Innovative nuclear fuels: results and strategy

International Nuclear Information System (INIS)

Stan, Marius

2009-01-01

To facilitate the discovery and design of innovative nuclear fuels, multi-scale models and simulations are used to predict irradiation effects on the thermal conductivity, oxygen diffusivity, and thermal expansion of oxide fuels. The multi-scale approach is illustrated using results on ceramic fuels with a focus on predictions of point defect concentrations, stoichiometry, and phase stability. The high performance computer simulations include coupled heat transport, diffusion, and thermal expansion, gas bubble formation and temperature evolution in a fuel element consisting of UO2 fuel and metallic cladding. The second part of the talk is dedicated to a discussion of an international strategy for developing advanced, innovative nuclear fuels. Four initiative are proposed to accelerate the discovery and design of new materials: (a) Develop an international pool of experts, (b) Create Institutes for Materials Discovery and Design, (c) Create an International Knowledge base for experimental data, models (mathematical expressions), and simulations (codes) and (d) Organize international workshops and conference sessions. The paper ends with a discussion of existing and emerging international collaborations.

Nuclear performance of target-blanket assemblies for electronuclear fuel production

International Nuclear Information System (INIS)

Alsmiller, R.G. Jr.; Gabriel, T.A.; Barish, J.

1978-01-01

The results of calculations of high energy transport carried out to evaluate the nuclear performance of several of the designs that have been proposed for electronuclear fuel production are presented. Topics covered include: results for 1-GeV protons and 1-GeV deuterons incident on a lithium target surrounded by a 238 U blanket; results for 1 GeV protons and 1-GeV deuterons incident on a thorium salt; results for 1-GeV protons, incident on a gas-cooled system fueled by either 238 UO 2 or 232 ThO 2 ; and results for protons in the energy range 0.5 to 10 GeV incident on a very large natural uranium target

Impact of indoor air pollution from the use of solid fuels on the incidence of life threatening respiratory illnesses in children in India.

Science.gov (United States)

Upadhyay, Ashish Kumar; Singh, Abhishek; Kumar, Kaushalendra; Singh, Ashish

2015-03-28

India contributes 24% of the global annual child deaths due to acute respiratory infections (ARIs). According to WHO, nearly 50% of the deaths among children due to ARIs is because of indoor air pollution (IAP). There is insufficient evidence on the relationship between IAP from the use of solid fuels and incidence of life threatening respiratory illnesses (LTRI) in children in India. Panel data of children born during 2001-02, from the Young Lives Study (YLS) conducted in India during 2002 and 2006-07 was used to estimate the impact of household use of solid fuels for cooking on LTRI in children. Multivariable two-stage random effects logistic regression model was used to estimate the odds of suffering from LTRI among children from households using solid fuels relative to children from households using other fuels (Gas/Electricity/Kerosene). Bivariate results indicate that the probability of an episode of LTRI was considerably higher among children from households using solid fuels for cooking (18%) than among children from households using other fuels (10%). Moreover, children from households using solid fuels in both the rounds of YLS were more likely to suffer from one or more than one episode of LTRI compared to children from households using solid fuels in only one round. Two-stage random effects logistic regression result shows that children from households using solid fuels were 1.78 (95% CI: 1.05-2.99) times as likely to suffer from LTRI as those from households using other fuels. The findings of this paper provide conclusive evidence on the harmful effects of the use of solid fuels for cooking on LTRI in India. The Government of India must make people aware about the health risks associated with the use of solid fuels for cooking and strive to promote the use of cleaner fuels.

Fuel handling problems at KANUPP

International Nuclear Information System (INIS)

Ahmed, I.; Mazhar Hasan, S.; Mugtadir, A.

1991-01-01

KANUPP experienced two abnormal fuel and fuel handling related problems during the year 1990. One of these had arisen due to development of end plate to end plate coupling between the two bundles at the leading end of the fuel string in channel HO2-S. The incident occurred when attempts were being made to fuel this channel. Due to pulling of sticking bundles into the acceptor fuelling machine (north) magazine, which was not designed to accommodate two bundles, a magazine rotary stop occurred. The forward motion of the charge tube was simultaneously discovered to be restricted. The incident led to stalling of fuelling machine locked on to the channel HO2, necessitating a reactor shut down. Removal of the fuelling machine was accomplished sometime later after draining of the channel. The second incident which made the fuelling of channel KO5-N temporarily inexecutable, occurred during attempts to remove its north end shield plug when this channel came up for fuelling. The incident resulted due to breaking of the lugs of the shield plug, making its withdrawal impossible. The Plant however kept operating with suspended fuelling of channel KO5, until it could no longer sustain a further increase in fuel burnup at the maximum rating position. Resolving both these problems necessitated draining of the respective channels, leaving the resident fuel uncovered for the duration of the associated operation. Due to substantial difference in the oxidation temperatures Of UO 2 and Zircaloy and its influence as such on the cooling requirement, it was necessary either to determine explicitly that the respective channels did not contain defective fuel bundles or wait for time long enough to allow the decay heat to reduce to manageable proportions. This had a significant bearing on the Plant down time necessary for the rectification of the problems. This paper describes the two incidents in detail and dwells upon the measures adopted to resolve the related problems. (author)

Fuel handling problems at KANUPP

Energy Technology Data Exchange (ETDEWEB)

Ahmed, I; Mazhar Hasan, S; Mugtadir, A [Karachi Nuclear Power Plant (KANUPP), Karachi (Pakistan)

1991-04-01

KANUPP experienced two abnormal fuel and fuel handling related problems during the year 1990. One of these had arisen due to development of end plate to end plate coupling between the two bundles at the leading end of the fuel string in channel HO2-S. The incident occurred when attempts were being made to fuel this channel. Due to pulling of sticking bundles into the acceptor fuelling machine (north) magazine, which was not designed to accommodate two bundles, a magazine rotary stop occurred. The forward motion of the charge tube was simultaneously discovered to be restricted. The incident led to stalling of fuelling machine locked on to the channel HO2, necessitating a reactor shut down. Removal of the fuelling machine was accomplished sometime later after draining of the channel. The second incident which made the fuelling of channel KO5-N temporarily inexecutable, occurred during attempts to remove its north end shield plug when this channel came up for fuelling. The incident resulted due to breaking of the lugs of the shield plug, making its withdrawal impossible. The Plant however kept operating with suspended fuelling of channel KO5, until it could no longer sustain a further increase in fuel burnup at the maximum rating position. Resolving both these problems necessitated draining of the respective channels, leaving the resident fuel uncovered for the duration of the associated operation. Due to substantial difference in the oxidation temperatures Of UO{sub 2} and Zircaloy and its influence as such on the cooling requirement, it was necessary either to determine explicitly that the respective channels did not contain defective fuel bundles or wait for time long enough to allow the decay heat to reduce to manageable proportions. This had a significant bearing on the Plant down time necessary for the rectification of the problems. This paper describes the two incidents in detail and dwells upon the measures adopted to resolve the related problems. (author)

Fuel Coolant Interaction Results in the Fuel Pins Melting Facility (PMF)

International Nuclear Information System (INIS)

Urunashi, H.; Hirabayashi, T.; Mizuta, H.

1976-01-01

The experimental work related to FCI at PNC has been concentrated into the molten UO 2 dropping test. After the completion of molten UO 2 drop experiments, emphasis is directed toward the FCI phenomena of the initiating conditions of the accident under the more realistic geometry. The experiments are conducted within the Pin Melt Facility (PMF) in which UO 2 pellets clad in stainless steel are melted by direct electric heating under the stagnant or flowing sodium. The primary objectives of the PMF test are to: - obtain detail experimental results (heat-input, clad temperature, sodium temperature, etc.) on the FCI under TOP and LOF conditions; - observe the movement of the fuel before and after the pin failure by the X-ray cinematography; - observe the degree of coherence of the pin failures; - accumulate the experience of the FCI experiment which is applicable to the subassembly or more larger scale; - simulate the fuel behavior of the in-pile test (GETR, CABRI). The preliminary conclusions can be drawn from the foregoing observations are as follows: - Although the fuel motion and FCI of the closed test section appeared to be different from those of the open test section, the conclusion of the effect of the inside pressure on FCI needs more experimental data. - The best heating condition of the UO 2 pellet for the FCI study with PMF is established as 40 w/cm at the steady state and 1680 J/g of UO 2 during the additional transient state. The total energy deposition of the UO 2 pellet is thus estimated in the range of 2400 J/g of UO 2 -2600 J/g of UO 2 . The analytical model of the fuel pin failure and the subsequent FCI are suggested to count the following parameters: - The fuel pin failure due to the fuel vaporization due to the rapid energy deposition; - Molten fuel, clad and sodium interaction in the fuel pin after the pin failure; - The upward flow of molten fuel with molten clad or vapor sodium, as well as the slumping of molten fuel

Interim results from UO2 fuel oxidation tests in air

International Nuclear Information System (INIS)

Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j.

1987-08-01

An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO 2 , fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO 2 pellets in the temperature range of 135 to 250 0 C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10 5 R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10 5 R/h gamma field. 33 refs., 51 figs., 6 tabs

WWER fuel: Results of post irradiation examination

International Nuclear Information System (INIS)

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

2006-01-01

Kozloduy NPP and advanced working assemblies at the Kola NPP demonstrate this. Now a high-priority task is to prolong the service life and increase a burnup of WWER fuel without decrease of high reliability. Design and engineering innovations in the course of fuel development require a fundamental analysis from the viewpoint of efficiency. Innovations are checked at the final stage based on the results of PIE of FAs after semi-commercial operation in the commercial reactor. PIE make it possible to get a detailed and unprejudiced evaluation of a new generation fuel for consistency with the above stipulated requirements. The present paper describes the results of FA PIE with regard to its main parameters and properties, which affect the serviceability of products and their structural components. The paper has the following contents: Introduction; Post-irradiation examination results; FA geometries and rigidity; State of fuel elements; Fuel- cladding interaction; Cladding corrosion; Fission gas release; Zr-spacer grid state; Fuel failure; Corrective actions; Conclusion. In conclusion the authors underline that the WWER-440 FAs provided with shroud demonstrate a high geometrical stability in the course of their operation up to 6 fuel cycles. The problem of FA bending in the WWER-1000 reactor core has been solved at the expense of FA-A and FA-2 development. These designs are provided with rigid skeleton that demonstrates permissible deformation if they are operated up to 6 fuel cycles. As for other parameters affecting the FE serviceability (geometries, cladding corrosion, mechanical properties, fission gas release, etc), the service life is not exhausted up to a fuel burnup of 70 MWd/kgU. The state of ZSGs is satisfactory on the whole, but there are some cases of higher corrosion demonstrating the necessity of further design and engineering work. E635 products (guide tubes, central tubes and FE claddings) demonstrate geometrical stability and strength properties as opposed to

Fuel integrity project: analysis of light water reactor fuel rods test results

International Nuclear Information System (INIS)

Dallongeville, M.; Werle, J.; McCreesh, G.

2004-01-01

BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

Fuel integrity project: analysis of light water reactor fuel rods test results

Energy Technology Data Exchange (ETDEWEB)

Dallongeville, M.; Werle, J. [COGEMA Logistics (AREVA Group) (France); McCreesh, G. [BNFL Nuclear Sciences and Technology Services (United Kingdom)

2004-07-01

BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

Predicting fuel poverty at a small-area level in England

International Nuclear Information System (INIS)

Fahmy, Eldin; Gordon, David; Patsios, Demi

2011-01-01

This paper describes the development of a series of models for predicting the incidence of fuel poverty in England at a small-area level and examines the adequacy of the modelled results in informing our understanding of the geography of fuel poverty. This paper summarises the development of alternative approaches to model specification based upon different approaches to the treatment of household income. Since 2003 small-area fuel poverty estimates have been widely used to inform affordable warmth policies and local targeting of fuel poverty programs. Whilst improvements in data sources and methods in recent years provide an opportunity to better understand the spatial distribution of fuel poverty, these analyses suggest that our understanding of the incidence and spatial distribution of fuel poverty is highly sensitive to the way in which household incomes are measured. - Highlights: → The proposed models estimate fuel poverty incidence at a small-area level. → This is necessary in order to accurately target local fuel poverty interventions. → Fuel poverty estimates are highly sensitive to differences in income measurement. → Fewer children and more pensioners are fuel poor using EHCS income measures. → More children and fewer pensioners are fuel poor using HBAI income measures.

Results of tests with open fuel in KNK II

International Nuclear Information System (INIS)

Schmitz, G.

1987-03-01

For the operation of Liquid Metal Cooled Fast Breeder Reactors with cladding failures the consequences of increased contamination by fission products and fuel and the possibility of failure propagation to adjacent fuel pins due to fuel swelling have to be envisaged. To clarify some of these problems a KNK II test program involving open fuel was defined with the first experiments of this program being performed between October 1981 and May 1984. After the description of the test equipment and of the test program, the results will be presented on delayed neutron measurements, fission gas measurements and post irradiation examinations. The report will conclude with a discussion of the results [de

Results of Microstructural Examinations of Irradiated LEU U-Mo Fuels

Energy Technology Data Exchange (ETDEWEB)

Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, ID 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organization (Australia)

2009-06-15

Introduction: The RERTR program is responsible for converting research reactors that use high-enriched uranium fuels to ones that use low-enriched uranium fuels [1]. As part of the development of LEU fuels, a variety of irradiation experiments are being conducted using the Advanced Test Reactor. Based on the results of initial fuel plate testing, adjustments have been made to the characteristics of fuel plates to improve the stability of the fuel microstructure. One improvement has been to add Si to the matrix of a dispersion fuel. This material is also being added at the fuel/cladding interface of a monolithic fuel. This paper will discuss the irradiation performance of these fuels, in terms of the stability of their microstructures during irradiation. Results and discussion: The post-irradiation examinations of fuel plates are performed at the Idaho National Laboratory. These examinations consist of visual examinations of fuel plates, gamma scanning, thickness measurements, oxide thickness measurements, and optical metallographic examinations of the fuel plate microstructures. Microstructural analysis is also performed using scanning electron microscopy. Overall, U-7Mo and U-10Mo alloy fuels have displayed the best irradiation performance, particularly, when a Si-containing Al alloy is used as the dispersion fuel matrix. The benefit of using this type of matrix is that the commonly observed fuel/cladding interaction that occurs during irradiation is reduced and the interaction layer that forms exhibit stable behavior during irradiation. Monolithic-type fuels, which consist of a U-Mo foil encased in Al alloy cladding, are also being developed. These types of fuels are also showing promise and will continue to be developed. One challenge with this type of fuel is in trying to maximize the bond strength at the foil/cladding interface. Fuel/cladding interactions can affect the quality of the boding at this interface. Si is being added to improve the characteristics

Loss of spent fuel pool cooling PRA: Model and results

International Nuclear Information System (INIS)

Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

1996-09-01

This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 x 10 -5 and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 x 10 -3 . Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible

SP-100 Fuel Pin Performance: Results from Irradiation Testing

Science.gov (United States)

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

1994-07-01

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

Results of trial operation of the WWER advanced fuel assemblies

International Nuclear Information System (INIS)

Vasilchenko, I.; Dragunov, Y.; Mikhalchuk, A.

2001-01-01

The paper describes results from experimental operation of advanced WWER-1000 fuel assemblies (AFA) at five units in Balakovo NPP. Advanced fuel is developed according to the concept of standard WWER-1000 fuel assembly (jacket-free). The new features includes: 1) zirconium guiding channels (alloy E-635 and E-110) and spacer grids (alloy E-110); 2) integrated burnable absorber gadolinium; 3) extended service life of fuel assemblies (FA) and absorber rods (possibility of repair of FA); 4) improved adoption to reactor conditions. Some results of AFA pilot operation of a three year operation are presented and analyses of effectiveness of improvements are made concerning application of zirconium channels and grids; application of integrated burnable absorbers; extension of FA and absorbing rods service life and FA repairability. These new features of WWER-1000 fuel design allow: 1) to reduce the average fuel enrichment to the 3.77% instead of 4.31% in U-235; 2) to reduce the FA axial load in reactor hot state by 40%,; 3) increasing of fuel operation in reactor to the 30000 effective days with possibility to have a 5-year residence time in the reactor. The design of new generation FA for WWER-440 reactors involves few key changes. Fuel inventory in new fuel design is increased due to elongation of fuel stack and reducing the diameter of the central hole. Vibration stability is enhanced as a result of: no-play junction of the fuel rod with the lower grid; change of SG arrangements; strengthening of the lower grid unit; secure of the central tube in the gap. Water-uranium ration is increased. Introduction of all these kinds of modernization in a 5-year fuel cycle reduces fuel component in the energy cost to the 7%

Incidents Associated with Gas Operation

Directory of Open Access Journals (Sweden)

Szer J.

2017-06-01

Full Text Available This article shows incidents associated with the use of gas as an energy carrier. It presents selected incidents which have occurred in Poland and around the world in recent decades. Based on this, consequences of gas and air mixture explosions were analysed as well. The article presents the main causes of gas incidents which have taken place, as per instances which are similar worldwide. Incidents associated with the use of gas are not frequent, but at the same time very tragic as they often lead to illness or even death. In Poland, in the last twenty years, construction area disasters caused by gas explosions account for only 5% of all which have occurred, but the number of fatalities resulting from these cases is approximately 14%. The number of individuals injured reached 39% of all construction disaster victims. Considering all these facts, it is necessary to undertake wide preventive measures in order to increase safety in the use of gaseous fuels.

Association of PM2.5 with diabetes, asthma, and high blood pressure incidence in Canada: A spatiotemporal analysis of the impacts of the energy generation and fuel sales.

Science.gov (United States)

Requia, Weeberb J; Adams, Matthew D; Koutrakis, Petros

2017-04-15

Numerous studies have reported an association between fine particulate matter (PM 2.5 ) and human health. Often these relationships are influenced by environmental factor that varies spatially and/or temporally. To our knowledge, there are no studies in Canada that have considered energy generation and fuel sales as PM 2.5 effects modifiers. Determining exposure and disease-specific risk factors over space and time is crucial for disease prevention and control. In this study, we evaluated the association of PM 2.5 with diabetes, asthma, and High Blood Pressure (HBP) incidence in Canada. Then we explored the impact of the energy generation and fuel sales on association changes. We fit an age-period-cohort as the study design, and we applied an over-dispersed Poisson regression model to estimate the risk. We conducted a sensitivity analysis to explore the impact of variation in clean energy rates and fuel sales on outcomes changes. The study included 117 health regions in Canada between 2007 and 2014. Our findings showed strong association of PM 2.5 with diabetes, asthma, and HBP incidence. A two-year increase of 10μg/m 3 in PM 2.5 was associated with an increased risk of 5.34% (95% CI: 2.28%; 12.53%) in diabetes incidence, 2.24% (95% CI: 0.93%; 5.38%) in asthma incidence, and 8.29% (95% CI: 3.44%; 19.98%) in HBP incidence. Our sensitivity analysis findings suggest higher risks of diabetes, asthma and HBP incidence when there is low clean energy generation. On the other hand, we found lower risk when we considered high rate of clean energy generation. For example, considering only diabetes incidence, we found that the risk in health regions with low rates of clean electricity is approximately 700% higher than the risk in health regions with high rates of clean electricity. Furthermore, our analysis suggested that the risk in regions with low fuel sales is 66% lower than the risk is health regions with low rates of clean electricity. Our study provides support for

Experimental irradiation of UMo fuel: Pie results and modeling of fuel behaviour

International Nuclear Information System (INIS)

Languille, A.; Plancq, D.; Huet, F.; Guigon, B.; Lemoine, P.; Sacristan, P.; Hofman, G.; Snelgrove, J.; Rest, J.; Hayes, S.; Meyer, M.; Vacelet, H.; Leborgne, E.; Dassel, G.

2002-01-01

Seven full-sized U Mo plates containing ca. 8 g/cm 3 of uranium in the fuel meat have been irradiated since the beginning of the French U Mo development program. The first three of them with 20% 235 U enrichment were irradiated at maximum surfacic power under 150 W/cm 2 in the OSIRIS reactor up to 50% burn-up and are under examination. Their global behaviour is satisfactory: no failure and a low swelling. The other four plates were irradiated in the HFR Petten at maximum surfacic power between 150 and 250 W/cm 2 with two enrichments 20 and 35%. The experiment was stopped after two cycles due to a fuel failure. The post- irradiation examinations were completed in 2001 in Petten. Examinations showed a correct behaviour of 20% enriched plates and an abnormal behaviour of the two other plates (35%-enriched) with a clad failure on the plate 4. The fuel failure appears to result from a combination of factors that led to high corrosion cladding and high fuel meat temperatures. (author)

Introduction of a prehospital critical incident monitoring system--pilot project results.

Science.gov (United States)

Stella, Julian; Davis, Anna; Jennings, Paul; Bartley, Bruce

2008-01-01

Hospital medical incident monitoring improves preventable morbidity and mortality rates. Error management systems have been adopted widely in this setting. Data relating to incident monitoring in the prehospital setting is limited. Implementation of an incident monitoring process in a prehospital setting. This is a prospective, descriptive study of the pilot phase of the implementation of an incident monitoring process in a regional prehospital setting, with a focus on trauma care. Paramedics and emergency department staff submitted anonymous incident reports, and a chart review was performed on patients who met major trauma criteria. Selected trauma cases were analyzed by a structured interview/debriefing process to elucidate undocumented incidents. A project committee coded and logged all incidents and developed recommendations. Of 4,429 ambulance responses, 41 cases were analyzed. Twenty-four (58.5%; 95% CI = 49.7-67.4%) were reported anonymously, and the rest were major trauma patients. A total of 77 incidents were identified (mean per case = 1.8; CI = 1.03-2.57). Anonymous cases revealed 26 incidents (mean = 1.1; CI = 0.98-1.22); eight trauma debriefings revealed 38 incidents (mean = 4.8; CI = 0.91-8.69) and nine trauma chart reviews revealed 13 incidents (mean = 1.6; CI = 1.04-2.16). A total of 56 of 77 (72.7%; CI = 65.5-80.0%) incidents related to system inadequacies, and 15 (57.7%; CI = 46.7-68.6%) anonymously reported incidents related to resource problems. A total of 35 of 77 (45.5%; CI = 40.4-50.5%) incidents had minimal or no impact on the patients' outcomes. Thirty-four of 77 (44.2%; CI = 39.3-49.1%) incidents were considered mitigated by circumstance. Incident monitoring led to generalized feedback in most cases (65 of 77; 84.4%; CI = 77.6-91.3%); in three cases (3.9%; CI = 3.7-4.1%), specific education occurred; two cases were reported to an external body (2.6%; CI = 2.5-2.7%); three cases resulted in remedial action (3.9%; CI = 3.7-4.1%); four for

Spent fuel drying system test results (first dry-run)

International Nuclear Information System (INIS)

Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first dry-run test, which was conducted without a fuel element. The empty test apparatus was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The data from this dry-run test can serve as a baseline for the first two fuel element tests, 1990 (Run 1) and 3128W (Run 2). The purpose of this dry-run was to establish the background levels of hydrogen in the system, and the hydrogen generation and release characteristics attributable to the test system without a fuel element present. This test also serves to establish the background levels of water in the system and the water release characteristics. The system used for the drying test series was the Whole Element Furnace Testing System, described in Section 2.0, which is located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in section 3.0, and the experimental

Modeling fuel succession

Science.gov (United States)

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

2009-01-01

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

Research on consequence analysis method for probabilistic safety assessment of nuclear fuel facilities (4). Investigation of safety evaluation method for fire and explosion incidents

International Nuclear Information System (INIS)

Abe, Hitoshi; Tashiro, Shinsuke; Ueda, Yoshinori

2010-01-01

A special committee on 'Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)' was organized by the Atomic Energy Society of Japan (AESJ) under the entrustment of Japan Atomic Energy Agency (JAEA). The committee aims to research on the state-of-the-art consequence analysis method for Probabilistic Safety Assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objective of this research is to obtain the useful information related to the establishment of quantitative performance objectives and to risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on the analysis method of consequences for postulated accidents with potentially large consequences in NFFs, e.g., events of criticality, spill of molten glass, hydrogen explosion, boiling of radioactive solution, and fire (including rapid decomposition of TBP complexes), resulting in the release of radio active materials into the environment. The results of the research were summarized in a series of six reports, which consist of a review report and five technical ones. In this technical report, the research results about basic experimental data and the method for safety evaluation of fire and explosion incidents were summarized. (author)

PBF severe fuel damage program: results and comparison to analysis

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Results of industrial tests of carbonate additive to fuel oil

Science.gov (United States)

Zvereva, E. R.; Dmitriev, A. V.; Shageev, M. F.; Akhmetvalieva, G. R.

2017-08-01

Fuel oil plays an important role in the energy balance of our country. The quality of fuel oil significantly affects the conditions of its transport, storage, and combustion; release of contaminants to atmosphere; and the operation of main and auxiliary facilities of HPPs. According to the Energy Strategy of Russia for the Period until 2030, the oil-refining ratio gradually increases; as a result, the fraction of straight-run fuel oil in heavy fuel oils consistently decreases, which leads to the worsening of performance characteristics of fuel oil. Consequently, the problem of the increase in the quality of residual fuel oil is quite topical. In this paper, it is suggested to treat fuel oil by additives during its combustion, which would provide the improvement of ecological and economic indicators of oil-fired HPPs. Advantages of this method include simplicity of implementation, low energy and capital expenses, and the possibility to use production waste as additives. In the paper, the results are presented of industrial tests of the combustion of fuel oil with the additive of dewatered carbonate sludge, which is formed during coagulation and lime treatment of environmental waters on HPPs. The design of a volume delivery device is developed for the steady additive input to the boiler air duct. The values are given for the main parameters of the condition of a TGM-84B boiler plant. The mechanism of action of dewatered carbonate sludge on sulfur oxides, which are formed during fuel oil combustion, is considered. Results of industrial tests indicate the decrease in the mass fraction of discharged sulfur oxides by 36.5%. Evaluation of the prevented damage from sulfur oxide discharged into atmospheric air shows that the combustion of the fuel oil of 100 brand using carbonate sludge as an additive (0.1 wt %) saves nearly 6 million rubles a year during environmental actions at the consumption of fuel oil of 138240 t/year.

Connecticut Transit (CTTRANSIT) Fuel Cell Transit Bus: Preliminary Evaluation Results

Energy Technology Data Exchange (ETDEWEB)

Chandler, K.; Eudy, L.

2008-10-01

This report provides preliminary results from a National Renewable Energy Laboratory evaluation of a protoptye fuel cell transit bus operating at Connecticut Transit in Hartford. Included are descriptions of the planned fuel cell bus demonstration and equipment; early results and agency experience are also provided.

Drying results of K-Basin fuel element 1990 (Run 1)

International Nuclear Information System (INIS)

Marschman, S.C.; Abrefah, J.; Klinger, G.S.; Oliver, B.M.; MacFarlan, P.J.; Ritter, G.A.

1998-06-01

The water-filled K-Basins in the Hanford 100-Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0

First qualitative analysis of fuel irradiation results carried out in the MR reactor on WWER-1000 fuel

Energy Technology Data Exchange (ETDEWEB)

Chantoin, P [International Atomic Energy Agency, Vienna (Austria); Dubrovin, K; Platonov, P [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Onufriev, V [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

1994-12-31

Four experiments carried out in the MR reactor are evaluated. They are aimed to assess the influence of burnup and the size of the pellet central hole on the fuel temperature and thus on the fuel swelling and fission gas release. The experiments have been performed at different linear rate and burnup of the fuel rods which are above the actual licensed values in WWER power stations. In this paper the results on WWER fuel rod behaviour are examined. The main fabrication and irradiation characteristics for each experiment are given. The main results from destructive and non-destructive examinations are summarized. They include: burnup determination by gamma spectroscopy, caesium shifting along fuel column and accumulation at the end of the fuel stack, fission gas release. fuel rod diameter and length change and macro-graphs showing the central hole size and the morphology after irradiation. From observation of fuel structure, Cs spectrometry and fission gas release, a large degradation of fuel thermal conductivity can be identified at high burnup. If the fuel burnup is the right parameter to be considered, burnup limits identified are: 0 70-75 MWd/kg for rods with large central hole; (2) 58-64 MWd/kg for rods with small central hole. As a general conclusion it is stressed the importance of the study due to irradiation beyond the usual linear rates at high burnup. Up to now the fuel life limiting factor was cladding corrosion when using Zircaloy-4. As the cladding corrosion situation improves, the next life limiting factor to be met could be the fuel itself. The decreasing fuel thermal conductivity is probably of prime importance and should be further studied and modelled. 5 tabs., 5 figs., 3 refs.

Sensitivity Analysis of FEAST-Metal Fuel Performance Code: Initial Results

International Nuclear Information System (INIS)

Edelmann, Paul Guy; Williams, Brian J.; Unal, Cetin; Yacout, Abdellatif

2012-01-01

This memo documents the completion of the LANL milestone, M3FT-12LA0202041, describing methodologies and initial results using FEAST-Metal. The FEAST-Metal code calculations for this work are being conducted at LANL in support of on-going activities related to sensitivity analysis of fuel performance codes. The objective is to identify important macroscopic parameters of interest to modeling and simulation of metallic fuel performance. This report summarizes our preliminary results for the sensitivity analysis using 6 calibration datasets for metallic fuel developed at ANL for EBR-II experiments. Sensitivity ranking methodology was deployed to narrow down the selected parameters for the current study. There are approximately 84 calibration parameters in the FEAST-Metal code, of which 32 were ultimately used in Phase II of this study. Preliminary results of this sensitivity analysis led to the following ranking of FEAST models for future calibration and improvements: fuel conductivity, fission gas transport/release, fuel creep, and precipitation kinetics. More validation data is needed to validate calibrated parameter distributions for future uncertainty quantification studies with FEAST-Metal. Results of this study also served to point out some code deficiencies and possible errors, and these are being investigated in order to determine root causes and to improve upon the existing code models.

Probability of spent fuel transportation accidents

International Nuclear Information System (INIS)

McClure, J.D.

1981-07-01

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

Results from the characterisation of the Futurix-FTA metal alloy transmutation fuels

International Nuclear Information System (INIS)

Rory Kennedy, J.; O'Holleran, Th.; Keiser, D.

2007-01-01

Full text of publication follows. Idaho National Laboratory has been developing and irradiation testing a number of fuels and fuel types for actinide transmutation as part of the Advanced Fuel Cycle Initiative (AFCI). Fuel types under consideration include both fertile (fast reactor systems) and fertile-free (accelerator-driven systems) metallic alloys. Most recently, fuel fabrication was completed and the fuel pins shipped to the fast flux Phenix reactor in Marcoule, France for irradiation testing as part of the FUTURIX-FTA experiment: an international experiment involving the USA, France, the European Commission and Japan. The metal alloy fuels for this experiment are the low-fertile U-29Pu-4Am-2Np-30Zr and the non-fertile Pu-12Am-40Zr. The fresh fuels have been fully characterised for chemical composition, phase, microstructure, thermal behaviour and fuel-cladding-chemical-interaction (FCCI). Preliminary FCCI results raised some safety concerns with respect to the formation of low melting phases and cladding degradation, which could preclude a fuel from consideration. Results from diffusion couple experiments between the non-fertile fuel Pu-12Am-40Zr and the ferritic HT9 and 422 stainless steels (SS) used in the AFC experiments in the ATR reactor (USA) compared to the austenitic AIM1 SS used in the FUTURIX-FTA experiments in the Phenix reactor (France) indicate significant inter-diffusion with the AIM1 SS. Up to about a 30-fold increase in the diffusion of iron (and accompanying Ni and Cr) into the fuel at 650 C was observed compared to the 422 SS studies. Comparable studies between the low-fertile U-29Pu-4Am-2Np-30Zr fuel alloy and the AIM1 SS show virtually no inter-diffusion. The Fe (along with small amounts of Ni and Cr) appears as small precipitates in the fuel alloy with only minor concentrations identified in the fuel alloy matrix. These results will be discussed in terms of mechanisms of the inter-diffusion and the difference in behaviour between the

Lessons learned from recent safety related incidents at A Canadian uranium conversion facility

International Nuclear Information System (INIS)

Jaferi, Jafir

2013-01-01

This paper presents the Canadian Nuclear Safety Commission's (CNSC) regulatory requirements for nuclear fuel facility licensees to report any situation or incident that results or is likely to result in a hazard to the health or safety of any person or the environment and to submit its incident investigation report with cause(s) of the incident and corrective actions taken or planned. In addition, the paper presents two recent safety-related incidents that occurred at a uranium conversion facility in Canada along with their consequences, causes, corrective actions and any lessons learned. The first incident resulted in a release of uranium hexafluoride (UF6) inside the UF6 cylinder filling station and the second one resulted in a spill of uranium tetrafluoride (UF 4 ) slurry inside the UF6 plant. Both incidents had no impact on the workers or the environment. (authors)

Retinoblastoma incidence patterns in the US Surveillance, Epidemiology, and End Results program.

Science.gov (United States)

Wong, Jeannette R; Tucker, Margaret A; Kleinerman, Ruth A; Devesa, Susan S

2014-04-01

IMPORTANCE Several studies have found no temporal or demographic differences in the incidence of retinoblastoma except for age at diagnosis, whereas other studies have reported variations in incidence by sex and race/ethnicity. OBJECTIVE To examine updated US retinoblastoma incidence patterns by sex, age at diagnosis, laterality, race/ethnicity, and year of diagnosis. DESIGN, SETTING, AND PARTICIPANTS The Surveillance, Epidemiology, and End Results (SEER) databases were examined for retinoblastoma incidence patterns by demographic and tumor characteristics. We studied 721 children in SEER 18 registries, 659 in SEER 13 registries, and 675 in SEER 9 registries. MAIN OUTCOMES AND MEASURES Incidence rates, incidence rate ratios (IRRs), and annual percent changes in rates. RESULTS During 2000-2009 in SEER 18, there was a significant excess of total retinoblastoma among boys compared with girls (IRR, 1.18; 95% CI, 1.02 to 1.36), in contrast to earlier reports of a female predominance. Bilateral retinoblastoma among white Hispanic boys was significantly elevated relative to white non-Hispanic boys (IRR, 1.81; 95% CI, 1.22 to 2.79) and white Hispanic girls (IRR, 1.75; 95% CI, 1.11 to 2.91) because of less rapid decreases in bilateral rates since the 1990s among white Hispanic boys than among the other groups. Retinoblastoma rates among white non-Hispanics decreased significantly since 1992 among those younger than 1 year and since 1998 among those with bilateral disease. CONCLUSIONS AND RELEVANCE Although changes in the availability of prenatal screening practices for retinoblastoma may have contributed to these incidence patterns, further research is necessary to determine their actual effect on the changing incidence of retinoblastoma in the US population. In addition, consistent with other cancers, an excess of retinoblastoma diagnosed in boys suggests a potential effect of sex on cancer origin.

Autoignition of liquid-fuel sprays

International Nuclear Information System (INIS)

Mitzutani, Y.

1991-01-01

This paper reports on the published autoignition data of liquid fuel sprays that were extensively reviewed by classifying them into the following three categories; liquid fuels injected into a stagnant hot atmosphere, liquid fuels injected into a hot air stream (vitiated or unvitiated), and droplet cluster ignited behind an incident or reflected shock. Comparison of these data with the counterparts of gaseous fuels and single droplets revealed that it was the ignition process dominated by droplet evaporation whereas it was the one dominated by chemical kinetics. It consisted, depending on the experimental condition, of the data and of the ignition process dominated by the shattering of droplets by an incident shock. In addition, theoretical works on spray autoignition were reviewed, pointing out that they were still far from universally predicting the ignition delays of liquid fuel sprays

Thermal and in-pile densification of MOX fuels: Some recent results

International Nuclear Information System (INIS)

Caillot, L.; Malgouyres, P.P.; Souchon, F.; Gotta, M.J.; Warin, D.; Chotard, A.; Couty, J.C.

1997-01-01

In-pile densification of PWR fuels is one of the main phenomena which determine the evolution of the pellet-clad gap during the first stage of the irradiation, and thus has consequences onto the thermo-mechanical behaviours of fuel rods. It can be predicted using the results of resintering tests and appropriate correlations. In this context, CEA, FRAMATOME and EDF have undertaken a joint research programme aiming to characterize the densification of MOX fuels. Different fuels were prepared by the MIMAS process using different UO 2 powders as matrix. After a detailed characterization, fuel pellets were submitted to isothermal resintering tests and analytical irradiations. Correlations between in-pile and thermal densification were established. This paper presents the results obtained with two types of MOX fuel: one fabricated wit the AUC UO 2 powder (ammonium uranyl carbonate conversion process) and another one fabricated with the SFEROX powder (peroxide conversion process). 8 refs, 8 figs

Spent fuel drying system test results (second dry-run)

International Nuclear Information System (INIS)

Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks have been detected in the basins and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the second dry-run test, which was conducted without a fuel element. With the concurrence of project management, the test protocol for this run, and subsequent drying test runs, was modified. These modifications were made to allow for improved data correlation with drying procedures proposed under the IPS. Details of these modifications are discussed in Section 3.0

Simulation of the Paks-2 incident. The CODEX-CT-1 experiment

International Nuclear Information System (INIS)

Windberg, P.; Hozer, Z.; Nagy, I.; Vimi, A.

2006-01-01

The Paks-2 cleaning tank incident was simulated with an electrically heated fuel bundle in the CODEX facility. The test conditions included seven hours of oxidation in hydrogen rich steam and final water quenching of the brittle fuel rods. The final state of the bundle showed similar picture that was observed after the incident at the power plant in 2003. (author)

Fire and explosion incident at bituminization demonstration facility of PNC Tokai works, on march 11, 1997

International Nuclear Information System (INIS)

Miura, A.; Sato, Y.; Koyama, T.; Omori, E.; Kato, Y.; Suzuki, H.; Norjiri, I.; Yamanouchi, T.

2001-01-01

On March 11, a fire and explosion incident occurred at the Bituminization Demonstration Facility (BDF) of Tokai Reprocessing Plant in Power Reactor and Nuclear Fuel Development Corporation (PNC). Soon after the incident, PNC (now reorganized to JNC) started to investigate the facility damage, operational records around the incident, technical notes including facility design and reviews of R and D results, operators witness and to perform several analysis, tests and calculations. This paper describes outline and cause of the incident which were concluded based on the results of continuous serious investigation, analysis and calculation. (author)

PND fuel handling decontamination program: specialized techniques and results

International Nuclear Information System (INIS)

Pan, R.; Hobbs, K.; Minnis, M.; Graham, K.

1995-01-01

The use of various decontamination techniques and equipment has become a critical part of Fuel Handling maintenance work at the Pickering Nuclear Station, an eight unit CANDU station located about 30 km east of Toronto. This paper presents an overview of the set up and techniques used for cleaning in the PND Fuel Handling Maintenance Facility, and the results achieved. (author)

Drying Results of K-Basin Fuel Element 6603M (Rune 5)

International Nuclear Information System (INIS)

Oliver, B.M.; Ritter, G.A.; Klinger, G.S.; Abrefah, J.; Greenwood, L.R.; MacFarlan, P.J.; Marschman, S.C.

1999-01-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium spent nuclear fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fifth of those tests conducted on an N-Reactor outer fuel element (6603M) which had been stored underwater in the Hanford 100 Area K-West basin from 1983 until 1996. This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments which were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0. The test conditions and methodologies are given in Section 3.0. Inspections on the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0. Discussion of the results is given in Section 6.0

Drying Results of K-Basin Fuel Element 6603M (Rune 5)

Energy Technology Data Exchange (ETDEWEB)

B.M. Oliver; G.A. Ritter; G.S. Klinger; J. Abrefah; L.R. Greenwood; P.J. MacFarlan; S.C. Marschman

1999-09-24

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium spent nuclear fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fifth of those tests conducted on an N-Reactor outer fuel element (6603M) which had been stored underwater in the Hanford 100 Area K-West basin from 1983 until 1996. This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments which were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0. The test conditions and methodologies are given in Section 3.0. Inspections on the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0. Discussion of the results is given in Section 6.0.

Results of modeling advanced BWR fuel designs using CASMO-4

International Nuclear Information System (INIS)

Knott, D.; Edenius, M.

1996-01-01

Advanced BWR fuel designs from General Electric, Siemens and ABB-Atom have been analyzed using CASMO-4 and compared against fission rate distributions and control rod worths from MCNP. Included in the analysis were fuel storage rack configurations and proposed mixed oxide (MOX) designs. Results are also presented from several cycles of SIMULATE-3 core follow analysis, using nodal data generated by CASMO-4, for cycles in transition from 8x8 designs to advanced fuel designs. (author)

Results of the implementation of a learning system with incidents in an radiotherapy department

International Nuclear Information System (INIS)

Radicchi, Lucas Augusto; Vilela, Ellen Pedroso Severino; Faustino, Fabio de Lima C.; Rodrigues, Fernanda Arantes C.; Gomes, Franciele N.; Souza, Guilherme Vicente de; Silva, Rose Marta S.; Toledo, Jose Carlos de

2016-01-01

An incident learning system (ILS) is an important tool for improving aspects of patient and staff safety. In radiation oncology, ILS has been implemented both at the institutional level as at the national level, allowing to share lessons learned from incidents that have already occurred. The objective of this study is to present the preliminary results of the ILS implemented in a radiation oncology department. In total, 128 incidents were reviewed by a multidisciplinary committee, and the professional groups that registered more were medical physicists, radiation oncologists and radiation therapists. In addition, incidents have occurred and have been detected mainly in the treatment step. The incident learning system proved to be an important process improvement tool, according to the results shown,the improvement actions proposed and the perception of the people involved. (author)

Test results for fuel cell operation on anaerobic digester gas

Science.gov (United States)

Spiegel, R. J.; Preston, J. L.

EPA, in conjunction with ONSI, embarked on a project to define, design, test, and assess a fuel cell energy recovery system for application at anaerobic digester waste water (sewage) treatment plants. Anaerobic digester gas (ADG) is produced at these plants during the process of treating sewage anaerobically to reduce solids. ADG is primarily comprised of methane (57-66%), carbon dioxide (33-39%), nitrogen (1-10%), and a small amount of oxygen (sulfur-bearing compounds (principally hydrogen sulfide) and halogen compounds (chlorides). The project has addressed two major issues: development of a cleanup system to remove fuel cell contaminants from the gas and testing/assessing of a modified ONSI PC25 C fuel cell power plant operating on the cleaned, but dilute, ADG. Results to date demonstrate that the ADG fuel cell power plant can, depending on the energy content of the gas, produce electrical output levels close to full power (200 kW) with measured air emissions comparable to those obtained by a natural gas fuel cell. The cleanup system results show that the hydrogen sulfide levels are reduced to below 10 ppbv and halides to approximately 30 ppbv.

Fuel conditioning facility electrorefiner start-up results

International Nuclear Information System (INIS)

Goff, K.M.; Mariani, R.D.; Vaden, D.; Bonomo, N.L.; Cunningham, S.S.

1996-01-01

At ANL-West, there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Conditioning Facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. The treatment operations will make use of an electrometallurgical process employing molten salts and liquid metals. The treatment equipment is presently undergoing testing with depleted uranium. Operations with irradiated fuel will commence when the environmental evaluation for FCF is complete

Lung function and incidence of chronic obstructive pulmonary disease after improved cooking fuels and kitchen ventilation: a 9-year prospective cohort study.

