via research at various national laboratories and universities. Reprocessing can also separate fissile and non-fissile radioactive elements for transmutation.

Neptunium: a bibliographic reference

International Nuclear Information System (INIS)

Mosley, R.E.

1979-06-01

A comprehensive bibliograhy of the literature on the element neptunium published prior to January 1976 is presented. A short abstract is given for each listed reference, with a few exceptions. The references are divided into sections categorized as General, Man-Made Sources (Reactors), Man-Made Sources (Fuel Reprocessing), Chemistry (Solubility), Chemistry (Compounds), Chemistry (Isotopes), Analyses (Instrumental), Analyses (Chemical), Chemical (Animal), Biological (Effects), Biological (Animal-Metabolism-Retention), Biological (Air Movement), Biological (Human Inhalation), Measurement, and Dosimetry. The bibliography contains author and keyword indexes and was compiled to serve as a quick reference source for neptunium-related work. 184 citations

Nuclear fuel production

International Nuclear Information System (INIS)

Randol, A.G.

1985-01-01

The production of new fuel for a power plant reactor and its disposition following discharge from the power plant is usually referred to as the ''nuclear fuel cycle.'' The processing of fuel is cyclic in nature since sometime during a power plant's operation old or ''depleted'' fuel must be removed and new fuel inserted. For light water reactors this step typically occurs once every 12-18 months. Since the time required for mining of the raw ore to recovery of reusable fuel materials from discharged materials can span up to 8 years, the management of fuel to assure continuous power plant operation requires simultaneous handling of various aspects of several fuel cycles, for example, material is being mined for fuel to be inserted in a power plant 2 years into the future at the same time fuel is being reprocessed from a discharge 5 years prior. Important aspects of each step in the fuel production process are discussed

Biodegradation of biodiesel fuels

International Nuclear Information System (INIS)

Zhang, X.; Haws, R.; Wright, B.; Reese, D.; Moeller, G.; Peterson, C.

1995-01-01

Biodiesel fuel test substances Rape Ethyl Ester (REE), Rape Methyl Ester (RME), Neat Rape Oil (NR), Say Methyl Ester (SME), Soy Ethyl Ester (SEE), Neat Soy Oil (NS), and proportionate combinations of RME/diesel and REE/diesel were studied to test the biodegradability of the test substances in an aerobic aquatic environment using the EPA 560/6-82-003 Shake Flask Test Method. A concurrent analysis of Phillips D-2 Reference Diesel was also performed for comparison with a conventional fuel. The highest rates of percent CO 2 evolution were seen in the esterified fuels, although no significant difference was noted between them. Ranges of percent CO 2 evolution for esterified fuels were from 77% to 91%. The neat rape and neat soy oils exhibited 70% to 78% CO 2 evolution. These rates were all significantly higher than those of the Phillips D-2 reference fuel which evolved from 7% to 26% of the organic carbon to CO 2 . The test substances were examined for BOD 5 and COD values as a relative measure of biodegradability. Water Accommodated Fraction (WAF) was experimentally derived and BOD 5 and COD analyses were carried out with a diluted concentration at or below the WAF. The results of analysis at WAF were then converted to pure substance values. The pure substance BOD 5 and COD values for test substances were then compared to a control substance, Phillips D-2 Reference fuel. No significant difference was noted for COD values between test substances and the control fuel. (p > 0.20). The D-2 control substance was significantly lower than all test substances for BCD, values at p 5 value

Preparation for commissioning of nuclear plant with reference to British Nuclear Fuels Plc fuel handling plant project

International Nuclear Information System (INIS)

Bamber, D.R.

1987-01-01

The new Fuel Handling Plant at British Nuclear Fuels Sellafield is part of a Pound 550M complex which provides facilities for the receipt, storage and mechanical preparation of both Magnox and A.G.R. fuel. The plant is very large and complex with considerable use of computer based process control systems, providing for physical and nuclear safety. The preparation of such plant for active commissioning necessitates a great many physical checks and technical evaluations in support of its safety case. This paper describes arrangements for plant commissioning checks, against the regulatory framework and explains the physical preparations necessary for their timely accomplishment. (author)

Preparation for commissioning of nuclear plant with reference to British Nuclear Fuels Plc fuel handling plant project

International Nuclear Information System (INIS)

Bamber, D.

1987-01-01

The new Fuel Handing Plant at British Nuclear Fuels Sellafield is part of a Pound 550M complex which provides facilities for the receipt, storage and mechanical preparation of both magnox and A.G.R. fuel. The plant is very large and complex with considerable use of computer based process control systems, providing for physical and nuclear safety. The preparation of such plant for ''active'' commissioning necessitates a great many physical checks and technical evaluations in support of its safety case. This paper describes arrangements for plant commissioning checks, against the regulatory framework and explains the physical preparations necessary for their timely accomplishment. (author)

A study on the direct use of spent PWR fuel in CANDU reactors -Fuel management and safety analysis-

Energy Technology Data Exchange (ETDEWEB)

Park, Hyun Soo; Lee, Boh Wook; Choi, Hang Bok; Lee, Yung Wook; Cho, Jae Sun; Huh, Chang Wook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

The reference DUPIC fuel composition was determined based on the reactor safety, thermal-hydraulics, economics, and refabrication aspects. The center pin of the reference DUPIC fuel bundle is poisoned with natural dysprosium. The worst LOCA analysis has shown that the transient power and heat deposition of the reference DUPIC core are the same as those of natural uranium CANDU core. The intra-code comparison has shown that the accuracy of DUPIC physics code system is comparable to the current CANDU core design code system. The sensitivity studies were performed for the refuelling schemes of DUPIC core and the 2-bundle shift refuelling scheme was selected as the standard refuelling scheme of the DUPIC core. The application of 4-bundle shift refuelling scheme will be studied in parallel as the auto-refuelling method is improved and the reference core parameters of the heterogeneous DUPIC core are defined. The heterogeneity effect was analyzed in a preliminary fashion using 33 fuel types and the random loading strategy. The refuelling simulation has shown that the DUPIC core satisfies the current CANDU 6 operating limits of channel and bundle power regardless of the fuel composition heterogeneity. The 33 fuel types used in the heterogeneity analysis was determined based on the initial enrichment and discharge burnup of the PWR fuel. 90 figs, 62 tabs, 63 refs. (Author).

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

International Nuclear Information System (INIS)

Huertas, F.; Ulibarri, A.

1993-01-01

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

Psychopathology of the General Population Referred by Primary Care Physicians for Urgent Assessment in Psychiatric Hospitals

Directory of Open Access Journals (Sweden)

Judith McLenan

2016-12-01

Full Text Available Objective: The aim of this study was to evaluate the type, severity and progression of psychiatric pathologies in a sample of 372 outpatients (age range 18–65 years referred by their primary general practitioners (GPs to an Urgent Referral Team (URT based in a psychiatric hospital in Aberdeen, Scotland. This team offers immediate appointments (1- to 7-day delays for rapid assessments and early interventions to the outpatients referred by their primary family doctors.Method: One-sample t-test and z statistic were used for data analysis. From the total population, a convenience sample of 40 people was selected and assessed to evaluate whether follow-up appointments after the first visit could reduce the severity of suicidal ideation, depression and anxiety in the outpatients seen by the URT. A two-sample t-test and a Wilcoxon signed-rank test were used to assess the variations in the scores during the follow-up visits.Results: We found a statistically significant prevalence of depressive disorders, comorbid with anxiety at first presentation in people who were females, white, never married, living with a partner, not studying and not in paid employment. The common presentation of borderline personality disorder and dysthymia in this population underscores its vulnerability to major socioeconomic challenges.Conclusion: The data confirmed the impact that primary care cooperation with psychiatric hospitals can have on the psychiatric system, and as a reflection, on the population’s mental health and well-being. In fact, active cooperation and early diagnosis and intervention will help detect cases at risk in the general population and reduce admissions into hospitals.

Theoretical study of the effects of pilot fuel quantity and its injection timing on the performance and emissions of a dual fuel diesel engine

International Nuclear Information System (INIS)

Papagiannakis, R.G.; Hountalas, D.T.; Rakopoulos, C.D.

2007-01-01

Various solutions have been proposed for improving the combustion process of conventional diesel engines and reducing the exhaust emissions without making serious modifications on the engine, one of which is the use of natural gas as a supplement for the conventional diesel fuel, the so called dual fuel natural gas diesel engines. The most common type of these is referred to as the pilot ignited natural gas diesel engine (PINGDE). Here, the primary fuel is natural gas that controls the engine power output, while the pilot diesel fuel injected near the end of the compression stroke auto-ignites and creates ignition sources for the surrounding gaseous fuel mixture to be burned. Previous research studies have shown that the main disadvantage of this dual fuel combustion is its negative impact on engine efficiency compared to the normal diesel operation, while carbon monoxide emissions are also increased. The pilot diesel fuel quantity and injection advance influence significantly the combustion mechanism. Then, in order to examine the effect of these two parameters on the performance and emissions, a comprehensive two-zone phenomenological model is employed and applied on a high-speed, pilot ignited, natural gas diesel engine located at the authors' laboratory. According to the results, the simultaneously increase of the pilot fuel quantity accompanied with an increase of its injection timing results to an improvement of the engine efficiency (increase) and of the emitted CO emissions (decrease) while it has a negative effect (increase) of NO emissions

Ignition of alkane-rich FACE gasoline fuels and their surrogate mixtures

KAUST Repository

Sarathy, Mani

2015-01-01

Petroleum derived gasoline is the most used transportation fuel for light-duty vehicles. In order to better understand gasoline combustion, this study investigated the ignition propensity of two alkane-rich FACE (Fuels for Advanced Combustion Engines) gasoline test fuels and their corresponding PRF (primary reference fuel) blend in fundamental combustion experiments. Shock tube ignition delay times were measured in two separate facilities at pressures of 10, 20, and 40 bar, temperatures from 715 to 1500 K, and two equivalence ratios. Rapid compression machine ignition delay times were measured for fuel/air mixtures at pressures of 20 and 40 bar, temperatures from 632 to 745 K, and two equivalence ratios. Detailed hydrocarbon analysis was also performed on the FACE gasoline fuels, and the results were used to formulate multi-component gasoline surrogate mixtures. Detailed chemical kinetic modeling results are presented herein to provide insights into the relevance of utilizing PRF and multi-component surrogate mixtures to reproduce the ignition behavior of the alkane-rich FACE gasoline fuels. The two FACE gasoline fuels and their corresponding PRF mixture displayed similar ignition behavior at intermediate and high temperatures, but differences were observed at low temperatures. These trends were mimicked by corresponding surrogate mixture models, except for the amount of heat release in the first stage of a two-stage ignition events, when observed. © 2014 The Combustion Institute.

Rethinking fat as a fuel for endurance exercise.

Science.gov (United States)

Volek, Jeff S; Noakes, Timothy; Phinney, Stephen D

2015-01-01

A key element contributing to deteriorating exercise capacity during physically demanding sport appears to be reduced carbohydrate availability coupled with an inability to effectively utilize alternative lipid fuel sources. Paradoxically, cognitive and physical decline associated with glycogen depletion occurs in the presence of an over-abundance of fuel stored as body fat that the athlete is apparently unable to access effectively. Current fuelling tactics that emphasize high-carbohydrate intakes before and during exercise inhibit fat utilization. The most efficient approach to accelerate the body's ability to oxidize fat is to lower dietary carbohydrate intake to a level that results in nutritional ketosis (i.e., circulating ketone levels >0.5 mmol/L) while increasing fat intake for a period of several weeks. The coordinated set of metabolic adaptations that ensures proper interorgan fuel supply in the face of low-carbohydrate availability is referred to as keto-adaptation. Beyond simply providing a stable source of fuel for the brain, the major circulating ketone body, beta-hydroxybutyrate, has recently been shown to act as a signalling molecule capable of altering gene expression, eliciting complementary effects of keto-adaptation that could extend human physical and mental performance beyond current expectation. In this paper, we review these new findings and propose that the shift to fatty acids and ketones as primary fuels when dietary carbohydrate is restricted could be of benefit for some athletes.

Fuels Performance | Transportation Research | NREL

Science.gov (United States)

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

New techniques for the characterization of refuse-derived fuels and solid recovered fuels.

Science.gov (United States)

Rotter, Vera Susanne; Lehmann, Annekatrin; Marzi, Thomas; Möhle, Edda; Schingnitz, Daniel; Hoffmann, Gaston

2011-02-01

Solid recovered fuel (SRF) today refers to a waste-derived fuel meeting defined quality specifications, in terms of both origin (produced from non-hazardous waste) and levels of certain fuel properties. Refuse-derived fuel (RDF) nowadays is more used for unspecified waste after a basic processing to increase the calorific value and therefore this term usually refers to the segregated, high calorific fraction of municipal solid waste (MSW), commercial or industrial wastes. In comparison with conventional fuels, both types of secondary fuel show waste of inherently varying quality and an increased level of waste-specific contaminants.The transition from RDF to SRF in the emerging national and European market requires a quality assurance system with defined quality parameters and analytical methods to ensure reliable fuel characterization. However, due to the quality requirements for RDF and SRF, the current standardized analysis methods often do not meet these practical demands. Fast test methods, which minimize personnel, financial and time efforts and which are applicable for producers as well as users can be an important supporting tool for RDF- and SRF-characterization. Currently, a fast test system based on incineration and correlation analyses which enable the determination of relevant fuel parameters is under development. Fast test methods are not aimed at replacing current standardized test methods, but have to be considered as practical supporting tools for the characterization of RDF and SRF.

Fuel performance annual report for 1986

International Nuclear Information System (INIS)

Bailey, W.J.; Wu, S.

1988-03-01

This annual report, the ninth in a series, provides a brief description of fuel performance during 1986 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related U.S. Nuclear Regulatory Commission evaluations are included. 550 refs., 12 figs., 31 tabs

Fuel performance: Annual report for 1987

International Nuclear Information System (INIS)

Bailey, W.J.; Wu, S.

1989-03-01

This annual report, the tenth in a series, provides a brief description of fuel performance during 1987 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related US Nuclear Regulator Commission evaluations are included. 384 refs., 13 figs., 33 tabs

Descriptions of reference LWR facilities for analysis of nuclear fuel cycles. Appendixes

International Nuclear Information System (INIS)

Schneider, K.J.; Kabele, T.J.

1979-09-01

The appendixes present the calculations that were used to derive the release factors discussed for each fuel cycle facility in Volume I. Appendix A presents release factor calculations for a surface mine, underground mine, milling facility, conversion facility, diffusion enrichment facility, fuel fabrication facility, PWR, BWR, and reprocessing facility. Appendix B contains additional release factors calculated for a BWR, PWR, and a reprocessing facility. Appendix C presents release factors for a UO 2 fuel fabrication facility

Assessment of ethanol-fueled IMHEX{reg_sign} fuel cell power plants in distributed generation

Energy Technology Data Exchange (ETDEWEB)

Woods, R. [M-C Power Corp., Burr Ridge, IL (United States); Lefeld, J. [PSI Energy, Plainfield, IN (United States)

1993-12-31

Ethanol-fueled cell power plants presents several significant opportunities for the power generation industry. The potential exists to reduce pollution, help the nation shift from its dependence on imported fuels, reduce global warming, and strengthen the economy. Two important developments can be merged to create a clean, high-technology, bio-based energy system: the production of ethanol fuels and the application of fuel cell power plants. Utilization of ethanol will be in dual-fueled applications initially, and evolve toward the primary fuel as the need for renewable energy sources increase and the economic competitiveness improves. This assessment addresses the major issues of this proposed concept and outlines the benefits anticipated to the environment, US agriculture, energy supplies, and electric power customers. Economic and technical aspects of the concept are also reviewed. One of PSI Energy`s primary interests is the utilization of renewable fuels supplied by their customer base. The IMHEX{reg_sign} fuel cell is an advanced electric power generation technology currently under development by M-C Power. Commercial applications within the power generation industry are scheduled to begin during the late 1990s.

Waste as a fuel

Energy Technology Data Exchange (ETDEWEB)

Lowes, T. M.; Lorimer, A. D.

1979-07-01

The methods of using the energy available in wastes to reduce the energy costs in processes are discussed. Special reference is made to the need for careful evaluation of the potential of the waste to reduce energy costs, before significant investment is made. Problems generally arise due to the effective balancing of the cost of the waste pretreatment with the disposal fee and prime fuel saving. Special reference is made to use of waste as a fuel in the cement industry. Municipal refuse is discussed as a typical successful application.

Analysis of performance losses of direct ethanol fuel cells with the aid of a reference electrode

Science.gov (United States)

Li, Guangchun; Pickup, Peter G.

The performances of direct ethanol fuel cells with different anode catalysts, different ethanol concentrations, and at different operating temperatures have been studied. The performance losses of the cell have been separated into individual electrode performance losses with the aid of a reference electrode, ethanol crossover has been quantified, and CO 2 and acetic acid production have been measured by titration. It has been shown that the cell performance strongly depends on the anode catalyst, ethanol concentration, and operating temperature. It was found that the cathode and anode exhibit different dependences on ethanol concentration and operating temperature. The performance of the cathode is very sensitive to the rate of ethanol crossover. Product analysis provides insights into the mechanisms of electro-oxidation of ethanol.

Analysis of performance losses of direct ethanol fuel cells with the aid of a reference electrode

Energy Technology Data Exchange (ETDEWEB)

Li, Guangchun; Pickup, Peter G. [Department of Chemistry, Memorial University of Newfoundland, Elizabeth Avenue, St. John' s, Newfoundland (Canada A 1B 3X7)

2006-10-20

The performances of direct ethanol fuel cells with different anode catalysts, different ethanol concentrations, and at different operating temperatures have been studied. The performance losses of the cell have been separated into individual electrode performance losses with the aid of a reference electrode, ethanol crossover has been quantified, and CO{sub 2} and acetic acid production have been measured by titration. It has been shown that the cell performance strongly depends on the anode catalyst, ethanol concentration, and operating temperature. It was found that the cathode and anode exhibit different dependences on ethanol concentration and operating temperature. The performance of the cathode is very sensitive to the rate of ethanol crossover. Product analysis provides insights into the mechanisms of electro-oxidation of ethanol. (author)

Operation experience of the advanced fuel assemblies at Unit 1 of Volgodonsk NPP within four fuel cycles

International Nuclear Information System (INIS)

Vasilchenko, I.; Kobelev, S.; Kushmanov, S.

2006-01-01

The first commissioning of Volgodonsk NPP Unit 1 with standard reactor WWER-1000 (project V-320) was in 2001. The reactor core, starting from the first fuel charge, was arranged completely with Advanced Fuel Assemblies (AFAs). In this way, it is possible to obtain the experience in startup and operation of the core, completely arranged with AFAs, and also to get a possibility of performing the comprehensive check for justification of newly commissioned units and justification of design solutions accepted in the design of reactor core for Taiwan NPP, Bushehr NPP and Kudankulam NPP. The first fuel charge of the Volgodonsk NPP Unit 1 is a reference and unified for Tiawan NPP (V-428), Bushehr NPP (V-446), Kudankulam NPP(V-412) with small differences caused by design features of RP V-320. The first core charge of Unit 1 of Volgodonsk NPP was arranged of 163 AFAs, comprising 61 CPS ARs and 42 BAR bundles. The subsequent fuel charges were arranged of AFAs with gadolinium oxide integrated into fuel instead of BAR. By 2005 the results of operation of the core at Unit 1 of Volgodonsk NPP during four fuel cycles showed that AFA is sufficiently reliable and serviceable. The activity of the primary coolant of the Volgodonsk NPP is at stable low level. During the whole time of the core operation of the Volgodonsk NPP Unit 1 no leaky AFAs were revealed. The modifications of the internals, made during pre-operational work, are reasonable and effective to provide for fuel mechanical stability in the course of operation. The modifications, made in AFA structure during operation of the Volgodonsk NPP Unit 1, are aimed at improving the service and operational reliability of its components. Correctness of the solutions taken is confirmed by AFAs operation experience both at the Volgodonsk NPP, and at other operating Russian NPPs

HTGR fuel and fuel cycle technology

International Nuclear Information System (INIS)

Lotts, A.L.; Homan, F.J.; Balthesen, E.; Turner, R.F.

1977-01-01

Significant advances have occurred in the development of HTGR fuel and fuel cycle. These accomplishments permit a wide choice of fuel designs, reactor concepts, and fuel cycles. Fuels capable of providing helium outlet temperatures of 750 0 C are available, and fuels capable of 1000 0 C outlet temperatures may be expected from extension of present technology. Fuels have been developed for two basic HTGR designs, one using a spherical (pebble bed) element and the other a prismatic element. Within each concept a number of variations of geometry, fuel composition, and structural materials are permitted. Potential fuel cycles include both low-enriched and high-enriched Th- 235 U, recycle Th- 233 U, and Th-Pu or U-Pu cycles. This flexibility offered by the HTGR is of great practical benefit considering the rapidly changing economics of power production. The inflation of ore prices has increased optimum conversion ratios, and increased the necessity of fuel recycle at an early date. Fuel element makeup is very similar for prismatic and spherical designs. Both use spherical fissile and fertile particles coated with combinations of pyrolytic carbon and silicon carbide. Both use carbonaceous binder materials, and graphite as the structural material. Weak-acid resin (WAR) UO 2 -UC 2 fissile fuels and sol-gel-derived ThO 2 fertile fuels have been selected for the Th- 233 U cycle in the prismatic design. Sol-gel-derived UO 2 UC 2 is the reference fissile fuel for the low-enriched pebble bed design. Both the United States and Federal Republic of Germany are developing technology for fuel cycle operations including fabrication, reprocessing, refabrication, and waste handling. Feasibility of basic processes has been established and designs developed for full-scale equipment. Fuel and fuel cycle technology provide the basis for a broad range of applications of the HTGR. Extension of the fuels to higher operating temperatures and development and commercial demonstration of fuel

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

International Nuclear Information System (INIS)

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

1980-10-01

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

Preliminary LEU fuel cycle analyses for the Belgian BR2 reactor

International Nuclear Information System (INIS)

Deen, J.R.; Snelgrove, J.L.

1986-01-01

Fuel cycle calculations have been performed with reference HEU fuel and LEU fuel using Cd wires or boron as burnable absorbers. The 235 U content in the LEU element has increased 20% to 480g compared to the reference HEU element. The number of fuel plates has remained unchanged while the fuel meat thickness has increased to 0.76 mm from 0.51 mm. The LEU meat density is 5.1 Mg U/m 3 . The reference fuel cycle was a 31 element core operating at 56 MW with a 19.8 day cycle length and eight fresh elements loaded per cycle. Comparable fuel cycle characteristics can be achieved using the proposed LEU fuel element with either Cd wires or boron burnable absorbers. The neutron flux for E/sub n/ > 1 eV changes very little (<5%) in LEU relative to HEU cores. Thermal flux reductions are 5 to 10% in non-fueled positions, and 20 to 30% in fuel elements

Introduction to selected references on fossil fuels of the central and southern Appalachian basin: Chapter H.1 in Coal and petroleum resources in the Appalachian basin: distribution, geologic framework, and geochemical character

Science.gov (United States)

Ruppert, Leslie F.; Lentz, Erika E.; Tewalt, Susan J.; Román Colón, Yomayra A.; Ruppert, Leslie F.; Ryder, Robert T.