Directory of Open Access Journals (Sweden)

Yumin Zhou

2014-03-01

Full Text Available BACKGROUND: Biomass smoke is associated with the risk of chronic obstructive pulmonary disease (COPD, but few studies have elaborated approaches to reduce the risk of COPD from biomass burning. The purpose of this study was to determine whether improved cooking fuels and ventilation have effects on pulmonary function and the incidence of COPD. METHODS AND FINDINGS: A 9-y prospective cohort study was conducted among 996 eligible participants aged at least 40 y from November 1, 2002, through November 30, 2011, in 12 villages in southern China. Interventions were implemented starting in 2002 to improve kitchen ventilation (by providing support and instruction for improving biomass stoves or installing exhaust fans and to promote the use of clean fuels (i.e., biogas instead of biomass for cooking (by providing support and instruction for installing household biogas digesters; questionnaire interviews and spirometry tests were performed in 2005, 2008, and 2011. That the interventions improved air quality was confirmed via measurements of indoor air pollutants (i.e., SO₂, CO, CO₂, NO₂, and particulate matter with an aerodynamic diameter of 10 µm or less in a randomly selected subset of the participants' homes. Annual declines in lung function and COPD incidence were compared between those who took up one, both, or neither of the interventions. Use of clean fuels and improved ventilation were associated with a reduced decline in forced expiratory volume in 1 s (FEV₁: decline in FEV₁ was reduced by 12 ml/y (95% CI, 4 to 20 ml/y and 13 ml/y (95% CI, 4 to 23 ml/y in those who used clean fuels and improved ventilation, respectively, compared to those who took up neither intervention, after adjustment for confounders. The combined improvements of use of clean fuels and improved ventilation had the greatest favorable effects on the decline in FEV₁, with a slowing of 16 ml/y (95% CI, 9 to 23 ml/y. The longer the duration of improved fuel use and

Test Results of PBMR Fuel Spheres

International Nuclear Information System (INIS)

Koshcheev, Konstantin; Diakov, Alexander; Beltyukov, Igor; Barybin, Andrey; Chernetsov, Mikhail

2014-01-01

Results of pre-irradiation testing of fuel spheres (FS) and coated particles (CP) manufactured by PBMR SOC (Republic of South Africa) are described. The stable high quality level of major characteristics (dimensions, CP coating structure, uranium-235 contamination of the FS matrix graphite and the outer PyC layer of the CP coating) are shown. Results of a methodical irradiation test of two FS in helium and neon medium at temperatures of 800 to 1300 °C with simultaneous determination of release-to-birth ratios for major gaseous fission products (GFP) are described. (author)

Preliminary test results for post irradiation examination on the HTTR fuel

International Nuclear Information System (INIS)

Ueta, Shohei; Umeda, Masayuki; Sawa, Kazuhiro; Sozawa, Shizuo; Shimizu, Michio; Ishigaki, Yoshinobu; Obata, Hiroyuki

2007-01-01

The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core. (author)

Westinghouse accident tolerant fuel program. Current results and future plans

Energy Technology Data Exchange (ETDEWEB)

Ray, Sumit; Xu, Peng; Lahoda, Edward; Hallstadius, Lars; Boylan, Frank [Westinghouse Electric Company LLC, Hopkins, SC (United States)

2016-07-15

This paper discusses the current status, results from initial tests, as well as the future direction of the Westinghouse's Accident Tolerant Fuel (ATF) program. The current preliminary testing is addressed that is being performed on these samples at the Massachusetts Institute of Technology (MIT) test reactor, initial results from these tests, as well as the technical learning from these test results. In the Westinghouse ATF approach, higher density pellets play a significant role in the development of an integrated fuel system.

Inspecting method for fuel rods

International Nuclear Information System (INIS)

Watanabe, Masaaki; Kogure, Sumio.

1976-01-01

Purpose: To precisely detect the response of flaw in clad tube and submerged fuel pellets from a relationship between the surface of fuel rod and internal signal. Constitution: Ultrasonic reflected waves from the surface of fuel rods and the interior are detected and either one of fuel rod or ultrasonic flaw detecting contact is rotated to thereby precisely detect the response of the flaw of clad tube and submerged fuel pellets from a relationship between said surface and the interior. It will be noted that the ultrasonic flaw detecting contact used is of the line-focus type, the incident angle of ultrasonic wave from the ultrasonic flaw detecting contact relative to the fuel rod is the angle of skew, that is, the ultrasonic flaw detecting contact is not perpendicular to a center axis of the fuel rod but is slightly displace. That is, the use of the aforesaid contact may facilitate discrimination between the surface flaw of the fuel rod and the response of submergence, and in addition, the employment of the aforesaid incident angle makes it hard to receive reflected waves from the surface of the fuel rod which is great in terms of energy to facilitate discrimination of waves responsive to submergence. (Kawakami, Y.)

Supporting staff recovery and reintegration after a critical incident resulting in infant death.

Science.gov (United States)

Roesler, Roberta; Ward, Debra; Short, Mary

2009-08-01

A critical incident is described as any sudden unexpected event that has the power to overwhelm the usual effective coping skills of an individual or a group and can cause significant psychological distress in usually healthy persons. A Just Culture model to deal with critical incidents is an approach that seeks to identify and balance system events and personal accountability. This article reports a critical incident that occurred at the Neonatal Intensive Care Unit, Methodist Hospital of Indianapolis, when 5 infants received an overdose of heparin that resulted in the death of 3 infants. Although care of the family after the critical incident was the immediate priority, the focus of this article was on the recovery and reintegration of the NICU staff after a critical incident based on the Just Culture philosophy.

Operational results of WWER fuel fabricated by MSZ (Elektrostal, Russia)

International Nuclear Information System (INIS)

Asatiani, I.; Balabanov, S.; Beglov, A.; Khryashchev, D.

2009-01-01

The presentation brings forth a statistical analysis of the WWER fuel manufactured by OAO MSZ, operational experience. A necessity of such an analysis is determined by the fact that objective operational results prove the appropriateness of the solutions and decisions made by vendor, designer, manufacturer and utility, as well as motivates further fuel improvements. (authors)

Drying results of K-Basin fuel element 5744U (Run 4)

International Nuclear Information System (INIS)

Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fourth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 5744U. This element (referred to as Element 5744U) was stored underwater in the K-West Basin from 1983 until 1996. Element 5744U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

Drying Results of K-Basin Fuel Element 2660M (Run 7)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1999-01-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the seventh of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 2660M. This element (referred to as Element 2660M) was stored underwater in the K-West Basin from 1983 until 1996. Element 2660M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

Drying Results of K-Basin Fuel Element 6513U (Run 8)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1999-01-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL)on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the eighth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 6513U. This element (referred to as Element 6513U) was stored underwater in the K-West Basin from 1983 until 1996. Element 6513U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0 and discussed in Section 6.0

Drying results of K-Basin fuel element 1164M (run 6)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-08-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the sixth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 1164 M. This element (referred to as Element 1164M) was stored underwater in the K-West Basin from 1983 until 1996. Element 1164M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

Main examination results of WWER-1000 fuel after its irradiation in power reactors

Energy Technology Data Exchange (ETDEWEB)

Bibiliashvili, Yu [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation); Dubrovin, K [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Vasilchenko, I [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation); Yenin, A; Kushmanov, A [AO Novosibirskij Zavod Khimcontsentratov, Novosibirsk (Russian Federation); Smirnov, A; Smirnov, V [Nauchno-Issledovatel` skij Inst. Atomnykh Reaktorov, Dimitrovgrad (Russian Federation)

1994-12-31

WWER-1000 fuel examination has been undertaken to specify the properties of fuel assembly members by defining the parameters of their materials and their interconnection in power reactor operation conditions. Nine fuel assemblies are examined. The examination program includes: visual inspection, measurement of overall dimensions, eddy-current test, gamma-scanning, X-ray and neutron radiography, analysis of gas pressure and composition inside fuel rods, ceramography/metallography, mass spectrometry, microanalysis and electron microscopy of fuel and fuel claddings. The examination results suggest that WWER-1000 fuel spent at steady-state operation conditions up to 50 Mwd/kg U of burnup is in satisfactory condition. The examination of all types of fuel cladding failures indicates that the reason lies in the interaction of cladding with coolant solid impurities. The nodular cladding corrosion of fuel assembly discharged from the South-Ukrainian NPP is caused by the graphite compounds deposited on the fuel rod. Those deposits are a result of the circulating pump damage and had accidental, non-typical character. Some of the rods were found to have a small cladding `fretting` of the spacer grid cell material. The values of the majority of parameters determining the fuel efficiency allow to assume that there is a potential for further extension of fuel burnup and operation length. 1 tab., 11 figs.

Main examination results of WWER-1000 fuel after its irradiation in power reactors

International Nuclear Information System (INIS)

Bibiliashvili, Yu.; Dubrovin, K.; Vasilchenko, I.; Yenin, A.; Kushmanov, A.; Smirnov, A.; Smirnov, V.

1994-01-01

WWER-1000 fuel examination has been undertaken to specify the properties of fuel assembly members by defining the parameters of their materials and their interconnection in power reactor operation conditions. Nine fuel assemblies are examined. The examination program includes: visual inspection, measurement of overall dimensions, eddy-current test, gamma-scanning, X-ray and neutron radiography, analysis of gas pressure and composition inside fuel rods, ceramography/metallography, mass spectrometry, microanalysis and electron microscopy of fuel and fuel claddings. The examination results suggest that WWER-1000 fuel spent at steady-state operation conditions up to 50 Mwd/kg U of burnup is in satisfactory condition. The examination of all types of fuel cladding failures indicates that the reason lies in the interaction of cladding with coolant solid impurities. The nodular cladding corrosion of fuel assembly discharged from the South-Ukrainian NPP is caused by the graphite compounds deposited on the fuel rod. Those deposits are a result of the circulating pump damage and had accidental, non-typical character. Some of the rods were found to have a small cladding 'fretting' of the spacer grid cell material. The values of the majority of parameters determining the fuel efficiency allow to assume that there is a potential for further extension of fuel burnup and operation length. 1 tab., 11 figs

MOX fuel irradiation behaviour: Results from X-ray microbeam analysis

International Nuclear Information System (INIS)

Walker, C.T.; Goll, W.; Matsumura, T.

1997-01-01

The behaviour of plutonium, xenon and caesium were investigated in two sections of irradiated MOX fuel produced by the OCOM process. In one fuel (OCOM30), the MOX agglomerates contained 18 wt% fissile plutonium, and had a low volume fraction of 0.17; in the other (OCOM15) the agglomerates contained 9 wt% fissile plutonium, and had a high volume fraction of 0.34. Both fuels had been irradiated under normal power reactor conditions to a burn-up of approximately 44 GWd/t. The main aim of the work was to establish whether the above differences in composition affected the percentage fission gas released by the fuels. Since U/Pu interdiffusion did not occurred during the irradiation, both fuels remained inhomogeneous on the microscopic scale. However, the concentration of plutonium in the MOX agglomerates decreases by about 50% as a result of fission, whereas the plutonium content of the UO 2 matrix increased by about a factor of four to approximately 2 wt% due to neutron capture by 238 U. The agglomerates in the OCOM15 fuel generally exhibited a finer structure due to the lower burn-up. More than 80% of the fission gas had been released from the oxide lattice of the MOX agglomerates in both fuels. However, a very high fraction of this gas precipitated and remained in the pore structure of the agglomerates. Consequently, puncturing revealed that for both fuels the percentage of gas released to the rod free volume increased from less than 0.5% at 10 GWd/t to a maximum of 3.5% at 45 GWd/t. The conclusion is that the percentage of gas released by MOX fuel is largely unaffected of the level of inhomogeneity of the fuel. In both fuels caesium showed near complete retention in both the MOX agglomerates and the UO 2 matrix. (author). 8 refs, 11 figs, 3 tabs

Some results on development, irradiation and post-irradiation examinations of fuels for fast reactor-actinide burner (MOX and inert matrix fuel)

International Nuclear Information System (INIS)

Poplavsky, V.; Zabudko, L.; Moseev, L.; Rogozkin, B.; Kurina, I.

1996-01-01

Studies performed have shown principal feasibility of the BN-600 and BN-800 cores to achieve high efficiency of Pu burning when MOX fuel with Pu content up to 45% is used. Valuable experience on irradiation behaviour of oxide fuel with high Pu content (100%) was gained as a result of operation of two BR-10 core loadings where the maximum burnup 14 at.% was reached. Post-irradiation examination (PIE) allowed to reveal some specific features of the fuel with high plutonium content. Principal irradiation and PIE results are presented in the paper. Use of new fuel without U-238 provides the maximum burning capability as in this case the conversion ratio is reduced to zero. Technological investigations of inert matrix fuels have been continued now. Zirconium carbide, zirconium nitride, magnesium oxide and other matrix materials are under consideration. Inert matrices selection criteria are discussed in the paper. Results of technological study, of irradiation in the BOR-60 reactor and PIE results of some inert matrix fuels are summarized in this report. (author). 2 refs, 1 fig., 3 tabs

Operation results of 3-rd generation nuclear fuel WWER-440 in initial period

International Nuclear Information System (INIS)

Adeev, V.; Panov, A.

2011-01-01

On unit 4 of Kola NPP trial operation of 3-rd generation's fuel began in 2010. Fuel assemblies of 3-rd generation (FA-3) have a number of design features that provide better operational characteristics. Concise description of a design and the basic advantages of fuel of 3-rd generation are described in articles. Increasing of efficiency of nuclear fuel usage will be achieved by reduction of the parasitic capture of thermal neutrons in constructional materials (weight of zirconium is reduced), optimization of uranium-water relation (increase in fuel elements step), increasing of uranium loading (usage of fuel pellets with increased diameter and without central hole in them). By results of trial operation mass transition to use of given type of assemblies in WWER-440 is possible. This report presents the basic outcomes of the trial operation, a brief survey of the obtained data. The basic characteristics of the reactor core with fuel of 3-rd generation are resulted in work. (authors)

Abstract of results of safety study. Nuclear fuel cycle field in fiscal 2003

International Nuclear Information System (INIS)

2004-11-01

This report descried the results of studies of nuclear fuel cycle field (nuclear fuel facilities, seismic design, all subjects of environmental radiation and waste disposal, and subjects on nuclear fuel cycle in probabilistic safety assessment) in fiscal 2003 on the basis of the principle project of safety study (from fiscal 2001 to 2005). It consists of four chapters; the first chapter is outline of the principle of project, the second is objects and subjects of safety study in the nuclear fuel cycle field, the third list of questionnaire of results of safety study and the forth investigation of results of safety study in fiscal 2003. There are 49 lists, which include 22 reports on the nuclear fuel facility, one on the seismic design, 4 on the probabilistic safety assessment, 7 on the environmental radiation and 15 on the waste disposal. (S.Y.)

Bio-fuels: results in progress, necessary adaptations

International Nuclear Information System (INIS)

2016-02-01

In this report, the French Court of Auditors examines whether its assessments and recommendations published some years before about the development and use of bio-fuels in France had been taken into account. It shows that the support to bio-fuels has multiple objectives, produced some interesting results, but at high cost. Production processes are described. The authors outline that instruments had not been always coherently implemented, and that the tax system had negative effects. They notice that the objective in terms of bio-diesel share has been reached, whereas that of bio-ethanol has not. They also outline that these results have also been obtained with the help of some palliative measures, and that cost remains high for the consumer. In a second part, the report outlines that the present context calls for adaptations, notably due to its uncertainty (instability of European rules, lack of European ambition, a less promising market, a lower priority for automotive manufacturers), and proposes some perspectives and approaches of adaptation, notably to reach quantitative objectives with a greater transparency for the consumer. The report also contains answers made by the different concerned ministers

Database for the OECD-IAEA Paks Fuel Project

International Nuclear Information System (INIS)

Szabo, Emese; Hozer, Zoltan; Gyori, Csaba; Hegyi, Gyoergy

2010-01-01

really necessary for the analytical work. The present version of the database can be extended according to the requests of project participants and considering the availability of requested data. The present version of the database was collected by the experts of AEKI in close cooperation with Paks NPP. The work was supported by the International Atomic Energy Agency (IAEA) and the Hungarian Atomic Energy Authority (HAEA). The database has been prepared to support three types of calculations: - Thermal-hydraulic calculations to describe how the inadequate cooling conditions could have established during the incident. - Fuel behaviour simulation to describe the oxidation and degradation mechanisms of fuel assemblies. - Activity release and transport calculations to simulate the release of fission products from the failed fuel rods. The database includes the following main parts: - Design characteristics of VVER-440 fuel assemblies (main geometrical data, some mechanical properties, oxidation kinetics of Zr1%Nb cladding, and integral data of assemblies). - Operational data of damaged fuel assemblies (power histories of fuel assemblies, burnup, fuel rod internal pressure, isotope inventories, decay heat and axial power distribution). - Design characteristic of the cleaning tank (main geometrical data). - Measured data during the incident: (temperature, water level measurements, cleaning tank outlet flowrates). - Activity measurements (measured coolant activity concentrations, activity release through the chimney, flowrate of water make-up system, released activities). - Reports (describing results of investigations, chronology). - Status of fuel (results of visuals observations). The database items were collected from different sources. Some of them were calculated by Paks NPP and AEKI using VVER-440 and Paks specific input data. The details of the present version of the database including the main information on calculational work will be described in the following

Spent fuel heatup following loss of water during storage

International Nuclear Information System (INIS)

Benjamin, A.S.; McCloskey, D.J.

1978-01-01

Spent fuel assemblies from light water reactors are typically stored for one year or more in the reactor spent fuel pool and then transported for long-term storage at an off-site location. Because of the design, construction, and operation features of spent fuel storage pools, an accident that might drain most of the water from a pool is assessed as being extremely improbable. As a limiting case, however, a hypothetical incident involving instantaneous draining of all the water from a storage pool has been postulated, and the subsequent heatup of the spent fuel elements has been evaluated. The model is analyzed, and results are summarized

Analysis of fuel-handling incidents (safety analysis detailed report no. 5). PEC Brasimone reactor design basis accidents

Energy Technology Data Exchange (ETDEWEB)

1988-01-15

The features covered by this report deal with the equipment and cells in which the handling, examination, measurement, conditioning and storage of core elements are carried out. The operations covered range from the receiving of new element shipments to their insertion in the vessel (excluding handling inside the vessel itself, which is covered in report no. 2) and removal of the spent-elements from the vessel, transfer to their final storage and their ultimate loading into containers for transport outside the plant. The incident analysis along the path of the spent fuel was conducted with the same method adopted for other plant systems. It is treated separately here because the operation of the handling system is practically autonomous from reactor operation.

Radioactive material (RAM) accident/incident data analysis program

International Nuclear Information System (INIS)

Emerson, E.L.; McClure, J.D.

1985-03-01

This report describes the development of the Radioactive Material Transportation Accident/Incident Data Base (RAM-AIDB), which contains information on the occurrences of transportation accidents and incidents, for radioactive materials (RAM) that are involved in the process of transportation, loading and unloading operation, or temporary storage. These transportation operations are in support of the nuclear fuel cycle for electrical energy generation. This study analyzes in some detail basic accident/incident statistical data, RAM packaging accident response data, and the health effects associated with RAM transport accidents/incidents. This report presents a summary of US RAM transport accident/incident experience for the period 1971 through December 1981. In addition, a sample annual summary of accident/incident experience is presented for the calendar year 1981

Zero Emission Bay Area (ZEBA) Fuel Cell Bus Demonstration Results. Fourth Report

Energy Technology Data Exchange (ETDEWEB)

Eudy, Leslie [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Post, Matthew [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

2015-07-02

This report presents results of a demonstration of fuel cell electric buses (FCEB) operating in Oakland, California. Alameda-Contra Costa Transit District (AC Transit) leads the Zero Emission Bay Area (ZEBA) demonstration, which includes 12 advanced-design fuel cell buses and two hydrogen fueling stations. The FCEBs in service at AC Transit are 40-foot, low-floor buses built by Van Hool with a hybrid electric propulsion system that includes a US Hybrid fuel cell power system and EnerDel lithium-based energy storage system. The buses began revenue service in May 2010.

Data base and postirradiation examination results of spent WWER-1000 fuel elements and assemblies

International Nuclear Information System (INIS)

Kanashov, B.A.; Polenok, V.S.; Smirnov, A.V.; Zhitelev, V.A.

1995-01-01

The report presents the results of the postirradiation shape change examination of standard fuel elements and fuel assemblies irradiated in standard conditions in Russian power reactors of the WWER-1000 type. The information is based on the results obtained at the Fuel Research Department of the Federal Scientific Centre Research Institute of Atomic Reactors (FSC RIAR, Dimitrovgrad, Russian Federation) within the period from 1987 to 1994. Emphasis is placed on such experimental and calculational data as: length, cross-section dimensions and shape of FAs with wrapper; change of standard FA skeleton members dimensions; fuel bundle elongation; change of the fuel cladding outer diameter; and elongation and change of the fuel stack outer diameter. (author)

Energy Harvesting From River Sediment Using a Microbial Fuel Cell: Preliminary Results

Directory of Open Access Journals (Sweden)

Philippe Namour

2014-05-01

Full Text Available We have built a sedimentary fuel cell or Sediment Microbial Fuel Cell (SMFC. The device works on the principle of microbial fuel cells by exploiting directly the energy contained in sedimentary organic matter. It converts in electricity the sediment potential, thanks to microorganisms able to waste electrons from their metabolism directly to a solid anode instead of their natural electron acceptors, such as oxygen or nitrate. The sediment microbial fuel cell was made of a non-corrodible anode (graphite buried in anoxic sediments layer and connected via an electrical circuit to a cathode installed in surface water. We present the first results of laboratory sedimentary fuel cell and a prototype installed in the river.

Fuel rod leak detector

International Nuclear Information System (INIS)

Womack, R.E.

1978-01-01

A typical embodiment of the invention detects leaking fuel rods by means of a radiation detector that measures the concentration of xenon-133 ( 133 Xe) within each individual rod. A collimated detector that provides signals related to the energy of incident radiation is aligned with one of the ends of a fuel rod. A statistically significant sample of the gamma radiation (γ-rays) that characterize 133 Xe is accumulated through the detector. The data so accumulated indicates the presence of a concentration of 133 Xe appropriate to a sound fuel rod, or a significantly different concentration that reflects a leaking fuel rod

Zero Emission Bay Area (ZEBA) Fuel Cell Bus Demonstration Results: Fifth Report

Energy Technology Data Exchange (ETDEWEB)

Eudy, Leslie [National Renewable Energy Lab. (NREL), Golden, CO (United States); Post, Matthew [National Renewable Energy Lab. (NREL), Golden, CO (United States); Jeffers, Matthew [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2016-06-01

This report presents results of a demonstration of fuel cell electric buses (FCEB) operating in Oakland, California. Alameda-Contra Costa Transit District (AC Transit) leads the Zero Emission Bay Area (ZEBA) demonstration, which includes 13 advanced-design fuel cell buses and two hydrogen fueling stations. The ZEBA partners are collaborating with the U.S. Department of Energy (DOE) and DOE's National Renewable Energy Laboratory (NREL) to evaluate the buses in revenue service. NREL has published four previous reports describing operation of these buses. This report presents new and updated results covering data from January 2015 through December 2015.

Zero Emission Bay Area (ZEBA) Fuel Cell Bus Demonstration Results: Sixth Report

Energy Technology Data Exchange (ETDEWEB)

Eudy, Leslie [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Post, Matthew B. [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Jeffers, Matthew A. [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

2017-09-11

This report presents results of a demonstration of fuel cell electric buses (FCEB) operating in Oakland, California. Alameda-Contra Costa Transit District (AC Transit) leads the Zero Emission Bay Area (ZEBA) demonstration, which includes 13 advanced-design fuel cell buses and two hydrogen fueling stations. The ZEBA partners are collaborating with the U.S. Department of Energy (DOE) and DOE's National Renewable Energy Laboratory (NREL) to evaluate the buses in revenue service. NREL has published five previous reports describing operation of these buses. This report presents new and updated results covering data from January 2016 through December 2016.

Behavior of pre-irradiated fuel under a simulated RIA condition. Results of NSRR Test JM-5

International Nuclear Information System (INIS)

Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Tanzawa, Sadamitsu; Ishijima, Kiyomi; Kobayashi, Shinsho; Kamata, Hiroshi; Homma, Kozo; Sakai, Haruyuki.

1995-11-01

This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-5, conducted in the Nuclear Safety Research Reactor (NSRR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and interpretations and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 25.7 MWd/kgU with average linear heat rate of 33.4 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a desposited energy of 223 ± 7 cal/g·fuel (0.93 ± 0.03 kJ/g·fuel) and a peak fuel enthalpy of 167 ± 5 cal/g·fuel (0.70 ± 0.02 kJ/g·fuel) under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The Test JM-5 resulted in cladding failure. More than twenty small cracks were found in the post-test cladding, and most of the defects located in pre-existing locally hydrided region. The result indicates an occurrence of fuel failure by PCMI (pellet/cladding mechanical interaction) in combination with decreased integrity of hydrided cladding. (author)

Fuel improvement and WWER-1000 FA main operational results

International Nuclear Information System (INIS)

Rozhkov, V.; Enin, A.; Bezborodov, Y.; Petrov, V.

2003-01-01

The JSC NCCP experience of WWER-1000 Fuel Assemblies (FAs) fabrication and operation confirms the adequate feasibility and efficiency of fuel operation in 3-4-x fuel cycles, high operating reliability and competitive capacity as compared with foreign analogues. The work on fuel improvement is aimed at an improvement of the operating reliability and an enhancement of the fuel use efficiency in WWER-1000 advanced FAs

Results of experimental investigations for substantiation of WWER cermet fuel pin performance

International Nuclear Information System (INIS)

Popov, V.V.; Karpin, A.D.; Isupov, I.A.; Rumyantsev, V.N.; Troyanov, V.M.; Subonyaev, V.N.; Melnichenko, N.A.

1997-01-01

The out-of-pile experiment results on interaction of the cladding and matrix materials and uranium dioxide at cermet fuel temperature for normal operating conditions of the WWER-440 reactor are analyzed. Cermet fuel element behaviour under the maximum designed damage of the WWER-440 reactor is considered. In the AM reactor loop a fission product output from the unsealed cermet fuel elements have been studied. (author). 6 figs, 3 tabs

Design and preliminary results of a fuel flexible industrial gas turbine combustor

Science.gov (United States)

Novick, A. S.; Troth, D. L.; Yacobucci, H. G.

1981-01-01

The design characteristics are presented of a fuel tolerant variable geometry staged air combustor using regenerative/convective cooling. The rich/quench/lean variable geometry combustor is designed to achieve low NO(x) emission from fuels containing fuel bound nitrogen. The physical size of the combustor was calculated for a can-annular combustion system with associated operating conditions for the Allison 570-K engine. Preliminary test results indicate that the concept has the potential to meet emission requirements at maximum continuous power operation. However, airflow sealing and improved fuel/air mixing are necessary to meet Department of Energy program goals.

Results of calculation of WWER-440 fuel rods (Kol`skaya-3 NPP) at high burnup

Energy Technology Data Exchange (ETDEWEB)

Scheglov, A; Proselkov, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Panin, M; Pitkin, Yu [Kol` skaya NPP, (Russian Federation); Tzibulya, V [AO Mashinostroitelnij Zavod Electrostal (Russian Federation)

1994-12-31

Thermal-physical characteristics of fuel rods of two fuel assemblies which were operated within 5 - 8 and 5 - 9 core fuel loadings of the Unit 3 of the Kol`skaya NPP are calculated. They have achieved deep burnup during 4-year (> 46 Mwd/kg U) and 5-year (> 48 Mwd/kg U) fuel cycle. Fuel assemblies have been unloaded off the reactor and subjected to a post-irradiation testing. PIN-mod2 code originally designed for modelling of WWER fuel rod behaviour in a quasi-steady-state operation is used. The average fuel rod in the fuel assembly and the fuel rod with maximum burnup are selected. The preliminary comparison of the calculation results with those of the post-irradiation examination shows a satisfactory agreement. On the basis of the results obtained in the post-irradiation experiments an improvement of the model for calculation of fission gas release and creep of the cladding is planned. The results of the analysis performed indicate that the fuel rod completely preserves its working ability; fuel temperature does not exceed 1300{sup o} C; fission gas release does not exceed 4%; maximum gas pressure inside the cladding at the end of campaign does not exceed 2 MPa. 2 tabs., 11 figs., 5 refs.

Comprehensive CFD analyses concerning the serious incident occurred in the Paks NPP, Spring 2003

International Nuclear Information System (INIS)

Legradi, Gabor; Boros, Ildiko; Aszodi, Attila

2007-01-01

On 10-11 April, 2003, a serious incident occurred in a special fuel assembly cleaning tank, which was installed into the service shaft of the 2. unit of the Paks NPP of Hungary. As a consequence of the incident, most of the 30 fuel assemblies put into the cleaning tank have seriously damaged. At the Institute of Nuclear Techniques, Budapest University of Technology and Economics, several CFD investigations were performed concerning the course of the incident, the post incidental conditions and the recovery work. The main reason of the incident can be originated from the defective design of the cleaning tank which resulted in the insufficient cooling of the system. First, the CFD calculations, with which the thermal-hydraulic deficiencies of the design were revealed, are presented in this paper. This investigation clearly showed how as strong temperature stratification could develop inside the cleaning tank that it was able to block the coolant flow through the fuel assemblies. After the blocking of the flow, the coolant started to boil, and the assemblies became uncovered. The temperature of the surfaces of the fuel assemblies rose above 1000 deg. C. With the aid of the radiative heat transfer model of the CFX-5.6 code, the surface temperatures and also the heat transfer between the shrouds of the assemblies and the inner surface of the cleaning tank mantle were analyzed. When the cleaning instrument got opened the fuel assemblies suffered a serious thermal shock and the assemblies highly damaged. For better understanding of the post-incident state inside the cleaning vessel and for serving as a basis for the safety analyses of the cooling systems, other detailed 3D thermal-hydraulic calculations were performed. These calculations also serve as a basis for the recovery work. The recovery work is planned to be started in the close future. During this work, the service shaft will be operated in a low-coolant-level operational mode. The operators of the damaged fuel

Probabilistic analysis of fuel pin behaviour during an eventual loss of coolant in PWR reactors

International Nuclear Information System (INIS)

1981-02-01

Brief description of the development of the coolant loss incident in a pressurized water reactor and analysis of its significance for the behaviour of the fuel rods. Description of a probalistic method for estimating the effects of the accident on the fuel rods and results obtained [fr

Joint studies of LOF and TOP incidents for a 1300 MW(E) LMFBR using the computer codes SAS3D/EPIC and FRAX-2

International Nuclear Information System (INIS)

Leslie, R.; Billington, D.E.; Mann, J.E.

1982-04-01

The results of joint studies carried out for a 1300MW(E) LMFBR are described. The incidents examined were a slow TOP (3c/s) and a LOF (pump rundown with 9s flow halving time), both with failure to trip. For the TOP incident a benign outcome was predicted largely as a consequence of the prediction of clad failure near the top of the core. For the LOF incident highly energetic outcomes were not predicted for the reference case because the incident was terminated by disassembly (by fuel vapour pressure) in voided channels and failures in low-rated flooded channels with MFCI potential were not predicted. In the variant cases where MFCIs were predicted before shutdown, and rapid enough extension of the clad rips was allowed, low energetics were still predicted as a consequence of fuel sweepout. The strength of the MFCIs (as represented by a Cho-Wright treatment) does not appear to be an important factor but the results are dependent on the prediction of negative reactivity addition through fuel sweepout. The physical conditions obtaining at the time of fuel failure are such as to suggest that internal fuel motion following failure should not have an important effect on accident energetics, unless the development of the initial rip is delayed by several milliseconds. This is an area where only limited experimental evidence is available. Other areas of uncertainty are associated with the position of failure, of clad rip propagation and the influence of incoherency on the progression of the incident. Clad motion effects were shown not to influence accident energetics significantly for the reactor model considered. (author)

Integration of post-irradiation examination results of failed WWER fuel rods

International Nuclear Information System (INIS)

Smirnov, A.; Markov, D.; Smirnov, V.; Polenok, V.; Perepelkin, S.

2003-01-01

The aim of the work is to investigate the causes of WWER fuel rod failures and to reveal the dependence of the failed fuel rod behaviour and state on the damage characteristics and duration of their operation in the core. The post-irradiation examination of 12 leaky fuel assemblies (5 for WWER-440 and 7 for WWER-1000) has been done at SSC RF RIAR. The results show that the main mechanism responsible for the majority of cases of the WWER fuel rod perforation is debris-damage of the claddings. Debris fretting of the claddings spread randomly over the fuel assembly cross-section and they are registered in the area of the bundle supporting grid or under the lower spacer grids along the fuel assembly height. In the WWER fuel rods, the areas of secondary hydrogenating of cladding are spaced from the primary defects by ∼2500-3000 mm, as a rule, and are often adjacent closely to the upper welded joints. There is no pronounced dependence of the distance between the primary and secondary cladding defects neither on the linear power, at which the fuel rods were operated, nor on the period of their operation in the leaky state. The time period of the significant secondary damage formation is about 250 ± 50 calendar days for the WWER fuel rods with slight through primary defects (∼0.1 - 0.5 mm 2 ) operated in the linear power range 170-215 W/cm. Cladding degradation, taking place due to the secondary hydrogenating, does not occur in case of large through debris-defects during operation up to 600 calendar days

Results on Technical and Consultants Service Meetings on Lessons Learned from Operating Experience in Wet and Dry Spent Fuel Storage

International Nuclear Information System (INIS)

White, B.; Zou, X.

2015-01-01

Spent fuel storage has been and will continue to be a vital portion of the nuclear fuel cycle, regardless of whether a member state has an open or closed nuclear fuel cycle. After removal from the reactor core, spent fuel cools in the spent fuel pool, prior to placement in dry storage or offsite transport for disposal or reprocessing. Additionally, the inventory of spent fuel at many reactors worldwide has or will reach the storage capacity of the spent fuel pool; some facilities are alleviating their need for additional storage capacity by utilizing dry cask storage. While there are numerous differences between wet and dry storage; when done properly both are safe and secure. The nuclear community shares lessons learned worldwide to gain knowledge from one another’s good practices as well as events. Sharing these experiences should minimize the number of incidents worldwide and increase public confidence in the nuclear industry. Over the past 60 years, there have been numerous experiences storing spent fuel, in both wet and dry mediums, that when shared effectively would improve operations and minimize events. These lessons learned will also serve to inform countries, who are new entrants into the nuclear power community, on designs and operations to avoid and include as best practices. The International Atomic Energy Agency convened a technical and several consultants’ meetings to gather these experiences and produce a technical document (TECDOC) to share spent fuel storage lessons learned among member states. This paper will discuss the status of the TECDOC and briefly discuss some lessons learned contained therein. (author)

Incidence of cervical cancer after several negative smear results by age 50: prospective observational study

DEFF Research Database (Denmark)

Rebolj, Matejka; van Ballegooijen, Marjolein; Lynge, Elsebeth

2009-01-01

/100,000 (95% confidence interval 33 to 51) in the younger group and 36/100,000 (24 to 52) in the older group (P=0.48). The cumulative incidence rate of cervical intraepithelial neoplasia grade I+ was twice as high in the younger than in the older group (Pcervical cancer......OBJECTIVE: To determine the incidence of cervical cancer after several negative cervical smear tests at different ages. DESIGN: Prospective observational study of incidence of cervical cancer after the third consecutive negative result based on individual level data in a national registry...... of histopathology and cytopathology (PALGA). SETTING: Netherlands, national data. Population 218,847 women aged 45-54 and 445,382 aged 30-44 at the time of the third negative smear test. MAIN OUTCOME MEASURES: 10 year cumulative incidence of interval cervical cancer. RESULTS: 105 women developed cervical cancer...

Results of Cesar II critical facility with low enriched fuel balls

Energy Technology Data Exchange (ETDEWEB)

Langlet, G; Guerange, J; Laponche, B; Morier, F; Neef, R D; Bock, H J; Kring, F J; Scherer, W

1972-06-15

The Cesar facility has been transformed to load in its center a pebble bed fuel. This new Cesar assembly is called Cesar II. The program for the measurements with HTR type fuel balls is managed under a cooperation between physicists of CEA/CADARACHE and KFA/JUELICH. A description of the measuring zones of Cesar II and of the experimental results is given.

Statement of nuclear incidents at nuclear installations. Third quarter 2001

International Nuclear Information System (INIS)

2002-01-01

A statement of nuclear incidents at nuclear installations in Britain during the third quarter of 2001 is published today by the Health and Safety Executive (copy attached). It covers the period 1 July to 30 September 2001. The statement is published under arrangements that came into effect from the first quarter of 1993, derived from the Health and Safety Commission's powers under section 11 of the Health and Safety at Work, etc. Act 1974. Normally each incident mentioned in HSE's Quarterly Incident Statements will already have been made public by the licensee or site operator either through a press statement or by inclusion in the newsletter for the site concerned. The locations of the installations mentioned in the statement are as follows: Heysham 1 (British Energy Generation plc), Sellafield (British Nuclear Fuels plc), Chapelcross (British Nuclear Fuels plc)

Physico-chemical characteristics of eight different biomass fuels and comparison of combustion and emission results in a small scale multi-fuel boiler

International Nuclear Information System (INIS)

Forbes, E.G.A.; Easson, D.L.; Lyons, G.A.; McRoberts, W.C.

2014-01-01

Highlights: • Physical parameters of the eight biomass fuels examined were all different. • Significant differences were found in Proximate, Ultimate and TGA results. • Energy outputs were not proportionate to dry matter energy content. • Highest flue ash production from fuels with highest fines content. • Flue gas emissions varied significantly, NOx levels correlated with fuel N content. - Abstract: This study describes the results from the investigation of 7 different biomass fuel types produced on a farm, and a commercial grade wood pellet, for their physical, chemical, thermo-gravimetric and combustion properties. Three types of short rotation coppice (SRC) willow, two species of conifers, forest residues (brash), commercially produced wood-pellets and a chop harvested energy grass crop Miscanthus giganteus spp., (elephant grass) were investigated. Significant differences (p < 0.05) were found in most of the raw fuel parameters examined using particle distribution, Thermogravimetric, Ultimate and Proximate analysis. Combustion tests in a 120 kW multi-fuel boiler revealed differences, some significant, in the maximum output, energy conversion efficiency, gaseous emission profiles and ash residues produced from the fuels. It was concluded that some of the combustion results could be directly correlated with the inherent properties of the different fuels. Ash production and gaseous emissions were the aspects of performance that were clearly and significantly different though effects on energy outputs were more varied and less consistent. The standard wood pellet fuel returned the best overall performance and miscanthus produced the largest amount of total ash and clinker after combustion in the boiler

Fuel models and results from the TRAC-PF1/MIMAS TMI-2 accident calculation

International Nuclear Information System (INIS)

Schwegler, E.C.; Maudlin, P.J.

1983-01-01

A brief description of several fuel models used in the TRAC-PF1/MIMAS analysis of the TMI-2 accident is presented, and some of the significant fuel-rod behavior results from this analysis are given. Peak fuel-rod temperatures, oxidation heat production, and embrittlement and failure behavior calculated for the TMI-2 accident are discussed. Other aspects of fuel behavior, such as cladding ballooning and fuel-cladding eutectic formation, were found not to significantly affect the accident progression

SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: Fourth Results Report

Energy Technology Data Exchange (ETDEWEB)

Eudy, L.; Chandler, K.

2013-01-01

SunLine Transit Agency, which provides public transit services to the Coachella Valley area of California, has demonstrated hydrogen and fuel cell bus technologies for more than 10 years. In May 2010, SunLine began demonstrating the advanced technology (AT) fuel cell bus with a hybrid electric propulsion system, fuel cell power system, and lithium-based hybrid batteries. This report describes operations at SunLine for the AT fuel cell bus and five compressed natural gas buses. The U.S. Department of Energy's National Renewable Energy Laboratory (NREL) is working with SunLine to evaluate the bus in real-world service to document the results and help determine the progress toward technology readiness. NREL has previously published three reports documenting the operation of the fuel cell bus in service. This report provides a summary of the results with a focus on the bus operation from February 2012 through November 2012.

Main results and status of the development of LEU fuel for Russian research reactors

International Nuclear Information System (INIS)

Vatulin, A.; Morozov, A.; Suprun, V.; Dobrikova, I.

2005-01-01

VNIINM develops low enrichment uranium (LEU) fuel on base U-Mo alloys and a novel design of pin-type fuel elements. The development is carried out both for existing reactors, and for new advanced designs of reactors. The work is carried on the following main directions: - irradiate LEU U-Mo dispersion fuel (the uranium density up to 6,0 g/cm 3 ) in two Russian research reactors: MIR (RIAR, Dimitrovgrad) as pin type fuel mini-elements and in WWR-M (PINP, Gatchina) within full-scaled fuel assembly (FA) with pin type fuel elements; - finalize development of design and fabrication process of IRT type FA with pin type fuel elements; - develop methods of reducing of U-Mo fuel --Al matrix interaction under irradiation; - develop fabricating methods of fuel elements on base of monolithic U-Mo fuel. The paper generally reviews the results of calculation, design and technology investigations accomplished by now. (author)

Postirradiation examination results from the LP-FP-2 center fuel module

International Nuclear Information System (INIS)

Jensen, S.M.; Akers, D.W.

1990-01-01

The LP-FP-2 experiment was conducted on July 9, 1985 in the Loss of Fluid Test (LOFT) facility located at the Idaho National Engineering Laboratory (INEL). The primary purpose of this experiment was to provide information of the release, transport, and deposition of fission products and aerosols during a sever core damage event performed in a large scale nuclear reactor facility. Postirradiation nondestructive and destructive examinations of the fuel bundle provided information to assist in achieving this objective, as well as providing information on the material behavior and interactions that occurred within the fuel bundle during this sever core damage experiment. This was a large-scale integral test, incorporating an 11 x 11 array of fuel rods, control rods, and instrumentation tubes, with an active core length of 1.68 m. Peak temperatures in the fuel bundle exceeded 2100 K or approximately 4.5 min, with localized peak temperatures exceeding the melting point of the UO 2 fuel (3120 K). Large amounts of zircaloy oxidation and material relocation occurred during the experiment. The transient phase was terminated by a rapid reflood of cooling water, which resulted in significant oxidation and hydrogen generation. Zircaloy oxidation during the reflood period caused a rapid temperature excursion to occur in the upper two-thirds of the fuel bundle. This article summarizes the data and analysis from the postirradiation examinations of the LP-FP-2 fuel bundle. 12 refs., 39 figs., 8 tabs

Fuel and fuel pin behaviour in a high burnup fast breeder fuel subassembly: Results of destructive post-irradiation examinations of the KNK II/1 fuel subassembly NY-205

International Nuclear Information System (INIS)

Patzer, G.

1991-05-01

The report gives a summarizing overview of the design characteristics, of the irradiation history and of the results of the destructive post-irradiation examinations of the fuel pins of the high-burnup fuel subassembly NY-205 of the KNK II first core. This element was operated for about 10 years and reached a maximum local burnup of 175 MWd/kg(HM) and a maximum neutron dose of 67 dpa-NRT. The main design data of this subassembly agree with those of the SNR 300 Mark-Ia, and it reached more than twice of the burnup and a similar neutron dose as foreseen for the SNR 300 fuel subassemblies [de

Solid fuel block as an alternate fuel for cooking and barbecuing: Preliminary results

International Nuclear Information System (INIS)

Sharma, Monikankana; Mukunda, H.S.; Sridhar, G.

2009-01-01

A large part of the rural people of developing countries use traditional biomass stoves to meet their cooking and heating energy demands. These stoves possess very low thermal efficiency; besides, most of them cannot handle agricultural wastes. Thus, there is a need to develop an alternate cooking contrivance which is simple, efficient and can handle a range of biomass including agricultural wastes. In this reported work, a highly densified solid fuel block using a range of low cost agro residues has been developed to meet the cooking and heating needs. A strategy was adopted to determine the best suitable raw materials, which was optimized in terms of cost and performance. Several experiments were conducted using solid fuel block which was manufactured using various raw materials in different proportions; it was found that fuel block composed of 40% biomass, 40% charcoal powder, 15% binder and 5% oxidizer fulfilled the requirement. Based on this finding, fuel blocks of two different configurations viz. cylindrical shape with single and multi-holes (3, 6, 9 and 13) were constructed and its performance was evaluated. For instance, the 13 hole solid fuel block met the requirement of domestic cooking; the mean thermal power was 1.6 kW th with a burn time of 1.5 h. Furthermore, the maximum thermal efficiency recorded for this particular design was 58%. Whereas, the power level of single hole solid fuel block was found to be lower but adequate for barbecue cooking application

NASA fuel cell applications for space: Endurance test results on alkaline fuel cell electrolyzer components

International Nuclear Information System (INIS)

Sheibley, D.W.