2014-01-01

The Appalachian basin contains abundant coal and petroleum resources that have been studied and extracted for at least 150 years. In this volume, U.S. Geological Survey (USGS) scientists describe the geologic framework and geochemical character of the fossil-fuel resources of the central and southern Appalachian basin. Separate subchapters (some previously published) contain geologic cross sections; seismic profiles; burial history models; assessments of Carboniferous coalbed methane and Devonian shale gas; distribution information for oil, gas, and coal fields; data on the geochemistry of natural gas and oil; and the fossil-fuel production history of the basin. Although each chapter and subchapter includes references cited, many historical or other important references on Appalachian basin and global fossil-fuel science were omitted because they were not directly applicable to the chapters.

Long-term storage of spent nuclear fuel

International Nuclear Information System (INIS)

Kempe, T.F.; Martin, A.; Thorne, M.C.

1980-06-01

This report presents the results of a study on the storage of spent nuclear fuel, with particular reference to the options which would be available for long-term storage. Two reference programmes of nuclear power generation in the UK are defined and these are used as a basis for the projection of arisings of spent fuel and the storage capacity which might be needed. The characteristics of spent fuel which are relevant to long-term storage include the dimensions, materials and physical construction of the elements, their radioactive inventory and the associated decay heating as a function of time after removal from the reactor. Information on the behaviour of spent fuel in storage ponds is reviewed with particular reference to the corrosion of the cladding. The review indicates that, for long-term storage, both Magnox and AGR fuel would need to be packaged because of the high rate of cladding corrosion and the resulting radiological problems. The position on PWR fuel is less certain. Experience of dry storage is less extensive but it appears that the rate of corrosion of cladding is much lower than in water. Unit costs are discussed. Consideration is given to the radiological impact of fuel storage. (author)

Effect of engine parameters and type of gaseous fuel on the performance of dual-fuel gas diesel engines. A critical review

Energy Technology Data Exchange (ETDEWEB)

Sahoo, B.B. [Centre for Energy, Indian Institute of Technology, Guwahati 781039 (India); Sahoo, N.; Saha, U.K. [Department of Mechanical Engineering, Indian Institute of Technology, Guwahati 781039 (India)

2009-08-15

Petroleum resources are finite and, therefore, search for their alternative non-petroleum fuels for internal combustion engines is continuing all over the world. Moreover gases emitted by petroleum fuel driven vehicles have an adverse effect on the environment and human health. There is universal acceptance of the need to reduce such emissions. Towards this, scientists have proposed various solutions for diesel engines, one of which is the use of gaseous fuels as a supplement for liquid diesel fuel. These engines, which use conventional diesel fuel and gaseous fuel, are referred to as 'dual-fuel engines'. Natural gas and bio-derived gas appear more attractive alternative fuels for dual-fuel engines in view of their friendly environmental nature. In the gas-fumigated dual-fuel engine, the primary fuel is mixed outside the cylinder before it is inducted into the cylinder. A pilot quantity of liquid fuel is injected towards the end of the compression stroke to initiate combustion. When considering a gaseous fuel for use in existing diesel engines, a number of issues which include, the effects of engine operating and design parameters, and type of gaseous fuel, on the performance of the dual-fuel engines, are important. This paper reviews the research on above issues carried out by various scientists in different diesel engines. This paper touches upon performance, combustion and emission characteristics of dual-fuel engines which use natural gas, biogas, producer gas, methane, liquefied petroleum gas, propane, etc. as gaseous fuel. It reveals that 'dual-fuel concept' is a promising technique for controlling both NO{sub x} and soot emissions even on existing diesel engine. But, HC, CO emissions and 'bsfc' are higher for part load gas diesel engine operations. Thermal efficiency of dual-fuel engines improve either with increased engine speed, or with advanced injection timings, or with increased amount of pilot fuel. The ignition

Transportation of nuclear fuel

International Nuclear Information System (INIS)

Prowse, D.R.

1979-01-01

Shipment of used fuel from nuclear reactors to a central fuel management facility is discussed with particular emphasis on the assessment of the risk to the public due to these shipments. The methods of transporting used fuel in large shipping containers is reviewed. In terms of an accident scenario, it is demonstrated that the primary risk of transport of used fuel is due to injury and death in common road accidents. The radiological nature of the used fuel cargo is, for all practical purposes, an insignificant factor in the total risk to the public. (author)

Reactor-specific spent fuel discharge projections, 1987-2020

International Nuclear Information System (INIS)

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

1988-03-01

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

World nuclear fuel cycle requirements 1990

International Nuclear Information System (INIS)

1990-01-01

This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under three nuclear supply scenarios. Two of these scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries with free market economies (FME countries). A No New Orders scenario is presented only for the United States. These nuclear supply scenarios are described in Commercial Nuclear Power 1990: Prospects for the United States and the World (DOE/EIA-0438(90)). This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the FME projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2030 for the Lower and Upper Reference cases and through 2040, the last year in which spent fuel is discharged, for the No New Orders case. These disaggregated projections are provided at the request of the Department of Energy's Office of Civilian Radioactive Waste Management

Fuel performance annual report for 1983. Volume 1

Energy Technology Data Exchange (ETDEWEB)

Bailey, W.J.; Dunenfeld, M.S.

1985-03-01

This annual report, the sixth in a series, provides a brief description of fuel performance during 1983 in commercial nuclear power plants. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to additional, more detailed information and related NRC evaluations are included.

Burnable poison option for DUPIC fuel

International Nuclear Information System (INIS)

Choi, Hang Bok; Cupta, H. P.

1996-08-01

The mechanisms of positive coolant void reactivity of CANDU natural uranium and DUPIC fuel have been studied. The design study of DUPIC fuel was performed using the burnable poison material in the center pin to reduce the coolant void reactivity. The amount of burnable poison was determined such that the prompt inverse period of DUPIC fuel upon full coolant voiding is the same as that of natural uranium fuel at equilibrium burnup. A parametric study on various burnable poisons has shown that natural dysprosium has more merit over other materials because it uniformly controls the void reactivity throughout the burnup with reasonable amount of poison. Additional studies on the option of using scattering or absorber material in the center pin position and the option using variable fuel density were performed. In any case of option using variable fuel density were performed. In any case of options to reduce the void reactivity, it was found that either the discharge burnup and/or the relative linear pin power are sacrificed. A preliminary study was performed for the evaluation of reference DUPIC fuel performance especially represented by Stress Corrosion Cracking(SCC) parameters which is mainly influenced by the refueling operations. For the reference 2-bundle shift refueling scheme, the predicted ramped power and power increment of the reference DUPIC fuel are below the SCC thresholds of CANDU natural uranium fuel. For a 4-bundle shift refueling scheme, the envelopes of element ramped power and power increment upon refueling are 8% and 44% higher than those of a 2-bundle shift refueling scheme on the average, respectively, but still have margins to the failure thresholds of natural uranium fuel. 23 tabs., 25 figs., 20 refs. (Author)

Fuel options for the fuel cell vehicle: hydrogen, methanol or gasoline?

International Nuclear Information System (INIS)

Thomas, C.E.; James, B.D.; Lomax, F.D. Jr.; Kuhn, I.F. Jr.

2000-01-01

Fuel cell vehicles can be powered directly by hydrogen or, with an onboard chemical processor, other liquid fuels such as gasoline or methanol. Most analysts agree that hydrogen is the preferred fuel in terms of reducing vehicle complexity, but one common perception is that the cost of a hydrogen infrastructure would be excessive. According to this conventional wisdom, the automobile industry must therefore develop complex onboard fuel processors to convert methanol, ethanol or gasoline to hydrogen. We show here, however, that the total fuel infrastructure cost to society including onboard fuel processors may be less for hydrogen than for either gasoline or methanol, the primary initial candidates currently under consideration for fuel cell vehicles. We also present the local air pollution and greenhouse gas advantages of hydrogen fuel cell vehicles compared to those powered by gasoline or methanol. (Author)

Hot Experimental Facility reference flowsheet

International Nuclear Information System (INIS)

North, E.D.

1982-01-01

This paper is a useful set of background information of HEF flowsheets, although many changes have been made in the past three years. The HEF reference flowsheet is a modified high-acid PUREX flowsheet capable of operating in the coprocessing mode or with full partitioning of U and Pu. Adequate decontamination factors are provided to purify high-burnup, fast breeder-reactor fuels to levels required for recycle back to a fuel fabrication facility. Product streams are mixed U-Pu oxide and uranium oxide. No contaminated liquid wastes are intentionally discharged to the environment. All wastes are solidified and packaged for appropriate disposal. Acid and water are recovered for internal recycle. Excess water is treated and discharged from the plant stack. Several changes have been made in the reference flowsheet since that time, and these are noted briefly

Certification for copper concentration in reference material for fuel anhydro ethylic alcohol; Certificacao da concentracao de cobre em material de referencia para alcool etilico anidro combustivel (AEC)

Energy Technology Data Exchange (ETDEWEB)

Reis, Lindomar Augusto dos; Rocha, Marcia Silva da; Mesko, Marcia Foster; Silva, Fagner Francisco da; Quaresma, Maria Cristina Baptista; Araujo, Thiago Oliveira [Instituto Nacional de Metrologia, Normalizacao e Qualidade Industrial (DIMCI/INMETRO), Duque de Caxias, RJ (Brazil). Diretoria de Metrologia Cientifica e Industrial], E-mail: lareis@inmetro.gov.br

2009-07-01

This work aiming to obtain the first certified reference material for fuel anhydro ethylic alcohol relative to the copper concentration, which has his maximum limit determined by the in force legislation providing traceability and reliability for the measurement results.

Ignition delay measurements of light naphtha: A fully blended low octane fuel

KAUST Repository

Javed, Tamour

2016-06-15

Light naphtha is a fully blended, low-octane (RON. = 64.5, MON. = 63.5), highly paraffinic (>. 90% paraffinic content) fuel, and is one of the first distillates obtained during the crude oil refining process. Light naphtha is an attractive low-cost fuel candidate for advanced low-temperature compression ignition engines where autoignition is the primary control mechanism. We measured ignition delay times for light naphtha in a shock tube and a rapid compression machine (RCM) over a broad range of temperatures (640-1250. K), pressures (20 and 40. bar) and equivalence ratios (0.5, 1 and 2). Ignition delay times were modeled using a two-component primary reference fuel (PRF) surrogate and a multi-component surrogate. Both surrogates adequately captured the measured ignition delay times of light naphtha under shock tube conditions. However, for low-temperature RCM conditions, simulations with the multi-component surrogate showed better agreement with experimental data. These simulated surrogate trends were confirmed by measuring the ignition delay times of the PRF and multi-component surrogates in the RCM at . P = 20. bar, . ϕ = 2. Detailed kinetic analyses were undertaken to ascertain the dependence of the surrogates\\' reactivity on their chemical composition. To the best of our knowledge, this is the first fundamental autoignition study on the reactivity of a low-octane fully blended fuel and the use of a suitably formulated multi-component surrogate to model its behavior.

Study of fuel element characteristic of SM and SMP (SM-PRIMA) fuel assemblies

International Nuclear Information System (INIS)

Klinov, A.V.; Kuprienko, V.A.; Lebedev, V.A.; Makhin, V.M.; Tuchnin, L.M.; Tsykanov, V.A.

1999-01-01

The paper discusses the techniques and results of reactor tests and post-reactor investigations of the SM reactor fuel elements and fuel elements developed in the process of designing the specialized PRIMA test reactor with the SM reactor fuel elements used as a prototype and which are referred to as the SMP fuel elements. The behavior of fuel elements under normal operating conditions and under deviation from normal operating conditions was studied to verify the calculation techniques, to check the calculation results during preparation of the SM reactor safety substantiation report and to estimate the possibility of using such fuel elements in other projects. During tests of fuel rods under deviation from normal operating conditions their advantages were shown over fuel elements, the components of which were produced using the Al-based alloys. (author)

The disposal of Canada's nuclear fuel waste: engineered barriers alternatives

International Nuclear Information System (INIS)

Johnson, L.H.; Tait, J.C.; Shoesmith, D.W.; Crosthwaite, J.L.; Gray, M.N.

1994-01-01

The concept for disposal of Canada's nuclear fuel waste involves emplacing the waste in a vault excavated at a depth of 500 to 1000 m in plutonic rock of the Canadian Shield. The solid waste would be isolated from the biosphere by a multibarrier system consisting of engineered barriers, including long-lived containers and clay and cement-based sealing materials, and the natural barrier provided by the massive geological formation. The technical feasibility of this concept and its impact on the environment and human health are being documented in an Environmental Impact Statement (EIS), which will be submitted for review under the federal Environmental Assessment and Review Process. This report, one of nine EIS primary references, describes the various alternative designs and materials for engineered barriers that have been considered during the development of the Canadian disposal concept and summarizes engineered barrier concepts being evaluated in other countries. The basis for the selection of a reference engineered barrier system for the EIS is presented. This reference system involves placing used CANDU (Canada Deuterium Uranium) fuel bundles in titanium containers, which would then be emplaced in boreholes drilled in the floor of disposal rooms. Clay-based sealing materials would be used to fill both the space between the containers and the rock and the remaining excavations. In the section on waste forms, the properties of both used-fuel bundles and solidified high-level wastes, which would be produced by treating wastes resulting from the reprocessing of used fuel, are discussed. Methods of solidifying the wastes and the chemical durability of the solidified waste under disposal conditions are reviewed. Various alternative container designs are reviewed, ranging from preliminary conceptual designs to designs that have received extensive prototype testing. Results of structural performance, welding and inspection studies are also summarized. The corrosion of

Energy in France. Reference data

International Nuclear Information System (INIS)

2001-01-01

This small guidebook summarizes the main statistical data on energy for France. Data are presented in tables and graphs for the last decades up to the year 2000: energy in economy (contribution of energy industries to the national gross product, employment and investments in the energy sector), overall energies (average monthly price of crude oil, public opinion about nuclear energy, supplies and consumptions, primary energy production, consumption per energy and per sector, energy bill, import crude oil prices), petroleum (primary production, sectoral end-consumption of refined petroleum products, consumption per type of product, automotive fuels demand, imports and exports of petroleum products, petroleum bill), natural gas (total production, sectoral consumption, imports per country of production, imports bill), coal (production, sectoral consumption, imports per country of production), electricity (gross production, classical thermal production per type of fuel, simplified status, sectoral final consumption, nuclear sites, electricity trades with foreign countries), renewable energies (total production, district heating networks, wood consumption, installed wind power, solar thermal and photovoltaic production), rational use of energy (primary energy intensity, compared evolutions of the national gross product with the energy consumption, cumulated energy savings), prices (energies in the industry and in the domestic sectors, automotive fuels, energy consumptions in households), environment (sectoral and net CO 2 emissions). An energy units conversion table is added at the end of the booklet. (J.S.)

Reference value standards and primary standards for pH measurements in D2O and aqueous-organic solvent mixtures: new accessions and assessments

International Nuclear Information System (INIS)

Mussini, P.R.; Mussini, T.; Rondinini, S.

1997-01-01

Recommended Reference Value Standards based on the potassium hydro-genphthalate buffer at various temperatures are reported for pH measurements in various binary solvent mixtures of water with eight organic solvents: methanol, ethanol, 2-propanol, 1,2-ethanediol, 2-methoxyethanol (''methylcellosolve''), acetonitrile, 1,4-dioxane, and dimethyl sulfoxide, together with Reference Value Standard based on the potassium deuterium phthalate buffer for pD measurements in D 2 O. In addition are reported Primary Standards for pH based on numerous buffers in various binary solvent mixtures of water with methanol, ethanol, and dimethyl sulfoxide, together with Primary Standards for pD in D 2 O based on the citrate, phosphate and carbonate buffers. (author)

New technology and fuel cycles

International Nuclear Information System (INIS)

Mooradian, A.J.

1979-06-01

The means of improving uranium utilization in nuclear power reactors are reviewed with respect to economic considerations, assurance of adequate fuel supplies and risk of weapons proliferation. Reference is made to what can be done to improve fuel economy in existing reactor systems operating on a once-through fuel cycle and the potential for improvement offered by fuel recycle in those systems. The state of development of new reactor systems that offer significant savings in uranium utilization is also reviewed and conclusions are made respecting the policy implications of the search for fuel economy. (author)

Using Sankey diagrams to map energy flow from primary fuel to end use

International Nuclear Information System (INIS)

Subramanyam, Veena; Paramshivan, Deepak; Kumar, Amit; Mondal, Md. Alam Hossain

2015-01-01

Highlights: • Energy flows from both supply and demand sides shown through Sankey diagrams. • Energy flows from reserves to energy end uses for primary and secondary fuels shown. • Five main energy demand sectors in Alberta are analyzed. • In residential/commercial sectors, highest energy consumption is in space heating. • In the industrial sector, highest energy use is in the mining subsector. - Abstract: The energy sector is the largest contributor to gross domestic product (GDP), income, employment, and government revenue in both developing and developed nations. But the energy sector has a significant environmental footprint due to greenhouse gas (GHG) emissions. Efficient production, conversion, and use of energy resources are key factors for reducing the environmental footprint. Hence it is necessary to understand energy flows from both the supply and the demand sides. Most energy analyses focus on improving energy efficiency broadly without considering the aggregate energy flow. We developed Sankey diagrams that map energy flow for both the demand and supply sides for the province of Alberta, Canada. The diagrams will help policy/decision makers, researchers, and others to understand energy flow from reserves through to final energy end uses for primary and secondary fuels in the five main energy demand sectors in Alberta: residential, commercial, industrial, agricultural, and transportation. The Sankey diagrams created for this study show total energy consumption, useful energy, and energy intensities of various end-use devices. The Long-range Energy Alternatives Planning System (LEAP) model is used in this study. The model showed that Alberta’s total input energy in the five demand sectors was 189 PJ, 186 PJ, 828.5PJ, 398 PJ, and 50.83 PJ, respectively. On the supply side, the total energy input and output were found to be 644.84 PJ and 239 PJ, respectively. These results, along with the associated energy flows were depicted pictorially using

Nuclear fuel assemblies

International Nuclear Information System (INIS)

Butterfield, R.S.; Garner, D.L.M.

1977-01-01

Reference is made to nuclear fuel assemblies designed for cooling on the 'tube-in-shell' principle in which the fuel is contained by a shell and is cooled by coolant passed through tubes extending through the shell. It has been proposed to employ coated particle fuel as a porous bed on the tube side and the bleed coolant from the tubes into direct contact with the fuel particles. In this way heat is extracted both by direct contact with the fuel and by heat transfer through the coolant tube walls. The system described aims to provide an improved structure of tube and shell for a fuel assembly of this kind and is particularly suitable for use in a gas cooled fast reactor, being able to withstand the neutron flux and high temperature conditions in these reactors. Constructional details are given. (U.K.)

New Search Space Reduction Algorithm for Vertical Reference Trajectory Optimization

Directory of Open Access Journals (Sweden)

Alejandro MURRIETA-MENDOZA

2016-06-01

Full Text Available Burning the fuel required to sustain a given flight releases pollution such as carbon dioxide and nitrogen oxides, and the amount of fuel consumed is also a significant expense for airlines. It is desirable to reduce fuel consumption to reduce both pollution and flight costs. To increase fuel savings in a given flight, one option is to compute the most economical vertical reference trajectory (or flight plan. A deterministic algorithm was developed using a numerical aircraft performance model to determine the most economical vertical flight profile considering take-off weight, flight distance, step climb and weather conditions. This algorithm is based on linear interpolations of the performance model using the Lagrange interpolation method. The algorithm downloads the latest available forecast from Environment Canada according to the departure date and flight coordinates, and calculates the optimal trajectory taking into account the effects of wind and temperature. Techniques to avoid unnecessary calculations are implemented to reduce the computation time. The costs of the reference trajectories proposed by the algorithm are compared with the costs of the reference trajectories proposed by a commercial flight management system using the fuel consumption estimated by the FlightSim® simulator made by Presagis®.

Advanced fuel cycles and burnup increase of WWER-440 fuel

International Nuclear Information System (INIS)

Proselkov, V.; Saprykin, V.; Scheglov, A.

2003-01-01

Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge

Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant

International Nuclear Information System (INIS)

Monteiro, Iara Arraes

1999-02-01

The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

Global climate impacts of country-level primary carbonaceous aerosol from solid-fuel cookstove emissions

International Nuclear Information System (INIS)

Lacey, Forrest; Henze, Daven

2015-01-01

Cookstove use is globally one of the largest unregulated anthropogenic sources of primary carbonaceous aerosol. While reducing cookstove emissions through national-scale mitigation efforts has clear benefits for improving indoor and ambient air quality, and significant climate benefits from reduced green-house gas emissions, climate impacts associated with reductions to co-emitted black (BC) and organic carbonaceous aerosol are not well characterized. Here we attribute direct, indirect, semi-direct, and snow/ice albedo radiative forcing (RF) and associated global surface temperature changes to national-scale carbonaceous aerosol cookstove emissions. These results are made possible through the use of adjoint sensitivity modeling to relate direct RF and BC deposition to emissions. Semi- and indirect effects are included via global scaling factors, and bounds on these estimates are drawn from current literature ranges for aerosol RF along with a range of solid fuel emissions characterizations. Absolute regional temperature potentials are used to estimate global surface temperature changes. Bounds are placed on these estimates, drawing from current literature ranges for aerosol RF along with a range of solid fuel emissions characterizations. We estimate a range of 0.16 K warming to 0.28 K cooling with a central estimate of 0.06 K cooling from the removal of cookstove aerosol emissions. At the national emissions scale, countries’ impacts on global climate range from net warming (e.g., Mexico and Brazil) to net cooling, although the range of estimated impacts for all countries span zero given uncertainties in RF estimates and fuel characterization. We identify similarities and differences in the sets of countries with the highest emissions and largest cookstove temperature impacts (China, India, Nigeria, Pakistan, Bangladesh and Nepal), those with the largest temperature impact per carbon emitted (Kazakhstan, Estonia, and Mongolia), and those that would provide the

Global climate impacts of country-level primary carbonaceous aerosol from solid-fuel cookstove emissions

Science.gov (United States)

Lacey, Forrest; Henze, Daven

2015-11-01

Cookstove use is globally one of the largest unregulated anthropogenic sources of primary carbonaceous aerosol. While reducing cookstove emissions through national-scale mitigation efforts has clear benefits for improving indoor and ambient air quality, and significant climate benefits from reduced green-house gas emissions, climate impacts associated with reductions to co-emitted black (BC) and organic carbonaceous aerosol are not well characterized. Here we attribute direct, indirect, semi-direct, and snow/ice albedo radiative forcing (RF) and associated global surface temperature changes to national-scale carbonaceous aerosol cookstove emissions. These results are made possible through the use of adjoint sensitivity modeling to relate direct RF and BC deposition to emissions. Semi- and indirect effects are included via global scaling factors, and bounds on these estimates are drawn from current literature ranges for aerosol RF along with a range of solid fuel emissions characterizations. Absolute regional temperature potentials are used to estimate global surface temperature changes. Bounds are placed on these estimates, drawing from current literature ranges for aerosol RF along with a range of solid fuel emissions characterizations. We estimate a range of 0.16 K warming to 0.28 K cooling with a central estimate of 0.06 K cooling from the removal of cookstove aerosol emissions. At the national emissions scale, countries’ impacts on global climate range from net warming (e.g., Mexico and Brazil) to net cooling, although the range of estimated impacts for all countries span zero given uncertainties in RF estimates and fuel characterization. We identify similarities and differences in the sets of countries with the highest emissions and largest cookstove temperature impacts (China, India, Nigeria, Pakistan, Bangladesh and Nepal), those with the largest temperature impact per carbon emitted (Kazakhstan, Estonia, and Mongolia), and those that would provide the

Potentials and limitations of alternative fuels for diesel engine

Directory of Open Access Journals (Sweden)

Gligorijević Radinko

2009-01-01

Full Text Available The primary energy consumption in the world has increased continuously. The most important primary energy source is oil. The supply of automotive fuels today is based almost entirely on oil, and the demand for liquid transportation fuels worldwide will rise significantly in the next fifty years. Growing energy consumption and decreasing fossil resources are reasons for increasing prices of fossil fuel. Besides limited availability, contribution to greenhouse effect and pollutant emission represent another problem of fossil fuel. Both of these problems can be overcome by increased application of renewable biofuels. Therefore, great effort is made to supplement the primary energy sources by including renewable energies. There are alternative fuels 1st and 2nd generation. Some of them show high potential for reduction of engine out emission. But there are economical and technical barriers when such fuels are applied. This paper shows both advantage and disadvantage of alternative fuels, especially when used for diesel engines.