1984-01-01

Fuel cells continue to play a major role in manned spacecraft power generation. The Gemini and Apollo programs used fuel cell power plants as the primary source of mission electrical power, with batteries as the backup. The current NASA use for fuel cells is in the Orbiter program. Here, low temperature alkaline fuel cells provide all of the on-board power with no backup power source. Three power plants per shipset are utilized; the original power plant contained 32-cell substacks connected in parallel. For extended life and better power performance, each power plant now contains three 32-cell substacks connected in parallel. One of the possible future applications for fuel cells will be for the proposed manned Space Station in low earth orbit (LEO)(1, 2, 3). By integrating a water electrolysis capability with a fuel cell (a regenerative fuel cell system), a multikilowatt energy storage capability ranging from 35 kW to 250 kW can be achieved. Previous development work on fuel cell and electrolysis systems would tend to minimize the development cost of this energy storage system. Trade studies supporting initial Space Station concept development clearly show regenerative fuel cell (RFC) storage to be superior to nickel-cadmium and nickel-hydrogen batteries with regard to subsystem weight, flexibility in design, and integration with other spacecraft systems when compared for an initial station power level ranging from 60 kW to 75 kW. The possibility of scavenging residual O 2 and H 2 from the Shuttle external tank for use in fuel cells for producing power also exists

Results of 200 KW fuel cell evaluation programs

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

Science.gov (United States)

Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.

1991-02-01

The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

Intermodal transfer of spent fuel

International Nuclear Information System (INIS)

Neuhauser, K.S.; Weiner, R.F.

1991-01-01

As a result of the international standardization of containerized cargo handling in ports around the world, maritime shipment handling is particularly uniform. Thus, handier exposure parameters will be relatively constant for ship-truck and ship-rail transfers at ports throughout the world. Inspectors' doses are expected to vary because of jurisdictional considerations. The results of this study should be applicable to truck-to-rail transfers. A study of the movement of spent fuel casks through ports, including the loading and unloading of containers from cargo vessels, afforded an opportunity to estimate the radiation doses to those individuals handling the spent fuels with doses to the public along subsequent transportation routes of the fuel. A number of states require redundant inspections and for escorts over long distances on highways; thus handlers, inspectors, escort personnel, and others who are not normally classified as radiation workers may sustain doses high enough to warrant concern about occupational safety. This paper addresses the question of radiation safety for these workers. Data were obtained during, observation of the offloading of reactor spent fuel (research reactor spent fuel, in this instance) which included estimates of exposure times and distances for handlers, inspectors and other workers during offloading and overnight storage. Exposure times and distance were also for other workers, including crane operators, scale operators, security personnel and truck drivers. RADTRAN calculational models and parameter values then facilitated estimation of the dose to workers during incident-free ship-to-truck transfer of spent fuel

Incident reporting to BfArM - regulatory framework, results and challenges.

Science.gov (United States)

Seidel, Robin; Stößlein, Ekkehard; Lauer, Wolfgang

2016-04-01

Medical devices are manifold and one of the most innovative fields of technology. As technologies advance, former limits cease to exist and complex devices become reality. Medical devices represent a very dynamic field with high economic relevance. The manufacturer of a medical device is obliged to minimize product-related risks as well as to demonstrate compliance with the so-called "essential requirements" regarding safety and performance before placing the device on the market. Any critical incident in relation to the application of a medical device has to be reported to the competent authority for risk assessment, which in Germany is either the Federal Institute for Drugs and Medical Devices (BfArM) or the Paul Ehrlich Institute (PEI) depending on the type of device. In this article, the German regulatory framework for medical devices and the resulting tasks for BfArM are described as well as the topics of its recently installed research and development group on prospective risk identification and application safety for medical devices. Results of failure mode and root cause analyses of incident data are presented as well as further data on cases with the result "root-cause analysis not possible". Finally an outlook is given on future challenges regarding risk assessment for medical devices.

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

Energy Technology Data Exchange (ETDEWEB)

David Petti

2014-06-01

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

Posttest examination results of recent treat tests on metal fuel

International Nuclear Information System (INIS)

Holland, J.W.; Wright, A.E.; Bauer, T.H.; Goldman, A.J.; Klickman, A.E.; Sevy, R.H.

1986-01-01

A series of in-reactor transient tests is underway to study the characteristics of metal-alloy fuel during transient-overpower-without-scam conditions. The initial tests focused on determining the margin to cladding breach and the axial fuel motions that would mitigate the power excursion. The tests were conducted in flowing-sodium loops with uranium - 5% fissium EBR-II Mark-II driver fuel elements in the TREAT facility. Posttest examination of the tests evaluated fuel elongation in intact pins and postfailure fuel motion. Microscopic examination of the intact pins studied the nature and extent of fuel/cladding interaction, fuel melt fraction and mass distribution, and distribution of porosity. Eutectic penetration and failure of the cladding were also examined in the failed pins

SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: First Results Report

Energy Technology Data Exchange (ETDEWEB)

Eudy, L.; Chandler, K.

2011-03-01

This report describes operations at SunLine Transit Agency for their newest prototype fuel cell bus and five compressed natural gas (CNG) buses. In May 2010, SunLine began operating its sixth-generation hydrogen fueled bus, an Advanced Technology (AT) fuel cell bus that incorporates the latest design improvements to reduce weight and increase reliability and performance. The agency is collaborating with the U.S. Department of Energy's (DOE) National Renewable Energy Laboratory (NREL) to evaluate the bus in revenue service. This report provides the early data results and implementation experience of the AT fuel cell bus since it was placed in service.

Assessment of radiation doses due to normal operation, incidents and accidents of the final disposal facility

International Nuclear Information System (INIS)

Rossi, J.; Raiko, H.; Suolanen, V.; Ilvonen, M.

1999-03-01

Radiation doses for workers of the encapsulation and disposal facility and for inhabitants in the environment caused by the facility during its operation were considered. The study covers both the normal operation of the plant and some hypothetical incidents and accidents. Occupational radiation doses inside the plant during normal operation are based on the design basis, assuming that highest permitted dose levels are prevailing in control rooms during fuel transfer and encapsulation processes. Release through the ventilation stack is assumed to be filtered both in normal operation and in hypothetical incident and accident cases. Calculation of the offsite doses from normal operation is based on the hypothesis that one fuel pin per 100 fuel bundles for all batches of spent fuel transported to the encapsulation facility is leaking. The release magnitude in incidents and accidents is based on the event chains, which lead to loss of fuel pin tightness followed by a discharge of radionuclides into the handling chamber and to some degree through the ventilation stack into atmosphere. The weather data measured at the Olkiluoto meteorological mast was employed for calculating of offsite doses. Therefore doses could be calculated in a large amount of different dispersion conditions, the statistical frequencies of which have, been measured. Finally doses were combined into cumulative distributions, from which a dose value representing the 99.5 % confidence level, is presented. The dose values represent the exposure of a critical group, which is assumed to live at the distance of 200 meters from the encapsulation and disposal plant and thus it will receive the largest doses in most dispersion conditions. Exposure pathways considered were: cloudsnine, inhalation, groundshine and nutrition (milk of cow, meat of cow, green vegetables, grain and root vegetables). Nordic seasonal variation is included in ingestion dose models. The results obtained indicate that offsite doses

Data analysis of grade crossing incidents : research results.

Science.gov (United States)

2011-12-01

Incidents and fatalities at highway-rail grade crossing in the United States have declined significantly over the past two decades despite a significant increase in both train and vehicle traffic. Therefore, to provide a more realistic comparison of ...

Descriptive Study of the Environmental Epidemiology of High Lung Cancer 
Incidence Rate in Qujing, Yunnan, China

Directory of Open Access Journals (Sweden)

Linlin ZHANG

2012-03-01

Full Text Available Background and objective Qujing, located in Southwest China, is an area with an extremely high lung cancer incidence. Combustion of coal has exposed local people to great health hazards. The aim of this study is to achieve a thorough understanding of the relationship between environmental pollution and the high incidence of lung cancer in Qujing, Yunnan Province, China. The results would provide a scientific basis and support for the etiology of lung cancer, as well as suggestions on improving the environmental conditions in the area. Methods A total of 280 rural villages were selected through stratified cluster random sampling. Environmental background and pollution were investigated, including details on fuel type, coking plant, metal smelting, and chemical plant, among others. Logistic regression analysis was used to analyze the investigated factors. Results Out of the total number of local villages studied, 78.1% of those with high incidence often use smoky coal and coking. On the other hand, 78.8% of the low-incidence areas use smokeless coal or wood. Logistic regression analysis indicated that the coal type used for everyday life was a main risk factor related to lung cancer (P<0.05. Using smoky and coking coals create an alarmingly high risk for developing lung cancer. Meanwhile, smokeless coals and wood seemed to have no significant relationship to the lung cancer incidence. Conclusion The fuel type used for everyday life is an important factor in the high incidence of lung cancer in Qujing. Evidently, the use of smoky coal and coke increased the incidence of lung cancer, whereas smokeless coal and wood seem to bring about the contrary.

Spent Fuel Drying System Test Results (Dry-Run in Preparation for Run 8)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1999-01-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL)(a)on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of a test ''dry-run'' conducted prior to the eighth and last of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister6513U. The system used for the dry-run test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. The experimental results are provided in Section 4.0 and discussed Section 5.0

Incident factor as a learning aspect to enhance safety culture in the experimental fuel element installation of PTBN - BATAN

International Nuclear Information System (INIS)

Heri Hardiyanti; Agus Sartono; Bambang Herutomo; AS Latief

2013-01-01

The safety of a nuclear facility depends not only on the fulfillment of all technical requirements, but also on the role of non-technical aspects. The primary causation of incidents or accidents in a nuclear facility is human error which is non-technical. Therefore, in order to enhance safety, efforts from the technical aspects are as important as efforts to deal with the human factor which can be done through the application of safety culture in the facility. Incidents that took place in the Experimental Fuel Element Installation (EFEI) of PTBN - BATAN from 2011 to 2012 were caused by aging instruments and human error. In order to prevent accidents and to enhance safety, non-technical efforts that were done in the EFEI were, interalia, the obligations on all personnel to attend the pre-lab briefing, to prepare a work proposal, to compose a HIRADC (hazard identification, risk assessment, and determining control) document, to utilize self protection devices, to perform a routine maintenance, and to practice safe behavior. All personnel were involved in all those efforts. Safety is the first priority and can always be improved in the facility. A strong commitment of and cooperation between the top management and the staff are needed. (author)

Description of a plutonium incident in the Hot Laboratory on May 24, 1983, and the resulting consequences

International Nuclear Information System (INIS)

Hausmann, W.; Francioni, W.; Ingold, F.; Ledergerber, G.

1985-08-01

At 16.15 on 24 May 1983 a rapid chemical reaction occured in an apparatus in which a routine waste sludge evaporation process was under way. The waste was produced during fuel fabrication. The resulting pressure wave ruptured a number of parts at the glove box in which the equipment was situated leading to an alpha contamination of that and neighboring laboratories. The alarm system activated by the fire alarm, the activity detectors and manualy by the operators functioned correctly. No one was physicaly injured and thanks to immediate protecture measures (e.g. gas masks) the amount of alpha-dust incorporated could be limited in degree and extent. Very time-consuming and costly cleaning and decontamination procedures were found to be necessary but these led to the encouraging result that the greater part of the contaminated equipment (particularly electrical and electronic apparatus) could be released for further unrestricted use in open labs. Detailed tests with the original waste solutions showed that at a certain ammonium nitrate content and at 235 C an exothermic chemical reaction can occur in the sludge. This temperature can only be reached due to a defective apparatus (e.g. a broken glass vessel) allowing the sludge to contact the heating element directly. The external expert agrees that this is the most likely cause of the incident. (author)

Development of incident progress prediction technologies for nuclear emergency preparedness. Current status and future subjects

International Nuclear Information System (INIS)

Yoshida, Yoshitaka; Yamamoto, Yasunori; Kusunoki, Takayoshi; Kawasaki, Ikuo; Yanagi, Chihiro; Kinoshita, Ikuo; Iwasaki, Yoshito

2014-01-01

Nuclear licensees are required to maintain a prediction system during normal condition for using a nuclear emergency by the Basic Plan for Disaster Prevention of government. With prediction of the incident progress, if the present condition of nuclear power plant is understood appropriately and it grows more serious with keeping the present situation, it is in predicting what kind of situation will be occurred in the near future, choosing the effective countermeasures against the coming threat, and understanding the time available of intervention time. Following the accident on September 30 1999 in the nuclear fuel fabrication facility in Tokai Village of Ibaraki Prefecture, the Institute of Nuclear Safety System started development of incident progress prediction technologies for nuclear emergency preparedness. We have performed technical applications and made improvements in nuclear emergency exercises and verified the developed systems using the observed values of the Fukushima Daiichi Nuclear Power Plant accident. As a result, our developed Incident Progress Prediction System was applied to nuclear emergency exercises and we accumulated knowledge and experience by which we improved the system to make predictions more rapidly and more precisely, including for example, the development of a prediction method for leak size of reactor coolant. On the other hand, if a rapidly progressing incident occurs, since end users need simple and quick predictions about the public's protection and evacuation areas, we developed the Radioactive Materials Release, Radiation Dose and Radiological Protection Area Prediction System which changed solving an inverse problem into a forward problem solution. In view of the water-level-decline incident of the spent fuel storage facility at the Fukushima Daiichi Nuclear Power Plant, the spent fuel storage facility water level and the water temperature evaluation tool were improved. Such incident progress prediction technologies were

Testing Systems and Results for Advanced Nuclear Fuel Materials

International Nuclear Information System (INIS)

Rooyen, I.J. van; Griffith, G.W.; Garnier, J.E.

2012-01-01

Light Water Reactor Sustainability (LWRS) Program Advanced LWR Nuclear Fuel Development (ALFD) Pathway. Development and testing of high performance fuel cladding identified as high priority to support: enhancement of fuel performance, reliability, and reactor safety. One of the technologies being examined is an advanced fuel cladding made from ceramic matrix composites (CMC) utilizing silicon carbide (SiC) as a structural material supplementing a commercial Zircaloy-4 (Zr-4) tube. A series of out-of-pile tests to fully characterize the SiC CMC hybrid design to produce baseline data. The planned tests are intended to either produce quantitative data or to demonstrate the properties required to achieve two initial performance conditions relative to standard zircaloybased cladding: decreased hydrogen uptake (corrosion) and decreased fretting of the cladding tube under normal operating and postulated accident conditions. These two failure mechanisms account for approximately 70% of all in-pile failures of LWR commercial fuel assemblies

Design of JMTR high-performance fuel element

International Nuclear Information System (INIS)

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

1999-01-01

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

SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: Third Results Reports

Energy Technology Data Exchange (ETDEWEB)

Eudy, L.; Chandler, K.

2012-05-01

This report describes operations at SunLine Transit Agency for their newest prototype fuel cell bus and five compressed natural gas (CNG) buses. In May 2010, SunLine began operating its sixth-generation hydrogen fueled bus, an Advanced Technology (AT) fuel cell bus that incorporates the latest design improvements to reduce weight and increase reliability and performance. The agency is collaborating with the U.S. Department of Energy's (DOE) National Renewable Energy Laboratory (NREL) to evaluate the bus in revenue service. NREL has previously published two reports documenting the operation of the fuel cell bus in service. This report provides a summary of the results with a focus on the bus operation from July 2011 through January 2012.

First results on the effect of fuel-cladding eccentricity

International Nuclear Information System (INIS)

Panka, I.; Kereszturi, A.

2009-01-01

In the traditional fuel-behaviour or hot channel calculations it is assumed that the fuel pellet is centered within the clad. However, in the real life the pellet could be positioned asymmetrically within the clad, which leads to asymmetric gap conductance and therefore it is worthwhile to investigate the magnitude of the effect on maximal fuel temperature and surface heat flux. In this paper our first experiences are presented on this topic. (Authors)

Similar herpes zoster incidence across Europe: results from a systematic literature review.

Science.gov (United States)

Pinchinat, Sybil; Cebrián-Cuenca, Ana M; Bricout, Hélène; Johnson, Robert W

2013-04-10

Herpes zoster (HZ) is caused by reactivation of the varicella-zoster virus (VZV) and mainly affects individuals aged ≥50 years. The forthcoming European launch of a vaccine against HZ (Zostavax®) prompts the need for a better understanding of the epidemiology of HZ in Europe. Therefore the aim of this systematic review was to summarize the available data on HZ incidence in Europe and to describe age-specific incidence. The Medline database of the National Library of Medicine was used to conduct a comprehensive literature search of population-based studies of HZ incidence published between 1960 and 2010 carried out in the 27 member countries of the European Union, Iceland, Norway and Switzerland. The identified articles were reviewed and scored according to a reading grid including various quality criteria, and HZ incidence data were extracted and presented by country. The search identified 21 studies, and revealed a similar annual HZ incidence throughout Europe, varying by country from 2.0 to 4.6/1 000 person-years with no clearly observed geographic trend. Despite the fact that age groups differed from one study to another, age-specific HZ incidence rates seemed to hold steady during the review period, at around 1/1 000 children European Union Member States and to monitor the impact of VZV immunization on the epidemiology of HZ. Available data in Europe have shortcomings which make an accurate assessment of HZ incidence and change over time impossible. However, data are indicative that HZ incidence is comparable, and increases with age in the same proportion across Europe.

The chemistry of water reactor fuel

International Nuclear Information System (INIS)

Potter, P.E.

1990-01-01

In this paper, the authors discuss features of the changes in chemical constitution which occur in fuel and fuel rods for water reactors during operation and in fault conditions. The fuel for water reactors consists of pellets of urania (UO 2 ) clad in Zircaloy. An essential step in the prediction of the fate of all the radionuclides in a fault or accident is to possess a detailed knowledge of their chemical behavior at all stages of the development of such incidents. In this paper, the authors consider: the chemical constitution of fuel during operation at temperatures corresponding to rather low ratings, together with a quite detailed discussion of the chemistry within the fuel-clad gap; the behavior of fuel subjected to higher temperatures and ratings than those of contemporary fuel; and the changes in constitution on failure of fuel rods in fault or accident conditions

Initial results for electrochemical dissolution of spent EBR-II fuel

International Nuclear Information System (INIS)

Li, S. X.

1998-01-01

Initial results are reported for the anode behavior of spent metallic nuclear fuel in an electrorefining process. The anode behavior has been characterized in terms of the initial spent fuel composition and the final composition of the residual cladding hulls. A variety of results have been obtained depending on the experimental conditions. Some of the process variables considered are average and maximum cell voltage, average and maximum anode voltage, amount of electrical charge passed (coulombs or amp-hours) during the experiment, and cell resistance. The main goal of the experiments has been the nearly complete dissolution of uranium with the retention of zirconium and noble metal fission products in the cladding hulls. Analysis has shown that the most indicative parameters for determining an endpoint to the process, recognizing the stated goal, are the maximum anode voltage and the amount of electrical charge passed. For the initial experiments reported here, the best result obtained is greater than 98% uranium dissolution with approximately 50% zirconium retention. Noble metal fission product retention appears to be correlated with zirconium retention

First ONPE report - French national observatory of fuel poverty. Definitions, indicators, first results and recommendations

International Nuclear Information System (INIS)

Cherel, Didier; Nolay, Pierre; Devaliere, Isolde; Teissier, Olivier; Maresca, Bruno; Guimard, Sebastien; Moisan, Marie; Rousseau, Nicolas; Jouffe, Yves; Poutrel, Severin; Buresi, Sandrine

2014-09-01

This first report from the French national observatory of fuel poverty (ONPE) aims at defining fuel poverty and at characterizing and measuring this phenomenon through indicators and inquiries. An additional dimension concerns the vulnerability linked with everyday mobility which is presented in a separate chapter. The national and local policies against fuel poverty are presented with their results, efficiency and possible improvements. A short glimpse on fuel poverty in Europe is given before the conclusion and recommendations

Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core

International Nuclear Information System (INIS)

Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

1985-01-01

The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO 2 and UO 2 /metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO 2 crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process

Standalone BISON Fuel Performance Results for Watts Bar Unit 1, Cycles 1-3

Energy Technology Data Exchange (ETDEWEB)

Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pawlowski, Roger [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Stimpson, Shane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-03-07

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is moving forward with more complex multiphysics simulations and increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable, and major advances have been made in analysis efforts, including the simulation of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. While this is a major achievement, the VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON fuel performance code. Several approaches are being implemented to consider fuel performance, including a more direct multiway coupling with Tiamat, as well as a more loosely coupled one-way approach with standalone BISON cases. Fuel performance typically undergoes an independent analysis using a standalone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This report summarizes the improvements made since the initial milestone to execute BISON from VERA-CS output. Many of these improvements were prompted through tighter collaboration with the BISON development team at Idaho National Laboratory (INL). A brief description of WBN1 and some of the VERA-CS data used to simulate it are presented. Data from a small mesh sensitivity study are shown, which helps justify the mesh parameters used in this work. The multi-cycle results are presented, followed by the results for the first three cycles of WBN1 operation, particularly the parameters of interest to pellet-clad interaction (PCI) screening (fuel-clad gap closure, maximum centerline fuel temperature, maximum/minimum clad hoop stress, and cumulative damage index). Once the mechanics of this capability are functioning, future work will target cycles with

Results of a conventional fuel tank blunt impact test

Science.gov (United States)

2015-03-23

The Federal Railroad Administrations Office of Research : and Development is conducting research into passenger : locomotive fuel tank crashworthiness. A series of impact tests is : being conducted to measure fuel tank deformation under two : type...

Fast reactor fuel failures and steam generator leaks: Transient and accident analysis approaches

International Nuclear Information System (INIS)

1996-10-01

This report consists of a survey of activities on transient and accident analysis for the LMFR. It is focused on the following subjects: Fuel transient tests and analyses in hypothetical incident/accident situations; sodium-water interaction in steam generators, and sodium fires: test and analyses. There are also sections dealing with the experimental and analytical studies of: fuel subassembly failures; sodium boiling, molten fuel-coolant interaction; molten material movement and relocation in fuel bundles; heat removal after an accident or incident; sodium-water reaction in steam generator; steam generator protection systems; sodium-water contact in steam generator building; fire-fighting methods and systems to deal with sodium fires. Refs, figs, tabs

Results of VVER fuel rods tests in the MIR.M1 reactor under power cycling conditions

International Nuclear Information System (INIS)

Burukin, A.; Izhutov, A.; Ovchinnikov, V.; Kalygin, V.; Markov, D.; Pimenov, Y.; Novikov, V.; Medvedev, A.; Nesterov, B.

2011-01-01

The paper presents the main results of the 50 ... 60 MWd/kgU burnup VVER fuel rods tests performed in the MIR.M1 reactor loop facilities under power cycling. The non-destructive PIE results are presented as well. A series of experiments was performed, including overall measurement of fuel rod parameters test, in one of which 300 cycles were done. Irradiation under power cycling conditions and PIE of high-burnup VVER fuel rods showed the following: 1) all fuel rods claddings preserved their integrity under irradiation at linear heat rate (LHR) higher than the NPP operating one; 2) experimental data were obtained on the axial and radial cladding strain and fission gas release (FGR) from 50 ... 60 MWd/kgU burnup VVER-440 and VVER-1000 fuel rods as well as on the kinetics of the change in these parameters and fuel temperature under the power cycling; 3) non-destructive PIE results are in a satisfactory correlation with the data obtained by means of in-pile measurement gages during irradiation. (authors)

Results of tests under normal and abnormal operating conditions concerning LMFBR fuel element behaviour

International Nuclear Information System (INIS)

Languille, A.; Bergeonneau, P.; Essig, C.; Guerin, Y.

1985-04-01

The objective of this paper is to improve the knowledge on LMFBR fuel element behaviour during protected and unprotected transients in RAPSODIE and PHENIX reactors in order to evaluate its reliability. The range of the tests performed in these reactors is sufficiently large to cover normal and also extreme off normal conditions such as fuel melting. Results of such tests allow to better establish transient design limits for reactor structural components in particular for fuel pin cladding which play a lead role in controlling the accident sequence. Three main topics are emphasized in this paper: fuel melting during slow over-power excursions; influence of the fuel element geometrical evolution on reactivity feedback effects and reactor dynamic behaviour; clad damage evaluation during a transient (essentially very severe loss of flow)

Outline of results of safety research (in nuclear fuel cycle field in fiscal year 1996)

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-01

The safety research in Power Reactor and Nuclear Fuel Development Corporation in fiscal year 1996 has been carried out based on the basic plan of safety research (from fiscal year 1996 to 2000) which was decided in March, 1996. In this report, on nuclear fuel cycle field, namely all the subjects in the fields of nuclear fuel facilities, environmental radioactivity and waste disposal, and the subjects related to nuclear fuel facilities among the fields of aseismatic and probabilistic safety assessments, the results of research in fiscal year 1996, the first year of the 5-year project, are summarized together with the outline of the basic plan of safety research. The basic policy, objective and system for promotion of the safety research are described. The objectives of the safety research are the advancement of safety technology, the safety of facilities, stable operation techniques, the safety design and the evaluation techniques of next generation facilities, and the support of transferring nuclear fuel cycle to private businesses. The objects of the research are uranium enrichment, fuel fabrication and reprocessing, and waste treatment and storage. 52 investigation papers of the results of the safety research in nuclear fuel cycle field in fiscal year 1996 are collected in this report. (K.I.)

Advanced fuel cycle cost estimation model and its cost estimation results for three nuclear fuel cycles using a dynamic model in Korea

Energy Technology Data Exchange (ETDEWEB)

Kim, Sungki, E-mail: sgkim1@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Ko, Wonil [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Youn, Saerom; Gao, Ruxing [University of Science and Technology, 217 Gajungro, Yuseong-gu, Daejeon 305-350 (Korea, Republic of); Bang, Sungsig, E-mail: ssbang@kaist.ac.kr [Korea Advanced Institute of Science and Technology, Department of Business and Technology Management, 291 Deahak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

2015-11-15

Highlights: • The nuclear fuel cycle cost using a new cost estimation model was analyzed. • The material flows of three nuclear fuel cycle options were calculated. • The generation cost of once-through was estimated to be 66.88 mills/kW h. • The generation cost of pyro-SFR recycling was estimated to be 78.06 mills/kW h. • The reactor cost was identified as the main cost driver of pyro-SFR recycling. - Abstract: The present study analyzes advanced nuclear fuel cycle cost estimation models such as the different discount rate model and its cost estimation results. To do so, an analysis of the nuclear fuel cycle cost of three options (direct disposal (once through), PWR–MOX (Mixed OXide fuel), and Pyro-SFR (Sodium-cooled Fast Reactor)) from the viewpoint of economic sense, focusing on the cost estimation model, was conducted using a dynamic model. From an analysis of the fuel cycle cost estimation results, it was found that some cost gap exists between the traditional same discount rate model and the advanced different discount rate model. However, this gap does not change the priority of the nuclear fuel cycle option from the viewpoint of economics. In addition, the fuel cycle costs of OT (Once-Through) and Pyro-SFR recycling based on the most likely value using a probabilistic cost estimation except for reactor costs were calculated to be 8.75 mills/kW h and 8.30 mills/kW h, respectively. Namely, the Pyro-SFR recycling option was more economical than the direct disposal option. However, if the reactor cost is considered, the economic sense in the generation cost between the two options (direct disposal vs. Pyro-SFR recycling) can be changed because of the high reactor cost of an SFR.

Zero Emission Bay Area (ZEBA) Fuel Cell Bus Demonstration: Second Results Report

Energy Technology Data Exchange (ETDEWEB)

Eudy, L.; Chandler, K.

2012-07-01

This report presents results of a demonstration of 12 new fuel cell electric buses (FCEB) operating in Oakland, California. The 12 FCEBs operate as a part of the Zero Emission Bay Area (ZEBA) Demonstration, which also includes two new hydrogen fueling stations. This effort is the largest FCEB demonstration in the United States and involves five participating transit agencies. The ZEBA partners are collaborating with the U.S. Department of Energy (DOE) and DOE's National Renewable Energy Laboratory (NREL) to evaluate the buses in revenue service. The first results report was published in August 2011, describing operation of these new FCEBs from September 2010 through May 2011. New results in this report provide an update through April 2012.

Medication incidents reported to an online incident reporting system.

LENUS (Irish Health Repository)

Alrwisan, Adel

2011-01-15

AIMS: Approximately 20% of deaths from adverse events are related to medication incidents, costing the NHS an additional £500 million annually. Less than 5% of adverse events are reported. This study aims to assess the reporting rate of medication incidents in NHS facilities in the north east of Scotland, and to describe the types and outcomes of reported incidents among different services. Furthermore, we wished to quantify the proportion of reported incidents according to the reporters\\' profession. METHODS: A retrospective description was made of medication incidents reported to an online reporting system (DATIX) over a 46-month-period (July 2005 to April 2009). Reports originated from acute and community hospitals, mental health, and primary care facilities. RESULTS: Over the study period there were 2,666 incidents reported with a mean monthly reporting rate of 78.2\\/month (SD±16.9). 6.1% of all incidents resulted in harm, with insulin being the most commonly implicated medication. Nearly three-quarters (74.2%, n=1,978) of total incidents originated from acute hospitals. Administration incidents were implicated in the majority of the reported medication incidents (59%), followed by prescribing (10.8%) and dispensing (9.9%), while the nondescript "other medication incidents" accounted for 20.3% of total incidents. The majority of reports were made by nursing and midwifery staff (80%), with medical and dental professionals reporting the lowest number of incidents (n=56, 2%). CONCLUSIONS: The majority of medication incidents in this study were reported by nursing and midwifery staff, and were due to administration incidents. There is a clear need to elucidate the reasons for the limited contribution of the medical and dental professionals to reporting medication incidents.

Some Main Results of Commissioning of the Dalat Research Reactor with Low Enriched Fuel

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Luong Ba Vien; Pham Van Lam; Le Vinh Vinh; Huynh Ton Nghiem

2014-01-01

After completion of design calculation of the Dalat Nuclear Research Reactor (DNRR) for conversion from high-enriched uranium fuel (HEU) to low-enriched uranium (LEU) fuel, the commissioning programme for DNRR with entire core loaded with LEU fuel was successfully carried out from 24 November 2011 to 13 January 2012. The experimental results obtained during the implementation of commissioning programme showed a good agreement with design calculations and affirmed that the DNRR with LEU core have met all safety and exploiting requirements. (author)

Loading procedures for shipment of irradiated fuel

Energy Technology Data Exchange (ETDEWEB)

Bates, E F; Feltz, D E; Sandel, P S; Schoenbucher, B [Texas A and M University (United States)

1974-07-01

The Nuclear Science Center at Texas A and M does not have proper equipment and facilities for transferring irradiated fuel from the reactor pool to the transport vehicle. To accomplish the transfer of 23 MTR type fuel elements procedures were developed using a modified fork lift and flex-lift obtained locally. The transfer was accomplished without incident and with negligible personnel exposure. (author)

Loading procedures for shipment of irradiated fuel

International Nuclear Information System (INIS)

Bates, E.F.; Feltz, D.E.; Sandel, P.S.; Schoenbucher, B.

1974-01-01

The Nuclear Science Center at Texas A and M does not have proper equipment and facilities for transferring irradiated fuel from the reactor pool to the transport vehicle. To accomplish the transfer of 23 MTR type fuel elements procedures were developed using a modified fork lift and flex-lift obtained locally. The transfer was accomplished without incident and with negligible personnel exposure. (author)

Results of fuel elements fabrication on the basis of increased concentration dioxide fuel for research reactors

International Nuclear Information System (INIS)

Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

1996-01-01

According to the Russian Reduced Enrichment for Research and Test Reactors (RERTR) program, that were constructed under the Russian projects, at the Novosibirsk Chemical Concentrates Plant the pilot series of different configuration (WR-M2, MR, IRT-4M) fuel elements, based on increased concentration uranium dioxide fuel, have been fabricated for reactor tests. Comprehensive fabricated fuel elements quality estimation has been carried out. (author)

Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

2009-06-15

Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

FY04 Inspection Results for Wet Uruguay Fuel in L-Basin

Energy Technology Data Exchange (ETDEWEB)

VORMELKER, PHILIP

2005-09-01

The 2004 visual inspection of four Uruguay nuclear fuel assemblies stored in L-Basin was completed. This was the third inspection of this wet stored fuel since its arrival in the summer of 1998. Visual inspection photographs of the fuel from the previous and the recent inspections were compared and no evidence of significant corrosion was found on the individual fuel plate photographs. Fuel plates that showed areas of pitting in the cladding during the original receipt inspection were also identified during the 2004 inspection. However, a few pits were found on the non-fuel aluminum clamping plates that were not visible during the original and 2001 inspections.

Results of the production of wood derived fuels; Puupolttoaineiden tuotantotekniikka - tutkimusalueen katsaus

Energy Technology Data Exchange (ETDEWEB)

Korpilahti, A [Metsaeteho, Helsinki (Finland)

1997-12-31

During the year 1995 there were over 30 projects concerning the production of wood derived fuels going on. Nearly half of them focused on integrated production of pulp wood and wood fuel. About in ten projects work was carried out to promote wood fuel production from logging residues. Other topics were fire wood production, production logistics and wood fuel resources. For production of fuel chips from logging residues, a new chipper truck, MOHA-SISU, was introduced. Having ability to move on terrain, and equipped with drum chipper, hook technic for interchangeable containers and a trailer, the whole production chain can be carried out by the same machine. In Mikkeli region three years of active work promoted the usage of wood fuel in a district power plant to the level of over 110 000 cubic metres of fuel chips. The production costs tend to be a little high in average, and the production chain still needs to be improved. In the field of integrated production a great stride was taken when the first pilot plant using the MASSAHAKE-method started up. Components of the production line and knowledge to operate the process have increased resulting in good performance of the plant. And even another concept for integrated production was introduced. In order to fully control the debarking of small sized trees, a production line of chain flail equipment and debarking drum followed by a chipper and screening facilities was built up. Equipment and machines for harvesting young stands in a way that increases substantially the yield of energy component are still mostly first prototypes. The development of them into well functioning, efficient tools is the most important task in integrated production

Results of the production of wood derived fuels; Puupolttoaineiden tuotantotekniikka - tutkimusalueen katsaus

Energy Technology Data Exchange (ETDEWEB)

Korpilahti, A. [Metsaeteho, Helsinki (Finland)

1996-12-31

During the year 1995 there were over 30 projects concerning the production of wood derived fuels going on. Nearly half of them focused on integrated production of pulp wood and wood fuel. About in ten projects work was carried out to promote wood fuel production from logging residues. Other topics were fire wood production, production logistics and wood fuel resources. For production of fuel chips from logging residues, a new chipper truck, MOHA-SISU, was introduced. Having ability to move on terrain, and equipped with drum chipper, hook technic for interchangeable containers and a trailer, the whole production chain can be carried out by the same machine. In Mikkeli region three years of active work promoted the usage of wood fuel in a district power plant to the level of over 110 000 cubic metres of fuel chips. The production costs tend to be a little high in average, and the production chain still needs to be improved. In the field of integrated production a great stride was taken when the first pilot plant using the MASSAHAKE-method started up. Components of the production line and knowledge to operate the process have increased resulting in good performance of the plant. And even another concept for integrated production was introduced. In order to fully control the debarking of small sized trees, a production line of chain flail equipment and debarking drum followed by a chipper and screening facilities was built up. Equipment and machines for harvesting young stands in a way that increases substantially the yield of energy component are still mostly first prototypes. The development of them into well functioning, efficient tools is the most important task in integrated production

Laser assisted decontamination of nuclear fuel elements

International Nuclear Information System (INIS)

Padma Nilaya, J.; Biswas, Dhruba J.; Kumar, Aniruddha

2010-04-01

Laser assisted removal of loosely bound fuel particulates from the clad surface following the process of pellet loading has decided advantages over conventional methods. It is a dry and noncontact process that generates very little secondary waste and can occur inside a glove box without any manual interference minimizing the possibility of exposure to personnel. The rapid rise of the substrate/ particulate temperature owing to the absorption of energy from the incident laser pulse results in a variety of processes that may lead to the expulsion of the particulates. As a precursor to the cleaning of the fuel elements, initial experiments were carried out on contamination simulated on commonly used clad surfaces to gain a first hand experience on the various laser parameters for which as efficient cleaning can be obtained without altering the properties of the clad surface. The cleaning of a dummy fuel element was subsequently achieved in the laboratory by integrating the laser with a work station that imparted simultaneous rotational and linear motion to the fuel element. (author)

Materials for solar fuels and chemicals.

Science.gov (United States)

Montoya, Joseph H; Seitz, Linsey C; Chakthranont, Pongkarn; Vojvodic, Aleksandra; Jaramillo, Thomas F; Nørskov, Jens K

2016-12-20

The conversion of sunlight into fuels and chemicals is an attractive prospect for the storage of renewable energy, and photoelectrocatalytic technologies represent a pathway by which solar fuels might be realized. However, there are numerous scientific challenges in developing these technologies. These include finding suitable materials for the absorption of incident photons, developing more efficient catalysts for both water splitting and the production of fuels, and understanding how interfaces between catalysts, photoabsorbers and electrolytes can be designed to minimize losses and resist degradation. In this Review, we highlight recent milestones in these areas and some key scientific challenges remaining between the current state of the art and a technology that can effectively convert sunlight into fuels and chemicals.

Drying results of K-Basin fuel element 3128W (run 2)

International Nuclear Information System (INIS)

Abrefah, J.; Klinger, G.S.; Oliver, B.M.; Marshman, S.C.; MacFarlan, P.J.; Ritter, G.A.; Flament, T.A.

1998-07-01

An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-East Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of N-Reactor spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from an open K-East canister (3128W) during the first fuel selection campaign conducted in 1995, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. Although it was judged to be breached during in-basin (i.e., K-Basin) examinations, visual inspection of this fuel element in the hot cell indicated that it was likely intact. Some scratches on the coating covering the cladding were identified before the furnace test. The drying test was conducted in the Whole Element Furnace Testing System located in G-Cell within the PTL. This test system is composed of three basic systems: the in-cell furnace equipment, the system gas loop, and the analytical instrument package. Element 3128W was subjected to the drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step. Results of the Pressure Rise and Gas Evolution Tests suggest that most of the free water in the system was released during the extended CVD cycle (68 hr versus 8 hr for the first run). An additional ∼0.34 g of water was released during the subsequent HVD phase, characterized by multiple water release peaks, with a principle peak at ∼180 C. This additional water is attributed to decomposition of a uranium hydrate (UO 4 ·4H 2 O/UO 4 ·2H 2 O) coating that was observed to be covering the surface of the fuel element to a thickness of ∼1.6 mg/cm 2 . A

Zero Emission Bay Area (ZEBA) Fuel Cell Bus Demonstration Results: Third Report

Energy Technology Data Exchange (ETDEWEB)

Eudy, L.; Post, M.

2014-05-01

This report presents results of a demonstration of 12 fuel cell electric buses (FCEB) operating in Oakland, California. The 12 FCEBs operate as a part of the Zero Emission Bay Area (ZEBA) Demonstration, which also includes two new hydrogen fueling stations. This effort is the largest FCEB demonstration in the United States and involves five participating transit agencies. The ZEBA partners are collaborating with the U.S. Department of Energy (DOE) and DOE's National Renewable Energy Laboratory (NREL) to evaluate the buses in revenue service. NREL has published two previous reports, in August 2011 and July 2012, describing operation of these buses. New results in this report provide an update covering eight months through October 2013.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-09-30

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

Statement of nuclear incidents at nuclear installations: first quarter 1994

International Nuclear Information System (INIS)

1994-06-01

As a requirement of the 1974 Health and Safety at Work Act, the United Kingdom Health and Safety Inspectorate is required to publish reports of nuclear incidents at nuclear installations. This report covers the period 1st January to 31st March 1994. Two incidents are reported for Dungeness A Power Station, one at the Amersham International building on site at Harwell Laboratory, one at AEA Technology's Windscale Plant and one at British Nuclear Fuels Limited's Sellafield site. (UK)

RESULTS OF AIRPLANE TU-204-300 OPERATED BY "VLADIVOSTOK AVIA" COMPANY FUEL FLOW MONITIRING

Directory of Open Access Journals (Sweden)

G. E. Maslennikova

2014-01-01

Full Text Available This article describes the results obtained from continuous monitoring of fuel flow on the airplanes Tu-204-300, operated by the aircompany "Vladivostok Avia". The reasons for the change in the cost of each copy of airplane and changes in fuel characteristics due to pilots manner of flying.

Results of a recent crud/corrosion fuel risk assessment at a U.S. PWR

International Nuclear Information System (INIS)

Lamanna, Larry; Pop, Mike; Gregorich, Carola; Harne, Richard; Jones, John

2012-09-01

In order to avoid potential fuel reliability issues, specifically crud-related issues, it is necessary to achieve and maintain a crud safe environment. Therefore, the ability to confidently predict risks associated with crud deposition on fuel becomes critically important. AREVA is applying its cutting-edge PWR Fuel Crud (Primary System corrosion products)/Corrosion Tools, i.e. COBRA-FLX (subchannel-by-subchannel T/H tool) coupled with FDIC (crud deposition tool) to subsequently perform PWR Fuel Crud /Corrosion risk assessments for operating plants in the US. After describing the method, the result of one of these assessments is presented for an operating plant in the US that has experienced recent crud observations/concerns. Both Crud Induced Localized Corrosion (CILC) and Crud Induced Power Shift (CIPS) risk assessment methods, as applied to the upcoming cycle (Cycle N), were compared to the current/on-going cycle (Cycle N-1) and to the previous cycle (Cycle N-2). The results allowed the Utility to consider crud risk management changes associated with the upcoming cycle (Cycle-N). Benchmarking of the AREVA tools, using the plant-specific crud information gained from the crud sampling/characterization for the Unit will be presented. The CIPS analysis references boron loading and the amount of insoluble iron-nickel-borates predicted for Cycles N-2, N-1, and N. The results of the CILC evaluation reference FDIC-predicted crud thickness, cladding temperature under deposit, evolution of CILC bearing species and lithium concentration in the zirconium oxide layer. The approach taken by AREVA during the evaluation was to consider both 'risk' and 'margin' to fuel performance impact caused by crud deposits. The conclusion of the assessment, illustrated by the results presented in this paper, is that the example Plant has sufficient margin in worst case conditions for CIPS and CILC risk in Cycle N, based on Cycle N-1 and Cycle N-2 conditions and behavior

Chemical analytical considerations on the determination of burnup in irradiated nuclear fuels

International Nuclear Information System (INIS)

Cretella, R.F.; Servant, R.E.

1989-01-01

Burnup in an irradiated nuclear fuel may be defined as the energy produced per mass unit, from the time the fuel is introduced into the reactor and until a given moment. It is usually shown in megawatt/day or megawatt/hour generated per ton or kilo of fuel. It is also indicated as the number of fission produced per volume unit (cm 3 ) or per every 100 initial fissionable atoms. The yield of a power plant is directly related to the burnup of its fuel load and knowing the latter contributes to optimizing the economy in reactor operation and the related technologies. The development of nuclear fuels and the operation of reactors require doing with exact and accurate methods allowing to know the burnup. Errors in this measurement have an incidence upon the fuel design, the physical and nuclear calculations, the shielding requirements, the design of vehicles for the transportation of irradiated fuels, the engineering of processing plants, etc. All these factors, in turn, have an incidence upon the cost of nuclear power generation. (Author) [es

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

International Nuclear Information System (INIS)

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

2000-01-01

of the assessment of the safety criteria, the TF members looked at various issues, as they relate to one or more criteria, that have become of special interest. These topics included high burnup, core management, MOX, mixed cores, incomplete control rod insertion, and axial offset anomaly. Also, an attempt was made to assess the current level of methods and codes, which are used to verify the criteria and margins. As code development activities are widespread, the Task Force could not identify all such activities but focused on those needed to adequately analyse the effects of new design elements. The Task Force did not extensively review all ongoing and future research programs. However, a few examples of research programs are given that contribute to investigating the phenomena and mechanisms of fuel behavior under transient / accident conditions. These include hot-cell testing at Argonne National Laboratory in the U.S., the Halden Reactor Project in Norway, research and development work by Belgonucleaire in Belgium, the Cabri test reactor and related programs in France, and the Nuclear Safety Research Reactor program in Japan. As a result of all above assessments, the Task Force considers that the current framework of fuel safety criteria remains generally applicable, being largely unaffected by the 'new' or modern design elements; the levels (numbers) in the individual safety criteria may, however, change in accordance with the particular fuel and core design features. Some of these levels have already been - or are continuously being - adjusted; level adjustments of several other criteria (RIA, LOCA) also appear to be needed, on the basis of experimental data and the analysis thereof. For this (re)assessment of fuel safety criteria, the following process is recommended: (a) Continue to develop best-estimate (nominal) analysis methods, together with a suitable uncertainty analysis, in all areas of safety analysis (b) Continue to perform experimental verification

Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

International Nuclear Information System (INIS)

Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.

1997-10-01

The Solution High-Energy Burst Assembly (SHEBA) was originally constructed during 1980 and was designed to be a clean free-field geometry, right-circular, cylindrically symmetric critical assembly employing U(5%)O 2 F 2 solution as fuel. A second version of SHEBA, employing the same fuel but equipped with a fuel pump and shielding pit, was commissioned in 1993. This report includes data and operating experience for the 1993 SHEBA only. Solution-fueled benchmark work focused on the development of experimental measurements of the characterization of SHEBA; a summary of the results are given. A description of the system and the experimental results are given in some detail in the report. Experiments were designed to: (1) study the behavior of nuclear excursions in a low-enrichment solution, (2) evaluate accidental criticality alarm detectors for fuel-processing facilities, (3) provide radiation spectra and dose measurements to benchmark radiation transport calculations on a low-enrichment solution system similar to centrifuge enrichment plants, and (4) provide radiation fields to calibrate personnel dosimetry. 15 refs., 37 figs., 10 tabs

Occupational safety data and casualty rates for the uranium fuel cycle

International Nuclear Information System (INIS)

O'Donnell, F.R.; Hoy, H.C.

1981-10-01

Occupational casualty (injuries, illnesses, fatalities, and lost workdays) and production data are presented and used to calculate occupational casualty incidence rates for technologies that make up the uranium fuel cycle, including: mining, milling, conversion, and enrichment of uranium; fabrication of reactor fuel; transportation of uranium and fuel elements; generation of electric power; and transmission of electric power. Each technology is treated in a separate chapter. All data sources are referenced. All steps used to calculate normalized occupational casualty incidence rates from the data are presented. Rates given include fatalities, serious cases, and lost workdays per 100 man-years worked, per 10 12 Btu of energy output, and per other appropriate units of output

Experimental results with hydrogen fueled internal combustion engines

Science.gov (United States)

De Boer, P. C. T.; Mclean, W. J.; Homan, H. S.

1975-01-01

The paper focuses on the most important experimental findings for hydrogen-fueled internal combustion engines, with particular reference to the application of these findings to the assessment of the potential of hydrogen engines. Emphasis is on the various tradeoffs that can be made, such as between maximum efficiency, maximum power, and minimum NO emissions. The various possibilities for induction and ignition are described. Some projections are made about areas in which hydrogen engines may find their initial application and about optimum ways to design such engines. It is shown that hydrogen-fueled reciprocal internal combustion engines offer important advantages with respect to thermal efficiency and exhaust emissions. Problems arising from preignition can suitably be avoided by restricting the fuel-air equivalence ratio to values below about 0.5. The direct cylinder injection appears to be a very attractive way to operate the engine, because it combines a wide range of possible power outputs with a high thermal efficiency and very low NO emissions at part loads.

Main results of post-irradiation examinations of new-generation fuel assemblies VVER-1000

International Nuclear Information System (INIS)

Zvir, E.; Markov, D.; Polenok, V.; Zhitelev, V.; Kobylyansky, G.

2009-01-01

To increase the competitiveness of Russian nuclear fuel at the foreign market and to improve its technical and economic performance in order to provide a necessary level of safety, it is necessary to solve certain important tasks: Increase of fuel burn-up; Extension of operational lifetime of fuel assemblies and operational reliability of nuclear fuel; Introduction of cost-beneficial and flexible fuel cycles. Alternative fuel assemblies TVSA VVER-1000 and TVS-2 are used as a basis to optimize the nuclear fuel and develop advanced fuel cycles for nuclear power plants with VVER-1000 reactor types. Four fuel assemblies TVSA operated during 1 and up to 6 reactor cycles, reference fuel assembly TVS-2 operated during three reactor cycles and achieved an average fuel burnup of 48MW·day/kgU as well as failed fuel assembly TVS-2 operated during one cycle were examined at RIAR in recent years. The main objectives of these examinations were to obtain experimental data in support of operational integrity of products or to find out reasons of their failure. The performed post-irradiation examinations confirmed the operational integrity of alternative fuel assemblies TVSA including their geometrical stability up to the average fuel burnup of 55 MW·day/kgU over the fuel assembly (FA) (up to the maximal fuel burnup of ∼73 MW·day/kgU in fuel rods) and of TVS-2 up to the average fuel burnup of 48 MW·day/kgU over the fuel assembly. The changes introduced in the design of VVER-1000 fuel assembly during the development of alternative fuel assembly TVSA and TVS-2 did not make any negative effect on fuel rods. It was proved that causes of fuel rod failure were not related to design features of fuel assemblies. The design features and operating conditions of fuel assemblies under examinations are briefly described. Post-irradiation examinations proved the geometrical stability of fuel assemblies TVSA and TVS-2 under operation up to the fuel burnup of ∼50 MW day/kgU, as for the

A Profile of Criminal Incidents at School: Results from the 2003-05 National Crime Victimization Survey Crime Incident Report NCES 2010-318

Science.gov (United States)

Ruddy, Sally A.; Bauer, Lynn; Neiman, Samantha

2010-01-01

This report provides estimates of criminal incidents that occur at school. Incident-level data were obtained from the National Crime Victimization Survey (NCVS), the nation's primary source of information on criminal victimization and criminal incidents in the United States. The NCVS collects demographic information on respondents in the NCVS…

Geographical variation in cardiovascular incidence: results from the British Women's Heart and Health Study

Directory of Open Access Journals (Sweden)

Ebrahim Shah

2010-11-01

Full Text Available Abstract Background Prevalence of cardiovascular disease (CVD in women shows regional variations not explained by common risk factors. Analysis of CVD incidence will provide insight into whether there is further divergence between regions with increasing age. Methods Seven-year follow-up data on 2685 women aged 59-80 (mean 69 at baseline from 23 towns in the UK were available from the British Women's Heart and Health Study. Time to fatal or non-fatal CVD was analyzed using Cox regression with adjustment for risk factors, using multiple imputation for missing values. Results Compared to South England, CVD incidence is similar in North England (HR 1.05 (95% CI 0.84, 1.31 and Scotland (0.93 (0.68, 1.27, but lower in Midlands/Wales (0.85 (0.64, 1.12. Event severity influenced regional variation, with South England showing lower fatal incident CVD than other regions, but higher non-fatal incident CVD. Kaplan-Meier plots suggested that regional divergence in CVD occurred before baseline (before mean baseline age of 69. Conclusions In women, regional differences in CVD early in adult life do not further diverge in later life. This may be due to regional differences in early detection, survivorship of women entering the study, or event severity. Targeting health care resources for CVD by geographic variation may not be appropriate for older age-groups.

Effects of sex on the incidence and prognosis of spinal meningiomas: a Surveillance, Epidemiology, and End Results study.

Science.gov (United States)

Westwick, Harrison J; Shamji, Mohammed F

2015-09-01

Most spinal meningiomas are intradural lesions in the thoracic spine that present with both local pain and myelopathy. By using the large prospective Surveillance, Epidemiology, and End Results (SEER) database, the authors studied the incidence of spinal meningiomas and examined demographic and treatment factors predictive of death. Using SEER*Stat software, the authors queried the SEER database for cases of spinal meningioma between 2000 and 2010. From the results, tumor incidence and demographic statistics were computed; incidence was analyzed as a function of tumor location, pathology, age, sex, and malignancy code. Survival was analyzed by using a Cox proportional hazards ratio in SPSS for age, sex, marital status, primary site, size quartile, treatment modality, and malignancy code. In this analysis, significance was set at a p value of 0.05. The 1709 spinal meningiomas reported in the SEER database represented 30.7% of all primary intradural spinal tumors and 7.9% of all meningiomas. These meningiomas occurred at an age-adjusted incidence of 0.193 (95% CI 0.183-0.202) per 100,000 population and were closely related to sex (337 [19.7%] male patients and 1372 [80.3%] female patients). The Cox hazard function for mortality in males was higher (2.4 [95% CI1.7-3.5]) and statistically significant, despite the lower lesion incidence in males. All-cause survival was lowest in patients older than 80 years. Primary site and treatment modality were not significant predictors of mortality. Spinal meningiomas represent a significant fraction of all primary intradural spinal tumors and of all meningiomas. The results of this study establish the association of lesion incidence and survival with sex, with a less frequent incidence in but greater mortality among males.

Drying results of K-Basin fuel element 0309M (Run 3)

International Nuclear Information System (INIS)

Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01

An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-West Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from K-West canister 0309M during the second fuel selection campaign, conducted in 1996, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. The fuel element was broken in two pieces, with a relatively clean fracture, and the larger piece was tested. A gray/white coating was observed. This was the first test of a damaged fuel element in the furnace. K-West canisters can hold up to seven complete fuel assemblies, but, for purposes of this report, the element tested here is designated as Element 0309M. Element 0309M was subjected to drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step

Comparison of the Transportation Risks Resulting from Accidents during the Transportation of the Spent Fuel

International Nuclear Information System (INIS)

Jeong Jong Tae; Cho, Dong Kuen; Choi, Heui Joo; Choi, Jong Won

2007-01-01

The safe, environmentally sound and publicly acceptable disposal of high level wastes and spent fuels is becoming a very important issue. The operational safety assessment of a repository including a transportation safety assessment is a fundamental part in order to achieve this goal. According to the long term management strategy for spent fuels in Korea, they will be transported from the spent fuel pools in each nuclear power plant to the central interim storage facility (CISF) which is to start operation in 2016. Therefore, we have to determine the safe and economical logistics for the transportation of these spent fuels by considering their transportation risks and costs. In this study, we developed four transportation scenarios by considering the type of transportation casks and transport means in order to suggest safe and economical transportation logistics for spent fuels. Also, we estimated and compared the transportation risks resulting from the accidents during the transportation of spent fuels for these four transportation scenarios

The Test for Flow Characteristics of Tubular Fuel Assembly(II) - Experimental results and CFD analysis

International Nuclear Information System (INIS)

Park, Jong Hark; Chae, H. T.; Park, C.; Kim, H.

2006-12-01

A test facility had been established for the experiment of velocity distribution and pressure drop in a tubular fuel. A basic test had been conducted to examine the performance of the test loop and to verify the accuracy of measurement by pitot-tube. In this report, test results and CFD analysis for the hydraulic characteristics of a tubular fuel, following the previous tests, are described. Coolant velocities in all channels were measured using pitot-tube and the effect of flow rate change on the velocity distribution was also examined. The pressure drop through the tubular fuel was measured for various flow rates in range of 1 kg/s to 21 kg/s to obtain a correlation of pressure drop with variation of flow rate. In addition, a CFD(Computational Fluid Dynamics) analysis was also done to find out the hydraulic characteristics of tubular fuel such as velocity distribution and pressure drop. As the results of CFD analysis can give us a detail insight on coolant flow in the tubular fuel, the CFD method is a very useful tool to understand the flow structure and phenomena induced by fluid flow. The CFX-10, a commercial CFD code, was used in this study. The two results by the experiment and the CFD analysis were investigated and compared with each other. Overall trend of velocity distribution by CFD analysis was somewhat different from that of experiment, but it would be reasonable considering measurement uncertainties. The CFD prediction for pressure drop of a tubular fuel shows a tolerably good agreement with experiment within 8% difference

Chemical incidents resulted in hazardous substances releases in the context of human health hazards

Directory of Open Access Journals (Sweden)

Anna Pałaszewska-Tkacz

2017-02-01

Full Text Available Objectives: The research purpose was to analyze data concerning chemical incidents in Poland collected in 1999–2009 in terms of health hazards. Material and Methods: The data was obtained, using multimodal information technology (IT system, from chemical incidents reports prepared by rescuers at the scene. The final analysis covered sudden events associated with uncontrolled release of hazardous chemical substances or mixtures, which may potentially lead to human exposure. Releases of unidentified substances where emergency services took action to protect human health or environment were also included. Results: The number of analyzed chemical incidents in 1999–2009 was 2930 with more than 200 different substances released. The substances were classified into 13 groups of substances and mixtures posing analogous risks. Most common releases were connected with non-flammable corrosive liquids, including: hydrochloric acid (199 cases, sulfuric(VI acid (131 cases, sodium and potassium hydroxides (69 cases, ammonia solution (52 cases and butyric acid (32 cases. The next group were gases hazardous only due to physico-chemical properties, including: extremely flammable propane-butane (249 cases and methane (79 cases. There was no statistically significant trend associated with the total number of incidents. Only with the number of incidents with flammable corrosive, toxic and/or harmful liquids, the regression analysis revealed a statistically significant downward trend. The number of victims reported was 1997, including 1092 children and 18 fatalities. Conclusions: The number of people injured, number of incidents and the high 9th place of Poland in terms of the number of Seveso establishments, and 4 times higher number of hazardous industrial establishments not covered by the Seveso Directive justify the need for systematic analysis of hazards and their proper identification. It is advisable enhance health risk assessment, both qualitative and

Storage of LWR spent fuel in air. Volume 3, Results from exposure of spent fuel to fluorine-contaminated air

Energy Technology Data Exchange (ETDEWEB)

Cunningham, M.E.; Thomas, L.E.

1995-06-01

The Behavior of Spent Fuel in Storage (BSFS) Project has conducted research to develop data on spent nuclear fuel (irradiated U0{sub 2}) that could be used to support design, licensing, and operation of dry storage installations. Test Series B conducted by the BSFS Project was designed as a long-term study of the oxidation of spent fuel exposed to air. It was discovered after the exposures were completed in September 1990 that the test specimens had been exposed to an atmosphere of bottled air contaminated with an unknown quantity of fluorine. This exposure resulted in the test specimens reacting with both the oxygen and the fluorine in the oven atmospheres. The apparent source of the fluorine was gamma radiation-induced chemical decomposition of the fluoro-elastomer gaskets used to seal the oven doors. This chemical decomposition apparently released hydrofluoric acid (HF) vapor into the oven atmospheres. Because the Test Series B specimens were exposed to a fluorine-contaminated oven atmosphere and reacted with the fluorine, it is recommended that the Test Series B data not be used to develop time-temperature limits for exposure of spent nuclear fuel to air. This report has been prepared to document Test Series B and present the collected data and observations.

Storage of LWR spent fuel in air. Volume 3, Results from exposure of spent fuel to fluorine-contaminated air

International Nuclear Information System (INIS)

Cunningham, M.E.; Thomas, L.E.

1995-06-01

The Behavior of Spent Fuel in Storage (BSFS) Project has conducted research to develop data on spent nuclear fuel (irradiated U0 2 ) that could be used to support design, licensing, and operation of dry storage installations. Test Series B conducted by the BSFS Project was designed as a long-term study of the oxidation of spent fuel exposed to air. It was discovered after the exposures were completed in September 1990 that the test specimens had been exposed to an atmosphere of bottled air contaminated with an unknown quantity of fluorine. This exposure resulted in the test specimens reacting with both the oxygen and the fluorine in the oven atmospheres. The apparent source of the fluorine was gamma radiation-induced chemical decomposition of the fluoro-elastomer gaskets used to seal the oven doors. This chemical decomposition apparently released hydrofluoric acid (HF) vapor into the oven atmospheres. Because the Test Series B specimens were exposed to a fluorine-contaminated oven atmosphere and reacted with the fluorine, it is recommended that the Test Series B data not be used to develop time-temperature limits for exposure of spent nuclear fuel to air. This report has been prepared to document Test Series B and present the collected data and observations

Critical incidence reporting systems - an option in equine anaesthesia? Results from a panel meeting.

Science.gov (United States)

Hartnack, Sonja; Bettschart-Wolfensberger, Regula; Driessen, Bernd; Pang, Daniel; Wohlfender, Franziska

2013-11-01

To provide a brief introduction into Critical Incident Reporting Systems (CIRS) as used in human medicine, and to report the discussion from a recent panel meeting discussion with 23 equine anaesthetists in preparation for a new CEPEF-4 (Confidential Enquiry into Perioperative Equine Fatalities) study. Moderated group discussions, and review of literature. The first group discussion focused on the definition of 'preventable critical incidents' and/or 'near misses' in the context of equine anaesthesia. The second group discussion focused on categorizing critical incidents according to an established framework for analysing risk and safety in clinical medicine. While critical incidents do occur in equine anaesthesia, no critical incident reporting system including systematic collection and analysis of critical incidents is in place. Critical incident reporting systems could be used to improve safety in equine anaesthesia - in addition to other study types such as mortality studies. © 2013 Association of Veterinary Anaesthetists and the American College of Veterinary Anesthesia and Analgesia.

Incidence of dementia in elderly Latin Americans: Results of the Maracaibo Aging Study.

Science.gov (United States)

Maestre, Gladys E; Mena, Luis J; Melgarejo, Jesus D; Aguirre-Acevedo, Daniel C; Pino-Ramírez, Gloria; Urribarrí, Milady; Chacon, Inara J; Chávez, Carlos A; Falque-Madrid, Luis; Gaona, Ciro A; Terwilliger, Joseph D; Lee, Joseph H; Scarmeas, Nikolaos

2018-02-01

There are few longitudinal studies of dementia in developing countries. We used longitudinal data from the Maracaibo Aging Study to accurately determine the age- and sex-specific incidence of dementia in elderly Latin Americans. The Diagnostic and Statistical Manual of Mental Disorders (Fourth Edition, Text Revision) was used to diagnose dementia, which was classified as Alzheimer's disease, vascular dementia, or other. Age- and sex-specific incidence was estimated as the number of new cases of dementia divided by person-years (p-y) of follow-up. The incidence of all dementia diagnoses was 9.10 per 1000 p-y (95% confidence interval [CI] 7.13-11.44; 8026 total p-y), 5.18 for Alzheimer's disease (95% CI 3.72-7.03; 7916 total p-y), and 3.35 for vascular dementia (95% CI 2.19-4.91; 7757 total p-y). Among Maracaibo Aging Study participants younger than 65 years, the incidence of dementia was higher than that of US Whites. Among individuals older than 65 years, the incidence was comparable to the mean of previous incidence estimates for other populations worldwide. Copyright © 2017 the Alzheimer's Association. Published by Elsevier Inc. All rights reserved.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Performance Test Results for the Laser-Powered Microthruster

International Nuclear Information System (INIS)

Phipps, Claude R.; Luke, James R.; Helgeson, Wesley; Johnson, Richard

2006-01-01

Microthrusters are useful for orienting and repositioning small craft above the atmosphere. We report technical results obtained during a successful 5-year program to develop a commercially-viable laser-powered microthruster. Its main advantage is the ability to generate a broad thrust range under programmable electronic control with minimal electrical power. The device applies millisecond-duration diode-laser pulses to a fuel tape to produce an ablation jet. By employing laser-initiated energetic polymers in our ablation fuel tapes, we obtained momentum coupling coefficients as large as 3mN/W of incident laser power, giving a continuous thrust range from 50μN to 10mN. With our standard 30m x 8mm fuel tape, fueled thruster mass is 0.5kg and 50N-s lifetime impulse is achieved. With an order-of-magnitude greater fuel mass, the thruster could accomplish re-entry or substantial orbit-raising of a 10-kg microsatellite. In its usual configuration, specific impulse is 200 seconds, and ablation efficiency, the ratio of exhaust kinetic energy to incident laser optical energy is 180%. We compare performance of several laser-initiated micropropellants which we studied, including polyvinyl nitrate (PVN), glycidyl azide polymer (GAP), and nitrocellulose (NC). All were doped with a laser-absorbing component, either carbon nanopearls with 10nm mean diameter or dyes tuned to the 920-nm laser wavelength but transparent at visible wavelengths. Our demonstrated momentum coupling coefficient is sufficient to levitate a 0.15-kg object with a 500-W laser beam having appropriate characteristics

Analysis of environmental contamination resulting from catastrophic incidents: part 2. Building laboratory capability by selecting and developing analytical methodologies.

Science.gov (United States)

Magnuson, Matthew; Campisano, Romy; Griggs, John; Fitz-James, Schatzi; Hall, Kathy; Mapp, Latisha; Mullins, Marissa; Nichols, Tonya; Shah, Sanjiv; Silvestri, Erin; Smith, Terry; Willison, Stuart; Ernst, Hiba

2014-11-01

Catastrophic incidents can generate a large number of samples of analytically diverse types, including forensic, clinical, environmental, food, and others. Environmental samples include water, wastewater, soil, air, urban building and infrastructure materials, and surface residue. Such samples may arise not only from contamination from the incident but also from the multitude of activities surrounding the response to the incident, including decontamination. This document summarizes a range of activities to help build laboratory capability in preparation for sample analysis following a catastrophic incident, including selection and development of fit-for-purpose analytical methods for chemical, biological, and radiological contaminants. Fit-for-purpose methods are those which have been selected to meet project specific data quality objectives. For example, methods could be fit for screening contamination in the early phases of investigation of contamination incidents because they are rapid and easily implemented, but those same methods may not be fit for the purpose of remediating the environment to acceptable levels when a more sensitive method is required. While the exact data quality objectives defining fitness-for-purpose can vary with each incident, a governing principle of the method selection and development process for environmental remediation and recovery is based on achieving high throughput while maintaining high quality analytical results. This paper illustrates the result of applying this principle, in the form of a compendium of analytical methods for contaminants of interest. The compendium is based on experience with actual incidents, where appropriate and available. This paper also discusses efforts aimed at adaptation of existing methods to increase fitness-for-purpose and development of innovative methods when necessary. The contaminants of interest are primarily those potentially released through catastrophes resulting from malicious activity

Results of Am isotopic ratio analysis in irradiated MOX fuels

Energy Technology Data Exchange (ETDEWEB)

Koyama, Shin-ichi; Osaka, Masahiko; Mitsugashira, Toshiaki; Konno, Koichi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Kajitani, Mikio

1997-04-01

For analysis of a small quantity of americium, it is necessary to separate from curium which has similar chemical property. As a chemical separation method for americium and curium, the oxidation of americium with pentavalent bismuth and subsequent co-precipitation of trivalent curium with BIP O{sub 4} were applied to analyze americium in irradiated MOX fuels which contained about 30wt% plutonium and 0.9wt% {sup 241}Am before irradiation and were irradiated up to 26.2GWd/t in the experimental fast reactor Joyo. The purpose of this study is to measure isotopic ratio of americium and to evaluate the change of isotopic ratio with irradiation. Following results are obtained in this study. (1) The isotopic ratio of americium ({sup 241}Am, {sup 242m}Am and {sup 243}Am) can be analyzed in the MOX fuels by isolating americium. The isotopic ratio of {sup 242m}Am and {sup 243}Am increases up to 0.62at% and 0.82at% at maximum burnup, respectively, (2) The results of isotopic analysis indicates that the contents of {sup 241}Am decreases, whereas {sup 242m}Am, {sup 243}Am increase linearly with increasing burnup. (author)

Safety issues on advanced fuel

International Nuclear Information System (INIS)

Gross, H.; Krebs, W.D.

1998-01-01

In the recent years a general discussion has started whether unsolved safety issues are related to advanced fuel. Advanced fuel is in this context a summary of features like high burnup, improved clad materials, low leakage loading pattern with high peaking factors etc. The design basis accidents RIA and Loca are of special interest for this discussion. From the Siemens point of view RIA is not a safety issue. There are sufficient margins between the enthalpy rise calculated by modern 3D methods and the fuel failures which occurred in RIA simulation tests when the effect of pulse width is taken into account. The evaluation of possible uncertainties for the established Loca criteria (17% equivalent corrosion, 1200 C clad temperature) for high burnup makes sense. But fuel with high burnup has significantly lower peaking factors than fuel with lower burnup. This gives sufficient margin counterbalancing possible uncertainties. In contrast to the above incomplete control rod insertion at higher burnup is potentially a real safety issue. Although Siemens fuel was not affected by the reported incidents they addressed the problem and checked that they have sufficient design margin for their fuel. (orig.) [de

Causal Factors and Adverse Events of Aviation Accidents and Incidents Related to Integrated Vehicle Health Management

Science.gov (United States)

Reveley, Mary S.; Briggs, Jeffrey L.; Evans, Joni K.; Jones, Sharon M.; Kurtoglu, Tolga; Leone, Karen M.; Sandifer, Carl E.

2011-01-01

Causal factors in aviation accidents and incidents related to system/component failure/malfunction (SCFM) were examined for Federal Aviation Regulation Parts 121 and 135 operations to establish future requirements for the NASA Aviation Safety Program s Integrated Vehicle Health Management (IVHM) Project. Data analyzed includes National Transportation Safety Board (NSTB) accident data (1988 to 2003), Federal Aviation Administration (FAA) incident data (1988 to 2003), and Aviation Safety Reporting System (ASRS) incident data (1993 to 2008). Failure modes and effects analyses were examined to identify possible modes of SCFM. A table of potential adverse conditions was developed to help evaluate IVHM research technologies. Tables present details of specific SCFM for the incidents and accidents. Of the 370 NTSB accidents affected by SCFM, 48 percent involved the engine or fuel system, and 31 percent involved landing gear or hydraulic failure and malfunctions. A total of 35 percent of all SCFM accidents were caused by improper maintenance. Of the 7732 FAA database incidents affected by SCFM, 33 percent involved landing gear or hydraulics, and 33 percent involved the engine and fuel system. The most frequent SCFM found in ASRS were turbine engine, pressurization system, hydraulic main system, flight management system/flight management computer, and engine. Because the IVHM Project does not address maintenance issues, and landing gear and hydraulic systems accidents are usually not fatal, the focus of research should be those SCFMs that occur in the engine/fuel and flight control/structures systems as well as power systems.

Conventional fuel tank blunt impact tests : test and analysis results

Science.gov (United States)

2014-04-02

The Federal Railroad Administrations Office of Research : and Development is conducting research into fuel tank : crashworthiness. A series of impact tests are planned to : measure fuel tank deformation under two types of dynamic : loading conditi...

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

International Nuclear Information System (INIS)

Patrick, W.C.

1986-01-01

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

Issues and decisions for nuclear power plant management after fuel damage events

International Nuclear Information System (INIS)

1997-04-01

Experience has shown that the on-site activities following an incident that results in severely damaged fuel at a nuclear power plant required extraordinary effort. Even in cases that are not extreme but in which fuel damage is greater than mentioned in the specifications for operation, the recovery will require extensive work. This publication includes information from several projects at the IAEA since 1989 that have resulted in a Technical Report, a TECDOC and a Workshop. While the initial purpose of the projects was focused on providing technical information transfer to the experts engaged in recovery work at the damaged unit of Chernobyl NPP, the results have led to a general approach to managing events in which there is substantial fuel damage. This TECDOC summarizes the work to focus on management issues that may be encountered in any such event whether small or large. 11 refs, 2 figs, 5 tabs

Non-Flow-Through Fuel Cell System Test Results and Demonstration on the SCARAB Rover

Science.gov (United States)

Scheidegger, Brianne, T.; Burke, Kenneth A.; Jakupca, Ian J.

2012-01-01

This paper describes the results of the demonstration of a non-flow-through PEM fuel cell as part of a power system on the SCARAB rover. A 16-cell non-flow-through fuel cell stack from Infinity Fuel Cell and Hydrogen, Inc. was incorporated into a power system designed to act as a range extender by providing power to the rover s hotel loads. This work represents the first attempt at a ground demonstration of this new technology aboard a mobile test platform. Development and demonstration were supported by the Office of the Chief Technologist s Space Power Systems Project and the Advanced Exploration System Modular Power Systems Project.

Development of a recovery process of scraps resulting from the manufacture of metallic uranium fuels

International Nuclear Information System (INIS)

Camilo, Ruth L.; Kuada, Terezinha A.; Forbicini, Christina A.L.G.O.; Cohen, Victor H.; Araujo, Bertha F.; Lobao, Afonso S.T.

1996-01-01

The study of the dissolution of natural metallic uranium fuel samples with aluminium cladding is presented, in order to obtain optimized conditions for the system. The aluminium cladding was dissolved in an alkaline solution of Na OH/Na NO 3 and the metallic uranium with HNO 3 . A fumeless dissolution with total recovery of nitrous gases was achieved. The main purpose of this project was the recovery of uranium from scraps resulting from the manufacture of the metallic uranium fuel or other non specified fuels. (author)

Connecticut Transit (CTTRANSIT) Fuel Cell Transit Bus Preliminary Evaluation Results

Science.gov (United States)

2008-10-16

This report describes operations at Connecticut Transit (CTTRANSIT) in Hartford for one prototype fuel cell bus and three new diesel buses operating from the same location. The report discusses the planned fuel cell bus demonstration and equipment us...

Statement of nuclear incidents at nuclear installations

International Nuclear Information System (INIS)

1993-10-01

Three incidents were reported in April-June 1993. The first was on the British Nuclear Fuel plc (BNFL) site at Sellafield and concerned leakage of 0.5 TBq of alpha activity from plutonium contaminated waste stored in a steel drum. This was subsequently double contained and moved so it could be inspected regularly. No contamination of personnel occurred. The second concerned the leakage of thorium liquor from a pipe at the UKAEA's Thorium reprocessing plant at Dounreay. Two temporary repairs were made and no personnel were contaminated. The third was at the Sellafield site where a small quantity (5 mls) of plutonium containing liquor had leaked from a package and released alpha activity. The bags were temporary containment of engineering debris which may have had sharp edges. The bags had been piled up and one of the bags had torn. Recommendations were made following inquiries into each of the incidents to improve procedures and prevent similar incidents occurring. (UK)

Results from Cycles 1 and 2 of NNWSI Series 2 spent fuel dissolution tests

International Nuclear Information System (INIS)

Wilson, C.N.

1987-05-01

PWR spent fuel rod segments from the H.B. Robinson Unit 2 and Turkey Point Unit 3 reactors were leach tested in Nevada Nuclear Waste Storage Investigations (NNWSI) reference J-13 water under ambient hot cell conditions. The test matrix included bare fuel plus the cladding, rod segments with artificially induced cladding defects, and undefected rod segments. Radionuclide release results are presented and discussed. The actinides Pu, Am, Cm and Np appear to have been released congruently as the UO 2 oxide fuel matrix dissolved. Preferential U release measured in certain tests may be related to dissolution of oxidized UO/sub 2+x/ from the fuel surface, and/or greater solubility (and mobility) of U relative to the other actinides within defected cladding specimens. Uranium solubility measured in the J-13 water was much greater then that measured in deionized water in previous tests. All of the principal fission products analyzed ( 137 Cs, 129 I, 99 Tc and 90 Sr) were released preferentially relative to the actinides. Preferential release of activation product 14 C was also measured, with a portion of the 14 C release appearing to originate from the cladding exterior surface. Much greater fractional fuel dissolution appeared to have occurred with bare fuel particles than from fuel contained in defected cladding. Actinide release from test specimens containing small (∼200 μm) laser-drilled holes through the cladding was not significantly greater than from undefected specimens

GREET 1.5 - transportation fuel-cycle model - Vol. 1 : methodology, development, use, and results

International Nuclear Information System (INIS)

Wang, M. Q.

1999-01-01

This report documents the development and use of the most recent version (Version 1.5) of the Greenhouse Gases, Regulated Emissions, and Energy Use in Transportation (GREET) model. The model, developed in a spreadsheet format, estimates the full fuel-cycle emissions and energy associated with various transportation fuels and advanced vehicle technologies for light-duty vehicles. The model calculates fuel-cycle emissions of five criteria pollutants (volatile organic compounds, carbon monoxide, nitrogen oxides, particulate matter with diameters of 10 micrometers or less, and sulfur oxides) and three greenhouse gases (carbon dioxide, methane, and nitrous oxide). The model also calculates total energy consumption, fossil fuel consumption, and petroleum consumption when various transportation fuels are used. The GREET model includes the following cycles: petroleum to conventional gasoline, reformulated gasoline, conventional diesel, reformulated diesel, liquefied petroleum gas, and electricity via residual oil; natural gas to compressed natural gas, liquefied natural gas, liquefied petroleum gas, methanol, Fischer-Tropsch diesel, dimethyl ether, hydrogen, and electricity; coal to electricity; uranium to electricity; renewable energy (hydropower, solar energy, and wind) to electricity; corn, woody biomass, and herbaceous biomass to ethanol; soybeans to biodiesel; flared gas to methanol, dimethyl ether, and Fischer-Tropsch diesel; and landfill gases to methanol. This report also presents the results of the analysis of fuel-cycle energy use and emissions associated with alternative transportation fuels and advanced vehicle technologies to be applied to passenger cars and light-duty trucks

HTGR fuel reprocessing pilot plant: results of the sequential equipment operation

International Nuclear Information System (INIS)

Strand, J.B.; Fields, D.E.; Kergis, C.A.

1979-05-01

The second sequential operation of the HTGR fuel reprocessing cold-dry head-end pilot plant equipment has been successfully completed. Twenty standard LHGTR fuel elements were crushed to a size suitable for combustion in a fluid bed burner. The graphite was combusted leaving a product of fissile and fertile fuel particles. These particles were separated in a pneumatic classifier. The fissile particles were fractured and reburned in a fluid bed to remove the inner carbon coatings. The remaining products are ready for dissolution and solvent extraction fuel recovery

Zero Emission Bay Area (ZEBA) Fuel Cell Bus Demonstration Results: Sixth Report

Science.gov (United States)

2017-09-01

This report presents results of a demonstration of fuel cell electric buses (FCEBs) operating in Oakland, California. Alameda-Contra Costa Transit District (AC Transit) leads the Zero Emission Bay Area (ZEBA) demonstration that includes 13 advanced-d...

Incidence and Prognosis of Spinal Hemangioblastoma: A Surveillance Epidemiology and End Results Study.

Science.gov (United States)

Westwick, Harrison J; Giguère, Jean-François; Shamji, Mohammed F

2016-01-01

Intradural spinal hemangioblastoma are infrequent, vascular, pathologically benign tumors occurring either sporadically or in association with von Hippel-Lindau disease along the neural axis. Described in fewer than 1,000 cases, literature is variable with respect to epidemiological factors associated with spinal hemangioblastoma and their treatment. The objective of this study was to evaluate the epidemiology of intradural spinal hemangioblastoma with the Surveillance, Epidemiology and End Results (SEER) database while also presenting an illustrative case. The SEER database was queried for cases of spinal hemangioblastoma between 2000 and 2010 with the use of SEER*Stat software. Incidence was evaluated as a function of age, sex and race. Survival was evaluated with the Cox proportionate hazards ratio using IBM SPSS software evaluating age, sex, location, treatment modality, pathology and number of primaries (p = 0.05). Descriptive statistics of the same factors were also calculated. The case of a 43-year-old patient with a surgical upper cervical intramedullary hemangioblastoma is also presented. In the data set between 2000 and 2010, there were 133 cases with an age-adjusted incidence of 0.014 (0.012-0.017) per 100,000 to the standard USA population. Hemangioblastoma was the tenth most common intradural spinal tumor type representing 2.1% (133 of 6,156) of all spinal tumors. There was no difference in incidence between men and women with an female:male rate ratio of 1.05 (0.73-1.50) with p = 0.86. The average age of patients was 48.0 (45.2-50.9) years, and a lower incidence was noted in patients incidence amongst the different races. Treatment included surgical resection in 106 (79.7%) cases, radiation with surgery in 7 (5.3%) cases, and radiation alone was used in only 1 (0.8%) case, and no treatment was performed in 17 (12.8%) cases. Mortality was noted in 12 (9%) cases, and median survival of 27.5 months (range 1-66 months) over the 10-year period. Mortality

Influence of FGR complexity modelling on the practical results in gas pressure calculation of selected fuel elements from Dukovany NPP

International Nuclear Information System (INIS)

Lahodova, M.

2001-01-01

A modernization fuel system and advanced fuel for operation up to the high burnup are used in present time in Dukovany NPP. Reloading of the cores are evaluated using computer codes for thermomechanical behavior of the most loaded fuel rods. The paper presents results of parametric calculations performed by the NRI Rez integral code PIN, version 2000 (PIN2k) to assess influence of fission gas release modelling complexity on achieved results. The representative Dukovany NPP fuel rod irradiation history data are used and two cases of fuel parameter variables (soft and hard) are chosen for the comparison. Involved FGR models where the GASREL diffusion model developed in the NRI Rez plc and standard Weisman model that is recommended in the previous version of the PIN integral code. FGR calculation by PIN2k with GASREL model represents more realistic results than standard Weisman's model. Results for linear power, fuel centre temperature, FGR and gas pressure versus burnup are given for two fuel rods

Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix C, marine transport and associated environmental impacts. Volume 2

International Nuclear Information System (INIS)

1995-03-01

This is Appendix C to a Draft Environmental Statement on a Proposed Nuclear Weapon Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. Shipment of any material via ocean transport entails risks to both the ship's crew and the environment. The risks result directly from transportation-related accidents and, in the case of radioactive or other hazardous materials, also include exposure to the effects of the material itself. This appendix provides a description of the approach used to assess the risks associated with the transport of foreign research reactor spent nuclear fuel from a foreign port to a U.S. port(s) of entry. This appendix also includes a discussion of the shipping configuration of the foreign research reactor spent nuclear fuel, the possible types of vessels that could be used to make the shipments, the risk assessment methodology (addressing both incident-free and accident risks), and the results of the analyses. Analysis of activities in the port(s) is described in Appendix D. The incident-free and accident risk assessment results are presented in terms of the per shipment risk and total risks associated with the basic implementation of Management Alternative 1and other implementation alternatives. In addition, annual risks from incident-free transport are developed

Novel technique for manipulating MOX fuel particles using radiation pressure of a laser light

International Nuclear Information System (INIS)

Omori, R.; Suzuki, A.

2001-01-01

We proposed two principles based on the laser manipulation technique for collecting MOX fuel particles floating in air. While Principle A was based on the acceleration of the MOX particles due to the radiation pressure of a visible laser light, Principle B was based on the gradient forces exerted on the particles when an infrared laser light was incident. Principle A was experimentally verified using MnO 2 particles. Numerical results also showed the possibility of collecting MOX fuel particles based on both the principles. (authors)

Improvement in operating incident experience at the Savannah River Burial Ground

International Nuclear Information System (INIS)

Cornman, W.R.

1979-01-01

Low-level radioactive wastes generated at the Savannah River Plant and Laboratory are stored at the Savannah River burial ground. These wastes have accumulated from >20 years of reprocessing nuclear fuels and materials for defense programs at the Savannah River Plant. Burial in earthen trenches and aboveground storage for transuranic materials are the principal modes of storage. The infrequent operating incidents that have occurred during the 20-year period have been analyzed. The incidents can be categorized as those causing airborne contamination, waterborne contamination, or vegetation contamination through penetration of plant roots into contaminated soil. Contamination was generally confined to the immediate area of the burial ground. Several incidents occurred because of unintentional burial or exhumation of material. The frequency of operating incidents decreased with operating experience of the burial ground, averaging only about two incidents per year during the last six years of operation

TVSA-T fuel assembly for 'Temelin' NPP. Main results of design and safety analyses. Trends of development

International Nuclear Information System (INIS)

Samojlov, O.B.; Kajdalov, V.B.; Falkov, A.A.; Bolnov, V.A.; Morozkin, O.N.; Molchanov, V.L.; Ugryumov, A.V.

2010-01-01

TVSA is a fuel assembly with rigid skeleton formed by 6 angle pieces and SG is successfully operated at 17 VVER-1000 power units of Kalinin NPP, as well as at Ukrainian and Bulgarian NPPs. Based on a contract for fuel supply to the Temelin NPP, the TVSA-T fuel assembly was developed, building on proven solutions confirmed by operation of TVSA modifications during 4-6 years and by the results of post-irradiation examination. The TVSA-T design includes combined spacer grids (SG+MG) and by fuel column elongation by 150 mm. A set of analyses and experiments was performed to validate the design, including thermal hydraulic tests, validation of critical heat flux correlation for TVSA-T, integrated mechanical, vibration and lifetime tests. A licence to use the fuel has been granted by the Czech State Office for Nuclear Safety. The TVSA-T core is currently in operation at the Temelin-1 reactor unit. The presentation is concluded as follows: TVSA-T fuel assembly for Temelin has been validated. The TVSA-T design is based on approved technical decisions and meets the current requirements for lifetime, operational maneuverability and safety. The results of post-irradiation examination of TVSA-T operated at the Kalinin-1 unit for 4 years confirm the assembly operability, skeleton stiffness, geometric stability and normal fuel rod cladding condition. The properties of the TVSA fuel with MG allow the core power to be increased up to 3300 MW to match the envisaged future VVER (MIR-1200) design, providing allowable fuel rod power FΔh =1.63 (to implement effective fuel cycles). (P.A.)

Deterioration of the fuel injection parameters as a result of Common Rail injectors deposit formation

Directory of Open Access Journals (Sweden)

Stępień Zbigniew

2017-01-01

Full Text Available The article describes external and internal Common Rail injectors deposits formed in dynamometer engine simulation tests. It discussed not only the key reasons and factors influencing injector deposit formation but also the resulting way of fuel preparation and engine test approaches. The effects of external coking deposit as well as internal deposits two most common form types that is carboxylic soaps and organic amides on deterioration of the fuel injection parameters were assessed. The assessments covered both deposits impacts on quantitative and qualitative changes of the injectors diagnostic parameters and as a result on deterioration of the injector performance. Finally the comparisons between characteristic of dosage of one fuel injector before test and characteristics few injectors after engine tests of simulated deposit formation were made.

PSA predicts development of incident lower urinary tract symptoms: results from the REDUCE study.

Science.gov (United States)

Patel, Devin N; Feng, Tom; Simon, Ross M; Howard, Lauren E; Vidal, Adriana C; Moreira, Daniel M; Castro-Santamaria, Ramiro; Roehrborn, Claus; Andriole, Gerald L; Freedland, Stephen J

2018-05-23

The relationship between baseline prostate-specific antigen (PSA) and development of lower urinary tract symptoms (LUTS) in asymptomatic and mildly symptomatic men is unclear. We sought to determine if PSA predicts incident LUTS in these men. A post-hoc analysis of the 4-year REDUCE study was performed to assess for incident LUTS in 1534 men with mild to no LUTS at baseline. The primary aim was to determine whether PSA independently predicted incident LUTS after adjusting for the key clinical variables of age, prostate size, and baseline International prostate symptom score (IPSS). Incident LUTS was defined as the first report of medical treatment, surgery, or sustained clinically significant symptoms (two IPSS >14). Cox proportional hazards, cumulative incidence curves, and the log-rank test were used to test our hypothesis. A total of 1534 men with baseline IPSS PSA 2.5-4 ng/mL, 589 with PSA 4.1-6 ng/mL, and 610 with PSA 6-10 ng/mL. During the 4-year study, 196 men progressed to incident LUTS (50.5% medical treatment, 9% surgery, and 40.5% new symptoms). As a continuous variable, higher PSA was associated with increased incident LUTS on univariable (HR 1.09, p = 0.019) and multivariable (HR 1.08, p = 0.040) analysis. Likewise, baseline PSA 6-10 ng/mL was associated with increased incident LUTS vs. PSA 2.5-4 ng/mL in adjusted models (HR 1.68, p = 0.016). This association was also observed in men with PSA 4.1-6 ng/mL vs. PSA 2.5-4 ng/mL (HR 1.60, p = 0.032). Men with mild to no LUTS but increased baseline PSA are at increased risk of developing incident LUTS presumed due to benign prostatic hyperplasia.