Development of challengeable reprocessing and fuel fabrication technologies for advanced fast reactor fuel cycle

International Nuclear Information System (INIS)

Nomura, S.; Aoshima, T.; Myochin, M.

2001-01-01

R and D in the next five years in Feasibility Study Phase-2 are focused on selected key technologies for the advanced fuel cycle. These are the reference technology of simplified aqueous extraction and fuel pellet short process based on the oxide fuel and the innovative technology of oxide-electrowinning and metal- electrorefining process and their direct particle/metal fuel fabrication methods in a hot cell. Automatic and remote handling system operation in both reprocessing and fuel manufacturing can handle MA and LLFP concurrently with Pu and U attaining the highest recovery and an accurate accountability of these materials. (author)

Assessment of the response of spent fuel transports to malevolent acts

International Nuclear Information System (INIS)

Sandoval, R.P.

1983-01-01

This paper describes the results of a program conducted at Sandia National Laboratories (SNL) for the US Department of Energy to provide an experimental data base for more accurately assessing the radiological consequences from a hypothetical sabotage attack on a spent fuel shipping cask. The primary objectives of the program were limited to: (1) evaluating the effectiveness of selected high explosive devices (HED) in breaching full-size spent fuel casks, (2) quantifying and characterizing relevant aerosol properties of the released fuel, and (3) using the resulting experimental data to evaluate the radiological health consequences resulting from a hypothetical sabotage attack on a spent fuel shipping cask in a densely populated area. Subscale and full-scale experiments in conjunction with an analytical modeling study were performed to meet the programmatic objectives. The data from this program indicate that the Urban Studies greatly overestimated the impact of malevolent acts directed at spent fuel casks in urban environs. From that standpoint this work could be the basis of additional regulatory revisions of the NRC physical protection requirements. In a larger sense this work can also be the basis of more credible worst case analyses since it defines the actual result of an event which is well beyond any expectations of cask failures in accident environments. 5 references

Fuel corrosion processes under waste disposal conditions

International Nuclear Information System (INIS)

Shoesmith, D.W.

2000-01-01

The release of the majority of radionuclides from spent nuclear fuel under permanent disposal conditions will be controlled by the rate of dissolution of the UO 2 fuel matrix. In this manuscript the mechanism of the coupled anodic (fuel dissolution) and cathodic (oxidant reduction) reactions which constitute the overall fuel corrosion process is reviewed, and the many published observations on fuel corrosion under disposal conditions discussed. The primary emphasis is on summarizing the overall mechanistic behaviour and establishing the primary factors likely to control fuel corrosion. Included are discussions on the influence of various oxidants including radiolytic ones, pH, temperature, groundwater composition, and the formation of corrosion product deposits. The relevance of the data recorded on unirradiated UO 2 to the interpretation of spent fuel behaviour is included. Based on the review, the data used to develop fuel corrosion models under the conditions anticipated in Yucca Mountain (NV, USA) are evaluated

Amenorrhea - primary

Science.gov (United States)

... of periods - primary Images Primary amenorrhea Normal uterine anatomy (cut section) Absence of menstruation (amenorrhea) References Bulun SE. The physiology and pathology of the female reproductive axis. In: ...

Fuel oil and kerosene sales 1994

International Nuclear Information System (INIS)

1995-01-01

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

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

International Nuclear Information System (INIS)

Sokolov, F.

2001-01-01

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

Development of nuclear fuel. Development of CANDU advanced fuel bundle

International Nuclear Information System (INIS)

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

1991-07-01

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

Genetics Home Reference: fragile X-associated primary ovarian insufficiency

Science.gov (United States)

... Share: Email Facebook Twitter Home Health Conditions FXPOI Fragile X-associated primary ovarian insufficiency Printable PDF Open All ... Javascript to view the expand/collapse boxes. Description Fragile X-associated primary ovarian insufficiency ( FXPOI ) is a condition ...

Reactor-specific spent fuel discharge projections: 1986 to 2020

International Nuclear Information System (INIS)

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

1987-03-01

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

DUPIC nuclear fuel manufacturing and process technology development

International Nuclear Information System (INIS)

Yang, Myung Seung; Park, J. J.; Lee, J. W.

2000-05-01

In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated

DUPIC nuclear fuel manufacturing and process technology development

Energy Technology Data Exchange (ETDEWEB)

Yang, Myung Seung; Park, J. J.; Lee, J. W. [and others

2000-05-01

In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated.

Effect of primary air content on formation of nitrogen oxides during combustion of Ehkibastuz coal

Energy Technology Data Exchange (ETDEWEB)

Kotler, V.R.; Imankulov, Eh.R.

1986-01-01

Investigations are discussed carried out in a pilot plant at the Kaz. Power Engineering Scientific Research Institute into the effect of the amount of primary air in coal-dust flame on the final concentration of nitrogen oxides in flue gases. The tests were carried out in a 7500 mm high, 1600 mm dia vertical cylindrical combustion chamber having type P-57 burner, and air dispersed fuel plus additional air supplies located at the top. Amounts of coal dust fed by a drum feeder along the air pipe varied from 100-600 kg/h. The required air was supplied by 5000 m/sup 3//h Type TK-700/5 blowers at 0.04 MPa. Ehkibastuz coal samples contained: 1.3% moisture; 48.1% ash; 38.02% carbon; 2.56% hydrogen; 0.73% sulfur; 0.60% nitrogen; heat of combustion was 14.3 MJ/kg. Results obtained indicate that variations in the amount of primary air in swirl flow burners affect formation of fuel nitrogen; there is an optimum volume at which minimum quantities of nitrogen oxides are formed. Either an increase or decrease in the primary air results in a rise in nitrogen oxide concentration. 3 references.

The energy future and the chemical fuels

International Nuclear Information System (INIS)

Bockris, J.O'M.

1976-01-01

An account is first given of the origin of present chemical fuels, with particular reference to the lastingness of coal. Methods of estimation of these fuels are discussed and the greenhouse effect arising from the burning of coal is described. Consideration is then given to methods available for extending the uses of chemical fuels, including interfacing them with new inexhaustible, clean energy sources. Finally, accounts are given of the Hydrogen Economy and of the production of chemical fuels from wind energy in massive wind belts. The paper includes references to the part that nuclear power was expected to play in future energy policy. Problems of breeder reactor development and the safety and management of plutonium and radioactive wastes are discussed. (author)

Improvements of MCOR: A Monte Carlo depletion code system for fuel assembly reference calculations

Energy Technology Data Exchange (ETDEWEB)

Tippayakul, C.; Ivanov, K. [Pennsylvania State Univ., Univ. Park (United States); Misu, S. [AREVA NP GmbH, An AREVA and SIEMENS Company, Erlangen (Germany)

2006-07-01

This paper presents the improvements of MCOR, a Monte Carlo depletion code system for fuel assembly reference calculations. The improvements of MCOR were initiated by the cooperation between the Penn State Univ. and AREVA NP to enhance the original Penn State Univ. MCOR version in order to be used as a new Monte Carlo depletion analysis tool. Essentially, a new depletion module using KORIGEN is utilized to replace the existing ORIGEN-S depletion module in MCOR. Furthermore, the online burnup cross section generation by the Monte Carlo calculation is implemented in the improved version instead of using the burnup cross section library pre-generated by a transport code. Other code features have also been added to make the new MCOR version easier to use. This paper, in addition, presents the result comparisons of the original and the improved MCOR versions against CASMO-4 and OCTOPUS. It was observed in the comparisons that there were quite significant improvements of the results in terms of k{sub inf}, fission rate distributions and isotopic contents. (authors)

Fuel management during Cirus refurbishing and re-commissioning

International Nuclear Information System (INIS)

Rai, K.K.; Srivastava, Alok; Ramesh, N.; Sharma, R.C.

2006-01-01

Cirus is a Heavy water moderated and Demineralised water cooled 40 MW(th) research reactor. Graphite is used as reflector. Natural uranium in metallic form and clad in aluminium is used as fuel. After over three decades of operation, signs of ageing started surfacing up. Refurbishment plan was drawn up based upon ageing studies and performance review. Core unloading was the foremost requirement for jobs like assessment of integrity of primary coolant underground pipelines. Refuelling is carried out during reactor shut down with primary coolant pumps in operation. During the fuel unloading with the gradual removal of assemblies, reduction in the gross flow of primary coolant was also envisaged. The scheme for core unloading was formulated with emphasis on optimization of fuel utilization, cooling to the fuel assemblies during transit and overall safety of equipment and personnel. After completion of refurbishing jobs, it was decided to install dummy assemblies in pile to facilitate commissioning of primary coolant system. A lot of difficulty was faced due to release of iron oxide flakes from the surface of primary coolant pipelines. The inlet feeders, valves and dummy assemblies had to be periodically flushed to get rid of iron oxide flakes deposited during the period. Various efforts made to get rid of iron oxide flakes included increasing the velocity by bypassing the core, installation of hollow dummy assemblies and installation of strainer at core inlet. The decision of fuel loading was made based upon the experience feedback with dummy assemblies and assessment of the pattern of release of iron oxide flakes. The dummy assemblies were replaced with uranium fuel assemblies. Fuelling work was carried out with Reactor hall crane with additional precautions. This paper describes the experience with handling of irradiated fuel assemblies to facilitate core unloading, experience with dummy assemblies and loading of fuel into the core and subsequent performance

A study of HANARO core conversion using high density U-Mo fuel

International Nuclear Information System (INIS)

Lee, K.H.; Lee, C.S.; Lee, B.C.; Park, S.J.; Kim, H.; Kim, C.K.

2002-01-01

Currently, HANARO is using 3.15gU/cc U3Si/Al as a driver fuel. HANARO has seven vertical irradiation holes in the core region. Three of them including a central trap are located in the inner region of the core and mainly being used for material irradiation tests. Four of them are located in the reflector tank but cooled by primary coolant. They are used for fuel irradiation tests or radioisotope development tests. For minimum core modification using high density U-Mo fuels, no dimension change is assumed in the current fuel rods and the cladding thickness remains the same in this study. The high density U-Mo fuel will have up to about twice the linear uranium loading of a current HANARO driver fuel. Using this high density fuel 8 fuel sites can be replaced with irradiation sites. Three kinds of conceptual cores are considered using 5 gU/cc U-7Mo/Al and 16 gU/cc U-7Mo. The increase of the linear heat generation rate due to the decrease of total fuel length can be overcome by more uniform radial and axial power distribution using different uranium densities and different fuel meat diameters are introduced into those cores. The new core has 4.54 times larger surface-to-volume ratio than the reference core. The core uranium loading, linear heat generation rate, excess reactivity, and control rod worth as well as the neutron spectra are analysed for each core. (author)

Study on HANARO core conversion using U-Mo fuel

International Nuclear Information System (INIS)

Lee, K.H.; Lee, C.S.; Seo, C.G.; Park, S.J.; Kim, H.; Kim, C.K.

2002-01-01

Two types of fuel rods with different fuel meat diameter and uranium density are considered for HANARO core conversion with high density U-Mo fuel. Arranging standard fuels of 5.0 g U/cc and 6.35 mm in diameter at the inner ring of an assembly and reduced fuels of 4.3 g U/cc and 5.49 mm in diameter at the outer ring of an assembly flattens the assembly power distribution and avoids the increase of linear heat generation rate due to using higher uranium density and less number of fuel rods. The maximum linear heat generation rate is similar with the current reference core and four fuel sites at the outer core in the reflector tank is converted to the irradiation sites to suit more demand on fuel tests and radioisotope production at outer core sites. This new core has 32% longer fuel cycle than the current reference core. (author)

The Nuclear Fuel Cycle Information System

International Nuclear Information System (INIS)

1987-02-01

The Nuclear Fuel Cycle Information System (NFCIS) is an international directory of civilian nuclear fuel cycle facilities. Its purpose is to identify existing and planned nuclear fuel cycle facilities throughout the world and to indicate their main parameters. It includes information on facilities for uranium ore processing, refining, conversion and enrichment, for fuel fabrication, away-from-reactor storage of spent fuel and reprocessing, and for the production of zirconium metal and Zircaloy tubing. NFCIS currently covers 271 facilities in 32 countries and includes 171 references

Fabrication of the instrumented fuel rods for the 3-Pin Fuel Test Loop at HANARO

International Nuclear Information System (INIS)

Sohn, Jae Min; Park, Sung Jae; Shin, Yoon Tag; Lee, Jong Min; Ahn, Sung Ho; Kim, Soo Sung; Kim, Bong Goo; Kim, Young Ki; Lee, Ki Hong; Kim, Kwan Hyun

2008-09-01

The 3-Pin Fuel Test Loop(hereinafter referred to as the '3-Pin FTL') facility has been installed at HANARO(High-flux Advanced Neutron Application Reactor) and the 3-Pin FTL is under a test operation. The purpose of this report is to fabricate the instrumented fuel rods for the 3-Pin FTL. The fabrication of these fuel rods was based on experiences and technologies of the instrumented fuel rods for an irradiation fuel capsule. The three instrumented fuel rods of the 3-Pin FTL have been designed. The one fuel rod(180 .deg. ) was designed to measure the centerline temperature of the nuclear fuels and the internal pressure of the fuel rod, and others(60 .deg. and 300 .deg. ) were designed to measure the centerline temperature of the fuel pellets. The claddings were made of the reference material 1 and 2 and new material 1 and 2. And nuclear fuel was used UO 2 (2.0w/o) pellet type with large grain and standard grain. The major procedures of fabrication are followings: (1) the assembling and weld of fuel rods with the pellet mockups and the sensor mockups for the qualification tests, (2) the qualification tests(dimension measurements, tensile tests, metallography examinations and helium leak tests) of weld, (3) the assembling and weld of instrumented fuel rods with the nuclear pellets and the sensors for the irradiation test, and (4) the qualification tests(the helium leak test, the dimensional measurement, electric resistance measurements of sensors) of test fuel rods. Satisfactory results were obtained for all the qualification tests of the instrumented fuel rods for the 3-Pin FTL. Therefore the three instrumented fuel rods for the 3-Pin FTL have been fabricated successfully. These will be installed in the In-Pile Section of 3-Pin FTL. And the irradiation test of these fuel rods is planned from the early next year for about 3 years at HANARO

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

International Nuclear Information System (INIS)

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

2005-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-10-15

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

Thermodynamic modeling of LPG combustion in dual-fuel engines; Modelisation thermodynamique de la combustion du GPL dans les moteurs dual-fuel

Energy Technology Data Exchange (ETDEWEB)

Bilcan, A.; Le Corre, O.; Tazerout, M. [Ecole des Mines de Nantes, 44 (France); Ramesh, A. [Indian Institute of Technology Madras (India)

2002-07-01

Dual-fuel engines are modified diesel engines burning simultaneously two fuels inside the cylinder: a gaseous one, called the primary fuel and a liquid one, called the pilot fuel. The thermal efficiency of the dual-fuel engine and of the diesel engine are comparable; the level of emissions is lower compared to the diesel one. This article presents a new procedure for the combustion modeling in a LPG-diesel dual-fuel engine. The procedures deals with the ignition delay period and with the rate of heat release inside the cylinder. This procedure is validated using experimental data issued front a collaboration with the Indian Institute of Technology from Madras, India. The used engine is a single-cylinder one, air-cooled. The pilot fuel is direct injected inside the cylinder The engine was run at constant load and with different diesel substitutions, i.e. for different air to fuel ratios of the primary fuel-air mixture. The general error of the procedure is below 10%. (authors)

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

International Nuclear Information System (INIS)

1983-03-01

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

Montana fuel tax refunds : draft final report.

Science.gov (United States)

2011-11-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Autoignition characterization of primary reference fuels and n-heptane/n-butanol mixtures in a constant volume combustion device and homogeneous charge compression ignition engine

KAUST Repository

Baumgardner, Marc E.; Sarathy, Mani; Má rchese, Anthony J.

2013-01-01

-octane, n-heptane, and n-butanol were tested in the FIT - 28 test runs with 25 ignition measurements for each test run, totaling 350 individual tests in all. These experimental results supported previous findings that fuel blends with high alcohol content

Fuel damage during off-normal transients in metal-fueled fast reactors

International Nuclear Information System (INIS)

Kramer, J.M.; Bauer, T.H.

1990-01-01

Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs

Checklist for transition to new highway fuel(s).

Energy Technology Data Exchange (ETDEWEB)

Risch, C.; Santini, D.J. (Energy Systems)

2011-12-15

Transportation is vital to the U.S. economy and society. As such, U.S. Presidents have repeatedly stated that the nation needs to reduce dependence on petroleum, especially for the highway transportation sector. Throughout history, highway transportation fuel transitions have been completed successfully both in United States and abroad. Other attempts have failed, as described in Appendix A: Historical Highway Fuel Transitions. Planning for a transition is critical because the changes can affect our nation's ability to compete in the world market. A transition will take many years to complete. While it is tempting to make quick decisions about the new fuel(s) of choice, it is preferable and necessary to analyze all the pertinent criteria to ensure that correct decisions are made. Doing so will reduce the number of changes in highway fuel(s). Obviously, changes may become necessary because of occurrences such as significant technology breakthroughs or major world events. With any and all of the possible transitions to new fuel(s), the total replacement of gasoline and diesel fuels is not expected. These conventional fuels are envisioned to coexist with the new fuel(s) for decades, while the revised fuel and vehicle infrastructures are implemented. The transition process must analyze the needs of the primary 'players,' which consist of the customers, the government, the fuel industry, and the automotive industry. To maximize the probability of future successes, the prime considerations of these groups must be addressed. Section 2 presents a succinct outline of the Checklist. Section 3 provides a brief discussion about the groupings on the Checklist.

Fuel shipment experience, fuel movements from the BMI-1 transport cask

International Nuclear Information System (INIS)

Bauer, Thomas L.; Krause, Michael G.

1986-01-01

The University of Texas at Austin received two shipments of irradiated fuel elements from Northrup Aircraft Corporation on April 11 and 16, 1985. A total of 59 elements consisting of standard and instrumented TRIGA fuel were unloaded from the BMI-1 shipping cask. At the time of shipment, the Northrup core burnup was approximately 50 megawatt days with fuel element radiation levels, after a cooling time of three months, of approximately 1.75 rem/hr at 3 feet. In order to facilitate future planning of fuel shipment at the UT facility and other facilities, a summary of the recent transfer process including several factors which contributed to its success are presented. Numerous color slides were made of the process for future reference by UT and others involved in fuel transfer and handling of the BMI-1 cask

The role of certified reference materials in material control and accounting

International Nuclear Information System (INIS)

Turel, S.P.

1979-01-01

One way of providing an adequate material control and accounting system for the nuclear fuel cycle is to calculate material unaccounted for (MUF) after a physical inventory and to compare the limit of error of the MUF value (LEMUF) against prescribed criteria. To achieve a meaningful LEMUF, a programme for the continuing determination of systematic and random errors is necessary. Within this programme it is necessary to achieve traceability of all Special Nuclear Material (SNM) control and accounting measurements to an International/National Measurement System by means of Certified Reference Materials. SNM measurements for control and accounting are made internationally on a great variety of materials using many diverse measurement procedures by a large number of facilities. To achieve valid overall accountability over this great variety of measurements there must be some means of relating all these measurements and their uncertainties to each other. This is best achieved by an International/National Measurement System (IMS/NMS). To this end, all individual measurement systems must be compatible to the IMS/NMS and all measurement results must be traceable to appropriate international/national Primary Certified Reference Materials. To obtain this necessary compatibility for any given SNM measurement system, secondary certified reference materials or working reference materials are needed for every class of SNM and each type of measurement system. Ways to achieve ''traceability'' and the various types of certified reference material are defined and discussed in this paper. (author)

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

Energy Technology Data Exchange (ETDEWEB)

Heidet, F.; Kim, T.; Grandy, C. (Nuclear Engineering Division)

2012-07-30

best core performance characteristics for each of them. With the exception of the fuel type and enrichment, the reference AFR-100 core design characteristics were kept unchanged, including the general core layout and dimensions, assembly dimensions, materials and power rating. In addition, the mass of {sup 235}U required was kept within a reasonable range from that of the reference AFR-100 design. The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code. The thermal margins were evaluated by comparing the peak temperatures to the design limits for parameters such as the fuel melting temperature and the fuel-cladding eutectic temperature. The inherent safety features of AFR-100 cores proposed were assessed using the integral reactivity parameters of the quasi-static reactivity balance analysis. The design objectives and requirements, the computation methods used as well as a description of the core concept are provided in Section 2. The three major approaches considered are introduced in Section 3 and the neutronics performances of those approaches are discussed in the same section. The orifice zoning strategies used and the steady-state thermal-hydraulic performance are provided in Section 4. The kinetics and reactivity coefficients, including the inherent safety characteristics, are provided in Section 5, and the Conclusions in Section 6. Other scenarios studied and sensitivity studies are provided in the Appendix section.

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Compatibility analysis of DUPIC fuel (part5) - DUPIC fuel cycle economics analysis

Energy Technology Data Exchange (ETDEWEB)

Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung

2000-08-01

This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the one-batch equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.55-6.72 mills/kWh for proposed DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.04-0.28 mills/kWh. Considering the uncertainty (0.45-0.51 mills/kWh) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by -20% and reduce the spent fuel arising by -65%, compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle possesses a strong advantage over the once-through fuel cycle from the viewpoint of the environmental effect.

Compatibility analysis of DUPIC fuel (part5) - DUPIC fuel cycle economics analysis

International Nuclear Information System (INIS)

Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung

2000-08-01

This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the one-batch equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.55-6.72 mills/kWh for proposed DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.04-0.28 mills/kWh. Considering the uncertainty (0.45-0.51 mills/kWh) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by -20% and reduce the spent fuel arising by -65%, compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle possesses a strong advantage over the once-through fuel cycle from the viewpoint of the environmental effect

Emissions from Diesel and Gasoline Vehicles Fuelled by Fischer-Tropsch Fuels and Similar Fuels

DEFF Research Database (Denmark)

Larsen, Ulrik; Lundorff, Peter; Ivarsson, Anders

2007-01-01

and an alkylate fuel (Aspen), which was taken to be the ultimate formula of FT gasoline. FT based diesel generally showed good emission performance, whereas the FT based gasoline not necessary lead to lower emissions. On the other hand, the Aspen fuel did show many advantages for the emissions from the gasoline...... vehicles fuelled by Fischer Tropsch (FT) based diesel and gasoline fuel, compared to the emissions from ordinary diesel and gasoline. The comparison for diesel fuels was based on a literature review, whereas the gasoline comparison had to be based on our own experiments, since almost no references were...