Incidents with hazardous radiation sources

International Nuclear Information System (INIS)

Schoenhacker, Stefan

2016-01-01

Incidents with hazardous radiation sources can occur in any country, even those without nuclear facilities. Preparedness for such incidents is supposed to fulfill globally agreed minimum standards. Incidents are categorized in incidents with licensed handling of radiation sources as for material testing, transport accidents of hazardous radiation sources, incidents with radionuclide batteries, incidents with satellites containing radioactive inventory, incidents wit not licensed handling of illegally acquired hazardous radiation sources. The emergency planning in Austria includes a differentiation according to the consequences: incidents with release of radioactive materials resulting in restricted contamination, incidents with release of radioactive materials resulting in local contamination, and incidents with the hazard of e@nhanced exposure due to the radiation source.

Reducing fuel subsidies and the implication on fiscal balance and poverty in Indonesia: A simulation analysis

International Nuclear Information System (INIS)

Dartanto, Teguh

2013-01-01

There is an urgent need for phasing out fuel subsidies in Indonesia due to a severe budget deficit and a worsening of income distribution. Fuel subsidies, of which almost 72% are enjoyed by the 30% of the richest income groups, have consumed on average 63.8% of the total subsidies between 1998 and 2013. This paper aims at evaluating the relationship between existing fuel subsidies and fiscal balance and at analysing the poverty impact of phasing out fuel subsidies. Applying a CGE-microsimulation, this study found that removing 25% of fuel subsidies increases the incidence of poverty by 0.259 percentage points. If this money were fully allocated to government spending, the poverty incidence would decrease by 0.27 percentage points. Moreover, the 100% removal of fuel subsidies and the reallocation of 50% of them to government spending, transfers and other subsidies could decrease the incidence of poverty by 0.277 percentage points. However, these reallocation policies might not be effective in compensating for the adverse impacts of the 100% removal of fuel subsidies if economic agents try to seek profit through mark-up pricing over the increase of production costs. - Highlights: ► Massive fuel subsidies reduce fiscal spaces used to alleviate poverty in Indonesia. ► Indonesia can avoid a budget deficit by 78% cutting of fuel subsidies. ► A CGE-microsimulation is applied to analyse the impacts of fuel subsidy reallocation. ► The 50% of reallocation fuel subsidies decreases the poverty by 0.277 percentage points. ► Mark-up pricing done by economic agents reduces the effectiveness of reallocation

Update on the Us Army’s Fuel Contamination Detection Efforts

Science.gov (United States)

2011-04-04

Dual Eng flameout Contaminated with water and particulate 14 2009 C-130 USAF Jet Fuel Contamination with Fatty Acid Methyl Ester ( FAME ) BioFuel 15 20...DRNEN. WARRGHTER FOCUSED. Summary of Incidents • US Air Force – 2 - T-37 Super Absorbent Polymer 1 C 130 FAME Fatty Methyl Ester

Nuclear fuels

International Nuclear Information System (INIS)

Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

2009-01-01

Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

Nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

2009-07-01

Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

Fuel cell powered vehicles using supercapacitors-device characteristics, control strategies, and simulation results

Energy Technology Data Exchange (ETDEWEB)

Zhao, H.; Burke, A.F. [Institute of Transportation Studies, University of California (United States)

2010-10-15

The fuel cell powered vehicle is one of the most attractive candidates for the future due to its high efficiency and capability to use hydrogen as the fuel. However, its relatively poor dynamic response, high cost and limited life time have impeded its widespread adoption. With the emergence of large supercapacitors (also know as ultracapacitors, UCs) with high power density and the shift to hybridisation in the vehicle technology, fuel cell/supercapacitor hybrid fuel cell vehicles are gaining more attention. Fuel cells in conjunction with supercapacitors can create high power with fast dynamic response, which makes it well suitable for automotive applications. Hybrid fuel cell vehicles with different powertrain configurations have been evaluated based on simulations performed at the Institute of Transportation Studies, University of California-Davis. The following powertrain configurations have been considered: (a)Direct hydrogen fuel cell vehicles (FCVs) without energy storage (b)FCVs with supercapacitors directly connected in parallel with fuel cells (c)FCVs with supercapacitors coupled in parallel with fuel cells through a DC/DC converter (d)FCVs with fuel cells connected to supercapacitors via a DC/DC converter. (Abstract Copyright [2010], Wiley Periodicals, Inc.)

Initial cathode processing experiences and results for the treatment of spent fuel

International Nuclear Information System (INIS)

Westphal, B.R.; Laug, D.V.; Brunsvold, A.R.; Roach, P.D.

1996-01-01

As part of the spent fuel treatment demonstration at Argonne National Laboratory, a vacuum distillation process is being employed for the recovery of uranium following an electrorefining process. Distillation of a salt electrolyte, primarily consisting of a eutectic mixture of lithium and potassium chlorides, from uranium is achieved by a batch operation termed ''cathode processing.'' Cathode processing is performed in a retort furnace which enables the production of a stable uranium product that can be isotopically diluted and stored. To date, experiments have been performed with two distillation units; one for prototypical testing and the other for actual spent fuel treatment operations. The results and experiences from these initial experiments with both units will be discussed as well as problems encountered and their resolution

Development of Metallic Fuels for Actinide Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hayes, Steven Lowe [Idaho National Laboratory; Fielding, Randall Sidney [Idaho National Laboratory; Benson, Michael Timothy [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory; Carmack, William Jonathan [Idaho National Laboratory

2015-09-01

Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimized for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10

Safety of operations in the manufacture of driver fuel for the first charge of the Dragon Reactor and modifications to the safety document for the Dragon Fuel Element Production Building

International Nuclear Information System (INIS)

Beutler, H.; Cross, J.; Flamm, J.

1965-01-01

The manufacture of the zirconium containing 'driver' fuel and fuel elements for the First Charge of the Dragon Reactor Experiment has been completed without incident. This is a report on the safety of operations in the Dragon Fuel Element Production Building during an approximately six month period when the 'driver' fuel was manufactured and 25 elements containing this fuel were assembled and exported to the Reactor Building. The opportunity is taken to bring the Safety Document up-to-date and to report on any significant operational failures of equipment. (author)

SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: Second Results Report and Appendices

Energy Technology Data Exchange (ETDEWEB)

Eudy, L.; Chandler, K.

2011-10-01

This report describes operations at SunLine Transit Agency for their newest prototype fuel cell bus and five compressed natural gas (CNG) buses. In May 2010, SunLine began operating its sixth-generation hydrogen fueled bus, an Advanced Technology (AT) fuel cell bus that incorporates the latest design improvements to reduce weight and increase reliability and performance. The agency is collaborating with the U.S. Department of Energy's (DOE) National Renewable Energy Laboratory (NREL) to evaluate the bus in revenue service. This is the second results report for the AT fuel cell bus since it was placed in service, and it focuses on the newest data analysis and lessons learned since the previous report. The appendices, referenced in the main report, provide the full background for the evaluation. They will be updated as new information is collected but will contain the original background material from the first report.

The French program on the spent nuclear fuel long term evolution: Major results, uncertainties and new requirements

International Nuclear Information System (INIS)

Ferry, Cecile; Poinssot, Christophe; Gras, Jean-Marie

2006-01-01

conditions; oxidation kinetics of UO 2 and spent fuel are assessed by coupling a microscopic approach with macroscopic classical methods; - in water, this boundary condition corresponds to the nominal scenario after the breaching of the canister during geological disposal; in this case, the effect of water a radiolysis on the spent fuel matrix dissolution is investigated. An overview of the state of knowledge on the long-term behaviour of spent fuel under these various conditions has been provided by Ferry et al. (2005). This report constitutes the scientific synthesis due at the term of the law. This state of the art was derived from the results obtained under the PRECCI project as well as from a review of the literature and some of the data issued from the 5. PRDC European project 'Spent Fuel Stability under Repository Conditions' (Poinssot et al, 2005). Regarding these results, this paper presents for each initial operational question and concept, the main scientific issues, the major results obtained since 1999, and the remaining uncertainties. The new requirements in the frame of a broader context which includes transport and in-pool storage issues are also identified. The contents of this paper is as follows: 1 Introduction; 2 The retrievability of spent fuel assemblies after storage; 2.1 Main scientific issues; 2.2 Major results; 2.2.1 Mechanical evolution of the cladding and structural materials; 2.2.2 Specific case of the defected fuel rods - Evolution of the spent fuel pellets; 2.3 Remaining uncertainties and new requirements for each operational context; 3 The treatment of spent fuel after a long period of storage; 3.1 Main scientific issues; 3.2 Major results; 3.3 Remaining uncertainties and new requirements; 4 The definition of the radionuclide source terms; 4.1 Main scientific issues; 4.2 Major results; 4.3 Remaining uncertainties and new requirements for each operational context; 5 The compatibility between dry storage and subsequent disposal; 5.1 Main

Results of a diesel multiple unit fuel tank blunt impact test

Science.gov (United States)

2017-04-04

The Federal Railroad Administrations Office of Research and Development is conducting research into passenger locomotive fuel tank crashworthiness. A series of impact tests is being conducted to measure fuel tank deformation under two types of dyn...

Fuel assembly cooling experience at the FFTF/IEM cell

International Nuclear Information System (INIS)

McGuinness, P.W.

1985-01-01

In the Fast Flux Test Facility (FFTF), sodium wetted irradiated fuel assemblies are discharged to the Interim Examination and Maintenance (IEM) Cell for disassembly and post-irradiation examination in an inert argon atmosphere. While in the IEM Cell, fuel assemblies are cooled by the IEM Cell Subassembly Cooling System. This paper describes the cooling system design, performance, and lessons learned, including a discussion of two overtemperature incidents. 2 refs., 6 figs

EUBIONET II. Efficient trading of biomass fuels and analysis of fuel supply chains and business models for market actors by networking. Final result-oriented report

Energy Technology Data Exchange (ETDEWEB)

Alakangas, E.; Wiik, C.; Vesterinen, P. (and others)

2008-02-15

The project aimed to increase deployment of biomass fuels into European market and match technology uptake by means of market and policy analysis and other well-defined tasks to meet European policy targets in renewable energy sector. The project is to result as increased use of biomass fuels and market uptake of innovative bioenergy technologies. The objectives of the project were the following: 1) To give a clear outlook on current and future biomass fuel market trends. 2) To give feedback on the suitability of CEN 335 biofuel standard for trading of solid biofuels. 3) To provide well-analysed estimation on techno-economic potential of the biomass fuel volumes until 2010 based on the existing studies and experts opinions. Regarding the forest biomass sector, co-operation will be done with forest industry stakeholders to find proper balance between forest industry raw material and bioenergy usage. 4) To enhance biomass fuel trade and technology transfer by networking among different actors. 5) To analyse, select and describe the most suitable trading and business models for small- and largescale biofuel supply chains for heat and power production by taking into account the environmental aspects and sustainability. 6) To enhance biomass usage by the means of co-operation and information dissemination among different market actors in the fuel-utilisation chain. Target groups were biomass fuel traders and users, fuel producers and suppliers of different scales, policy makers in both current and new member states. Key associations, i.e. AEBIOM and CEPI, were participating in the project and disseminating information to various groups. The project has been structured in 5 workpackages. Project was carried out by 16 partners, which are the key national bioenergy organisations in the European countries and have a long co-operation relationship in previous bioenergy networks. The project has published summary reports and national report of each WP and this report is

Reasons for discrepancy between incidence and prevalence of epilepsy in lower income countries: Epilepsia's survey results.

Science.gov (United States)

Mathern, Gary W; Beninsig, Laurie; Nehlig, Astrid

2015-02-01

From July to August 2014, Epilepsia conducted an online survey seeking opinions that explained the discrepancy between the incidence and prevalence of epilepsy in lower income countries. Data on cumulative incidence suggest a higher rate of active epilepsy than reported in lifetime prevalence surveys. This study reports the findings of that poll addressing the proposal in our Controversy in Epilepsy series that it could be from increased death rates. The survey consisted of a question addressing possible reasons to explain the discrepancy between the incidence and prevalence of epilepsy. Another four questions addressed demographic information. There were 34 responders who completed the survey. Half (50%) of the responders felt that the discrepancy between cumulative incidence and lifetime prevalence was due to lack of uniform definitions and misclassification of patients in study design, 23.5% said the discrepancy was due to a higher mortality from diseases and conditions such as trauma and infections associated with epilepsy, 23.5% indicated that the stigma of epilepsy prevented people from acknowledging their disease in prevalence surveys, and 2.9% felt it was from poor access to qualified medical personal and utilization of medical treatments that increased death rates directly related to epilepsy. Within the limitations of sample size, the results of this survey support that the discrepancy between the incidence and prevalence of epilepsy in lower income regions of the world is due to problems in acquiring the data and stigma rather than higher mortality from diseases associated with epilepsy and repeated seizures. Wiley Periodicals, Inc. © 2014 International League Against Epilepsy.

Argonne Fuel Cycle Facility ventilation system -- modeling and results

International Nuclear Information System (INIS)

Mohr, D.; Feldman, E.E.; Danielson, W.F.

1995-01-01

This paper describes an integrated study of the Argonne-West Fuel Cycle Facility (FCF) interconnected ventilation systems during various operations. Analyses and test results include first a nominal condition reflecting balanced pressures and flows followed by several infrequent and off-normal scenarios. This effort is the first study of the FCF ventilation systems as an integrated network wherein the hydraulic effects of all major air systems have been analyzed and tested. The FCF building consists of many interconnected regions in which nuclear fuel is handled, transported and reprocessed. The ventilation systems comprise a large number of ducts, fans, dampers, and filters which together must provide clean, properly conditioned air to the worker occupied spaces of the facility while preventing the spread of airborne radioactive materials to clean am-as or the atmosphere. This objective is achieved by keeping the FCF building at a partial vacuum in which the contaminated areas are kept at lower pressures than the other worker occupied spaces. The ventilation systems of FCF and the EBR-II reactor are analyzed as an integrated totality, as demonstrated. We then developed the network model shown in Fig. 2 for the TORAC code. The scope of this study was to assess the measured results from the acceptance/flow balancing testing and to predict the effects of power failures, hatch and door openings, single-failure faulted conditions, EBR-II isolation, and other infrequent operations. The studies show that the FCF ventilation systems am very controllable and remain stable following off-normal events. In addition, the FCF ventilation system complex is essentially immune to reverse flows and spread of contamination to clean areas during normal and off-normal operation

Strategy of fuel management

International Nuclear Information System (INIS)

Guesdon, B.; Le Bars, M.; Mathonniere, G.

1996-01-01

The management of nuclear fuels in PWR type reactors has been adapted to improve the safety and the competitiveness of brackets. The economic optimum, at the park level, depends on many parameters, variable with time and in function of them, we favour the annual campaigns and the economy won on the cost of cycle, or long campaigns with benefit on availability. The reduction of the number of stopping improves the availability, limits the doses integrated by the personnel of intervention and reduces the number of incidents during the stopping. An other determining factor is connected to the policy of closed cycle with the the principle of equality between the reprocessing flux and the valorization of reprocessed fuels: plutonium and reprocessed uranium. The progress of fuel have allowed significant improvements in the managements of cores. With the safety, the aim is also to keep if not improve the competitiveness of the Nuclear park by valorizing the matter coming from reprocessing. (N.C.)

Proceedings of the Topical Meeting on the safety of nuclear fuel cycle intermediate storage facilities

International Nuclear Information System (INIS)

1998-01-01

The CSNI Working Group on Fuel Cycle Safety held an International Topical Meeting on safety aspects of Intermediate Storage Facilities in Newby Bridge, England, from 28 to 30 October 1997. The main purpose of the meeting was to provide a forum for the exchange of information on the technical issues on the safety of nuclear fuel cycle facilities (intermediate storage). Titles of the papers are: An international view on the safety challenges to interim storage of spent fuel. Interim storage of intermediate and high-level waste in Belgium: a description and safety aspects. Encapsulated intermediate level waste product stores at Sellafield. Safety of interim storage facilities of spent fuel: the international dimension and the IAEA's activities. Reprocessing of irradiated fuel and radwaste conditioning at Belgoprocess site: an overview. Retrieval of wastes from interim storage silos at Sellafield. Outline of the fire and explosion of the bituminization facility and the activities of the investigation committee (STAIJAERI). The fire and explosion incident of the bituminization facility and the lessons learned from the incident. Study on the scenario of the fire incident and related analysis. Study on the scenario of the explosion incident and related analysis. Accident investigation board report on the May 14, 1997 chemical explosion at the plutonium reclamation facility, Hanford site, Richland, Washington. Dry interim storage of spent nuclear fuel elements in Germany. Safe and effective system for the bulk receipt and storage of light water reactor fuel prior to reprocessing. Receiving and storage of glass canisters at vitrified waste storage center of Japan Nuclear Fuel Ltd. Design and operational experience of dry cask storage systems. Sellafield MOX plant; Plant safety design (BNFL). The assessment of fault studies for intermediate term waste storage facilities within the UK nuclear regulatory regime. Non-active and active commissioning of the thermal oxide

Results from Tight and Loose Coupled Multiphysics in Nuclear Fuels Performance Simulations using BISON

Energy Technology Data Exchange (ETDEWEB)

S. R. Novascone; B. W. Spencer; D. Andrs; R. L. Williamson; J. D. Hales; D. M. Perez

2013-05-01

The behavior of nuclear fuel in the reactor environment is affected by multiple physics, most notably heat conduction and solid mechanics, which can have a strong influence on each other. To provide credible solutions, a fuel performance simulation code must have the ability to obtain solutions for each of the physics, including coupling between them. Solution strategies for solving systems of coupled equations can be categorized as loosely-coupled, where the individual physics are solved separately, keeping the solutions for the other physics fixed at each iteration, or tightly coupled, where the nonlinear solver simultaneously drives down the residual for each physics, taking into account the coupling between the physics in each nonlinear iteration. In this paper, we compare the performance of loosely and tightly coupled solution algorithms for thermomechanical problems involving coupled thermal and mechanical contact, which is a primary source of interdependence between thermal and mechanical solutions in fuel performance models. The results indicate that loosely-coupled simulations require significantly more nonlinear iterations, and may lead to convergence trouble when the thermal conductivity of the gap is too small. We also apply the tightly coupled solution strategy to a nuclear fuel simulation of an experiment in a test reactor. Studying the results from these simulations indicates that perhaps convergence for either approach may be problem dependent, i.e., there may be problems for which a loose coupled approach converges, where tightly coupled won’t converge and vice versa.

Results of the irradiation of mixed UO2 - PuO2 oxide fuel elements

International Nuclear Information System (INIS)

Mikailoff, H.; Mustelier, J.P.; Bloch, J.; Ezran, L.; Hayet, L.

1966-01-01

In order to study the behaviour of fuel elements used for the first charge of the reactor Rapsodie, a first batch of eleven needles was irradiated in the reactor EL3 and then examined. These needles (having a shape very similar lo that of the actual needles to be used) were made up of a stack of sintered mixed-oxide pellets: UO 2 containing about 10 per cent of PuO 2 . The density was 85 to 97 per cent of the theoretical, value. The diametral gap between the oxide and the stainless steel can was between 0,06 and 0,27 mm. The specific powers varied from 1230 to 2700 W/cm 3 and the can temperature was between 450 and 630 C. The maximum burn-up attained was 22000 MW days/tonne. Examination of the needles (metrology, radiography and γ-spectrography) revealed certain macroscopic changes, and the evolution of the fuel was shown by micrographic studies. These observations were used, together with flux measurements results, to calculate the temperature distribution inside the fuel. The volume of the fission gas produced was measured in some of the samples; the results are interpreted taking into account the temperature distribution in the oxide and the burn-up attained. Finally a study was made both of the behaviour of a fuel element whose central part was molten during irradiation, and of the effect of sodium which had penetrated into some of the samples following can rupture. (author) [fr

Results from tight and loose coupled multiphysics in nuclear fuels performance simulations using BISON

International Nuclear Information System (INIS)

Novascone, S. R.; Spencer, B. W.; Andrs, D.; Williamson, R. L.; Hales, J. D.; Perez, D. M.

2013-01-01

The behavior of nuclear fuel in the reactor environment is affected by multiple physics, most notably heat conduction and solid mechanics, which can have a strong influence on each other. To provide credible solutions, a fuel performance simulation code must have the ability to obtain solutions for each of the physics, including coupling between them. Solution strategies for solving systems of coupled equations can be categorized as loosely-coupled, where the individual physics are solved separately, keeping the solutions for the other physics fixed at each iteration, or tightly coupled, where the nonlinear solver simultaneously drives down the residual for each physics, taking into account the coupling between the physics in each nonlinear iteration. In this paper, we compare the performance of loosely and tightly coupled solution algorithms for thermomechanical problems involving coupled thermal and mechanical contact, which is a primary source of interdependence between thermal and mechanical solutions in fuel performance models. The results indicate that loosely-coupled simulations require significantly more nonlinear iterations, and may lead to convergence trouble when the thermal conductivity of the gap is too small. We also apply the tightly coupled solution strategy to a nuclear fuel simulation of an experiment in a test reactor. Studying the results from these simulations indicates that perhaps convergence for either approach may be problem dependent, i.e., there may be problems for which a loose coupled approach converges, where tightly coupled won't converge and vice versa. (authors)

Mixed-oxide (MOX) fuel performance benchmark. Summary of the results for the PRIMO MOX rod BD8

International Nuclear Information System (INIS)

Ott, L.J.; Sartori, E.; Costa, A.; ); Sobolev, V.; Lee, B-H.; Alekseev, P.N.; Shestopalov, A.A.; Mikityuk, K.O.; Fomichenko, P.A.; Shatrova, L.P.; Medvedev, A.V.; Bogatyr, S.M.; Khvostov, G.A.; Kuznetsov, V.I.; Stoenescu, R.; Chatwin, C.P.

2009-01-01

The OECD/NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, nuclear fuel performance, and fuel cycle issues related to the disposition of weapons-grade plutonium as MOX fuel. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close cooperation with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A major part of these activities includes benchmark studies. This report describes the results of the PRIMO rod BD8 benchmark exercise, the second benchmark by the TFRPD relative to MOX fuel behaviour. The corresponding PRIMO experimental data have been released, compiled and reviewed for the International Fuel Performance Experiments (IFPE) database. The observed ranges (as noted in the text) in the predicted thermal and FGR responses are reasonable given the variety and combination of thermal conductivity and FGR models employed by the benchmark participants with their respective fuel performance codes

Operating results and simulations on a fuel cell for residential energy systems

International Nuclear Information System (INIS)

Hamada, Yasuhiro; Goto, Ryuichiro; Nakamura, Makoto; Kubota, Hideki; Ochifuji, Kiyoshi

2006-01-01

This paper describes the performance evaluation of a polymer electrolyte fuel cell (PEFC) prototype and demonstration experiments of the electric power and domestic hot water system using it from a pragmatic view-point. Three types of demonstration experiments were carried out applying standard electric power and hot water demands. It was shown that the primary energy reduction rate of this system as compared to the conventional system reached up to 24% under double daily start and stop (DSS) operation. The amount of primary energy reduction in experiments using the energy demand of a household in Sapporo in winter exceeded the experimental results of the standard energy demand, demonstrating that the effects of the introduction of a fuel cell in cold regions could be considerable, in particular, during the winter season

Radiation heat transfer calculations for the uranium fuel-containment region of the nuclear light bulb engine.

Science.gov (United States)

Rodgers, R. J.; Latham, T. S.; Krascella, N. L.

1971-01-01

Calculation results are reviewed of the radiant heat transfer characteristics in the fuel and buffer gas regions of a nuclear light bulb engine based on the transfer of energy by thermal radiation from gaseous uranium fuel in a neon vortex, through an internally cooled transparent wall, to seeded hydrogen propellant. The results indicate that the fraction of UV energy incident on the transparent walls increases with increasing power level. For the reference engine power level of 4600 megw, it is necessary to employ space radiators to reject the UV radiated energy absorbed by the transparent walls. This UV energy can be blocked by employing nitric oxide and oxygen seed gases in the fuel and buffer gas regions. However, this results in increased UV absorption in the buffer gas which also requires space radiators to reject the heat load.

Nevada commercial spent nuclear fuel transportation experience

International Nuclear Information System (INIS)

1991-09-01

The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed

Main results on pilot operation during 5 years of the 3rd generation fuel in VVER-440 reactors of Kola NPP

International Nuclear Information System (INIS)

Saprykin, V.; Sumarokov, M.; Gagarinskiy, A.; Sumarokova, A.; Adeev, V.

2015-01-01

In the report the results of comparison of main neutron-physical data of exploitation of nuclear fuel are presented for the average enrichment (on U - 235) of 4.87 for the 2nd and 3rd (12 piece) generations with the results of calculations by the complex of the programs KASKAD for 5 fuel loadings of Kola NPP Unit 4 with the reactor VVER- 440. The basic feature of fuel of the 3rd generation as compared with the 2nd is a presence of ribs of inflexibility at corners instead of cover of the fuel assembly and also the increased amount of uranium. The arrangement of fuel rods with different enrichment in fuel assemblies of the 2nd and 3rd generations is chosen identical for the convenient comparison of neutronic and thermohydraulic characteristics of the fuel of different generations. The fuel of 3rd generation was situated in the core symmetrically to the fuel of 2nd one

CRBRP design and test results for fuel handling systems, plugs, and seals

International Nuclear Information System (INIS)

Berg, G.E.

1977-01-01

The fuel handling system and reactor rotating plugs for the Clinch River Breeder Reactor Plant (CRBRP) are based primarily on existing technology and, in many respects, follow the concept developed for the Fast Flux Test Facility (FFTF). The equipment and the development programs initiated to verify its performance are described. Test results obtained from the development program, and the extent to which these results verified original design selections, or suggested potential improvements, are discussed

Study of core characteristics on fuel and coolant type. Results of F/S phase-I

International Nuclear Information System (INIS)

Ikegami, Tetsuo; Hayashi, Hideyuki; Sasaki, Makoto; Mizuno, Tomoyasu; Yamadate, Megumi; Takaki, Naoyuki; Kurosawa, Norifumi; Sakashita, Yoshiaki; Naganuma, Masayuki

2001-03-01

The phase-I of the Feasibility Study of Commercialized Fast Reactor Cycle Systems (F/S) were started from July, 1999 and terminated at the end of FY2000 in order to executed examination about technology alternatives of various commercialized fast reactor (FR) recycle concepts, in response to the JNC middle long term enterprise plan. In the phase-I of this F/S, a number of conceptual candidates have been selected from the following 5 viewpoints: a) ensuring safety, b) economic competitiveness to future LWRs, c) efficient utilization of resources, d) reduction of environmental burden, e) enhancement of nuclear non-proliferation. As for this study from the above viewpoints, core characteristics of many kinds of reactors have been investigated, analyzed and examined a core / a fuel characteristic in the combinations of fuel and coolant types and power output scales. Based on these results, R and D plans of the phase-II to be performed have been proposed, and a database to select candidate reactor concepts has been prepared. The conclusions have been obtained in the phase-I are as follows: (1) Evaluation of a fuel form in every each coolant was compared. A promising fuel form was extracted as follows: an oxide and a metal fuel for sodium coolant cores, a metal and a nitride fuel for heavy metal coolant cores, an oxide and a nitride fuel for carbon dioxide coolant cores and a nitride fuel for He gas coolant cores. (2) As the general idea that performance of a core nucleus can be compatible with re-criticality evasion in sodium coolant large-sized oxide fuel cores, a axial blanket particle elimination radial heterogeneous core is one influential candidate. (3) In case of Pb-Bi coolant nature circulation medium size core with an oxide fuel, it is difficult to simultaneously achieve higher discharged burn-up and higher breeding ratio according to the viewpoints of the phase-I. (4) Core characteristics of a carbon dioxide coolant core shows to be almost equivalent to that of

Infant brain tumors: incidence, survival, and the role of radiation based on Surveillance, Epidemiology, and End Results (SEER) Data.

Science.gov (United States)

Bishop, Andrew J; McDonald, Mark W; Chang, Andrew L; Esiashvili, Natia

2012-01-01

To evaluate the incidence of infant brain tumors and survival outcomes by disease and treatment variables. The Surveillance, Epidemiology, and End Results (SEER) Program November 2008 submission database provided age-adjusted incidence rates and individual case information for primary brain tumors diagnosed between 1973 and 2006 in infants less than 12 months of age. Between 1973 and 1986, the incidence of infant brain tumors increased from 16 to 40 cases per million (CPM), and from 1986 to 2006, the annual incidence rate averaged 35 CPM. Leading histologies by annual incidence in CPM were gliomas (13.8), medulloblastoma and primitive neuroectodermal tumors (6.6), and ependymomas (3.6). The annual incidence was higher in whites than in blacks (35.0 vs. 21.3 CPM). Infants with low-grade gliomas had the highest observed survival, and those with atypical teratoid rhabdoid tumors (ATRTs) or primary rhabdoid tumors of the brain had the lowest. Between 1979 and 1993, the annual rate of cases treated with radiation within the first 4 months from diagnosis declined from 20.5 CPM to incidence of infant brain tumors has been stable since 1986. Survival outcomes varied markedly by histology. For infants with medulloblastoma and ATRTs, improved survival was observed in patients treated with both surgery and early radiation compared with those treated with surgery alone. Copyright © 2012 Elsevier Inc. All rights reserved.

Protection of spent aluminum-clad research reactor fuels during extended wet storage

International Nuclear Information System (INIS)

Fernandes, Stela M.C.; Correa, Olandir V.; Souza, Jose A.; Ramanathan, Lalgudi V.; Antunes, Renato A.

2013-01-01

Aluminum-clad spent nuclear fuel from research reactors (RR) is stored in light water filled pools or basins worldwide. Many incidences of pitting corrosion of the fuel cladding has been reported and attributed to synergism in the effect of certain water parameters. Protection of spent Al-clad RR fuel with a conversion coating was proposed in 2008. Preliminary results revealed increased pitting corrosion resistance of cerium oxide coated aluminum alloys AA 1050 and AA 6061, used as RR fuel plate cladding. Further development of conversion coatings for Al alloys was carried out and this paper presents: (a) the preparation and characterization of hydrotalcite (HTC) coatings; (b) the results of laboratory tests in which the corrosion behavior of coated Al alloys in NaCl solutions was determined; (c) the results of field tests in which un-coated, boehmite coated, HTC coated and cerium modified boehmite / HTC coated AA 1050 and AA 6061 coupons were exposed to the IEA-R1 reactor spent fuel basin for extended periods. In these field tests the coupons coated with HTC from a high temperature (HT) bath and subsequently modified with Ce were the most resistant to pitting corrosion. In laboratory tests also, HT- hydrotalcite + Ce coated specimens were the most corrosion resistant in 0.01 M NaCl. The role of cerium in increasing the corrosion resistance imparted by the different conversion coatings of spent Al-clad RR fuel elements is presented. (author)

Dry vault for spent fuel depository. Basic outsets, operating results and safety of the CASCAD plant

International Nuclear Information System (INIS)

Bardelle, P.

1997-01-01

Reprocessing and recycling of fissile materials is the preferred approach to spent fuel management in France. However, a number of spent fuel elements from prototype and experimental nuclear reactors cannot be reprocessed in the existing industrial facilities, either because such facilities are booked to full capacity, or due to technical factors such as non standard nature of fuel or limited series of fuel. The CEA therefore built a facility in which spent fuel can be stored for a few decades (50 years), until favourable conditions prevail for its disposal. The main features of this project consist in a dry repository, which presents a low cost of working, against a wet one which is more expensive due to the circulation and the continuous control of the water. Therefore, this is a fair solution because the experimental fuels will present a rather low residual heat power after decay in the nuclear reactor. At this stage, it becomes possible to cool the fuel elements by a fully passive air circulation. This process allows a good efficiency without mechanical equipment and works all the better as the amount of heat to exhaust is great, in the limits of the design. However, we will see that this concept may be extended to a repository of standard spent fuel elements. This facility, known as 'CASCAD' (= CASemate CADarache) started up in 1990, and received its first canister of fuel on May 29, 1990. This paper reviews the basic design data of the facility, outlines the main techniques used for its construction, draws the safety concept and presents the first results determined by a looking-back over 4 years of working. (author)

Dry vault for spent fuel depository. Basic outsets, operating results and safety of the ''CASCAD'' plant

International Nuclear Information System (INIS)

Bardelle, P.

1994-01-01

Reprocessing and recycling of fissile materials is the preferred approach to spent fuel management in France. However, a number of spent fuel elements from prototype and experimental nuclear reactors cannot be reprocessed in the existing industrial facilities, either because such facilities are booked to full capacity, or due to technical factors such as non standard nature of fuel or limited series of fuel. The CEA therefore built a facility in which spent fuel can be stored for a few decades (50 years), until favourable conditions prevail for its disposal. The main features of this project consist in a dry depositary, which presents a low cost of working, against a wet one which is more expensive due to the circulation and the continuous controls of the water. Therefore, this is a fair solution because the experimental fuels will present a rather low residual heat power after decay in the nuclear reactor. At this stage, it becomes possible to cool the fuel elements by a fully passive air circulation. This process allows a good efficiency without mechanical equipment and works all the better as the amount of heat to exhaust is great, in the limits of the design. However, we will see that this concept may be extended to a depository of standard spent fuel elements. This facility, known as ''CASCAD'' (shortening for CASemate (=vault) CADarache) started up in 1990, and received its first canister of fuel on May 29 th 1990. This paper reviews the basic design data of the facility, outlines the main techniques used for its construction, draws the safety concepts and presents the first results determined by a looking-back over 4 years of working. (author)

Motor fuel taxation, energy conservation, and economic development: A regional approach

International Nuclear Information System (INIS)

England, Richard W.

2007-01-01

Combustion of motor fuels has a variety of environmental impacts on local, regional and global scales. Taxing motor fuels more heavily would mitigate those environmental impacts. However, many governments are reluctant to increase motor fuel taxes because they fear that the tax incidence will be regressive and that economic development will be impeded. Using data for the New England region of the United States, this paper argues that an oil-importing region can conserve energy, avoid regressive impacts and encourage economic development by taxing motor fuels more heavily and rebating the incremental revenues to owners of motor vehicles. (author)

Fuel performance of DOE fuels in water storage

International Nuclear Information System (INIS)

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

1993-01-01

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

Cancer incidence in North West Algeria (Mascara) 2000-2010: results from a population-based cancer registry

Science.gov (United States)

Benarba, Bachir; Meddah, Boumedienne; Hamdani, Houria

2014-01-01

Cancer is a leading cause of death worldwide accounting for 7.4 million deaths. Cancer has become a major public health concern in Algeria. The aim of the present study was to estimate cancer incidence in Mascara Province based on the population-based cancer registry. We analyzed data from the cancer registry of Mascara covering all cancer cases diagnosed by all methods and included in the registry from 1st January 2000 to 31st December 2010. The results are presented as incidence rates of cases by site, sex, age, and crude rate. Age-standardized rates per 100,000 person-years (ASRs) were calculated, using the direct method of standardization to the world population. A total of 1875 cases of invasive cancer were recorded. The mean age of diagnosis for all cancers was 52.66 ± 0.5 in men and 59.18 ± 0.6 in women. The ASR for all cancers in females was 27.8 per 100,000, and that for males was 23.6 per 100,000. The most important finding of the present study was the high incidence of liver cancer among males and females in Mascara. Among females, breast cancer was the most frequently reported followed by Cervix uteri, liver and colon. The most frequent cancer types in males were lung, colon, esophagus and stomach and liver. Cancer incidence in Mascara province was lower than that reported in other national and regional registries. Findings of the present study revealed high incidence of liver cancer in the province, the highest in Algeria, suggesting high prevalence of risk factors. PMID:26417294

Postulated accident scenarios for the on-site transport of spent nuclear fuel

International Nuclear Information System (INIS)

Morandin, G.; Sauve, R.

2004-01-01

Once a spent fuel container is loaded with spent fuel it typically travels on-site to a processing building for permanent lid attachment. During on-site transport a lid clamp is utilized to ensure the container lid remains in place. The safe on-site transport of spent nuclear fuel must rely on the structural integrity of the transport container and system of transport. Regard for on-site traffic and safe, efficient travel routes are important and manageable with well thought-out planning. Non-manageable incidences, such as flying debris from tornado force winds or postulated blasts in proximity to the transport container, that may result in high velocity impact and shock loading on the transport system must be considered. This paper consists of simulations that consider these types of postulated accident scenarios using detailed nonlinear finite element techniques

Complementary results of radioactivity analysis made in Corsica following the navigation incident of the Uss Hartford submarine

International Nuclear Information System (INIS)

2004-01-01

The results of analysis made by the Institute of radiation protection and nuclear safety in the area of the Bonifacio estuary do not reveal any increase of artificial radioactivity following the navigation incident of the Uss Hartford submarine. They confirm the previous published results. (N.C.)

Incidence of type 2 diabetes in Mexico. Results of The Mexico City Diabetes Study after 18 years of follow-up.

Directory of Open Access Journals (Sweden)

Clicerio González-Villalpando

2014-01-01

Full Text Available Objective. To estimate the incidence of type 2 diabetes (T2D in Mexican population. Materials and methods. Population based prospective study. At baseline (1990, the population at risk (1939 non-diabetic adults 35-64 years was evaluated with oral glucose tolerance test. Subsequent similar evaluations were done (1994, 1998, 2008. American Diabetes Association diagnostic criteria were applied. Results. The period of observation was 27842 person-years, the cumulative incidence of T2D was 14.4 and 13.7 per 1000 person-years for men and women, respectively. Incidence was 15.8, 15.7 and 12.7 per 1 000 person-years for the second (1994, third (1998 and fourth (2008 follow-up phases, respectively. The mean age at diagnosis was 44 years for prevalent cases and 56 years for incident cases. Conclusions. This is the first estimate of long-term incidence of T2D in Mexican population. The incidence is among the highest reported worldwide. It remained with few changes throughout the study period.

PHEBUS program: first results on PWR fuel behaviour in LOCA conditions

International Nuclear Information System (INIS)

Del Negro, R.; Reocreux, M.; Pelce, J.; Legrand, B.; Berna, P.

1982-09-01

In the first PHEBUS test with pressurized rods some rods burst and clad temperature reached 1100 0 C in the 25 rods bundle. There is now a lot of valuable experimental results and their analysis is in progress. The phase II on fuel behaviour in case of a large LOCA will start at the beginning of 83. The onset of the SFD program is foreseen to take place on the first months of 85

Concussion Incidence and Return-to-Play Time in National Basketball Association Players: Results From 2006 to 2014.

Science.gov (United States)

Padaki, Ajay S; Cole, Brian J; Ahmad, Christopher S

2016-09-01

Various research efforts have studied concussions in the National Football League, Major League Baseball, and the National Hockey League. However, no study has investigated the incidence and return-to-play trends in the National Basketball Association (NBA), which this study aims to do. Increased media scrutiny and public awareness, in addition to the institution of a league-wide concussion protocol, may have resulted in more conservative return-to-play practices. Descriptive epidemiology study. All concussions to NBA players that were publicly reported in the media from the beginning of the 2006 NBA season to the end of the 2014 season were included. The incidence and return-to-play statistics were generated by synthesizing information from publicly available records. There were 134 publicly reported concussions to NBA players from the beginning of the 2006 season to the conclusion of the 2014 season, resulting in an average of 14.9 concussions per season. The incidence has not changed significantly during this time span. The average games missed after a concussion from 2006 to 2010 was 1.6, significantly less than the 5.0 games missed from 2011 to 2014, following the institution of the NBA concussion protocol (P = .023). Although the incidence of publicly reported concussions in the NBA has not changed appreciably over the past 9 seasons, the time missed after a concussion has. While players often returned in the same game in the 2006 season, the combination of implemented policy, national coverage, medical staff awareness, and player education may have contributed to players now missing an average of 4 to 6 games after a concussion. A multitude of factors has resulted in more conservative return-to-play practices for NBA players after concussions. © 2016 The Author(s).

Tracing and analytical results of the dioxin contamination incident in 2008 originating from the Republic of Ireland.

Science.gov (United States)

Heres, L; Hoogenboom, R; Herbes, R; Traag, W; Urlings, B

2010-12-01

High levels of dioxins (PCDD/Fs) in pork were discovered in France and the Netherlands at the end of 2008. The contamination was rapidly traced back to a feed stock in the Republic of Ireland (RoI). Burning oil, used for the drying of bakery waste, appeared to be contaminated with PCBs. Consequently, very high levels up to 500 pg TEQ g⁻¹ fat were found in pork. The congener pattern clearly pointed to PCB-oil as a source, but the ratio between the non-dioxin-like indicator PCBs (PCBs 28, 52, 101, 138, 152 and 180) and PCDD/Fs was much lower than observed during the Belgian incident, thereby limiting the suitability of indicator PCBs as a marker for the presence of dioxins and dioxin-like PCBs. This paper describes the tracking and tracing of the incident, the public-private cooperation, the surveillance activities and its results. A major lesson to be learned from this incident is the importance of good private food safety systems. In this incident, it was the private surveillance systems that identified the origin of contamination within 10 days after the first signal of increased dioxin levels in a product. On the other hand, retrospective analyses showed that signals were missed that could have led to an earlier detection of the incident and the source. Above all, the incident would not have occurred when food safety assurance systems had been effectively implemented in the involved feed chain. It is discussed that besides primary responsibility for effective private food safety systems, the competent authorities have to supervise whether the food safety procedures are capable of coping with these kinds of complex food safety issues, while private food companies need to implement the law, and public authorities should supervise and enforce them. Finally, it is discussed whether the health risks derived from consumption of the contaminated batches of meat may have been underestimated during the incident due to the unusually high intake of dioxins.

Status and results of the theoretical and experimental investigations on the LWR fuel rod behavior under accident conditions

International Nuclear Information System (INIS)

Bocek, M.; Hofmann, P.; Leistikow, S.; Class, G.; Meyder, R.; Raff, S.; Erbacher, F.; Hofmann, G.; Ihle, P.; Karb, E.; Fiege, A.

1978-09-01

In this report the status of knowledge is described which has been gathered up to the end of 1977 of the LWR fuel rod behavior in loss-of-coolant accidents. The majority of results indicated have been derived from studies on the fuel rod behavior performed within the framework of the Nuclear Safety Project (PNS); partly, also the results of cooperating research establishments and fm international exchange of experience are referred to. The report has been subdivided into two complete parts: Part I provides a survey of the most significant results of the theoretical and experimental research projects on fuel rod behavior. Part II describes by detailed individual presentations the status as well as the results with respect to the major central subjects. (orig.) 891 RW 892 AP [de

Dry vault for spent fuel depository, basic outsets, operating results and safety of the ''CASCAD'' plant

International Nuclear Information System (INIS)

Mercier, J.P.; Bardelle, P.

1994-01-01

Reprocessing and recycling of fissile materials is the preferred approach to spent fuel management in France. However, a number of spent fuel elements from prototype and experimental nuclear reactors cannot be reprocessed in the existing industrial facilities, either because such facilities are booked to full capacity, or due to technical factors such as non standard nature of fuel or limited series of fuel. The CEA therefore built a facility in which spent fuel can be stored for a few decades (50 years), until favourable conditions prevail for its disposal. The main features of this project consist of a dry depository, which presents a low cost of operation, against a wet on which is more expensive due to the circulation and the continuous controls of the water. Therefore, this is a fair solution because the experimental fuels will present a rather low residual heat power after decay in the nuclear reactor. At this stage, it becomes possible to cool the fuel elements by a fully passive air circulation. This process allows a good efficiency without mechanical equipment and works all the better as the amount of heat to exhaust is great, in the limits of the design. However, we will see that this concept may be extended to a depository of standard spent fuel elements. This facility, known as ''CASCAD'' (shortening for CASemate (=vault) CADarache) started up in 1990, and received its first canister of fuel on May 29th 1990. This paper reviews the basic design data of the facility, outlines the main techniques used for its construction, draws the safety concepts and presents the first results determined by a looking-back over 4 years of working. (author). 1 fig

Results of the German alternative fuel cycle evaluation and further efforts geared toward demonstration of direct disposal

International Nuclear Information System (INIS)

Papp, R.; Closs, K.D.

1986-01-01

In a comparative study initiated by the German Federal Ministry for Research and Technology which was carried out by Karlsruhe Nuclear Research Center in the period from 1981 to 1985, direct disposal of spent fuel was contrasted to the traditional fuel cycle with reprocessing and recycle. The results of the study did not exhibit decisive advantages of direct disposal over fuel reprocessing. Due to this face and legal requirements of the German Atomic Energy Act, the cabinet concluded to continue to adhere to fuel reprocessing as the preferred version of ''Entsorgung''. But the door was left ajar for the direct disposal alternative that, under present atomic law, is permissible for fuel for which reprocessing is neither technically feasible nor economically justified. An ambitious program has been launched in the Federal Republic of Germany (FRG), geared to bring direct disposal to a point of technical maturity

Shipment of Taiwanese research reactor spent nuclear fuel (Phase 2): Environmental assessment

International Nuclear Information System (INIS)

1988-06-01

The proposed action is to transport approximately 1100 spent fuel rods from a foreign research reactor in Taiwan by sea to Hampton Roads, Virginia, and then overland by truck to the receiving basin for offsite fuels at the Savannah River Plant (SRP) for reprocessing to recover uranium and plutonium. The analysis of the impacts of the proposed action have been evaluated and shown to have negligible impact on the local environments. The calculations have been completed using the RADTRAN III code. PWR spent fuel was analyzed as a benchmark to link the calculations in this analysis to those in earlier environmental documentation. Cumulative total, maximum annual, and per shipment risks were calculated. The results indicate that the PWR spent fuel shipment risks are somewhat lower than those previously estimated. The cumulative and maximum annual normal, or incident-free, risks associated with the shipment of Taiwanese research reactor spent fuel is a factor of 10 lower than that for PWR fuel, and the cumulative and maximum annual accident radiological risks are a factor of about 2.2 lower than that for PWR spent fuel. As a result, the port risks are about a factor of 10 larger than the risk of overland transport. All of the risks calculated are small. The PWR risk values are similar to those judged by the NRC to be small enough not to warrant increased stringency in regulations. The Taiwanese research reactor spent fuel shipment risk values are smaller yet. 51 refs., 22 tabs

Theoretical analysis of the temperature changes and resultant loss of fuel integrity in the IEA-R1 research reactor fuel elements following a loss of coalant accident

International Nuclear Information System (INIS)

Garone, J.G.M.

1983-01-01

The IEA-R1 core following a loss of coolant accident (LOCA) is analysed. THe AIRLOCA code was used to calculate fuel temperatures, heat generation due to fission product decay and convective and radiative heat transfer from the fuel elements to the surrounding air both during and following the loss of coolant. The influence of certain critical parameters, such as log time, specific power was studied in detail. Representative results are presented and suggestions made to ensure that fuel integrity is maintained following a LOCA. (Author) [pt

Effects of using winter grazing as a fuel treatment on Wyoming big sagebrush plant communities

Science.gov (United States)

More frequent wildfires and incidences of mega-fires have increased the pressure for fuel treatments in sagebrush (Artemisia) communities. Winter grazing has been one of many fuel treatments proposed for Wyoming big sagebrush (A. tridentata Nutt. subsp. wyomingensis Beetle and A. Young) communitie...

In-pile test results of U-silicide or U-nitride coated U-7Mo particle dispersion fuel in Al

Science.gov (United States)

Kim, Yeon Soo; Park, J. M.; Lee, K. H.; Yoo, B. O.; Ryu, H. J.; Ye, B.

2014-11-01

U-silicide or U-nitride coated U-Mo particle dispersion fuel in Al (U-Mo/Al) was in-pile tested to examine the effectiveness of the coating as a diffusion barrier between the U-7Mo fuel kernels and Al matrix. This paper reports the PIE data and analyses focusing on the effectiveness of the coating in terms of interaction layer (IL) growth and general fuel performance. The U-silicide coating showed considerable success, but it also provided evidence for additional improvement for coating process. The U-nitride coated specimen showed largely inefficient results in reducing IL growth. From the test, important observations were also made that can be utilized to improve U-Mo/Al fuel performance. The heating process for coating turned out to be beneficial to suppress fuel swelling. The use of larger fuel particles confirmed favorable effects on fuel performance.

In-pile test results of U-silicide or U-nitride coated U-7Mo particle dispersion fuel in Al

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Park, J.M.; Lee, K.H.; Yoo, B.O. [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Ryu, H.J. [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Ye, B. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2014-11-15

U-silicide or U-nitride coated U-Mo particle dispersion fuel in Al (U-Mo/Al) was in-pile tested to examine the effectiveness of the coating as a diffusion barrier between the U-7Mo fuel kernels and Al matrix. This paper reports the PIE data and analyses focusing on the effectiveness of the coating in terms of interaction layer (IL) growth and general fuel performance. The U-silicide coating showed considerable success, but it also provided evidence for additional improvement for coating process. The U-nitride coated specimen showed largely inefficient results in reducing IL growth. From the test, important observations were also made that can be utilized to improve U-Mo/Al fuel performance. The heating process for coating turned out to be beneficial to suppress fuel swelling. The use of larger fuel particles confirmed favorable effects on fuel performance.

Report of lower endplug welding, and testing and inspecting result for MONJU 1{sup th} reload core fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Kajiyama, Takasi; Numata, Kazuaki; Ohtani, Seiji [Quality Assuranse Section, Technical Administration Division, Plutonium Fuel Center, Tokai Works, Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan); Kobayashi, Hiromi; Watanabe, Hiroaki; Goto, Tatsuro; Takahashi, Hideki; Nagasaku, Katsuhiko [Inspection Development Campany Ltd., Tokai, Ibaraki (Japan)

2000-02-01

The procedure and result of lower endplugwelding, Test and Inspection and Shipment of the 1{sup th} reload core fuel assembly (80 Fuel Assemblies) for the fast breeder reactor MONJU are reported, which had been examined and inspected in Tamatsukuri Branch, Material Insurance Office, Quality Assurance Section, Technical Administration Division, Plutonium Fuel Center (before: Inspection Section, Plutonium Fuel Division), from June 1994 to January 1996. The number of cladding tubes welded to the endplug were totally 13,804: 7,418 for Core - Inside of 43 fuel Assemblies and 6,836 for Core-Outside of 37 fuel Assemblies. 13,794 of them, 7,414 Core-Inside and 6,379 Core-Outside, were approved by the test and sent to Plutonium Fuel Center. 10 of them weren't approved mainly because of default welding. Disapproval rating was 0.07%. (author)

Infant Brain Tumors: Incidence, Survival, and the Role of Radiation Based on Surveillance, Epidemiology, and End Results (SEER) Data

International Nuclear Information System (INIS)

Bishop, Andrew J.; McDonald, Mark W.; Chang, Andrew L.; Esiashvili, Natia

2012-01-01

Purpose: To evaluate the incidence of infant brain tumors and survival outcomes by disease and treatment variables. Methods and Materials: The Surveillance, Epidemiology, and End Results (SEER) Program November 2008 submission database provided age-adjusted incidence rates and individual case information for primary brain tumors diagnosed between 1973 and 2006 in infants less than 12 months of age. Results: Between 1973 and 1986, the incidence of infant brain tumors increased from 16 to 40 cases per million (CPM), and from 1986 to 2006, the annual incidence rate averaged 35 CPM. Leading histologies by annual incidence in CPM were gliomas (13.8), medulloblastoma and primitive neuroectodermal tumors (6.6), and ependymomas (3.6). The annual incidence was higher in whites than in blacks (35.0 vs. 21.3 CPM). Infants with low-grade gliomas had the highest observed survival, and those with atypical teratoid rhabdoid tumors (ATRTs) or primary rhabdoid tumors of the brain had the lowest. Between 1979 and 1993, the annual rate of cases treated with radiation within the first 4 months from diagnosis declined from 20.5 CPM to <2 CPM. For infants with medulloblastoma, desmoplastic histology and treatment with both surgery and upfront radiation were associated with improved survival, but on multivariate regression, only combined surgery and radiation remained associated with improved survival, with a hazard ratio for death of 0.17 compared with surgery alone (p = 0.005). For ATRTs, those treated with surgery and upfront radiation had a 12-month survival of 100% compared with 24.4% for those treated with surgery alone (p = 0.016). For ependymomas survival was higher in patients treated in more recent decades (p = 0.001). Conclusion: The incidence of infant brain tumors has been stable since 1986. Survival outcomes varied markedly by histology. For infants with medulloblastoma and ATRTs, improved survival was observed in patients treated with both surgery and early radiation

Analysis of cold flow fluidization test results for various biomass fuels

Energy Technology Data Exchange (ETDEWEB)

Abdullah, M.Z.; Husain, Z.; Pong, S.L.Y. [University Sains Malaysia, Penang (Malaysia). School of Mechanical Engineering

2003-07-01

A systematic theoretical and experimental study was conducted to obtain hydrodynamic properties such as particle size diameter, bulk density, fluidizing velocity, etc. for locally available biomass residue fuels in Malaysia like rice husk, sawdust, peanut shell, coconut shell, palm fiber as well as coal and bottom ash. The tests were carried out in a cold flow fluidization bed chamber of internal diameter 60 mm with air as fluidizing medium. Bed-pressure drop was measured as a function of superficial air velocity over a range of bed heights for each individual type of particle. The data were used to determine minimum fluidization velocity, which could be used to compare with theoretical values. The particle size of biomass residue fuel was classified according to Gildart's distribution diagram. The results show that Gildart's particle size (B) for sawdust, coal bottom ash, coconut shell have good fluidizing properties compared to rice husk, type (D) or palm fiber, type (A). The bulk density and voidage are found to be main factors contributing to fluidizing quality of the bed.

Fuel elements and safety engineering goals

International Nuclear Information System (INIS)

Schulten, R.; Bonnenberg, H.

1990-01-01

There are good prospects for silicon carbide anti-corrosion coatings on fuel elements to be realised, which opens up the chance to reduce the safety engineering requirements to the suitable design and safe performance of the ceramic fuel element. Another possibility offered is combined-cycle operation with high efficiencies, and thus good economic prospects, as with this design concept combining gas and steam turbines, air ingress due to turbine malfunction is an incident that can be managed by the system. This development will allow economically efficient operation also of nuclear power reactors with relatively small output, and hence contribute to reducing CO 2 emissions. (orig./DG) [de

Intermodal transfer of spent fuel

International Nuclear Information System (INIS)

Neuhauser, K.S.; Weiner, R.F.

1993-01-01

This paper discusses RADTRAN calculational models and parameter values for describing dose to workers during incident-free ship-to-truck transfer of spent fuel. Data obtained during observation of the offloading of research reactor spent fuel at Newport News Terminal in the Port of Hampton Roads, Virginia, are described. These data include estimates of exposure times and distances for handlers, inspectors, and other workers during offloading and overnight storage. Other workers include crane operators, scale operators, security personnel, and truck drivers. The data are compared to the default data in RADTRAN 4, and the latter are found to be conservative. The casks were loaded under IAEA supervision at their point of origin, and three separate radiological inspections of each cask were performed at the entry to the port (Hampton Roads) by the U.S. Coast Guard, the state of Virginia, and the shipping firm. As a result of the international standardization of containerized cargo handling in ports around the world, maritime shipment handling is particularly uniform. Thus, handler exposure parameters will be relatively constant for ship-truck and ship-rail transfers at ports throughout the world. Inspectors' doses are expected to vary because of jurisdictional considerations. The results of this study should be applicable to truck-to-rail transfers. (author)

Activity release from the damaged spent VVER-fuel during long-term wet storage

International Nuclear Information System (INIS)

Slonszki, E.; Hozer, Z.; Pinter, T.; Baracska Varju, I.

2010-01-01

An ex-core fuel damage incident took place at Unit 2 of Paks Nuclear Power Plant in Hungary on the 10 th April 2003. After this event the damaged fuel assemblies were stored under water for four years. During wet storage a continuous activity release was observed. The evaluation of the measured activity concentration showed that the UO 2 mass released from the fuel into the coolant was ∼ 1.8% of the total fuel mass. Furthermore this paper contains the calculation methods and the calculated activity release of the main analysed isotopes. (orig.)

Operation of marine diesel engines on biogenic fuels: modification of emissions and resulting climate effects.

Science.gov (United States)

Petzold, Andreas; Lauer, Peter; Fritsche, Uwe; Hasselbach, Jan; Lichtenstern, Michael; Schlager, Hans; Fleischer, Fritz

2011-12-15

The modification of emissions of climate-sensitive exhaust compounds such as CO(2), NO(x), hydrocarbons, and particulate matter from medium-speed marine diesel engines was studied for a set of fossil and biogenic fuels. Applied fossil fuels were the reference heavy fuel oil (HFO) and the low-sulfur marine gas oil (MGO); biogenic fuels were palm oil, soybean oil, sunflower oil, and animal fat. Greenhouse gas (GHG) emissions related to the production of biogenic fuels were treated by means of a fuel life cycle analysis which included land use changes associated with the growth of energy plants. Emissions of CO(2) and NO(x) per kWh were found to be similar for fossil fuels and biogenic fuels. PM mass emission was reduced to 10-15% of HFO emissions for all low-sulfur fuels including MGO as a fossil fuel. Black carbon emissions were reduced significantly to 13-30% of HFO. Changes in emissions were predominantly related to particulate sulfate, while differences between low-sulfur fossil fuels and low-sulfur biogenic fuels were of minor significance. GHG emissions from the biogenic fuel life cycle (FLC) depend crucially on energy plant production conditions and have the potential of shifting the overall GHG budget from positive to negative compared to fossil fuels.

Results on safety research for five years (from fiscal year 1996 to 2000). A field of nuclear fuel cycle

International Nuclear Information System (INIS)

2001-10-01

This safety research carried out by the Japan Nuclear Cycle Development Institute (JNC) for five years ranged from fiscal year 1996 to 2000, was performed according to the safety research basic plan (from fiscal year 1996 to 2000) established on March, 1996 (revised again on May, 2000). This report was arranged on a field on nuclear fuel cycle (all subjects on fields of nuclear fuel facility, environmental radioactivity and radioactive wastes and a subject on nuclear fuel cycle in a field of seismic resistant and probabilistic safety assessment) by combining its research results for five years ranged from 1996 to 2000 fiscal year with general outlines on the safety research basic plan. Here were shown outlines on the safety research basic plan, aims and subjects on safety research at a field of nuclear fuel cycle, a list of survey sheets on safety research result, and survey sheets on safety research results. The survey sheets containing research field, title, organization, researcher name, researching period, names of cooperative organization, using facilities, research outline, research results, established contents, application, and research trends, are ranged to 21 items on nuclear fuel facility, 1 item on seismic resistance, 2 items on probabilistic safety assessment, 8 items on environmental radioactivity, and 20 items on radioactive wastes. (G.K.)

Experimental results on the MOX fuel. Study of the calculus/measures divergences

International Nuclear Information System (INIS)

Martin, S.

1997-01-01

For each nuclear plant unit restart, all safety criterion have to be respected. Various parameters as boron concentration, temperature coefficient, worth or power and activity distributions related to fuel assemblies, have to be calculated. To compute these parameters Framatome uses the neutronic channel Science. Before the validation they are compare to experimental measures. For UO 2 fuel the divergence calculus/measures are correct. But for MOX fuels the divergence worsening. This paper discusses tis divergence and research the origin. (A.L.B.)

Nuclear fuel tax in court

International Nuclear Information System (INIS)

Leidinger, Tobias

2014-01-01

Besides the 'Nuclear Energy Moratorium' (temporary shutdown of eight nuclear power plants after the Fukushima incident) and the legally decreed 'Nuclear Energy Phase-Out' (by the 13th AtG-amendment), also the legality of the nuclear fuel tax is being challenged in court. After receiving urgent legal proposals from 5 nuclear power plant operators, the Hamburg fiscal court (4V 154/13) temporarily obliged on 14 April 2014 respective main customs offices through 27 decisions to reimburse 2.2 b. Euro nuclear fuel tax to the operating companies. In all respects a remarkable process. It is not in favour of cleverness to impose a political target even accepting immense constitutional and union law risks. Taxation 'at any price' is neither a statement of state sovereignty nor one for a sound fiscal policy. Early and serious warnings of constitutional experts and specialists in the field of tax law with regard to the nuclear fuel tax were not lacking. (orig.)

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

Energy Technology Data Exchange (ETDEWEB)

Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

2016-04-17

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated melting of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.

Reactivity change measurements on plutonium-uranium fuel elements in hector experimental techniques and results

International Nuclear Information System (INIS)

Tattersall, R.B.; Small, V.G.; MacBean, I.J.; Howe, W.D.

1964-08-01

The techniques used in making reactivity change measurements on HECTOR are described and discussed. Pile period measurements were used in the majority of oases, though the pile oscillator technique was used occasionally. These two methods are compared. Flux determinations were made in the vicinity of the fuel element samples using manganese foils, and the techniques used are described and an error assessment made. Results of both reactivity change and flux measurements on 1.2 in. diameter uranium and plutonium-uranium alloy fuel elements are presented, these measurements being carried out in a variety of graphite moderated lattices at temperatures up to 450 deg. C. (author)

Fuel rod failure during film boiling (PCM-1 test in the PBF)

International Nuclear Information System (INIS)

Domenico, W.F.; Stanley, C.J.; Mehner, A.S.

1978-01-01

The Power-Cooling-Mismatch (PCM) Test, PCM-1 was conducted in the Power Burst Facility (PFB) in March of 1978. The PCM Test Series is being conducted at the Idaho National Engineering Laboratory by EG and G Idaho, Inc., under contract to the USNRC and is designed to characterize the behavior of nuclear fuel rods operating under conditions of high power or low coolant flow or both leading to departure from nucleate boiling. The PCM-1 test was performed to provide in-pile data for a ''worst case'' PCM incident. The objective of this experiment was to study the behavior of a single pressurized water reactor (PWR) fuel rod subjected to a high-power and low flow environment which would result in cladding failure at full power. The ''worst case'' conditions established for the experiment consisted of a rod peak power of 78.7 kW/m and a coolant mass flux of 1356 kg/s.m 2 . Fuel temperatures at the stipulated operating conditions were such that a significant volume of molten fuel was present when failure occurred which produced a high probability of molten fuel-coolant interaction (MFCI) with the possibility of a vapor explosion

Diesel fuel long term storage and treatment- recommended tests and practices (U)

Energy Technology Data Exchange (ETDEWEB)

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

2009-06-05

The Clean Air Act (1970) is the comprehensive federal law that regulates air emissions from stationary and mobile sources. Among other things, this law authorized the Environmental Protection Agency (EPA) to establish National Ambient Air Quality Standards to protect public health and public welfare and to regulate emissions of hazardous air pollutants. In recent years, EPA regulations have forced oil refineries into producing a very low sulfur diesel fuel and incentives for adding up to 5% bio-diesel. These changes to the fuel oil formulation are beneficial to air quality and to energy conservation, but adversely impact heat content, long term storage stability, engine power, and injection system reliability. Diesel engines typically have a high incidence of injector failure resulting from poor diesel fuel quality. Since standby diesel engines do not run continuously it is necessary to implement periodic surveillance's to ensure the quality of diesel fuel is acceptable for reliable operation when a loss of power occurs. The information contained in this document is a compilation of best practices to be used as a guide for maintenance of a reliable diesel fuel system.

British Nuclear Fuels - a dirty business

International Nuclear Information System (INIS)

Bunyard, P.

1983-01-01

The radioactive discharges from British Nuclear Fuels Sellafield, Cumbria, reprocessing plant to the sea are discussed. Statements that have been made by various individuals and groups about the contamination of the sea, the coast and places inland, and the biological effects of plutonium and americium, are discussed in detail. Particular stress is placed on statements about increased incidence of cancers. (U.K.)

Major results on the development of high density U-Mo fuel and pin-type fuel elements executed under the Russian RERTR program and in cooperation with ANL (USA)

International Nuclear Information System (INIS)

Vatulin, A.; Morozov, A.; Stetsky, Y.; Suprun, V.; Dobrikova, I.; Trifonov, Y.; Mishunin, V.; Sorokin, V.

2003-01-01

VNIINM is active participant of 'Russian program on Reduced Enrichment for Research and Test Reactors'. Institute Works in two main directions: 1) development of new high-density fuels (HDF) and 2) development of new design of fuel elements with LEU. The development of the new type fuel element is carried out both for existing reactors, and for developing new advanced reactors. The 'TVEL' concern is coordinator of works of this program. The majority enterprises of branch (NIIAR, PIYaF, RRC KI, NZChK) take part in this work. Since 2000 these works are being conducted in cooperation with Argonne National Laboratory (USA) within the RERTR program under VNIINM with ANL contract. At the present, a large set of pre-pile investigations has been completed. All necessary fabrication procedures have been developed for utilization of U-Mo dispersion fuel in Russian-designed research reactors. For irradiation tests the pin-type mini-fuel elements with HDF dispersion fuel with LEU and the uranium density equaled to 4,0 and 6,0 g/cm 3 (up to 40 vol.%) have been manufactured. Their irradiation began in August 2003 in the MIR reactor (NIIAR, Dimitrovgrad). A large set of works for preparation of lifetime tests (WWR-M reactor in Gatchina) of two full-scale fuel assemblies with new pin-type fuel elements on basis LEU UO 2 -Al and UMo-Al fuels has been completed. The in-pile tests of fuel assemblies began in September 2003. The summary of important results of performed works and their near-term future are presented in paper. (author)

General Atomic HTGR fuel reprocessing pilot plant: results of initial sequential equipment operation

International Nuclear Information System (INIS)

1978-09-01

In September 1977, the processing of 20 large high-temperature gas-cooled reactor (LHTGR) fuel elements was completed sequentially through the head-end cold pilot plant equipment. This report gives a brief description of the equipment and summarizes the results of the sequential operation of the pilot plant. 32 figures, 15 tables

Airflow Obstruction and Use of Solid Fuels for Cooking or Heating: BOLD Results.

Science.gov (United States)

Amaral, André F S; Patel, Jaymini; Kato, Bernet S; Obaseki, Daniel O; Lawin, Hervé; Tan, Wan C; Juvekar, Sanjay K; Harrabi, Imed; Studnicka, Michael; Wouters, Emiel F M; Loh, Li-Cher; Bateman, Eric D; Mortimer, Kevin; Buist, A Sonia; Burney, Peter G J

2017-09-12

Evidence supporting the association of COPD or airflow obstruction with use of solid fuels is conflicting and inconsistent. To assess the association of airflow obstruction with self-reported use of solid fuels for cooking or heating. We analysed 18,554 adults from the BOLD study, who had provided acceptable post-bronchodilator spirometry measurements and information on use of solid fuels. The association of airflow obstruction with use of solid fuels for cooking or heating was assessed by sex, within each site, using regression analysis. Estimates were stratified by national income and meta-analysed. We carried out similar analyses for spirometric restriction, chronic cough and chronic phlegm. We found no association between airflow obstruction and use of solid fuels for cooking or heating (ORmen=1.20, 95%CI 0.94-1.53; ORwomen=0.88, 95%CI 0.67-1.15). This was true for low/middle and high income sites. Among never smokers there was also no evidence of an association of airflow obstruction with use of solid fuels (ORmen=1.00, 95%CI 0.57-1.76; ORwomen=1.00, 95%CI 0.76-1.32). Overall, we found no association of spirometric restriction, chronic cough or chronic phlegm with the use of solid fuels. However, we found that chronic phlegm was more likely to be reported among female never smokers and those who had been exposed for ≥20 years. Airflow obstruction assessed from post-bronchodilator spirometry was not associated with use of solid fuels for cooking or heating.

The influence of a non-uniform radial temperature distribution in the fuel on the results of calculation of transients

International Nuclear Information System (INIS)

Goltsev, A.O.; Davidenko, V.D.; Tsibulsky, V.F.; Lekomtsev, A.A.

2003-01-01

The paper is devoted to the discussion of results of computational studies of transients for different ways of accounting the temperature of the fuel in the full-scale comprehensive calculations of neutron physics. The paper demonstrates that in calculation of the neutron physics, it is necessary to use the effective temperature of the fuel in order to provide for correct accounting of the fuel temperature feedback, since the value of volume-averaged temperature being used in calculations of neutron physics with feedbacks would result in underestimation of consequences of accidents, especially accidents involving the dispersion of radiation

Incidence of syphilis seroconversion among HIV-infected persons in Asia: results from the TREAT Asia HIV Observational Database.

Science.gov (United States)

Ahn, Jin Young; Boettiger, David; Kiertiburanakul, Sasisopin; Merati, Tuti Parwati; Huy, Bui Vu; Wong, Wing Wai; Ditangco, Rossana; Lee, Man Po; Oka, Shinichi; Durier, Nicolas; Choi, Jun Yong

2016-01-01

Outbreaks of syphilis have been described among HIV-infected men who have sex with men (MSM) in Western communities, whereas reports in Asian countries are limited. We aimed to characterize the incidence and temporal trends of syphilis among HIV-infected MSM compared with HIV-infected non-MSM in Asian countries. Patients enrolled in the TREAT Asia HIV Observational Database cohort and with a negative non-treponemal test since enrolment were analyzed. Incidence of syphilis seroconversion, defined as a positive non-treponemal test after previously testing negative, was evaluated among patients at sites performing non-treponemal tests at least annually. Factors associated with syphilis seroconversion were investigated at sites doing non-treponemal testing in all new patients and subsequently testing routinely or when patients were suspected of having syphilis. We included 1010 patients from five sites that performed non-treponemal tests in all new patients; those included had negative non-treponemal test results during enrolment and subsequent follow-ups. Among them, 657 patients were from three sites conducting regular non-treponemal testing. The incidence of syphilis seroconversion was 5.38/100 person-years (PY). Incidence was higher in MSM than non-MSM (7.64/100 PY vs. 2.44/100 PY, psyphilis diagnosis (IRR 5.15, 95% CI 3.69-7.17) and younger age (IRR 0.84 for every additional 10 years, 95% CI 0.706-0.997) were significantly associated with syphilis seroconversion. We observed a higher incidence of syphilis seroconversion among HIV-infected MSM and a trend to increasing annual incidence. Regular screening for syphilis and targeted interventions to limit transmission are needed in this population.

Public information circular for shipments of irradiated reactor fuel

International Nuclear Information System (INIS)

1983-07-01

This publication contains basically three kinds of information: routes approved by the Commission for the shipment of irradiated reactor fuel, information regarding any safeguards-significant incidents which have been reported to occur during shipments along such routes, and cumulative amounts of material shipped

Status of LEU fuel development and conversion of NRU

International Nuclear Information System (INIS)

Sears, D.F.; Herbert, L.N.; Vaillancourt, K.D.

1989-11-01

The status of the low-enrichment uranium (LEU) fuel development and NRU conversion program at Chalk River Nuclear Laboratories is reviewed. Construction of a new fuel fabrication facility is essentially completed and installation of LEW fuel manufacturing equipment has begun. The irradiation of 31 prototype Al-61 wt% U 3 Si dispersion fuel rods, approximately one third of a full NRU core, is continuing without incident. Recent post-irradiation examination of spent fuel rods revealed that the prototype LEU fuel achieved the design burnup (80 at%) in excellent condition, confirming that the Al-U 3 Si 2 dispersion fuel to complement out Al-U 3 Si capability. Three full-size NRU rods containing Al-U 3 Si 2 dispersion fuel have been fabricated for a qualification irradiation in NRU. Post-irradiation examinations of mini-elements containing Al-U 3 Si 2 fuel revealed that the U 3 Si 2 behaved similarly to U 3 Si 2 fuel revealed that the U 3 Si 2 particles and the aluminum matrix, and fission gas bubbles up to 10 μm in diameter, could be seen in the particles after 60 at% and 80 at% burnup. The mini-elements contained a variety of silicide particle sizes; however, no significant swelling dependence on particle size distribution was observed

Results and comments on the gamma spectrometry examinations carried out on PWR and fast neutron fuel elements

International Nuclear Information System (INIS)

Pineira, Thomas; Mouchnino, Michel; Juste, Guy; Vignesoult, Nicole.

1980-05-01

The gamma spectrometry analyses on the fuel elements of PWR and fast neutron systems have experienced a significant growth in the CEA. This nondestructive, quick, inexpensive and quantitative method, seems to us particularly advantageous for qualifying the behavior of fuel under irradiation. However, in order to use it to the maximum, it must have reached a high degree of automation and the interpretation of the results must be the outcome of a coherent team that includes gamma spectrometry and fuel element specialists, since the growth of hot cell gamma spectrometry involves the processing of a considerable number of data upon which the quality of the results depends (large number of spectra per pencil analyzed, dimension of the 2000 or 4000 channel spectra, number of lines studied, etc.). Therefore the need to make the most of the information and, in particular, to present the results in a form suitable for direct processing in a minimum response time, requires a highly automated system. Further, the more specific results of gamma spectrometry correlated to the metallurgical results obtained in the laboratories should contribute efficiently to obtaining major information [fr

Exploration for fossil and nuclear fuels from orbital altitudes. [results of ERTS program for oil exploration

Science.gov (United States)

Short, N. M.

1974-01-01

Results from the ERTS program pertinent to exploration for oil, gas, and uranium are discussed. A review of achievements in relevant geological studies from ERTS, and a survey of accomplishments oriented towards exploration for energy sources are presented along with an evaluation of the prospects and limitations of the space platform approach to fuel exploration, and an examination of continuing programs designed to prove out the use of ERTS and other space system in exploring for fuel resources.

Nuclear fuels technologies Fiscal Year 1996 research and development test results

International Nuclear Information System (INIS)

Beard, C.A.; Blair, H.T.; Buksa, J.J.; Butt, D.P.; Chidester, K.; Eaton, S.L.; Farish, T.J.; Hanrahan, R.J.; Ramsey, K.B.

1996-11-01

During fiscal year 1996, the Department of Energy's Office of Fissile Materials Disposition (OFMD) funded Los Alamos National Laboratory (LANL) to investigate issues associated with the fabrication of plutonium from dismantled weapons into mixed-oxide (MOX) nuclear fuel for disposition in nuclear power reactors. These issues can be divided into two main categories: issues associated with the fact that the plutonium from dismantled weapons contains gallium, and issues associated with the unique characteristics of the PuO[sub 2] produced by the dry conversion process that OFMD is proposing to convert the weapons material. Initial descriptions of the experimental work performed in fiscal year 1996 to address these issues can be found in Nuclear Fuels Technologies Fiscal Year 1996 Research and Development Test Matrices. However, in some instances the change in programmatic emphasis towards the Parallex program either altered the manner in which some of these experiments were performed (i.e., the work was done as part of the Parallex fabrication development and not as individual separate-effects tests as originally envisioned) or delayed the experiments into Fiscal Year 1997. This report reviews the experiments that were conducted and presents the results

Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

Energy Technology Data Exchange (ETDEWEB)

Firnhaber, M. [Gesellschaft fuer Anlagen-und Reaktorsicherheit, Koeln (Germany); Yegorova, L. [Nuclear Safety Institute of Russian Research Center, Moscow (Russian Federation); Brockmeier, U. [Ruhr-Univ. of Bochum (Germany)] [and others

1995-09-01

International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement.

Activity release from the damaged spent VVER-fuel during long-term wet storage

Energy Technology Data Exchange (ETDEWEB)

Slonszki, E.; Hozer, Z. [Hungarian Academy of Sciences, KFKI Atomic Energy Research Inst., Budapest (Hungary); Pinter, T.; Baracska Varju, I. [Nuclear Power Plant Paks, Paks (Hungary)

2010-07-01

An ex-core fuel damage incident took place at Unit 2 of Paks Nuclear Power Plant in Hungary on the 10{sup th} April 2003. After this event the damaged fuel assemblies were stored under water for four years. During wet storage a continuous activity release was observed. The evaluation of the measured activity concentration showed that the UO{sub 2} mass released from the fuel into the coolant was {approx} 1.8% of the total fuel mass. Furthermore this paper contains the calculation methods and the calculated activity release of the main analysed isotopes. (orig.)

Adverse incidents resulting in exposure to body fluids at a UK dental teaching hospital over a 6-year period

Directory of Open Access Journals (Sweden)

Hughes A

2012-10-01

Full Text Available A Hughes,1 L Davies,1 R Hale,1 JE Gallagher21Kings College Hospital NHS Foundation Trust, 2King's College London Dental Institute, London, United KingdomBackground: The safety and protection of patients and health care workers is of paramount importance in dentistry, and this includes students in training who provide clinical care. Given the nature of dental care, adverse incidents can and do occur, exposing health care workers to body fluids and putting them at risk of infection, including contracting a blood-borne virus. The aim of this research was to analyze trends in the volume, rate, nature, management, and outcome of adverse incidents reported at one dental teaching hospital from 2005 to 2010.Methods: Descriptive analysis of trends in the volume, rate, nature, management, and outcome of adverse incidents reported at one dental teaching hospital over a six-year period was undertaken in relation to the level of outpatient and day surgery activity.Results: In total, 287 incidents were reported over a six-year period, which amounted to 0.039% of outpatient or day surgery appointments. Nearly three quarters of all the incidents (n = 208, 72% took place during treatment or whilst clearing away after the appointment. The most frequent incidents were associated with administration of local anesthetic (n = 63, 22%, followed by burs used in dental hand pieces (n = 51, 18%.Conclusion: This research confirms that adverse incidents are a feature of dental hospitals and reports the common sources. The importance of accurate and consistent reporting of data to ensure that these issues are monitored to inform action and reduce risks to staff, students, and patients are highlighted.Keywords: risk management, blood-borne virus, dental hospital, body fluids exposure, adverse event reporting

The prevalence, incidence, and gender and age-specific incidence of problem gambling: results of the Swedish longitudinal gambling study (Swelogs).

Science.gov (United States)

Abbott, Max; Romild, Ulla; Volberg, Rachel

2018-04-01

To estimate the prevalence, incidence and gender and age-specific incidence of problem gambling in the Swedish adult population. Longitudinal cohort study with linkage to register data. Sweden. Stratified random sample aged 16-84 years at baseline (n = 8165) re-assessed a year later (n = 6021). Problem gambling (life-time and past 12 months) was measured by the South Oaks Gambling Screen-Revised (SOGS-R). Past 12-month (current) problem gambling was also measured by the Problem Gambling Severity Index (PGSI). The SOGS-R combined current pathological and problem gambling prevalence rate (PR) was 2.1 [95% confidence interval (CI) = 1.8-2.4] at baseline and 1.7 (1.4-2.0) at follow-up, approximately half the corresponding life-time estimates.[Correction added on 22 Dec 2017, after first online publication: In the preceding sentence, the SOGS-R combined current pathological and problem gambling prevalence rate (PR) was incorrectly reported as being double the corresponding life-time rate. It has been corrected in this version.] PGSI combined current problem and moderate-risk gambling PRs were 2.2 (1.9-2.5) at baseline and 1.9 (1.6-2.2) at follow-up. Combined incidence rates (IRs) were 1.0 (0.8-1.3) (SOGS-R) and 1.4 (1.1-1.7) (PGSI), with more than three-quarters being new cases. While first-time IRs did not vary by gender, males had a higher relapse IR and proportionately more females were new cases. The young adult IR was more than double the older adult IR; similar proportions were new cases. The actual incidence of problem gambling relapse in Sweden is likely to be higher than estimated. The profile of problem gambling in Sweden is likely to change over time, with increased proportions of women and older adults. © 2017 Society for the Study of Addiction.

Post-test investigation result on the WWER-1000 fuel tested under severe accident conditions

International Nuclear Information System (INIS)

Goryachev, A.; Shtuckert, Yu.; Zwir, E.; Stupina, L.

1996-01-01

The model bundle of WWER-type were tested under SFD condition in the out-of-pile CORA installation. The objective of the test was to provide an information on the WWER-type fuel bundles behaviour under severe fuel damage accident conditions. Also it was assumed to compare the WWER-type bundle damage mechanisms with these experienced in the PWR-type bundle tests with aim to confirm a possibility to use the various code systems, worked our for PWR as applied to WWER. In order to ensure the possibility of the comparison of the calculated core degradation parameters with the real state of the tested bundle, some parameters have been measured on the bundle cross-sections under examination. Quantitative parameters of the bundle degradation have been evaluated by digital image processing of the bundle cross-sections. The obtained results are shown together with corresponding results obtained by the other participants of this investigation. (author). 3 refs, 13 figs

Leak testing fuel stored in the ICPP fuel storage basin

International Nuclear Information System (INIS)

Lee, J.L.; Rhodes, D.W.

1977-06-01

Irradiated fuel to be processed at the Idaho Chemical Processing Plant is stored under water at the CPP-603 Fuel Storage Facility. Leakage of radionuclides through breaks in the cladding of some of the stored fuels contaminates the water with radionuclides resulting in radiation exposure to personnel during fuel handling operations and contamination of the shipping casks. A leak test vessel was fabricated to test individual fuel assemblies which were suspected to be leaking. The test equipment and procedures are described. Test results demonstrated that a leaking fuel element could be identified by this method; of the eleven fuel assemblies tested, six were estimated to be releasing greater than 0.5 Ci total radionuclides/day to the basin water

Mechanical guided waves for fuel level monitoring system

Directory of Open Access Journals (Sweden)

Tiberiu Adrian SALAORU

2017-09-01

Full Text Available The mechanical guided waves have a wide range of applications in many types of equipment and devices. The fuel level is an important parameter which needs to be monitored for a vehicle which can be a space vehicle, an aircraft or any other. For this purpose mechanical guided waves can be used as they have several major advantages over any other methods. There are a wide ultrasonic sensors used for this purpose but in the most cases the mechanical waves are traveling through air or fuel for measuring their level. In general the wave propagation through a single media at a time is utilized. The method described in this work uses the propagation of the mechanical guided waves through two different media in the same time. The propagating media is the container wall and the other is the fuel. One of the advantages of this method is the reduction of the measurement errors when the incident angle to the fuel level surface is different from 90 degree. These situations could occur when the fuel tank is tilted or when the fuel surface is not flat. This measurement method will not be affected by these conditions.

Accelerated fuel depreciation as an economic incentive for low-leakage fuel management

International Nuclear Information System (INIS)

Downar, T.J.

1986-01-01

An analysis is presented which evaluates the tax depreciation advantage which results from the increased rate of fuel depletion achieved in the current low-leakage fuel-management LWR core reload designs. An analytical fuel-cycle cost model is used to examine the important cost parameters which are then validated using the fuel-cycle cost code CINCAS and data from the Maine Yankee PWR. Results show that low-leakage fuel management, through the tax depreciation advantage from accelerated fuel depletion, provides an improvement of several percent in fuel-cycle costs compared to traditional out-in fuel management and a constant fuel depletion rate. (author)

A fuel thermal conductivity correlation based on the latest experimental results

International Nuclear Information System (INIS)

Sontheimer, F.; Landskron, H.; Billaux, M.R.

2000-01-01

A new fuel thermal conductivity (ftc) correlation for UO 2 and (U,Gd)O 2 is presented, which is based on the relaxation-time theory of Klemens. The correlation is chosen because of its validity in a wide range of defect concentrations as for instance encountered in fuel with a wide range of burnup and gadolinia additions, as has been shown by Ishimoto. The phonon term of the new correlation has the form 1/x·arctan(x) , where x is a measure of the defect concentration introduced by burnup and gadolinia additions. For low defect concentrations, this term is identical with the classical form for the phonon term 1/(A+B.T). At high defect concentrations, however, when phonon-point defect scattering starts dominating over phonon-phonon scattering, the new correlation deviates from the classical formulation and has a distinctly weaker dependence on temperature and defect concentration than the classical form. The new arctan correlation in combination with an appropriate electronic ftc term is fitted to the Halden data base of fuel centre-line temperature measurements (represented by the ''Halden ftc correlation recommendation''). Agreement is very good up to a burnup of about 60 MWd/kgU; beyond, the arctan form has a saturating burnup degradation. The new arctan correlation in combination with an appropriate electronic ftc term is also shown to describe very well our latest ftc measurements on unirradiated gadolinia fuel up to 9% gadolinia content. Application to Halden measurements up to very high burnup is successful, when combined with the so-called ''rim-effect'', which counteracts the saturation tendency of the new correlation at high burnup. Latest laser thermal diffusivity measurements on irradiated gadolinia fuel in the frame of the NFIR program, although not yet open for literature and not discussed in the paper, indicate very good agreement with the new arctan correlation. (author)

Fuel cycle cost comparisons with oxide and silicide fuels

Energy Technology Data Exchange (ETDEWEB)

Matos, J E; Freese, K E [RERTR Program, Argonne National Laboratory (United States)

1983-09-01

This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. The status of the development and demonstration of the oxide and silicide fuels are presented in several papers in these proceedings. Routine utilization of these fuels with the uranium densities considered here requires that they are successfully demonstrated and licensed. Thermal-hydraulic safety margins, shutdown margins, mixed cores, and transient analyses are not addressed here, but analyses of these safety issues are in progress for a limited number of the most promising design options. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data is presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed. All safety criteria for the reactor with these fuel element designs need to be satisfied as well. With LEU oxide fuel, 31 g U/cm{sup 3} 1 and 0.76 mm--thick fuel meat, elements with 18-22 plates 320-391 g {sup 235}U) result in the same or lower total costs than with the HEU element 23 plates, 280 g {sup 235}U). Higher LEU loadings (more plates per element) are needed for larger excess reactivity requirements. However, there is little cost advantage to using more than 20 of these plates per element. Increasing the fuel meat thickness from 0.76 mm to 1.0 mm with 3.1 g U/cm{sup 3} in the design with 20 plates per element could result in significant cost reductions if the

Emergency response arrangements for the transport of irradiated nuclear fuel from Japan to Europe in Japanese territorial waters

International Nuclear Information System (INIS)

Ikeda, T.; Inada, T.; Narahara, S.; Cheshire, R.D.; Lee, G.

1993-01-01

About 90 % of nuclear fuel irradiated in Japanese nuclear power stations is transported to UK and France for reprocessing. Pacific Nuclear Transport Ltd (PNTL), a subsidiary of British Nuclear Fuels plc (BNFL), owns and operates its own fleet of 5 purpose built ships specially designed for the transport of flasks containing irradiated fuel from Japan to Europe. These vessels sail to Japan on 8 to 10 voyages per year from the BNFL's Marine Terminal at Barrow in UK via Cherbourg Port in France. On arrival in Japan empty flasks are delivered to Japanese nuclear power stations, and full flasks are collected for the return journey to Europe. Whilst the probability of a serious flask incident involving the release of radioactivity is very small, it is nevertheless important to plan for such an emergency. In the case of an incident BNFL will provide an emergency response. If an incident occurs in Japanese territorial waters, the initial response will be provided by Nuclear Services Company (NSC), who are based in Japan (the head office in Tokyo, Tokai Office in Ibaraki Prefecture and Tsuruga Office in Fukui Prefecture) and contracted to BNFL to provide a similar response to that available from UK. This paper describes the communication links which have been established between UK and Japan and the internal communication within Japan. It also describes the emergency equipent held in Japan, the training of teams and the results of exercises jointly carried out with BNFL. (J.P.N.)

Initial results from dissolution rate testing of N-Reactor spent fuel over a range of potential geologic repository aqueous conditions

International Nuclear Information System (INIS)

Gray, W.J.; Einziger, R.E.