Aids to determining fuel models for estimating fire behavior

Science.gov (United States)

Hal E. Anderson

1982-01-01

Presents photographs of wildland vegetation appropriate for the 13 fuel models used in mathematical models of fire behavior. Fuel model descriptions include fire behavior associated with each fuel and its physical characteristics. A similarity chart cross-references the 13 fire behavior fuel models to the 20 fuel models used in the National Fire Danger Rating System....

Fast breeder fuel cycle

International Nuclear Information System (INIS)

1978-07-01

This contribution is prepared for the answer to the questionnaire of working group 5, subgroup B. B.1. is the short review of the fast breeder fuel cycles based on the reference large commercial Japanese LMFBR. The LMFBRs are devided into two types. FBR-A is the reactor to be used before 2000, and its burnup and breeding ratio are relatively low. The reference fuel cycle requirement is calculated based on the FBR-A. FBR-B is the one to be used after 2000, and its burnup and breeding ratio are relatively high. B.2. is basic FBR fuel reprocessing scheme emphasizing the differences with LWR reprocessing. This scheme is based on the conceptual design and research and development work on the small scale LMFBR reprocessing facility of Japan. The facility adopts a conventional PUREX process except head end portions. The report also describes the effects of technical modifications of conventional reprocessing flow sheets, and the problems to be solved before the adoption of these alternatives

Control of degradation of spent LWR [light-water reactor] fuel during dry storage in an inert atmosphere

International Nuclear Information System (INIS)

Cunningham, M.E.; Simonen, E.P.; Allemann, R.T.; Levy, I.S.; Hazelton, R.F.

1987-10-01

Dry storage of Zircaloy-clad spent fuel in inert gas (referred to as inerted dry storage or IDS) is being developed as an alternative to water pool storage of spent fuel. The objectives of the activities described in this report are to identify potential Zircaloy degradation mechanisms and evaluate their applicability to cladding breach during IDS, develop models of the dominant Zircaloy degradation mechanisms, and recommend cladding temperature limits during IDS to control Zircaloy degradation. The principal potential Zircaloy cladding breach mechanisms during IDS have been identified as creep rupture, stress corrosion cracking (SCC), and delayed hydride cracking (DHC). Creep rupture is concluded to be the primary cladding breach mechanism during IDS. Deformation and fracture maps based on creep rupture were developed for Zircaloy. These maps were then used as the basis for developing spent fuel cladding temperature limits that would prevent cladding breach during a 40-year IDS period. The probability of cladding breach for spent fuel stored at the temperature limit is less than 0.5% per spent fuel rod. 52 refs., 7 figs., 1 tab

Factors controlling metal fuel lifetime

International Nuclear Information System (INIS)

Porter, D.L.; Hofman, G.L.; Seidel, B.R.; Walters, L.C.

1986-01-01

The reliability of metal fuel elements is determined by a fuel burnup at which a statistically predicted number of fuel breaches would occur, the number of breaches determined by the amount of free fission gas which a particular reactor design can tolerate. The reliability is therefore measured using experimentally determined breach statistics, or by modelling fuel element behavior and those factors which contribute to cladding breach. The factors are fuel/cladding mechanical and chemical interactions, fission gas pressure, fuel phase transformations involving volume changes, and fission product effects on cladding integrity. Experimental data for EBR-II fuel elements has shown that the primary, and perhaps the only significant factor affecting metal fuel reliability, is the pressure-induced stresses caused by fission gas release. Other metal fuel/cladding systems may perform similarly

FUEL FORMULATION EFFECTS ON DIESEL FUEL INJECTION, COMBUSTION, EMISSIONS AND EMISSION CONTROL

Energy Technology Data Exchange (ETDEWEB)

Boehman, A; Alam, M; Song, J; Acharya, R; Szybist, J; Zello, V; Miller, K

2003-08-24

This paper describes work under a U.S. DOE sponsored Ultra Clean Fuels project entitled ''Ultra Clean Fuels from Natural Gas,'' Cooperative Agreement No. DE-FC26-01NT41098. In this study we have examined the incremental benefits of moving from low sulfur diesel fuel and ultra low sulfur diesel fuel to an ultra clean fuel, Fischer-Tropsch diesel fuel produced from natural gas. Blending with biodiesel, B100, was also considered. The impact of fuel formulation on fuel injection timing, bulk modulus of compressibility, in-cylinder combustion processes, gaseous and particulate emissions, DPF regeneration temperature and urea-SCR NOx control has been examined. The primary test engine is a 5.9L Cummins ISB, which has been instrumented for in-cylinder combustion analysis and in-cylinder visualization with an engine videoscope. A single-cylinder engine has also been used to examine in detail the impacts of fuel formulation on injection timing in a pump-line-nozzle fueling system, to assist in the interpretation of results from the ISB engine.

Small global effect on terrestrial net primary production due to increased fossil fuel aerosol emissions from East Asia during the last decade.

Science.gov (United States)

O'Sullivan, Michael; Rap, Alex; Reddington, Carly; Spracklen, Dominick; Buermann, Wolfgang

2016-04-01

The global terrestrial carbon sink has increased since the start of this century at a time of rapidly growing carbon dioxide emissions from fossil fuel burning. Here we test the hypothesis that increases in atmospheric aerosols from fossil fuel burning have increased the diffuse fraction of incoming solar radiation and the efficiency of photosynthesis leading to increased plant carbon uptake. Using a combination of atmospheric and biospheric models, we find that changes in diffuse light associated with fossil fuel aerosol emission accounts for only 2.8% of the increase in global net primary production (1.221 PgC/yr) over the study period 1998 to 2007. This relatively small global signal is however a result of large regional compensations. Over East Asia, the strong increase in fossil fuel emissions contributed nearly 70% of the increased plant carbon uptake (21 TgC/yr), whereas the declining fossil fuel aerosol emissions in Europe and North America contributed negatively (-16% and -54%, respectively) to increased plant carbon uptake. At global scale, we also find the CO2 fertilization effect on photosynthesis to be the dominant driver of increased plant carbon uptake, in line with previous studies. These results suggest that further research into alternative mechanisms by which fossil fuel emissions could increase carbon uptake, such as nitrogen deposition and carbon-nitrogen interactions, is required to better understand a potential link between the recent changes in fossil fuel emissions and terrestrial carbon uptake.

Small Volume Fuel Testers Report

Energy Technology Data Exchange (ETDEWEB)

Schoegl, I. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); McNenly, M. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Killingsworth, N. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2016-10-31

Micro-liter fuel ignition testing (μ-FIT) is based on the premise that characteristics FREI (Flames with Repetitive Extinction and Ignition, i.e. cyclically occurring combustion events within heated capillaries), are linked to fuel properties. In early FY16, proof-of-concept measurements with primary reference fuel (PRF) mixtures, i.e. blends of n-heptane and iso-octane, yielded clear evidence for the feasibility of the approach. Our experiments showed that it is critical to accurately link observed flame positions to local temperatures, which provides information on ignition, extinction and flame propagation, all of which are known to be impacted by fuel properties. In FY16, one major hurdle was uncertainty of temperature calibration, which required significant efforts for corrective action that were not included in the original scope of work. Temperature calibrations are obtained by translating a thermocouple within the capillary in absence of a flame. While measurements have good repeatability when accounting for transient and insertion effects, results from nominally identical thermocouples reveal unacceptable uncertainty (up to ±50K), which is attributed to variations in thermocouple placement and manufacturing tolerances. This issue is currently being resolved by switching to non-intrusive optical temperature measurements. Updates are expected to yield uncertainties of less than ±10K, while also eliminating transient and insertion effects. The experimental work was complemented by computational efforts where it was shown that a simplified Lagrangian zero-D model with detailed kinetics yields fuelspecific differentiation of ignition temperatures for simple fuels that are consistent with experiments. Further, a 2D transient model was implemented in OpenFOAM to investigate combustion behavior of simple fuels at elevated pressure. In an upcoming visit to LLNL, more advanced simulations using LLNL’s computational tools (e.g. zero-RK) are planned, which will

Airline view of LH2 as a fuel for commercial aircraft

Energy Technology Data Exchange (ETDEWEB)

Wilkinson, K.G.

1983-01-01

Fuel for air transport is a particularly important matter, because it is an energy-intensive form of transport. Price and energy density are therefore correspondingly critical. Airlines views tend to be short term because there is no future if they do not survive tomorrow. Thus, although petroleum supply and price has been unstable since 1973, airlines' primary concern has been with economizing the use of fuel and specifying more fuel economic equipment for the future, rather than serious pursuit of more radical solutions. LH2 gives promise of being a serious candidate for the longer term, but there is not yet convincing proof that cheaper seat mile costs could be produced in today's circumstances. The situation could be changed by the realization of laminar flow by surface cooling at subsonic speeds. There is a strong case for a modest programme of carefully aimed research on an international basis to fill the most important gaps in our knowledge on the design of LH2 fueled aircraft. Future crises in energy supplies may well raise questions of transport fuel strategy which will heed this background for wise decision-making. Particular significance attaches to the place of nuclear power in world energy policy since this dominates the question of LH2 availability and price. 2 references, 2 figures.

Thermodynamic analysis of biofuels as fuels for high temperature fuel cells

Science.gov (United States)

Milewski, Jarosław; Bujalski, Wojciech; Lewandowski, Janusz

2013-02-01

Based on mathematical modeling and numerical simulations, applicativity of various biofuels on high temperature fuel cell performance are presented. Governing equations of high temperature fuel cell modeling are given. Adequate simulators of both solid oxide fuel cell (SOFC) and molten carbonate fuel cell (MCFC) have been done and described. Performance of these fuel cells with different biofuels is shown. Some characteristics are given and described. Advantages and disadvantages of various biofuels from the system performance point of view are pointed out. An analysis of various biofuels as potential fuels for SOFC and MCFC is presented. The results are compared with both methane and hydrogen as the reference fuels. The biofuels are characterized by both lower efficiency and lower fuel utilization factors compared with methane. The presented results are based on a 0D mathematical model in the design point calculation. The governing equations of the model are also presented. Technical and financial analysis of high temperature fuel cells (SOFC and MCFC) are shown. High temperature fuel cells can be fed by biofuels like: biogas, bioethanol, and biomethanol. Operational costs and possible incomes of those installation types were estimated and analyzed. A comparison against classic power generation units is shown. A basic indicator net present value (NPV) for projects was estimated and commented.

Thermodynamic analysis of biofuels as fuels for high temperature fuel cells

Directory of Open Access Journals (Sweden)

Milewski Jarosław

2013-02-01

Full Text Available Based on mathematical modeling and numerical simulations, applicativity of various biofuels on high temperature fuel cell performance are presented. Governing equations of high temperature fuel cell modeling are given. Adequate simulators of both solid oxide fuel cell (SOFC and molten carbonate fuel cell (MCFC have been done and described. Performance of these fuel cells with different biofuels is shown. Some characteristics are given and described. Advantages and disadvantages of various biofuels from the system performance point of view are pointed out. An analysis of various biofuels as potential fuels for SOFC and MCFC is presented. The results are compared with both methane and hydrogen as the reference fuels. The biofuels are characterized by both lower efficiency and lower fuel utilization factors compared with methane. The presented results are based on a 0D mathematical model in the design point calculation. The governing equations of the model are also presented. Technical and financial analysis of high temperature fuel cells (SOFC and MCFC are shown. High temperature fuel cells can be fed by biofuels like: biogas, bioethanol, and biomethanol. Operational costs and possible incomes of those installation types were estimated and analyzed. A comparison against classic power generation units is shown. A basic indicator net present value (NPV for projects was estimated and commented.

Alternative fuel cycles and non-proliferation aspects

International Nuclear Information System (INIS)

Kessler, G.

1980-10-01

The most important physical characteristics of the U/Pu and the Th/U fuel cycles and the technical data of the most significant converter reactors operating with Th/U fuel are outlined in the report. Near breeders as well as breeders with a thermal neutron spectrum are briefly discussed, and the potential of breeders with fast neutron spectra in the Th/U fuel is outlined. The essential criteria for the comparison of the alternative fuel cycles with the reference Pu/U cycle are the consumption of natural uranium, the numbers of U-233 producing and U-233 consuming converter reactors and the amounts of fission material transported and handled within the fuel cycle (reprocessing, refabrication). Although the alternative U/Th fuel cycles are feasible with some advantages and some disadvantages as compared to the reference U/Pu cycle, not much experience has so far been gathered with pilot plants of the fuel cycle. The respective status in reprocessing, refabrication and waste disposal is briefly discussed. Finally, a comparison of the risk potential inherent in secular storage is presented and questions of resistance to proliferation and of safeguards of the U/Th cycle are discussed

An integrated methodology to evaluate a spent nuclear fuel storage system

International Nuclear Information System (INIS)

Yoon, Jeong Hyoun

2008-02-01

This study introduced a methodology that can be applied for development of a dry storage system for spent nuclear fuels. It consisted of several design activities that includes development of a simplified program to analyze the amount of spent nuclear fuels from reflecting the practical situation in spent nuclear fuel management and a simplified program to evaluate the cost of 4 types of representing storage system to choose the most competitive option considering economic factor. As verification of the implementation of the reference module to practical purpose, a simplified thermal analysis code was suggested that can see fulfillment of limitation of temperature in long term storage and oxidation analysis. From the thermal related results, the reference module can accommodate full range of PHWR spent nuclear fuels and significant portion of PWR ones too. From the results, the reference storage system can be concluded that has fulfilled the important requirements in terms of long term integrity and radiological safety. Also for the purpose of solving scattered radiation along with deep penetration problems in cooling storage system, small but efficient design alternation was suggested together with its efficiency that can reduce scattered radiation by 1/3 from the original design. Along with the countermeasure for the shielding problem, in consideration of PWR spent nuclear fuels, simplified criticality analysis methodology retaining conservativeness was proposed. The results show the reference module is efficient low enrichment PWR spent nuclear fuel and even relatively high enrichment fuels too if burnup credit is taken. As conclusive remark, the methodology is simple but efficient to plan a concept design of convective cooling type of spent nuclear fuels storage. It can be also concluded that the methodology derived in this study and the reference module has feasibility in practical implementation to mitigate the current complex situation in spent fuel

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

International Nuclear Information System (INIS)

Greenspan, E

2006-01-01

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

Proceedings of the Topical meeting on the reactor fuel performance - TopFuel 2012 Transactions

International Nuclear Information System (INIS)

2012-01-01

TopFuel is an annual topical meeting organised by ENS, the American Nuclear Society and the Atomic Energy Society of Japan. TopFuel's primary objective is to bring together leading specialists in the field from around the world to analyse advances in nuclear fuel management technology and to use the findings of the latest cutting-edge research to help manufacture the high performance nuclear fuels of today and tomorrow. Aim is to discuss the challenges facing the developers and manufacturers of new high-performance nuclear fuels - fuels that will help meet current and future energy demand and reduce man's over dependence upon CO 2 -emitting fossil fuels. The technical scope of Top Fuel 2012 includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. (authors)

Krakow clean fossil fuels and energy efficiency project

Energy Technology Data Exchange (ETDEWEB)

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

1995-11-01

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

LEU fuel cycle analyses for the Belgian BR2 Research Reactor

International Nuclear Information System (INIS)

Deen, J.R.; Snelgrove, J.L.

1988-01-01

Equilibrium fuel cycle characteristics were calculated for reference HEU and two proposed LEU fuel cycles using an 11-group diffusion-theory neutron flux solution in hexagonal-Z geometry. The diffusion theory model was benchmarked with a detailed Monte Carlo core model. The two proposed LEU fuel designs increased the 235 U loading 20% and the fuel meat volume 51%. The first LEU design used 10 B as a burnable absorber. Either proposed LEU fuel element would provide equilibrium fuel cycle characteristics similar to those of the HEU fuel cycle. Irradiation rates of Co control followers and Ir disks in the center of the core were reduced 6 ± 1% in the LEU equilibrium core compared to reference HEU core. 11 refs., 4 figs., 5 tabs

Shock wave calibration of under-expanded natural gas fuel jets

Science.gov (United States)

White, T. R.; Milton, B. E.

2008-10-01

Natural gas, a fuel abundant in nature, cannot be used by itself in conventional diesel engines because of its low cetane number. However, it can be used as the primary fuel with ignition by a pilot diesel spray. This is called dual-fuelling. The gas may be introduced either into the inlet manifold or, preferably, directly into the cylinder where it is injected as a short duration, intermittent, sonic jet. For accurate delivery in the latter case, a constant flow-rate from the injector is required into the constantly varying pressure in the cylinder. Thus, a sonic (choked) jet is required which is generally highly under-expanded. Immediately at the nozzle exit, a shock structure develops which can provide essential information about the downstream flow. This shock structure, generally referred to as a “barrel” shock, provides a key to understanding the full injection process. It is examined both experimentally and numerically in this paper.

Fuel assembly leak tightness control on WWER-1000 reactor

International Nuclear Information System (INIS)

Ivanova, R.; Gerchev, N.; Mateev, A.

2001-01-01

The main index for integrity of the fuel rods cladding is the specific activity value of the primary coolant. This value determines the safe operation of the reactor. The limit for safe operation of WWER-1000 reactor is the value of the total activity of Iodine isotopes in the primary coolant 5.0x10 -3 Ci/l. The paper briefly describes the methodology for performing a fuel tightness test (sipping test) and shows the results from these tests performed during the period 1987 -1999 in units 5 and 6 at the Kozloduy NPP. An additional index related to the safe operation is defined to characterize the fuel cladding integrity Fuel Reliability Index (FRI). The FRI is defined as value of the average activity of 131 I in the primary coolant, corrected with a part of precipitated 235 U migration and fixed to the general permanent purification frequency. Two criteria (quantitative and statistic) are determined to qualify the fuel cladding integrity. The results from sipping tests show good reliability of the fuel irradiated in unit 5 and 6 at the Kozloduy NPP

Updated reference design of a liquid metal cooled tandem mirror fusion breeder

International Nuclear Information System (INIS)

Berwald, D.H.; Whitley, R.H.; Garner, J.K.

1985-09-01

Detailed studies of key techinical issues for liquid metal cooled fusion breeder (fusion-fission hybrid blankets) have been performed during the period 1983-4. Based upon the results of these studies, the 1982 reference liquid metal cooled tandem mirror fusion breeder blanket design was updated and is described. The updated reference blankets provides increased breeding and lower technological risk in comparison with the original reference blanket. In addition to the blanket design revisions, a plant concept, cost, and fuel cycle economics assessment is provided. The fusion breeder continues to promise an economical source of fissile fuel for the indefinite future

Updated reference design of a liquid metal cooled tandem mirror fusion breeder

Energy Technology Data Exchange (ETDEWEB)

Berwald, D.H.; Whitley, R.H.; Garner, J.K.; Gromada, R.J.; McCarville, T.J.; Moir, R.W.; Lee, J.D.; Bandini, B.R.; Fulton, F.J.; Wong, C.P.C.; Maya, I.; Hoot, C.G.; Schultz, K.R.; Miller, L.G.; Beeston, J.M.; Harris, B.L.; Westman, R.A.; Ghoniem, N.M.; Orient, G.; Wolfer, M.; DeVan, J.H.; Torterelli, P.

1985-09-01

Detailed studies of key techinical issues for liquid metal cooled fusion breeder (fusion-fission hybrid blankets) have been performed during the period 1983-4. Based upon the results of these studies, the 1982 reference liquid metal cooled tandem mirror fusion breeder blanket design was updated and is described. The updated reference blankets provides increased breeding and lower technological risk in comparison with the original reference blanket. In addition to the blanket design revisions, a plant concept, cost, and fuel cycle economics assessment is provided. The fusion breeder continues to promise an economical source of fissile fuel for the indefinite future.

Carbon Tolerant Fuel Electrodes for Reversible Sofc Operating on Carbon Dioxide

Directory of Open Access Journals (Sweden)

Papazisi Kalliopi Maria

2017-01-01

Full Text Available A challenging barrier for the broad, successful implementation of Reversible Solid Oxide Fuel Cell (RSOFC technology for Mars application utilizing CO2 from the Martian atmosphere as primary reactant, remains the long term stability by the effective control and minimization of degradation resulting from carbon built up. The perovskitic type oxide material La0.75Sr0.25Cr0.9Fe0.1O3-δ (LSCF has been developed and studied for its performance and tolerance to carbon deposition, employed as bi-functional fuel electrode in a Reversible SOFC operating on the CO2 cycle (Solid Oxide Electrolysis Cell/SOEC: CO2 electrolysis, Solid Oxide Fuel Cell/SOFC: power generation through the electrochemical reaction of CO and oxygen. A commercial state-of-the-art NiO-YSZ (8% mol Y2O3 stabilized ZrO2 cermet was used as reference material. CO2 electrolysis and fuel cell operation in 70% CO/CO2 were studied in the temperature range of 900-1000°C. YSZ was used as electrolyte while LSM-YSZ/LSM (La0.2Sr0.8MnO3 as oxygen electrode. Results showed that LSCF had high and stable performance under RSOFC operation.

Performance limits of coated particle fuel. Part I. The significance of empirical performance diagrams and mathematical models in fuel development and power reactor studies

Energy Technology Data Exchange (ETDEWEB)

Graham, L. W.; Hick, H.

1973-06-15

This report introduces a general survey of our present knowledge and understanding of coated particle fuel performance. It defines first the reference power reactor conditions and the reference coated particle design on which the survey is centred. It describes then the typical strategy which has been followed in coated particle fuel development by the Dragon Project R & D Branch. Finally it shows the priorities which have governed the time scale and scope of fuel development and of the present review.

Part 5. Fuel cycle options

International Nuclear Information System (INIS)

Lineberry, M.J.; McFarlane, H.F.; Amundson, P.I.; Goin, R.W.; Webster, D.S.

1980-01-01

The results of the FBR fuel cycle study that supported US contributions to the INFCE are presented. Fuel cycle technology is reviewed from both generic and historical standpoints. Technology requirements are developed within the framework of three deployment scenarios: the reference international, the secured area, and the integral cycle. Reprocessing, fabrication, waste handling, transportation, and safeguards are discussed for each deployment scenario. Fuel cycle modifications designed to increase proliferation defenses are described and assessed for effectiveness and technology feasibility. The present status of fuel cycle technology is reviewed and key issues that require resolution are identified

Fuel oil and kerosene sales 1992

International Nuclear Information System (INIS)

1993-01-01

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

Innovative inert matrix-thoria fuels for in-reactor plutonium disposition

International Nuclear Information System (INIS)

Vettraino, F.; Padovani, E.; Luzzi, L.; Lombardi, C.; Thoresen, H.; Oberlander, B.; Iversen, G.; Espeland, M.