1998-04-01

Hanford N-Reactor spent nuclear fuel (HSNF) may ultimately be placed in a geologic repository for permanent disposal. To determine whether the engineered barrier system that will be designed for emplacement of light-water-reactor (LWR) spent fuel will also suffice for HSNF, aqueous dissolution rate measurements were conducted on the HSNF. The purpose of these tests was to determine whether HSNF dissolves faster or slower than LWR spent fuel under some limited repository-relevant water chemistry conditions. The tests were conducted using a flowthrough method that allows the dissolution rate of the uranium matrix to be measured without interference by secondary precipitation reactions that would confuse interpretation of the results. Similar tests had been conducted earlier with LWR spent fuel, thereby allowing direct comparisons. Two distinct corrosion modes were observed during the course of these 12 tests. The first, Stage 1, involved no visible corrosion of the test specimen and produced no undissolved corrosion products. The second, Stage 2, resulted in both visible corrosion of the test specimen and left behind undissolved corrosion products. During Stage 1, the rate of dissolution could be readily determined because the dissolved uranium and associated fission products remained in solution where they could be quantitatively analyzed. The measured rates were much faster than has been observed for LWR spent fuel under all conditions tested to date when normalized to the exposed test specimen surface areas. Application of these results to repository conditions, however, requires some comparison of the physical conditions of the different fuels. The surface area of LWR fuel that could potentially be exposed to repository groundwater is estimated to be approximately 100 times greater than HSNF. Therefore, when compared on the basis of mass, which is more relevant to repository conditions, the HSNF and LWR spent fuel dissolve at similar rates

Public information circular for shipments of irradiated reactor fuel

International Nuclear Information System (INIS)

1982-06-01

This publication is the third in a proposed series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. Subsequent issues in this series will update the information contained herein. This publication contains basically three kinds of information: (1) routes approved by the Commission for the shipment of irradiated reactor fuel, (2) information regarding any safeguards-significant incidents which have been reported to occur during shipments along such routes, and (3) cumulative amounts of material shipped

Nuclear fuel elements

International Nuclear Information System (INIS)

Nakai, Keiichi

1983-01-01

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

Chemical incidents resulted in hazardous substances releases in the context of human health hazards.

OpenAIRE

Palaszewska-Tkacz, Anna; Czerczak, Sławomir; Konieczko, Katarzyna

2017-01-01

Objectives: The research purpose was to analyze data concerning chemical incidents in Poland collected in 1999–2009 in terms of health hazards. Material and Methods: The data was obtained, using multimodal information technology (IT) system, from chemical incidents reports prepared by rescuers at the scene. The final analysis covered sudden events associated with uncontrolled release of hazardous chemical substances or mixtures, which may potentially lead to human exposure. Releases of uniden...

Some elaborating methods of gamma scanning results on irradiated nuclear fuels

International Nuclear Information System (INIS)

Sternini, E.

1979-01-01

Gamma scanning, as a post-irradiation examination, is a technique which provides a large number of informations on irradiated nuclear fuels. Power profile, fission products distribution, average and local burn-up of single elements structural and nuclear behaviour of fuel materials are examples of the obtained informations. In the present work experimental methods and theoretical calculations used at the CNEN hot cell laboratory for the mentioned purposes are described. Errors arising from the application of the gamma scanning technique are also discussed

FFTF initial fuel loading, preanalyses, and comparison with preliminary results

International Nuclear Information System (INIS)

Rothrock, R.B.; Daughtry, J.W.; Zimmerman, B.D.; Petrowicz, N.E.; Bennett, R.A.; Ombrellaro, P.A.

1980-02-01

Disadvantages of conventional loading from the center out were circumvented by loading one trisector at a time, and connecting the control rod drivelines in each sector after it was loaded so that the rods could be operated during the loading of subsequent trisectors. This sequence was interrupted once during the loading of the final sector, to achieve initial criticality at an approximately minimum critical loading and to measure absolute subcriticality by the rod drop technique. An in-core detector was preferable to the standard FTR ex-core detectors for monitoring the initial fuel loading. Consequently, special fission chambers were installed in an instrument thimble near the core center to monitor the initial fuel loading

Fuel cycle cost study with HEU and LEU fuels

International Nuclear Information System (INIS)

Matos, J.E.; Freese, K.E.

1984-01-01

Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors

The Cost of SOx Limits to Marine Operators; Results from Exploring Marine Fuel Prices

Directory of Open Access Journals (Sweden)

Orestis Schinas

2013-06-01

Full Text Available Marine operators are confronted with the new air emissions regulations that determine the limits of sulfur content in marine fuels. The low-sulfur (LS marine fuels have a higher price, and their fluctuation is almost similar to the fluctuation of high-sulfur (HS fuels. The price difference between HS and LS might also determine the decision of operators for alternative technical means, such as scrubbers, in order to comply with the new limits. This paper aims to provide a thorough statistical analysis of the currently available LS and HS marine fuels time series, as well as to present the analysis of the differential of the HS and LS fuel prices. The paper concludes with suggestions for further research.

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)

Communicating risk: news media reportage of a significant nuclear contamination incident in the UK

International Nuclear Information System (INIS)

Walker, G.P.; Macgill, S.M.

1988-02-01

The way the mass media reported a controversial discharge incident at British Nuclear Fuel's reprocessing plant at Sellafield in Cumbria in November 1983 is discussed. Apart from the Windscale fire in 1957 this was the most serious radiation incident at any nuclear installation in the United Kingdom. The aim of the study was to use that incident to develop a more complete understanding of the way information about risk is communicated through the news media. Two issues, in particular, are assessed. First, the accuracy of the reports of events and how factual, technical or scientific aspects of risk are portrayed through the news media. Secondly, risk evaluation is studied: what value judgements are made, what sentiments of alarm, danger and hazard are presented. (author)

Operation of Marine Diesel Engines on Biogenic Fuels: Modification of Emissions and Resulting Climate Effects

OpenAIRE

Petzold, A.; Lauer, P.; Fritsche, U.; Hasselbach, J.; Lichtenstern, M.; Schlager, H.; Fleischer, F.

2011-01-01

The modification of emissions of climate-sensitive exhaust compounds such as CO2, NOx, hydrocarbons, and particulate matter from medium-speed marine diesel engines was studied for a set of fossil and biogenic fuels. Applied fossil fuels were the reference heavy fuel oil (HFO) and the low-sulfur marine gas oil (MGO); biogenic fuels were palm oil, soybean oil, sunflower oil, and animal fat. Greenhouse gas (GHG) emissions related to the production of biogenic fuels were treated by means of a fue...

Statement of incidents at nuclear installations: third quarter 1992

International Nuclear Information System (INIS)

1993-01-01

Three incidents are reported for the third quarter of 1992. During a radiological survey of British Nuclear Fuel's site at Sellafield in June, contamination of the ground under a cracked pipebridge was found. Contamination of two workers was removed by washing; the contaminated soil was removed and contained in drums. In September on the same site, a pipe failure occurred and plutonium nitrate leaked into the secondary containment cell leading to a shutdown of the reprocessing plant. However, no discharge of radioactivity to the environment and no additional radiation exposure to workers occurred. This was subsequently classified as a level 3 incident. 25 spots of radioactive contamination of a service road at the United Kingdom Atomic Energy Authority's Winfrith site were removed and disposed of without injury or contamination. Recommendations to improve the site roads and car parks were made. (UK)

Nuclear fuel management via fuel quality factor averaging

International Nuclear Information System (INIS)

Mingle, J.O.

1978-01-01

The numerical procedure of prime number averaging is applied to the fuel quality factor distribution of once and twice-burned fuel in order to evolve a fuel management scheme. The resulting fuel shuffling arrangement produces a near optimal flat power profile both under beginning-of-life and end-of-life conditions. The procedure is easily applied requiring only the solution of linear algebraic equations. (author)

Fuel Cell Electric Bus Evaluations | Hydrogen and Fuel Cells | NREL

Science.gov (United States)

Bus Evaluations Fuel Cell Electric Bus Evaluations NREL's technology validation team evaluates fuel cell electric buses (FCEBs) to provide comprehensive, unbiased evaluation results of fuel cell bus early transportation applications for fuel cell technology. Buses operate in congested areas where

Fuel and fuel cycles with high burnup for WWER reactors

International Nuclear Information System (INIS)

Chernushev, V.; Sokolov, F.

2002-01-01

The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

Increased fuel burn-up and fuel cycle equilibrium

International Nuclear Information System (INIS)

Debes, M.

2001-01-01

Improvement of nuclear competitiveness will rely mainly on increased fuel performance, with higher burn-up, and reactors sustained life. Regarding spent fuel management, the EDF current policy relies on UO 2 fuel reprocessing (around 850 MTHM/year at La Hague) and MOX recycling to ensure plutonium flux adequacy (around 100 MTHM/year, with an electricity production equivalent to 30 TWh). This policy enables to reuse fuel material, while maintaining global kWh economy with existing facilities. It goes along with current perspective to increase fuel burn-up up to 57 GWday/t mean in 2010. The following presentation describes the consequences of higher fuel burn-up on fuel cycle and waste management and implementation of a long term and global equilibrium for decades in spent fuel management resulting from this strategy. (author)

Fuel-cycle cost comparisons with oxide and silicide fuels

International Nuclear Information System (INIS)

Matos, J.E.; Freese, K.E.

1982-01-01

This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data are presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed

AGR-1 Fuel Compact 6-3-2 Post-Irradiation Examination Results

Energy Technology Data Exchange (ETDEWEB)

Paul demkowicz; jason Harp; Scott Ploger

2012-12-01

Destructive post-irradiation examination was performed on fuel Compact 6-3-2, which was irradiated in the AGR-1 experiment to a final compact average burnup of 11.3% FIMA and a time-average, volume-average temperature of 1070°C. The analysis of this compact was focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, measurement of fuel burnup by several methods, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy. A single particle with a defective SiC layer was identified during deconsolidation-leach-burn-leach analysis, which is in agreement with previous measurements showing elevated cesium in the Capsule 6 graphite fuel holder associated with this fuel compact. The fraction of the compact europium inventory released from the particles and retained in the matrix was relatively high (approximately 6E-3), indicating release from intact particle coatings. The Ag-110m inventory in individual particles exhibited a very broad distribution, with some particles retaining =80% of the predicted inventory and others retaining less than 25%. The average degree of Ag-110m retention in 60 gamma counted particles was approximately 50%. This elevated silver release is in agreement with analysis of silver on the Capsule 6 components, which indicated an average release of 38% of the Capsule 6 inventory from the fuel compacts. In spite of the relatively high degree of silver release from the particles, virtually none of the Ag-110m released was found in the compact matrix, and presumably migrated out of the compact and was deposited on the irradiation capsule components. Release of all other fission products from the particles appears to be less than a single

Safety aspects of LWR fuel reprocessing and mixed oxide fuel fabrication plants

International Nuclear Information System (INIS)

Fischer, M.; Leichsenring, C.H.; Herrmann, G.W.; Schueller, W.; Hagenberg, W.; Stoll, W.

1977-01-01

The paper is focused on the safety and the control of the consequences of credible accidents in LWR fuel reprocessing plants and in mixed oxide fuel fabrication plants. Each of these plants serve for many power reactor (about 50.000 Mwel) thus the contribution to the overall risk of nuclear energy is correspondingly low. Because of basic functional differences between reprocessing plants, fuel fabrication plants and nuclear power reactors, the structure and safety systems of these plants are different in many respects. The most important differences that influence safety systems are: (1) Both fuel reprocessing and fabrication plants do not have the high system pressure that is associated with power reactors. (2) A considerable amount of the radioactivity of the fuel, which is in the form of short-lived radionuclides has decayed. Therefore, fuel reprocessing plants and mixed oxide fuel fabrication plants are designed with multiple confinement barriers for control of radioactive materials, but do not require the high-pressure containment systems that are used in LWR plants. The consequences of accidents which may lead to the dispersion of radioactive materials such as chemical explosions, nuclear excursions, fires and failure of cooling systems are considered. A reasonable high reliability of the multiple confinement approach can be assured by design. In fuel reprocessing plants, forced cooling is necessary only in systems where fission products are accumulated. However, the control of radioactive materials can be maintained during normal operation and during the above mentioned accidents, if the dissolver off-gas and vessel off-gas treatment systems provide for effective removal of radioactive iodine, radioactive particulates, nitrogen oxides, tritium and krypton 85. In addition, the following incidents in the dissolver off-gas system itself must be controlled: failures of iodine filters, hydrogen explosion in O 2 - and NOsub(x)-reduction component, decomposition of

Fuel motion in overpower tests of metallic integral fast reactor fuel

International Nuclear Information System (INIS)

Rhodes, E.A.; Bauer, T.H.; Stanford, G.S.; Regis, J.P.; Dickerman, C.E.

1992-01-01

In this paper results from hodoscope data analyses are presented for transient overpower (TOP) tests M5, M6, and M7 at the Transient Reactor Test Facility, with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding breach and prefailure elongation of D9-clad ternary (U-Pu-Zr) integral fast reactor-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT-9-clad binary (U-Zr) Fast Flux Test Facility driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure

Protection system for minimizing the consequences of a flow blockage incident at a pool-type research reactor

International Nuclear Information System (INIS)

de Vries, J.W.; van Dam, H.; Gysler, G.

1990-01-01

Safety analysis activities were performed for the HOR, a pool-type research reactor with plate-type fuel elements and a maximum licensed power of 3 MW. Following internationally accepted guidelines, a wide variety of possible process disturbances has been considered. For the HOR the most aggravating accident conditions could result from a sudden flow blockage of cooling channels. If this event occurs in the high power density region of the core, a decrease of the hot channel flow either causes flow reversal or prompts burnout. Unless the reactor is scrammed in time, the fuel plates will heat up rapidly and local melting will occur with possible propagation of voiding and burnout to adjacent channels. In the analysis, melting of the cladding has been considered by using a simplified model approach. The number of voided coolant channels, as well as the propagation rate of fuel plates reaching locally the melting temperature, were calculated for different conditions of operation. In order to reduce the risk of a fuel melt accident occurring at the HOR, the protection system features a special design option. The system recognizes cooling channel voiding by detection of a sudden decrease of neutron flux. In the present work, it has been shown that a flow blockage incident can be detected in the early stages of development. Also, in accordance with the results of experimental tests, it can be concluded that in many cases melting of fuel plates will be effectively prevented. If such an accident occurs on a very fast time scale, at least the radiological consequences are significantly mitigated by preventing propagation, thus limiting the number of molten fuel plates

Detecting Terrorism Incidence Type from News Summary

DEFF Research Database (Denmark)

Nizamani, Sarwat; Memon, Nasrullah

2012-01-01

The paper presents the experiments to detect terrorism incidence type from news summary data. We have applied classification techniques on news summary data to analyze the incidence and detect the type of incidence. A number of experiments are conducted using various classification algorithms...... and results show that a simple decision tree classifier can learn incidence type with satisfactory results from news data....

Increasing TRIGA fuel lifetime with 12 wt.% U TRIGA fuel

Energy Technology Data Exchange (ETDEWEB)

Naughton, W F; Cenko, M J; Levine, S H; Witzig, W F [Pennsylvania State University (United States)

1974-07-01

In-core fuel management studies have been performed for the Penn State Breazeale Reactor (PSBR) wherein 12 wt % U fuel elements are used to replace the standard 8.5 wt % U TRIGA fuel. The core configuration used to develop a calculational model was a 90-element hexagonal array, which is representative of the PSBR core, and consists of five hexagonal rings surrounding a central thimble containing water. The technique employed for refueling the core fully loaded with 8.5 wt % U fuel involves replacing 8.5 wt % U fuel with 12 wt % U fuel using an in-out reloading scheme. A batch reload consists of 6 new 12 wt % U fuel elements. Placing the 12 wt % U fuel in the B ring produces fuel temperatures ({approx}450 {sup o}C) that are well below the 800{sup o}C maximum limitation when the PSBR is operating at its maximum allowed power of 1 Megawatt. The advantages of using new 12 wt % U fuel to replace the burned up 8.5 wt % U fuel in the B ring over refueling strictly with 8.5 wt % U-Zr TRIGA fuel are clearly delineated in Table 1 where cost calculations used the General Atomic pre-1972 prices for TRIGA fuel, i.e., $1500 and $1650 for an 8.5 and 12 wt % U fuel element, respectively. Experimental results obtained to date utilizing the 12 wt % U fuel elements agree with the computed results. (author)

Savannah River Laboratory data banks for risk assessment of fuel reprocessing plants

International Nuclear Information System (INIS)

Durant, W.S.

1981-10-01

The Savannah River Laboratory maintains a series of computerized data banks primarily as an aid in probabilistic risk assessment studies in the fuel reprocessing facilities. These include component failure rates, generic incidents, and reports of specific deviations from normal operating conditions. In addition to providing data for probability studies, these banks, have served as a valuable aid in trend analysis, equipment histories, process hazards analysis, consequence assessments, incident audit, process problem solving, and training

Transient feedback from fuel motion in metal IFR [Integral Fast Reactor] fuel

International Nuclear Information System (INIS)

Rhodes, E.A.; Stanford, G.S.; Regis, J.P.; Bauer, T.H.; Dickerman, C.E.

1990-01-01

Results from hodoscope data analyses are presented for TREAT transient-overpower tests M5 through M7 with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding branch and prefailure elongation of D9-clad ternary (U-Pu-Zr) IFR-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT9-clad binary (U-Zr) FFTF-driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure. 4 refs., 6 figs

Results from NNWSI [Nevada Nuclear Waste Storage Investigations] Series 2 bare fuel dissolution tests

International Nuclear Information System (INIS)

Wilson, C.N.

1990-09-01

The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Two bare spent fuel specimens plus the empty cladding hulls were tested in NNWSI J-13 well water in unsealed fused silica vessels under ambient hot cell air conditions (25 degree C) in the currently reported tests. One of the specimens was prepared from a rod irradiated in the H. B. Robinson Unit 2 reactor and the other from a rod irradiated in the Turkey Point Unit 3 reactor. Results indicate that most radionuclides of interest fall into three groups for release modeling. The first group principally includes the actinides (U, Np, Pu, Am, and Cm), all of which reached solubility-limited concentrations that were orders of magnitude below those necessary to meet the NRC 10 CFR 60.113 release limits for any realistic water flux predicted for the Yucca Mountain repository site. The second group is nuclides of soluble elements such as Cs, Tc, and I, for which release rates do not appear to be solubility-limited and may depend on the dissolution rate of fuel. In later test cycles, 137 Cs, 90 Sr, 99 Tc, and 129 I were continuously released at rates between about 5 x 10 -5 and 1 x 10 -4 of inventory per year. The third group is radionuclides that may be transported in the vapor phase, of which 14 C is of primary concern. Detailed test results are presented and discussed. 17 refs., 15 figs., 21 tabs

In-reactor fuel cladding external corrosion measurement process and results

International Nuclear Information System (INIS)

Thomazet, J.; Musante, Y.; Pigelet, J.

1999-01-01

Analysis of the zirconium alloy cladding behaviour calls for an on-site corrosion measurement device. In the 80's, a FISCHER probe was used and allowed oxide layer measurements to be taken along the outer generating lines of the peripheral fuel rods. In order to allow measurements on inner rods, a thin Eddy current probe called SABRE was developed by FRAMATOME. The SABRE is a blade equipped with two E.C coils is moved through the assembly rows. A spring allows the measurement coil to be clamped on each of the generating lines of the scanned rods. By inserting this blade on all four assembly faces, measurements can also be performed along several generating lines of the same rod. Standard rings are fitted on the device and allow on-line calibration for each measured row. Signal acquisition and processing are performed by LAGOS, a dedicated software program developed by FRAMATOME. The measurements are generally taken at the cycle outage, in the spent fuel pool. On average, data acquisition calls for one shift per assembly (eight hours): this corresponds to more than 2500 measurement points. These measurements are processed statistically by the utility program SAN REMO. All the results are collected in a database for subsequent behaviour analysis: examples of investigated parameters are the thermal/hydraulic conditions of the reactors, the irradiation history, the cladding material, the water chemistry This analysis can be made easier by comparing the behaviour measurement and prediction by means of the COROS-2 corrosion code. (author)

Analysis of environmental contamination resulting from catastrophic incidents: part 1. Building and sustaining capacity in laboratory networks.

Science.gov (United States)

Magnuson, Matthew; Ernst, Hiba; Griggs, John; Fitz-James, Schatzi; Mapp, Latisha; Mullins, Marissa; Nichols, Tonya; Shah, Sanjiv; Smith, Terry; Hedrick, Elizabeth

2014-11-01

Catastrophic incidents, such as natural disasters, terrorist attacks, and industrial accidents, can occur suddenly and have high impact. However, they often occur at such a low frequency and in unpredictable locations that planning for the management of the consequences of a catastrophe can be difficult. For those catastrophes that result in the release of contaminants, the ability to analyze environmental samples is critical and contributes to the resilience of affected communities. Analyses of environmental samples are needed to make appropriate decisions about the course of action to restore the area affected by the contamination. Environmental samples range from soil, water, and air to vegetation, building materials, and debris. In addition, processes used to decontaminate any of these matrices may also generate wastewater and other materials that require analyses to determine the best course for proper disposal. This paper summarizes activities and programs the United States Environmental Protection Agency (USEPA) has implemented to ensure capability and capacity for the analysis of contaminated environmental samples following catastrophic incidents. USEPA's focus has been on building capability for a wide variety of contaminant classes and on ensuring national laboratory capacity for potential surges in the numbers of samples that could quickly exhaust the resources of local communities. USEPA's efforts have been designed to ensure a strong and resilient laboratory infrastructure in the United States to support communities as they respond to contamination incidents of any magnitude. The efforts include not only addressing technical issues related to the best-available methods for chemical, biological, and radiological contaminants, but also include addressing the challenges of coordination and administration of an efficient and effective response. Laboratory networks designed for responding to large scale contamination incidents can be sustained by applying

Energies and media nr 27. Conditions for the nuclear sector. Incidents this summer in France, Belgium and Germany

International Nuclear Information System (INIS)

2008-10-01

After some comments on recent events in the nuclear sector (agreement between India and the USA, new programs in different countries, activity of French companies abroad), this issue comments the worrying accumulation of incidents which occurred in several French nuclear installations in France (notably on the Tricastin site). It comments the actions of the French nuclear safety authority (ASN), the incident classification, delays in public information, problems faced on fuel assemblies, the issue of subcontracting. Incidents which occurred in Belgium (leakage of radioactive iodine) and Germany (water in an ancient salt and potash mine) are also commented

Fuel performance under normal PWR conditions: A review of relevant experimental results and models

Science.gov (United States)

Charles, M.; Lemaignan, C.

1992-06-01

Experiments conducted at Grenoble (CEA/DRN) over the past 20 years in the field of nuclear fuel behaviour are reviewed. Of particular concern is the need to achieve a comprehensive understanding of and subsequently overcome the limitations associated with high burnup and load-following conditions (pellet-cladding interaction (PCI), fission gas release (FGR), water-side corrosion). A general view is given of the organization of research work as well as some experimental details (irradiation, postirradiation examination — PIE). Based on various experimental programmes (Cyrano, Medicis, Anemone, Furet, Tango, Contact, Cansar, Hatac, Flog, Decor), the main contributions of the thermomechanical behaviour of a PWR fuel rod are described: thermal conductivity, in-pile densification, swelling, fission gas release in steady state and moderate transient conditions, gap thermal conductance, formation of primary and secondary ridges under PCI conditions. Specific programmes (Gdgrif, Thermox, Grimox) are devoted to the behaviour of particular fuels (gadolinia-bearing fuel, MOX fuel). Moreover, microstructure-based studies have been undertaken on fission gas release (fine analysis of the bubble population inside irradiated fuel samples), and on cladding behaviour (PCI related studies on stress-corrosion cracking (SCO, irradiation effects on zircaloy microstructure).

Results from Nevada Nuclear Waste Storage Investigations (NNWSI) Series 3 spent fuel dissolution tests

International Nuclear Information System (INIS)

Wilson, C.N.

1990-06-01

The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Yucca Mountain Project (YMP), formerly the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Specimens prepared from pressurized water reactor fuel rod segments were tested in sealed stainless steel vessels in Nevada Test Site J-13 well water at 85 degree C and 25 degree C. The test matrix included three specimens of bare-fuel particles plus cladding hulls, two fuel rod segments with artificially defected cladding and water-tight end fittings, and an undefected fuel rod section with watertight end fittings. Periodic solution samples were taken during test cycles with the sample volumes replenished with fresh J-13 water. Test cycles were periodically terminated and the specimens restarted in fresh J-13 water. The specimens were run for three cycles for a total test duration of 15 months. 22 refs., 32 figs., 26 tabs

Sustainable thorium nuclear fuel cycles: A comparison of intermediate and fast neutron spectrum systems

International Nuclear Information System (INIS)

Brown, N.R.; Powers, J.J.; Feng, B.; Heidet, F.; Stauff, N.E.; Zhang, G.; Todosow, M.; Worrall, A.; Gehin, J.C.; Kim, T.K.; Taiwo, T.A.

2015-01-01

Highlights: • Comparison of intermediate and fast spectrum thorium-fueled reactors. • Variety of reactor technology options enables self-sustaining thorium fuel cycles. • Fuel cycle analyses indicate similar performance for fast and intermediate systems. • Reproduction factor plays a significant role in breeding and burn-up performance. - Abstract: This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems

Sustainable thorium nuclear fuel cycles: A comparison of intermediate and fast neutron spectrum systems

Energy Technology Data Exchange (ETDEWEB)

Brown, N.R., E-mail: nbrown@bnl.gov [Brookhaven National Laboratory, Upton, NY (United States); Powers, J.J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Feng, B.; Heidet, F.; Stauff, N.E.; Zhang, G. [Argonne National Laboratory, Argonne, IL (United States); Todosow, M. [Brookhaven National Laboratory, Upton, NY (United States); Worrall, A.; Gehin, J.C. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Kim, T.K.; Taiwo, T.A. [Argonne National Laboratory, Argonne, IL (United States)

2015-08-15

Highlights: • Comparison of intermediate and fast spectrum thorium-fueled reactors. • Variety of reactor technology options enables self-sustaining thorium fuel cycles. • Fuel cycle analyses indicate similar performance for fast and intermediate systems. • Reproduction factor plays a significant role in breeding and burn-up performance. - Abstract: This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10{sup 5} eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.

Impact of Fuel Failure on Criticality Safety of Used Nuclear Fuel

International Nuclear Information System (INIS)

Marshall, William J.; Wagner, John C.

2012-01-01

Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., 45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k eff . Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

Climate consequences of low-carbon fuels: The United States Renewable Fuel Standard

International Nuclear Information System (INIS)

Hill, Jason; Tajibaeva, Liaila; Polasky, Stephen

2016-01-01

A common strategy for reducing greenhouse gas (GHG) emissions from energy use is to increase the supply of low-carbon alternatives. However, increasing supply tends to lower energy prices, which encourages additional fuel consumption. This “fuel market rebound effect” can undermine climate change mitigation strategies, even to the point where efforts to reduce GHG emissions by increasing the supply of low-carbon fuels may actually result in increased GHG emissions. Here, we explore how policies that encourage the production of low-carbon fuels may result in increased GHG emissions because the resulting increase in energy use overwhelms the benefits of reduced carbon intensity. We describe how climate change mitigation strategies should follow a simple rule: a low-carbon fuel with a carbon intensity of X% that of a fossil fuel must displace at least X% of that fossil fuel to reduce overall GHG emissions. We apply this rule to the United States Renewable Fuel Standard (RFS2). We show that absent consideration of the fuel market rebound effect, RFS2 appears to reduce GHG emissions, but once the fuel market rebound effect is factored in, RFS2 actually increases GHG emissions when all fuel GHG intensity targets are met. - Highlights: • Low-carbon fuels partially displace petroleum via fuel market rebound effect. • Synthesis of recent analyses shows incomplete petroleum displacement by biofuels. • Fuel market rebound effect can reduce or reverse climate benefit of low-carbon fuels. • Fossil fuel displacement must exceed relative carbon footprint of a low-carbon fuel. • The Renewable Fuel Standard increases greenhouse gas emissions when mandate is met.

Nuclear fuels technologies fiscal year 1996 research and research development test results

International Nuclear Information System (INIS)

Beard, C.A.; Blair, H.T.; Buksa, J.J.; Butt, D.P.

1996-01-01

During fiscal year 1996, the Department of Energy's Office of Fissile Materials Disposition (OFMD) funded Los Alamos National Laboratory (LANL) to investigate issues associated with the fabrication of plutonium from dismantled weapons into mixed-oxide (MOX) nuclear fuel for disposition in nuclear power reactors. These issues can be divided into two main categories: issues associated with the fact that the plutonium from dismantled weapons contains gallium, and issues associated with the unique characteristics of the PuO 2 produced by the dry conversion process that OFMD is proposing to convert the weapons material. Initial descriptions of the experimental work performed in fiscal year 1996 to address these issues can be found in Nuclear Fuels Technologies Fiscal Year 1996 Research and Development Test Matrices'. However, in some instances the change in programmatic emphasis towards the Parallex program either altered the manner in which some of these experiments were performed (i.e., the work was done as part of the Parallex fabrication development and not as individual separate-effects tests as originally envisioned) or delayed the experiments into Fiscal Year 1997. This report reviews the experiments that were conducted and presents the results. 7 figs., 14 tabs

Multi-objective regulations on transportation fuels: Comparing renewable fuel mandates and emission standards

International Nuclear Information System (INIS)

Rajagopal, D.; Plevin, R.; Hochman, G.; Zilberman, D.

2015-01-01

We compare two types of fuel market regulations — a renewable fuel mandate and a fuel emission standard — that could be employed to simultaneously achieve multiple outcomes such as reduction in fuel prices, fuel imports and greenhouse gas (GHG) emissions. We compare these two types of regulations in a global context taking into account heterogeneity in carbon content of both fossil fuels and renewable fuels. We find that although neither the ethanol mandate nor the emission standard is certain to reduce emissions relative to a business-as-usual baseline, at any given level of biofuel consumption in the policy region, a mandate, relative to an emission standard, results in higher GHG emissions, smaller expenditure on fuel imports, lower price of ethanol-blended gasoline and higher domestic fuel market surplus. This result holds over a wide range of values of model parameters. We also discuss the implications of this result to a regulation such as the US Renewable Fuel Standard given recent developments within the US such as increase in shale and tight oil production and large increase in average vehicle fuel economy of the automotive fleet. - Highlights: • Biofuel mandates and fuel GHG emission standards are analyzed from a multiple criteria perspective • An emission-standard always results in lower global emissions while requiring less biofuel relative to a biofuel mandate • An emission-standard results in higher fuel price in the home region relative to a biofuel mandate • Emission standards lead to more shuffling of both fossil fuels and biofuels between home and abroad • The relative impact of the policies on fuel imports depends on the relative cost-effectiveness of domestic & imported biofuel • Recent developments oil production and fuel economy increase the net benefits of an LCFS approach relative to RFS

A fuel cycle cost study with HEU and LEU fuels

International Nuclear Information System (INIS)

Matos, J.E.; Freese, K.E.

1985-01-01

Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

A fuel cycle cost study with HEU and LEU fuels

Energy Technology Data Exchange (ETDEWEB)

Matos, J E; Freese, K E [Argonne National Laboratory, Argonne, IL (United States)

1985-07-01

Fuel cycle costs are compared for a range of {sup 235}U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

Proceedings of fuel strategies for conventional and unconventional fuels

International Nuclear Information System (INIS)

Mahr, D.; Nechvatal, T.T.

1991-01-01

Fuel selection is a major decision for a power engineer. It is the single largest item in the power plant operating budget and has a major effect on power plant economics. Fuel determines plant design requirements and the types of systems that are provided. As a result, it affects capital budgets and financing requirements. In the last few decades, we have seen different fuels of choice at any one time. Coal has always been a staple for power generation. During the 1950s and 1960s, oil became an attractive alternative. Nuclear fuel became a popular choice due to its very low energy cost. After Three-Mile-Island, however, capital budgets went through the roof, resulting in severe financial constraints. Natural gas, which was rationed in some regions a few years ago, is now a popular choice. Some sources predict that its cost will increase faster than other fuels. To mitigate the relative variations in energy cost for different fuels, a balanced energy plan is required. A balanced power generation plan with fuel options provides the flexibility to react to unpredictable changes. The papers in this book are a continuation of the Fuel Strategies theme. Three technical topics are covered: Converting to Orimulsion, A Replacement Fuel for Heavy Oil; Innovations in Handling Conventional and Unconventional Fuels for Power Plants; and Pacific Rim Experience With Coal

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.

Cesium relocation in mixed-oxide fuel pins resulting from increased temperature reirradiation

International Nuclear Information System (INIS)

Lawrence, L.A.; Woodley, R.E.; Weber, E.T.

1976-06-01

Mixed-oxide fuel pins from EBR-II test subassemblies PNL-3 and PNL-4 were reirradiated in the GETR to study effects of increased fuel and cladding temperatures on chemical and thermomechanical behavior. Radial and axial distributions of cesium were obtained using postirradiation nondestructive precision gamma-scanning techniques. Data presented relate to the dependence of cesium distribution and transport processes on temperature gradients which were altered after substantial steady-state operation

Fuel Property, Emission Test, and Operability Results from a Fleet of Class 6 Vehicles Operating on Gas-to-Liquid Fuel and Catalyzed Diesel Particle Filters

Energy Technology Data Exchange (ETDEWEB)

Alleman, T. L.; Eudy, L.; Miyasato, M.; Oshinuga, A.; Allison, S.; Corcoran, T.; Chatterjee, S.; Jacobs, T.; Cherrillo, R. A.; Clark, R.; Virrels, I.; Nine, R.; Wayne, S.; Lansing, R.

2005-11-01

A fleet of six 2001 International Class 6 trucks operating in southern California was selected for an operability and emissions study using gas-to-liquid (GTL) fuel and catalyzed diesel particle filters (CDPF). Three vehicles were fueled with CARB specification diesel fuel and no emission control devices (current technology), and three vehicles were fueled with GTL fuel and retrofit with Johnson Matthey's CCRT diesel particulate filter. No engine modifications were made.

The cancer mortality and incidence experience of workers at British Nuclear Fuels plc, 1946–2005

International Nuclear Information System (INIS)

Gillies, Michael; Haylock, Richard

2014-01-01

The aim of this study was to estimate cancer mortality and incidence risk associated with external radiation exposure in the BNFL cohort of nuclear workers and to determine if these risks are modified by potential for internal exposure. The cohort comprised 64 956 individuals who were employed at the four study sites between 1946 and 2002, followed up to 2005. External radiation exposures as measured by personal dosimeters (generally ‘film badges’) were available for 42 431 individuals classified as ‘radiation workers’. Poisson regression models were used to investigate cancer mortality and incidence in relation to cumulative external radiation exposure using relative risk models. The cohort showed the expected ‘healthy worker’ effect. This analysis found an increased risk of all cancers associated with external occupational radiation exposure (ERR/Gy = 0.34 90% CI: 0.07; 0.64), with significant excess risks observed for all solid cancers (ERR/Gy = 0.29 90% CI: 0.02; 0.59) and leukaemia excluding CLL (ERR/Gy = 2.60 90% CI: 0.28; 7.01). The overall cancer risk estimates are consistent with values used by national and international bodies in setting radiation protection standards. The slopes of the dose response relationships for all cancer mortality and incidence were found to be significantly less steep for workers exposed to both external radiation and potentially to internal radiation (ERR/Gy = 0.09 90% CI: −0.17; 0.39) when compared to those workers only exposed to external radiation (ERR/Gy = 1.14 90% CI: 0.49; 1.89). Analyses of individual cancer types indicate that this overall result is mainly driven by that for digestive cancers and in particular cancers of the oesophagus. Categorical analyses also revealed that the difference in the dose response relationship between the two groups is only apparent for those exposed to cumulative external doses in excess of 200 mGy. Such differences have also been observed for non-cancer mortality

Childhood sarcoidosis in Denmark 1979-1994: incidence, clinical features and laboratory results at presentation in 48 children

DEFF Research Database (Denmark)

Hoffmann, A L; Milman, N; Byg, K E

2004-01-01

.7-15). In 1979-1994 the incidence was 0.29 per 100000 person-years incidence was 0.06 in children weight loss, abdominal discomfort, respiratory symptoms, lymphadenopathy......AIM: To describe the incidence, clinical presentation and paraclinical findings in childhood sarcoidosis in Denmark, 1979-1994. METHODS: Patients (n = 5536) with a diagnosis of sarcoidosis were drawn from the nationwide Patient Registry; 81 patients were ... examination (glucose, albumin, haemoglobin) was normal in 96% of the patients; the patient with nephrocalcinosis had albuminuria and haematuria. CONCLUSION: The incidence of sarcoidosis in Danish children is low and increases with age. Sarcoidosis in young children may present clinical features...

Apparatus for loading fuel pellets in fuel rods

International Nuclear Information System (INIS)

Tedesco, R.J.

1976-01-01

An apparatus is disclosed for loading fuel pellets into fuel rods for a nuclear reactor including a base supporting a table having grooves therein for holding a multiplicity of pellets. Multiple fuel rods are placed in alignment with grooves in the pellet table and a guide member channels pellets from the table into the corresponding fuel rods. To effect movement of pellets inside the fuel rods without jamming, a number of electromechanical devices mounted on the base have arms connected to the lower surface of the fuel rod table which cyclically imparts a reciprocating arc motion to the table for moving the fuel pellets longitudinally of and inside the fuel rods. These electromechanical devices include a solenoid having a plunger therein connected to a leaf type spring, the arrangement being such that upon energization of the solenoid coil, the leaf spring moves the fuel rod table rearwardly and downwardly, and upon deenergization of the coil, the spring imparts an upward-forward movement to the table which results in physical displacement of fuel pellets in the fuel rods clamped to the table surface. 8 claims, 6 drawing figures

A five-year history of hazardous materials incidents in Chester County, PA

International Nuclear Information System (INIS)

Shorten, C.V.; McNamara, J.

1993-01-01

The Emergency Planning and Community Right-to-Know Act (EPCRA) of 1986 established Local Emergency Planning Committees (LEPCs) to oversee emergency response planning at the community level. In Pennsylvania, each county was assigned its own LEPC, and Chester County held its first LEPC meeting on October 15, 1987. From the data of that meeting through September 1992, 300 hazardous materials incidents have been reported. The majority of these incidents were met with fire department response, but several warranted response by hazardous materials teams. This report presents an analysis of the database of reported hazardous materials incidents in Chester County, including chemical identification, amount released, type of response, location, and trends. Over 235 of the reported spills were either gasoline, diesel fuel, home heating oil, or kerosene, often in five to 50 gallon amounts from transportation accidents. A number of extremely hazardous substance (EHS) incidents were reported, however, including sulfuric acid, chlorine, ammonia, phosphorus, formaldehyde, bromine, methyl mercaptan, and hydrofluoric acid. The most commonly released EHS's were ammonia and chlorine. The number of hazardous materials incidents reported in Chester County increased from only 14 in 1988 to 95 in 1991, with 67 in 1992 through September. This dramatic increase is attributable to both increased reporting and an increased number of incidents. This database clearly indicates both the success of EPCRA reporting system and the magnitude of hazardous materials incidents in this part of Pennsylvania

Fuel Cycle Concept with Advanced METMET and Composite Fuel in LWRs

International Nuclear Information System (INIS)

Savchenko, A.; Skupov, M.; Vatulin, A.; Glushenkov, A.; Kulakov, G.; Lipkina, K.

2014-01-01

The basic factor that limits the serviceability of fuel elements developing in the framework of RERTR Program (transition from HEU to LEU fuel of research reactors) is interaction between U10Mo fuel and aluminium matrix . Interaction results in extra swelling of fuels, disappearance of a heat conducting matrix, a temperature rise in the fuel centre, penetration porosity, etc. Several methods exist to prevent fuel-matrix interaction. In terms of simplifying fuel element fabrication technology and reducing interaction, doping of fuel is the most optimal version

Results of oscillation experiments on the Cesar and Marius piles - Uranium-Plutonium fuels

International Nuclear Information System (INIS)

Laponche, Bernard; Brunet, Max; Menessier, Denise; Morier, Francis; Basiuk, Marie-Jose; Tonolli, Jacky; Vanuxeem, Jacqueline

1969-05-01

The authors present, comment and discuss results obtained during three measurement campaigns performed on the Cesar and Marius atomic piles between 1965 and 1967 for the determination of some physical quantities (like the Plutonium η or its cross sections) from measurements of two signals which characterize the pile response to a central disturbance caused by the fuel to be studied. They more particularly address mass-corrected signals, the Uranium-235 and Boron calibration of the reactor, the local signal of the equivalent sample to a measured UPu sample. They indicate the different steps of interpretation of these results, present and discuss the measured results

Impact of fuel fabrication and fuel management technologies on uranium management

International Nuclear Information System (INIS)

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

1994-01-01

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

Methanol fuel processor and PEM fuel cell modeling for mobile application

Energy Technology Data Exchange (ETDEWEB)

Chrenko, Daniela [ISAT, University of Burgundy, Rue Mlle Bourgoise, 58000 Nevers (France); Gao, Fei; Blunier, Benjamin; Bouquain, David; Miraoui, Abdellatif [Transport and Systems Laboratory (SeT) - EA 3317/UTBM, Fuel cell Laboratory (FCLAB), University of Technology of Belfort-Montbeliard, Rue Thierry Mieg 90010, Belfort Cedex (France)

2010-07-15

The use of hydrocarbon fed fuel cell systems including a fuel processor can be an entry market for this emerging technology avoiding the problem of hydrogen infrastructure. This article presents a 1 kW low temperature PEM fuel cell system with fuel processor, the system is fueled by a mixture of methanol and water that is converted into hydrogen rich gas using a steam reformer. A complete system model including a fluidic fuel processor model containing evaporation, steam reformer, hydrogen filter, combustion, as well as a multi-domain fuel cell model is introduced. Experiments are performed with an IDATECH FCS1200 trademark fuel cell system. The results of modeling and experimentation show good results, namely with regard to fuel cell current and voltage as well as hydrogen production and pressure. The system is auto sufficient and shows an efficiency of 25.12%. The presented work is a step towards a complete system model, needed to develop a well adapted system control assuring optimized system efficiency. (author)

Results of post-irradiation examination of WWER fuel assembly structural components made of E110 and E635 alloys

International Nuclear Information System (INIS)

Smirnov, A.; Markov, D.; Smirnov, V.; Polenok, V.; Ivashchenko, A.; Strozhuk, A.

2006-01-01

The paper presents the main examination results on the condition of fuel rods claddings, guide tubes and spacer grids of the WWER FA made of E110 and E635 alloys operated under standard operating conditions. The paper is based on the data obtained during the examination of 28 WWER-1000 FA and 12 WWER-400 FA. E110 alloy is shown to be suitable material for the WWER fuel rod claddings under the normal operating conditions. E635 alloy is attractive to manufacturing of the skeleton components. The currently used combination (E110 as a material of fuel rods claddings and E635 - as a material of the skeleton components) is the optimal solution for the WWER fuel assembly because the advantages of the both alloys are used. (authors)

The prediction of the-circumferential fuel-temperature distribution under ballonian condition. Vol. 3

Energy Technology Data Exchange (ETDEWEB)

Abdallah, A M; El-Sherbiny, E M [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

Swelling and thermal distortion of nuclear fuel elements due to depressurization of reactor coolant may cause contracts in points or finite regions between adjacent fuel elements in square and triangle lattices. This is very probable in Advanced Pressurized Water Reactors where the clearance between fuel elements is about 1 mm. This results in partial blocking of the coolant flow and formation of hot spots in the contact regions. In these regions, absence of coolant results in nonuniform clad circumferential temperature distribution. This causes excessive thermal stresses which may produce local melting or clad failure. An accurate prediction of the clad circumferential temperature distribution during these severe incidents is very important. This problem was studied numerically during transient and steady state conditions. Recently, a semi analytical solution for the underlying problem was derived assuming the heat transfer coefficient to vary linearly with the circumferential distance measured from the cusp point, and the heat flux at the fuel-clad interface to be a constant quantity. In the present work, an approximate analytic solution is obtained. The accuracy is tested by solving the problem numerically. Also the problem is reanalyzed by considering the heat flux at the fuel-clad interface to be a power function of the angular distance along the clad surface. Moreover, the heat transfer coefficient is assumed to be a function of both the circumferential coordinate and temperature of the clad. Discussion of the analytical solution and the assumptions are rationalized in the text. 4 figs.