1999-01-01

The present leading option for plutonium disposition, either civilian or weapons Pu, is to burn it in LWRs after having converted it to MOX fuel. However, among the possible types of fuel which can be envisaged to burn plutonium in LWRs, innovative U-free fuels such as inert matrix and thoria fuel are novel concept in view of a more effective and ultimate solution from both security and safety standpoint. Inert matrix fuel is an non-fertile oxide fuel consisting of PuO 2 , either weapon-grade or reactor-grade, diluted in inert oxides such as for ex. stabilized ZrO 2 or MgAl 2 O 4 , its primary advantage consisting in no-production of new plutonium during irradiation, because it does not contain uranium (U-free fuel) whose U-238 isotope is the departure nuclide for breeding Pu-239. Some thoria addition in the matrix (thoria-doped fuel) may be required for coping with reactivity feedback needs. The full thoria-plutonia fuel though still a U-free variant cannot be defined non-fertile any more because the U-233 generation. The advantage of such a fuel option consisting basically on a remarkable already existing technological background and a potential acceleration in getting rid of the Pu stocks. All U-free fuels are envisaged to be operated under a once-through cycle scheme being the spent fuel outlooked to be sent directly to the final disposal in deep geological formations without requiring any further reprocessing treatment, thanks to the quality-poor residual Pu and a very high chemical stability under the current fuel reprocessing techniques. Besides, inert matrix-thoria fuel technology is suitable for in-reactor MAs transmutation. An additional interest in Th containing fuel refers to applicability in ADS, the innovative accelerated driven subcritical systems, specifically aimed at plutonium, minor actnides and long lived fission products transmutation in a Th-fuel cycle scheme which enables to avoid generations of new TRUs. A first common irradiation experiment

Fast breeder fuel cycle

International Nuclear Information System (INIS)

1978-09-01

Basic elements of the ex-reactor part of the fuel cycle (reprocessing, fabrication, waste handling and transportation) are described. Possible technical and proliferation measures are evaluated, including current methods of accountability, surveillance and protection. The reference oxide based cycle and advanced cycles based on carbide and metallic fuels are considered utilizing conventional processes; advanced nonaqueous reprocessing is also considered. This contribution provides a comprehensive data base for evaluation of proliferation risks

Thirteenth annual report of the Technical Advisory Committee on the Nuclear Fuel Waste Management Program

Energy Technology Data Exchange (ETDEWEB)

1993-05-15

This report details activities since the last reporting period by the Technical Advisory Committee (TAC). The emphasis of the work in the Canadian Nuclear Fuel Waste Management Program (CNFWMP) has been on the writing of the Environmental Impact Statement (EIS) and the associated set of the primary reference document as well as supporting documents. These are in preparation for submission to the Environmental Assessment Review Panel who will lead the national evaluation of the disposal concept under the auspices of the Federal Environmental Assessment Review Office (FEARO).

Aims of failed fuel detection and substantiation of radiation safety at implementation of new kinds of nuclear fuel and fuel cycles on NPP with WWER

International Nuclear Information System (INIS)

Miglo, V.; Luzanova, L.

2011-01-01

Limiting of number of leaking fuel rods in a core during reactor operation in the analyses which are carried out for a substantiation of radiating safety for NPP with WWER as well as problems and possibilities of FFD at implementation of new kinds of fuel and fuel cycles are the main topics discussed in this paper. Available experience of designing of the NPP with WWER shows, that for ensuring of implementation of the RS criteria regarding limiting radioactive emissions from the NPP and doses of an irradiation of the population living near to NPP, it is required to regulate more rigidly number of failed fuel rods in comparison with requirements of Rules of nuclear safety NP-082-07. The reason of it is necessity to consider a technical condition of all safety barriers on a path of radioactive FP extension in a complex, first and foremost of uncontrolled leakage of the primary coolant to the NPP premises and efficiency of filters of ventilating systems, and also spike-effect on activity of isotopes of iodine after a power unit shutdown for fuel reloading and openings of a cover of a reactor. Depending on the project of NPP, parameters of fuel loading, a place of placing of the NPP and other factors the limit level of activity of isotopes of iodine in the primary coolant will be reached at various number of leaking fuel rods which can be unequal for various power units and the NPP with WWER, constructed on one design. The quantity of leaking fuel rods at which the design limit on FP-activity in the primary coolant of operating reactor is reached, can be essential below an operational limit on number of failed fuel rods established by Rules of nuclear safety. However the reached quality of fabrication of the WWER fuel rods providing their high reliability (the probability of fuel rod failure in the course of one operation year is not higher than 10 -5 ) as well as due to the levels of the WWER fuel rod depressurization actually attainable in the normal conditions of

Fuel production for LWRs - MOX fuel aspects

International Nuclear Information System (INIS)

Deramaix, P.

2005-01-01

Plutonium recycling in Light Water Reactors is today an industrial reality. It is recycled in the form of (U, Pu)O 2 fuel pellets (MOX), fabricated to a large extent according to UO 2 technology and pellet design. The similarity of physical, chemical, and neutron properties of both fuels also allows MOX fuel to be burnt in nuclear plants originally designed to burn UO 2 . The industrial processes presently in use or planned are all based on a mechanical blending of UO 2 and PuO 2 powders. To obtain finely dispersed plutonium and to prevent high local concentration of plutonium, the feed materials are micronised. In the BNFL process, the whole (UO 2 , PuO 2 ) blend is micronised by attrition milling. According to the MIMAS process, developed by BELGONUCLEAIRE, a primary blend made of UO 2 containing about 30% PuO 2 is micronised in a ball mill, afterwards this primary blend is mechanically diluted in UO 2 to obtain the specified Pu content. After mixing, the (U, Pu)O 2 powder is pressed and the pellets are sintered. The sintering cover gas contains moisture and 5 v/o H 2 . Moisture increases the sintering process and the U-Pu interdiffusion. After sintering and grinding, the pellets are submitted to severe controls to verify conformity with customer specifications (fissile content, Pu distribution, surface condition, chemical purity, density, microstructure). (author)

High Temperature PEM Fuel Cells and Organic Fuels

DEFF Research Database (Denmark)

Vassiliev, Anton

of the products. The observation of internal reforming was indirectly confirmed by electrochemical impedance spectroscopy, where the best fits were obtained when a Gerischer element describing preceding chemical reaction and diffusion was included in the equivalent circuit of a methanol/air operated cell...... evaporated liquid stream supply to either of the electrodes. A large number of MEAs with different component compositions have been prepared and tested in different conditions using the constructed setups to obtain a basic understanding of the nature of direct DME HT-PEM FC, to map the processes occurring...... inside the cells and to determine the lifetime. Additionally, comparison was made with methanol as fuel, which is the main competitor to DME in direct oxidation of organic fuels in fuel cells. For the reference, measurements have also been done with conventional hydrogen/air operation. All...

Recent advances in ratio primary reference measurement procedures (definitive methods) and their use in certification of reference materials and controlling assigned values in proficiency testing

International Nuclear Information System (INIS)

Dybczyñski, R.S.; Polkowska-Motrenko, H.; Chajduk, E.; Danko, B.; Pyszynska, M.

2014-01-01

The idea of definitive methods based on radiochemical neutron activation analysis (RNAA), consists in combination of neutron activation with the highly selective and quantitative post-irradiation isolation of the desired radionuclide by column chromatography followed by γ-ray spectrometric measurement. The principles of construction of such methods, which were devised in the Institute of Nuclear Chemistry and Technology, are reminded and the significance of these methods for analytical quality assurance is emphasized. According to VIM 3 nomenclature these methods may be called: ratio primary reference measurement procedures (RPRMPs). RPRMP for the determination of Se is briefly presented and its use for checking the accuracy of 'assigned values' established by expert laboratories in some proficiency tests, is demonstrated

The properties of spherical fuel elements and its behavior in the modular HTR

International Nuclear Information System (INIS)

Lohnert, G.H.; Ragoss, H.

1985-01-01

The reference fuel element for all future HTR applications in the Federal Republic of Germany as developed by NUKEM/HOBEG in the framework of the 'High temperature Fuel-Cycle Project' had to be scrutinised for its compatibility with all the other design principles of the modular HTR, or possibly for restrictions forced upon reactor layout. This reference fuel element can be characterized by the following features: moulded spherical fuel element of 60 mm in diameter with fuel free shell of 5 mm thickness, based on carbon matrix; low enriched uranium (U/Pu fuel cycle); UO 2 fuel kernels; TRISO coating (pyrocarbon and additional SiC layers)

NWTS conceptual reference repository description (CRRD). Volume 1. Summary

International Nuclear Information System (INIS)

1981-05-01

This report is a Conceptual Reference Repository Description (CRRD) for the terminal storage of spent, unreprocessed fuel assemblies and low-level transuranic waste. It is based on the following three previously prepared conceptual design reports: National Waste Terminal Storage in a Bedded Salt Formation for Spent Unreprocessed Fuel (NWTS-R2), prepared by Kaiser Engineers; National Waste Terminal Storage Repository Number 1 (NWTS-R1), prepared by Stearns-Roger; Spent Fuel Receiving and Packaging Facility Conceptual Design, prepared by Rockwell Hanford Operations and Kaiser Engineers. The CRRD is not itself a conceptual design report, but combines the key features of the above three reports into a description of a reference repository. This description is to be used as a basis for the preparation of a preliminary information report for the repositories. The CRRD discusses from a conceptual engineering standpoint the structures, systems, equipment, and operations necessary to (1) receive unreprocessed spent fuel assemblies via standard shipping casks, (2) offload these assemblies, (3) place the assemblies in canisters, (4) transport the canisters to underground storage locations in the salt dome, and (5) place these canisters in terminal storage. The CRRD also elaborates on the concepts for the retrieval and recovery of the spent fuel after burial; describes the development of the shafts and the underground areas,as well as the supporting operational utility and administrative features of the repository; and discusses the impacts on the reference repository of (1) changes in the waste package concept, (2) dry receiving and dry temporary storage of waste, (3) terminal storage of reprocessing waste, and (4) alternative underground development schemes

Reactivity Measurements On Burnt And Reference Fuel Samples In LWR-PROTEUS Phase II

International Nuclear Information System (INIS)

Murphy, M.; Jatuff, F.; Grimm, P.; Seiler, R.; Luethi, A.; Van Geemert, R.; Brogli, R.; Chawla, R.; Meier, G.; Berger, H.-D.

2003-01-01

During the year 2002, the PROTEUS research reactor was used to make a series of reactivity measurements on Pressurised Water Reactor (PWR) burnt fuel samples, and on a series of specially prepared standards. These investigations have been made in two different neutron spectra. In addition, the intrinsic neutron emissions of the burnt fuel samples have been determined. (author)

Near surface spent fuel storage: environmental issues

International Nuclear Information System (INIS)

Nelson, I.C.; Shipler, D.B.; McKee, R.W.; Glenn, R.D.

1979-01-01

Interim storage of spent fuel appears inevitable because of the lack of reprocessing plants and spent fuel repositories. This paper examines the environmental issues potentially associated with management of spent fuel before disposal or reprocessing in a reference scenario. The radiological impacts of spent fuel storage are limited to low-level releases of noble gases and iodine. Water needed for water basin storage of spent fuel and transportation accidents are considered; the need to minimize the distance travelled is pointed out. Resource commitments for construction of the storage facilities are analyzed

Fuel bundle for nuclear reactor

International Nuclear Information System (INIS)

Long, J.W.; Flora, B.S.; Ford, K.L.

1977-01-01

The invention concerns a new, simple and inexpensive system for assembling and dismantling a nuclear reactor fuel bundle. Several fuel rods are fitted in parallel rows between two retaining plates which secure the fuel rods in position and which are maintained in an assembled position by means of several stays fixed to the two end plates. The invention particularly refers to an improved apparatus for fixing the stays to the upper plate by using locking fittings secured to rotating sleeves which are applied against this plate [fr

Transportation of irradiated fuel elements

International Nuclear Information System (INIS)

1980-01-01

A critique is presented of current methods of transporting spent nuclear fuel and the inadequacies of the associated contingency plans, with particular reference to the transportation of irradiated fuel through London. Anti-nuclear and pro-nuclear arguments are presented on a number of factors, including tests on flasks, levels of radiation exposure, routine transport arrangements and contingency arrangements. (U.K.)

Standard Test Method for Calibration of Non-Concentrator Photovoltaic Secondary Reference Cells

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This test method covers calibration and characterization of secondary terrestrial photovoltaic reference cells to a desired reference spectral irradiance distribution. The recommended physical requirements for these reference cells are described in Specification E1040. Reference cells are principally used in the determination of the electrical performance of a photovoltaic device. 1.2 Secondary reference cells are calibrated indoors using simulated sunlight or outdoors in natural sunlight by reference to a primary reference cell previously calibrated to the same desired reference spectral irradiance distribution. 1.3 Secondary reference cells calibrated according to this test method will have the same radiometric traceability as the of the primary reference cell used for the calibration. Therefore, if the primary reference cell is traceable to the World Radiometric Reference (WRR, see Test Method E816), the resulting secondary reference cell will also be traceable to the WRR. 1.4 This test method appli...

The relationship between income and personal vehicle fuel efficiency and associated equity concerns for the fuel tax.

Science.gov (United States)

2011-03-01

The fuel tax, which is assessed on the physical amount of fuel purchased by the consumer, is the primary : means of funding roadway development at the state and national level. However, because it is assessed on a : gallon basis, drivers of vehicles ...

Full fuel-cycle comparison of forklift propulsion systems

International Nuclear Information System (INIS)

Gaines, L.L.; Elgowainy, A.; Wang, M.Q.

2008-01-01

Hydrogen has received considerable attention as an alternative to fossil fuels. The U.S. Department of Energy (DOE) investigates the technical and economic feasibility of promising new technologies, such as hydrogen fuel cells. A recent report for DOE identified three near-term markets for fuel cells: (1) Emergency power for state and local emergency response agencies, (2) Forklifts in warehousing and distribution centers, and (3) Airport ground support equipment markets. This report examines forklift propulsion systems and addresses the potential energy and environmental implications of substituting fuel-cell propulsion for existing technologies based on batteries and fossil fuels. Industry data and the Argonne Greenhouse Gases, Regulated Emissions, and Energy Use in Transportation (GREET) model are used to estimate full fuel-cycle emissions and use of primary energy sources, back to the primary feedstocks for fuel production. Also considered are other environmental concerns at work locations. The benefits derived from using fuel-cell propulsion are determined by the sources of electricity and hydrogen. In particular, fuel-cell forklifts using hydrogen made from the reforming of natural gas had lower impacts than those using hydrogen from electrolysis

Full fuel-cycle comparison of forklift propulsion systems.

Energy Technology Data Exchange (ETDEWEB)

Gaines, L. L.; Elgowainy, A.; Wang, M. Q.; Energy Systems

2008-11-05

Hydrogen has received considerable attention as an alternative to fossil fuels. The U.S. Department of Energy (DOE) investigates the technical and economic feasibility of promising new technologies, such as hydrogen fuel cells. A recent report for DOE identified three near-term markets for fuel cells: (1) Emergency power for state and local emergency response agencies, (2) Forklifts in warehousing and distribution centers, and (3) Airport ground support equipment markets. This report examines forklift propulsion systems and addresses the potential energy and environmental implications of substituting fuel-cell propulsion for existing technologies based on batteries and fossil fuels. Industry data and the Argonne Greenhouse Gases, Regulated Emissions, and Energy Use in Transportation (GREET) model are used to estimate full fuel-cycle emissions and use of primary energy sources, back to the primary feedstocks for fuel production. Also considered are other environmental concerns at work locations. The benefits derived from using fuel-cell propulsion are determined by the sources of electricity and hydrogen. In particular, fuel-cell forklifts using hydrogen made from the reforming of natural gas had lower impacts than those using hydrogen from electrolysis.

Continuity of care: what matters to women when they are referred from primary to secondary care during labour? a qualitative interview study in the Netherlands.

Science.gov (United States)

de Jonge, Ank; Stuijt, Rosan; Eijke, Iva; Westerman, Marjan J

2014-03-17

Continuity of care during labour is important for women. Women with an intrapartum referral from primary to secondary care look back more negatively on their birh experience compared to those who are not referred. It is not clear which aspects of care contribute to this negative birth experience. This study aimed to explore in-depth the experiences of women who were referred during labour from primary to secondary care with regard to the different aspects of continuity of care. A qualitative interview study was conducted in the Netherlands among women who were in primary care at the onset of labour and were referred to secondary care before the baby was born. Through purposive sampling 27 women were selected. Of these, nine women planned their birth at home, two in an alongside midwifery unit and 16 in hospital. Thematic analysis was used. Continuity of care was a very important issue for women because it contributed to their feeling of safety during labour. Important details were sometimes not handed over between professionals within and between primary and secondary care, in particular about women's personal preferences. In case of referral of care from primary to secondary care, it was important for women that midwives handed over the care in person and stayed until they felt safe with the hospital team. Personal continuity of care, in which case the midwife stayed until the end of labour, was highly appreciated but not always expected.Fear of transportion during or after labour was a reason for women to choose hospital birth but also to opt for home birth. Choice of place of birth emerged as a fluid concept; most women planned their place of birth during pregnancy and were aware that they would spend some time at home and possibly some time in hospital. In case of referral from primary to secondary care during labour, midwives should hand over their care in person and preferrably stay with women throughout labour. Planned place of birth should be regarded as a

A comparative study of the oxidation characteristics of two gasoline fuels and an n-heptane/iso-octane surrogate mixture

KAUST Repository

Javed, Tamour

2015-01-01

Ignition delay times and CO, H2O, OH and CO2 time-histories were measured behind reflected shock waves for two FACE (Fuels for Advanced Combustion Engines) gasolines and one PRF (Primary Reference Fuel) blend. The FACE gasolines chosen for this work are primarily paraffinic and have the same octane rating (∼RON = 84) as the PRF blend, but contain varying amounts of iso- and n-paraffins. Species time-histories and ignition delay times were measured using laser absorption methods over a temperature range of 1350-1550 K and pressures near 2 atm. Measured species time-histories and ignition delay times of the PRF blend and the two FACE fuels agreed reasonably well. However, when compared to recent gasoline surrogate mechanisms, the simulations did not capture some of the kinetic trends found in the species profiles. To our knowledge, this work provides some of the first shock tube species time-history data for gasoline fuels and PRF surrogates and should enable further improvements in detailed kinetic mechanisms of gasoline fuels.

ITER fuel cycle

International Nuclear Information System (INIS)

Leger, D.; Dinner, P.; Yoshida, H.

1991-01-01

Resulting from the Conceptual Design Activities (1988-1990) by the parties involved in the International Thermonuclear Experimental Reactor (ITER) project, this document summarizes the design requirements and the Conceptual Design Descriptions for each of the principal subsystems and design options of the ITER Fuel Cycle conceptual design. The ITER Fuel Cycle system provides for the handling of all tritiated water and gas mixtures on ITER. The system is subdivided into subsystems for fuelling, primary (torus) vacuum pumping, fuel processing, blanket tritium recovery, and common processes (including isotopic separation, fuel management and storage, and processes for detritiation of solid, liquid, and gaseous wastes). After an introduction describing system function and conceptual design procedure, a summary of the design is presented including a discussion of scope and main parameters, and the fuel design options for fuelling, plasma chamber vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary and common processes. Design requirements are defined and design descriptions are given for the various subsystems (fuelling, plasma vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary/common processes). The document ends with sections on fuel cycle design integration, fuel cycle building layout, safety considerations, a summary of the research and development programme, costing, and conclusions. Refs, figs and tabs

Reference data about petroleum fiscality

International Nuclear Information System (INIS)

2006-01-01

This paper explains the different taxes existing in France for the petroleum products (domestic tax on petroleum products, added-value tax), the share of taxes in the retail price, the differences with other European countries, the French Government fiscal receipts and budget. Some information forms are attached to this document and concern: the formation of fuel prices (upstream, refining and transport-distribution margins), the evolution of annual average transport-distribution margins, some reference data about world petroleum markets (supply and demand, prices, market data), and some reference data about the role of oil companies on the petroleum market. (J.S.)

Nuclear fuels and development of nuclear fuel elements

International Nuclear Information System (INIS)

Sundaram, C.V.; Mannan, S.L.

1989-01-01

Safe, reliable and economic operation of nuclear fission reactors, the source of nuclear power at present, requires judicious choice, careful preparation and specialised fabrication procedures for fuels and fuel element structural materials. These aspects of nuclear fuels (uranium, plutonium and their oxides and carbides), fuel element technology and structural materials (aluminium, zircaloy, stainless steel etc.) are discussed with particular reference to research and power reactors in India, e.g. the DHRUVA research reactor at BARC, Trombay, the pressurised heavy water reactors (PHWR) at Rajasthan and Kalpakkam, and the Fast Breeder Test Reactor (FBTR) at Kalpakkam. Other reactors like the gas-cooled reactors operating in UK are also mentioned. Because of the limited uranium resources, India has opted for a three-stage nuclear power programme aimed at the ultimate utilization of her abundant thorium resources. The first phase consists of natural uranium dioxide-fuelled, heavy water-moderated and cooled PHWR. The second phase was initiated with the attainment of criticality in the FBTR at Kalpakkam. Fast Breeder Reactors (FBR) utilize the plutonium and uranium by-products of phase 1. Moreover, FBR can convert thorium into fissile 233 U. They produce more fuel than is consumed - hence, the name breeders. The fuel parameters of some of the operating or proposed fast reactors in the world are compared. FBTR is unique in the choice of mixed carbides of plutonium and uranium as fuel. Factors affecting the fuel element performance and life in various reactors e.g. hydriding of zircaloys, fuel pellet-cladding interaction etc. in PHWR and void swelling; irradiation creep and helium embrittlement of fuel element structural materials in FBR are discussed along with measures to overcome some of these problems. (author). 15 refs., 9 tabs., 23 figs

Prevention of nuclear fuel cladding materials corrosion

International Nuclear Information System (INIS)

Yang, K.R.; Yang, J.C.; Lee, I.C.; Kang, H.D.; Cho, S.W.; Whang, C.K.

1983-01-01

The only way which could be performed by the operator of nuclear power plant to minimizing the degradation of nuclear fuel cladding material is to control the water quality of primary coolant as specified standard conditions which dose not attack the cladding material. If the water quality of reactor coolant does not meet far from the specification, the failure will occure not only cladding material itself but construction material of primary system which contact with the coolant. The corrosion product of system material are circulate through the whole primary system with the coolant and activated by the neutron near the reactor core. The activated corrosion products and fission products which released from fuel rod to the coolant, so called crud, will repeate deposition and redeposition continuously on the fuel rod and construction material surface. As a result we should consider heat transfer problem. In this study following activities were performed; 1. The crud sample was taken from the spent fuel rod surface of Kori unit one and analized for radioactive element and non radioactive chemical species. 2. The failure mode of nuclear fuel cladding material was estimated by the investigation of releasing type of fission products from the fuel rod to the reactor coolant using the iodine isotopes concentration of reactor coolants. 3. A study was carried out on the sipping test results of spent fuel and a discussion was made on the water quality control records through the past three cycle operation period of Kori unit one plant. (Author)

The Influence of fuel Diversification Concerning Vapour and Gas Electric Power Performance

International Nuclear Information System (INIS)

Amiral-Aziz; Panca-Porakusuma

2007-01-01

Nowadays consumption of electricity in Indonesia has increased. The oil fuel as the primary fuel for power plant installation has decreased the amount of resources and makes the oil fuel prices very high. In the recent years it has been developed the Combined Cycle technology with Fuel Diversification from the oil fuel (HSD) to natural gas in power plant installation. From the investigation it could be concluded that in the Combined Cycle power plant the production cost using natural gas for the primary fuel cheaper than HSD fuel. The production cost of the combined cycle with configuration of 1-1-1 for 199 MW load is Rp. 172/kWh for natural gas fuel and Rp 941/kWh for HSD fuel. (author)

Mineralogic and petrologic investigation of pre-test core samples from the spent fuel test-climax

International Nuclear Information System (INIS)

Ryerson, F.J.; Qualheim, B.J.

1983-12-01

Pre-test samples obtained from just inside the perimeter of the canister emplacement holes of the Spent Fuel Test-Climax have been characterized by petrographic and microanalytical techniques. The primary quartz monzonite has undergone various degrees of hydrothermal alteration as a result of natural processes. Alteration is most apparent on primary plagioclase and biotite. The most common secondary phases on plagioclase are muscovite and calcite, while the most common secondary phases on biotite are epidote and chlorite. The major alteration zones encountered are localized along filled fractures, i.e. veins. The thickness and mineralogy of the alteration zones can be correlated with the vein mineralogy, becoming wider and more complex mineralogically when the veins contain calcite. 7 references, 10 figures, 4 tables

The provision of Primary Health Care in two rural districts of the Eastern Cape Province with particular reference to human resources and accessibility

Directory of Open Access Journals (Sweden)

M. Thipanyana

1998-09-01

Full Text Available The provision of Primary Health Care Services (PHC is still a problem in developing countries like South Africa. In other countries, one finds enough human resources whilst in other countries there may be enough material resources. A both qualitative and quantitative research was conducted at Mqanduli and part of the Eastern Elliotdale districts in the Eastern Cape Province with the aim of investigating the provision of Primary health Care Services, reference was made to the availability of human resources and accessibility of PHC services.

Design considerations, operating and maintenance experience with wet storage of Ontario Hydro's used fuel

International Nuclear Information System (INIS)

Frost, C.R.

1989-01-01

The characteristics of Ontario Hydro's fuel and at-reactor used fuel storage water pools (or used fuel bays) are described. There are two types of bay, known respectively as primary bays and auxiliary bays, used for at- reactor used fuel storage. Used fuel is discharged remotely from Ontario Hydro's reactors to the primary bays for initial storage and cooling. The auxiliary bays are used to receive and store fuel after its initial cooling in the primary bay, and provide additional storage capacity as needed. With on- power fueling of reactors, each reactor of greater than 500 MW(e) net discharges an average of 10 or more used fuel bundles to bay storage every full power day. The logistics of handling such large quantities of used fuel bundles (corresponding to about 300 te/year of uranium for a 4 unit station) present a challenge to designers and operators. The major considerations in used fuel bay design, including site- specific requirements, reliability and quality assurance, are discussed

Consideration of black carbon and primary organic carbon emissions in life-cycle analysis of Greenhouse gas emissions of vehicle systems and fuels.

Science.gov (United States)

Cai, Hao; Wang, Michael Q

2014-10-21

The climate impact assessment of vehicle/fuel systems may be incomplete without considering short-lived climate forcers of black carbon (BC) and primary organic carbon (POC). We quantified life-cycle BC and POC emissions of a large variety of vehicle/fuel systems with an expanded Greenhouse gases, Regulated Emissions, and Energy use in Transportation model developed at Argonne National Laboratory. Life-cycle BC and POC emissions have small impacts on life-cycle greenhouse gas (GHG) emissions of gasoline, diesel, and other fuel vehicles, but would add 34, 16, and 16 g CO2 equivalent (CO2e)/mile, or 125, 56, and 56 g CO2e/mile with the 100 or 20 year Global Warming Potentials of BC and POC emissions, respectively, for vehicles fueled with corn stover-, willow tree-, and Brazilian sugarcane-derived ethanol, mostly due to BC- and POC-intensive biomass-fired boilers in cellulosic and sugarcane ethanol plants for steam and electricity production, biomass open burning in sugarcane fields, and diesel-powered agricultural equipment for biomass feedstock production/harvest. As a result, life-cycle GHG emission reduction potentials of these ethanol types, though still significant, are reduced from those without considering BC and POC emissions. These findings, together with a newly expanded GREET version, help quantify the previously unknown impacts of BC and POC emissions on life-cycle GHG emissions of U.S. vehicle/fuel systems.

In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

Energy Technology Data Exchange (ETDEWEB)

Caprioli, Sara

2004-04-01

A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

Discussion on the re-irradiated fuel assembly with damaged guide vanes

International Nuclear Information System (INIS)

Li Ligang

2013-01-01

In January 2011, during the second plant of CNNC Nuclear Power Operations Management Co., Ltd.(hereinafter referred to as the second plant) refueling outage, the visual inspection found the guide vanes of fuel assembly A had felling off. After the National Nuclear Safety Administration (NNSA) estimated and approved, the fuel assembly A was reloaded in the specified location of reactor core. During the refueling outage in March 2012, the fuel assembly A was removed again from the reactor core. Visual inspection confirmed that the fuel assembly A was complete and without abnormal changes. The practice provides reference for re-irradiated of fuel assembly with the same type of damaged guide vanes, and provides case support for standard development for the same type of re-irradiated fuel assembly with damaged guide vanes. (author)

Towards a reference architecture of fuel-based carbon management systems in the logistics industry

NARCIS (Netherlands)

Iacob, Maria Eugenia; van Sinderen, Marten J.; Steenwijk, M.; Verkroost, P.

2013-01-01

The current practice in the logistics industry is to calculate the carbon footprint of transportation activities based on the distance covered, using long-term fuel consumption averages per kilometer. However, fuel consumption may actually vary over time, because of differences in road

BATING A REFERENCE INSTALLATION BASED ON CONTROLLED-POTENTIAL COULOMETRY METOD IN THE FRAME OF IMPROVING THE STATE PRIMARY STANDARD GET 176 AND ITS MEASUREMENT CAPABILITIES

Directory of Open Access Journals (Sweden)

V. M. Zyskin

2016-01-01

Full Text Available The results of developing of reference installation, based on a controlled-potential coulometry, in the frame of improving the State primary standard of the units of mass (molar fraction and mass (molar concentration of a component in the liquid and solid substances and materials GET 176 are presented. The physical principles of controlled-potential coulometry, content and metrological characteristics of the developed installation are considered. Measurement results of copper, iron and lead contents in the certified reference materials of metals' solutions and CRM of brass produced by BAM, Germany, obtained using reference installation are given.

High Burnup Fuel Performance and Safety Research

Energy Technology Data Exchange (ETDEWEB)

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

2007-03-15

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

Improved capability of U-ZrH fuels

Energy Technology Data Exchange (ETDEWEB)

Gietzen, A J [General Atomic Company, San Diego, CA (United States)

1983-08-01

This paper provides a brief background on TRIGA fuels and a summary of the development of U-ZrH fuels utilizing higher uranium loadings in the alloy. Most of the development work was reported two years ago; however, in this paper some emphasis will be placed upon applications that General Atomic Company is now making with these higher alloy fuels. They are referred to as higher uranium content alloys because their development was not really LEU development; the TRIGA fuel has used low enrichment since its inception.

Influence of hydrazine primary water chemistry on corrosion of fuel cladding and primary circuit components

International Nuclear Information System (INIS)

Iourmanov, V.; Pashevich, V.; Bogancs, J.; Tilky, P.; Schunk, J.; Pinter, T.

1999-01-01

Earlier at Paks 1-4 NPP standard ammonia chemistry was in use. The following station performance indicators were improved when hydrazine primary water chemistry was introduced: occupational radiation exposures of personnel; gamma-radiation dose rates near primary system components during refuelling and maintenance outages. The reduction of radiation exposures and radiation fields were achieved without significant expenses. Recent results of experimental studies allowed to explain the mechanism of hydrazine dosing influence on: corrosion rate of structure materials in primary coolant; behaviour of soluble and insoluble corrosion products including long-life corrosion-induced radionuclides in primary system during steady-state and transient operation modes; radiolytic generation of oxidising radiolytic products in core and its corrosion activity in primary system; radiation situation during refuelling and maintenance outages; foreign material degradation and removal (including corrosion active oxidant species) from primary system during abnormal events. Operational experience and experimental data have shown that hydrazine primary water chemistry allows to reduce corrosion wear and thereby makes it possible to extend the life-time of plant components in primary system. (author)

Integrated risk assessment for spent fuel transportation using developed software

International Nuclear Information System (INIS)

Yun, Mi Rae; Christian, Robby; Kim, Bo Gyung; Almomani, Belal; Ham, Jae Hyun; Kang, Gook Hyun; Lee, Sang hoon

2016-01-01

As on-site spent fuel storage meets limitation of their capacity, spent fuel need to be transported to other place. In this research, risk of two ways of transportation method, maritime transportation and on-site transportation, and interim storage facility were analyzed. Easier and integrated risk assessment for spent fuel transportation will be possible by applying this software. Risk assessment for spent fuel transportation has not been researched and this work showed a case for analysis. By using this analysis method and developed software, regulators can get some insights for spent fuel transportation. For example, they can restrict specific region for preventing ocean accident and also they can arrange spend fuel in interim storage facility avoiding most risky region which have high risk from aircraft engine shaft. Finally, they can apply soft material on the floor for specific stage for on-site transportation. In this software, because we targeted Korea, we need to use Korean reference data. However, there were few Korean reference data. Especially, there was no food chain data for Korean ocean. In MARINRAD, they used steady state food chain model, but it is far from reality. Therefore, to get Korean realistic reference data, dynamic food chain model for Korean ocean need to be developed

Integrated risk assessment for spent fuel transportation using developed software

Energy Technology Data Exchange (ETDEWEB)

Yun, Mi Rae; Christian, Robby; Kim, Bo Gyung; Almomani, Belal; Ham, Jae Hyun; Kang, Gook Hyun [KAIST, Daejeon (Korea, Republic of); Lee, Sang hoon [Keimyung University, Daegu (Korea, Republic of)

2016-05-15

As on-site spent fuel storage meets limitation of their capacity, spent fuel need to be transported to other place. In this research, risk of two ways of transportation method, maritime transportation and on-site transportation, and interim storage facility were analyzed. Easier and integrated risk assessment for spent fuel transportation will be possible by applying this software. Risk assessment for spent fuel transportation has not been researched and this work showed a case for analysis. By using this analysis method and developed software, regulators can get some insights for spent fuel transportation. For example, they can restrict specific region for preventing ocean accident and also they can arrange spend fuel in interim storage facility avoiding most risky region which have high risk from aircraft engine shaft. Finally, they can apply soft material on the floor for specific stage for on-site transportation. In this software, because we targeted Korea, we need to use Korean reference data. However, there were few Korean reference data. Especially, there was no food chain data for Korean ocean. In MARINRAD, they used steady state food chain model, but it is far from reality. Therefore, to get Korean realistic reference data, dynamic food chain model for Korean ocean need to be developed.

Evaluation model for PWR irradiated fuel

International Nuclear Information System (INIS)

Gomes, I.C.

1983-01-01

The individual economic value of the plutonium isotopes for the recycle of the PWR reactor is investigated, assuming the existence of an market for this element. Two distinct market situations for the stages of the fuel cycle are analysed: one for the 1972 costs and the other for costs of 1982. Comparisons are made for each of the two market situations concerning enrichment of the U-235 in the uranium fuel that gives the minimum cost in the fuel cycle. The method adopted to establish the individual value of the plutonium isotopes consists on the economical analyses of the plutonium fuel cycle for four different isotopes mixtures refering to the uranium fuel cycle. (Author) [pt

Reprocessing of fast neutron reactor fuel

International Nuclear Information System (INIS)

Bourgeois, M.

1981-05-01

A PUREX process specially adapted to fast neutron reactor fuels is employed. The results obtained indicate that the aqueous process can be applied to this type of fuel: almost 10 years operation at the AT 1 plant which processes fuel from RAPSODIE; the good results obtained at the MARCOULE pilot plant on large batches of reference fuels. The CEA is continuing its work to transfer this technology onto an industrial scale. Industrial prototypes and the launching of the TOR (traitement d'oxydes rapides) project will facilitate this transfer. In 1984, it is expected that fast fuels will be able to be processed on a significant scale and that supplementary R and D facilities will be available [fr

The nuclear fuel management in a new market: The Spanish Case

International Nuclear Information System (INIS)

Gonzalez, P.

2001-01-01

The modification of traditional policies in nuclear business of Spanish Electrical Utilities, to meet new regulations is presented in this paper. It would be remarkable the influence of a new environment of significant economical transformation, which show us utilities affected by reorganization processes to face market demands. We will pay special attention to those activities involved in nuclear fuel management. Focusing in the field of Enriched Uranium supply, ENUSA, a state owned company, according to legal framework, played the role of sole supplier. The liberalization of primary energy sources was the turning point that triggered a process of changes. Now, the utilities rule the overall strategy and, not surprisingly they do definitively it is not. The new regulations tend to create conditions for a truly free market. So, competitiveness has become the magic word for almost all electricity producers. But in our case, the first question to be answered should be if free market does really mean a threat for us, or if it is just and external reference that puts no real pressure in our future plans. Lets remember that nuclear fuel cost in Spain is much lower than any other fuel cost. The working model we have defined is oriented to get benefit from market opportunities, improve flexibility and avoiding uncertainties in planning fuel cost strategy. (Author)

Irradiation experiments of 3rd, 4th and 5th fuel assemblies by an in-pile gas loop, OGL-1

International Nuclear Information System (INIS)

Fukuda, Kousaku; Kobayashi, Fumiaki; Hayashi, Kimio; Minato, Kazuo; Kikuchi, Teruo; Adachi, Mamoru; Iwamoto, Kazumi; Ikawa, Katsuichi; Itami, Hiroharu.

1986-07-01

Three irradiation experiments for 3rd, 4th and 5th fuel assemblies which had been composed of VHTR reference coated particle fuels and graphite components were carried out by an in-pile gas loop, OGL-1 during 1979 and 1982. The main purposes of these experiments were to study on bowing of the fuel rod by irradiation for the 3rd fuel assembly, to study on fuel behavior under relatively low burnup irradiation for the 4th fuel assembly, and to study on fuel behavior up to full burnup of VHTR design for the 5th fuel assembly. For understanding in-pile fuel behavior, fractional releases of fission gases from each fuel assembly were estimated by measuring the fission gas concentrations in the primary loop of OGL-1. The post-irradiation examination (PIE) was carried out extensively on the fuel block, the fuel rods and the fuel compacts in Tokai Hot Laboratory. Also, made were the measurements of metallic fission product distributions in the fuel assemblies and the fuel rods. The results in these experiments were given as follows ; bowing of the fuel rod in the 3rd fuel assembly was 0.7 mm, but integrity of the rod was kept under irradiation. Fractional release of the fission gas from the 4th fuel assembly remained in the order of 10 -7 during irradiation, suggesting that the fuel performance was excellent. The fractional release from the 5th fuel assembly, on the other hand, was in the order of 10 -5 which was the same level in the VHTR design. (author)

Compatibility analysis of DUPIC fuel (part 3) - radiation physics analysis

Energy Technology Data Exchange (ETDEWEB)

Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Park, Byung Yun; Koh, Young Kown

2000-04-01

As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points and by -10% for the corner of the calandria subshells and annular plates in the calandria, respectivdely. Finally, the feasibility study of the DUPIC fuel handling was performed, which has shown that all handling and inspection of the DUPIC fuel bundles be done remotely and behind a shielding wall. For the transportation of the DUPIC fuel, the preliminary study has shown that there shold be no technical problem th design a transportation cask for the fresh and spent DUPIC fuel bundles. For the storage of the fresh and spent DUPIC fuels, there is no the criticality safety problem unless the fuel bundle geometry is destroyed.

Compatibility analysis of DUPIC fuel (part 3) - radiation physics analysis

International Nuclear Information System (INIS)

Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Park, Byung Yun; Koh, Young Kown

2000-04-01

As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points and by -10% for the corner of the calandria subshells and annular plates in the calandria, respectivdely. Finally, the feasibility study of the DUPIC fuel handling was performed, which has shown that all handling and inspection of the DUPIC fuel bundles be done remotely and behind a shielding wall. For the transportation of the DUPIC fuel, the preliminary study has shown that there shold be no technical problem th design a transportation cask for the fresh and spent DUPIC fuel bundles. For the storage of the fresh and spent DUPIC fuels, there is no the criticality safety problem unless the fuel bundle geometry is destroyed

Special topics reports for the reference tandem mirror fusion breeder. Volume 2. Reactor safety assessment

International Nuclear Information System (INIS)

Maya, I.; Hoot, C.G.; Wong, C.P.C.; Schultz, K.R.; Garner, J.K.; Bradbury, S.J.; Steele, W.G.; Berwald, D.H.

1984-09-01

The safety features of the reference fission suppressed fusion breeder reactor are presented. These include redundancy and overcapacity in primary coolant system components to minimize failure probability, an improved valve location logic to provide for failed component isolation, and double-walled coolant piping and steel guard vessel protection to further limit the extent of any leak. In addition to the primary coolant and decay heat removal system, reactor safety systems also include an independent shield cooling system, the module safety/fuel transfer coolant system, an auxiliary first wall cooling system, a psssive dump tank cooling system based on the use of heat pipes, and several lithium fire suppression systems. Safety system specifications are justified based on the results of thermal analysis, event tree construction, consequence calculations, and risk analysis. The result is a reactor design concept with an acceptably low probability of a major radioactivity release. Dose consequences of maximum credible accidents appear to be below 10CFR100 regulatory limits

Fusion reactor fuel processing

International Nuclear Information System (INIS)

Johnson, E.F.

1972-06-01

For thermonuclear power reactors based on the continuous fusion of deuterium and tritium the principal fuel processing problems occur in maintaining desired compositions in the primary fuel cycled through the reactor, in the recovery of tritium bred in the blanket surrounding the reactor, and in the prevention of tritium loss to the environment. Since all fuel recycled through the reactor must be cooled to cryogenic conditions for reinjection into the reactor, cryogenic fractional distillation is a likely process for controlling the primary fuel stream composition. Another practical possibility is the permeation of the hydrogen isotopes through thin metal membranes. The removal of tritium from the ash discharged from the power system would be accomplished by chemical procedures to assure physiologically safe concentration levels. The recovery process for tritium from the breeder blanket depends on the nature of the blanket fluids. For molten lithium the only practicable possibility appears to be permeation from the liquid phase. For molten salts the process would involve stripping with inert gas followed by chemical recovery. In either case extremely low concentrations of tritium in the melts would be desirable to maintain low tritium inventories, and to minimize escape of tritium through unwanted permeation, and to avoid embrittlement of metal walls. 21 refs

Substantial improvements of fuel economy. Potentials of electric and hybrid electric vehicles

Energy Technology Data Exchange (ETDEWEB)

Joergensen, K [Technical Univ. of Denmark (Denmark); Nielsen, L H [Forskningscenter Risoe (Denmark)

1996-12-01

This paper evaluates the scope for improvement of the energy and environmental impacts of road traffic by means of electrical and hybrid electric propulsion. These technologies promise considerable improvements of the fuel economy of vehicles compared to the present vehicle types as well as beneficial effects for the energy and traffic system. The paper - based on work carried out in the project `Transportation fuel based on renewable energy`, funded by the National Energy Agency of Denmark and carried out by Department of Buildings and Energy, Technical University of Denmark and System Analysis Department, Risoe National Laboratory - assesses the potentials for reduction of the primary energy consumption and emissions, and points to the necessary technical development to reap these benefits. A case study concerning passenger cars is analysed by means of computer simulations, comparing electric and hybrid electric passenger car to an equivalent reference vehicle (a conventional gasoline passenger car). (au) 10 refs.

In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched 235U fuel pins

International Nuclear Information System (INIS)

Caprioli, Sara

2004-04-01

A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched 235 U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are rather

Comparability of reference values

International Nuclear Information System (INIS)

Rossbach, M.; Stoeppler, M.

1993-01-01

Harmonization of certified values in Reference Materials (RMs) can be carried out by applying nuclear analytical techniques to RMs of various matrix types and concentration levels. Although RMs generally should not be used as primary standards the cross evaluation of concentrations in RMs leads to better compatibility of reference values and thus to a greater agreement between analytical results from different laboratories using these RMs for instrument calibration and quality assurance. (orig.)

FY2015 ceramic fuels development annual highlights

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-22

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

FY2016 Ceramic Fuels Development Annual Highlights

Energy Technology Data Exchange (ETDEWEB)

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

2017-01-24

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

Reviewing reactor engineering and fuel handling

International Nuclear Information System (INIS)

1991-12-01

Experience has shown that the better operating nuclear power plants have well defined and effectively administered policies and procedures for governing reactor engineering and fuel handling (RE and FH) activities. This document provides supplementary guidance to OSART experts for evaluating the RE and FH programmes and activities at a nuclear power plant and assessing their effectiveness and adequacy. It is in no way intended to conflict with existing regulations and rules, but rather to exemplify those characteristics and features that are desirable for an effective, well structured RE and FH programme. This supplementary guidance addresses those aspects of RE and FH activities that are required in order to ensure optimum core operation for a nuclear reactor without compromising the limits imposed by the design, safety considerations of the nuclear fuel. In the context of this document, reactor engineering refers to those activities associated with in-core fuel and reactivity management, whereas fuel handling refers to the movement, storage, control and accountability of unirradiated and irradiated fuel. The document comprises five main sections and several appendices. In Section 2 of this guide, the essential aspects of an effective RE and FH programme are discussed. In Section 3, the various types of documents and reference materials needed for the preparatory work and investigation are listed. In Section 4, specific guidelines for investigation of RE and FH programmes are presented. In Section 5, the essential attributes of an excellent RE and FH programme are listed. The supplementary guidance is concluded with a series of appendices exemplifying the various qualities and attributes of a sound, well defined RE and FH programme

Spent fuel storage requirements

International Nuclear Information System (INIS)

Fletcher, J.

1982-06-01

Spent fuel storage requirements, as projected through the year 2000 for U.S. LWRs, were calculated using information supplied by the utilities reflecting plant status as of December 31, 1981. Projections through the year 2000 combined fuel discharge projections of the utilities with the assumed discharges of typical reactors required to meet the nuclear capacity of 165 GWe projected by the Energy Information Administration (EIA) for the year 2000. Three cases were developed and are summarized. A reference case, or maximum at-reactor (AR) capacity case, assumes that all reactor storage pools are increased to their maximum capacities as estimated by the utilities for spent fuel storage utilizing currently licensed technologies. The reference case assumes no transshipments between pools except as currently licensed by the Nuclear Regulatory Commission (NRC). This case identifies an initial requirement for 13 MTU of additional storage in 1984, and a cumulative requirement for 14,490 MTU additional storage in the year 2000. The reference case is bounded by two alternative cases. One, a current capacity case, assumes that only those pool storage capacity increases currently planned by the operating utilities will occur. The second, or maximum capacity with transshipment case, assumes maximum development of pool storage capacity as described above and also assumes no constraints on transshipment of spent fuel among pools of reactors of like type (BWR, PWR) within a given utility. In all cases, a full core discharge capability (full core reserve or FCR) is assumed to be maintained for each reactor, except that only one FCR is maintained when two reactors share a common pool. For the current AR capacity case the indicated storage requirements in the year 2000 are indicated to be 18,190 MTU; for the maximum capacity with transshipment case they are 11,320 MTU

EPRI PWR primary water chemistry guidelines revision

International Nuclear Information System (INIS)

McElrath, Joel; Fruzzetti, Keith

2014-01-01

EPRI periodically updates the PWR Primary Water Chemistry Guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). The last revision of the PWR water chemistry guidelines identified an optimum primary water chemistry program based on then-current understanding of research and field information. This new revision provides further details with regard to primary water stress corrosion cracking (PWSCC), fuel integrity, and shutdown dose rates. A committee of industry experts, including utility specialists, nuclear steam supply system (NSSS) and fuel vendor representatives, Institute of Nuclear Power Operations (INPO) representatives, consultants, and EPRI staff collaborated in reviewing the available data on primary water chemistry, reactor water coolant system materials issues, fuel integrity and performance issues, and radiation dose rate issues. From the data, the committee updated the water chemistry guidelines that all PWR nuclear plants should adopt. The committee revised guidance with regard to optimization to reflect industry experience gained since the publication of Revision 6. Among the changes, the technical information regarding the impact of zinc injection on PWSCC initiation and dose rate reduction has been updated to reflect the current level of knowledge within the industry. Similarly, industry experience with elevated lithium concentrations with regard to fuel performance and radiation dose rates has been updated to reflect data collected to date. Recognizing that each nuclear plant owner has a unique set of design, operating, and corporate concerns, the guidelines committee has retained a method for plant-specific optimization. Revision 7 of the Pressurized Water Reactor Primary Water Chemistry Guidelines provides guidance for PWR primary systems of all manufacture and design. The guidelines continue to emphasize plant

Relationship between basic nuclear data and LWR fuel cycle parameters

International Nuclear Information System (INIS)

Becker, M.; Harris, D.R.; Quan, B.; Ryskamp, J.M.

1979-01-01

An interactive system has been developed at RPI to analyze the sensitivity of water reactor fuel cycle parameters and costs to uncertainties in nuclear data. A sequence of batch depletion, core analysis, and fuel cost codes (referred to as Path B) determines the changes in fuel cycle parameters and costs for changes in few-group microscopic cross sections, in fission yields, and in decay data. For cases that are found to be significant from Part B analysis, the sensitivities of few-group data to basic nuclear data are determined by detailed calculations (referred to as Path A). Analyses of pressurized and boiling water reactors with recycle and throwaway options show substantial sensitivities of fuel cycle parameters and costs, particularly to thermal and resonance nuclear data for fissile nuclides. The results bring out the importance for power reactor sensitivity analysis of dealing with the full fuel cycle including depletion of initially-loaded fuel and the building-in of actinides and fission products

Physical and chemical effects of low octane gasoline fuels on compression ignition combustion

KAUST Repository

Badra, Jihad

2016-09-30

Gasoline compression ignition (GCI) engines running on low octane gasoline fuels are considered an attractive alternative to traditional spark ignition engines. In this study, three fuels with different chemical and physical characteristics have been investigated in single cylinder engine running in GCI combustion mode at part-load conditions both experimentally and numerically. The studied fuels are: Saudi Aramco light naphtha (SALN) (Research octane number (RON) = 62 and final boiling point (FBP) = 91 °C), Haltermann straight run naphtha (HSRN) (RON = 60 and FBP = 140 °C) and a primary reference fuel (PRF65) (RON = 65 and FBP = 99 °C). Injection sweeps, where the start of injection (SOI) is changed between −60 and −11 CAD aTDC, have been performed for the three fuels. Full cycle computational fluid dynamics (CFD) simulations were executed using PRFs as chemical surrogates for the naphtha fuels. Physical surrogates based on the evaporation characteristics of the naphtha streams have been developed and their properties have been implemented in the engine simulations. It was found that the three fuels have similar combustion phasings and emissions at the conditions tested in this work with minor differences at SOI earlier than −30 CAD aTDC. These trends were successfully reproduced by the CFD calculations. The chemical and physical effects were further investigated numerically. It was found that the physical characteristics of the fuel significantly affect the combustion for injections earlier than −30 CAD aTDC because of the low evaporation rates of the fuel because of the higher boiling temperature of the fuel and the colder in-cylinder air during injection. © 2016 Elsevier Ltd

Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

International Nuclear Information System (INIS)

Weaver, K.D.; Sterbentz, J.; Meyer, M.; Lowden, R.; Hoffman, E.; Wei, T.Y.C.

2004-01-01

The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO 2 ) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

Overview of fuel conversion

International Nuclear Information System (INIS)

Green, A.E.S.

1991-01-01

The conversion of solid fuels to cleaner-burning and more user-friendly solid liquid or gaseous fuels spans many technologies. In this paper, the authors consider coal, residual oil, oil shale, tar sends tires, municipal oil waste and biomass as feedstocks and examine the processes which can be used in the production of synthetic fuels for the transportation sector. The products of mechanical processing to potentially usable fuels include coal slurries, micronized coal, solvent refined coal, vegetable oil and powdered biomall. The thermochemical and biochemical processes considered include high temperature carbide production, liquefaction, gasification, pyrolysis, hydrolysis-fermentation and anaerobic digestion. The products include syngas, synthetic natural gas, methanol, ethanol and other hydrocarbon oxygenates synthetic gasoline and diesel and jet engine oils. The authors discuss technical and economic aspects of synthetic fuel production giving particular attention and literature references to technologies not discussed in the five chapters which follow. Finally the authors discuss economic energy, and environmental aspects of synthetic fuels and their relationship to the price of imported oil

Block Copolymers for Alkaline Fuel Cell Membrane Materials

Science.gov (United States)

2014-07-30

temperature fuel cells including proton exchange membrane fuel cell ( PEMFC ) and alkaline fuel cell (AFC) with operation temperature usually lower than 120...advantages over proton exchange membrane fuel cells ( PEMFCs ) resulting in the popularity of AFCs in the US space program.[8-11] The primary benefit AFC...offered over PEMFC is better electrochemical kinetics on the anode and cathode under the alkaline environment, which results in the ability to use

Fuel Cells in the Waste-to-Energy Chain Distributed Generation Through Non-Conventional Fuels and Fuel Cells

CERN Document Server

McPhail, Stephen J; Moreno, Angelo

2012-01-01

As the availability of fossils fuels becomes more limited, the negative impact of their consumption becomes an increasingly relevant factor in our choices with regards to primary energy sources. The exponentially increasing demand for energy is reflected in the mass generation of by-products and waste flows which characterize current society’s development and use of fossil sources. The potential for recoverable material and energy in these ever-increasing refuse flows is huge, even after the separation of hazardous constituent elements, allowing safe and sustainable further exploitation of an otherwise 'wasted' resource.  Fuel Cells in the Waste-to-Energy Chain explores the concept of waste-to-energy through a 5 step process which reflects the stages during the transformation of  refuse flows to a valuable commodity such as clean energy. By providing selected, integrated alternatives to the current centralized, wasteful, fossil-fuel based infrastructure, Fuel Cells in the Waste-to-Energy Chain explores ho...

Electro-optical fuel pin identification system

International Nuclear Information System (INIS)

Kirchner, T.L.

1978-09-01

A prototype Electro-Optical Fuel Pin Identification System referred to as the Fuel Pin Identification System (FPIS) has been developed by the Hanford Engineering Development Laboratory (HEDL) in support of the Fast Flux Test Facility (FFTF) presently under construction at HEDL. The system is designed to remotely read an alpha-numeric identification number that is roll stamped on the top of the fuel pin end cap. The prototype FPIS consists of four major subassemblies: optical read head, digital compression electronics, video display, and line printer

Hydrogen as automotive fuel

International Nuclear Information System (INIS)

Ambrosini, G.; Ciancia, A.; Pede, G.; Brighigna, M.

1993-01-01

Hydrogen fueled vehicles may just be the answer to the air pollution problem in highly polluted urban environments where the innovative vehicle's air pollution abatement characteristics would justify its high operating costs as compared with those of conventional automotive alternatives. This paper examines the feasibility of hydrogen as an automotive fuel by analyzing the following aspects: the chemical-physical properties of hydrogen in relation to its use in internal combustion engines; the modifications necessary to adapt internal combustion engines to hydrogen use; hydrogen fuel injection systems; current production technologies and commercialization status of hydrogen automotive fuels; energy efficiency ratings; environmental impacts; in-vehicle storage systems - involving the use of hydrides, high pressure systems and liquid hydrogen storage systems; performance in terms of pay-load ratio; autonomous operation; and operating costs. With reference to recent trial results being obtained in the USA, an assessment is also made of the feasibility of the use of methane-hydrogen mixtures as automotive fuels. The paper concludes with a review of progress being made by ENEA (the Italian Agency for New Technology, Energy and the Environment) in the development of fuel storage and electronic fuel injection systems for hydrogen powered vehicles

Methanol fuel update

International Nuclear Information System (INIS)

Colledge, R.; Spacek, J.

1992-01-01

An overview is presented of methanol fuel developments, with particular reference to infrastructure, supply and marketing. Methanol offers reduced emissions, easy handling, is cost effective, can be produced from natural gas, coal, wood, or municipal waste, is a high performance fuel, is safer than gasoline, and contributes to energy security. Methanol supply, environmental benefits, safety/health issues, economics, passenger car economics, status of passenger car technology, buses, methanol and the prosperity initiative, challenges to implementation, and the role of government and original equipment manufacturers are discussed. Governments must assist in the provision of methanol refuelling infrastructure, and in providing an encouraging regulatory atmosphere. Discriminatory and inequitable taxing methods must be addressed, and an air quality agenda must be defined to allow the alternative fuel industry to respond in a timely manner

Fuel cycle studies for the Dragon HTR

Energy Technology Data Exchange (ETDEWEB)

Desoisa, J A; Nunn, R M; Twitchin, A E

1971-02-15

This note reports the progress made at B.N.L. in the study of the fuel cycle for the HTR design described by Daub (1970). The primary purpose of the study is to examine the special problems of the approach to equilibrium fuel cycle.

Reference design description for a geologic repository: Revision 01

International Nuclear Information System (INIS)

1997-09-01

This document describes the current design expectations for a potential geologic repository that could be located at Yucca Mountain in Nevada. This Reference Design Description (RDD) looks at the surface and subsurface repository and disposal container design. Additionally, it reviews the expected long-term performance of the potential repository. In accordance with current legislation, the reference design for the potential repository does not include an interim storage option. The reference design presented allows the disposal of highly radioactive material received from government-owned spent fuel custodian sites; produces high-level waste sites, and commercial spent fuel sites. All design elements meet current federal, state, and local regulations governing the disposal of high-level radioactive waste and protection of the public and the environment. Due to the complex nature of developing a repository, the design will be created in three phases to support Viability Assessment, License Application, and construction. This document presents the current reference design. It will be updated periodically as the design progresses. Some of the details presented here may change significantly as more cost-effective solutions, technical advancements, or changes to requirements are identified

Corrosion control in nuclear fuel reprocessing

International Nuclear Information System (INIS)

Steele, D.F.

1986-01-01

This article looks in detail at tribology-related hazards of corrosion in irradiated fuel reprocessing plants and tries to identify and minimize problems which could contribute to disaster. First, the corrosion process is explained. Then the corrosion aspects at each of four stages in reprocessing are examined, with particular reference to oxide fuel reprocessing. The four stages are fuel receipt and storage, fuel breakdown and dissolution, solvent extraction and product concentration and waste management. Results from laboratory and plant corrosion trails are used at the plant design stage to prevent corrosion problems arising. Operational procedures which minimize corrosion if it cannot be prevented at the design stage, are used. (UK)

Assessment of the prediction capability of the TRANSURANUS fuel performance code on the basis of power ramp tested LWR fuel rods

International Nuclear Information System (INIS)

Pastore, G.; Botazzoli, P.; Di Marcello, V.; Luzzi, L.

2009-01-01

The present work is aimed at assessing the prediction capability of the TRANSURANUS code for the performance analysis of LWR fuel rods under power ramp conditions. The analysis refers to all the power ramp tested fuel rods belonging to the Studsvik PWR Super-Ramp and BWR Inter-Ramp Irradiation Projects, and is focused on some integral quantities (i.e., burn-up, fission gas release, cladding creep-down and failure due to pellet cladding interaction) through a systematic comparison between the code predictions and the experimental data. To this end, a suitable setup of the code is established on the basis of previous works. Besides, with reference to literature indications, a sensitivity study is carried out, which considers the 'ITU model' for fission gas burst release and modifications in the treatment of the fuel solid swelling and the cladding stress corrosion cracking. The performed analyses allow to individuate some issues, which could be useful for the future development of the code. Keywords: Light Water Reactors, Fuel Rod Performance, Power Ramps, Fission Gas Burst Release, Fuel Swelling, Pellet Cladding Interaction, Stress Corrosion Cracking

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

Genetics Home Reference: primary coenzyme Q10 deficiency

Science.gov (United States)

... mutations have occurred in the COQ2 , COQ4 , COQ6 , COQ8A , and COQ8B genes. Smaller numbers of mutations in other COQ genes have also been found ... primary coenzyme Q10 deficiency ... Related Information What is a gene? What is a gene mutation and how do mutations occur? How can gene ...

An Overview of the EPRI PWR Primary Chemistry Program

International Nuclear Information System (INIS)

Perkins, David; Fruzzetti, Keith; Haas, Carey; Wells, Dan

2012-09-01

Primary chemistry controls continue to evolve, impacting long term equipment reliability goals, optimized core designs, and radiation dose management practices. Chemistry initiatives include increased primary system pH (T) , zinc injection, and optimization of primary system hydrogen concentration. Nevertheless, utilities are faced with ever changing challenges as fuel vendors continue to optimize core power densities coupled with longer operating cycles and material replacement efforts. These challenges must be collaboratively addressed by the plant chemists, engineers, and operators. Operational chemistry has changed dramatically over the years with increased primary pH (T) programs requiring some utilities to operate with up to 6 ppm lithium or slightly higher. Coupled with primary pH (T) program optimization, are ongoing EPRI research efforts attempting to develop an optimized hydrogen control program balancing material issues associated with primary water stress corrosion cracking (PWSCC) crack growth rate against fuel concerns associated with increased hydrogen concentrations. One of the most significant primary chemistry changes that effectively balances the demands of materials, fuels, chemistry and dose management strategies is zinc injection into the primary coolant. Since 1994 when Farley initiated zinc injection, zinc injection has been successfully injected at over 70 pressurized water reactors world-wide. Combining operational chemistry with shutdown chemistry controls provides the plant chemist with a technically based and balanced approach to fuel and material integrity as well as dose management strategies. Shutdown chemistry has continually evolved since the 1970's when the chemist was primarily concerned with fission products. Now the chemist must manage corrosion product release, and support Outage Management and Radiation Protection through the performance of a controlled shutdown. In part, this change was driven as plant materials evolved

Software in support of fuel operation in WWERS

International Nuclear Information System (INIS)

Evdokimov, I.A; Novikov, V.V; Ugrumov, A.V; Shishkin, A.A

2013-01-01

A software package comprising computer codes and fuel monitoring tools is under development in Russia in support of WWER fuel operation. The software package includes an expert computer system designed for failure diagnosis in course of reactor operation, prediction of activity evolution in primary coolant and express analysis of pellet-to-cladding mechanical interaction (PCMI) on rod-by-rod basis under normal and transient modes of operation. Coupled with the expert system, the first version of a graphical interface computer program is developed for NPP operating bodies. One of the features of this program is to launch automatically a fuel performance code for a series of detailed calculations for fuel rods with severe PCMI. The particular rods for calculations are determined by the expert system during the express core analysis. A greater attention is paid to recent results in prediction of fuel behavior after a primary failure has occurred. One of the major risks to further operation of leaking fuel comes from secondary fuel degradation due to massive cladding hydriding. Threshold conditions for initiation of secondary hydriding have been found on the basis of physical modeling. Final criteria of secondary failure occurrence were deduced by applying the model to analysis of post-irradiation examinations of leaking WWER fuel. (authors)

Fuel temperature prediction during high burnup HTGR fuel irradiation test. US-JAERI irradiation test for HTGR fuel

International Nuclear Information System (INIS)

Sawa, Kazuhiro; Fukuda, Kousaku; Acharya, R.

1995-01-01

This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for an irradiation test in a removable beryllium position of the High Flux Isotope Reactor(HFIR) at Oak Ridge National Laboratory. This test is being carried out under Annex 2 of the Arrangement between the U.S. Department of Energy and the Japan Atomic Energy Research Institute on Cooperation in Research and Development regarding High-Temperature Gas-cooled Reactors. The fuel used in the test is an advanced type. The advanced fuel was designed aiming at burnup of about 10%FIMA(% fissions per initial metallic atom) which was higher than that of the first charge fuel for the High Temperature Engineering Test Reactor(HTTR) and was produced in Japan. CACA-2, a heavy isotope and fission product concentration calculational code for experimental irradiation capsules, was used to determine time-dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries(HEATING) code was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body, which contains the fuel compacts, and of the primary pressure vessel were determined such that the requirements of running the fuel compacts at an average temperature less than 1250degC and of not exceeding a maximum fuel temperature of 1350degC were met throughout the four cycles of irradiation. The detail design of the capsule was carried out based on this analysis. (author)

Determining reactor fuel elements broken by Cerenkov counting

International Nuclear Information System (INIS)

Guo Juhao; Dong Shiyuan; Feng Yuying

1996-01-01

The basis and method of determining fuel elements broken in a reactor by Cerenkov counting measured with liquid scintillation spectrometer are introduced. The radioactive characteristic of the radiation nuclides generating Cherenkov radiation in the primary water of 200 MW nuclear district heating reactor is analyzed. The activity of the activation products in the primary water and the fission products in the fuel elements are calculated. A feasibility of Cerenkov counting measure was analyzed. This method is simple and quick

Radioactive certified reference materials

International Nuclear Information System (INIS)

Watanabe, Kazuo

2010-01-01

Outline of radioactive certified reference materials (CRM) for the analysis of nuclear materials and radioactive nuclides were described. The nuclear fuel CRMs are supplied by the three institutes: NBL in the US, CETAMA in France and IRMM in Belgium. For the RI CRMs, the Japan Radioisotope Association is engaged in activities concerning supply. The natural-matrix CRMs for the analysis of trace levels of radio-nuclides are prepared and supplied by NIST in the US and the IAEA. (author)

Nuclear-fuel-cycle optimization: methods and modelling techniques

International Nuclear Information System (INIS)

Silvennoinen, P.

1982-01-01

This book present methods applicable to analyzing fuel-cycle logistics and optimization as well as in evaluating the economics of different reactor strategies. After an introduction to the phases of a fuel cycle, uranium cost trends are assessed in a global perspective. Subsequent chapters deal with the fuel-cycle problems faced by a power utility. The fuel-cycle models cover the entire cycle from the supply of uranium to the disposition of spent fuel. The chapter headings are: Nuclear Fuel Cycle, Uranium Supply and Demand, Basic Model of the LWR (light water reactor) Fuel Cycle, Resolution of Uncertainties, Assessment of Proliferation Risks, Multigoal Optimization, Generalized Fuel-Cycle Models, Reactor Strategy Calculations, and Interface with Energy Strategies. 47 references, 34 figures, 25 tables

Wood fuel production technologies in EU countries

Energy Technology Data Exchange (ETDEWEB)

Hakkila, P [Finnish Forest Research Institute, Vantaa (Finland)

1998-12-31

The presentation reviews the major technologies used for the production of fuel chips for heating plants in Europe. Three primary options are considered: production of whole-tree chips from young trees for fuel; integrated harvesting of fiber and energy from thinning based on tree-section system; and production of fuel chips from logging residue in clear-cut areas after fully mechanized logging. The characteristics of the available biomass reserve and proven technology for its recovery are discussed. The employment effects of fuel chip production and the costs of wood fuels are also briefly discussed. (author) 3 refs., 3 figs.

Wood fuel production technologies in EU countries

Energy Technology Data Exchange (ETDEWEB)

Hakkila, P. [Finnish Forest Research Institute, Vantaa (Finland)

1997-12-31

The presentation reviews the major technologies used for the production of fuel chips for heating plants in Europe. Three primary options are considered: production of whole-tree chips from young trees for fuel; integrated harvesting of fiber and energy from thinning based on tree-section system; and production of fuel chips from logging residue in clear-cut areas after fully mechanized logging. The characteristics of the available biomass reserve and proven technology for its recovery are discussed. The employment effects of fuel chip production and the costs of wood fuels are also briefly discussed. (author) 3 refs., 3 figs.

Bioethanol E85 as a fuel for dual fuel diesel engine

International Nuclear Information System (INIS)

Tutak, Wojciech

2014-01-01

Highlights: • An increase in the E85 fraction is changing the nature of the HRR course. • Change of combustion phasing with E85 fraction. • The ignition delay of dual fuel engine decreases with increasing E85 fuel. • Premixed combustion of E85 fuel reduces smoke emissions. - Abstract: This study investigates the potential of E85 fuelling in a diesel engine. Researches were performed using a three-cylinder a direct injection diesel engine. A dual-fuelling technology is implemented such that E85 is introduced into the intake manifold using a port-fuel injector while diesel is injected directly into the cylinder. The primary aim of the study was to determine the operating parameters of the engine powered on E85 bioethanol fuel in dual fuel system. The parameters that were taken into account are: engine efficiency, indicated mean effective pressure, heat release rate, combustion duration and ignition delay, combustion phasing and exhaust toxicity. With E85 fuel participation, NO x and soot emissions were reduced, whereas CO and HC emissions increased considerably. It was found that E85 participation in a combustible mixture reduced the excess air factor for the engine and this led to increased emissions of CO and HC, but decreased emissions of nitrogen oxides and soot

Primary energy savings in desiccant and evaporative cooling-assisted 100% outdoor air system combined with a fuel cell

International Nuclear Information System (INIS)

Kim, Min-Hwi; Dong, Hae-Won; Park, Joon-Young; Jeong, Jae-Weon

2016-01-01

Highlights: • A LD-IDECOAS integrated with a PEMFC was proposed. • A pilot system was installed and tested during cooling operation. • The proposed system powered by the PEMFC saved 21% of the primary energy consumption during cooling. - Abstract: The main purpose of this study involved investigating the primary energy saving potential of a liquid desiccant and evaporative cooling-assisted 100% outdoor air system (LD-IDECOAS) integrated with a proton exchange membrane fuel cell (PEMFC). During the cooling season, the heat produced by the PEMFC was used to regenerate a weak desiccant solution, and the electricity generated was used to operate the LD-IDECOAS. A pilot LD-IDECOAS powered by a PEMFC was installed and operated in an office space to experimentally verify the annual operating energy savings of the proposed system. The findings indicated that the heat reclaimed from the PEMFC saved 42% of the desiccant solution regenerating energy when compared to that in the case of a conventional gas-fired water heater. The results also suggested that the LD-IDECOAS combined with a PEMFC consumed 21% less primary energy when compared with that of a system powered by grid electricity and a conventional gas-fired water heater.

Fuel-motion diagnostics and cineradiography

International Nuclear Information System (INIS)

DeVolpi, A.

1982-09-01

Nuclear and non-nuclear applications of cineradiography are reviewed, with emphasis on diagnostic instrumentation for in-pile transient-reactor safety testing of nuclear fuel motion. The primary instrument for this purpose has been the fast-neutron hodoscope, which has achieved quantitative monitoring of time, location, mass, and velocity of fuel movement under the difficult conditions associated with transient-reactor experiments. Alternative diagnostic devices that have been developed have not matched the capabilities of the hodoscope. Other applications for the fuel-motion diagnostic apparatus are also evolving, including time-integrated radiography and direct time- and space-resolved fuel-pin power monitoring. Although only two reactors are now actively equipped with high-resolution fuel-motion diagnostic systems, studies and tests have been carried out in and for many other reactors

Waste package reference conceptual designs for a repository in salt

International Nuclear Information System (INIS)

1986-02-01

This report provides the reference conceptual waste package designs for the Office of Nuclear Waste Isolation to baseline these designs, thereby establishing the configuration and interface controls necessary, within the Civilian Radioactive Waste Management Program, formerly the National Waste Terminal Storage Program, to proceed in an orderly manner with preliminary design. Included are designs for the current reference defense high-level waste form from the Savannah River Plant, an optimized commercial high-level waste form, and spent fuel which has been disassembled and compacted into a circular bundle containing either 12 pressurized-water reactor or 30 boiling-water reactor assemblies. For compacted spent fuel, it appears economically attractive to standardize the waste package diameter for all fuel types. The reference waste packages consist of the containerized waste form, a low carbon steel overpack, and, after emplacement, a cover of salt. The overpack is a hollow cylinder with a flat head welded to each end. Its design thickness is the sum of the structural thickness required to resist the 15.4-MPa lithostatic pressure plus the corrosion allowance necessary to assure the required structural thickness will exist through the 1000-year containment period. Based on available data and completed analyses, the reference concepts described in this report satisfy all requirements of the US Department of Energy and the US Nuclear Regulatory Commission with reasonable assurance. In addition, sufficient design maturity exists to form a basis for preliminary design; these concepts can be brought under configuration control to serve as reference package designs. Development programs are identified that will be required to support these designs during the licensing process. 19 refs., 37 figs., 31 tabs

Neutron radiographic findings in light water reactor fuel

International Nuclear Information System (INIS)

Domanus, J.C.

1979-06-01

The assessment of neutron radiographs of nuclear fuel elements can be much easier, faster and simpler if reference can be made to typical defects, which can be revealed by neutron radiography. In other fields of industrial radiography collections of reference radiographs, showing typical defects in welding, or casting have been completed and published long ago. Since 1974 neutron radiography is routinely used at Risoe National Laboratory, Denmark, for the quality and performance control of nuclear fuel. About 2000 neutron radiographs were taken, mainly during the post irradiation examination of light water reactor fuel. During assessment of neutron radiographs some typical defects of the fuel were found and it was felt that a classification of such defects will help to speed up the assessment procedure. Therefore an attempt was made to establish such a classification, which is currently used at Risoe now. This classification is presented in this atlas, which contains 36 neutron radiographs reproduced on film (in original size) and on paper (twice enlarged). (author)

Biomass fuel exposure and respiratory diseases in India.

Science.gov (United States)

Prasad, Rajendra; Singh, Abhijeet; Garg, Rajiv; Giridhar, Giridhar B

2012-10-01

One half of the world's population relies on biomass fuel as the primary source of domestic energy. Biomass fuel exposure causes a high degree of morbidity and mortality in humans. This is especially true in the context of developing countries, which account for 99% of the world's biomass fuel use. Biomass fuel consists of fire wood, dung cakes, agricultural crop residues such as straw, grass, and shrubs, coal fuels and kerosene. Together, they supply 75% of the domestic energy in India. An estimated three-quarters of Indian households use biomass fuel as the primary means for domestic cooking. Ninety percent of rural households and 32% of urban households cook their meals on a biomass stove. There are wide variations between the rural and urban households regarding the specific type of biomass fuel used. Globally, almost 2 million deaths per year are attributable to solid fuel use, with more than 99% of these occurring in developing countries. Biomass fuel accounts for 5-6% of the national burden of disease. Burning biomass fuels emits toxic fumes into the air that consist of small solid particles, carbon monoxide, polyorganic and polyaromatic hydrocarbons, and formaldehyde. Exposure to biomass fuels has been found to be associated with many respiratory diseases such as acute lower respiratory infections, chronic obstructive pulmonary disease, lung cancer, pulmonary tuberculosis, and asthma. Biomass fuel exposure is closely related to the burden of disease in India. Hopes are that future studies will examine the morbidity associated with biomass exposure and seek to prevent it. Concerted efforts to improve stove design and transition to high-efficiency low-emission fuels may reduce respiratory disease associated with biomass fuel exposure.

Blending Octane Number of Toluene with Gasoline-like and PRF Fuels in HCCI Combustion Mode

KAUST Repository

Waqas, Muhammad Umer

2018-04-03

Future internal combustion engines demand higher efficiency but progression towards this is limited by the phenomenon called knock. A possible solution for reaching high efficiency is Octane-on-Demand (OoD), which allows to customize the antiknock quality of a fuel through blending of high-octane fuel with a low octane fuel. Previous studies on Octane-on-Demand highlighted efficiency benefits depending on the combination of low octane fuel with high octane booster. The author recently published works with ethanol and methanol as high-octane fuels. The results of this work showed that the composition and octane number of the low octane fuel is significant for the blending octane number of both ethanol and methanol. This work focuses on toluene as the high octane fuel (RON 120). Aromatics offers anti-knock quality and with high octane number than alcohols, this work will address if toluene can provide higher octane enhancement. Our aim is to investigate the impact of three gasoline-like fuels and two Primary Reference Fuels (PRFs). More specifically, fuels are FACE (Fuels for Advanced Combustion Engines) I, FACE J, FACE A, PRF 70 and PRF 84. A CFR engine was used to conduct the experiments in HCCI mode. For this combustion mode, the engine operated at four specific conditions based on RON and MON conditions. The octane numbers corresponding to four HCCI numbers were obtained for toluene concentration of 0, 2, 5, 10, 15 and 20%. Results show that the blending octane number of toluene varies non-linearly and linearly with the increase in toluene concentration depending on the base fuel, experimental conditions and the concentration of toluene. As a result, the blending octane number can range from close to 150 with a small fraction of toluene to a number closer to that of toluene, 120, with larger fractions.

Advanced compressed hydrogen fuel storage systems

International Nuclear Information System (INIS)

Jeary, B.

2000-01-01

Dynetek was established in 1991 by a group of private investors, and since that time efforts have been focused on designing, improving, manufacturing and marketing advanced compressed fuel storage systems. The primary market for Dynetek fuel systems has been Natural Gas, however as the automotive industry investigates the possibility of using hydrogen as the fuel source solution in Alternative Energy Vehicles, there is a growing demand for hydrogen storage on -board. Dynetek is striving to meet the needs of the industry, by working towards developing a fuel storage system that will be efficient, economical, lightweight and eventually capable of storing enough hydrogen to match the driving range of the current gasoline fueled vehicles

Fuel oil and kerosene sales, 1990

International Nuclear Information System (INIS)

1991-01-01

Sales data is presented for kerosene and fuel oils. This is the second year that the survey data have appeared in a separate publication. Prior to the 1989 report, the statistics appeared in the Petroleum Marketing Annual (PMA) for reference year 1988 and the Petroleum Marketing Monthly (PMM) for reference years 1984 through 1987. 4 figs., 24 tabs

Fuel oil and kerosene sales, 1990

Energy Technology Data Exchange (ETDEWEB)

1991-10-10

Sales data is presented for kerosene and fuel oils. This is the second year that the survey data have appeared in a separate publication. Prior to the 1989 report, the statistics appeared in the Petroleum Marketing Annual (PMA) for reference year 1988 and the Petroleum Marketing Monthly (PMM) for reference years 1984 through 1987. 4 figs., 24 tabs.

Development of a skeletal multi-component fuel reaction mechanism based on decoupling methodology

International Nuclear Information System (INIS)

Mohan, Balaji; Tay, Kun Lin; Yang, Wenming; Chua, Kian Jon

2015-01-01

Highlights: • A compact multi-component skeletal reaction mechanism was developed. • Combined bio-diesel and PRF mechanism was proposed. • The mechanism consists of 68 species and 183 reactions. • Well validated against ignition delay times, flame speed and engine results. - Abstract: A new coupled bio-diesel surrogate and primary reference fuel (PRF) oxidation skeletal mechanism has been developed. The bio-diesel surrogate sub-mechanism consists of oxidation sub-mechanisms of Methyl decanoate (MD), Methyl 9-decenoate (MD9D) and n-Heptane fuel components. The MD and MD9D are chosen to represent the saturated and unsaturated methyl esters respectively in bio-diesel fuels. Then, a reduced iso-Octane oxidation sub-mechanism is added to the bio-diesel surrogate sub-mechanism. Then, all the sub-mechanisms are integrated to a reduced C_2–C_3 mechanism, detailed H_2/CO/C_1 mechanism and reduced NO_x mechanism based on decoupling methodology. The final mechanism consisted of 68 species and 183 reactions. The mechanism was well validated with shock-tube ignition delay times, laminar flame speed and 3D engine simulations.

Fast neutron activation analysis of fossil fuels and liquefaction products

International Nuclear Information System (INIS)

Ehmann, W.D.; Khalil, S.R.; Koppenaal, D.W.

1982-01-01

The problems associated with neutron absorption/thermalization, gamma-ray self-absorption, and variable irradiation and counting geometries associated with the composition, densities and physical states of the samples and standards of fossil fuels are considered. Two sets of liquid organic reagent primary standards and several solid standards are selected and evaluated for use in the determiation of oxygen and nitrogen in coals, coal conversion liquids, and residual solids. Analyses of a number of coals, conversion products and NBS reference standards are presented. Problems associated with selecting a reproducible pre-analysis drying procedure for oxygen determinations in coal and discussed. It is suggested that a brief freeze-drying procedure may result in minimal matrix alternation and yield reproducible values for bulk oxygen contents of coals

In-core fuel management benchmarks for PHWRs

International Nuclear Information System (INIS)

1996-06-01

Under its in-core fuel management activities, the IAEA set up two co-ordinated research programmes (CRPs) on complete in-core fuel management code packages. At a consultant meeting in November 1988 the outline of the CRP on in-core fuel management benchmars for PHWRs was prepared, three benchmarks were specified and the corresponding parameters were defined. At the first research co-ordination meeting in December 1990, seven more benchmarks were specified. The objective of this TECDOC is to provide reference cases for the verification of code packages used for reactor physics and fuel management of PHWRs. 91 refs, figs, tabs

Spent fuel management

International Nuclear Information System (INIS)

2005-01-01

The production of nuclear electricity results in the generation of spent fuel that requires safe, secure and efficient management. Appropriate management of the resulting spent fuel is a key issue for the steady and sustainable growth of nuclear energy. Currently about 10,000 tonnes heavy metal (HM) of spent fuel are unloaded every year from nuclear power reactors worldwide, of which 8,500 t HM need to be stored (after accounting for reprocessed fuel). This is the largest continuous source of civil radioactive material generated, and needs to be managed appropriately. Member States have referred to storage periods of 100 years and even beyond, and as storage quantities and durations extend, new challenges arise in the institutional as well as in the technical area. The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs

Reference Material Properties and Standard Problems to Verify the Fuel Performance Models Ver 1.0

International Nuclear Information System (INIS)

Yang, Yong Sik; Kim, Jae Yong; Koo, Yang Hyun

2010-12-01

All fuel performance models must be validated by in-pile and out-pile tests. However, the model validation requires much efforts and times to confirm its exactness. In many fields, new performance models and codes are confirmed by code-to-code benchmarking process under simplified standard problem analysis. At present, the DUOS, which is the steady state fuel performance analysis code for dual cooled annular fuel, development project is progressing and new FEM module is developed to analyze the fuel performance during transient period. In addition, the verification process is planning to examine the new models and module's rightness by comparing with commercial finite element analysis such as a ADINA, ABAQUS and ANSYS. This reports contains the result of unification of material properties and establishment of standard problem to verify the newly developed models with commercial FEM code

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

International Nuclear Information System (INIS)

Dobrevski, I.; Zaharieva, N.

2008-01-01

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

VVER fuel. Results of post irradiation examination

International Nuclear Information System (INIS)

Smirnov, V.P.; Markov, D.V.; Smirnov, A.V.; Polenok, V.S.; Perepelkin, S.O.; Ivashchenko, A.A.

2005-01-01

The present paper presents the main results of post-irradiation examination of more than 40 different fuel assemblies (FA) operated in the cores of VVER-1000 and VVER-440-type power reactors in a wide range of fuel burnup. The condition of fuel assembly components from the viewpoint of deformation, corrosion resistance and mechanical properties is described here. A serviceability of the FA design as a whole and interaction between individual FA components under vibration condition and mechanical load received primary emphasis. The reasons of FA damage fuel element failure in a wide range of fuel burnup are also analyzed. A possibility and ways of fuel burnup increase have been proved experimentally for the case of high-level serviceability maintenance of fuel elements to provide for advanced fuel cycles. (author)

Design considerations and operating experience with wet storage of Ontario Hydro's irradiated fuel

International Nuclear Information System (INIS)

Frost, C.R.; Naqvi, S.J.; McEachran, R.A.

1987-01-01

The characteristics of Ontario Hydro's fuel and at-reactor irradiated fuel storage water pools (or irradiated fuel bays) are described. There are two types of bay known respectively as primary bays and auxiliary bays, used for at-reactor irradiated fuel storage. Irradiated fuel is discharged remotely from Ontario Hydro's reactors to the primary bays for initial storage and cooling. The auxiliary bays are used to receive and store fuel after its initial cooling in the primary bay, and provide additional storage capacity as needed. The major considerations in irradiated fuel bay design, including site-specific requirements, reliability and quality assurance, are discussed. The monitoring of critical fuel bay components, such as bay liners, the development of high storage density fuel containers, and the use of several irradiated fuel bays at each reactor site have all contributed to the safe handling of the large quantities of irradiated fuel over a period of about 25 years. Routine operation of the irradiated fuel bays and some unusual bay operational events are described. For safety considerations, the irradiated fuel in storage must retain its integrity. Also, as fuel storage is an interim process, likely for 50 years or more, the irradiated fuel should be retrievable for downstream fuel management phases such as reprocessing or disposal. A long-term experimental program is being used to monitor the integrity of irradiated fuel in long-term wet storage. The well characterized fuel, some of which has been in wet storage since 1962 is periodically examined for possible deterioration. The evidence from this program indicates that there will be no significant change in irradiated fuel integrity (and retrievability) over a 50 year wet storage period

Nuclear fuels accounting interface: River Bend experience

International Nuclear Information System (INIS)

Barry, J.E.

1986-01-01

This presentation describes nuclear fuel accounting activities from the perspective of nuclear fuels management and its interfaces. Generally, Nuclear Fuels-River Bend Nuclear Group (RBNG) is involved on a day-by-day basis with nuclear fuel materials accounting in carrying out is procurement, contract administration, processing, and inventory management duties, including those associated with its special nuclear materials (SNM)-isotopics accountability oversight responsibilities as the Central Accountability Office for the River Bend Station. As much as possible, these duties are carried out in an integrated, interdependent manner. From these primary functions devolve Nuclear Fuels interfacing activities with fuel cost and tax accounting. Noting that nuclear fuel tax accounting support is of both an esoteric and intermittent nature, Nuclear Fuels-RBNG support of developments and applications associated with nuclear fuel cost accounting is stressed in this presentation

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

Korean Reference HLW Disposal System

Energy Technology Data Exchange (ETDEWEB)

Choi, Heui Joo; Lee, J. Y.; Kim, S. S. (and others)

2008-03-15

This report outlines the results related to the development of Korean Reference Disposal System for High-level radioactive wastes. The research has been supported around for 10 years through a long-term research plan by MOST. The reference disposal method was selected via the first stage of the research during which the technical guidelines for the geological disposal of HLW were determined too. At the second stage of the research, the conceptual design of the reference disposal system was made. For this purpose the characteristics of the reference spent fuels from PWR and CANDU reactors were specified, and the material and specifications of the canisters were determined in term of structural analysis and manufacturing capability in Korea. Also, the mechanical and chemical characteristics of the domestic Ca-bentonite were analyzed in order to supply the basic design parameters of the buffer. Based on these parameters the thermal and mechanical analysis of the near-field was carried out. Thermal-Hydraulic-Mechanical behavior of the disposal system was analyzed. The reference disposal system was proposed through the second year research. At the final third stage of the research, the Korean Reference disposal System including the engineered barrier, surface facilities, and underground facilities was proposed through the performance analysis of the disposal system.

Reactor core with rod-shaped fuel cells

International Nuclear Information System (INIS)

Dworak, A.

1976-01-01

The proposal refers to the optimization of the power distribution in a reactor core which is provided with several successive rod-shaped fuel cells. A uniform power output - especially in radial direction - is aimed at. This is achieved by variation of the dwelling periods of the fuel cells, which have, for this purpose, a fuel mixture changing from layer to layer. The fuel cells with the shortest dwelling period are arranged near the coolant inlet side of the reactor core. The dwelling periods of the fuel cells are adapted to the given power distribution. As neighboring cells have equal dwelling periods, the exchange can be performed much easier then with the composition currently known. (UWI) [de

Reference radiology in nephroblastoma: accuracy and relevance for preoperative chemotherapy

International Nuclear Information System (INIS)

Schenk, J.P.; Schrader, C.; Zieger, B.; Ley, S.; Troeger, J.; Furtwaengler, R.; Graf, N.; Leuschner, I.

2006-01-01

Purpose: A reference radiologic diagnosis was carried out for the purpose of quality control and in order to achieve high diagnostic accuracy in the ongoing trial and study SIOP 2001/GPOH for renal tumors during childhood. The aim of the present study is to evaluate the value of diagnostic imaging and the benefit of reference evaluation at a pediatric radiology center. Materials and Methods: In 2004 the imaging studies of 97 patients suspected of having a renal tumor were presented at the beginning of therapy. Diagnostic imaging was compared to the primary imaging results and the histological findings and was analyzed in regard to the therapeutic consequence (primary chemotherapy without prior histology). 77 MRI, 35 CT and 67 ultrasound examinations of 47 girls and 50 boys (mean age 4 years; one day to 15.87 years old) were analyzed. In addition to the histological findings, the reference pathological results were submitted in 86 cases. Results from the primary imaging corresponding to the histology and results from the reference radiology corresponding to the histology and results from the reference radiology corresponding to the histology were statistically compared in a binomial test. Results: In 76 of the reference-diagnosed Wilms' tumors, 67 were confirmed histologically. In 72 cases preoperative chemotherapy was initiated. In 5 cases neither a Wilms' tumor nor a nephroblastomatosis was found. 16 of 21 cases (76%) with reference-diagnosed non-Wilms' tumors were selected correctly. The results of the primary imaging corresponded to the histology in 71 cases, and those of the reference radiology in 82 cases. The statistical evaluation showed that the results of the reference radiology were significantly better (p=0.03971). (orig.)

Evaluation of fuel release rate and mechanism tests under RBCB conditions

International Nuclear Information System (INIS)

Adamson, M.G.

1981-09-01

This task includes theoretical evaluation of fuel/fission product release behavior from failed LMFBR fuel elements as well as an on-going experimental investigation of the mechanism of oxide fuel dispersal into flowing liquid sodium. The primary objectives of this work are to develop a fuel source term that can be used in predictive models for primary heat transfer system contamination and to understand the separate influences of important system variables (such as flow rate, oxygen impurity level) on this source term. The present report is written in two parts: the first, in condensed form, is an updated evaluation of fuel (U,Pu) and fission product release data, and the second describes the current status of supporting experimental work at General Electric's Vallecitos Laboratory

Superphenix 1 primary handling system fabrication and testing

International Nuclear Information System (INIS)

Branchu, J.; Ebbinghaus, K.; Gigarel, C.

1985-01-01

Primary handling covers the operations performed for spent fuel removal, new fuel insertion, and the insodium storage outside the new or spent fuel vessel. This equipment typifies many of the difficulties encountered with the project as a whole: fabrication coordination when several countries are involved and design and construction of very large, relatively complex components. Detailed design studies were mainly influenced by thermal and seismic requirements, as applicable to sodium-immersed structures. Where possible, well-tried mechanical solutions were used, but widely differing techniques were involved, ranging from the high precision fabrication of structures and mechanisms comprising numerous component parts, implying complex machining operations. No particular problems were encountered during the sodium testing of the primary handling equipment. Trends for the 1500-MW (electric) breeder include investigation of the advisability of fuel storage in the core lattice and the possibility of handling system simplification

Adrenal Vein Catecholamine Levels and Ratios: Reference Intervals Derived from Patients with Primary Aldosteronism.

Science.gov (United States)

Sze, Candy W C; O'Toole, Samuel Matthew; Tirador, Roger Kent; Akker, Scott A; Matson, Matthew; Perry, Leslie; Druce, Maralyn Rose; Dekkers, Tanja; Deinum, Jaap; Lenders, Jacques W M; Eisenhofer, Graeme; Drake, William Martyn

2017-06-01

Phaeochromocytoma localisation is generally reliably achieved with modern imaging techniques, particularly in sporadic cases. On occasion, however, there can be diagnostic doubt due to the presence of bilateral adrenal abnormalities, particularly in patients with mutations in genes predisposing them to the development of multiple phaeochromocytomas. In such cases, surgical intervention is ideally limited to large or functional lesions due to the long-term consequences associated with hypoadrenalism. Adrenal venous sampling (AVS) for catecholamines has been used in this situation to guide surgery, although there are few data available to support diagnostic thresholds. Retrospective analyses of AVS results from 2 centres were carried out. A total of 172 patients (88 men, 84 women) underwent AVS under cosyntropin stimulation for the diagnosis of established primary aldosteronism (PA) with measurement of adrenal and peripheral venous cortisol, aldosterone and catecholamines. Six patients (3 men, 3 women) with phaeochromocytoma underwent AVS for diagnostic purposes with subsequent histological confirmation. Reference intervals for the adrenal venous norepinephrine to epinephrine ratio were created from the PA group. Using the 97.5th centile (1.21 on the left, 1.04 on the right), the false negative rate in the phaeochromocytoma group was 0%. In conclusion, this study describes the largest dataset of adrenal venous catecholamine measurements and provides reference intervals in patients without phaeochromocytoma. This strengthens the certainty with which conclusions related to adrenal venous sampling for catecholamines can be drawn, acknowledging the procedure is not part of the routine diagnostic workup and is an adjunct for use only in difficult clinical cases. © Georg Thieme Verlag KG Stuttgart · New York.

The effect of water injection on nitric oxide emissions of a gas turbine combustor burning ASTM Jet-A fuel

Science.gov (United States)

Marchionna, N. R.; Diehl, L. A.; Trout, A. M.

1973-01-01

Tests were conducted to determine the effect of water injection on oxides of nitrogen (NOx) emissions of a full annular, ram induction gas turbine combustor burning ASTM Jet-A fuel. The combustor was operated at conditions simulating sea-level takeoff and cruise conditions. Water at ambient temperature was injected into the combustor primary zone at water-fuel ratios up to 2. At an inlet-air temperature of 589 K (600 F) water injection decreased the NOx emission index at a constant exponential rate: NOx = NOx (o) e to the -15 W/F power (where W/F is the water-fuel ratio and NOx(o) indicates the value with no injection). The effect of increasing combustor inlet-air temperature was to decrease the effect of the water injection. Other operating variables such as pressure and reference Mach number did not appear to significantly affect the percent reduction in NOx. Smoke emissions were found to decrease with increasing water injection.

Safety assessment of OPG's used fuel for dry storage

International Nuclear Information System (INIS)

Roman, H.; Khan, A.

2005-01-01

'Full text:' Ontario Power Generation (OPG) operates the Pickering Waste Management Facility (PWMF) and Western Waste Management Facility (WWMF) where OPG has been storing 10-year or older used fuel in the Dry Storage Containers (DSCs) since 1996 and 2003 respectively. The construction licence for the Darlington Used Fuel Dry Storage Facility (DUFDSF) was obtained in August 2004. Safety assessment of the used fuel for dry storage is required to support each request for regulatory approval to construct and operate a dry storage facility. The objective of the safety assessment is to assess the used fuel performance under normal operation and postulated credible accident scenarios. A reference used fuel bundle is defined based on the operating history and data on fuel discharged from the reactors of the specific nuclear generating station. The characteristics of the reference used fuel bundle are used to calculate the nuclide inventory, source term and decay heat used for the assessment. When assessing malfunctions and accidents, postulated external and internal events are considered. Consideration is also given to the design basis accidents of the specific nuclear generating station that could affect the used fuel under dry storage. For those events deemed credible (i.e. probability > 10 -7 ), a bounding fuel failure consequence is predicted. Given the chemical characteristics of the radionuclides in used fuel, the design of the CANDU fuel and the conditions inside the DSC, in the event that a used fuel bundle should become damaged during used fuel dry storage operations, the only significant radionuclides species that are volatile are krypton-85 and tritium. Release of these radionuclides is considered in calculating public and worker doses. (author)