Fuel pattern recognition device

International Nuclear Information System (INIS)

Sato, Tomomi.

1995-01-01

The device of the present invention monitors normal fuel exchange upon fuel exchanging operation carried out in a reactor of a nuclear power plant. Namely, a fuel exchanger is movably disposed to the upper portion of the reactor and exchanges fuels. An exclusive computer receives operation signals of the fuel exchanger during operation as inputs, and outputs reactor core fuel pattern information signals to a fuel arrangement diagnosis device. An underwater television camera outputs image signals of a fuel pattern in the reactor core to an image processing device. If there is any change in the image signals for the fuel pattern as a result of the fuel exchange operation of the fuel exchanger, the image processing device outputs the change as image signals to the fuel pattern diagnosis device. The fuel pattern diagnosis device compares the pattern information signals from the exclusive computer with the image signals from the image processing device, to diagnose the result of the fuel exchange operation performed by the fuel exchanger and inform the diagnosis by means of an image display. (I.S.)

Methanol Fuel Cell

Science.gov (United States)

Voecks, G. E.

1985-01-01

In proposed fuel-cell system, methanol converted to hydrogen in two places. External fuel processor converts only part of methanol. Remaining methanol converted in fuel cell itself, in reaction at anode. As result, size of fuel processor reduced, system efficiency increased, and cost lowered.

Fuel consumption and emission on fuel mixer low-grade bioethanol fuelled motorcycle

Directory of Open Access Journals (Sweden)

Abikusna Setia

2017-01-01

Full Text Available Bioethanol is currently used as an alternative fuel for gasoline substitute (fossil fuel because it can reduce the dependence on fossil fuel and also emissions produced by fossil fuel which are CO2, HO, NOx. Bioethanol is usually used as a fuel mixed with gasoline with certain comparison. In Indonesia, the usage is still rare. Bioethanol that is commonly used is bioethanol anhydrous 99.5%. In the previous studies, bioethanol was distilled from low to high grade to produce ethanol anhydrous. But the result is only able to reach 95% or ethanol hydrous. This study is objected to design a simple mechanism in the mixing of bioethanol hydrous with the gasoline using a fuel mixer mechanism. By this mechanism, the fuel consumption and the resulting emissions from combustion engine can be analyzed. The fuel blend composition is prepared as E5, E10, and E15/E20, the result of fuel consumption and emission will be compared with pure gasoline. The using of bioethanol hydrous as a fuel mixture was tended to produce more stable bioethanol fuel consumption. However, the utilization of the mixture was found able to reduce the exhaust emissions (CO, HC, and NOx.

Cancer incidence among waiters

DEFF Research Database (Denmark)

Reijula, Jere; Kjaerheim, Kristina; Lynge, Elsebeth

2015-01-01

AIMS: To study cancer risk patterns among waiters in the Nordic countries. METHODS: We identified a cohort of 16,134 male and 81,838 female waiters from Denmark, Finland, Iceland, Norway and Sweden. During the follow-up period from 1961 to 2005, we found that 19,388 incident cancer cases were...... diagnosed. Standardised incidence ratio (SIR) was defined as the observed number of cancer cases divided by the expected number, based on national age, time period and gender-specific cancer incidence rates in the general population. RESULTS: The SIR of all cancers in waiters, in the five countries combined...... INCIDENCE IN SOME CANCER SITES CAN LIKELY BE EXPLAINED BY HIGHER ALCOHOL CONSUMPTION, THE PREVALENCE OF SMOKING AND OCCUPATIONAL EXPOSURE TO TOBACCO SMOKE HOPEFULLY, THE INCIDENCE OF CANCER AMONG WAITERS WILL DECREASE IN THE FUTURE, DUE TO THE BANNING OF TOBACCO SMOKING IN RESTAURANTS AND BARS IN THE NORDIC...

Public information circular for shipments of irradiated reactor fuel. Revision 4

International Nuclear Information System (INIS)

1984-06-01

This publication is the fifth in a series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. This publication contains basically three kinds of information: (1) routes recently approved (18 months) by the Commission for the shipment of irradiated reactor fuel; (2) information regarding any safeguards-significant incidents that may be (to date none have) reported during shipments along such routes; and (3) cumulative amounts of material shipped

Nuclear Fuel elements

International Nuclear Information System (INIS)

Hirakawa, Hiromasa.

1979-01-01

Purpose: To reduce the stress gradient resulted in the fuel can in fuel rods adapted to control the axial power distribution by the combination of fuel pellets having different linear power densities. Constitution: In a fuel rod comprising a first fuel pellet of a relatively low linear power density and a second fuel pellet of a relatively high linear power density, the second fuel pellet is cut at its both end faces by an amount corresponding to the heat expansion of the pellet due to the difference in the linear power density to the adjacent first fuel pellet. Thus, the second fuel pellet takes a smaller space than the first fuel pellet in the fuel can. This can reduce the stress produced in the portion of the fuel can corresponding to the boundary between the adjacent fuel pellets. (Kawakami, Y.)

Diabetes and risk of cancer incidence: results from a population-based cohort study in northern Italy.

Science.gov (United States)

Ballotari, Paola; Vicentini, Massimo; Manicardi, Valeria; Gallo, Marco; Chiatamone Ranieri, Sofia; Greci, Marina; Giorgi Rossi, Paolo

2017-10-25

Aim of this study was to compare cancer incidence in populations with and without diabetes by cancer site. Furthermore, we aimed at comparing excess risk of cancer according to diabetes type, diabetes duration and treatment, the latter as regards Type 2 diabetes. By use of the Reggio Emilia diabetes registry we classified the resident population aged 20-84 at December 31 st 2009 into two groups: with and without diabetes. By linking with the cancer registry we calculated the 2010-2013 cancer incidence in both groups. The incidence rate ratios (IRR) by cancer site, type of diabetes, diabetes duration, and as concerns Type 2 diabetes, by treatment regimen were computed using Poisson regression model and non-diabetic group as reference. The cohort included 383,799 subjects without diabetes and 23,358 with diabetes. During follow-up, we identified 1464 cancer cases in subjects with diabetes and 9858 in the remaining population. Overall cancer incidence was higher in subjects with diabetes than in those without diabetes (IRR = 1.22, 95%CI 1.15-1.29), with similar results focusing on subjects with at least 2-year diabetes duration. Cancer sites driving overall increased risk were liver, pancreas, Colon rectum, and bladder in both sexes, corpus uteri for females. There was also suggestion of an increased risk for kidney cancer in females and a decreased risk for prostate cancer. Excess risk was found in patients with Type 2 diabetes, more marked among insulin users, especially with combined therapy. We observed an increasing risk for diabetes duration up to 10 years from diagnosis (IRR = 1.44, 95%CI 1.29-1.61) and a subsequent decrease to moderate-higher risk (IRR = 1.15, 95%CI 1.04-1.30). Our study indicates that the strength of association depends on specific cancer site. Insulin, monotherapy or combined therapy, per se or as an indication of poor blood glucose control, in addition to diabetes duration, may play a role in the association of diabetes and

Reforming options for hydrogen production from fossil fuels for PEM fuel cells

Energy Technology Data Exchange (ETDEWEB)

Ersoz, Atilla; Olgun, Hayati [TUBITAK Marmara Research Center, Institute of Energy, Gebze, 41470 Kocaeli (Turkey); Ozdogan, Sibel [Marmara University Faculty of Engineering, Goztepe, 81040 Istanbul (Turkey)

2006-03-09

PEM fuel cell systems are considered as a sustainable option for the future transport sector in the future. There is great interest in converting current hydrocarbon based transportation fuels into hydrogen rich gases acceptable by PEM fuel cells on-board of vehicles. In this paper, we compare the results of our simulation studies for 100kW PEM fuel cell systems utilizing three different major reforming technologies, namely steam reforming (SREF), partial oxidation (POX) and autothermal reforming (ATR). Natural gas, gasoline and diesel are the selected hydrocarbon fuels. It is desired to investigate the effect of the selected fuel reforming options on the overall fuel cell system efficiency, which depends on the fuel processing, PEM fuel cell and auxiliary system efficiencies. The Aspen-HYSYS 3.1 code has been used for simulation purposes. Process parameters of fuel preparation steps have been determined considering the limitations set by the catalysts and hydrocarbons involved. Results indicate that fuel properties, fuel processing system and its operation parameters, and PEM fuel cell characteristics all affect the overall system efficiencies. Steam reforming appears as the most efficient fuel preparation option for all investigated fuels. Natural gas with steam reforming shows the highest fuel cell system efficiency. Good heat integration within the fuel cell system is absolutely necessary to achieve acceptable overall system efficiencies. (author)

Reforming options for hydrogen production from fossil fuels for PEM fuel cells

Science.gov (United States)

Ersoz, Atilla; Olgun, Hayati; Ozdogan, Sibel

PEM fuel cell systems are considered as a sustainable option for the future transport sector in the future. There is great interest in converting current hydrocarbon based transportation fuels into hydrogen rich gases acceptable by PEM fuel cells on-board of vehicles. In this paper, we compare the results of our simulation studies for 100 kW PEM fuel cell systems utilizing three different major reforming technologies, namely steam reforming (SREF), partial oxidation (POX) and autothermal reforming (ATR). Natural gas, gasoline and diesel are the selected hydrocarbon fuels. It is desired to investigate the effect of the selected fuel reforming options on the overall fuel cell system efficiency, which depends on the fuel processing, PEM fuel cell and auxiliary system efficiencies. The Aspen-HYSYS 3.1 code has been used for simulation purposes. Process parameters of fuel preparation steps have been determined considering the limitations set by the catalysts and hydrocarbons involved. Results indicate that fuel properties, fuel processing system and its operation parameters, and PEM fuel cell characteristics all affect the overall system efficiencies. Steam reforming appears as the most efficient fuel preparation option for all investigated fuels. Natural gas with steam reforming shows the highest fuel cell system efficiency. Good heat integration within the fuel cell system is absolutely necessary to achieve acceptable overall system efficiencies.

Blending Biodiesel in Fishing Boat Fuels for Improved Fuel Characteristics

International Nuclear Information System (INIS)

Lin, Cherng-Yuan

2014-01-01

Biodiesel is a renewable, clean, alternative energy source with advantages, such as excellent lubricity, superior biodegradability, and high combustion efficiency. Biodiesel is considered for mixing with fishing boat fuels to adjust their fuel characteristics so that toxic pollutants and greenhouse-effect gas emissions from such shipping might be reduced. The effects of blending fishing boat fuels A and B with various weight proportions of biodiesel are experimentally investigated in this study. The results show that biodiesel blending can significantly improve the inferior fuel properties of both fishing boat fuels and particularly fuel B. The flash points of both of these fuels increases significantly with the addition of biodiesel and thus enhances the safety of transporting and storing these blended fuels. The flash point of fishing boat fuel B even increases by 16% if 25 wt.% biodiesel is blended. The blending of biodiesel with no sulfur content is found to be one of the most effective ways to reduce the high sulfur content of fishing boat fuel, resulting in a reduction in the emission of sulfur oxides. The addition of only 25 wt.% biodiesel decreased the sulfur content of the fishing boat fuel by 37%. The high kinematic viscosity of fishing boat fuel B was also observed to be reduced by 63% with the blending of just 25 wt.% biodiesel. However, biodiesel blending caused a slight decrease in heating value around 1–4.5%.

Blending Biodiesel in Fishing Boat Fuels for Improved Fuel Characteristics

Energy Technology Data Exchange (ETDEWEB)

Lin, Cherng-Yuan, E-mail: lin7108@ntou.edu.tw [Department of Marine Engineering, National Taiwan Ocean University, Keelung, Taiwan (China)

2014-02-24

Biodiesel is a renewable, clean, alternative energy source with advantages, such as excellent lubricity, superior biodegradability, and high combustion efficiency. Biodiesel is considered for mixing with fishing boat fuels to adjust their fuel characteristics so that toxic pollutants and greenhouse-effect gas emissions from such shipping might be reduced. The effects of blending fishing boat fuels A and B with various weight proportions of biodiesel are experimentally investigated in this study. The results show that biodiesel blending can significantly improve the inferior fuel properties of both fishing boat fuels and particularly fuel B. The flash points of both of these fuels increases significantly with the addition of biodiesel and thus enhances the safety of transporting and storing these blended fuels. The flash point of fishing boat fuel B even increases by 16% if 25 wt.% biodiesel is blended. The blending of biodiesel with no sulfur content is found to be one of the most effective ways to reduce the high sulfur content of fishing boat fuel, resulting in a reduction in the emission of sulfur oxides. The addition of only 25 wt.% biodiesel decreased the sulfur content of the fishing boat fuel by 37%. The high kinematic viscosity of fishing boat fuel B was also observed to be reduced by 63% with the blending of just 25 wt.% biodiesel. However, biodiesel blending caused a slight decrease in heating value around 1–4.5%.

Fuel assembly

International Nuclear Information System (INIS)

Nomata, Terumitsu.

1993-01-01

Among fuel pellets to be loaded to fuel cans of a fuel assembly, fuel pellets having a small thermal power are charged in a region from the end of each of spacers up to about 50mm on the upstream of coolants that flow vertically at the periphery of fuel rods. Coolants at the periphery of fuel rods are heated by the heat generation, to result in voids. However, since cooling effect on the upstream of the spacers is low due to influences of the spacers. Further, since the fuel pellets disposed in the upstream region have small thermal power, a void coefficient is not increased. Even if a thermal power exceeding cooling performance should be generated, there is no worry of causing burnout in the upstream region. Even if burnout should be caused, safety margin and reliability relative to burnout are improved, to increase an allowable thermal power, thereby enabling to improve integrity and reliability of fuel rods and fuel assemblies. (N.H.)

OECD-IAEA Paks Fuel Project. Final Report

International Nuclear Information System (INIS)

2010-05-01

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

Study on using acetylene in dual fuel mode with exhaust gas recirculation

International Nuclear Information System (INIS)

Lakshmanan, T.; Nagarajan, G.

2011-01-01

Interest in employing gaseous fuels to internal combustion (IC) engines whether for stationary or mobile automotive applications has gained importance because of the economical, sustainable and environmental technical features associated with their usage. However, the incidence of preignition and knock remains a significant barrier in achieving their optimum performance potential. With the advent of latest technologies, the above barriers are eliminated. One such technique is timed manifold injection (TMI) of the gaseous fuel, which is controlled electronically to precisely monitor the induction of fuel to overcome the preignition problem in the intake. In the present investigation, acetylene was injected in the intake manifold in a single cylinder diesel engine, with a gas flow rate of 240 g/h, start of injection time is 10 o aTDC and 90 o CA (9.9 ms) duration, operated in dual fuel mode. In order to decrease the NOx emissions from acetylene-diesel engine, cooled EGR was employed. The cylinder pressure, brake thermal efficiency and emissions such as NOx, smoke, CO, HC, CO 2 and exhaust gas temperature were studied. Dual fuel operation with acetylene induction coupled with cooled EGR results in lowered NOx emissions and improved part load performance. -- Highlights: → Acetylene was tried in SI engines, but due to backfire further research was hindered as an alternative fuel. → But it is not tried in CI engine. Timed manifold injection was tried in diesel engine in the present work to combat backfire. → Author was successful in running the diesel engine in dual fuel mode. → 21% maximum diesel replacement was achieved. Author is confident that acetylene will be commercialised as a fuel for diesel engine in future.

Focused ultrasonic wave testing, in immersion of spent fuel cans

International Nuclear Information System (INIS)

Poinboeuf, P.; Furlan, J.

1984-10-01

To detect weak and very weak damage of the fuel can, ultrasonic testing has been used. For that, a simple mechanical device, allowing to maintain an optimal ultrasonic focussing on irradiated cans, is presented. Its aim is to correct the variation of the incidence angle due to the possible ovalization of pins. After a description of the device, the results obtained with tests carried out on non-irradiated cans, including artificial ovalized regions, standard defects, are presented. After the description of the adaptation of this mechanism on a test bench which allows an helicoidal exploration of pins, some results obtained in hot cell during examinations experimental pins and previously tested by Foucault current [fr

A grey incidence algorithm to detect high-Z material using cosmic ray muons

Science.gov (United States)

He, W.; Xiao, S.; Shuai, M.; Chen, Y.; Lan, M.; Wei, M.; An, Q.; Lai, X.

2017-10-01

Muon scattering tomography (MST) is a method for using cosmic muons to scan cargo containers and vehicles for special nuclear materials. However, the flux of cosmic ray muons is low, in the real life application, the detection has to be done a short timescale with small numbers of muons. In this paper, we present a novel approach to detection of special nuclear material by using cosmic ray muons. We use the degree of grey incidence to distinguish typical waste fuel material, uranium, from low-Z material, medium-Z material and other high-Z materials of tungsten and lead. The result shows that using this algorithm, it is possible to detect high-Z materials with an acceptable timescale.

Run-Beyond-Cladding-Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program

International Nuclear Information System (INIS)

Batte, G.L.; Hoffman, G.L.

1990-01-01

In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab

Modeling the highway transportation of spent fuel

International Nuclear Information System (INIS)

Harrison, I.G.

1986-01-01

There will be a substantial increase in the number of spent fuel shipments on the nation's highway system in the next thirty years. Most of the spent fuel will be moving from reactors to a spent fuel repository. This study develops two models that evaluate the risk and cost of moving the spent fuel. The Minimum Total Transport Risk Model (MTTRM) seeks an efficient solution for this problem by finding the minimum risk path through the network and sending all the spent fuel shipments over this one path. The Equilibrium Transport Risk Model (ETRM) finds an equitable solution by distributing the shipments over a number of paths in the network. This model decreases the risk along individual paths, but increases society's risk because the spent fuel shipments are traveling over more links in the network. The study finds that there is a trade off between path risk and societal risk. As path risk declines, societal risk rises. The cost of shipping also increases as the number of paths expand. The cost and risk of shipping spent fuel from ten reactors to four potential repository sites are evaluated using the MTTRM. The temporary monitored retrievable storage (MRS) facility in Tennessee is found to be the minimum cost and minimum risk solution. When direct shipment to the permanent sites is considered, Deaf Smith, Texas is the least cost and least incident free transport risk location. Yucca Mountain, Nevada is the least risk location when the focus is placed on the potential consequences of an accident

Design of durability test protocol for vehicular fuel cell systems operated in power-follow mode based on statistical results of on-road data

Science.gov (United States)

Xu, Liangfei; Reimer, Uwe; Li, Jianqiu; Huang, Haiyan; Hu, Zunyan; Jiang, Hongliang; Janßen, Holger; Ouyang, Minggao; Lehnert, Werner

2018-02-01

City buses using polymer electrolyte membrane (PEM) fuel cells are considered to be the most likely fuel cell vehicles to be commercialized in China. The technical specifications of the fuel cell systems (FCSs) these buses are equipped with will differ based on the powertrain configurations and vehicle control strategies, but can generally be classified into the power-follow and soft-run modes. Each mode imposes different levels of electrochemical stress on the fuel cells. Evaluating the aging behavior of fuel cell stacks under the conditions encountered in fuel cell buses requires new durability test protocols based on statistical results obtained during actual driving tests. In this study, we propose a systematic design method for fuel cell durability test protocols that correspond to the power-follow mode based on three parameters for different fuel cell load ranges. The powertrain configurations and control strategy are described herein, followed by a presentation of the statistical data for the duty cycles of FCSs in one city bus in the demonstration project. Assessment protocols are presented based on the statistical results using mathematical optimization methods, and are compared to existing protocols with respect to common factors, such as time at open circuit voltage and root-mean-square power.

Alameda-Contra Costa Transit District (AC Transit) Fuel Cell Transit Buses: Preliminary Evaluation Results

Energy Technology Data Exchange (ETDEWEB)

Chandler, K.; Eudy, L.

2007-03-01

This report provides an evaluation of three prototype fuel cell-powered transit buses operating at AC Transit in Oakland, California, and six baseline diesel buses similar in design to the fuel cell buses.

Incidence and survival patterns of rare anal canal neoplasms using the surveillance epidemiology and end results registry.

Science.gov (United States)

Metildi, Cristina; McLemore, Elisabeth C; Tran, Thuy; Chang, David; Cosman, Bard; Ramamoorthy, Sonia L; Saltzstein, Sidney L; Sadler, Georgia Robins

2013-10-01

Small cell, neuroendocrine tumors, and melanoma of the anus are rare. Limited data exist on the incidence and management for these rare tumors. A large, prospective, population-based database was used to determine incidence and survival patterns of rare anal neoplasms. The Surveillance, Epidemiology and End Results registry was queried to identify patients diagnosed with anal canal neoplasms. Incidence and survival patterns were evaluated with respect to age, sex, race, histology, stage, and therapy. We identified 7078 cases of anal canal neoplasms: melanoma (n = 149), neuroendocrine (n = 61), and small cell neuroendocrine (n = 26). Squamous cell carcinoma (SCC) (n = 6842) served as the comparison group. Anal melanoma (AM) demonstrated the lowest survival rate at 2.5 per cent. Neuroendocrine tumors (NETs) demonstrated similar survival as SCC (10-year survival for regional disease of 25 and 22.3%, respectively). Ten-year survival of small cell NETs resembled AM (5.3 vs 2.5%). Age 60 years or older, sex, black race, stage, and surgery were independent predictors of survival. This study presents the largest patient series of rare anal neoplasms. NETs of the anal canal demonstrate similar survival patterns to SCC, whereas small cell NETs more closely resemble AM. Accurate histologic diagnosis is vital to determine treatment and surgical management because survival patterns can differ among rare anal neoplasms.

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

Energy Technology Data Exchange (ETDEWEB)

Michael A. Pope

2011-10-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

Association of tobacco habits, including bidi smoking, with overall and site-specific cancer incidence: results from the Mumbai cohort study

Science.gov (United States)

Pednekar, Mangesh S.; Gupta, Prakash C.; Yeole, Balkrishna B.; Hébert, James R.

2013-01-01

Objective Bidis are hand-rolled cigarettes commonly smoked in South Asia and are marketed to Western populations as a safer alternative to conventional cigarettes. This study examined the association between bidis and other forms of tobacco use and cancer incidence in an urban developing country population. Methods Using data from the large, well-characterized Mumbai cohort study, adjusted hazard ratios (HRs) and 95% confidence intervals (CIs) were computed from Cox proportional hazards regression models in order to compare the relative effect of various forms of tobacco use on cancer incidence. Results During 649,228 person-years of follow-up 1,267 incident cancers occurred in 87,222 male cohort members. Incident oral cancer in bidi smokers (HR = 3.55; 95% CI = 2.40,5.24) was 42% higher than in cigarette smokers (HR = 2.50;95% CI = 1.65,3.78). For all respiratory and intrathoracic organs combined, the increase was 69% (HR = 5.54; 95% CI = 3.46,8.87 vs. HR = 3.28; 95% CI = 1.99,5.39); for lung and larynx, the increases were 35 and 112%, respectively. Smokeless tobacco use was associated with cancers of the lip, oral cavity, pharynx, digestive, respiratory, and intrathoracic organs. Conclusions Despite marketing claims to the contrary, we found that smokeless tobacco use and bidi smoking are at least as harmful as cigarette smoking for all incident cancers and are associated with increased risk of oral and respiratory/intrathoracic cancers. PMID:21431915

Short term endurance results on a single cylinder diesel engine fueled with upgraded bio oil biodiesel emulsion

Science.gov (United States)

Prakash, R.; Murugan, S.

2017-11-01

This paper deliberates the endurance test outcomes obtained from a single cylinder, diesel engine fueled with an upgraded bio oil biodiesel emulsion. In this investigation a bio oil obtained by pyrolysis of woody biomass was upgraded with acid treatment. The resulted bio oil was emulsified with addition of biodiesel and suitable surfactant which is termed as ATJOE15. The main objective of the endurance test was to evaluate the wear characteristics of the engine components and lubrication oil properties, when the engine is fueled with the ATJOE15 emulsion. The photographic views taken before and after the end of 100 hrs endurance test, and visual inspection of the engine components, wear and carbon deposit results, are discussed in this paper.

Cumulative incidence, distribution, and determinants of catastrophic health expenditure in Nepal: results from the living standards survey.

Science.gov (United States)

Ghimire, Mamata; Ayer, Rakesh; Kondo, Masahide

2018-02-14

Nepal has committed to the global community to achieve universal health coverage by 2030. Nevertheless, Nepal still has a high proportion of out-of-pocket health payment and a limited risk-pooling mechanism. Out-of-pocket payment for the healthcare services could result in catastrophic health expenditure (CHE). Evidence is required to effectively channel the efforts to lower those expenses in order to achieve universal health coverage. However, little is known about CHE and its determinants in a broad national context in Nepal. Therefore, this study was conducted to explore the cumulative incidence, distribution, and determinants of CHE in Nepal. Data were obtained from the nationally representative survey, the Nepal Living Standards Survey-third undertaken in 2010/11. Information from 5988 households was used for the analyses. Households were classified as having CHE when their out-of-pocket health payment was greater than or equal to 40% of their capacity to pay. Remaining households were classified as not having CHE. Logistic regression analyses were used to identify determinants of CHE. Based on household-weighted sample, the cumulative incidence of CHE was 10.3% per month in Nepal. This incidence was concentrated in the far-western region and households in the poorer expenditure quartiles. Multivariable logistic regression revealed that households were more likely to face CHE if they; consisted of chronically ill member(s), have a higher burden of acute illness and injuries, have elderly (≥60 years) member(s), belonged to the poor expenditure quartile, and were located in the far-western region. In contrast, households were less likely to incur CHE when their household head was educated. Having children (≤5 years) in households did not significantly affect catastrophic health expenditure. This study identified a high cumulative incidence of CHE. CHE was disproportionately concentrated in the poor households and households located in the far

JP-8 jet fuel can promote auditory impairment resulting from subsequent noise exposure in rats.

Science.gov (United States)

Fechter, Laurence D; Gearhart, Caroline; Fulton, Sherry; Campbell, Jerry; Fisher, Jeffrey; Na, Kwangsam; Cocker, David; Nelson-Miller, Alisa; Moon, Patrick; Pouyatos, Benoit

2007-08-01

We report on the transient and persistent effects of JP-8 jet fuel exposure on auditory function in rats. JP-8 has become the standard jet fuel utilized in the United States and North Atlantic Treaty Organization countries for military use and it is closely related to Jet A fuel, which is used in U.S. domestic aviation. Rats received JP-8 fuel (1000 mg/m(3)) by nose-only inhalation for 4 h and half of them were immediately subjected to an octave band of noise ranging between 97 and 105 dB in different experiments. The noise by itself produces a small, but permanent auditory impairment. The current permissible exposure level for JP-8 is 350 mg/m(3). Additionally, a positive control group received only noise exposure, and a fourth group consisted of untreated control subjects. Exposures occurred either on 1 day or repeatedly on 5 successive days. Impairments in auditory function were assessed using distortion product otoacoustic emissions and compound action potential testing. In other rats, tissues were harvested following JP-8 exposure for assessment of hydrocarbon levels or glutathione (GSH) levels. A single JP-8 exposure by itself at 1000 mg/m(3) did not disrupt auditory function. However, exposure to JP-8 and noise produced an additive disruption in outer hair cell function. Repeated 5-day JP-8 exposure at 1000 mg/m(3) for 4 h produced impairment of outer hair cell function that was most evident at the first postexposure assessment time. Partial though not complete recovery was observed over a 4-week postexposure period. The adverse effects of repeated JP-8 exposures on auditory function were inconsistent, but combined treatment with JP-8 + noise yielded greater impairment of auditory function, and hair cell loss than did noise by itself. Qualitative comparison of outer hair cell loss suggests an increase in outer hair cell death among rats treated with JP-8 + noise for 5 days as compared to noise alone. In most instances, hydrocarbon constituents of the fuel

78 FR 77606 - Security Requirements for Facilities Storing Spent Nuclear Fuel

Science.gov (United States)

2013-12-24

... NUCLEAR REGULATORY COMMISSION 10 CFR Parts 72 and 73 [NRC-2009-0558] RIN 3150-AI78 Security... rulemaking that would revise the security requirements for storing spent nuclear fuel (SNF) in an independent... Nuclear Security and Incident Response, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001...

Measurement of fuel importance distribution in non-uniformly distributed fuel systems

International Nuclear Information System (INIS)

Yamane, Yoshihiro; Hirano, Yasushi; Yasui, Hazime; Izima, Kazunori; Shiroya, Seiji; Kobayashi, Keiji.

1995-01-01

A reactivity effect due to a spatial variation of nuclear fuel concentration is an important problem for nuclear criticality safety in a reprocessing plant. As a theory estimating this reactivity effect, the Goertzel and fuel importance theories are well known. It has been shown that the Goertzel's theory is valid in the range of our experiments based on measurements of reactivity effect and thermal neutron flux in non-uniformly distributed fuel systems. On the other hand, there have been no reports concerning systematic experimental studies on the flatness of fuel importance which is a more general index than the Goertzel's theory. It is derived from the perturbation theory that the fuel importance is proportional to the reactivity change resulting from a change of small amount of fuel mass. Using a uniform and three kinds of nonuniform fuel systems consisting of 93.2% enriched uranium plates and polyethylene plates, the fuel importance distributions were measured. As a result, it was found experimentally that the fuel importance distribution became flat, as its reactivity effect became large. Therefore it was concluded that the flatness of fuel importance distribution is the useful index for estimating reactivity effect of non-uniformly distributed fuel system. (author)

The conversion of NRU from HEU to LEU fuel

International Nuclear Information System (INIS)

Sears, D.F.; Atfield, M.D.; Kennedy, I.C.

1990-01-01

The program at Chalk River Nuclear Laboratories (CRNL) to develop and test low-enriched uranium fuel (LEU, 3 Si, USiAl, USi Al and U 3 Si 2 (U-3.96 wt% Si; U-3.5 wt% Si-1.5 wt% AL; U-3.2 wt%; Si-3 wt% Al; U-7.3 wt% Si, respectively). Fuel elements were fabricated with uranium loadings suitable for NRU, 3.15 gU/cm 3 , and for NRX, 4.5 gU/cm 3 , and were irradiated under normal fuel-operating conditions. Eight experimental irradiations involving 100 mini-elements and 84 full-length elements (7X12-element rods) were completed to qualify the LEU fuel and the fabrication technology. Post irradiation examinations confirmed that the performance of the LEU fuel, and that of a medium enrichment uranium (MEU, 45% U-235) alloy fuel tested as a back-up, was comparable to the HEU fuel. The uranium silicide dispersion fuel swelling was approximately linear up to burnups exceeding NRU's design terminal burnup (80 at%). NRU was partially converted to LEU fuel when the first 31 prototype fuel rods manufactured with industrial scale production equipment were installed in the reactor. The rods were loaded in NRU at a fuelling rate of about two rods per week over the period 1988 September to December. This partial LEU core (one third of a full NRU core) has allowed the reactor engineers and physicists to evaluate the bulk effects of the LEU conversion on NRU operations. As expected, the irradiation is proceeding without incident

Preparation results for lifetime test of conversion LEU fuel in plutonium production reactors

International Nuclear Information System (INIS)

Vatulin, A.; Stetskiy, Yu.; Kukharkin, N.; Kalougin, A.; Gavrilov, P.; Ivanov, A.

1999-01-01

The program of converting Russian production reactors for the purpose to stop their plutonium fabrication is currently in progress. The program also provides for operation of these reactors under the conversion mode with using of low-enriched fuel (LEU). LEU fuel elements were developed and activities related to their preparation for reactor tests were carried out. (author)

75 FR 42237 - Regulation of Fuels and Fuel Additives: 2011 Renewable Fuel Standards

Science.gov (United States)

2010-07-20

... Based on a Thermochemical Platform 3. Hybrid Thermochemical/Biochemical Processes 4. Pyrolysis and Depolymerization a. Pyrolysis Diesel Fuel and Gasoline b. Catalytic Depolymerization 5. Catalytic Reforming of... result would be that additional volumes of conventional renewable fuel, such as corn-starch ethanol...

Results of physics start-up tests of Mochovce and Bohunice units with 2-nd generation Gd fuel (average enrichment 4.87 %)

International Nuclear Information System (INIS)

Polakovic, F.

2015-01-01

There are presented main features of the fuel and the list of experimental neutron-physical characteristics measured during physics start-up tests.All together there were carried out 14 physics start-ups at Bohunice and Mochovce Units with the new type of fuel. Differences between theoretical and experimental neutron-physical characteristics were statistically processed and compared with the tests acceptance criteria. There are summarized results of reactor physics start-ups with 2-nd generation Gd fuel usage [ru

Measurement of gamma ray from fuel of high temperature engineering test reactor. Method of measurement and results

Energy Technology Data Exchange (ETDEWEB)

Fujimoto, Nozomu; Nojiri, Naoki; Takada, Eiji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

2001-02-01

To obtain information in the HTTR core directly, gamma ray from fuel blocks was measured when fuel blocks were discharged from the core and reloaded to the core. Gamma ray was measured using GM detector, CZT semiconductor detector installed in a door valve and area monitors installed in a stand pipe compartment. The measurement was carried out for 20 fuel blocks in 4 columns considering the symmetry of uranium enrichment distribution in the core. Relative axial distribution in each column obtained by the GM detector and CZT detector agreed with calculated results. However, calculation values showed higher values than measured values in upper region of the core, lower those in lower region of the core. The axial distributions were also evaluated by the area monitors. The measured values agreed with calculated values. It became clear that it was possible to obtain the data inside the core by this method. (author)

Incidents with hazardous radiation sources; Zwischenfaelle mit gefaehrlichen Strahlenquellen

Energy Technology Data Exchange (ETDEWEB)

Schoenhacker, Stefan [Bundesministerium fuer Inneres, Traiskirchen (Austria). Abt. 1/9 - Zivilschutzschule

2016-07-01

Incidents with hazardous radiation sources can occur in any country, even those without nuclear facilities. Preparedness for such incidents is supposed to fulfill globally agreed minimum standards. Incidents are categorized in incidents with licensed handling of radiation sources as for material testing, transport accidents of hazardous radiation sources, incidents with radionuclide batteries, incidents with satellites containing radioactive inventory, incidents wit not licensed handling of illegally acquired hazardous radiation sources. The emergency planning in Austria includes a differentiation according to the consequences: incidents with release of radioactive materials resulting in restricted contamination, incidents with release of radioactive materials resulting in local contamination, and incidents with the hazard of e@nhanced exposure due to the radiation source.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

Peak-Seeking Control For Reduced Fuel Consumption: Flight-Test Results For The Full-Scale Advanced Systems Testbed FA-18 Airplane

Science.gov (United States)

Brown, Nelson

2013-01-01

A peak-seeking control algorithm for real-time trim optimization for reduced fuel consumption has been developed by researchers at the National Aeronautics and Space Administration (NASA) Dryden Flight Research Center to address the goals of the NASA Environmentally Responsible Aviation project to reduce fuel burn and emissions. The peak-seeking control algorithm is based on a steepest-descent algorithm using a time-varying Kalman filter to estimate the gradient of a performance function of fuel flow versus control surface positions. In real-time operation, deflections of symmetric ailerons, trailing-edge flaps, and leading-edge flaps of an F/A-18 airplane are used for optimization of fuel flow. Results from six research flights are presented herein. The optimization algorithm found a trim configuration that required approximately 3 percent less fuel flow than the baseline trim at the same flight condition. This presentation also focuses on the design of the flight experiment and the practical challenges of conducting the experiment.

Dry vault for spent fuel depository, basic outsets, operating results and safety of the 'CASCAD' plant

International Nuclear Information System (INIS)

Bardelle, Ph.; Mercier, J.P.

1995-01-01

CEA built a facility in which spent fuel can be stored for a few decades (50 years), until favourable conditions prevail for its disposal. The main features of this project consist of a dry depository, which presents a low cost of operation, against a wet one which is more expensive due to the circulation and the continuous controls of the water. Fuel elements are cooled by a fully passive air circulation. This process allows a good efficiency without mechanical equipment and works all the better as the amount of heat to exhaust is great, in the limits of the design. This facility, known as 'CASCAD' (shortening for CASemate (=vault) CADarache) started up in 1990, and received its first canister of fuel on May 29th 1990. The basic design data of the facility, are reviewed the main techniques used for its construction are outlined the safety concepts are drawn and the first results determined by a looking-back over 4 years of working are presented. (K.A.). 1 tab

Inert matrix fuel in dispersion type fuel elements

Energy Technology Data Exchange (ETDEWEB)

Savchenko, A.M. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)]. E-mail: sav@bochvar.ru; Vatulin, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Morozov, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Sirotin, V.L. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Dobrikova, I.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kulakov, G.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Ershov, S.A. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kostomarov, V.P. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Stelyuk, Y.I. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)

2006-06-30

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg{sup -1} (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Inert matrix fuel in dispersion type fuel elements

Science.gov (United States)

Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

2006-06-01

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Transient and fuel performance analysis with VTT's coupled code system

International Nuclear Information System (INIS)

Daavittila, A.; Hamalainen, A.; Raty, H.

2005-01-01

VTT (technical research center of Finland) maintains and further develops a comprehensive safety analysis code system ranging from the basic neutronic libraries to 3-dimensional transient analysis and fuel behaviour analysis codes. The code system is based on various types of couplings between the relevant physical phenomena. The main tools for analyses of reactor transients are presently the 3-dimensional reactor dynamics code HEXTRAN for cores with a hexagonal fuel assembly geometry and TRAB-3D for cores with a quadratic fuel assembly geometry. HEXTRAN has been applied to safety analyses of VVER type reactors since early 1990's. TRAB-3D is the latest addition to the code system, and has been applied to BWR and PWR analyses in recent years. In this paper it is shown that TRAB-3D has calculated accurately the power distribution during the Olkiluoto-1 load rejection test. The results from the 3-dimensional analysis can be used as boundary conditions for more detailed fuel rod analysis. For this purpose a general flow model GENFLO, developed at VTT, has been coupled with USNRC's FRAPTRAN fuel accident behaviour model. The example case for FRAPTRAN-GENFLO is for an ATWS at a BWR plant. The basis for the analysis is an oscillation incident in the Olkiluoto-1 BWR during reactor startup on February 22, 1987. It is shown that the new coupled code FRAPTRAN/GENFLO is quite a promising tool that can handle flow situations and give a detailed analysis of reactor transients

Ohio's First Ethanol-Fueled Light-Duty Fleet: Final Study Results

Energy Technology Data Exchange (ETDEWEB)

Battelle

1998-10-01

In 1996, the State of Ohio established a project to demonstrate the use of an ethanol blend (E85, which is 85% transportation-grade ethanol and 15% gasoline) as a transportation fuel in flexible-fuel vehicles (FFVs). The study included ten FFVs and three gasoline vehicles (used as control vehicles) operated by five state agencies. The project included 24 months of data collection on vehicle operations. This report presents the data collection and analysis from the study, with a focus on the last year.

Fuel Burn Estimation Model

Science.gov (United States)

Chatterji, Gano

2011-01-01

Conclusions: Validated the fuel estimation procedure using flight test data. A good fuel model can be created if weight and fuel data are available. Error in assumed takeoff weight results in similar amount of error in the fuel estimate. Fuel estimation error bounds can be determined.

Considerations for handling failed fuel at the Barnwell Nuclear Fuel Plant

International Nuclear Information System (INIS)

Anderson, R.T.; Cholister, R.J.

1982-05-01

The impact of failed fuel receipt on reprocessing operations is qualitatively described. It appears that extended storage of fuel, particularly with advanced storage techniques, will increase the quantity of failed fuel, the nature and possibly the configuration of the fuel. The receipt of failed fuel at the BNFP increases handling problems, waste volumes, and operator exposure. If it is necessary to impose special operating precautions to minimize this impact, a loss in plant throughput will result. Hence, ideally, the reprocessing plant operator would take every reasonable precaution so that no failed fuel is received. An alternative policy would be to require that failed fuel be placed in a sealed canister. In the latter case the canister must be compatible with the shipping cask and suitable for in-plant storage. A required inspection of bare fuel would be made at the reactor prior to shipping off-site. This would verify fuel integrity. These requirements are obviously idealistic. Due to the current uncertain status of reprocessing and the need to keep reactors operating, business or governmental policy may be enacted resulting in the receipt of a negotiated quantity of non-standard fuel (including failed fuel). In this situation, BNFP fuel receiving policy based soley on fuel cladding integrity would be difficult to enforce. There are certain areas where process incompatibility does exist and where a compromise would be virtually impossible, e.g., canned fuel for which material or dimensional conflicts exist. This fuel would have to be refused or the fuel would require recanning prior to shipment. In other cases, knowledge of the type and nature of the failure may be acceptable to the operator. A physical inspection of the fuel either before shipment or after the cask unloading operation would be warranted. In this manner, concerns with pool contamination can be identified and the assembly canned if deemed necessary

DUPIC fuel compatibility assessment

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Rho, G. H.; Park, J. W. [and others

2000-03-01

The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition.

DUPIC fuel compatibility assessment

International Nuclear Information System (INIS)

Choi, Hang Bok; Rho, G. H.; Park, J. W. and others

2000-03-01

The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition

Syphilis incidence among men who have sex with men in China: results from a meta-analysis.

Science.gov (United States)

Chen, Guohong; Cao, Ya; Yao, Yuan; Li, Ming; Tang, Weiming; Li, Jianjun; Babu, Giridhara R; Jia, Yue; Huan, Xiping; Xu, Genxing; Yang, Haitao; Fu, Gengfeng; Li, Lei

2017-02-01

The recent upsurge of syphilis infections among men who have sex with men (MSM) is one of the major challenges facing China. However, the overall burden is still not clear. This study aims to summarize the incidence of syphilis among MSM in China by using meta-analysis. We comprehensively searched PubMed-MEDLINE, China National Knowledge Infrastructure and Chinese Wanfang databases. Articles published between December 2009 and March 2015 that met the inclusion criteria were considerably involved in this meta-analysis. Two reviewers performed a quality assessment of the studies and extracted data for estimating the overall syphilis incidence. STATA 12.0 was used to summarize the overall incidence of syphilis. In all, 14 studies from 13 papers were included in this study. Follow-up duration of these studies ranged from six to 36 months, while drop-out rates ranged from 11.9% to 83.6%. The individual incidence rates of the included studies varied from 3.1/100 person-years (95% CI, 0.8-5.3/100 person-years) to 38.5/100 person-years (95% CI, 28.9-48.1/100 person-years), with a pooled incidence of 9.6/100 person-years (95% CI, 7.0-12.2/100 person-years). The subgroup meta-analysis revealed that incidence estimates were 38.5/100 person-years (95% CI, 28.9-48.1/100 person-years), 12.1/100 person-years (95% CI, 7.0-17.2/100 person-years), 11.2/100 person-years (95% CI, 0.7-23.1/100 person-years), 8.9/100 person-years (95% CI, 6.5-11.2/100 person-years), 5.7/100 person-years (95% CI, 3.4-8.0/100 person-years) and 3.1/100 person-years (95% CI, 0.8-5.3/100 person-years) in Northeast, North, Southwest, East, South and Northwest China, respectively. Syphilis incidence among Chinese MSM is high, and this may increase the spread of other sexually transmitted infections, including human immunodeficiency virus. It is essential to integrate syphilis control programs with HIV control programs. This can be achieved by establishing public health response systems to monitor and control

Fuel rod behaviour at high burnup WWER fuel cycles

International Nuclear Information System (INIS)

Medvedev, A.; Bogatyr, S.; Kouznetsov, V.; Khvostov, G.; Lagovsky; Korystin, L.; Poudov, V.

2003-01-01

The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles