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Sample records for spent medium analysis

  1. Safety analysis of spent fuel packaging

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

    1987-01-01

    Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

  2. Cost analysis methodology of spent fuel storage

    International Nuclear Information System (INIS)

    1994-01-01

    The report deals with the cost analysis of interim spent fuel storage; however, it is not intended either to give a detailed cost analysis or to compare the costs of the different options. This report provides a methodology for calculating the costs of different options for interim storage of the spent fuel produced in the reactor cores. Different technical features and storage options (dry and wet, away from reactor and at reactor) are considered and the factors affecting all options defined. The major cost categories are analysed. Then the net present value of each option is calculated and the levelized cost determined. Finally, a sensitivity analysis is conducted taking into account the uncertainty in the different cost estimates. Examples of current storage practices in some countries are included in the Appendices, with description of the most relevant technical and economic aspects. 16 figs, 14 tabs

  3. An analysis of spent fuel characteristics for reactor RA spent fuel elements

    International Nuclear Information System (INIS)

    Milosevic, M.; Vukadin, Z.

    2001-01-01

    The need for reducing the intrinsic conservatism in the criticality safety assessment has led to the development of validated methods for assessing the reactivity effects associated with fuel burnup. This paper gives an overview of the work being performed to investigate the spent fuel characteristics and nuclear criticality safety of the reactor RA spent fuel storage. An analysis methodology is presented along with information representing the validation of the methods and geometrical models. Finally, the criticality safety analysis of the stainless steel containers and aluminum barrels, filled with spent fuel elements, is presented to demonstrate that an adequate margin of subcriticality is proved for the reactor RA spent fuel storage.(author)

  4. Development of advanced spent fuel management process. System analysis of advanced spent fuel management process

    International Nuclear Information System (INIS)

    Ro, S.G.; Kang, D.S.; Seo, C.S.; Lee, H.H.; Shin, Y.J.; Park, S.W.

    1999-03-01

    The system analysis of an advanced spent fuel management process to establish a non-proliferation model for the long-term spent fuel management is performed by comparing the several dry processes, such as a salt transport process, a lithium process, the IFR process developed in America, and DDP developed in Russia. In our system analysis, the non-proliferation concept is focused on the separation factor between uranium and plutonium and decontamination factors of products in each process, and the non-proliferation model for the long-term spent fuel management has finally been introduced. (Author). 29 refs., 17 tabs., 12 figs

  5. Hydrometallurgical route to recover molybdenum, nickel, cobalt and aluminum from spent hydrotreating catalysts in acid medium

    International Nuclear Information System (INIS)

    Valverde Junior, Ivam Macedo; Paulino, Jessica Frontino; Afonso, Julio Carlos

    2008-01-01

    This work describes a hydrometallurgical route for processing spent commercial catalysts (CoMo and NiMo/Al 2 O 3 ). Samples were preoxidized (500 deg C, 5 h) in order to eliminate coke and other volatile species present. The calcined solid was dissolved in concentrated H 2 SO 4 and water (1:1 vol/vol) at 90 de C; the insoluble matter was separated from the solution. Molybdenum was recovered by solvent extraction using tertiary amines at pH around 1.8. Cobalt (or nickel) was separated by addition of aqueous ammonium oxalate at the above pH. Phosphorus was removed by passing the liquid through a strong anion exchange column. Aluminum was recovered by neutralizing the solution with NaOH. The route presented in this work generates less final aqueous wastes because it is not necessary to use alkaline medium during the metal recovery steps. (author)

  6. Thermal analysis of spent nuclear fuels repository

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, F.; Salome, J.; Cardoso, F.; Velasquez, C.E.; Pereira, C. [Departamento de Engenharia Nuclear - Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Pampulha, Belo Horizonte MG, CEP 31270-901 (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores - CNPq, Asa Norte, Brazilia (Brazil); Viana, C. [Departamento de Engenharia Nuclear - Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Pampulha, Belo Horizonte MG, CEP 31270-901 (Brazil); Barros, G.P. [Comissao Nacional de Energia Nuclear-CNEN, Rua Gal Severiano, n 90 - Botafogo, 22290-901, Rio de Janeiro, RJ (Brazil)

    2016-07-01

    In the first part, Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) spent fuels (SF) were evaluated to the thermal of the spent fuel pool (SFP) without an external cooling system. The goal is to compare the water boiling time of the pool storing different types of spent nuclear fuels. This study used the software ANSYS Workbench 16.2 - student version. For the VHTR, two types of fuel were analyzed: (Th,TRU)O{sub 2} and UO{sub 2}. This part of the studies were performed for wet storage condition using a single type of SF and decay heat values at times t=0 and t=10 years after the reactor discharge. The ANSYS CFX module was used and the results show that the time that water takes to reach the boiling point varies from 2.4 minutes for the case of VHTR-(Th,TRU)O{sub 2} SF at time t=0 year after reactor discharge until 32.4 hours for the case of PWR SF at time t=10 years after the discharge reactor. The second part of this work consists of modeling a geological repository. Firstly, the temperature evaluation of the spent fuel from a PWR was analyzed. A PWR canister was simulated using the ANSYS transient thermal module. Then the temperature of canister could be computed during the time spent on a portion of a geological repository. The mean temperature on the canister surface increased during the first nine years, reaching a plateau at 35.5 C. degrees between the tenth and twentieth years after the geological disposal. The idea is to extend this study for the other systems analyzed in the first part. The idea is to include in the study, the spent fuels from VHTR and ADS and to compare the canister behavior using different spent fuels. (authors)

  7. Spent Nuclear Fuel Alternative Technology Decision Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, C.B.

    1999-11-29

    The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

  8. Spent Nuclear Fuel Alternative Technology Decision Analysis

    International Nuclear Information System (INIS)

    Shedrow, C.B.

    1999-01-01

    The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology

  9. Spent nuclear fuel project - criteria document spent nuclear fuel final safety analysis report

    International Nuclear Information System (INIS)

    MORGAN, R.G.

    1999-01-01

    The criteria document provides the criteria and planning guidance for developing the Spent Nuclear Fuel (SNF) Final Safety Analysis Report (FSAR). This FSAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, installation acceptance testing, startup, and operation of the SNF Project facilities (K Basins, Cold Vacuum Drying Facility, and Canister Storage Building)

  10. Impact analysis of spent fuel jacket assemblies

    International Nuclear Information System (INIS)

    Aramayo, G.A.

    1994-01-01

    As part of the analyses performed in support of the reracking of the High Flux Isotope Reactor pool, it became necessary to prove the structural integrity of the spent fuel jacket assemblies subjected to gravity drop that result from postulated accidents associated with the handling of these assemblies while submerged in the pool. The spent fuel jacket assemblies are an integral part of the reracking project, and serve to house fuel assemblies. The structure integrity of the jacket assemblies from loads that result from impact from a height of 10 feet onto specified targets has been performed analytically using the computer program LS-DYNA3D. Nine attitudes of the assembly at the time of impact have been considered. Results of the analyses show that there is no failure of the assemblies as a result of the impact scenarios considered

  11. Long-term outdoor cultivation by perfusing spent medium for biodiesel production from Chlorella minutissima.

    Science.gov (United States)

    Oh, Sung Ho; Kwon, Min Chul; Choi, Woon Yong; Seo, Yong Chang; Kim, Ga Bin; Kang, Do Hyung; Lee, Shin Young; Lee, Hyeon Yong

    2010-08-01

    A unique perfusion process was developed to maintain high concentrations of marine alga, Chlorella minutissima. This method is based on recycling cells by continuous feeding with warm spent sea water from nuclear power plants, which has very similar properties as sea water. A temperature of at least 30 degrees C in a 200 L photo-bioreactor was maintained in this system by perfusion of the thermal plume for 80 days in the coldest season. The maximum cell concentration and total lipid content was 8.3 g-dry wt./L and 23.2 %, w/w, respectively, under mixotrophic conditions. Lipid production was found to be due to a partially or non-growth related process, which implies that large amounts of biomass are needed for a high accumulation of lipids within the cells. At perfusion rates greater than 1.5 L/h, the temperature of the medium inside the reactor was around 30 degrees C, which was optimal for cell growth. For this system, a perfusion rate of 2.8 L/h was determined to be optimal for maintaining rapid cell growth and lipid production during outdoor cultivation. It was absolutely necessary to maintain the appropriate perfusion rate so that the medium temperature was optimal for cell growth. In addition, the lipids produced using this process were shown to be feasible for biodiesel production since the lipid composition of C. minutissima grown under these conditions consisted of 17 % (w/w) of C(16) and 47% (w/w) of C(18). The combined results of this study clearly demonstrated that the discharged energy of the thermal plume could be reused to cultivate marine alga by maintaining a relatively constant temperature in an outdoor photo-bioreactor without the need for supplying any extra energy, which could allow for cheap production of biodiesel from waste energy. Copyright 2010 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  12. Spent fuel packaging and its safety analysis

    International Nuclear Information System (INIS)

    Takada, Kimitaka; Nakaoki, Kozo; Tamamura, Tadao; Matsuda, Fumio; Fukudome, Kazuyuki

    1983-01-01

    An all stainless steel B(U) type packaging is proposed to transport spent fuels discharged from research reactors and other radioactive materials. The package is used dry and provided with surface fins to absorb drop shock and to dissipate decay heat. Safety was analyzed for structural, thermal, containment shielding and criticality factors, and the integrity of the package was confirmed with the MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, and KENO computer codes. (author)

  13. Time history analysis method for spent fuel racks

    International Nuclear Information System (INIS)

    Li Chen; Qian Hao; Zhang Kai; Xu Dinggen; Xie Yongcheng

    2013-01-01

    Background: Spent fuel racks are important facilities to store the spent fuel which are free standing in the spent fuel pool. The response of racks to seismic load is highly nonlinear and involves a complex combination of motions: sliding, impact, twisting and turning. Purpose: An analysis method should be built to accurately replicate these nonlinear responses. Methods: The whole pool multi-rack FEA model was developed and time history analysis was performed which contains the consideration of effect of sliding, impact and friction and the fluid structure interaction effect. Results: The analysis results such as displacement and force under seismic loads were obtained. Conclusion: The method can be used to the seismic analysis for spent fuel racks. (authors)

  14. Thermal analysis for spent fuel casks

    International Nuclear Information System (INIS)

    Wells, A.H.

    1986-01-01

    Thermal analyses for spent fuel storage or transportation must demonstrate that the heat produced in the fuel will be removed without causing excessive fuel cladding temperatures or cask surface temperatures. The time required to ship fuel from a reactor to another site is a matter of days so temperature control during shipment is a short term concern and temperatures of 1100-1200 degrees F are acceptable in a hypothetical fire scenario, but storage is envisioned for twenty years or more and temperatures are limited to 380 degrees C (716 F). Elevated temperatures do more that weaken fuel cladding: the materials of the fuel basket also weaken with higher temperature, thermal stresses may cause cracks to form in welds, and impact limiter materials may soften or degrade. The high temperatures of a fire cause loss of liquid neutron shields and may cause solid materials to char and insulate a cask after the fire

  15. Cultivation of medicinal caterpillar fungus, Cordyceps militaris (Ascomycetes), and production of cordycepin using the spent medium from levan fermentation.

    Science.gov (United States)

    Wu, Fang-Chen; Chen, Yi-Lin; Chang, Shu-Ming; Shih, Ing-Lung

    2013-01-01

    A process of tandem cultivation for the production of green and invaluable bioproducts (levan and Cordycepes militaris) useful for medical applications has been successfully developed. The process involves first cultivating Bacillus subtilis strain natto in sucrose medium to produce levan, followed by the subsequent cultivation of C. militaris in liquid- and solid-state cultures using the spent medium from levan fermentation as substrates. The factors affecting the cell growth and production of metabolites of C. militaris were investigated, and the various metabolites produced in the culture filtrate, mycelia, and fruiting body were analyzed. In addition, cordycepin was prepared from the solid waste medium of C. militaris. This is an excellent example in the development of cost effective biorefineries that maximize useful product formation from the available biomass. The preparation of cordycepin from solid waste medium of C. militaris using a method with high extraction efficiency and minimum solvent usage is also environmentally friendly.

  16. Gold biorecovery from e-waste: An improved strategy through spent medium leaching with pH modification.

    Science.gov (United States)

    Natarajan, Gayathri; Ting, Yen-Peng

    2015-10-01

    Rapid technological advancement and relatively short life time of electronic goods have resulted in an alarming growth rate of electronic waste which often contains significant quantities of toxic and precious metals. Compared to conventional recovery methods, bioleaching is an environmentally friendly process for metal extraction. Gold was bioleached from electronic scrap materials (ESM) via gold-cyanide complexation using cyanide produced from pure and mixed cultures of cyanogenic bacteria Chromobacterium violaceum, Pseudomonas aeruginosa and Pseudomonas fluorescens. As ESM was toxic to the bacteria, a two-step bioleaching approach was adopted where the solid waste was added to the bacterial culture after it has reached maximum growth and cyanide production during early stationary phase. Pure culture of C. violaceum showed the highest cyanide production, yielding maximum gold recovery of 11.3% at 0.5% w/v pulp density of ESM in two-step bioleaching. At the same pulp density of ESM, spent medium bioleaching using bacterial cell-free metabolites achieved gold recovery of 18%. Recovery increased to 30% when the pH of the spent medium was increased to shift the equilibrium in favor of cyanide ions production. It is demonstrated for the first time that pH modification of spent medium further improved metal solubilization and yielded higher metal recovery (compared to two-step bioleaching). Copyright © 2015 Elsevier Ltd. All rights reserved.

  17. Cost analysis of spent nuclear fuel management

    International Nuclear Information System (INIS)

    Robertson, D.L.M.; Ford, L.M.

    1993-01-01

    The Department of Energy Civilian Radioactive Waste Management System (CRWMS) is chartered to develop a waste management system for the safe disposal of spent nuclear fuel (SNF) from the 131 nuclear power reactors in the United States and a certain amount of high level waste (HLW) from reprocessing operations. The current schedule is to begin accepting SNF in 1998 for storage at a Monitored Retrievable Storage (MRS) facility. Subsequently, beginning in 2010, the system is scheduled to begin accepting SNF at a permanent geologic repository in 2010 and HLW in 2015. At this time, a MRS site has not been selected. Yucca Mountain, Nevada is currently being evaluated as the candidate site for the repository for permanent geologic disposal of SNF. All SNF, with the possible exception of the SNF from the western reactors, is currently planned to be shipped to or through the MRS site en route to the repository. The repository will operate in an acceptance and performance confirmation phase for a 50 year period beginning in 2010 with an additional nine year closure and five year decontamination and decommissioning period. The MRS has a statutory maximum capacity of 15,000 Metric Tons Uranium (MTU), with a further restriction that it may not store more than 10,000 MTU until the repository begins accepting waste. The repository is currently scheduled to store 63,000 MTU of SNF and an additional 7,000 MTU equivalent of HLW for a total capacity of 70,000 MTU. The amended act specified the MRS storage limits and identified Yucca Mountain as the only site to be characterized. Also, an Office of the Nuclear Waste Negotiator was established to secure a voluntary host site for the MRS. The MRS, the repository, and all waste containers/casks will go through a Nuclear Regulatory Commission licensing process much like the licensing process for a nuclear power plant. Environmental assessments and impact statements will be prepared for both the MRS and repository

  18. Bacteriological analysis of spent engine oil contaminated soil ...

    African Journals Online (AJOL)

    The bacteriological analysis of soil contaminated with spent engine oil (SEO) planted with cowpea was investigated. The aim of this study was to detect the microbial degradation of SEO in soil and how it affects the microbial activity and the effects of SEO on the growth of cowpea. SEO collected from a mechanic workshop in ...

  19. Thermal analysis of transportation packaging for nuclear spent fuel

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki

    1989-01-01

    Safety analysis of transportation packaging for nuclear spent fuel comprises structural, thermal, containment, shielding and criticality factors, and the safety of a packaging is verified by these analyses. In thermal analysis, the temperature of each part of the packaging is calculated under normal and accident test conditions. As an example of thermal analysis, the temperature distribution of a packaging being subjected to a normal test was calculated by the TRUMP code and compared with measured data. (author)

  20. Risk analysis for nuclear spent fuel storage facility

    International Nuclear Information System (INIS)

    Dina, Dumitru; Andrei, Veronica; Ghita, Sorin; Glodeanu, Florin

    2004-01-01

    In June 2003, the first capacity of the Intermediate Dry Spent Fuel Storage Facility (DICA) was commissioned at Cernavoda Nuclear Power Plant (Cernavoda NPP). The facility is a dry system type facility; its designed lifetime is for a minimum of 50 years and capacity for two nuclear power units' lifetime. The storage structures are monolith reinforced concrete modules offering a very good isolation of the spent fuel from the environment. The spent fuel is confined by a system of double barriers that prevents radioactive emissions and ensures protection of the population and environment. The security functions of the facility are operational through passive means. In Romania, the National Commission for Nuclear Activities Control, CNCAN, is the authority that licenses the nuclear activities. CNCAN issued the commissioning and operating licenses for DICA following a complex process. The Final Nuclear Safety Report represents basic documentation for licensing and one of its main chapters presents the risk analysis results. The risk analysis performed for DICA covers normal operational regimes and accident cases considered as design basis events (DBE). The results of risk analysis for Cernavoda NNP DICA demonstrates that risks for the population and environment are much lower than the authorization limits established by CNCAN and in agreement with values for proven safe spent fuel storage technologies from European Union and worldwide. (authors)

  1. Spent nuclear fuel storage pool thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Gay, R.R.

    1984-01-01

    Storage methods and requirements for spent nuclear fuel at U.S. commercial light water reactors are reviewed in Section 1. Methods of increasing current at-reactor storage capabilities are also outlined. In Section 2 the development of analytical methods for the thermal-hydraulic analysis of spent fuel pools is chronicled, leading up to a discussion of the GFLOW code which is described in Section 3. In Section 4 the verification of GFLOW by comparisons of the code's predictions to experimental data taken inside the fuel storage pool at the Maine Yankee nuclear power plant is presented. The predictions of GFLOW using 72, 224, and 1584 node models of the storage pool are compared to each other and to the experimental data. An example of thermal licensing analysis for Maine Yankee using the GFLOW code is given in Section 5. The GFLOW licensing analysis is compared to previous licensing analysis performed by Yankee Atomic using the RELAP-4 computer code

  2. Safety analysis of disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Vieno, T.

    1994-04-01

    The spent fuel from the Olkiluoto NPP (TVO I and II) is planned to be disposed of in a repository to be constructed at a depth of about 500 meters in the crystalline bedrock. The thesis is dealing with the safety analysis of the disposal. The main topics presented in the thesis are: (1) The amount of radioactive properties of the spent fuel, (2) The canister design and the planned disposal concept, (3) The results of the preliminary site investigations, (4) Discussion of the multi-barrier principle, (5) The general principles and methodology of the TVO-92 safety analysis, (6) Groundwater flow analysis, (7) Durability and behaviour of the canister, (8) Biosphere analysis and reference scenario, and (9) The sensitivity and uncertainty analyses. (246 refs., 75 figs., 44 tabs.)

  3. Pyrolysis of FeCl3-pretreated spent coffee grounds using CO2 as a reaction medium

    International Nuclear Information System (INIS)

    Cho, Dong-Wan; Lee, Jechan; Yoon, Kwangsuk; Ok, Yong Sik; Kwon, Eilhann E.; Song, Hocheol

    2016-01-01

    Highlights: • CO 2 enhances the thermal cracking of VOCs in pyrolysis of spent coffee grounds. • Reaction between CO 2 and VOCs increases syngas yield and subsequent tar reduction. • FeCl 3 treatment strengthens CO 2 effects in pyrolysis of spent coffee grounds. - Abstract: Pyrolysis of spent coffee grounds (SCG) was performed to achieve the multiple purposes of waste disposal and energy recovery. This study placed great emphasis on pretreatment of SCG with FeCl 3 (Fe-SCG) and utilizing carbon dioxide (CO 2 ) as a reaction medium to enhance the generation of syngas while reducing condensable hydrocarbons (e.g., tar). For example, the principal effect of CO 2 was the enhanced generation of syngas via the CO 2 -induced thermal cracking of volatile organic compounds (VOCs) and the reaction between CO 2 and VOCs, which resulted in subsequent reduction of tar. These identified effects on pyrolysis of SCG were more pronounced in pyrolysis of Fe-SCG, which could be attributable to the catalytic effect of the Fe mineral formed from phase transition of FeCl 3 during pyrolysis. The generation of CO in pyrolysis of Fe-SCG in the presence of CO 2 increased up to 8000% as compared to pyrolysis of SCG in N 2 .

  4. Thermal test and analysis of a spent fuel storage cask

    International Nuclear Information System (INIS)

    Yamakawa, H.; Gomi, Y.; Ozaki, S.; Kosaki, A.

    1993-01-01

    A thermal test simulated with full-scale cask model for the normal storage was performed to verify the storage skill of the spent fuels of the cask. The maximum temperature at each point in the test was lower than the allowable temperature. The integrity of the cask was maintained. It was observed that the safety of containment system was also kept according to the check of the seal before and after the thermal test. Therefore it was shown that using the present skill, it is possible to store spent fuels in the dry-type cask safely. Moreover, because of the good agreement between analysis and experimental results, it was shown that the analysis model was successfully established to estimate the temperature distribution of the fuel cladding and the seal portion. (J.P.N.)

  5. A cost-benefit analysis of spent fuel management

    International Nuclear Information System (INIS)

    Lamorlette, G.

    2001-01-01

    The back end of the fuel cycle is an area of economic risk for utilities having nuclear power plants to generate electricity. A cost-benefit analysis is a method by which utilities can evaluate advantages and drawbacks of alternative back end fuel cycle strategies. The present paper analyzes how spent fuel management can influence the risks and costs incurred by a utility over the lifetime of its power plants and recommends a recycling strategy. (author)

  6. Probabilistic safety analysis of transportation of spent fuel

    International Nuclear Information System (INIS)

    Subramaniam, Chitra

    1999-11-01

    The report presents the results of the study carried out to estimate the accident risk involved in the transport of spent fuel from Rajasthan Atomic Power Station near Kota to the fuel reprocessing plant at Tarapur. The technique of probabilistic safety analysis is used. The fuel considered is the Indian pressurised heavy water reactor fuel with a minimum cooling period of 485 days. The spent fuel is transported in a cuboidal, naturally-cooled shipping cask over a distance of 822 km by rail. The Indian rail accident statistics are used to estimate the basic rail accident frequency. The possible ways in which a release of radioactive material can occur from the spent fuel cask are identified by the fault tree analysis technique. The release sequences identified are classified into eight accident severity categories, and release fractions are assigned to each. The consequences resulting from the release are estimated by the computer code RADTRAN 4. Results of the risk analysis indicate that the accident risk values are very low and hence acceptable. Parametric studies show that the risk would continue to be small even if the controlling parameters were to simultaneously take extreme adverse values. (author)

  7. Thermal analysis of cold vacuum drying of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Piepho, M.G.

    1998-07-20

    The thermal analysis examined transient thermal and chemical behavior of the Multi canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with Mark IV N, Reactor spent fuel in four fuel baskets and one scrap basket. This analysis provides the basis for the MCO thermal behavior at the CVD Facility for its Phase 2 Safety Analysis Report (revision 4).

  8. Prototype spent-fuel canister design, analysis, and test

    International Nuclear Information System (INIS)

    Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

    1982-03-01

    Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included

  9. TVO-92 safety analysis of spent fuel disposal

    International Nuclear Information System (INIS)

    Vieno, T.; Hautojaervi, A.; Koskinen, L.; Nordman, H.

    1993-08-01

    The spent fuel from the TVO I and TVO II reactors at the Olkiluoto nuclear power plant is planned to be disposed in a repository constructed at a depth of about 500 meters in crystalline bedrock. Teollisuuden Voima Oy (TVO) has carried out preliminary site investigations for spent fuel disposal between 1987 and 1992 at five areas in Finland (Olkiluoto, Kivetty, Romuvaara, Syyry and Veitsivaara). The Safety analysis of the disposal system is presented in the report. Spent fuel will be encapsulated in composite copper-steel canisters. The canister design (ACP canister) consists of an inner container of steel as a load-bearing element and an outer container of oxygen-free copper to provide a shield against corrosion. In the repository the canisters will be emplaced in vertical holes drilled in the floors of horizontal deposition tunnels. The annulus between the canister and the rock is filled with compacted bentonite. The results of the safety analysis attest that the planned disposal system fulfils the safety requirements. Suitable places for the repository can be found at each of the five investigation sites

  10. The analysis of the RA reactor irradiated fuel cooling in the spent fuel pool

    International Nuclear Information System (INIS)

    Vrhovac, M.; Afgan, N.; Spasojevic, D.; Jovic, V.

    1985-01-01

    According to the RA reactor exploitation plan the great quantity of the irradiated spent fuel will be disposed in the reactor spent fuel pool after each reactor campaign which will including the present spent fuel inventory increase the residual power level in the pool and will soon cause the pool capacity shortage. To enable the analysis of the irradiated fuel cooling the pool and characteristic spent fuel canister temperature distribution at the residual power maximum was done. The results obtained under the various spent fuel cooling conditions in the pit indicate the normal spent fuel thermal load even in the most inconvenient cooling conditions. (author)

  11. Analysis of burnup credit on spent fuel storage

    International Nuclear Information System (INIS)

    Matsumura, T.; Sasahara, A.

    1999-01-01

    Chemical analyses were carried out on high burnup UO 2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins. Measured data of the composition of nuclides from 234 U to 242 Pu were used for evaluation of ORIGEN-2/82 code. Criticality calculations were executed for the casks which were being designed to store 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for (1) axial and horizontal profiles of burnup, and void history (BWR), (2) operational histories such as control rod insertion history, BPR insertion history and others, and (3) calculational accuracy of ORIGEN-2/82 code on the composition of nuclides. Present evaluation shows that introduction of burnup credit has a substantial merit in criticality safety analysis of the cask, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for present reactivity bias evaluation and showed a possibility of simplifying the reactivity bias evaluation in burnup credit. Finally, adapting procedures of burnup credit such as the burnup meter were evaluated. (author)

  12. Development of spent fuel remote handling technology - Kinematic analysis of bilateral arms for abnormal spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyu Won; Yoo, Ju Sang; Kim, Jong Yoon [Chungbuk National University, Chongju (Korea)

    2000-03-01

    In the project of 'Development of Spent Fuel Remote Handling Technology', Preprocessing technique, mechanism and teleoperation technique are being developed. One of the mechanisms is a device for disassembling of the spent fuel bundle. However, there may be abnormal fuel bar among the fuel bundle, In this case the unpacking task will be difficult and dangerous. So, in that case, a force reflected teleoperation manipulator is desirable. The system is composed of a anthropomorphic input device at control site, power manipulator at remote site and control system. In this research, the forward and inverse kinematic equations of input device and manipulators has been solved, respectively. In addition, the mapping algorithm is proposed and shown using computer simulation. The reaction force of the telemanipulator with the environmental object is reflected through control system. The reaction force is decomposed into joint torque of the input device based on the jacobian equation. The obtained theoretical relations are verified through computer simulation and they will be used effectively in the spent fuel remote handling technology. 6 refs., 26 figs., 7 tabs. (Author)

  13. Development of advanced spent fuel management process / criticality safety analysis for integrated mockup and metallized spent fuel storage

    International Nuclear Information System (INIS)

    Ro, Seong Gy; Shin, Hee Sung; Shin, Young Joon; Bae, Kang Mok

    1999-02-01

    Benchmark calculation for SCALE4.3 CSAS6 module and burnup credit criticality analysis performed by CSAS6 module are described in this report. Calculation biases by the SCALE4.3 CSAS6 module for PWR spent fuel, metallized spent fuel and aqueous nuclear materials have been determined on the basis of the benchmark to be 0.011, 0.023 and 0.010, respectively. The maximum allowable multiplication factor for an integrated mockup and metallized spent fuel storage is conservatively determined to be 0.927. With the aid of this code system, K eff values as a function of metallization ratio for the integrated mockup have been calculated. The maximum values of K eff for normal and hypothetical accident conditions are 0.346 and 0.598, respectively, much less than the maximum allowable multiplication factor of 0.927. Besides, burnup credit criticality analysis has been performed for infinite arrays of square and hexagonal canisters containing metallized spent fuel rods with different canister wall thickness, canister surface-to-surface distance and water content. It is revealed that the effective multiplication factor for canister arrays as mentioned above is well below the subcritical limit regardless of external conditions when its wall thickness is over 9 mm. (Author). 37 refs., 27 tabs., 64 figs

  14. Commercial Spent Nuclear Fuel Waste Package Misload Analysis

    International Nuclear Information System (INIS)

    J.K. Knudson

    2003-01-01

    The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M and O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis

  15. Commercial Spent Nuclear Fuel Waste Package Misload Analysis

    Energy Technology Data Exchange (ETDEWEB)

    A. Alsaed

    2005-07-28

    the Preclosure Safety Analysis Department. Before using the results of this calculation, the reader is cautioned to verify that the assumptions made in this calculation regarding the waste stream, the loading process, and the staging of the spent nuclear fuel assemblies are applicable.

  16. ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

    International Nuclear Information System (INIS)

    Chang, Sunyoung; Jeong, Yon Hong; Han, Jae-Jun; Lee, Aeri; Hwang, Yong-Soo

    2015-01-01

    The disposal facility of the spent nuclear fuel will be operated from 2051. This paper presents the ENVI code developed by GoldSim Software to simulate options for managing spent nuclear fuel (SNF) in South Korea. The ENVI is a simulator to allow decision-makers to assist to evaluate the performance for spent nuclear fuel management. The multiple options for managing the spent nuclear fuel including the storage and transportation are investigated into interim storage, permanent disposal in geological repositories and overseas and domestic reprocessing. The ENVI code uses the GoldSim software to simulate the logistics of the associated activities. The result by the ENVI model not only produces the total cost to compare among the multiple options but also predict the sizes and timings of different facilities required. In order to decide the policy for spent nuclear management this purpose of this paper is to draw the optimum management plan to solve the nuclear spent fuel issue in the economical aspects. This paper is focused on the development of the ENVI's logic and calculations to simulate four options(No Reprocessing, Overseas Reprocessing, Domestic Reprocessing, and Overseas and Domestic Reprocessing) for managing the spent nuclear fuel in South Korea. The time history of the spent nuclear fuel produced from both the existing and future NPP's can be predicted, based on the Goldsim software made available very user friendly model. The simulation result will be used to suggest the strategic plans for the spent nuclear fuel management

  17. ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Sunyoung; Jeong, Yon Hong; Han, Jae-Jun; Lee, Aeri; Hwang, Yong-Soo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

    2015-10-15

    The disposal facility of the spent nuclear fuel will be operated from 2051. This paper presents the ENVI code developed by GoldSim Software to simulate options for managing spent nuclear fuel (SNF) in South Korea. The ENVI is a simulator to allow decision-makers to assist to evaluate the performance for spent nuclear fuel management. The multiple options for managing the spent nuclear fuel including the storage and transportation are investigated into interim storage, permanent disposal in geological repositories and overseas and domestic reprocessing. The ENVI code uses the GoldSim software to simulate the logistics of the associated activities. The result by the ENVI model not only produces the total cost to compare among the multiple options but also predict the sizes and timings of different facilities required. In order to decide the policy for spent nuclear management this purpose of this paper is to draw the optimum management plan to solve the nuclear spent fuel issue in the economical aspects. This paper is focused on the development of the ENVI's logic and calculations to simulate four options(No Reprocessing, Overseas Reprocessing, Domestic Reprocessing, and Overseas and Domestic Reprocessing) for managing the spent nuclear fuel in South Korea. The time history of the spent nuclear fuel produced from both the existing and future NPP's can be predicted, based on the Goldsim software made available very user friendly model. The simulation result will be used to suggest the strategic plans for the spent nuclear fuel management.

  18. Multi-Detector Analysis System for Spent Nuclear Fuel Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Reber, Edward Lawrence; Aryaeinejad, Rahmat; Cole, Jerald Donald; Drigert, Mark William; Jewell, James Keith; Egger, Ann Elizabeth; Cordes, Gail Adele

    1999-09-01

    The Spent Nuclear Fuel (SNF) Non-Destructive Analysis (NDA) program at INEEL is developing a system to characterize SNF for fissile mass, radiation source term, and fissile isotopic content. The system is based on the integration of the Fission Assay Tomography System (FATS) and the Gamma-Neutron Analysis Technique (GNAT) developed under programs supported by the DOE Office of Non-proliferation and National Security. Both FATS and GNAT were developed as separate systems to provide information on the location of special nuclear material in weapons configuration (FATS role), and to measure isotopic ratios of fissile material to determine if the material was from a weapon (GNAT role). FATS is capable of not only determining the presence and location of fissile material but also the quantity of fissile material present to within 50%. GNAT determines the ratios of the fissile and fissionable material by coincidence methods that allow the two prompt (immediately) produced fission fragments to be identified. Therefore, from the combination of FATS and GNAT, MDAS is able to measure the fissile material, radiation source term, and fissile isotopics content.

  19. Sensitivity analysis of characteristics of spent mixed oxide fuel

    International Nuclear Information System (INIS)

    Hagura, Naoto; Yoshida, Tadashi

    2008-01-01

    Prediction error was evaluated for decay heat and nuclide generation in spent mixed oxide (MOX) fuels on the basis of error files in JENDL-3.3. This computational analysis was performed using SWAT code system, ORIGEN2 code, and ERRORJ code. The results of nuclide generation error evaluation were compared with some discrepancies in the calculated values to experimental values (C/E ratio) which were already published and were obtained by analysis of post irradiated experiments (PIE) data. Though the discrepancies of some C/E values, especially those of americium and curium isotopes, ranged from a half to twice, the present error evaluation based on the error file of nuclide generation became 10% or less. We conclude that the discrepancy between calculation and the PIE data is almost factor 5 larger than that evaluated from the covariance data in JENDL-3.3. Therefore the practical error value of total decay heat should be 20% or more on 1 σ basis. (authors)

  20. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    International Nuclear Information System (INIS)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-01-01

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  1. Three-dimensional thermal analysis of a baseline spent fuel repository

    International Nuclear Information System (INIS)

    Altenbach, T.J.; Lowry, W.E.

    1980-01-01

    A three-dimensional thermal analysis has been performed using finite difference techniques to determine the near-field response of a baseline spent fuel repository in a deep geologic salt medium. A baseline design incorporates previous thermal modeling experience and OWI recommendations for areal thermal loading in specifying the waste form properties, package details, and emplacement configuration. The base case in this thermal analysis considers one 10-year old PWR spent fuel assembly emplaced to yield a 36 kw/acre (8.9 w/m 2 ) loading. A unit cell model in an infinite array is used to simplify the problem and provide upper-bound temperatures. Boundary conditions are imposed which allow simulations to 1000 years. Variations studied include a comparison of ventilated and unventilated storage room conditions, emplacement packages with and without air gaps surrounding the canister, and room cool-down scenarios with ventilation following an unventilated state for retrieval purposes. At this low power level ventilating the emplacement room has an immediate cooling influence on the canister and effectively maintains the emplacement room floor near the temperature of the ventilating air. The annular gap separating the canister and sleeve causes the peak temperature of the canister surface to rise by 10 0 F (5.6 0 C) over that from a no gap case assuming perfect thermal contact. It was also shown that the time required for the emplacement room to cool down to 100 0 F (38 0 C) from an unventilated state ranged from 2 weeks to 6 months; when ventilation initiated after times of 5 years to 50 years, respectively. As the work was performed for the Nuclear Regulatory Commission, these results provide a significant addition to the regulatory data base for spent fuel performance in a geologic repository

  2. An economic analysis of spent fuel management and storage

    International Nuclear Information System (INIS)

    Nagano, Koji

    1998-01-01

    Spent fuel management is becoming a key issue not only in the countries that have already experienced years of nuclear operation but also in the Asian countries that started nuclear utilization rather lately. This paper summarizes the key aspects that essentially determine optimal conditions for desired spent fuel management strategies from the engineering-economic point of view, in both national and regional perspectives. The term 'desired' is intended to highlight positive and beneficial aspects of such strategies, namely mobile and timely exploitation of spent fuel storage. Among all, the economy of scale, the economy of scope, the learning-by-doing effect, and benefits of R and D are reviewed theoretically and empirically, and the paper overviews to what extent these factors are implemented in solving spent fuel management strategy optimization problem. (author)

  3. Spent fuel pool accident analysis and accident management

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gil; Cho, Cheon Hwey [ACT CO., Daejeon (Korea, Republic of); Lee, Jae Young; Sung, Joon Young; Maeng, Yun Hwan [Handong Global University, Pohang (Korea, Republic of); Jerng, Dong Wook [Chung-Ang University, Seoul (Korea, Republic of)

    2016-05-15

    The spent fuel pool(SFP) in unit 4 of the Fukushima Daiichi NPPs was damaged by an extreme seismic event and subsequent flooding by a tsunami. In order to investigate a progression of spent fuel pool accident scenarios, the well-defined MELCOR 1.8.6 code input deck was prepared and validated by experimental data of the OECD/NEA Sandia Fuel Project. Based on the validated MELCOR code input, three types of spent fuel pool accident scenarios were analyzed. In the complete loss of coolant accident (LOCA) scenarios, sensitivity studies were conducted to identify the modeling boundary conditions to initiate a zirconium fire in the spent fuel assemblies. A series of MELCOR code calculations were performed to investigate a consequence of each SFP accident scenario. Based on findings from the calculations, the recommended operator actions were proposed to manage the SFP accident progressions.

  4. Seismic analysis of submerged spent fuel storage structure

    International Nuclear Information System (INIS)

    Chang, S.J.

    1994-01-01

    The purpose of this calculation is to provide structural integrity analysis for the loaded new spent fuel rack arrays against possible seismic excitation. The seismic design calculation is based on the UCRL-15910 spectrum with peak ground acceleration of 0.32g and 5% damping. This spectrum may be considered as an upper bound of the newly developed Oak Ridge site-specific spectrum with 0.29g peak ground acceleration and 5% damping. Both are more conservative than the current design basis seismic acceleration of 0.15g for HFIR. The calculation is carried out by using ABAQUS version 5.2 and the response spectrum option. Since the new racks are to be submerged in HFIR pool, the pool water induced virtual mass has been conservatively taken into consideration. The result shows that if the silo buckling is regarded as failure than, with 95% confidence, the 5% probability of failure ground acceleration is as much as 2.334g. As compared with the design basis of 0.32g, the structure is very safe against earthquake

  5. Time Spent on Social Network Sites and Psychological Well-Being: A Meta-Analysis.

    Science.gov (United States)

    Huang, Chiungjung

    2017-06-01

    This meta-analysis examines the relationship between time spent on social networking sites and psychological well-being factors, namely self-esteem, life satisfaction, loneliness, and depression. Sixty-one studies consisting of 67 independent samples involving 19,652 participants were identified. The mean correlation between time spent on social networking sites and psychological well-being was low at r = -0.07. The correlations between time spent on social networking sites and positive indicators (self-esteem and life satisfaction) were close to 0, whereas those between time spent on social networking sites and negative indicators (depression and loneliness) were weak. The effects of publication outlet, site on which users spent time, scale of time spent, and participant age and gender were not significant. As most included studies used student samples, future research should be conducted to examine this relationship for adults.

  6. Safety Analysis of Spent Nuclear Fuel and Radwaste Facilities

    International Nuclear Information System (INIS)

    Poskas, P.; Ragaisis, V.

    2001-01-01

    The overview of the activities in the Laboratory of Heat Transfer in Nuclear Reactors related with the assessment of thermal, neutronic and radiation characteristics in spent nuclear fuel and radwaste facilities are performed. Activities related with decommissioning of Ignalina NPP are also reviewed. (author)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  8. VHTR, ADS, and PWR spent nuclear fuel analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salome, J.A.D.; Cardoso, F.; Velasquez, C.E.; Pereira, F.; Pereira, C. [Departamento de Engenharia Nuclear - Escola de Engenharia Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Pampulha, Belo Horizonte MG, CEP: 31270-901 (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores - CNPq, Rio de Janeiro (Brazil); Barros, G.P. [Comissao Nacional de Energia Nuclear - CNEN, Rua General Severiano 82, Botafogo, Rio de Janeiro, RJ, CEP: 22290-040 (Brazil)

    2016-07-01

    The aim of this study is to analyze and compare the discharged-spent fuel of 3 types of nuclear systems: a Very High-Temperature Gas Reactor (VHTR), a lead-cooled Accelerator-Driven System (ADS) and a standard Pressurized Water Reactor (PWR). The two first systems, VHTR, and ADS were designed to use reprocessed fuels. UREX+ and GANEX techniques were used for the reprocessing processes respectively. The fuel burnup simulated for the systems in other works have been used to obtain the final composition of the spent fuel discharged. After discharge, the radioactivity, the radiotoxicity, and the decay heat were evaluated through the ORIGEN 2.1 code until 10{sup 7} years and compared to the literature. The spent nuclear waste (SNF) coming from reprocessing techniques and burned up in advanced reactors show that the radiotoxicity decreases below a conventional SNF from a typical PWR for the time studied. The VHTR and ADs have higher values of radioactivity, radiotoxicity and decay heat, because of the greater concentrations of plutonium and curium in these reactors than in the PWR. Fission products have the greatest contribution for the first 25 years over the parameters studied for a PWR. The most harmful fission products are: Ba{sup 137m}, Tc{sup 99}, I{sup 129} and Nb{sup 93m} and for actinides is the plutonium and curium.

  9. Enhancement in lipid content of Chlorella sp. MJ 11/11 from the spent medium of thermophilic biohydrogen production process.

    Science.gov (United States)

    Ghosh, Supratim; Roy, Shantonu; Das, Debabrata

    2017-01-01

    The present study investigates the effect of spent media of acetogenic dark fermentation for mixotrophic algal cultivation for biodiesel production. Mixotrophic growth conditions were optimized in culture flask (250mL) using Chlorella sp. MJ 11/11. Maximum lipid accumulation (58% w/w) was observed under light intensity, pH, nitrate and phosphate concentration of 100μmolm -2 s -1 , 7, 2.7mM and 1.8mM, respectively. Air lift (1.4L) and flat panel (1.4L) reactors were considered for algal cultivation. Air lift showed significant improvement in biomass and lipid production as compared to flat panel reactor. The results could help in development of sustainable technology involving acetogenic hydrogen production integrated with sequential mitigation of spent media by algal cultivation for improved energy recovery. Copyright © 2016 Elsevier Ltd. All rights reserved.

  10. A spent fuel assemblies monitoring device by nondestructive analysis 'PYTHON'

    International Nuclear Information System (INIS)

    Saad, M.; Broeskamp, M.; Hahn, H.; Bignan, G.; Boisset, M.; Silie, P.

    1995-01-01

    The monitoring of spent fuel assemblies (16 x 16 UOX) in KWG-reactor pool with the use of non-destructive methods (total Gamma and neutron counting) allow the control of average burn-up and the extremity burn-up. The measurements allow a safety-criticality control before loading the fuel assemblies into the transport casks. A device called PYTHON has been tested and qualified in France. This paper presents a description of the industrial PYTHON device and the results of the measurements. (orig.)

  11. English Medium of Instruction: A Situation Analysis

    Science.gov (United States)

    Uys, Mandie; van der Walt, Johann; van den Berg, Ria; Botha, Sue

    2007-01-01

    The majority of learners in southern Africa receive their education through the medium of a second language, English. Although teachers of English play a crucial role in helping learners to acquire language skills in the medium of instruction, we argue that subject content teachers' lack of attention to the teaching of the four language skills may…

  12. Impact analysis of spent nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Huerta, M.; Dennis, A.W.; Yoshimura, R.H.

    1978-07-01

    A full-scale testing program involving impact tests of spent-nuclear-fuel shipping systems is described. The program is being conducted by Sandia Laboratories for the Environmental Control Technology Division of the U.S. Energy Research and Development Administration. The paper describes the analytical and scale modeling techniques being employed to predict the response of the full-scale system in the very severe impact tests. The analytical techniques include lumped parameter modeling of the vehicle and cask system and finite modeling of isolated shipping casks. Some preliminary results from the mathematical analyses and scale model tests demonstrate close agreement between these two techniques. Scale models of the systems are also described and some results presented

  13. Brewer's spent grain and corn steep liquor as alternative culture medium substrates for proteinase production by Streptomyces malaysiensis AMT-3.

    Science.gov (United States)

    do Nascimento, Rodrigo Pires; Junior, Nelson Alves; Coelho, Rosalie Reed Rodrigues

    2011-10-01

    Brewer's spent grain and corn steep liquor or yeast extract were used as the sole organic forms for proteinase production by Streptomyces malaysiensis in submerged fermentation. The influence of the C and N concentrations, as well as the incubation periods, were assessed. Eight proteolytic bands were detected through gelatin-gel-electrophoresis in the various extracts obtained from the different media and after different incubation periods, with apparent molecular masses of 20, 35, 43, 50, 70, 100, 116 and 212 kDa. The results obtained suggest an opportunity for exploring this alternative strategy for proteinases production by actinomycetes, using BSG and CSL as economically feasible substrates.

  14. Brewer's spent grain and corn steep liquor as alternative culture medium substrates for proteinase production by Streptomyces malaysiensis AMT-3

    Directory of Open Access Journals (Sweden)

    Rodrigo Pires do Nascimento

    2011-12-01

    Full Text Available Brewer's spent grain and corn steep liquor or yeast extract were used as the sole organic forms for proteinase production by Streptomyces malaysiensis in submerged fermentation. The influence of the C and N concentrations, as well as the incubation periods, were assessed. Eight proteolytic bands were detected through gelatin-gel-electrophoresis in the various extracts obtained from the different media and after different incubation periods, with apparent molecular masses of 20, 35, 43, 50, 70, 100, 116 and 212 kDa. The results obtained suggest an opportunity for exploring this alternative strategy for proteinases production by actinomycetes, using BSG and CSL as economically feasible substrates.

  15. Feasibility and incentives for the consideration of spent fuel operating histories in the criticality analysis of spent fuel shipping casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Westfall, R.M.; Jones, R.H.

    1987-08-01

    Analyses have been completed that indicate the consideration of spent fuel histories (''burnup credit'') in the design of spent fuel shipping casks is a justifiable concept that would result in cost savings and public risk benefits in the transport of spent nuclear fuel. Since cask capacities could be increased over those of casks without burnup credit, the number of shipments necessary to transport a given amount of fuel could be reduced. Reducing the number of shipments would increase safety benefits by reducing public and occupational exposure to both radiological and nonradiological risks associated with the transport of spent fuel. Economic benefits would include lower in-transit shipping, reduced transportation fleet capital costs, and reduced numbers of cask handling operations at both shipping and receiving facilities. 44 refs., 66 figs., 28 tabs

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  17. Risk analysis methodology designed for small and medium enterprises

    OpenAIRE

    Ladislav Beránek; Radim Remeš

    2009-01-01

    The aim of this paper is to present risk analysis procedures successfully applied by several Czech small and medium enterprises. The paper presents in detail the individual steps we use in risk analysis of small and medium enterprises in the Czech Republic. Suggested method to risk analysis is based on the modification of the FRAP methodology and the BITS recommendation. Modifications of both methodologies are described in detail. We propose modified risk analysis methodology which is quick a...

  18. Systems analysis of spent fuel management in Japan. (2). Methodologies for economic analyses of spent fuel storage

    International Nuclear Information System (INIS)

    Nagano, Koji

    2003-01-01

    Analytical methods for economic analyses of spent fuel storage are categorized in three layers; (1) static engineering-economic cost estimates, (2) dynamic strategy analyses, and (3) specific project financing assessments. After describing recent legal and institutional evolution in spent fuel management and storage in Japan, the report summarizes each of the three methods with numerical examples of applications. As a conclusion, the author maintains that users should choose the most suitable type of method or calculating tool in accordance with their specific purposes. General guidelines of choosing methodologies are elaborated as the conclusion. (author)

  19. Criticality safety analysis of spent fuel storage for NPP Mochovce using MCNP5

    International Nuclear Information System (INIS)

    Farkas, G.; Hascik, J.; Lueley, J.; Vrban, B.; Petriska, M.; Slugen, V.; Urban, P.

    2011-01-01

    The paper presents results of nuclear criticality safety analysis of spent fuel storage for the first and second unit of NPP Mochovce. The spent fuel storage pool (compact and reserve grid) was modeled using the Monte Carlo code MCNP5. Conservative approach was applied and calculation of k eff values was performed for normal and various postulated emergency conditions in order to evaluate the final maximal k eff values. The requirement of current safety regulations to ensure 5% subcriticality was met except one especially conservative case. (Authors)

  20. Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term-Disposal Criticality Safety

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1999-01-01

    Utilization of burnup credit in criticality safety analysis for long-term disposal of spent nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile material that will be present in the repository. Burnup-credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents (in terms of criticality potential), followed by criticality calculations to assess the value of the effective neutron multiplication factor (k(sub)eff) for the a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models

  1. Analysis of spent fuel assay with a lead slowing down spectrometer

    International Nuclear Information System (INIS)

    Gavron, A.; Smith, L. Eric; Ressler, Jennifer J.

    2009-01-01

    Assay of fissile materials in spent fuel that are produced or depleted during the operation of a reactor, is of paramount importance to nuclear materials accounting, verification of the reactor operation history, as well as for criticality considerations for storage. In order to prevent future proliferation following the spread of nuclear energy, we must develop accurate methods to assay large quantities of nuclear fuels. We analyze the potential of using a Lead Slowing Down Spectrometer for assaying spent fuel. We conclude that it possible to design a system that will provide around 1% statistical precision in the determination of the 239 Pu, 241 Pu and 235 U concentrations in a PWR spent-fuel assembly, for intermediate-to-high burnup levels, using commercial neutron sources, and a system of 238 U threshold fission detectors. Pending further analysis of systematic errors, it is possible that missing pins can be detected, as can asymmetry in the fuel bundle. (author)

  2. K Basins Spent Nuclear Fuel (SNF) Project Safety Analysis Report for Packaging (SARP) approval plan

    International Nuclear Information System (INIS)

    1995-01-01

    This document delineates the plan for preparation, review, and approval of the K Basins Spent Nuclear Fuel (SNF) Packaging Design Criteria (PDC) document and the on-site Safety Analysis Report for Packaging (SARP). The packaging addressed in these documents is used to transport SNF in a Multi- canister Overpack (MCO) configuration

  3. Hazards Analysis for the Spent Nuclear Fuel L-Experimental Facility

    International Nuclear Information System (INIS)

    Blanchard, A.

    1999-01-01

    The purpose of this Hazard Analysis (HA) is to identify and assess potential hazards associated with the operations of the Spent Nuclear Fuels (SNF) Treatment and Storage Facility LEF. Additionally, this HA will be used for identifying and assessing potential hazards and specifying functional attributes of SSCs for the LEF project

  4. Seismic hazard analysis for the NTS spent reactor fuel test site

    International Nuclear Information System (INIS)

    Campbell, K.W.

    1980-01-01

    An experiment is being directed at the Nevada Test Site to test the feasibility for storage of spent fuel from nuclear reactors in geologic media. As part of this project, an analysis of the earthquake hazard was prepared. This report presents the results of this seismic hazard assessment. Two distinct components of the seismic hazard were addressed: vibratory ground motion and surface displacement

  5. A simplified computational scheme for thermal analysis of LWR spent fuel dry storage and transportation casks

    International Nuclear Information System (INIS)

    Kim, Chang Hyun

    1997-02-01

    A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. Although a lot of computer codes and analytical models have been developed for application to the fields of thermal analysis of dry storage and/or transportation casks, some difficulties in its analysis arise from the complexity of the geometry including the rod bundles of spent fuel and the heat transfer phenomena in the cavity of cask. Particularly, if the disk-type structures such as fuel baskets and aluminium heat transfer fins are included, the thermal analysis problems in the cavity are very complex. To overcome these difficulties, cylindrical coordinate system is adopted to calculate the temperature profile of a cylindrical cask body using the multiple cylinder model as the step-1 analysis of the present study. In the step-2 analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three- dimensional conduction analysis model. The effective thermal conductivity for homogenized spent fuel assembly based on Manteufel and Todreas model is incorporated in step-2 analysis to predict the maximum fuel temperature. The presented two-step computational scheme has been performed using an existing HEATING 7.2 code and the effective thermal conductivity for the homogenized spent fuel assembly has been calculated by additional numerical analyses. Sample analyses of five cases are performed for NAC-STC including normal transportation condition to examine the applicability of the presented simplified computational scheme for thermal analysis of the large LWR spent fuel dry storage and transportation casks and heat transfer characteristics in the cavity of the cask with the disk-type structures

  6. Analysis of the risk of transporting spent nuclear fuel by train

    Energy Technology Data Exchange (ETDEWEB)

    Elder, H.K.

    1981-09-01

    This report uses risk analyses to analyze the safety of transporting spent nuclear fuel for commercial rail shipping systems. The rail systems analyzed are those expected to be used in the United States when the total electricity-generating capacity by nuclear reactors is 100 GW in the late 1980s. Risk as used in this report is the product of the probability of a release of material to the environment and the consequences resulting from the release. The analysis includes risks in terms of expected fatalities from release of radioactive materials due to transportation accidents involving PWR spent fuel shipped in rail casks. The expected total risk from such shipments is 1.3 x 10/sup -4/ fatalities per year. Risk spectrums are developed for shipments of spent fuel that are 180 days and 4 years out-of-reactor. The risk from transporting spent fuel by train is much less (by 2 to 4 orders of magnitude) than the risk to society from other man-caused events such as dam failure.

  7. Two-dimensional radiation shielding optimization analysis of spent fuel transport container

    International Nuclear Information System (INIS)

    Tian Yingnan; Chen Yixue; Yang Shouhai

    2013-01-01

    The intelligent radiation shielding optimization design software platform is a one-dimensional multi-target radiation shielding optimization program which is developed on the basis of the genetic algorithm program and one-dimensional discrete ordinate program-ANISN. This program was applied in the optimization design analysis of the spent fuel transport container radiation shielding. The multi-objective optimization calculation model of the spent fuel transport container radiation shielding was established, and the optimization calculation of the spent fuel transport container weight and radiation dose rate was carried by this program. The calculation results were checked by Monte-Carlo program-MCNP/4C. The results show that the weight of the optimized spent fuel transport container decreases to 81.1% of the origin and the radiation dose rate decreases to below 65.4% of the origin. The maximum deviation between the calculated values from the program and the MCNP is below 5%. The results show that the optimization design scheme is feasible and the calculation result is correct. (authors)

  8. Environmental analysis in small and medium enterprises

    International Nuclear Information System (INIS)

    Luciani, R.; Andriola, L.; Di Franco, N.

    2001-01-01

    An environmental analysis is considered to be one of the primary goals for an enterprise environmental management. Nevertheless the complexity of the environmental problems and of its regulations prevents the small enterprises the possibility to perform an environmental policy (EMAS, ISO 14001). One of the most correct evaluation instrument for creating a datum-point standard could be the filling up a questionnaire, built up in according to the industrial enterprises need. It has the function of creating a primary step for a subsequent environmental management [it

  9. Impact response analysis of cask for spent fuel by dimensional analysis and mode superposition method

    International Nuclear Information System (INIS)

    Kim, Y. J.; Kim, W. T.; Lee, Y. S.

    2006-01-01

    Full text: Full text: Due to the potentiality of accidents, the transportation safety of radioactive material has become extremely important in these days. The most important means of accomplishing the safety in transportation for radioactive material is the integrity of cask. The cask for spent fuel consists of a cask body and two impact limiters generally. The impact limiters are attached at the upper and the lower of the cask body. The cask comprises general requirements and test requirements for normal transport conditions and hypothetical accident conditions in accordance with IAEA regulations. Among the test requirements for hypothetical accident conditions, the 9 m drop test of dropping the cask from 9 m height to unyielding surface to get maximum damage becomes very important requirement because it can affect the structural soundness of the cask. So far the impact response analysis for 9 m drop test has been obtained by finite element method with complex computational procedure. In this study, the empirical equations of the impact forces for 9 m drop test are formulated by dimensional analysis. And then using the empirical equations the characteristics of material used for impact limiters are analysed. Also the dynamic impact response of the cask body is analysed using the mode superposition method and the analysis method is proposed. The results are also validated by comparing with previous experimental results and finite element analysis results. The present method is simpler than finite element method and can be used to predict the impact response of the cask

  10. Fabrication of magnetic biochar as a treatment medium for As(V) via pyrolysis of FeCl3-pretreated spent coffee ground.

    Science.gov (United States)

    Cho, Dong-Wan; Yoon, Kwangsuk; Kwon, Eilhann E; Biswas, Jayanta Kumar; Song, Hocheol

    2017-10-01

    This study investigated the preparation of magnetic biochar from N 2 - and CO 2 -assisted pyrolysis of spent coffee ground (SCG) for use as an adsorption medium for As(V), and the effects of FeCl 3 pretreatment of SCG on the material properties and adsorption capability of the produced biochar. Pyrolysis of FeCl 3 -pretreated SCG in CO 2 atmosphere produced highly porous biochar with its surface area ∼70 times greater than that produced in N 2 condition. However, despite the small surface area, biochar produced in N 2 showed greater As(V) adsorption capability. X-ray diffraction and X-ray photoelectron spectrometer analyses identified Fe 3 C and Fe 3 O 4 as dominant mineral phases in N 2 and CO 2 conditions, with the former being much more adsorptive toward As(V). The overall results suggest functional biochar can be facilely fabricated by necessary pretreatment to expand the applicability of biochar for specific purposes. Copyright © 2017 Elsevier Ltd. All rights reserved.

  11. Further Evaluation of the Neutron Resonance Transmission Analysis (NRTA) Technique for Assaying Plutonium in Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    J. W. Sterbentz; D. L. Chichester

    2011-09-01

    This is an end-of-year report (Fiscal Year (FY) 2011) for the second year of effort on a project funded by the National Nuclear Security Administration's Office of Nuclear Safeguards (NA-241). The goal of this project is to investigate the feasibility of using Neutron Resonance Transmission Analysis (NRTA) to assay plutonium in commercial light-water-reactor spent fuel. This project is part of a larger research effort within the Next-Generation Safeguards Initiative (NGSI) to evaluate methods for assaying plutonium in spent fuel, the Plutonium Assay Challenge. The second-year goals for this project included: (1) assessing the neutron source strength needed for the NRTA technique, (2) estimating count times, (3) assessing the effect of temperature on the transmitted signal, (4) estimating plutonium content in a spent fuel assembly, (5) providing a preliminary assessment of the neutron detectors, and (6) documenting this work in an end of the year report (this report). Research teams at Los Alamos National Laboratory (LANL), Lawrence Berkeley National Laboratory (LBNL), Pacific Northwest National Laboratory (PNNL), and at several universities are also working to investigate plutonium assay methods for spent-fuel safeguards. While the NRTA technique is well proven in the scientific literature for assaying individual spent fuel pins, it is a newcomer to the current NGSI efforts studying Pu assay method techniques having just started in March 2010; several analytical techniques have been under investigation within this program for two to three years or more. This report summarizes work performed over a nine month period from January-September 2011 and is to be considered a follow-on or add-on report to our previous published summary report from December 2010 (INL/EXT-10-20620).

  12. Thermal safety analysis of a dry storage cask for the Korean standard spent fuel - 16159

    International Nuclear Information System (INIS)

    Cha, Jeonghun; Kim, S.N.; Choi, K.W.

    2009-01-01

    A conceptual dry storage facility, which is based on a commercial dry storage facility, was designed for the Korea standard spent nuclear fuel (SNF) and preliminary thermal safety analysis was performed in this study. To perform the preliminary thermal analysis, a thermal analysis method was proposed. The thermal analysis method consists of 2 parts. By using the method, the surface temperature of the storage canister corresponding to the SNF clad temperature was calculated and the adequate air duct area was decided using the calculation result. The initial temperature of the facility was calculated and the fire condition and half air duct blockage were analyzed. (authors)

  13. Using New Fission Data with the Multi-detector Analysis System for Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    A. V. Ramayya; A.V. Daniel (Joint Institute for Nuclear Research); C. J. Beyer (Vanderbilt Univ.); E. L. Reber; G. M. Ter-Akopian; G.S. Popeko; J. D. Cole; J. H. Hamilton; J. K. Jewell (INEEL); M. W. Drigert; R. Aryaeinejad; Ts.Yu. Oganessian

    1998-11-01

    New experiments using an array of high purity germanium detectors and fast liquid scintillation detectors has been performed to observe the radiation emitted from the induced fission of 235U with a beam of thermal neutrons. The experiment was performed at the Argonne National Laboratory Intense Pulsed Neutron Source. Preliminary observations of the data are presented. A nondestructive analysis system for the characterization of DOE spent nuclear fuel based on these new data is presented.

  14. Using New Fission Data with the Multi-detector Analysis System for Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cole, Jerald Donald

    1998-11-01

    New experiments using an array of high purity germanium detectors and fast liquid scintillation detectors has been performed to observe the radiation emitted from the induced fission of 235U with a beam of thermal neutrons. The experiment was performed at the Argonne National Laboratory Intense Pulsed Neutron Source. Preliminary observations of the data are presented. A nondestructive analysis system for the characterization of DOE spent nuclear fuel based on these new data is presented.

  15. Use of Grass and Spent Mushroom Compost as a Growing Medium of Local Tomato (Lycopersicon Esculentum Miller) Seedling in the Nursery

    OpenAIRE

    Priadi, Dody; Arfani, Agus; Saskiawan, Iwan; Mulyaningsih, Enung Sri

    2016-01-01

    The objective of this study was to investigate the response of local tomato (Lycopersicon esculentum Miller) seedlings growth on media containing grass and spent mushroom compost in the nursery. The grass compost (GC) was produced by Research Center for Biotechnology-LIPI. Whereas the spent oyster mushroom (Pleurotus ostreatus) compost (OC) and spent paddy straw mushroom (Volvariella volvacea) compost (PC) were produced by Research Center for Biology-LIPI. Growing media of tomato seedling was...

  16. Analysis of burnup credit on spent fuel transport / storage casks - estimation of reactivity bias

    International Nuclear Information System (INIS)

    Mat sumura, T.; Sasahara, A.; Takei, M.; Takekawa, T.; Kagehira, K.; Nicolaou, G.; Betti, M.

    1998-01-01

    Chemical analyses of high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins were carried out. Measured data of nuclides' composition from U234 to P 242 were used for evaluation of ORIGEN-2/82 code and a nuclear fuel design code (NULIF). Critically calculations were executed for transport and storage casks for 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for axial and horizontal profiles of burnup, and historical void fraction (BWR), operational histories such as control rod insertion history, BPR insertion history and others, and calculational accuracy of ORIGEN-2/82 on nuclides' composition. This study shows that introduction of burnup credit has a large merit in criticality safety analysis of casks, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for the present reactivity bias evaluation and showed the possibility of simplifying the reactivity bias evaluation in burnup credit. (authors)

  17. Thermal analysis model for the temperature distribution of the CANDU spent fuel assembly

    International Nuclear Information System (INIS)

    Choi, Hae Yun; Kwon, Jong Soo; Park, Seong Hoon; Kim, Seong Rea; Lee, Gi Won

    1996-01-01

    The purpose of this technical is to introduce the methodology and experimental process for the experimental research work with the mock-up test performed to verify and validate the MAXROT code which is a thermal analysis method for Wolsong (CANDU) spent fuel dry storage canister. The experiment was conducted simulating the heat transfer characteristics of combinations of equilateral triangular and square pitch arrays of heater rods, similar to a CANDU spent fuel bundle. After assembly of the heater rod bundle into the containment vessel, the experimental apparatus was operated under the same operating and boundary conditions as an interim dry storage condition at the nuclear power plant site. The reduced data from this experiment has been utilized to verity a model developed to predict the maximum fuel rod surface temperature in a fuel bundle. These test procedures and the experiment can be utilized to establish the fine thermal analysis method applicable to dry storage system for the spent fuel. 12 figs., 5 tabs., 36 refs. (Author) .new

  18. An analysis of the properties of levelized cost analysis of storage or recycling of spent nuclear fuel

    International Nuclear Information System (INIS)

    Vergueiro, Sophia M. C.; Ramos, Alexandre F.

    2017-01-01

    The demand for reduction of carbon dioxide emissions in the processes of electricity generation, plus the demand for firm energy matrices, make the nuclear matrix a central component to occupy the energy mix during the next hundred years. Increasing the share of nuclear power in electricity production in a multiple developing countries will lead to increased spent fuel production. Thus, the managing radioactive waste aiming to decide about storing or recycling it is a central issue to be addressed by environmental management and nuclear energy communities. In this manuscript we present our studies aiming to understand the levelized analysis of cost of electricity generation comparing storage or recycling of the spent fuel. (author)

  19. Analysis of gamma irradiator dose rate using spent fuel elements with parallel configuration

    International Nuclear Information System (INIS)

    Setiyanto; Pudjijanto MS; Ardani

    2006-01-01

    To enhance the utilization of the RSG-GAS reactor spent fuel, the gamma irradiator using spent fuel elements as a gamma source is a suitable choice. This irradiator can be used for food sterilization and preservation. The first step before realization, it is necessary to determine the gamma dose rate theoretically. The assessment was realized for parallel configuration fuel elements with the irradiation space can be placed between fuel element series. This analysis of parallel model was choice to compare with the circle model and as long as possible to get more space for irradiation and to do manipulation of irradiation target. Dose rate calculation were done with MCNP, while the estimation of gamma activities of fuel element was realized by OREGEN code with 1 year of average delay time. The calculation result show that the gamma dose rate of parallel model decreased up to 50% relatively compared with the circle model, but the value still enough for sterilization and preservation. Especially for food preservation, this parallel model give more flexible, while the gamma dose rate can be adjusted to the irradiation needed. The conclusion of this assessment showed that the utilization of reactor spent fuels for gamma irradiator with parallel model give more advantage the circle model. (author)

  20. Global Spent Fuel Logistics Systems Study (GSFLS). Volume 3A. GSFLS technical analysis (appendix). Interim report

    International Nuclear Information System (INIS)

    1978-01-01

    This report is a part of the interim report documentation for the Global Spent Fuel Logistics System (GSFLS) study. The technical and financial considerations underlying a global spent fuel logistics systems have been studied and are reported. The Pacific Basin is used as a model throughout this report; however the stated methodology and, in many cases, considerations and conclusions are applicable to other global regions. Spent fuel discharge profiles for Pacific Basin Countries were used to determine the technical systems requirements for alternative concepts. Functional analyses and flows were generated to define both system design requirements and logistics parameters. A technology review was made to ascertain the state-of-the-art of relevant GSFLS technical systems. Modular GSFLS facility designs were developed using the information generated from the functional analysis and technology review. The modular facility designs were used as a basis for siting and cost estimates for various GSFLS alternatives. Various GSFLS concepts were analyzed from a financial and economic perspective in order to provide total concepts costs and ascertain financial and economic sensitivities to key GSFLS variations. Results of the study include quantification of GSFLS facility and hardware requirements; drawings of relevant GSFLS facility designs; system cost estimates; financial reports - including user service charges; and comparative analyses of various GSFLS alternatives

  1. Spent fuel acceptance scenarios devoted to shutdown reactors: A preliminary analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wood, T.W.; Plummer, A.M.; Dippold, D.G.; Short, S.M. (Pacific Northwest Lab., Richland, WA (USA); Battelle Memorial Inst., Columbus, OH (USA). Office of Transportation Systems and Planning; Pacific Northwest Lab., Richland, WA (USA))

    1989-10-01

    Spent fuel acceptance schedules and the allocation of federal acceptance capacity among commercial nuclear power reactors have important operational and cost consequences for reactor operators. Alternative allocation schemes were investigated to some extent in DOE's MRS Systems Study. The current study supplements these analyses for a class of acceptance schemes in which the acceptance capacity of the federal radioactive waste management system is allocated principally to shutdown commercial power reactors, and extends the scope of analysis to include considerations of at-reactor cask loading rates. The operational consequences of these schemes for power reactors, as measured in terms of quantity of spent fuel storage requirement above storage pool capacities and number of years of pool operations after last discharge, are estimated, as are the associated utility costs. This study does not attempt to examine the inter-utility equity considerations involved in departures from the current oldest-fuel-first (OFF) allocation rule as specified in the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste.'' In the sense that the alternative allocations are more economically efficient than OFF, however, they approximate the allocations that could result from free exchange of acceptance rights among utilities. Such a process would result in the preservation of inter-utility equity. 13 refs., 9 figs., 9 tabs.

  2. Cna 1 spent fuel element interim dry storage system thermal analysis

    International Nuclear Information System (INIS)

    Hilal, R. E; Garcia, J. C; Delmastro, D. F

    2006-01-01

    At the moment, the Atucha I Nuclear Power Plant (Cnea-I) located in the city of Lima, has enough room to store its spent fuel (Sf) in their two pools spent fuel until about 2015.In case of life extension a spend fuel element interim dry storage system is needed.Nucleolectrica Argentina S.A. (N A-S A) and the Comision Nacional de Energia Atomica (Cnea), have proposed different interim dry storage systems.These systems have to be evaluated in order to choose one of them.The present work's objective is the thermal analysis of one dry storage alternative for the Sf element of Cna 1.In this work a simple model was developed and used to perform the thermal calculations corresponding to the system proposed by Cnea.This system considers the store of sealed containers with 37 spent fuels in concrete modules.Each one of the containers is filled in the pool houses and transported to the module in a transference cask with lead walls.Fulfill the maximum cladding temperature requirement ( [es

  3. The feasibility of modelling coupled processes in safety analysis of spent nuclear fuel disposal

    Energy Technology Data Exchange (ETDEWEB)

    Rasilainen, K. [VTT Energy, Espoo (Finland); Luukkonen, A.; Niemi, A.; Poellae, J. [VTT Communities and Infrastructure, Espoo (Finland); Olin, M. [VTT Chemical Technology, Espoo (Finland)

    1999-07-01

    The potential of applying coupled modelling in the Finnish safety analysis programme has been reviewed. The study focused on the migration of radionuclides escaping from a spent fuel repository planned to be excavated in fractured bedrock. Two effects that can trigger various couplings in and around a spent fuel repository in Finland were studied in detail; namely heat generation in the spent fuel and the presence of deep, saline groundwaters. The latter have been observed in coastal areas. A systematic survey of the requirements of coupled modelling identified features that render such migration calculations a challenging task. In groundwater flow modelling there appears to be wide ranging uncertainty related to conceptualisation of flow systems and to the corresponding input data. In terms of migration related chemistry there appear to be large gaps in the underlying thermodynamic database for geochemical systems. Rock mechanical predictions are heavily dependent on knowing the location, structure and properties of dominant fractures; information which is extremely difficult to obtain. Conduction and convection of heat is understood well in principle. On the basis of this review, it appears that coupled migration modelling may not yet be at the stage of development that would allow its use as a standard modelling tool in performance assessments. However, a firmer basis for the conclusions reached can only be obtained after a systematic modelling exercise on a relevant and real migration problem has been carried out. (orig.)

  4. Thermal analysis on NAC-STC spent fuel transport cask under different transport conditions

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Yumei [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Yang, Jian, E-mail: zdhjkz@zju.edu.cn [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Xu, Chao; Wang, Weiping [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Ma, Zhijun [Department of Material Engineering, South China University of Technology, Guangzhou (China)

    2013-12-15

    Highlights: • Spent fuel cask was investigated as a whole instead of fuel assembly alone. • The cask was successfully modeled and meshed after several simplifications. • Equivalence method was used to calculate the properties of parts. • Both the integral thermal field and peak values are captured to verify safety. • The temperature variations of key parts were also plotted. - Abstract: Transport casks used for conveying spent nuclear fuel are inseparably related to the safety of the whole reprocessing system for spent nuclear fuel. Thus they must be designed according to rigorous safety standards including thermal analysis. In this paper, for NAC-STC cask, a finite element model is established based on some proper simplifications on configurations and the heat transfer mechanisms. Considering the complex components and gaps, the equivalence method is presented to define their material properties. Then an equivalent convection coefficient is introduced to define boundary conditions. Finally, the temperature field is captured and analyzed under both normal and accident transport conditions by using ANSYS software. The validity of numerical calculation is given by comparing its results with theoretical calculation. Obtaining the integral distribution laws of temperature and peak temperature values of all vital components, the security of the cask can be evaluated and verified.

  5. Examination of the Properties of a Spent Fuel based Electricity Generation System - Scintillator Performance Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Haneol; Yim, Man-Sung [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    Gammavoltaic was proposed by Karl Scharf in 1960. The low efficiency resulted in gammavoltaic being used as a radiation detector. In the 1990s the efficiency of gammavoltaic increased by the use of a scintillator. Gammavoltaic was further studied as a power source for spent fuel transportation and a nuclear battery in the 2000s Haneol Lee and Man-Sung Yim also suggested electricity generation system based on spent fuel stored inside the fuel pool of a nuclear power plant. This study proposed the systematic design of an electricity conversion system using CsI(Tl) scintillator and a-Si photovoltaic cell. As such, this study is selected to be a reference paper. The results of this paper indicate a self-absorption effect from the reference model. This effect is negligible while the irradiation degradation has to be considered. Two main ways to reduce radiation induced degradation are scintillator shielding and replacing scintillator material with a material having higher radiation resistance. The analysis of the scintillator used in the 'electricity generation system using gamma radiation from spent fuel' was performed to evaluate the ideal electricity generation in the reference research.

  6. The feasibility of modelling coupled processes in safety analysis of spent nuclear fuel disposal

    International Nuclear Information System (INIS)

    Rasilainen, K.; Luukkonen, A.; Niemi, A.; Poellae, J.; Olin, M.

    1999-01-01

    The potential of applying coupled modelling in the Finnish safety analysis programme has been reviewed. The study focused on the migration of radionuclides escaping from a spent fuel repository planned to be excavated in fractured bedrock. Two effects that can trigger various couplings in and around a spent fuel repository in Finland were studied in detail; namely heat generation in the spent fuel and the presence of deep, saline groundwaters. The latter have been observed in coastal areas. A systematic survey of the requirements of coupled modelling identified features that render such migration calculations a challenging task. In groundwater flow modelling there appears to be wide ranging uncertainty related to conceptualisation of flow systems and to the corresponding input data. In terms of migration related chemistry there appear to be large gaps in the underlying thermodynamic database for geochemical systems. Rock mechanical predictions are heavily dependent on knowing the location, structure and properties of dominant fractures; information which is extremely difficult to obtain. Conduction and convection of heat is understood well in principle. On the basis of this review, it appears that coupled migration modelling may not yet be at the stage of development that would allow its use as a standard modelling tool in performance assessments. However, a firmer basis for the conclusions reached can only be obtained after a systematic modelling exercise on a relevant and real migration problem has been carried out. (orig.)

  7. Global Spent Fuel Logistics Systems Study (GSFLS). Volume 3A. GSFLS technical analysis (appendix). Interim report

    Energy Technology Data Exchange (ETDEWEB)

    Kriger, A.

    1978-01-31

    This report is a part of the interim report documentation for the Global Spent Fuel Logistics System (GSFLS) study. The technical and financial considerations underlying a global spent fuel logistics systems have been studied and are reported. The Pacific Basin is used as a model throughout this report; however the stated methodology and, in many cases, considerations and conclusions are applicable to other global regions. Spent fuel discharge profiles for Pacific Basin Countries were used to determine the technical systems requirements for alternative concepts. Functional analyses and flows were generated to define both system design requirements and logistics parameters. A technology review was made to ascertain the state-of-the-art of relevant GSFLS technical systems. Modular GSFLS facility designs were developed using the information generated from the functional analysis and technology review. The modular facility designs were used as a basis for siting and cost estimates for various GSFLS alternatives. Various GSFLS concepts were analyzed from a financial and economic perspective in order to provide total concepts costs and ascertain financial and economic sensitivities to key GSFLS variations. Results of the study include quantification of GSFLS facility and hardware requirements; drawings of relevant GSFLS facility designs; system cost estimates; financial reports - including user service charges; and comparative analyses of various GSFLS alternatives.

  8. Global Spent Fuel Logistics Systems Study (GSFLS). Volume 3. GSFLS technical and financial analysis. Interim report

    International Nuclear Information System (INIS)

    1978-01-01

    This report is a part of the interim report documentation for the Global Spent Fuel Logistics System (GSFLS) study. The technical and financial considerations underlying a global spent fuel logistics systems have been studied and are reported herein. The Pacific Basin is used as a model throughout this report; however the stated methodology and, in many cases, considerations and conclusions are applicable to other global regions. Spent fuel discharge profiles for Pacific Basin Countries were used to determine the technical systems requirements for alternative concepts. Functional analyses and flows were generated to define both system design requirements and logistics parameters. A technology review was made to ascertain the state-of-the-art of relevant GSFLS technical systems. Modular GSFLS facility designs were developed using the information generated from the functional analysis and technology review. The modular facility designs were used as a basis for siting and cost estimates for various GSFLS alternatives. Various GSFLS concepts were analyzed from a financial and economic perspective in order to provide total concepts costs and ascertain financial and economic sensitivities to key GSFLS variations. Results of the study include quantification of GSFLS facility and hardware requirements; drawings of relevant GSFLS facility designs; system cost estimates; financial reports - including user service charges; and comparative analyses of various GSFLS alternatives

  9. Examination of the Properties of a Spent Fuel based Electricity Generation System - Scintillator Performance Analysis

    International Nuclear Information System (INIS)

    Lee, Haneol; Yim, Man-Sung

    2016-01-01

    Gammavoltaic was proposed by Karl Scharf in 1960. The low efficiency resulted in gammavoltaic being used as a radiation detector. In the 1990s the efficiency of gammavoltaic increased by the use of a scintillator. Gammavoltaic was further studied as a power source for spent fuel transportation and a nuclear battery in the 2000s Haneol Lee and Man-Sung Yim also suggested electricity generation system based on spent fuel stored inside the fuel pool of a nuclear power plant. This study proposed the systematic design of an electricity conversion system using CsI(Tl) scintillator and a-Si photovoltaic cell. As such, this study is selected to be a reference paper. The results of this paper indicate a self-absorption effect from the reference model. This effect is negligible while the irradiation degradation has to be considered. Two main ways to reduce radiation induced degradation are scintillator shielding and replacing scintillator material with a material having higher radiation resistance. The analysis of the scintillator used in the 'electricity generation system using gamma radiation from spent fuel' was performed to evaluate the ideal electricity generation in the reference research

  10. Analysis of radiation characteristics for casks loaded with spent RBMK-1500 nuclear fuel

    International Nuclear Information System (INIS)

    Smaizys, A.; Poskas, P.

    2001-01-01

    The objective of this paper is to present the analysis of radiation characteristics for the ductile cast iron CASTOR RBMK-1500 and heavy concrete CONSTOR RBMK-1500 casks loaded with spent nuclear fuel from Ignalina NPP RBMK-1500 reactors. These casks are designed for an interim storage (up to 50 years) of spent nuclear fuel at Ignalina NPP. Computer calculations have been performed using SCALE4.3 computer codes system. The dose rate calculations have been performed on the sidelong, upper and lower surface of the casks and for certain distance at the beginning of spent nuclear fuel storage in the casks and after 50 years of interim dry storage. The results obtained results show that dose rate values on the surface of the cask are much less than the design criteria value 1000 μSv/h when the average burn-up of fuel assembly is 20 GWd/tU. It was revealed that CONSTOR RBMK- 1500 cask has better shielding characteristics than CASTOR RBMK-1500 cask.(author)

  11. ANALYSIS OF ROMANIAN SMALL AND MEDIUM ENTERPRISES BANKRUPTCY RISK

    Directory of Open Access Journals (Sweden)

    Kulcsar Edina

    2014-07-01

    Full Text Available Considering the fundamental role of small and medium enterprises in Romanian economy, this paper aims to quantify the level of their bankruptcy risk for 2009 and 2012 period, after debuting of financial crisis. The main reason of selecting this type of companies is that they represent the backbone of national economy. They have an indispensable role, because they offer jobs for great part of population and their contribution for GDP stimulation is considerable. In this paper it was applied two default risk models, namely the well known Altman’s Z-score model, based on five financial ratios and a bankruptcy predictor model developed by Teti et. al (2012 used firstly exclusively for Italian small and medium-sized enterprise for 2006-2009 period. The model proposed by Teti et. is based on the investigation of financially distressed and financially non-distressed Italian small and medium-sized enterprises during the financial crisis by using a discriminant analysis model. They conclude that there are four financial ratios, which characterized well the small and medium-sized enterprises bankruptcy risk. These variables are financial ratios, like: Debt/Total Assets, Return on Sales (ROS, EBIT/Interest Expenses and Working capital/EBIDTA. They consider that small and medium-sized enterprises require a particular approach in terms of bankruptcy risk analysis. In present study I try to compare the efficiency of traditional bankruptcy risk model with a small and medium-sized specific model. The necessary database for present analysis is ensured by simplified financial reports of 120 small and medium-sized enterprises registered in Bihor County. The selected enterprises are operating in manufacturing industry (21,67% and trading (78,33%. Present investigation has an important value in actual economic background, where the healthiness and sustainability of small and medium-sized enterprises is a great issue. The results of study shows contradictory

  12. Whey proteins analysis in aqueous medium and in artificial gastric ...

    African Journals Online (AJOL)

    Whey proteins isolates (WPI) were treated in aqueous medium at various pH values. Zeta potential, turbidity and particle size measurement were determined as a function of pH. FTIR analysis was performed in ATR mode (attenuated total reflectance). Digestibility was assessed by treating whey proteins with artificial gastric ...

  13. Criticality analysis of PWR spent fuel storage facilities inside nuclear power plants

    International Nuclear Information System (INIS)

    Neuber, J.C.

    1999-01-01

    This paper describes some of the main features of the actinide plus fission product burnup credit methodology used by Siemens for criticality safety design analysis of wet PWR storage pools with soluble boron in the pool water. Application of burnup credit requires knowledge of the isotopic inventory of the irradiated fuel for which burnup credit is taken. This knowledge is gained by using depletion codes. The results of the depletion analysis are a necessary input to the criticality analysis. Siemens performs depletion calculations for PWR fuel burnup credit applications with the aid of the Siemens standard design procedure SAV90. The quality of this procedure relies on statistics on the differences between calculation and measurement extracted from in-core measurement data and chemical assay data. Siemens performs criticality safety calculations with the aid of the criticality calculation modules of the SCALE code package. These modules are verified many times with the aid of various kinds of critical experiments and configurations: Application of these modules to spent LWR fuel assembly storage pools was verified by analyzing critical experiments simulating such storage pools. Actinide plus fission product burnup credit applications of these modules were verified by analyzing PWR reactor critical configurations. The result of performing a burnup credit analysis is the determination of a burnup, credit loading curve for the spent fuel storage racks designed for burnup credit. This curve specifies the loading criterion by indicating the minimum burnup necessary for the fuel assembly with a specific initial enrichment to be placed in the storage racks designed for burnup credit. The loading of the spent fuel storage racks designed for burnup credit requires the implementation of controls to ensure that the loading curve is met. The controls include the determination of fuel assembly burnup based on reactor records. (author)

  14. Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications

    International Nuclear Information System (INIS)

    Parks, C.V.; Broadhead, B.L.; Hermann, O.W.; Tang, J.S.; Cramer, S.N.; Gauthey, J.C.; Kirk, B.L.; Roussin, R.W.

    1988-07-01

    This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, and point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs

  15. Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications. [Radiation dose rates from shielded spent fuels and high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Parks, C.V.; Broadhead, B.L.; Hermann, O.W.; Tang, J.S.; Cramer, S.N.; Gauthey, J.C.; Kirk, B.L.; Roussin, R.W.

    1988-07-01

    This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, and point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs.

  16. Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

    International Nuclear Information System (INIS)

    Lee, J.C.; Bang, K.S.; Seo, K.S.; Kim, H.D.; Choi, B.I.; Lee, H.Y.; Song, M.J.

    2004-01-01

    The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 27 □ under the normal condition. The off-normal condition has an environmental temperature of 40 □. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of ventilation system have been carried out for the determination of the optimum duct size and shape. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. In the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition and off-normal conditions

  17. Feasibility of subcriticality and NDA measurements for spent fuel by frequency analysis techniques with 252Cf

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; Valentine, T.E.; Mattingly, J.K.

    1996-01-01

    The 252 Cf-source-driven frequency analysis method can be used for measuring the subcritical neutron multiplication factor of arrays of LWR fuel and as little as a single PWR fuel assembly. These measurements can be used to verify the criticality safety margins of spent LWR fuel configurations and thus could be a means of obtaining the information to justify burnup credit for spent LWR transportation/storage casks. In addition, the data can be used to validate calculational methods for criticality safety. These measurements provide parameters that have a higher sensitivity to changes in fissile mass than neutron multiplication factor and thus serve as a better test of calculational methods. The analysis have also shown that measurement of the cross power spectral density (CPSD) between detectors on one side of a single fuel assembly and an internal or external 252 Cf source driving the fission chain multiplication process can be used for nondestructive assay of fissile mass along the length of the assembly. This CPSD is a smooth function of fissile mass and does not depend on the varying inherent source in the fuel assembly and thus is ideal for fissile mass assay

  18. A simplified modeling technique for the thermal analysis of spent nuclear assemblies

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-01-01

    A major design concern of a cask for the transport of spent fuel assemblies is the determination of the maximum fuel pin temperature. Normally the determination of the maximum steady state fuel pin temperature requires the thermal analysis (two-dimensional or three-dimensional) of the cask to be performed in conjunction with the fuel assemblies. This, in general, would require extensive effort and computer time. However, by performing a sensitivity analysis for fuel pin temperature for various boundary conditions it was possible to decouple the heat transfer analysis of the cask from that of the fuel assemblies. This paper presents the scaling law that was obtained for maximum fuel pin temperature as a function of emissivity, heat generation rate and assembly wrapper temperature for hex-shaped fuel assemblies

  19. Analysis of the Processes in Spent Fuel Pools in Case of Loss of Heat Removal due to Water Leakage

    Directory of Open Access Journals (Sweden)

    Algirdas Kaliatka

    2013-01-01

    Full Text Available The safe storage of spent fuel assemblies in the spent fuel pools is very important. These facilities are not covered by leaktight containment; thus, the consequences of overheating and melting of fuel in the spent fuel pools can be very severe. On the other hand, due to low decay heat of fuel assemblies, the processes in pools are slow in comparison with processes in reactor core during LOCA accident. Thus, the accident management measures play a very important role in case of some accidents in spent fuel pools. This paper presents the analysis of possible consequences of fuel overheating due to leakage of water from spent fuel pool. Also, the accident mitigation measure, the late injection of water was evaluated. The analysis was performed for the Ignalina NPP Unit 2 spent fuel pool, using system thermal hydraulic code for severe accident analysis ATHLET-CD. The phenomena, taking place during such accident, are discussed. Also, benchmarking of results of the same accident calculation using ASTEC and RELAP/SCDAPSIM codes is presented here.

  20. Water Quality Analysis Study Pond and Interim Storage for Spent Fuel

    International Nuclear Information System (INIS)

    Dyah Sulistyani R; Husen Zamroni; Sudiyati

    2007-01-01

    Purification system of Storage facility of spent fuel which there is in Indonesia is integrated purification system. Reservoir pond of fuel contains approximately 995 m 3 demin water and in pond equipped with some of reservoir racks of spent fuel which must always avoid from factor-factor causing corrosion. In process of this purification system, water impurity which has been activation and also which is not is activation before will filtered and catch by passing of ion exchange so that will reduce conductivity and fuel coolant water activity. Water quality pond and canals links must fulfill specifications, among other: degree of acidity (pH) primary cooling water ranges from 5.5 and 6.5 ; its conductivity 1 - 8 μ S/cm, content analysis CI 0.03 - 0.06 ppm and NO 3 0.1 - 0.2 ppm, radionuclide activity Cs 137 742 Bq/l and Co 60 657 Bq/l and the temperature be kept of less than 40℃ to avoid from corrosion speed. (author)

  1. Sampling and analysis plan for the preoperational environmental survey of the spent nuclear fuel project facilities

    Energy Technology Data Exchange (ETDEWEB)

    MITCHELL, R.M.

    1999-04-01

    This sampling and analysis plan will support the preoperational environmental monitoring for construction, development, and operation of the Spent Nuclear Fuel (SNF) Project facilities, which have been designed for the conditioning and storage of spent nuclear fuels; particularly the fuel elements associated with the operation of N-Reactor. The SNF consists principally of irradiated metallic uranium, and therefore includes plutonium and mixed fission products. The primary effort will consist of removing the SNF from the storage basins in K East and K West Areas, placing in multicanister overpacks, vacuum drying, conditioning, and subsequent dry vault storage in the 200 East Area. The primary purpose and need for this action is to reduce the risks to public health and safety and to the environment. Specifically these include prevention of the release of radioactive materials into the air or to the soil surrounding the K Basins, prevention of the potential migration of radionuclides through the soil column to the nearby Columbia River, reduction of occupational radiation exposure, and elimination of the risks to the public and to workers from the deterioration of SNF in the K Basins.

  2. Sampling and analysis plan for the preoperational environmental survey of the spent nuclear fuel project facilities

    International Nuclear Information System (INIS)

    MITCHELL, R.M.

    1999-01-01

    This sampling and analysis plan will support the preoperational environmental monitoring for construction, development, and operation of the Spent Nuclear Fuel (SNF) Project facilities, which have been designed for the conditioning and storage of spent nuclear fuels; particularly the fuel elements associated with the operation of N-Reactor. The SNF consists principally of irradiated metallic uranium, and therefore includes plutonium and mixed fission products. The primary effort will consist of removing the SNF from the storage basins in K East and K West Areas, placing in multicanister overpacks, vacuum drying, conditioning, and subsequent dry vault storage in the 200 East Area. The primary purpose and need for this action is to reduce the risks to public health and safety and to the environment. Specifically these include prevention of the release of radioactive materials into the air or to the soil surrounding the K Basins, prevention of the potential migration of radionuclides through the soil column to the nearby Columbia River, reduction of occupational radiation exposure, and elimination of the risks to the public and to workers from the deterioration of SNF in the K Basins

  3. PWR core and spent fuel pool analysis using scale and nestle

    International Nuclear Information System (INIS)

    Murphy, J. E.; Maldonado, G. I.; St Clair, R.; Orr, D.

    2012-01-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  4. PWR core and spent fuel pool analysis using scale and nestle

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

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

    International Nuclear Information System (INIS)

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

    2008-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-15

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

  7. Development of MAAP5.0.3 Spent Fuel Pool Model for Severe Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Mi Ro [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    After the Fukushima accident, the severe accident phenomena in the Spent Fuel Pool (SFP) have been the great issues in the nuclear industry. Generally, during full power operation status, the decay heat of the spent fuel in the SFP is not high enough to cause the severe accident that is the say, the melting of fuel and fuel rack. In addition to this, the SFP of the PWR is not isolated within the containment like the SFP of the old BWR plant, there are so many possible measures to prevent and mitigate severe accidents in the SFP. On the other hand, in the low power shutdown status (fuel refueling), all the core is transferred into the SFP during the refueling period. At this period, if some accidents happen such as the loss of SFP cooling and the failure of SFP integrity then the accidents may be developed into severe accident because the decay heat is high enough. So, the analysis of severe accidents in the SFP during low power shutdown state is greatly affected to the establishment of the major strategies in the severe accident management guideline (SAMG). However, the status of the domestic technical background for those analyses is very weak. it is known that the decay heat of the spent fuel in the SFP is not high enough to cause the severe accident qualitatively. However, there are some possibilities that can cause the severe accidents in the SFP if the loss of SFP cooling and integrity happens simultaneously. The severe accident phenomena in SFP themselves are not much different from those in the containment. However, since the structure of SFP cannot be isolated during the accidents like the containment, the consequence can be extremely significant. So, in terms of the establishment of the severe accident management strategy, it is necessary that the quantitative analysis for the severe accident progression in the SFP should be performed. In this study, the general behavior which can be appeared during the severe accidents in the SFP was analyzed using the

  8. ORIGAMI Automator Primer. Automated ORIGEN Source Terms and Spent Fuel Storage Pool Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wieselquist, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Thompson, Adam B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    Source terms and spent nuclear fuel (SNF) storage pool decay heat load analyses for operating nuclear power plants require a large number of Oak Ridge Isotope Generation and Depletion (ORIGEN) calculations. SNF source term calculations also require a significant amount of bookkeeping to track quantities such as core and assembly operating histories, spent fuel pool (SFP) residence times, heavy metal masses, and enrichments. The ORIGEN Assembly Isotopics (ORIGAMI) module in the SCALE code system provides a simple scheme for entering these data. However, given the large scope of the analysis, extensive scripting is necessary to convert formats and process data to create thousands of ORIGAMI input files (one per assembly) and to process the results into formats readily usable by follow-on analysis tools. This primer describes a project within the SCALE Fulcrum graphical user interface (GUI) called ORIGAMI Automator that was developed to automate the scripting and bookkeeping in large-scale source term analyses. The ORIGAMI Automator enables the analyst to (1) easily create, view, and edit the reactor site and assembly information, (2) automatically create and run ORIGAMI inputs, and (3) analyze the results from ORIGAMI. ORIGAMI Automator uses the standard ORIGEN binary concentrations files produced by ORIGAMI, with concentrations available at all time points in each assembly’s life. The GUI plots results such as mass, concentration, activity, and decay heat using a powerful new ORIGEN Post-Processing Utility for SCALE (OPUS) GUI component. This document includes a description and user guide for the GUI, a step-by-step tutorial for a simplified scenario, and appendices that document the file structures used.

  9. Development of Methodology for Spent Fuel Pool Severe Accident Analysis Using MELCOR Program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won-Tae; Shin, Jae-Uk [RETech. Co. LTD., Yongin (Korea, Republic of); Ahn, Kwang-Il [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The general reason why SFP severe accident analysis has to be considered is that there is a potential great risk due to the huge number of fuel assemblies and no containment in a SFP building. In most cases, the SFP building is vulnerable to external damage or attack. In contrary, low decay heat of fuel assemblies may make the accident processes slow compared to the accident in reactor core because of a great deal of water. In short, its severity of consequence cannot exclude the consideration of SFP risk management. The U.S. Nuclear Regulatory Commission has performed the consequence studies of postulated spent fuel pool accident. The Fukushima-Daiichi accident has accelerated the needs for the consequence studies of postulated spent fuel pool accidents, causing the nuclear industry and regulatory bodies to reexamine several assumptions concerning beyond-design basis events such as a station blackout. The tsunami brought about the loss of coolant accident, leading to the explosion of hydrogen in the SFP building. Analyses of SFP accident processes in the case of a loss of coolant with no heat removal have studied. Few studies however have focused on a long term process of SFP severe accident under no mitigation action such as a water makeup to SFP. USNRC and OECD have co-worked to examine the behavior of PWR fuel assemblies under severe accident conditions in a spent fuel rack. In support of the investigation, several new features of MELCOR model have been added to simulate both BWR fuel assembly and PWR 17 x 17 assembly in a spent fuel pool rack undergoing severe accident conditions. The purpose of the study in this paper is to develop a methodology of the long-term analysis for the plant level SFP severe accident by using the new-featured MELCOR program in the OPR-1000 Nuclear Power Plant. The study is to investigate the ability of MELCOR in predicting an entire process of SFP severe accident phenomena including the molten corium and concrete reaction. The

  10. Analysis of a hypothetical dropped spent nuclear fuel shipping cask impacting a floor mounted crush pad

    International Nuclear Information System (INIS)

    Hawkes, B.D.; Uldrich, E.D.

    1998-03-01

    A crush pad has been designed and analyzed to absorb the kinetic energy of a hypothetically dropped spent nuclear fuel shipping cask into a 44-ft. deep cask unloading pool at the Idaho Chemical Processing Plant. The 110-ton Large Cell Cask was assumed to be accidentally dropped onto the parapet of the unloading pool, causing the cask to tumble through the pool water and impact the floor mounted crush pad with the cask's top corner. The crush pad contains rigid polyurethane foam, which was modeled in a separate computer analysis to simulate the manufacturer's testing of the foam and to determine the foam's stress and strain characteristics. This computer analysis verified that the foam was accurately represented in the analysis to follow. A detailed non-linear, dynamic finite element analysis was then performed on the crush pad and adjacent pool structure to assure that a drop of this massive cask does not result in unacceptable damage to the storage facility. Additionally, verification was made that the crush pad adequately protects the cask from severe impact loading. At impact, the cask has significant vertical, horizontal and rotational velocities. The crush pad absorbs much of the energy of the cask through plastic deformation during primary and secondary impacts. After the primary impact with the crush pad, the cask still has sufficient energy to rebound and rotate until it impacts the pool wall. An assessment is made of the damage to the crush pad and pool wall and of the impact loading on the cask

  11. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 3, Validation assessments

    International Nuclear Information System (INIS)

    Lombardo, N.J.; Cuta, J.M.; Michener, T.E.; Rector, D.R.; Wheeler, C.L.

    1986-12-01

    This report presents the results of the COBRA-SFS (Spent Fuel Storage) computer code validation effort. COBRA-SFS, while refined and specialized for spent fuel storage system analyses, is a lumped-volume thermal-hydraulic analysis computer code that predicts temperature and velocity distributions in a wide variety of systems. Through comparisons of code predictions with spent fuel storage system test data, the code's mathematical, physical, and mechanistic models are assessed, and empirical relations defined. The six test cases used to validate the code and code models include single-assembly and multiassembly storage systems under a variety of fill media and system orientations and include unconsolidated and consolidated spent fuel. In its entirety, the test matrix investigates the contributions of convection, conduction, and radiation heat transfer in spent fuel storage systems. To demonstrate the code's performance for a wide variety of storage systems and conditions, comparisons of code predictions with data are made for 14 runs from the experimental data base. The cases selected exercise the important code models and code logic pathways and are representative of the types of simulations required for spent fuel storage system design and licensing safety analyses. For each test, a test description, a summary of the COBRA-SFS computational model, assumptions, and correlations employed are presented. For the cases selected, axial and radial temperature profile comparisons of code predictions with test data are provided, and conclusions drawn concerning the code models and the ability to predict the data and data trends. Comparisons of code predictions with test data demonstrate the ability of COBRA-SFS to successfully predict temperature distributions in unconsolidated or consolidated single and multiassembly spent fuel storage systems

  12. Investigation on structural analysis computer program of spent nuclear fuel shipping cask, (2)

    International Nuclear Information System (INIS)

    Yagawa, Ganki; Ikushima, Takeshi.

    1987-10-01

    This report describes the results (II) done by the Sub-Committee of Research Cooperation Committee (RC-62) of the Japan Society of Mechanical Engineers under the trust of the Japan Atomic Energy Research Institute. The principal fulfilments and accomplishments are summarized as follows: (1) Regarding the survey of structural analysis methods of spent fuel shipping cask, several documents, which explain the features and applications of the exclusive computer programs for impact analysis on the basis of 2 or 3 dimensional finite element or difference methods, were reviewed. (2) In comparative evaluation of the existing computer programs, the common benchmark test problems for drop impact of the axisymmetric cylinder and plate were adopted the calculational evaluations for taking into account the strain rate effect of material properties, effect of artificial viscosity and effect of time integration step size were carried out. (3) Evaluation of impact analysis algorithm of computer programs were conducted and the requirements for computer programs to be developed in future and an index for further studies have been clarified. (author)

  13. Investigation on structural analysis computer program of spent nuclear fuel shipping cask

    International Nuclear Information System (INIS)

    Yagawa, Ganki; Ikushima, Takeshi.

    1987-10-01

    This report describes the results done by the Sub-Committee of Research Cooperation Committee (RC-62) of the Japan Society of Mechanical Engineers under the trust of the Japan Atomic Energy Research Institute. The principal fulfilments and accomplishments are summarized as follows: (1) Regarding the survey of structural analysis methods of spent fuel shipping cask, several documents, which explain the features and applications of the exclusive computer programs for impact analysis on the basis of 2 or 3 dimensional finite element or difference methods such as HONDO, STEALTH and DYNA-3D, were reviewed. (2) In comparative evaluation of the existing computer programs, the common benchmark test problems for 9 m vertical drop impact of the axisymmetric lead cylinder with and without stainless steel clads were adopted where the calculational evaluations for taking into account the strain rate effect were carried out. (3) Evaluation of impact analysis algorithm of computer programs were conducted and the requirements for computer programs to be developed in future and an index for further studies have been clarified. (author)

  14. Analysis of raft foundations for spent fuel pool in nuclear facilities

    International Nuclear Information System (INIS)

    Subramanian, K.V.; Kashikar, A.V.; Nath, C.; Shintre, C.C.

    2005-01-01

    Foundation rafts are analysed as a plate on elastic foundation with the representation of the foundation media using the Winkler idealisation i.e. series of linear uncoupled springs. The elastic constant of the Winkler springs is derived using the sub-grade modulus. However, the Winkler approach has limitations due to incompatibility of the deflections at raft-soil interface. The deflection of the raft at the point of contact and the deformation of the foundation media at this point of contact are incompatible in this approach. This particularly influences flexible rafts and further if the foundation media is soil. This paper discusses the analysis of raft, in general, and the analysis of the foundation raft for a Spent Fuel pool facility using 'variable k approach' where deformations at a node and influencing nodes are computed using Boussinesq's theory. The limitations stated above are overcome in this approach. Some studies on the sensitivity of parameters were carried out in the form of variation of moduli of elasticity of concrete and deformation modulus of soil. Analysis is also performed with conventional method using 'Winkler' soil springs. It is concluded that the Winkler model does not correctly predict the behaviour of the mat both qualitatively and quantitatively and could lead to underestimation of soil pressures leading to unconservative design. The approach involving soil structure interaction like the one presented here is hence recommended for important structures like those involved in Nuclear facilities. (authors)

  15. Seismic tipping analysis of a spent nuclear fuel shipping cask sitting on a crush pad

    International Nuclear Information System (INIS)

    Uldrich, E.D.; Hawkes, B.D.

    1998-04-01

    A crush pad has been designed and analyzed to absorb the kinetic energy of an accidentally dropped spent nuclear fuel shipping cask into a 44 ft. deep cask unloading pool. Conventional analysis techniques available for evaluating a cask for tipping due to lateral seismic forces assume that the cask rests on a rigid surface. In this analysis, the cask (110 tons) sits on a stainless steel encased (0.25 in. top plate), polyurethane foam (4 ft. thick) crush pad. As the cask tends to rock due to horizontal seismic forces, the contact area between the cask and the crush pad is reduced, increasing the bearing stress, and causing the pivoting corner of the cask to depress into the crush pad. As the crush pad depresses under the cask corner, the pivot point shifts from the corner toward the cask center, which facilitates rocking and potential tipping of the cask. Subsequent rocking of the cask may deepen the depression, further contributing to the likelihood of cask tip over. However, as the depression is created, the crush pad is absorbing energy from the rocking cask. Potential tip over of the cask was evaluated by performing a non-linear, dynamic, finite element analysis with acceleration time history input. This time history analysis captured the effect of a deforming crush pad, and also eliminated conservatisms of the conventional approaches. For comparison purposes, this analysis was also performed with the cask sitting on a solid stainless steel crush pad. Results indicate that the conventional methods are quite conservative relative to the more exacting time history analysis. They also indicate that the rocking motion is less on the foam crush pad than on the solid stainless steel pad

  16. Infrared analysis of clay bricks incorporated with spent shea waste from the shea butter industry.

    Science.gov (United States)

    Adazabra, A N; Viruthagiri, G; Shanmugam, N

    2017-04-15

    The peculiar challenge of effective disposing abundant spent shea waste and the excellent compositional variation tolerance of clay material offered an impetus to examine the incorporation of spent shea waste into clay material as an eco-friendly disposal route in making clay bricks. For this purpose, the chemical constituent, mineralogical compositions and thermal behavior of both clay material and spent shea waste were initially characterized from which modelled brick specimens incorporating 5-20 wt% of the waste into the clay material were prepared. The clay material showed high proportions of SiO 2 (52.97 wt%) and Al 2 O 3 (27.10 wt%) indicating their rich kaolinitic content: whereas, the inert nature of spent shea waste was exhibited by their low oxide content. The striking similarities in infrared absorption bands of pristine clay material and clay materials incorporated with 15 wt% of spent shea waste showed that the waste incorporation had no impact on bond formation of the clay bricks. Potential performance benefits of developing bricks from clay material incorporated with spent shea waste included improved fluxing agents, economic sintering and making of sustainable bricks. Consequently, the analytical results authenticate the incorporation of spent shea waste into clay materials for various desired benefits aside being an environmental correct route of its disposal. Copyright © 2017 Elsevier Ltd. All rights reserved.

  17. Regulatory analysis for the resolution of Generic Issue 82, ''Beyond design basis accidents in spent fuel pools''

    International Nuclear Information System (INIS)

    Throm, E.D.

    1989-04-01

    Generic Issue 82, ''Beyond Design Basis Accidents in Spent Fuel Pools,'' addresses the concerns with the use of high density storage racks for the storage of spent fuel, and is applicable to all Light Water Reactor spent fuel pools. This report presents the regulatory analysis for Generic Issue 82. It includes (1) a summary of the issue, (2) a summary of the technical findings, (3) the proposed technical resolution, (4) alternative resolutions considered by the Nuclear Regulatory Commission, (5) an assessment of the benefits and cost of the alternatives considered, (6) the decision rationale, and (7) the relationships between Generic Issue 82 and other NRC programs and requirements. Based on this evaluation, the NRC staff concludes that no new regulatory requirements are warranted concerning the use of high density storage racks. 48 refs., 32 tabs

  18. Criticality and Its Uncertainty Analysis of Spent Fuel Storage Rack for Research Reactor

    International Nuclear Information System (INIS)

    Han, Tae Young; Park, Chang Je; Lee, Byung Chul

    2011-01-01

    For evaluating the criticality safety of spent fuel storage rack in an open pool type research reactor, a permissible upper limit of criticality should be determined. It can be estimated from the criticality upper limit presented by the regulatory guide and an uncertainty of criticality calculation. In this paper, criticalities for spent fuel storage rack are carried out at various conditions. The calculation uncertainty of MCNP system is evaluated from the calculation results for the benchmark experiments. Then, the upper limit of criticality is determined from the uncertainties and the calculated criticality of the spent fuel storage rack is evaluated

  19. Properties of the LiCl-KCl-Li2O system as operating medium for pyro-chemical reprocessing of spent nuclear fuel

    Science.gov (United States)

    Mullabaev, Albert; Tkacheva, Olga; Shishkin, Vladimir; Kovrov, Vadim; Zaikov, Yuriy; Sukhanov, Leonid; Mochalov, Yuriy

    2018-03-01

    Crystallization temperatures (liquidus and solidus) in the LiCl-Li2O and (LiCl-KCl)-Li2O systems with the KCl content of 10 and 20 mol.% were obtained with independent methods of thermal analysis using cooling curves, isothermal saturation, and differential scanning calorimetry. The linear sweep voltammetry was applied to control the time of the equilibrium establishment in the molten system after the Li2O addition, which depended on the composition of the base melt and the concentration of Li2O. The fragments of the binary LiCl-Li2O and quazi-binary [LiCl-KCl(10 mol.%)]-Li2O and [LiCl-KCl(20 mol.%)]-Li2O phase diagrams in the Li2O concentration range from 0 to 12 mol.% were obtained. The KCl presence in the LiCl-KCl-Li2O molten mixture in the amount of 10 and 20 mol.% reduces the liquidus temperature by 30 and 80°, respectively, but the region of the homogeneous molten state of the system is considerably narrowed, which complicates its practical application. The Li2O solubility in the molten LiCl, LiCl-KCl(10 mol.%) and LiCl-KCl(20 mol.%) decreases with increasing the KCl content and is equal to 11.5, 7.7 and 3.9 mol.% at 650°С, respectively. The LiCl-KCl melt with 10 mol.% KCl can be recommended for practical use as a medium for the SNF pyro-chemical reprocessing at temperature below 700 °C.

  20. Radiological pathways analysis for spent solvents from the boiler chemical cleaning at the Pickering Nuclear Site

    International Nuclear Information System (INIS)

    Garisto, N.C.; Eslami, Z.; Hodgins, S.; Beaman, T.; Von Svoboda, S.; Marczak, J.

    2006-01-01

    Spent solvents are generated as a result of Boiler Chemical Cleanings (BCC) at CANDU reactor sites. These solutions contain small amount of radioactivity from a number of different sources including: Cut tubes - short sections of boiler tubes are infrequently removed from the boilers for a detailed characterization. These tubes are typically only plugged at the tubesheet allowing the primary side deposits to be exposed to BCC solvents. Tube leaks - primary to secondary side leaks also occur infrequently as a result of tube degradation. Radioactivity from the leaking fluid can consequently be deposited in the sludge on the secondary side of the tubes. Diffusion of tritium - during normal operation of the reactor units, tritium slowly diffuses from the heavy water in the primary heat-transfer system to the light-water coolant on the secondary side. Some of this tritium is retained in the secondary side deposits. The Pickering Nuclear Generating Station (PNGS) would like the flexibility to have several options for handling the spent solvent waste and associated rinse water from BCC. To this end, a radiological pathways analysis was undertaken to determine dose consequences associated with each option. Sample results from this study are included in this paper. The pathways analysis is used in this study to calculate dose to hypothetical receptors including individuals such as truck drivers, incinerator workers, residue (ash) handlers, residents who live near the landfill, inadvertent intruders into the landfill after closure and residents who live near the outfall. This dose is compared to a de minimis dose. A de minimis dose or dose rate represents a level of risk, which is generally accepted as being of no significance. Shipments of spent solvents and rinse water with corresponding doses below de minimis can be sent to conventional (i.e., non-radioactive) landfills for incineration and disposal as the radioactive dose associated with them is much less than natural

  1. Spent fuel pool thermal-hydraulic analysis using RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Ramos, M. C.; Fernandes, G.H.N.; Costa, A.L.; Pereira, F.; Pereira, C., E-mail: marc5663@gmail.com, E-mail: ghnfernandes@pq.cnpq.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    In order to analyze the thermo-hydraulic behavior of spent fuel pools, and taking as reference a hypothetic PWR nuclear plant, a model of RELAP-3D for a spent fuel pool has been built. This model has been used to simulate a loss of coolant in SPF. This study focuses on the loss of coolant flow accident in spent fuel storage pool which is modelled by using RELAP5-3D code to observe the coolant level reduction and fuel uncovery because of decay heat generation of the spent fuel in the pool. The results have been compared with the available data. The developed model demonstrated that the RELAP5-3D is capable of reproduce the thermal behavior of SPF in a transient scenario. (author)

  2. Design and analysis of free-standing spent fuel racks in nuclear power plants

    International Nuclear Information System (INIS)

    Ashar, H.; DeGrassi, G.

    1989-01-01

    With the prohibition on reprocessing of spent fuel in the late 1970's the pools which were supposed to be short term storage became quasi-permanent storage spaces for spent fuel. Recognizing a need to provide permanent storage facilities for such nuclear wastes, the US Congress enacted a law cited as the Nuclear Waste Policy Act of 1982. The Act, in essence, required the Department of Energy to find ways for long term storage of high level waste. However, it also is required the owners of nuclear power plants to provide for interim storage of their spent fuel. The permanent government owned repositories are not scheduled to be operational until the year 2005. In order to accommodate the increasing inventory of spent fuel, the US utilities started looking for various means to store spent fuel at the reactor sites. One of the most economical ways to accommodate more spent fuel is to arrange storage locations as closely as possible at the same time making sure that the fuel remains subcritical and that there are adequate means to cope with the heat load. The free standing high density rack configuration is an outcome of efforts to accommodate to more fuel in the limited space. 3 refs., 3 figs

  3. CANDU spent fuel shielding analysis during intermediate dry storage by using Monte Carlo methodology

    International Nuclear Information System (INIS)

    Margeanu, Cristina Alice; Ilie, Petre

    2006-01-01

    Almost all the countries that operate or construct nuclear power plants have r and d programs for spent nuclear fuel and radioactive waste management. In these programs, optimal solutions for nuclear fuel cycle management are to be identified, geological disposal being one of the main goals here. Romania did not yet adopt a decision for final disposal. Nevertheless, researches and studies are in progress in order to select and characterize the geological formation for spent fuel final disposal. Currently, although there is no comprehensive EU policy in the field of safe spent fuel and radioactive waste management it is desirable to bring and keep the safety of radioactive waste management on a uniform high level among the Member States and the accession countries. The Romanian Cernavoda NPP, of CANDU type, has the following spent fuel management facilities: a spent fuel bay (for spent fuel wet storage) and a spent fuel interim dry storage facility. The dry storage technology is based on MACSTOR system consisting of storage modules located outdoors in the storage site, and equipment operated at the spent fuel storage bay for preparing the spent fuel for dry storage. The spent fuel is transferred from the preparation area to the storage site in a transfer flask. The concrete storage modules have two sealed barriers for storing the spent fuel: a seal welded stainless steel basket containing 60 spent fuel bundles and a seal welded cylinder containing 10 baskets. Twenty storage cylinders are in one storage module for a total capacity of 12,000 bundles per module. In 2003, the first storage module has become operational. The paper has the following contents: Introduction; The paper objectives; Theoretical model Set-Up; Results; Conclusions. In conclusions one notifies that SEU fuel leads to higher burnup degrees associated both with spent fuel and actinides mass reduction for 1 MWh generated electric power (from 7100 MWD/tU for UNAT to 20000 MWD/tU for SEU43

  4. Standard review plan for reviewing safety analysis reports for dry metallic spent fuel storage casks

    International Nuclear Information System (INIS)

    1988-01-01

    The Cask Standard Review Plan (CSRP) has been prepared as guidance to be used in the review of Cask Safety Analysis Reports (CSARs) for storage packages. The principal purpose of the CSRP is to assure the quality and uniformity of storage cask reviews and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The CSRP also sets forth solutions and approaches determined to be acceptable in the past by the NRC staff in dealing with a specific safety issue or safety-related design area. These solutions and approaches are presented in this form so that reviewers can take consistent and well-understood positions as the same safety issues arise in future cases. An applicant submitting a CSAR does not have to follow the solutions or approaches presented in the CSRP. However, applicants should recognize that the NRC staff has spent substantial time and effort in reviewing and developing their positions for the issues. A corresponding amount of time and effort will probably be required to review and accept new or different solutions and approaches

  5. Techno-economic analysis for brewer's spent grains use on a biorefinery concept: the Brazilian case.

    Science.gov (United States)

    Mussatto, Solange I; Moncada, Jonathan; Roberto, Inês C; Cardona, Carlos A

    2013-11-01

    A techno-economic analysis for use of brewer's spent grains (BSG) on a biorefinery concept for the Brazilian case is presented. Four scenarios based on different levels of heat and mass integration for the production of xylitol, lactic acid, activated carbon and phenolic acids are shown. A simulation procedure using the software Aspen Plus and experimental yields was used. Such procedure served as basis for the techno-economic and environmental assessment according to the Brazilian conditions. Full mass integration on water and full energy integration was the configuration with the best economic and environmental performance. For this case, the obtained economic margin was 62.25%, the potential environmental impact was 0.012 PEI/kg products, and the carbon footprint of the processing stage represented 0.96 kg CO2-e/kg of BSG. This result served as basis to draw recommendations on the technological, economic and environmental feasibility for implementation of such type of biorefinery in Brazil. Copyright © 2013 Elsevier Ltd. All rights reserved.

  6. COMPARATIVE ANALYSIS OF TRANSPORT AIRCRAFT, BACKROUND FOR SHORT/ MEDIUM COURIER TRANSPORT AIRCRAFT PROCUREMENT

    Directory of Open Access Journals (Sweden)

    Matei POPA

    2010-03-01

    Full Text Available In accordance with Air Force requirements, the comparative analysis of short/medium transport aircraft comes to sustain procurement decision of short/medium transport aircraft. This paper presents, in short, the principles and the results of the comparative analysis for short/medium military transport aircraft.

  7. Cost Estimation and Efficiency Analysis of Korean CANDU Spent Fuel Disposal Alternatives in Consideration of Future Price Volatility

    Directory of Open Access Journals (Sweden)

    Sungsig Bang

    2016-01-01

    Full Text Available In Korea, spent fuel is temporarily stored in spent fuel pools at nuclear reactor sites and it is predicted to become saturated between 2020 and 2024. For this reason, four disposal alternatives (KRS-1, A-KRS-1, A-KRS-21, and A-KRS-22 have been developed in order to carry out the direct disposal of the CANDU spent fuel. The objective of this study is to conduct cost efficiency analysis of the disposal alternatives in consideration of price volatility for the radioactive waste repository. To derive future price volatility, this study used the ARIMA model. As a result, A-KRS-1 is the most efficient in terms of price per bundle using 2015 price. As for the results using ARIMA model, except in the case of KRS-1, the cost per bundle of A-KRS-1, A-KRS-21, and A-KRS-22 is decreased. Cost estimation using ARIMA model shows little change or decreases in cost while cost estimation using inflation rates for 2020 resulted in approximately 7.2% increases compared to 2015 for all options. As for the results of scenario analysis, A-KRS-1 earned 8,160 points, while A-KRS-22 followed closely behind with 7,980 points among the total 24,300 points. The results of this study provide invaluable policy data for any nation considering the construction of spent nuclear fuel repository.

  8. Comparison and Analysis of Regulatory and Derived Requirements for Certain DOE Spent Nuclear Fuel Shipments; Lessons Learned for Future Spent Fuel Transportation Campaigns

    International Nuclear Information System (INIS)

    Kramer, George L.; Fawcett, Rick L.; Rieke, Philip C.

    2003-01-01

    Radioactive materials transportation is stringently regulated by the Department of Transportation and the Nuclear Regulatory Commission to protect the public and the environment. As a Federal agency, however, the U.S. Department of Energy (DOE) must seek State, Tribal and local input on safety issues for certain transportation activities. This interaction has invariably resulted in the imposition of extra-regulatory requirements, greatly increasing transportation costs and delaying schedules while not significantly enhancing the level of safety. This paper discusses the results an analysis of the regulatory and negotiated requirements established for a July 1998 shipment of spent nuclear fuel from foreign countries through the west coast to the Idaho National Engineering and Environmental Laboratory (INEEL). Staff from the INEEL Nuclear Materials Engineering and Disposition Department undertook the analysis in partnership with HMTC, to discover if there were instances where requirements derived from stakeholder interactions duplicate, contradict, or otherwise overlap with regulatory requirements. The study exhaustively lists and classifies applicable Department of Transportation (DOT) and Nuclear Regulatory Commission (NRC) regulations. These are then compared with a similarly classified list of requirements from the Environmental Impact Statements (EIS) and those developed during stakeholder negotiations. Comparison and analysis reveals numerous attempts to reduce transportation risk by imposing more stringent safety measures than those required by DOT and NRC. These usually took the form of additional inspection, notification and planning requirements. There are also many instances of overlap with, and duplication of regulations. Participants will gain a greater appreciation for the need to understand the risk-oriented basis of the radioactive materials regulations and their effectiveness in ensuring safety when negotiating extra-regulatory requirements

  9. Passive Tomography for Spent Fuel Verification: Analysis Framework and Instrument Design Study

    Energy Technology Data Exchange (ETDEWEB)

    White, Timothy A.; Svard, Staffan J.; Smith, Leon E.; Mozin, Vladimir V.; Jansson, Peter; Davour, Anna; Grape, Sophie; Trellue, H.; Deshmukh, Nikhil S.; Wittman, Richard S.; Honkamaa, Tapani; Vaccaro, Stefano; Ely, James

    2015-05-18

    The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly is being assessed through a collaboration of Support Programs to the International Atomic Energy Agency (IAEA). In the first phase of this study, two safeguards verification objectives have been identified. The first is the independent determination of the number of active pins that are present in the assembly, in the absence of a priori information. The second objective is to provide quantitative measures of pin-by-pin properties, e.g. activity of key isotopes or pin attributes such as cooling time and relative burnup, for the detection of anomalies and/or verification of operator-declared data. The efficacy of GET to meet these two verification objectives will be evaluated across a range of fuel types, burnups, and cooling times, and with a target interrogation time of less than 60 minutes. The evaluation of GET viability for safeguards applications is founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types are used to produce simulated tomographer responses to large populations of “virtual” fuel assemblies. Instrument response data are processed by a variety of tomographic-reconstruction and image-processing methods, and scoring metrics specific to each of the verification objectives are defined and used to evaluate the performance of the methods. This paper will provide a description of the analysis framework and evaluation metrics, example performance-prediction results, and describe the design of a “universal” GET instrument intended to support the full range of verification scenarios envisioned by the IAEA.

  10. Thermal analysis of dry concrete canister storage system for CANDU spent fuel

    International Nuclear Information System (INIS)

    Ryu, Yong Ho

    1992-02-01

    This paper presents the results of a thermal analysis of the concrete canisters for interim dry storage of spent, irradiated Canadian Deuterium Uranium(CANDU) fuel. The canisters are designed to contain 6-year-old fuel safely for periods of 50 years in stainless steel baskets sealed inside a steel-lined concrete shield. In order to assure fuel integrity during the storage, fuel rod temperature shall not exceed the temperature limit. The contents of thermal analysis include the following : 1) Steady state temperature distributions under the conservative ambient temperature and insolation load. 2) Transient temperature distributions under the changes in ambient temperature and insolation load. Accounting for the coupled heat transfer modes of conduction, convection, and radiation, the computer code HEATING5 was used to predict the thermal response of the canister storage system. As HEATING5 does not have the modeling capability to compute radiation heat transfer on a rod-to-rod basis, a separate calculating routine was developed and applied to predict temperature distribution in a fuel bundle. Thermal behavior of the canister is characterized by the large thermal mass of the concrete and radiative heat transfer within the basket. The calculated results for the worst case (steady state with maximum ambient temperature and design insolation load) indicated that the maximum temperature of the 6 year cooled fuel reached to 182.4 .deg. C, slightly above the temperature limit of 180 .deg. C. However,the thermal inertia of the thick concrete wall moderates the internal changes and prevents a rise in fuel temperature in response to ambient changes. The maximum extent of the transient zone was less than 75% of the concrete wall thickness for cyclic insolation changes. When transient nature of ambient temperature and insolation load are considered, the fuel temperature will be a function of the long term ambient temperature as opposed to daily extremes. The worst design

  11. English medium of instruction: A situation analysis | Uys | South ...

    African Journals Online (AJOL)

    The majority of learners in southern Africa receive their education through the medium of a second language, English. Although teachers of English play a crucial role in helping learners to acquire language skills in the medium of instruction, we argue that subject content teachers' lack of attention to the teaching of the four ...

  12. Isogeometric frictionless contact analysis with the third medium method

    Science.gov (United States)

    Kruse, R.; Nguyen-Thanh, N.; Wriggers, P.; De Lorenzis, L.

    2018-01-01

    This paper presents an isogeometric formulation for frictionless contact between deformable bodies, based on the recently proposed concept of the third medium. This concept relies on continuum formulations not only for the contacting bodies but also for a fictitious intermediate medium in which the bodies can move and interact. Key to the formulation is a suitable definition of the constitutive behavior of the third medium. In this work, based on a number of numerical tests, the role of the material parameters of the third medium is systematically assessed. We also assess the rate of spatial convergence for higher-order discretizations, stemming from the regularization of the non-smooth contact problem inherent to the third medium approach. Finally, problems with self contact are considered and turn out to be an attractive application of the method.

  13. Thermal-hydraulic experiment and analysis for interim dry storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Yoo, Seung Hun

    2011-02-01

    The experimental and numerical studies of interim storages for nuclear spent fuels have been performed to investigate thermal-hydraulic characteristics of the dry storage systems and to propose new methodologies for the analysis and the design. Three separate researches have been performed in the present study: (a) Development of a scaling methodology and thermal-hydraulic experiment of a single spent fuel assembly simulating a dry storage cask: (b) Full-scope simulation of a dry storage cask by the use of Computational Fluid Dynamics (CFD) code: (c) Thermal-hydraulic design of a tunnel-type interim storage facility. In the first study, a scaling methodology has been developed to design a scaled-down canister. The scaling was performed in two steps. For the first step, the height of a spent fuel assembly was reduced from full height to half height. In order to consider the effect of height reduction on the natural convection, the scaling law of Ishii and Kataoka (1984) was employed. For the second step, the quantity of spent fuel assemblies was reduced from multiple assemblies to a single assembly. The scaling methodology was validated through the comparison with the experiment of the TN24P cask. The Peak Cladding Temperature (PCT), temperature gradients, and the axial and radial temperature distribution in the nondimensional forms were in good agreement with the experimental data. Based on the developed methodology, we have performed a single assembly experiment which was designed to simulate the full scale of the TN24P cask. The experimental data was compared with the CFD calculations. It turns out that their PCTs were less than the maximum allowable temperature for the fuel cladding and that the differences of their PCTs were agreed within 3 .deg. C, which was less than measurement uncertainty. In the second study, the full-scope simulations of the TN24P cask were performed by FLUENT. In order to investigate the sensitivity of the numerical and physical

  14. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1991-08-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications

  15. Measurement and Analysis of Gamma-Rays Emitted From Spent Nuclear Fuel Above 3 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, Douglas C.; Anderson, Elaina R.; Anderson, Kevin K.; Campbell, Luke W.; Fast, James E.; Jarman, Kenneth D.; Kulisek, Jonathan A.; Orton, Christopher R.; Runkle, Robert C.; Stave, Sean C.

    2013-12-01

    The gamma-ray spectrum of spent nuclear fuel in the 3- to 6-MeV energy range is important for active interrogation since emitted gamma rays emitted from nuclear decay are not expected to interfere with measurements in this energy region. There is, unfortunately, a dearth of empirical measurements from spent nuclear fuel in this region. This work is an initial attempt to partially ll this gap by presenting an analysis of gamma-ray spectra collected from a set of spent nuclear fuel sources using a high-purity germanium detector array. This multi-crystal array possesses a large collection volume, providing high energy resolution up to 16 MeV. The results of these measurements establish the continuum count-rate in the energy region between 3- and 6-MeV. Also assessed is the potential for peaks from passive emissions to interfere with peak measurements resulting from active interrogation delayed emissions. As one of the first documented empirical measurements of passive emissions from spent fuel for energies above 3 MeV, this work provides a foundation for active interrogation model validation and detector development.

  16. Hydrodemetallation and Hydrodesulfurization Spent Catalysts Elemental Analysis: Comparison of Wavelength Dispersive X-ray Fluorescence and Atomic Emission Spectrometries.

    Science.gov (United States)

    Garoux, Laetitia; Gourhand, Sébastien; Hébrant, Marc; Schneider, Michel; Diliberto, Sébastien; Meux, Eric

    2017-08-01

    Petroleum industries continuously consume catalysts on very large scales. The recycling of spent catalysts is thus of major economic and environmental importance and its first step consists of the characterization of the valuable metal content. Wavelength dispersive X-ray fluorescence (WDXRF) analysis is compared with inductively coupled plasma atomic emission spectrometry (ICP-AES) for the analysis of five samples of spent hydrodesulphurization (HDS) and hydrodemetallization (HDM) catalysts. The elements are considered for their economic interest (Co, Ni, Mo, and V) or for the problems that can arise when they are present in the sample in significant quantities (Al, As, P, Fe). First, the systematic comparison of the analysis of known synthetic samples was performed. The originality here is that the samples were first beaded with lithium tetraborate (Li 2 B 4 O 7 ) for WDXRF analysis and then dissolved in hot HCl 6M for ICP-AES measurements. With this processing, we were able to clearly identify the origin of analytical problems when they arose. Second, the semi-quantitative protocol of WDXRF is compared with the quantitative procedure. Finally, the analysis of the spent catalysts is presented and the information gained by the systematic comparison of ICP-AES and WDXRF is shared. The interest of the simultaneous determination by the two techniques when such complicated heterogeneous matrices are involved is clearly demonstrated.

  17. Spent fuel performance data: An analysis of data relevant to the NNWSI Project

    International Nuclear Information System (INIS)

    Oversby, V.M.; Shaw, H.F.

    1987-08-01

    This paper summarizes the physical and chemical properties of spent light water reactor fuel that might influence its performance as a waste form under geologic disposal conditions at Yucca Mountain, Nevada. Results obtained on the dissolution testing of spent fuel conducted by the NNWSI Project are presented and discussed. Work published by other programs, in particular those of Canada and Sweden, are reviewed and compared with the NNWSI testing results. An attempt is made to relate all of the results to a common basis of presentation and to rationalize apparent conflicts between sets of results obtained under different experimental conditions

  18. The probabilistic risk analysis of external hazards of an interim storage for spent nuclear fuel in Olkiluoto

    International Nuclear Information System (INIS)

    Puukka, Tiia

    2014-01-01

    Due to natural disasters occurred in the world and the experiences perceived of the Fukushima nuclear accident, the particular knowledge of the role and influence of external hazards in the safety of interim storage of spent nuclear fuel has been emphasized. For that reason it is substantial that they are included in the probabilistic risk assessment (PRA) of the interim storage facility. This is also required by the Regulatory Guides issued by The Finnish Radiation and Nuclear Safety Authority STUK. To enhance safety culture and nuclear safety in Olkiluoto, The Finnish utility Teollisuuden Voima Oyj has recently completed an analysis of external natural (seismic events are studied as a separate analysis) and unintentional human-induced risks associated with the spent fuel pool cooling and decay heat removal systems as part of the full-scope PRA study for the interim storage of spent fuel (KPA store). The analysis had four goals to achieve: (1) to determine the definition of an initiating event in the context of the KPA store, (2) to identify all potential external hazards and hazard combinations, (3) to perform a qualitative screening analysis based on frequency-strength analysis and detailed plant responses analysis and (4) to model the hazards passed the screening analysis so that model can be used as a risk analysis tool in the risk informed decision making and operating procedures. The assessment carried out included the analysis of operation procedures of decay heat removal, the study of external hazards related initiating events included in the PRA of the OL1 and OL2 nuclear power plants and their dependencies on the initiating events of the KPA store. All external hazards related initiating events were modeled using fault tree linking method. The main result and conclusion of this study was that using the screening analysis, initiating events caused by external hazards that could lead to leakage of the spent fuel pools or that could pose a threat to the

  19. Encapsulation technology of MR6 spent fuel and quality analysis of the EK-10 and WWR-SM spent fuel stored more than 30 years in wet conditions

    International Nuclear Information System (INIS)

    Borek-Kruszewska, E.; Bykowski, W.; Chwaszczewski, S.; Czajkowski, W.; Madry, M.

    2002-01-01

    The research reactor MARIA has been in operation for more than twenty years and all the spent fuel assemblies used since the first commissioning of the reactor are stored in wet facility on site. The present paper deals with the spent fuel MR-6 encapsulation technology in MARIA reactor. The encapsulated spent MR-6 fuel will be stored under water in the same pool unless some other solution is available. The capsules made of stainless steel are capable to accommodate one MR-6 fuel assembly. The encapsulation process is performed in the hot cell by the MARIA reactor. The spent fuel having its leg cut off is loaded to the transport cylinder manually and next transferred to a trolley. The trolley is moving to a position directly below the entrance to the hot cell and the spent fuel is entering the hot cell. The spent fuel assembly is then put into the drying cell. Dried out spent fuel is moved into the capsule mounted on the grip of the machine. Next, the capsule lid is pressed in and welded. After the leak test and filling up with helium the capsule returns from the hot cell to the pool. The hermetic capsule is sunk back into the water and positioned in the separator . The results presented earlier show, that the limiting time of WWR-SM and Ek-10 type spent fuel residence in wet storage is about 40-45 years. Therefore, the systematic quality investigation of all Ek-10 fuel elements and WWR-SM fuel assemblies discharged from EWA reactor in the period of 1959-1969 was performed. Altogether, about 2500 Ek-10 fuel elements and 47 WWR-SM fuel assemblies were investigated. The results of these investigations are presented in the present work. The sipping test, visual investigation and ultrasonic techniques were used for that purpose. The radioactive isotope Cs-137 was used as the indicator of fission product release from the fuel assembly. Taking into account the value of Cs-137 release from damaged WWR-SM fuel assembly the criteria of damaged fuel assembly were proposed. It

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

    International Nuclear Information System (INIS)

    Henrikson, D.J.

    1999-01-01

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

  1. Neutron Resonance Transmission Analysis (NRTA): Initial Studies of a Method for Assaying Plutonium in Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    David L. Chichester; James W. Sterbentz

    2011-05-01

    Neutron Resonance Transmission Analysis (NRTA) is an analytical technique that uses neutrons to assay the isotopic content of bulk materials. The technique uses a pulsed accelerator to produce an intense, short pulse of neutrons in a time-of-flight configuration. These neutrons, traveling at different speeds according to their energy, can be used to interrogate a spent fuel (SF) assembly to determine its plutonium content. Neutron transmission through the assembly is monitored as a function of neutron energy (time after the pulse), similar to the way neutron cross-section data is often collected. The transmitted neutron intensity is recorded as a function of time, with faster (higher-energy) neutrons arriving first and slower (lower-energy) neutrons arriving later. The low-energy elastic scattering and absorption resonances of plutonium and other isotopes modulate the transmitted neutron spectrum. Plutonium content in SF can be determined by analyzing this attenuation. Work is currently underway at Idaho National Laboratory, as a part of United States Department of Energy's Next Generation Safeguards Initiative (NGSI), to investigate the NRTA technique and to assess its feasibility for quantifying the plutonium content in SF and for determining the diversion of SF pins from assemblies. Preliminary results indicate that NRTA has great potential for being able to assay intact SF assemblies. Operating in the 1-40 eV range, it can identify four plutonium isotopes (239, 240, 241, & 242Pu), three uranium isotopes (235, 236, & 238U), and six resonant fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm). It can determine the areal density or mass of these isotopes in single- or multiple-pin integral transmission scans. Further, multiple observables exist to allow the detection of material diversion (pin defects) including fast-neutron and x-ray radiography, gross-transmission neutron counting, plutonium resonance absorption analysis, and fission

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

    International Nuclear Information System (INIS)

    1984-09-01

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

  3. Metallurgical analysis of a 304L stainless steel canister from the Spent Fuel Test - Climax

    International Nuclear Information System (INIS)

    Weiss, H.; Van Konynenburg, R.A.; McCright, R.D.

    1985-01-01

    Results of a metallurgical examination of a type 304L stainless steel canister that had been used to store spent nuclear fuel in an underground granite formation for about three years are reported. No observable corrosion or cracking were found. The results are applied to waste packages in a potential high level nuclear waste repository in tuff. 10 refs., 9 figs., 2 tabs

  4. Analysis of near-term spent fuel transportation hardware requirements and transportation costs

    International Nuclear Information System (INIS)

    Daling, P.M.; Engel, R.L.

    1983-01-01

    A computer model was developed to quantify the transportation hardware requirements and transportation costs associated with shipping spent fuel in the commercial nucler fuel cycle in the near future. Results from this study indicate that alternative spent fuel shipping systems (consolidated or disassembled fuel elements and new casks designed for older fuel) will significantly reduce the transportation hardware requirements and costs for shipping spent fuel in the commercial nuclear fuel cycle, if there is no significant change in their operating/handling characteristics. It was also found that a more modest cost reduction results from increasing the fraction of spent fuel shipped by truck from 25% to 50%. Larger transportation cost reductions could be realized with further increases in the truck shipping fraction. Using the given set of assumptions, it was found that the existing spent fuel cask fleet size is generally adequate to perform the needed transportation services until a fuel reprocessing plant (FRP) begins to receive fuel (assumed in 1987). Once the FRP opens, up to 7 additional truck systems and 16 additional rail systems are required at the reference truck shipping fraction of 25%. For the 50% truck shipping fraction, 17 additional truck systems and 9 additional rail systems are required. If consolidated fuel only is shipped (25% by truck), 5 additional rail casks are required and the current truck cask fleet is more than adequate until at least 1995. Changes in assumptions could affect the results. Transportation costs for a federal interim storage program could total about $25M if the FRP begins receiving fuel in 1987 or about $95M if the FRP is delayed until 1989. This is due to an increased utilization of federal interim storage facility from 350 MTU for the reference scenario to about 750 MTU if reprocessing is delayed by two years

  5. Analysis of near-term spent fuel transportation hardware requirements and transportation costs

    Energy Technology Data Exchange (ETDEWEB)

    Daling, P.M.; Engel, R.L.

    1983-01-01

    A computer model was developed to quantify the transportation hardware requirements and transportation costs associated with shipping spent fuel in the commercial nucler fuel cycle in the near future. Results from this study indicate that alternative spent fuel shipping systems (consolidated or disassembled fuel elements and new casks designed for older fuel) will significantly reduce the transportation hardware requirements and costs for shipping spent fuel in the commercial nuclear fuel cycle, if there is no significant change in their operating/handling characteristics. It was also found that a more modest cost reduction results from increasing the fraction of spent fuel shipped by truck from 25% to 50%. Larger transportation cost reductions could be realized with further increases in the truck shipping fraction. Using the given set of assumptions, it was found that the existing spent fuel cask fleet size is generally adequate to perform the needed transportation services until a fuel reprocessing plant (FRP) begins to receive fuel (assumed in 1987). Once the FRP opens, up to 7 additional truck systems and 16 additional rail systems are required at the reference truck shipping fraction of 25%. For the 50% truck shipping fraction, 17 additional truck systems and 9 additional rail systems are required. If consolidated fuel only is shipped (25% by truck), 5 additional rail casks are required and the current truck cask fleet is more than adequate until at least 1995. Changes in assumptions could affect the results. Transportation costs for a federal interim storage program could total about $25M if the FRP begins receiving fuel in 1987 or about $95M if the FRP is delayed until 1989. This is due to an increased utilization of federal interim storage facility from 350 MTU for the reference scenario to about 750 MTU if reprocessing is delayed by two years.

  6. Linear analysis of active-medium two-beam accelerator

    Directory of Open Access Journals (Sweden)

    Miron Voin

    2015-07-01

    Full Text Available We present detailed development of the linear theory of wakefield amplification by active medium and its possible application to a two-beam accelerator (TBA is discussed. A relativistic train of triggering microbunches traveling along a vacuum channel in an active medium confined by a cylindrical waveguide excites Cherenkov wake in the medium. The wake is a superposition of azimuthally symmetric transverse magnetic modes propagating along a confining waveguide, with a phase velocity equal to the velocity of the triggering bunches. The structure may be designed in such a way that the frequency of one of the modes is close to active-medium resonant frequency, resulting in amplification of the former and domination of a single mode far behind the trigger bunches. Another electron bunch placed in proper phase with the amplified wakefield may be accelerated by the latter. Importantly, the energy for acceleration is provided by the active medium and not the drive bunch as in a traditional TBA. Based on a simplified model, we analyze extensively the impact of various parameters on the wakefield amplification process.

  7. Combined Effects of Time Spent in Physical Activity, Sedentary Behaviors and Sleep on Obesity and Cardio-Metabolic Health Markers: A Novel Compositional Data Analysis Approach.

    Science.gov (United States)

    Chastin, Sebastien F M; Palarea-Albaladejo, Javier; Dontje, Manon L; Skelton, Dawn A

    2015-01-01

    The associations between time spent in sleep, sedentary behaviors (SB) and physical activity with health are usually studied without taking into account that time is finite during the day, so time spent in each of these behaviors are codependent. Therefore, little is known about the combined effect of time spent in sleep, SB and physical activity, that together constitute a composite whole, on obesity and cardio-metabolic health markers. Cross-sectional analysis of NHANES 2005-6 cycle on N = 1937 adults, was undertaken using a compositional analysis paradigm, which accounts for this intrinsic codependence. Time spent in SB, light intensity (LIPA) and moderate to vigorous activity (MVPA) was determined from accelerometry and combined with self-reported sleep time to obtain the 24 hour time budget composition. The distribution of time spent in sleep, SB, LIPA and MVPA is significantly associated with BMI, waist circumference, triglycerides, plasma glucose, plasma insulin (all pblood pressure (pphysical inactivity.

  8. Heat transfer analysis of the geologic disposal of spent fuel and high-level waste storage canisters

    International Nuclear Information System (INIS)

    Allen, G.K.

    1980-08-01

    Near-field temperatures resulting from the storage of high-level waste canisters and spent unreprocessed fuel assembly canisters in geologic formations were determined. Preliminary design of the repository was modeled for a heat transfer computer code, HEATING5, which used the Crank-Nicolson finite difference method to evaluate transient heat transfer. The heat transfer system was evaluated with several two- and three-dimensional models which transfer heat by a combination of conduction, natural convention, and radiation. Physical properties of the materials in the model were based upon experimental values for the various geologic formations. The effects of canister spacing, fuel age, and use of an overpack were studied for the analysis of the spent fuel canisters; salt, granite, and basalt were considered as the storage media for spent fuel canisters. The effects of canister diameter and use of an overpack were studied for the analysis of the high-level waste canisters; salt was considered as the only storage media for high-level waste canisters. Results of the studies on spent fuel assembly canisters showed that the canisters could be stored in salt formations with a maximum heat loading of 134 kw/acre without exceeding the temperature limits set for salt stability. The use of an overpack had little effect on the peak canister temperatures. When the total heat load per acre decreased, the peak temperatures reached in the geologic formations decreased; however, the time to reach the peak temperatures increased. Results of the studies on high-level waste canisters showed that an increased canister diameter will increase the canister interior temperatures considerably; at a constant areal heat loading, a 381 mm diameter canister reached almost a 50 0 C higher temperature than a 305 mm diameter canister. An overpacked canister caused almost a 30 0 C temperature rise in either case

  9. Deep geological disposal system development; thermal stress analysis and nonlinear structural analysis of spent nuclear fuel disposal canister under sudden rock movement

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Joo; Kim, Jin An; Ha, Jun Yong [Hongik University, Seoul (Korea)

    2002-04-01

    This work constitutes a summary of research and development made for design and dimensioning of the spent nuclear fuel disposal canister. Since the spent nuclear fuel disposal emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for the spent nuclear fuel disposal should be secured. Usually this repository is expected to locate at a depth of 500m underground. In this work the thermal stress analysis of the spent nuclear fuel disposal canister in a deep repository at 500m underground is performed for the underground pressure variation. Thermal stresses of the canister due to thermal loads of the heat generation of spent nuclear fuels inside baskets are computed. The thermal stress analysis result shows that even though some high thermal stresses occur due to the heat generation of nuclear fuels inside baskets, the canister is still structurally safe because the maximum stress occurred in the canister is smaller than the yield strength of the cast iron. In this work, the nonlinear structural analysis for the composite structure of the spent nuclear fuel disposal canister and the 50cm thick bentonite buffer is also carried out to predict the collapse of the canister while the sudden rock movement of 10cm is applied on the composite structure. Elastoplastic material model is adopted. Drucker-Prager yield criterion is used for the material yield prediction of the bentonite buffer and von-Mises yield criterion is used for the material yield prediction of the canister(cast iron insert, copper outer shell and lid and bottom). The analysis result shows that even though very large deformations occur beyond the yield point in the bentonite buffer, the canister structure still endures elastic small strains and stresses below the yield strength. Analysis results also show that bending deformations occur in the canister structure due to the shear deformation of the bentonite buffer. 24

  10. Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

    International Nuclear Information System (INIS)

    Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

    1999-01-01

    The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses

  11. Sensitivity Analysis of Depletion Parameters for Heat Load Evaluation of PWR Spent Fuel Storage Pool

    International Nuclear Information System (INIS)

    Kim, In Young; Lee, Un Chul

    2011-01-01

    As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

  12. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Gauld, Ian C [ORNL

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  13. Removal of cesium and separation of strontium the analysis of the leachate of spent fuel

    International Nuclear Information System (INIS)

    Kim, Seung Soo; Chun, Kwan Sik; Kang, Chul Hyung

    2002-01-01

    The selective removal of cesium by ammonium molybdophosphate (AMP) was studied in order to reduce an interference by high radioactivity of cesium on the determination of low radioactive elements in leachate of spent fuel. The removal of Cs, U, Ce, La, Co, Na Sr and K was investigated for the leachate and the bentonite in contact with a spent fuel. More than 90% of cesium was removed by AMP and Ca, Na, Co and Sr was remained in 0.1M HNO 3 . However, three valence elements such as La and Ce were also removed by AMP. Though a little of potassium of the bentonite components was adsorbed on AMP, the potassium in the bentonite solution diluted to its concentration in a real sample would not affect the capacity of AMP greatly. From another experiment for the separation of strontium as a leaching indicator of spent fuel, the recovery of strontium in 8.0 M HNO 3 solution by using Sr-resin (Eichrom, P/N SR-B50-A) was more than 95% by eluting with 0.05 M HNO 3

  14. An Analysis of Medium Loss of Coolant Sequence for the Severe Accident Analysis DB

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Song, Yong Mann

    2007-12-15

    This report contains analysis methodologies and calculation results of medium loss of Coolant sequences for the severe accident analysis database system. The Korean standard nuclear power plant has been selected as a reference plant. Based on the probabilistic safety analysis of the corresponding plant, 10 accident scenarios, which was predicted to have more than 10{sup -10} /ry occurrence frequency, have been analyzed as base cases for the medium loss of Coolant sequence database. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results in this report will be utilized as input data of the severe accident analysis database system.

  15. Internal And External Environment Analysis On The Performance Of Small And Medium Industries Smes In Indonesia

    OpenAIRE

    Sofyan Indris; Ina Primiana

    2015-01-01

    Abstract The purpose of this study was to determine the influence of internal and external environment analysis on the performance of small and medium industries SMEs in Indonesia. The theoretical results showed that internal and external environment analysis have a significant effect on the performance of small and medium industries SMEs in Indonesia.

  16. Analysis of photonic band-gap structures in stratified medium

    DEFF Research Database (Denmark)

    Tong, Ming-Sze; Yinchao, Chen; Lu, Yilong

    2005-01-01

    Purpose - To demonstrate the flexibility and advantages of a non-uniform pseudo-spectral time domain (nu-PSTD) method through studies of the wave propagation characteristics on photonic band-gap (PBG) structures in stratified medium Design/methodology/approach - A nu-PSTD method is proposed...... in solving the Maxwell's equations numerically. It expands the temporal derivatives using the finite differences, while it adopts the Fourier transform (FT) properties to expand the spatial derivatives in Maxwell's equations. In addition, the method makes use of the chain-rule property in calculus together...... with the transformed space technique in order to make the algorithm flexible in terms of non-uniform spatial sampling. Findings - Through the studies of the wave propagation characteristics on PBG structures in stratified medium, it has been found that the proposed method retains excellent accuracy in the occasions...

  17. Basic Considerations for Dry Storage of Spent Nuclear Fuels and Revisited CFD Thermal Analysis on the Concrete Cask

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Jae Soo [ACT Co. Ltd., Daejeon (Korea, Republic of); Park, Younwon; Song, Sub Lee [BEES Inc., Daejeon (Korea, Republic of); Kim, Hyeun Min [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    The integrity of storage facility and also of the spent nuclear fuel itself is considered very important. Storage casks can be located in a designated area on a site or in a designated storage building. A number of different designs for dry storage have been developed and used in different countries. Dry storage system was classified into two categories by IAEA. One is container including cask and silo, the other one is vault. However, there is various way of categorization for dry storage system. Dry silo and cask are usually classified separately, so the dry storage system can be classified into three different types. Furthermore, dry cask storage can be categorized into two types based on the type of the materials, concrete cask and metal cask. In this paper, the design characteristics of dry storage cask are introduced and computational fluid dynamics (CFD) based thermal analysis for concrete cask is revisited. Basic principles for dry storage cask design were described. Based on that, thermal analysis of concrete dry cask was introduced from the study of H. M. Kim et al. From the CFD calculation, the temperature of concrete wall was maintained under the safety criteria. From this fundamental analysis, further investigations are expected. For example, thermal analysis on the metal cask, thermal analysis on horizontally laid spent nuclear fuel assemblies for transportation concerns, and investigations on better performance of natural air circulation in dry cask can be promising candidates.

  18. Development of a Reliable Fuel Depletion Methodology for the HTR-10 Spent Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Kiwhan [Los Alamos National Laboratory; Beddingfield, David H. [Los Alamos National Laboratory; Geist, William H. [Los Alamos National Laboratory; Lee, Sang-Yoon [unaffiliated

    2012-07-03

    A technical working group formed in 2007 between NNSA and CAEA to develop a reliable fuel depletion method for HTR-10 based on MCNPX and to analyze the isotopic inventory and radiation source terms of the HTR-10 spent fuel. Conclusions of this presentation are: (1) Established a fuel depletion methodology and demonstrated its safeguards application; (2) Proliferation resistant at high discharge burnup ({approx}80 GWD/MtHM) - Unfavorable isotopics, high number of pebbles needed, harder to reprocess pebbles; (3) SF should remain under safeguards comparable to that of LWR; and (4) Diversion scenarios not considered, but can be performed.

  19. An analysis of the technical status of high level radioactive waste and spent fuel management systems

    Science.gov (United States)

    English, T.; Miller, C.; Bullard, E.; Campbell, R.; Chockie, A.; Divita, E.; Douthitt, C.; Edelson, E.; Lees, L.

    1977-01-01

    The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.

  20. An analysis of the technical status of high level radioactive waste and spent fuel management systems

    Energy Technology Data Exchange (ETDEWEB)

    English, T.; Miller, C.; Bullard, E.; Campbell, R.; Chockie, A.

    1977-12-01

    The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.

  1. Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository - Volume 3: Appendices

    International Nuclear Information System (INIS)

    Taylor, L.L.; Wilson, J.R.; Sanchez, L.C.; Aguilar, R.; Trellue, H.R.; Cochrane, K.; Rath, J.S.

    1998-01-01

    The United States Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3)

  2. A comparative analysis of pretreatment strategies on the properties and hydrolysis of brewers' spent grain.

    Science.gov (United States)

    Ravindran, Rajeev; Jaiswal, Swarna; Abu-Ghannam, Nissreen; Jaiswal, Amit K

    2018-01-01

    In this study, brewer's spent grain (BSG) was subjected to a range pretreatments to study the effect on reducing sugar yield. Glucose and xylose were found to be the predominant sugars in BSG. Brewers spent grain was high in cellulose (19.21g/100g of BSG) and lignin content (30.84g/100g of BSG). Microwave assisted alkali (MAA) pretreatment was found to be the most effective pretreatment for BSG, where the pretreatment was conducted at 400W for 60s. A maximum reducing yield was observed with high biomass loading (1g/10ml), cellulase (158.76μl/10ml), hemicellulase (153.3μl/10ml), pH (5.4) and an incubation time (120h). Upon enzymatic hydrolysis, MAA pretreated BSG yielded 228.25mg of reducing sugar/g of BSG which was 2.86-fold higher compared to native BSG (79.67mg/g of BSG); simultaneously BSG was de-lignified significantly. The changes in functional groups, crystallinity and thermal behaviour was studies by means of FTIR, XRD and DSC, respectively. Copyright © 2017 Elsevier Ltd. All rights reserved.

  3. Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G.; Morrow, Charles.

    2013-01-01

    The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

  4. Analysis of photonic band-gap structures in stratified medium

    DEFF Research Database (Denmark)

    Tong, Ming-Sze; Yinchao, Chen; Lu, Yilong

    2005-01-01

    in solving the Maxwell's equations numerically. It expands the temporal derivatives using the finite differences, while it adopts the Fourier transform (FT) properties to expand the spatial derivatives in Maxwell's equations. In addition, the method makes use of the chain-rule property in calculus together...... in electromagnetic and microwave applications once the Maxwell's equations are appropriately modeled. Originality/value - The method validates its values and properties through extensive studies on regular and defective 1D PBG structures in stratified medium, and it can be further extended to solving more...

  5. Regular Advisory Group on Spent Fuel Management

    International Nuclear Information System (INIS)

    1993-01-01

    The Regular Advisory Group on Spent Fuel Management (RAGSFM) was established in accordance with the recommendations of the Expert Group on International Spent Fuel Management in 1982. The Advisory Group consists of nominated experts from countries with considerable experience and/or requirements in such aspects of the back-end of the fuel cycle as storage, safety, transportation and treatment of spent fuel. The RAGSFM activities cover the following main topics: a) Analysis and summary of spent fuel arisings and storage facilities; b) Interface between spent fuel storage and transportation activities; c) Spent fuel storage process and technology and related safety issues; d)Treatment of spent fuel

  6. Analysis of transportation and handling system for advanced spent fuel management process

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Dong Hee; Yoon, J. S.; Park, B. S.; Ahn, S. H.; Kim, Y. H.; Jung, J. H.; Jin, J. H.; Park, G. Y.; Song, T. G. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    In this report, several devices which are used to safely transport and handle nuclear materials without scattering have been derived by analyzing the Advanced Management Process, object nuclear material and modules of process equipment and performing graphical simulation of transportation/handling by computers for the demonstration of the Advanced Spent Fuel Management Process. For verification, powder transportation vessel and handling device have been designed and manufactured. And several tests such as transporting, grappling, rotating the vessel have been performed. Also, the functional requirements of transportation/handling equipment have been analyzed based on test results and process studies. The developed functional requirements in this research will be used as the design data for the Advanced Management Process. 6 refs., 25 figs., 6 tabs. (Author)

  7. Disposal of spent fuel

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  8. A geographic information system and multi criteria analysis method for site selection of spent nuclear fuel disposal

    International Nuclear Information System (INIS)

    Martins, Vivian Borges

    2009-01-01

    This thesis aims to develop a site selection methodology for the construction of final repository for the spent nuclear fuel disposal, by using geographic information systems (GIS) and multi-criteria decision analysis. Decision making processes of this kind are often complex, given the great number of space parameters to consider and also the typically conflicting opinions of the diverse stake holders. By using GIS, data from different space parameters can be quickly and reliably stored, treated and analyzed. Multi-criteria techniques allow for the incorporation of different stake holders' opinions. These tools, when jointly used, allow for the decision process to be more transparent, quick and reliable. The method developed was applied to the particular case of the state of Rio de Janeiro. Weights obtained from an expert panel and also by using the Hierarchical Analysis Method and cartographic data were combined in the GIS. The application showed that it is possible not only to select and classify areas as to their aptness for the proposed objective, but also to exclude those clearly inadequate areas, thus optimizing the selection process by reducing the search space and consequently minimizing costs and the time spent in the search. (author)

  9. Analysis of Bacteriophage Motion Through a Non-Static Medium

    Science.gov (United States)

    Dickey, Samuel A.

    In this work, I investigated the motion of bacteriophages (phages) through their mucosal environment. Recently, biologists here at San Diego State University have proposed a model in which phages move sub-diffusively through mucosal fibers in their hunt for bacteria to prey upon. Through a Hoc protein located upon the capsid of the wild type phages, these phages are allowed to bind to mucosal fibers, and extend the amount of time spent in a single location hunting for bacteria. Contrarily, the delta hoc phages are unable to. The ability of the wild type phages to attach itself to mucosal fibers is what enables its subdiffusive behavior. This study investigates the diffusive behavior of these phages in different mucus concentrations. It expands on previous studies in which only short tracks could be observed. In the study at hand, phages are imaged in a highly doped optical fiber with varying concentrations of mucus present in solution. Through rigorous image processing techniques, trajectories of these phages are created with a minimized noise level. We developed code that created position-versus-time files for each phage present in the experimental data. These files were then further analyzed. The sub-diffusive behavior is investigated via mean squared displacement versus time. The diffusive exponent can be obtained from fits to these data. For large enough time intervals, I always obtained an exponent of one for space and time averaged data. This indicates that the diffusion is normal, or sub-diffusive of the CTRW type. CTRW sub-diffusive motion is characterized by waiting times that resemble a power law distribution and have long tails. I investigate these stuck time distributions, however am unable to determine if a power law or exponential fits the data best. Moreover, the distribution gives the same power law exponent for phages moving through water, or mucus, for wild type and delta hoc phages. These exponents would predict super-diffusive instead of sub

  10. Medium Duty ARRA Data Reporting and Analysis (Presentation)

    Energy Technology Data Exchange (ETDEWEB)

    Walkowicz, K.

    2014-06-01

    This project compiles medium-duty (MD) aggregated deployment data and provides the compiled detailed analyses to industry. The U.S. Department of Energy's (DOE's) American Recovery and Reinvestment Act (ARRA) deployment and demonstration projects are helping to commercialize technologies for all-electric vehicles, electrified accessories, and electric charging infrastructure. Over 3.2 million miles of in-service all-electric MD truck data from 560 different vehicles have been collected since 2011, and usage data from over 1,000 truck electrification sites have been collected since 2013. Through the DOE's Vehicle Technologies Office, NREL is working to analyze real-time data from these deployment and demonstration projects to quantify the benefits: results and summary statistics are made available through the NREL website as quarterly and annual reports; 23 aggregated reports have been published on the performance and operation of these vehicles; and detailed data are being extracted to help further understand battery use and performance.

  11. Measuring the noble metal and iodine composition of extracted noble metal phase from spent nuclear fuel using instrumental neutron activation analysis

    International Nuclear Information System (INIS)

    Palomares, R.I.; Dayman, K.J.; Landsberger, S.; Biegalski, S.R.; Soderquist, C.Z.; Casella, A.J.; Brady Raap, M.C.; Schwantes, J.M.

    2015-01-01

    Masses of noble metal and iodine nuclides in the metallic noble metal phase extracted from spent fuel are measured using instrumental neutron activation analysis. Nuclide presence is predicted using fission yield analysis, and radionuclides are identified and the masses quantified using neutron activation analysis. The nuclide compositions of noble metal phase derived from two dissolution methods, UO 2 fuel dissolved in nitric acid and UO 2 fuel dissolved in ammonium-carbonate and hydrogen-peroxide solution, are compared. - Highlights: • The noble metal phase was chemically extracted from spent nuclear fuel and analyzed non-destructively. • Noble metal phase nuclides and long-lived iodine were identified and quantified using neutron activation analysis. • Activation to shorter-lived radionuclides allowed rapid analysis of long-lived fission products in spent fuel using gamma spectrometry

  12. Measurement and Analysis Plan for Investigation of Spent-Fuel Assay Using Lead Slowing-Down Spectroscopy

    International Nuclear Information System (INIS)

    Smith, Leon E.; Haas, Derek A.; Gavron, Victor A.; Imel, G.R.; Ressler, Jennifer J.; Bowyer, Sonya M.; Danon, Y.; Beller, D.

    2009-01-01

    Under funding from the Department of Energy Office of Nuclear Energy's Materials, Protection, Accounting, and Control for Transmutation (MPACT) program (formerly the Advanced Fuel Cycle Initiative Safeguards Campaign), Pacific Northwest National Laboratory (PNNL) and Los Alamos National Laboratory (LANL) are collaborating to study the viability of lead slowing-down spectroscopy (LSDS) for spent-fuel assay. Based on the results of previous simulation studies conducted by PNNL and LANL to estimate potential LSDS performance, a more comprehensive study of LSDS viability has been defined. That study includes benchmarking measurements, development and testing of key enabling instrumentation, and continued study of time-spectra analysis methods. This report satisfies the requirements for a PNNL/LANL deliverable that describes the objectives, plans and contributing organizations for a comprehensive three-year study of LSDS for spent-fuel assay. This deliverable was generated largely during the LSDS workshop held on August 25-26, 2009 at Rensselaer Polytechnic Institute (RPI). The workshop itself was a prominent milestone in the FY09 MPACT project and is also described within this report.

  13. Analysis for seismic response of dry storage facility for spent fuel

    International Nuclear Information System (INIS)

    Ko, Y.-Y.; Hsu, S.-Y.; Chen, C.-H.

    2009-01-01

    Most of the dry storage systems for spent fuel are freestanding, which leads to stability concerns in an earthquake. In this study, as a safety check, the ABAQUS/Explicit code is adopted to analyse the seismic response of the dry storage facility planned to be installed at Nuclear Power Plant no. 1 (NPP1) in Taiwan. A 3D coupled finite element (FE) model was established, which consisted of a freestanding cask, a concrete pad, and underneath soils interacting with frictional contact interfaces. The scenario earthquake used in the model included an artificial earthquake compatible to the design spectrum of NPP1, and a strong ground motion modified from the time history recorded during the Chi-Chi earthquake. The results show that the freestanding cask will slide, but not tip over, during strong earthquakes. The scale of the sliding is very small and a collision between casks will not occur. In addition, the differential settlement of the foundation pad that takes place due to the weight of the casks increases the sliding potential of the casks during earthquakes

  14. Anaerobic digestion of spent mushroom substrate under thermophilic conditions: performance and microbial community analysis.

    Science.gov (United States)

    Xiao, Zheng; Lin, Manhong; Fan, Jinlin; Chen, Yixuan; Zhao, Chao; Liu, Bin

    2018-01-01

    Spent mushroom substrate (SMS) is the residue of edible mushroom production occurring in huge amounts. The SMS residue can be digested for biogas production in the mesophilic anaerobic digestion. In the present study, performance of batch thermophilic anaerobic digestion (TAD) of SMS was investigated as well as the interconnected microbial population structure changes. The analyzed batch TAD process lasted for 12 days with the cumulative methane yields of 177.69 mL/g volatile solid (VS). Hydrolytic activities of soluble sugar, crude protein, and crude fat in SMS were conducted mainly in the initial phase, accompanied by the excessive accumulation of volatile fatty acids and low methane yield. Biogas production increased dramatically from days 4 to 6. The degradation rates of cellulose and hemicellulose were 47.53 and 55.08%, respectively. The high-throughput sequencing of 16S rRNA gene amplicons revealed that Proteobacteria (56.7%-62.8%) was the dominant phylum in different fermentative stages, which was highly specific compared with other anaerobic processes of lignocellulosic materials reported in the literature. Crenarchaeota was abundant in the archaea. The most dominant genera of archaea were retrieved as Methanothermobacter and Methanobacterium, but the latter decreased sharply with time. This study shows that TAD is a feasible method to handle the waste SMS.

  15. Analysis of frame structure of medium and small truck crane

    Science.gov (United States)

    Cao, Fuyi; Li, Jinlong; Cui, Mengkai

    2018-03-01

    Truck crane is an important part of hoisting machinery. Frame, as the support component of the quality of truck crane, determines the safety of crane jib load and the rationality of structural design. In this paper, the truck crane frame is a box structure, the three-dimensional model is established in CATIA software, and imported into Hyperworks software for finite element analysis. On the base of doing constraints and loads for the finite element model of the frame, the finite element static analysis is carried out. And the static stress test verifies whether the finite element model and the frame structure design are reasonable; then the free modal analysis of the frame and the analysis of the first 8 - order modal vibration deformation are carried out. The analysis results show that the maximum stress value of the frame is greater than the yield limit value of the material, and the low-order modal value is close to the excitation frequency value, which needs to be improved to provide theoretical reference for the structural design of the truck crane frame.

  16. IMPACT ANALYSIS OF SPENT FUEL DRY CASKS UNDER ACCIDENTAL DROP SCENARIOS

    International Nuclear Information System (INIS)

    BRAVERMAN, J.I.; MORANTE, R.J.; XU, J.; HOFMAYER, C.H.; SHAUKAT, S.K.

    2003-01-01

    A series of analyses were performed to assess the structural response of spent nuclear fuel dry casks subjected to various handling and on-site transfer events. The results of these analyses are being used by the Nuclear Regulatory Commission (NRC) to perform a probabilistic risk assessment (PRA). Although the PRA study is being performed for a specific nuclear plant, the PRA study is also intended to provide a framework for a general methodology that could also be applied to other dry cask systems at other nuclear plants. The dry cask system consists of a transfer cask, used for handling and moving the multi-purpose canister OLIIpC that contains the fuel, and a storage cask, used to store the MPC and fuel on a concrete pad at the site. This paper describes the analyses of the casks for two loading events. The first loading consists of dropping the transfer cask while it is lowered by a crane to a concrete floor at ground elevation. The second loading consists of dropping the storage cask while it is being transferred to the concrete storage pad outdoors. Three dimensional finite element models of the transfer cask and storage cask, containing the MPC and fuel, were utilized to perform the drop analyses. These models were combined with finite element models of the target structures being impacted. The transfer cask drop analyses considered various drop heights for the cask impacting the reinforced concrete floor at ground level. The finite element model of the target included a section of the concrete floor and concrete wall supporting the floor. The storage cask drop analyses evaluated a 30.5 cm (12 in.) drop of the cask impacting three different surfaces: reinforced concrete, asphalt, and gravel

  17. Analysis of the transportation logistics for spent nuclear fuel in Korea

    International Nuclear Information System (INIS)

    Lee, Hyo Jik; Ko, Won Il; Seo, Ki Seok

    2010-01-01

    As a part of the back-end fuel cycle, transportation of spent nuclear fuel (SNF) from nuclear power plants (NPP s ) to a fuel storage facility is very important in establishing a nuclear fuel cycle. In Korea, the accumulated amount of SNF in the NPP pools is troublesome since the temporary storage facilities at these NPP pools are expected to be full of SNF within ten years. Therefore, Korea cannot help but plan for the construction of an interim storage facility to solve this problem in the near future. Especially, a decision on several factors, such as where the interim storage facility should be located, how many casks a transport ship can carry at a time and how many casks are initially required, affect the configuration of the transportation system. In order to analyze the various possible candidate scenarios, we assumed four cases for the interim storage facility location, three cases for the load capacity that a transport ship can carry and two cases for the total amount of casks used for transportation. First, this study considered the currently accumulated amount of SNF in Korea, and the amount of SNF generated from NPP s until all NPP s are shut down. Then, how much SNF per year must be transported from theNPP s to an interim storage facility was calculated during an assumed transportation period. Second, 24 candidate transportation scenarios were constructed by a combination of the decision factors. To construct viable yearly transportation schedules for the selected 24 scenarios, we created a spreadsheet program named TranScenario, which was developed by using MS EXCEL. TranScenario can help schedulers input shipping routes and allocate transportation casks. Also,TranScenario provides information on the cask distribution in the NPP s and in the interim storage facility automatically, by displaying it in real time according to the shipping routes, cask types and cask numbers that the user generates. Once a yearly transportation schedule is established

  18. Dynamic analysis of multibody system immersed in a fluid medium

    International Nuclear Information System (INIS)

    Wu, R.W.; Liu, L.K.; Levy, S.

    1977-01-01

    This paper is concerned primarily with the development and evaluation of an analysis method for the reponse prediction of immersed systems to seismic and other dynamic excitations. For immersed multibody systems, the hydrodynamic interaction causes coupled motion among the solid bodies. Also, under intense external excitations, impact between bodies may occur. The complex character of such systems inhibit the use of conventional analytical solutions in closed form. Therefore, approximate numerical schemes have been devised. For an incompressible, inviscid fluid, the hydrodynamic forces exerted by the fluid on solid bodies are determined to be linearly proportional to the acceleration of the vibrating solid bodies; i.e., the presence of the fluid only affects the inertia of the solid body system. A finite element computer program has been developed for computing this hydrodynamic (or added) mass effect. This program can be used to determine the hydrodynamic mass of a two-dimensional fluid field with solid bodies of arbitrary geometry. Triangular elements and linear pressure interpolation function are used to discretize the fluid region. The component element method is used to determine the dynamic response of the multibody system to externally applied mechanical loading or support excitation. The present analysis method for predicting the dynamic response of submerged multibody system is quite general and pertains to any number of solid bodies. However in this paper, its application is demonstrated only for 4 and 25 body systems. (Auth.)

  19. A medium energy facility for variable temperature implantation and analysis

    International Nuclear Information System (INIS)

    Chaumont, J.; Lalu, F.; Salome, M.; Lamoise, A.M.; Bernas, H.

    1981-01-01

    We describe the new ion implantation system at Orsay, which operates from 5 to 190 kV. Sixty-five elements from H to U have been implanted in insulators, semiconductors or metals. Significant currents (several μA) of three-fold ionized elements have been implanted at energies up to 570 keV. Details are provided on the target-holders used, particularly on a variable temperature (1.7-300 K) cryostat and a variable temperature (80-300 K) goniometer, and on an in situ Rutherford back-scattering analysis set-up (using the 380 keV He 2+ beam) used in conjunction with all these target-holders. The latter system is used for studies of metastable low-temperature implanted alloys: specific examples will be given. (orig.)

  20. Small-scale analysis of exopolysaccharides from Streptococcus thermophilus grown in a semi-defined medium

    Directory of Open Access Journals (Sweden)

    Rådström Peter

    2001-09-01

    Full Text Available Abstract Background Exopolysaccharides (EPSs produced by lactic acid bacteria are important for the texture of fermented foods and have received a great deal of interest recently. However, the low production levels of EPSs in combination with the complex media used for growth of the bacteria have caused problems in the accurate analysis of the EPS. The purpose of this study was to find a growth medium for physiological studies of the lactic acid bacterium Streptococcus thermophilus, and to develop a simple method for qualitative and quantitative analysis of EPSs produced in this medium. Results A semi-defined polysaccharide medium was developed and evaluated on six strains of Streptococcus thermophilus. The EPSs were analysed using a novel protocol incorporating ultracentrifugation for the removal of interfering sugars, hydrolysis and analysis of the monomer composition by High Performance Anion-Exchange Chromatography with pulsed amperometric detection. The medium and analysis method allowed accurate quantification and monomer analysis of 0.5 ml samples of EPSs from tube cultures. Conclusions The presented medium should be useful for physiological studies of S. thermophilus, and, in combination with the method of analysis of EPS, will allow downscaling of physiological studies and screening for EPSs.

  1. Additional EIPC Study Analysis: Interim Report on Medium Priority Topics

    Energy Technology Data Exchange (ETDEWEB)

    Hadley, Stanton W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gotham, Douglas J. [Purdue Univ., West Lafayette, IN (United States)

    2014-03-01

    Between 2010 and 2012 the Eastern Interconnection Planning Collaborative (EIPC) conducted a major long-term resource and transmission study of the Eastern Interconnection (EI). With guidance from a stakeholder steering committee (SSC) that included representatives from the Eastern Interconnection States’ Planning Council (EISPC) among others, the project was conducted in two phases. The first was a 2015–2040 analysis that looked at a broad array of possible future scenarios, while the second focused on a more detailed examination of the grid in 2030. The studies provided a wealth of information on possible future generation, demand, and transmission alternatives. However, at the conclusion there were still unresolved questions and issues. The US Department of Energy, which had sponsored the study, asked Oak Ridge National Laboratory researchers and others who worked on the project to conduct an additional study of the data to provide further insights for stakeholders and the industry. This report documents the second part of that follow-on study [an earlier report (Hadley 2013) covered the first part, and a subsequent report will address the last part].

  2. Dynamic analysis of multibody system immersed in a fluid medium

    International Nuclear Information System (INIS)

    Wu, R.W.; Liu, L.K.; Levy, S.

    1977-01-01

    This paper is concerned primarily with the development and evaluation of an analysis method for the response prediction of immersed multibody systems to seismic and other dynamic excitations. For immersed multibody systems to seismic and other dynamic excitations. For immersed multibody systems, the hydrodynamic interaction causes coupled motion among the solid bodies. Also, under intense external excitations, impact between bodies may occur. The complex character of such systems inhibit the use of conventional analytical solutions in closed form. Therefore, approximate numerical schemes have been devised. For an incompressible, inviscid fluid, the hydrodynamic forces exerted by the fluid on solid bodies are determined to be linearly proportional to the acceleration of the vibrating solid bodies; i.e., the presence of the fluid only affects the inertia of the solid body system. A finite element computer program has been developed for computing this hydrodynamic (or added) mass effect. This program can be used to determine the hydrodynamic mass of a two-dimensional fluid field with solid bodies of arbitrary geometry. Triangular elements and linear pressure interpolation function are used to discretize the fluid region. The component element method is used to determine the dynamic response of the multibody system to externally applied mechanical loading or support excitation. In the component element method, each structural body is modeled by component elements of conceptual spring-damper type. This method is particularly advantageous for systems having nonlinearities. Direct timewise numerical integration scheme is used to solve the governing dynamic equation of the system. Analytical study is carried out and compared with an experimental result for the forced vibration of 4 simply supported beams in water. Also studied is a case of 25 (5x5) beams within a square fluid-filled container by using two different approaches

  3. THE ANALYSIS OF MEASUREMENTS FOR SMALL AND MEDIUM BUSINESSES SUPPORT

    Directory of Open Access Journals (Sweden)

    E. A. Tabolova

    2015-01-01

    Full Text Available Market transformations in economic system of the Russian Federation from 1991 conditioned the change of consumers drug support organization principles, and the market formation led to the business development in this field. Pharmaceutical market is a socially significant market, which leads to the necessity of active realization of competitive business environment development in the conditions of effective state regulation and support. The problem of creation and development of small businesses has a special value for the national economy. These businesses provide the production of significant, and in some cases dominating part of a gross domestic product in many developed countries. State policy of small business support in Russia has become an individual system direction of social and economical state policy. It is formed by the principle of favorable conditions creation for small business development, especially in the fields of activity which give maximum social and economical effect. The actions coordination of federal executive authorities, executive authorities of federal subjects of the Russian Federation and local governments, social organizations, and entrepreneurs associations was enhanced to create the conditions of effective development of small business, form a flexible system of its state support. Analysis of the measurements implemented in the Republic of North Ossetia – Alania, aimed to the small business support for the evaluation of possibility of their use for pharmacy organization opening was the purpose of the study.

  4. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - IV: DUPIC Fuel Cycle Cost

    International Nuclear Information System (INIS)

    Ko, Won Il; Choi, Hangbok; Yang, Myung Seung

    2001-01-01

    This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.21 to 6.34 mills/kW.h for DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.07 to 0.27 mills/kW.h. Considering the uncertainty (0.40 to 0.44 mills/kW.h) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by ∼20% and reduce the spent fuel arising by ∼65% compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle is comparable with the once-through fuel cycle from the viewpoint of FCC. In the future, it should be important to consider factors such as the environmental benefit owing to natural uranium savings, the capability of reusing spent pressurized water reactor fuel, and the safeguardability of the fuel cycle when deciding on an advanced nuclear fuel cycle option

  5. Safety analysis methodology for Chinshan nuclear power plant spent fuel pool under Fukushima-like accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Hao-Tzu [Institute of Nuclear Energy Research, Taoyuan, Taiwan (China). Research Atomic Energy Council; Li, Wan-Yun; Wang, Jong-Rong; Tseng, Yung-Shin; Chen, Hsiung-Chih; Shih, Chunkuan; Chen, Shao-Wen [National Tsing Hua Univ., HsinChu, Taiwan (China). Inst. of Nuclear Engineering and Science

    2017-03-15

    Chinshan nuclear power plant (NPP), a BWR/4 plant, is the first NPP in Taiwan. After Fukushima NPP disaster occurred, there is more concern for the safety of NPPs in Taiwan. Therefore, in order to estimate the safety of Chinshan NPP spent fuel pool (SFP), by using TRACE, MELCOR, CFD, and FRAPTRAN codes, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of Chinshan NPP SFP. There were two main steps in this research. The first step was the establishment of Chinshan NPP SFP models. And the transient analysis under the SFP cooling system failure condition (Fukushima-like accident) was performed. In addition, the sensitive study of the time point for water spray was also performed. The next step was the fuel rod performance analysis by using FRAPTRAN and TRACE's results. Finally, the animation model of Chinshan NPP SFP was presented by using the animation function of SNAP with MELCOR analysis results.

  6. Thermal analysis of thermo-gravimetric measurements of spent nuclear fuel oxidation rates

    International Nuclear Information System (INIS)

    Cramer, E.R.

    1997-01-01

    A detailed thermal analysis was completed of the sample temperatures in the Thermo-Gravimetric Analysis (TGA) system used to measure irradiated N Reactor fuel oxidation rates. Sample temperatures during the oxidation process did not show the increase which was postulated as a result of the exothermic reactions. The analysis shows the axial conduction of heat in the sample holder effectively removes the added heat and only a very small, i.e., <10 C, increase in temperature is calculated. A room temperature evaporation test with water showed the sample thermocouple sensitivity to be more than adequate to account for a temperature change of approximately 5 C. Therefore, measured temperatures in the TGA are within approximately 10 C of the actual sample temperatures and no adjustments to reported data to account for the heat input from the oxidation process are necessary

  7. The process system analysis for advanced spent fuel management technology (I)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Lee, J. R.; Kang, D. S.; Seo, C. S.; Shin, Y. J.; Park, S. W.

    1997-12-01

    Various pyrochemical processes were evaluated, and viable options were selected in consideration of the proliferation safety, technological feasibility and compatibility to the domestic nuclear power system. Detailed technical analysis were followed on the selected options such as unit process flowsheet including physico-chemical characteristics of the process systems, preliminary concept development, process design criteria and materials for equipment. Supplementary analysis were also carried out on the support technologies including sampling and transport technologies of molten salt, design criteria and equipment for glove box systems, and remote operation technologies. (author). 40 refs., 49 tabs., 37 figs.

  8. Critical analysis of the Hanford spent nuclear fuel project activity based cost estimate

    Energy Technology Data Exchange (ETDEWEB)

    Warren, R.N.

    1998-09-29

    In 1997, the SNFP developed a baseline change request (BCR) and submitted it to DOE-RL for approval. The schedule was formally evaluated to have a 19% probability of success [Williams, 1998]. In December 1997, DOE-RL Manager John Wagoner approved the BCR contingent upon a subsequent independent review of the new baseline. The SNFP took several actions during the first quarter of 1998 to prepare for the independent review. The project developed the Estimating Requirements and Implementation Guide [DESH, 1998] and trained cost account managers (CAMS) and other personnel involved in the estimating process in activity-based cost (ABC) estimating techniques. The SNFP then applied ABC estimating techniques to develop the basis for the December Baseline (DB) and documented that basis in Basis of Estimate (BOE) books. These BOEs were provided to DOE in April 1998. DOE commissioned Professional Analysis, Inc. (PAI) to perform a critical analysis (CA) of the DB. PAI`s review formally began on April 13. PAI performed the CA, provided three sets of findings to the SNFP contractor, and initiated reconciliation meetings. During the course of PAI`s review, DOE directed the SNFP to develop a new baseline with a higher probability of success. The contractor transmitted the new baseline, which is referred to as the High Probability Baseline (HPB), to DOE on April 15, 1998 [Williams, 1998]. The HPB was estimated to approach a 90% confidence level on the start of fuel movement [Williams, 1998]. This high probability resulted in an increased cost and a schedule extension. To implement the new baseline, the contractor initiated 26 BCRs with supporting BOES. PAI`s scope was revised on April 28 to add reviewing the HPB and the associated BCRs and BOES.

  9. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - II: DUPIC Fuel-Handling Cost

    International Nuclear Information System (INIS)

    Choi, Hangbok; Ko, Won Il; Yang, Myung Seung; Namgung, Ihn; Na, Bok-Gyun

    2001-01-01

    The Direct Use of spent Pressurized water reactor fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel-handling technique has been investigated through a conceptual design study to estimate the unit cost that can be used for the DUPIC fuel cycle cost calculation. The conceptual design study has shown that fresh DUPIC fuel can be transferred to the core following the existing spent-fuel discharge route, provided that new fuel-handling equipment, such as the manipulator, opening/sealing tool of shipping casks, new fuel magazine, new fuel ram, dryer, gamma-ray detector, etc., are installed. The reverse path loading option is known to minimize the number of additional pieces of equipment for fuel handling, because it utilizes the existing spent-fuel handling equipment, and the discharge of spent DUPIC fuel can be done through the existing spent-fuel handling system without any modification. However, because the decay heat of spent DUPIC fuel is much higher than that of spent natural uranium fuel, the extra cooling capacity should be supplemented in the spent-fuel storage bay. Based on the conceptual design study, the capital cost for DUPIC fuel handling and extra storage cooling capacity was estimated to be $3 750 000 (as of December 1999) per CANDU plant. The levelized unit cost of DUPIC fuel handling was then obtained by considering the amount of fuel that will be required during the lifetime of a plant, which is 5.13 $/kg heavy metal. Compared with the other unit costs of the fuel cycle components, it is expected that DUPIC fuel handling has only a minor effect on the overall fuel cycle cost

  10. A novel application of microwave-assisted extraction of polyphenols from brewer's spent grain with HPLC-DAD-MS analysis.

    Science.gov (United States)

    Moreira, Manuela M; Morais, Simone; Barros, Aquiles A; Delerue-Matos, Cristina; Guido, Luís F

    2012-05-01

    This paper reports a novel application of microwave-assisted extraction (MAE) of polyphenols from brewer's spent grains (BSG). A 2(4) orthogonal composite design was used to obtain the optimal conditions of MAE. The influence of the MAE operational parameters (extraction time, temperature, solvent volume and stirring speed) on the extraction yield of ferulic acid was investigated through response surface methodology. The results showed that the optimal conditions were 15 min extraction time, 100 °C extraction temperature, 20 mL of solvent, and maximum stirring speed. Under these conditions, the yield of ferulic acid was 1.31 ± 0.04% (w/w), which was fivefold higher than that obtained with conventional solid-liquid extraction techniques. The developed new extraction method considerably reduces extraction time, energy and solvent consumption, while generating fewer wastes. HPLC-DAD-MS analysis indicated that other hydroxycinnamic acids and several ferulic acid dehydrodimers, as well as one dehydrotrimer were also present, confirming that BSG is a valuable source of antioxidant compounds.

  11. Heat transfer analysis of consolidated dry storage system for CANDU spent fuel considering environmental conditions of Wolsong site

    International Nuclear Information System (INIS)

    Lee, K. H.; Yoon, J. H.; Choi, B. I.; Lee, H. Y.

    2004-01-01

    The purpose of the present paper is to perform heat transfer analysis of the MACSTOR/KN-400 dry storage system for CANDU spent fuel in order to predict maximum concrete temperatures and temperature gradients. This module has twice the capacity of the existing MACSTOR-200, which is in operation at Gentilly-2. In the thermal design of the MACSTOR/KN-400, Thermal Insulation Panels(TIP) were introduced to reduce concrete temperatures and temperature gradients in the module caused by the high fuel heat loads. Environmental factors such as solar heat, daily temperature variations and ambient temperatures in summer and winter at Wolsong site and the assumed presence of hot baskets were taken into consideration in the simulations. Two cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter. The maximum local concrete temperatures were predicted to be 63 .deg. C for the off-normal case. The temperature gradients in the concrete walls and roof are predicted to be 28C and 25C for off-normal operation in summer, incorporating a 3C uncertainty. In conclusion, this paper shows that the maximum temperature for the module is expected to meet the temperature limitations of ACI 349

  12. ASCHFLOW - A dynamic landslide run-out model for medium scale hazard analysis

    Czech Academy of Sciences Publication Activity Database

    Quan Luna, B.; Blahůt, Jan; van Asch, T.W.J.; van Westen, C.J.; Kappes, M.

    2016-01-01

    Roč. 3, 12 December (2016), č. článku 29. E-ISSN 2197-8670 Institutional support: RVO:67985891 Keywords : landslides * run-out models * medium scale hazard analysis * quantitative risk assessment Subject RIV: DE - Earth Magnetism, Geodesy, Geography

  13. Maternal Time Use and Nurturing: Analysis of the Association Between Breastfeeding Practice and Time Spent Interacting with Baby.

    Science.gov (United States)

    Smith, Julie P; Forrester, Robert

    2017-06-01

    Breastfeeding supports child development through complex mechanisms that are not well understood. Numerous studies have compared how well breastfeeding and nonbreastfeeding mothers interact with their child, but few examine how much interaction occurs. Our study of weekly time use among 156 mothers of infants aged 3-9 months investigated whether lactating mothers spend more time providing emotional support or cognitive stimulation of their infants than nonbreastfeeding mothers, and whether the amount of such interactive time is associated with breastfeeding intensity. Mothers were recruited via mother's and baby groups, infant health clinics, and childcare services, and used an electronic device to record their 24-hour time use for 7 days. Sociodemographic and feeding status data were collected by questionnaire. Statistical analysis using linear mixed modeling and residual maximum likelihood analysis compared maternal time use for those giving "some breastfeeding" and those "not breastfeeding." Analysis was also conducted for more detailed feeding subgroups. Breastfeeding and nonbreastfeeding mothers had broadly similar socioeconomic and demographic characteristics. Breastfeeding was found to be associated with more mother-child interaction time, a difference only partially explained by weekly maternal employment hours or other interactive care activities such as play or reading. This study presents data suggesting that lactating mothers spent significantly more hours weekly on milk feeding and on carrying, holding, or soothing their infant than nonlactating mothers; and on providing childcare. Understanding the mechanisms by which child mental health and development benefits from breastfeeding may have important implications for policies and intervention strategies, and could be usefully informed by suitably designed time use studies.

  14. A minimal growth medium for the basidiomycetePleurotus sapidusfor metabolic flux analysis.

    Science.gov (United States)

    Fraatz, Marco A; Naeve, Stefanie; Hausherr, Vanessa; Zorn, Holger; Blank, Lars M

    2014-01-01

    Pleurotus sapidus secretes a huge enzymatic repertoire including hydrolytic and oxidative enzymes and is an example for higher basidiomycetes being interesting for biotechnology. The complex growth media used for submerged cultivation limit basic physiological analyses of this group of organisms. Using undefined growth media, only little insights into the operation of central carbon metabolism and biomass formation, i.e. , the interplay of catabolic and anabolic pathways, can be gained. The development of a chemically defined growth medium allowed rapid growth of P. sapidus in submerged cultures. As P. sapidus grew extremely slow in salt medium, the co-utilization of amino acids using 13 C-labelled glucose was investigated by gas chromatography-mass spectrometry (GC-MS) analysis. While some amino acids were synthesized up to 90% in vivo from glucose ( e.g., alanine), asparagine and/or aspartate were predominantly taken up from the medium. With this information in hand, a defined yeast free salt medium containing aspartate and ammonium nitrate as a nitrogen source was developed. The observed growth rates of P. sapidus were well comparable with those previously published for complex media. Importantly, fast growth could be observed for 4 days at least, up to cell wet weights (CWW) of 400 g L -1 . The chemically defined medium was used to carry out a 13 C-based metabolic flux analysis, and the in vivo reactions rates in the central carbon metabolism of P. sapidus were investigated. The results revealed a highly respiratory metabolism with high fluxes through the pentose phosphate pathway and TCA cycle. The presented chemically defined growth medium enables researchers to study the metabolism of P. sapidus , significantly enlarging the analytical capabilities. Detailed studies on the production of extracellular enzymes and of secondary metabolites of P. sapidus may be designed based on the reported data.

  15. Long-term safety of radioactive waste disposal. Radioactive analysis of samples from spent fuel leaching experiments

    International Nuclear Information System (INIS)

    Geckeis, H.; Degering, D.; Goertzen, A.; Geyer, F.W.; Dressler, P.

    1995-09-01

    In order to assess the long-term performance of spent fuel during direct disposal, high burnup fuel (50 MWd/kg U) has been exposed to non-buffered brine solutions and to deionized water under static anaerobic conditions at 25 C. The leaching behaviour of several radionuclides has been observed over periods of approximately 500 d. Currently used radiometric methods (α-, β-, γ-spectrometry) were applied to the analysis of sample solutions. Due to its low specific activity, uranium was determined using ICP-mass-spectrometry (ICP-MS) or laser induced fluorescence spectrometry (LFS). In order to determine radionuclide concentrations without interferences a preceeding radiochemical separation by ion-exchange, solvent-extraction or extraction chromatography was necessary in most cases. The Sc-isotopes 134/137, which are present in a high excess over other γ-emitting nuclides, were separated using the inorganic ion exchanger ammonium molybdato phosphate (AMP). This step allowed the subsequent γ-spectrometric determination of Am-241, Ag-110m, Ru-106, Sb-125 and Eu-154/155. Activity concentrations of pure β-emitters like Sr-90, Tc-99, I-129 and Pu-241 were determined by liquid scintillation counting (LSC) after selective separation using extraction chromatography or solvent extraction. The actinides Am-241, Cm-242/244, Pu-238/239/240 and Np-237 were analysed by α-spectrometry again after selective separation. The direct analysis of uranium by LFS or ICP-MS was hampered by high salt concentrations. Therefore a separation by extraction chromatography turned out to be necessary, too. The analytical procedures used throughout this work are described in detail. (orig.) [de

  16. Analysis of Spent Moulding Sands with Binders of Various Thermal Degradations, in an Aspect of a Possibility of a Directional Mould Degassing

    Directory of Open Access Journals (Sweden)

    Łucarz M.

    2016-06-01

    Full Text Available The results of investigations of spent moulding sands taken from the mould in which the metal core cooling system - to increase the cooling rate of the ladle casting - was applied, are presented in the hereby paper. The changes of the spent moulding sand at the casting external side being the result of degradation and destruction processes of organic binder, were analysed in this publication. Since the reclaimed material, obtained as a result of the mechanical reclamation of spent sands of the same type, is used as a grain matrix of the moulding sand, the amount of a binder left from the previous technological cycle is essential for the sound castings production. On the bases of investigations of the thermal analysis, ignition losses, dusts contents and pH values of the samples taken from the spent sand the conditions under which the process of gases displacing in the casting mould was realised as well as factors limiting the efficient mould degassing - were considered in this study. The possible reason of a periodical occurrence of an increased number of casting defects due to changing gas volume emission, being the reason of the realised technological process, was indicated.

  17. Development of a Computer Program for an Analysis of the Logistics and Transportation Costs of the PWR Spent Fuels in Korea

    International Nuclear Information System (INIS)

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

    2009-01-01

    It is expected that a substantial amount of spent fuels will be transported from the four nuclear power plant (NPP) sites in Korea to a hypothetical centralized interim storage facility or a final repository in the near future. The cost for the transportation is proportional to the amount of spent fuels. In this paper, a cost estimation program is developed based on the conceptual design of a transportation system and a logistics analysis. Using the developed computer program, named as CASK, the minimum capacity of a centralized interim storage facility (CISF) and the transportation cost for PWR spent fuels are calculated. The PWR spent fuels are transported from 4 NPP sites to a final repository (FR) via the CISF. Since NPP sites and the CISF are located along the coast, a sea-transportation is considered and a road-transportation is considered between the CISF and the FR. The result shows that the minimum capacity of the interim storage facility is 15,000 MTU

  18. Deep geological disposal system development; mechanical structural stability analysis of spent nuclear fuel disposal canister under the internal/external pressure variation

    Energy Technology Data Exchange (ETDEWEB)

    Kwen, Y. J.; Kang, S. W.; Ha, Z. Y. [Hongik University, Seoul (Korea)

    2001-04-01

    This work constitutes a summary of the research and development work made for the design and dimensioning of the canister for nuclear fuel disposal. Since the spent nuclear fuel disposal emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for spent fuel disposal should be securred. Usually this repository is expected to locate at a depth of 500m underground. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for spent nuclear fuel disposal in a deep repository in the crystalline bedrock, which entails an evenly distributed load of hydrostatic pressure from undergroundwater and high pressure from swelling of bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables array type of inner baskets and thicknesses of outer shell and lid and bottom are tried to be determined through the mechanical linear structural analysis, thicknesses of outer shell is determined through the nonlinear structural analysis, and the bentonite buffer analysis for the rock movement is conducted through the of nonlinear structural analysis Also the thermal stress effect is computed for the cast iron insert. The canister types studied here are one for PWR fuel and another for CANDU fuel. 23 refs., 60 figs., 23 tabs. (Author)

  19. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1995-01-01

    The possible allowance of reactivity credit for the exposure history, of power reactor fuel has spurred interest because of the potential of greatly reduced risk and cost when applied to the design and certification of spent-fuel casks used for transportation and storage. Previous pressurized water reactor feasibility assessments are extended to boiling water reactor fuel

  20. Analysis, scale modeling, and full-scale test of a railcar and spent-nuclear-fuel shipping cask in a high-velocity impact against a rigid barrier

    International Nuclear Information System (INIS)

    Huerta, M.

    1981-06-01

    This report describes the mathematical analysis, the physical scale modeling, and a full-scale crash test of a railcar spent-nuclear-fuel shipping system. The mathematical analysis utilized a lumped-parameter model to predict the structural response of the railcar and the shipping cask. The physical scale modeling analysis consisted of two crash tests that used 1/8-scale models to assess railcar and shipping cask damage. The full-scale crash test, conducted with retired railcar equipment, was carefully monitored with onboard instrumentation and high-speed photography. Results of the mathematical and scale modeling analyses are compared with the full-scale test. 29 figures

  1. Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository: accident event analysis and mechanical failure probabilities. Final report

    International Nuclear Information System (INIS)

    Bhaskaran, G.; McCleery, J.E.

    1979-10-01

    This report provides support in developing an accident prediction event tree diagram, with an analysis of the baseline design concept for the retrieval of emplaced spent unreprocessed fuel (SURF) contained in a degraded Canister. The report contains an evaluation check list, accident logic diagrams, accident event tables, fault trees/event trees and discussions of failure probabilities for the following subsystems as potential contributors to a failure: (a) Canister extraction, including the core and ram units; (b) Canister transfer at the hoist area; and (c) Canister hoisting. This report is the second volume of a series. It continues and expands upon the report Retrieval System for Emplaced Spent Unreprocessed Fuel (SURF) in Salt Bed Depository: Baseline Concept Criteria Specifications and Mechanical Failure Probabilities. This report draws upon the baseline conceptual specifications contained in the first report

  2. Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository: accident event analysis and mechanical failure probabilities. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bhaskaran, G.; McCleery, J.E.

    1979-10-01

    This report provides support in developing an accident prediction event tree diagram, with an analysis of the baseline design concept for the retrieval of emplaced spent unreprocessed fuel (SURF) contained in a degraded Canister. The report contains an evaluation check list, accident logic diagrams, accident event tables, fault trees/event trees and discussions of failure probabilities for the following subsystems as potential contributors to a failure: (a) Canister extraction, including the core and ram units; (b) Canister transfer at the hoist area; and (c) Canister hoisting. This report is the second volume of a series. It continues and expands upon the report Retrieval System for Emplaced Spent Unreprocessed Fuel (SURF) in Salt Bed Depository: Baseline Concept Criteria Specifications and Mechanical Failure Probabilities. This report draws upon the baseline conceptual specifications contained in the first report.

  3. Time/motion observations and dose analysis of reactor loading, transportation, and dry unloading of an overweight truck spent fuel shipment

    International Nuclear Information System (INIS)

    Hostick, C.J.; Lavender, J.C.; Wakeman, B.H.

    1992-04-01

    This document presents observed activity durations and radiation dose analyses for an overweight truck shipment of pressurized water reactor (PWR) spent fuel from the Surry Power Station in Virginia to the Idaho National Engineering Laboratory. The shipment consisted of a TN-8L shipping cask carrying three 9-year-old PWR spent fuel assemblies. Handling times and dose analyses for at-reactor activities were completed by Virginia Electric and Power Company (Virginia Power) personnel. Observations of in-transit and unloading activities were made by Pacific Northwest Laboratory (PNL) personnel, who followed the shipment for approximately 2800 miles and observed cask unloading activities. In-transit dose estimates were calculated using dose rate maps provided by Virginia Power for a fully loaded TN-8L shipping cask. The dose analysis for the cask unloading operations is based on the observations of PNL personnel

  4. Collectibility Analysis for Small and Medium Enterprise – Bank BUKOPIN Case

    Directory of Open Access Journals (Sweden)

    Deddy P. Koesrindartoto

    2012-01-01

    Full Text Available In Indonesia, small and medium enterprises (SME have experienced 68% growth in 2008 and47% growth in 2007. The simple structure made it able to respond quickly to changingeconomic conditions and meet local customers’ needs, growing sometimes into large andpowerful corporations or failing within a short time of the firm’s inception. Unfortunately,lack of access to finance has been cited as an important problem for SMEs, being theconstraint in the creation, development or diversification of their economic activities. Lendinginstitution such as banks conduct an intensive assessment (usually called credit scoring tofind out the creditworthiness of applicants in order to mitigate the risk. This paper,“Collectability Analysis in Small and Medium Enterprise”, is purposed for creatingappropriate credit scoring model to support the judgmental analysis approach in small andmedium enterprise, finding out how generic and plafond-specific variables affect thecollectability of the debtors from different plafond level (< Rp. 500 Million and > Rp. 500Million, and providing early detector for the bank to predict about future loan performanceof its debtors, thus the bank can be more careful in selecting qualified debtors.Key words: small and medium enterprises, credit rating, loan plafond, collectability, genericvariables, plafond-specific variables.

  5. Thermal Stress Analysis of Medium-Voltage Converters for Smart Transformers

    DEFF Research Database (Denmark)

    Andresen, Markus; Ma, Ke; De Carne, Giovanni

    2017-01-01

    A smart transformer (ST) can take over an important managing role in the future electrical distribution grid system and can provide many advanced grid services compared to the traditional transformer. However, the risk is that the advanced functionality is balanced out by a lower reliability....... To address this concern, this work conducts thermal stress analysis for a modular multilevel converter (MMC), which is a promising solution for the medium voltage stage of the ST. The focus is put on the mission profiles of the transformer and the impact on the thermal stress of power semiconductor devices....... Normal operation at different power levels and medium voltage grid faults in a feeder fed by a traditional transformer are considered as well as the electrical and the thermal stress of the disconnection and the reconnection procedures. For the validation, the thermal stress of one MMC cell is reproduced...

  6. External Determinants of the Development of Small and Medium-Sized Enterprises – Empirical Analysis

    Directory of Open Access Journals (Sweden)

    Renata Lisowska

    2015-01-01

    Full Text Available The paper aims to identify external determinants of the development of small and medium-sized enterprises and assess their impact on the functioning of these entities in Poland. Meeting this objective required: identifying determinants of the development of SMEs, determining the current development situation of the surveyed enterprises and examining the impact of external determinants on the development of SMEs. The implementation of the above-presented goals was based on the following assumptions: (i the current situation of the surveyed enterprises is determined with the use of quantitative indicators (turnover volume, number of employees, market share, profit levels (ii the analysis of external determinants encompasses three components of the environment: the macro-environment, the meso-environment and the micro-environment, (iii in each analysed area there are separate analyses conducted for micro, small and medium-sized enterprises, enabling greater precision in the identification of external determinants of development for each category of businesses.

  7. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost

    International Nuclear Information System (INIS)

    Choi, Hangbok; Ko, Won Il; Yang, Myung Seung

    2001-01-01

    A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (∼49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC

  8. Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix A, environmental justice analysis. Volume 2

    International Nuclear Information System (INIS)

    1995-03-01

    This is Appendix A to a draft Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. This appendix addresses environmental justice for the acceptance of foreign research reactor spent nuclear fuel containing uranium enriched in the United States. Analyses of environmental justice concerns are provided in three areas: (1) potential ports of entry, (2) potential transportation routes from candidate ports of entry to interim management sites, and (3) areas surrounding potential interim management sites. These analyses lead to the conclusion that the alternatives analyzed in this Environmental Impact Statement (EIS) would result in no disproportionate adverse effects on minority populations or low-income communities surrounding the candidate ports, transport routes, or interim management sites

  9. The Analytical Repository Source-Term (AREST) model: Analysis of spent fuel as a nuclear waste form

    International Nuclear Information System (INIS)

    Apted, M.J.; Liebetrau, A.M.; Engel, D.W.

    1989-02-01

    The purpose of this report is to assess the performance of spent fuel as a final waste form. The release of radionuclides from spent nuclear fuel has been simulated for the three repository sites that were nominated for site characterization in accordance with the Nuclear Waste Policy Act of 1982. The simulation is based on waste package designs that were presented in the environmental assessments prepared for each site. Five distinct distributions for containment failure have been considered, and the release for nuclides from the UO 2 matrix, gap (including grain boundary), crud/surface layer, and cladding has been calculated with the Analytic Repository Source-Term (AREST) code. Separate scenarios involving incongruent and congruent release from the UO 2 matrix have also been examined using the AREST code. Congruent release is defined here as the condition in which the relative mass release rates of a given nuclide and uranium from the UO 2 matrix are equal to their mass ratios in the matrix. Incongruent release refers to release of a given nuclide from the UO 2 matrix controlled by its own solubility-limiting solid phase. Release of nuclides from other sources within the spent fuel (e.g., cladding, fuel/cladding gap) is evaluated separately from either incongruent or congruent matrix release. 51 refs., 200 figs., 9 tabs

  10. Proof of concept: preimplantation genetic screening without embryo biopsy through analysis of cell-free DNA in spent embryo culture media.

    Science.gov (United States)

    Shamonki, Mousa I; Jin, Helen; Haimowitz, Zachary; Liu, Lian

    2016-11-01

    To assess whether preimplantation genetic screening (PGS) is possible by testing for free embryonic DNA in spent IVF media from embryos undergoing trophectoderm biopsy. Prospective cohort analysis. Academic fertility center. Seven patients undergoing IVF and 57 embryos undergoing trophectoderm biopsy for PGS. On day 3 of development, each embryo was placed in a separate media droplet. All biopsied embryos received a PGS result by array comparative genomic hybridization. Preimplantation genetic screening was performed on amplified DNA extracted from media and results were compared with PGS results for the corresponding biopsy. [1] Presence of DNA in spent IVF culture media. [2] Correlation between genetic screening result from spent media and corresponding biopsy. Fifty-five samples had detectable DNA ranging from 2-642 ng/μL after a 2-hour amplification. Six samples with the highest DNA levels underwent PGS, rendering one result with a derivative log ratio SD (DLRSD) of media and a result that is consistent with trophectoderm biopsy. Improvements in DNA collection, amplification, and testing may allow for PGS without biopsy in the future. Copyright © 2016 American Society for Reproductive Medicine. Published by Elsevier Inc. All rights reserved.

  11. Spent fuels program

    International Nuclear Information System (INIS)

    Shappert, L.B.

    1983-01-01

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

  12. The cost of spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Palacios H, J. C.; Badillo, V.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    Spent fuel is one of the most important issues in the nuclear industry, currently spent fuel management is been cause of great amount of research, investments, constructing repositories or constructing the necessary facilities to reprocess the fuel, and later to recycle the plutonium recovered in thermal reactors. What is the best solution?, or What is the best technology for an specific solution? Many countries have deferred the decision on selecting an option, while others works actively constructing repositories and others implementing the reprocessing facilities to recycle the plutonium obtained from nuclear spent fuel. In Mexico the nuclear power is limited to two reactors BWR type and medium size. So the nuclear spent fuel discharged has been accommodated at reactor's spent fuel pools. Originally these pools have enough capacity to accommodate spent fuel for the 40 years of designed plant operation. However currently, the plants are under a process for extended power up-rate to 20% of original power and also there are plans to extended operational life for 20 more years. Under these conditions there will not be enough room for spent fuel in the pools. (Author)

  13. Analysis of low and medium energy physics records in databases. Science and technology indicators in low and medium energy physics. With particular emphasis on nuclear data

    International Nuclear Information System (INIS)

    Hillebrand, C.D.

    1998-12-01

    An analysis of the literature on low and medium energy physics, with particular emphasis on nuclear data, was performed on the basis of the contents of the bibliographic database INIS (International Nuclear Information System). Quantitative data were obtained on various characteristics of relevant INIS records such as subject categories, language and country of publication, publication types, etc. Rather surprisingly, it was found that the number of records in nuclear physics has remained nearly constant over the last decade. The analysis opens up the possibility of further studies, e.g. on international research co-operation and on publication patterns. (author)

  14. Three-dimensional visualization of objects in scattering medium using integral imaging and spectral analysis

    Science.gov (United States)

    Lee, Yeonkyung; Yoo, Hoon

    2016-02-01

    This paper presents a three-dimensional visualization method of 3D objects in a scattering medium. The proposed method employs integral imaging and spectral analysis to improve the visual quality of 3D images. The images observed from 3D objects in the scattering medium such as turbid water suffer from image degradation due to scattering. The main reason is that the observed image signal is very weak compared with the scattering signal. Common image enhancement techniques including histogram equalization and contrast enhancement works improperly to overcome the problem. Thus, integral imaging that enables to integrate the weak signals from multiple images was discussed to improve image quality. In this paper, we apply spectral analysis to an integral imaging system such as the computational integral imaging reconstruction. Also, we introduce a signal model with a visibility parameter to analyze the scattering signal. The proposed method based on spectral analysis efficiently estimates the original signal and it is applied to elemental images. The visibility-enhanced elemental images are then used to reconstruct 3D images using a computational integral imaging reconstruction algorithm. To evaluate the proposed method, we perform the optical experiments for 3D objects in turbid water. The experimental results indicate that the proposed method outperforms the existing methods.

  15. Review of the technical basis and verification of current analysis methods used to predict seismic response of spent fuel storage racks

    International Nuclear Information System (INIS)

    DeGrassi, G.

    1992-10-01

    This report presents the results of a literature review on spent fuel rack seismic analysis methods and modeling procedures. The analysis of the current generation of free standing high density spent fuel racks requires careful consideration of complex phenomena such as rigid body sliding and tilting motions; impacts between adjacent racks, between fuel assemblies and racks, and between racks and pool walls and floor; fluid coupling and frictional effects. The complexity of the potential seismic response of these systems raises questions regarding the levels of uncertainty and ranges of validity of the analytical results. BNL has undertaken a program to investigate and assess the strengths and weaknesses of current fuel rack seismic analysis methods. The first phase of this program involved a review of technical literature to identify the extent of experimental and analytical verification of the analysis methods and assumptions. Numerous papers describing analysis methods for free standing fuel racks were reviewed. However, the extent of experimental verification of these methods was found to be limited. Based on the information obtained from the literature review, the report provides an assessment of the significance of the issues of concern and makes recommendations for additional studies

  16. Neutron analysis of spent fuel storage installation using parallel computing and advance discrete ordinates and Monte Carlo techniques

    International Nuclear Information System (INIS)

    Shedlock, D.; Haghighat, A.

    2005-01-01

    In the United States, the Nuclear Waste Policy Act of 1982 mandated centralised storage of spent nuclear fuel by 1988. However, the Yucca Mountain project is currently scheduled to start accepting spent nuclear fuel in 2010. Since many nuclear power plants were only designed for ∼10 y of spent fuel pool storage, >35 plants have been forced into alternate means of spent fuel storage. In order to continue operation and make room in spent fuel pools, nuclear generators are turning towards independent spent fuel storage installations (ISFSIs). Typical vertical concrete ISFSIs are ∼6.1 m high and 3.3 m in diameter. The inherently large system, and the presence of thick concrete shields result in difficulties for both Monte Carlo (MC) and discrete ordinates (S N ) calculations. MC calculations require significant variance reduction and multiple runs to obtain a detailed dose distribution. S N models need a large number of spatial meshes to accurately model the geometry and high quadrature orders to reduce ray effects, therefore, requiring significant amounts of computer memory and time. The use of various differencing schemes is needed to account for radial heterogeneity in material cross sections and densities. Two P 3 , S 12 , discrete ordinate, PENTRAN (parallel environment neutral-particle Transport) models were analysed and different MC models compared. A multigroup MCNP model was developed for direct comparison to the S N models. The biased A 3MCNP (automated adjoint accelerated MCNP) and unbiased (MCNP) continuous energy MC models were developed to assess the adequacy of the CASK multigroup (22 neutron, 18 gamma) cross sections. The PENTRAN S N results are in close agreement (5%) with the multigroup MC results; however, they differ by ∼20-30% from the continuous-energy MC predictions. This large difference can be attributed to the expected difference between multigroup and continuous energy cross sections, and the fact that the CASK library is based on the

  17. Medium Duty ARRA Data Reporting and Analysis; NREL (National Renewable Energy Laboratory)

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Kenneth; Duran, Adam; Ragatz, Adam; Prohaska, Robert; Walkowicz, Kevin

    2015-06-11

    Medium-duty (MD) electric vehicle (EV) data collection and analysis will help drive design, purchase, and research investments. Over 4 million miles and 160,000 driving days of EV driving data were collected under this project. Publicly available data help drive technology research, development, and deployment. Feeding the vocational database for future analysis will lead to a better understanding of usage and will result in better design optimization and technology implementation. The performance of a vehicle varies with drive cycle and cargo load - MD vehicles are 'multi-functional.' Environment and accessory loads affect vehicle range and in turn add cost by adding battery capacity. MD EV vehicles can function in vocations traditionally serviced by gasoline or diesel vehicles. Facility implications (i.e., demand charges) need to be understood as part of site-based analysis for EV implementation.

  18. Reference analysis on the use of engineered barriers for isolation of spent nuclear fuel in granite and basalt

    Energy Technology Data Exchange (ETDEWEB)

    Cloninger, M.O.; Cole, C.R.

    1981-08-01

    This report evaluates the effectiveness of engineered barriers in delaying or reducing the rate of release of radionuclides from spent fuel in geologic respositories in granite and basalt. It was assumed that the major exposure pathway from the respository to humans would be the ground-water system overlying or underlying a site. Hence, this report focuses on ground-water pathways. A geosphere transport model, GETOUT, and the biosphere transport/dose models, ALLDOS and PABLM, were integrated and used to calculate the potential radiological dose that might be received by humans at various times after repository closure.

  19. A Review and Analysis of European Industrial Experience in Handling LWR Spent Fuel and Vitrified High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Blomeke, J.O.

    2001-07-10

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performance of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States.

  20. Regional statistical and economic analysis of small and medium-sized businesses development in Zhytomyr region

    Directory of Open Access Journals (Sweden)

    S.I. Pavlova

    2017-08-01

    Full Text Available Small and medium-sized businesses play an important role in the development of the regional economic system in particular and in solving a number of the following local problems: developing competition, developing the market for goods and services, providing jobs for the able-bodied population, raising living standards and improving the social environment in society. The purpose of this paper is to analyze the state and development of small and medium-sized businesses in the Zhytomyr region, to analyze its contribution to the economic development of the region, and to identify the main problems existing in the region. According to the indicators of state statistics, the author presents the general characteristics of enterprises in the Zhytomyr region from 2012 to 2016 in the context of indicators of the number of enterprises, the number of employed workers and the volume of the products sold, highlighting the activities of small enterprises and assessing their share in general levels. In addition, the paper provides the description of the activities of individual entrepreneurs. The structural comparison for the above-listed indicators of the distribution of influence on the economic system of the Zhytomyr region in terms of enterprises by size is presented. In terms of quantity 93,5 % are small enterprises that provide 31,4 % of the total number of employees with work and make up 23,1 % of the total volume of sales. Average enterprises in these indicators have 6,4 %, 62,0 % and 54,8 % respectively. The statistical and economic analysis of the structure of small enterprises by types of economic activity, by indicators of the number of registered enterprises, and by the volumes of sold products is carried out. The uniformity of the distribution is estimated using the index of the concentration coefficient. The indicators of revenues to budgets of different levels from small and medium-sized businesses are set. The paper presents and summarizes the

  1. Method for storing spent nuclear fuel in repositories

    Science.gov (United States)

    Schweitzer, D.G.; Sastre, C.; Winsche, W.

    A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

  2. Method for storing spent nuclear fuel in repositories

    Science.gov (United States)

    Schweitzer, Donald G.; Sastre, Cesar; Winsche, Warren

    1981-01-01

    A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

  3. Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository--Volume 2: Methodology and Results

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, L.L.; Wilson, J.R.; Sanchez, L.C.; Aguilar, R.; Trellue, H.R.; Cochrane, K.; Rath, J.S.

    1998-10-01

    The US Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3).

  4. Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository--Volume 2: Methodology and Results

    International Nuclear Information System (INIS)

    Taylor, L.L.; Wilson, J.R.; Sanchez, L.C.; Aguilar, R.; Trellue, H.R.; Cochrane, K.; Rath, J.S.

    1998-01-01

    The US Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3)

  5. Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository--Volume 1: Executive Summary

    International Nuclear Information System (INIS)

    Taylor, L.L.; Wilson, J.R.; Sanchez, L.Z.; Aguilar, R.; Trellue, H.R.; Cochrane, K.; Rath, J.S.

    1998-01-01

    The US Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3)

  6. Analysis of Medium-Scale Solar Thermal Systems and Their Potential in Lithuania

    Directory of Open Access Journals (Sweden)

    Rokas Valančius

    2015-06-01

    Full Text Available Medium-scale solar hot water systems with a total solar panel area varying from 60 to 166 m2 have been installed in Lithuania since 2002. However, the performance of these systems varies depending on the type of energy users, equipment and design of the systems, as well as their maintenance. The aim of this paper was to analyse operational SHW systems from the perspective of energy production and economic benefit as well as to outline the differences of their actual performance compared to the numerical simulation results. Three different medium-scale solar thermal systems in Lithuania were selected for the analysis varying in both equipment used (flat type solar collectors, evacuated tube collectors and type of energy user (swimming pool building, domestic hot water heating, district heating. The results of the analysis showed that in the analysed cases the gap between measured and modelled data of heat energy produced by SHW systems was approx. 11%. From the economical perspective, the system with flat type solar collectors used for domestic hot water production was proved to be most efficient. However, calculation of Internal Rate of Return showed that a grant of 35% is required for this project to be fully profitable.

  7. Costing of spent nuclear fuel storage

    International Nuclear Information System (INIS)

    2009-01-01

    This report deals with economic analysis and cost estimation, based on exploration of relevant issues, including a survey of analytical tools for assessment and updated information on the market and financial issues associated with spent fuel storage. The development of new storage technologies and changes in some of the circumstances affecting the costs of spent fuel storage are also incorporated. This report aims to provide comprehensive information on spent fuel storage costs to engineers and nuclear professionals as well as other stakeholders in the nuclear industry. This report is meant to provide informative guidance on economic aspects involved in selecting a spent fuel storage system, including basic methods of analysis and cost data for project evaluation and comparison of storage options, together with financial and business aspects associated with spent fuel storage. After the review of technical options for spent fuel storage in Section 2, cost categories and components involved in the lifecycle of a storage facility are identified in Section 3 and factors affecting costs of spent fuel storage are then reviewed in the Section 4. Methods for cost estimation and analysis are introduced in Section 5, and other financial and business aspects associated with spent fuel storage are discussed in Section 6.

  8. Dynamic Stability Analysis of Autonomous Medium-Voltage Mixed-Source Microgrid

    DEFF Research Database (Denmark)

    Zhao, Zhuoli; Yang, Ping; Guerrero, Josep M.

    2015-01-01

    In an autonomous microgrid without connection to the interconnected grid to support the frequency and voltage, it is more complex to control and manage. Thus in order to investigate the dynamic stability of the medium-voltage (MV) microgrid in Dongao Island, a detailed small-signal state......-space model of the autonomous MV mixed-source microgrid containing diesel generator set (DGS), grid-supporting battery energy storage system (BESS), squirrel cage induction generator (SCIG) wind turbine and network is developed. Sensitivity analysis is carried out to reveal the dynamic stability margin...... of the MV microgrid and identify the permissible variations range of the droop coefficients. Finally, the theoretical analysis is verified with time-domain simulation and experimental results....

  9. Systems report on the analysis of spent, highly enriched U-235 reactor fuel by delayed neutron interrogation

    International Nuclear Information System (INIS)

    Piper, T.C.; Kirkham, R.J.

    1990-05-01

    Design aspects are briefly given of a neutron source shuffler used to measure fissile material content of spent, highly enriched reactor fuel. The mode of operation used, results of analyzing 176 fuel packages and recommended system improvements are discussed. Four measurements were made on each of the fuel packages with the mean of the 176 standard deviations being 2.03 percent of value. The maximum individual standard deviation was 9.27 percent. Appendixes concerning imprecisions introduced by counting statistics and crane speed irregularities are given. Use of an improved neutron source shuffler, an improved fuel package motion system and modernized computer system should permit system performance to be limited mainly by counting statistics, to about 1.5 percent of measured value. A stronger source could then be installed to further enhance system operation. 16 figs., 3 tabs

  10. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2005-01-01

    When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

  11. Analysis of continuous solvent extraction of nickel from spent electroless nickel plating baths by a mixer-settler

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Ying, E-mail: huang-ying@aist.go.jp [Metals Recycling Group, Research Institute for Environmental Management Technology, National Institute of Advanced Industrial Science and Technology (AIST), 16-1 Onogawa, Tsukuba, Ibaraki 305-8569 (Japan); Tanaka, Mikiya, E-mail: mky-tanaka@aist.go.jp [Metals Recycling Group, Research Institute for Environmental Management Technology, National Institute of Advanced Industrial Science and Technology (AIST), 16-1 Onogawa, Tsukuba, Ibaraki 305-8569 (Japan)

    2009-05-30

    It is urgent to develop an effective technique to treat the large amount of spent electroless nickel plating bath and recycle the high concentration nickel. In our previous study, high recycling efficiency of nickel from the model spent bath was obtained by continuous solvent extraction with 2-hydroxy-5-nonylacetophenone oxime (LIX84I) as the extractant and 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (PC88A) as the accelerator using a mixer-settler extractor. It was observed that the extraction efficiency was affected by the operation parameters such as the flow rates of the aqueous and organic phases and the total stage number. In the present study, the effects of the operation parameters on the extraction efficiency were quantitatively studied on the basis of the pseudo-first-order interfacial extraction rate equation together with the hydrodynamic properties in the mixer. The organic phase holdup, measured under varying conditions of the flow rates of both phases, was analyzed by the Takahashi-Takeuchi holdup model in order to estimate the specific interfacial area. The overall extraction rate coefficients defined by the product of the interfacial extraction rate constant and the specific interfacial area were evaluated using the experimental data and ranged from 3.5 x 10{sup -3} to 6.7 x 10{sup -3} s{sup -1}, which was close to the value of 3.4 x 10{sup -3} s{sup -1} obtained by batch extraction. Finally, an engineering simulation method was established for assessing the extraction efficiency of nickel during a multistage operation.

  12. Analysis of continuous solvent extraction of nickel from spent electroless nickel plating baths by a mixer-settler

    International Nuclear Information System (INIS)

    Huang, Ying; Tanaka, Mikiya

    2009-01-01

    It is urgent to develop an effective technique to treat the large amount of spent electroless nickel plating bath and recycle the high concentration nickel. In our previous study, high recycling efficiency of nickel from the model spent bath was obtained by continuous solvent extraction with 2-hydroxy-5-nonylacetophenone oxime (LIX84I) as the extractant and 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (PC88A) as the accelerator using a mixer-settler extractor. It was observed that the extraction efficiency was affected by the operation parameters such as the flow rates of the aqueous and organic phases and the total stage number. In the present study, the effects of the operation parameters on the extraction efficiency were quantitatively studied on the basis of the pseudo-first-order interfacial extraction rate equation together with the hydrodynamic properties in the mixer. The organic phase holdup, measured under varying conditions of the flow rates of both phases, was analyzed by the Takahashi-Takeuchi holdup model in order to estimate the specific interfacial area. The overall extraction rate coefficients defined by the product of the interfacial extraction rate constant and the specific interfacial area were evaluated using the experimental data and ranged from 3.5 x 10 -3 to 6.7 x 10 -3 s -1 , which was close to the value of 3.4 x 10 -3 s -1 obtained by batch extraction. Finally, an engineering simulation method was established for assessing the extraction efficiency of nickel during a multistage operation.

  13. Analysis of continuous solvent extraction of nickel from spent electroless nickel plating baths by a mixer-settler.

    Science.gov (United States)

    Huang, Ying; Tanaka, Mikiya

    2009-05-30

    It is urgent to develop an effective technique to treat the large amount of spent electroless nickel plating bath and recycle the high concentration nickel. In our previous study, high recycling efficiency of nickel from the model spent bath was obtained by continuous solvent extraction with 2-hydroxy-5-nonylacetophenone oxime (LIX84I) as the extractant and 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (PC88A) as the accelerator using a mixer-settler extractor. It was observed that the extraction efficiency was affected by the operation parameters such as the flow rates of the aqueous and organic phases and the total stage number. In the present study, the effects of the operation parameters on the extraction efficiency were quantitatively studied on the basis of the pseudo-first-order interfacial extraction rate equation together with the hydrodynamic properties in the mixer. The organic phase holdup, measured under varying conditions of the flow rates of both phases, was analyzed by the Takahashi-Takeuchi holdup model in order to estimate the specific interfacial area. The overall extraction rate coefficients defined by the product of the interfacial extraction rate constant and the specific interfacial area were evaluated using the experimental data and ranged from 3.5 x 10(-3) to 6.7 x 10(-3)s(-1), which was close to the value of 3.4 x 10(-3)s(-1) obtained by batch extraction. Finally, an engineering simulation method was established for assessing the extraction efficiency of nickel during a multistage operation.

  14. A Composite Modeling Analysis of the Deformation Behavior of Medium Manganese Steels

    Energy Technology Data Exchange (ETDEWEB)

    Rana, Radhakanta [CSM/ASPPRC; Gibbs, Paul J [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); De Moor, Emmanuel [CSM/ASPPRC; Speer, John G [CSM/ASPPRC; Matlock, David K [CSM/ASPPRC

    2014-09-01

    The deformation behavior of medium manganese steels was evaluated with uniaxial tensile testing and the results were correlated with predictions of a composite model shown previously to provide design insight into the development of multi-phase steels with third-generation advanced high strength steel (3GAHSS) properties. An equilibrium thermodynamic-based methodology to design microstructures containing systematic amounts of metastable austenite with controlled stability against transformation is presented. The analysis is based on Mn enrichment of austenite during intercritical annealing of medium Mn (7 and 10 wt pct.) low carbon (0.1 and 0.15 wt pct) steels. The steels were produced as laboratory heats that were hot and cold rolled prior to annealing. After annealing the microstructures consisted primarily of either a matrix of fine grained ferrite with austenite contents between 32.6 and 45.2 wt pct (7Mn, 0.1C steels) or a matrix of martensite with various amounts of austenite in the higher Mn steel. The different intercritical annealing conditions produced steels with wide variations in austenite contents and austenite compositions (Mn and C contents) resulting in steels with significant variations in austenite stability. Predictions based on the composite analysis with different assumed flow behaviors for the individual constituents and stability functions for the meta-stable austenite are presented and shown to accurately predict strength-ductility combinations over a range of austenite volume fractions for the 7Mn steel. Applicability of the composite analysis is extended to consider the deformation behavior of the 10Mn steel and evaluate other possible microstructural combinations leading to 3GAHSS properties.

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

    International Nuclear Information System (INIS)

    1997-05-01

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

  16. Report on the ANSTO application for a licence to construct a Replacement Research Reactor, addressing seismic analysis and seismic design accident analysis, spent fuel and radioactive wastes

    International Nuclear Information System (INIS)

    2002-02-01

    The Report of the Nuclear Safety Committee (NSC) covers specific terms of reference as requested by the Chief Executive Officer of ARPANSA. The primary issue for the Working Group(WG) consideration was whether ANSTO had demonstrated: (i) that the overall approach to seismic analysis and its implementation in the design is both conservative and consistent with the international best practice; (ii) whether the full accident analysis in the Probabilistic Safety Assesment Report (PSAR) satisfies the radiation dose/frequency criteria specified in ARPANSA's regulatory assessment principle 28 and the assumptions used in the Reference Accident for the siting assessment have been accounted for in the PSAR; and (iii) the adequacy of the strategies for managing the spent fuel as proposed to be used in the Replacement Research Reactor and other radioactive waste (including emissions, taking into account the ALARA criterion) arising from the operation of the proposed replacement reactor and radioisotope production. The report includes a series of questions that were asked of the Applicant in the course of working group deliberations, to illustrate the breadth of inquiries that were made. The Committee noted that replies to some questions remain outstanding at the date of this document. The NSC makes a number of recommendations that appear in each section of the document, which has been compiled in three parts representing the work of each group. The NSC notes some lack of clarity in what was needed to be considered at this approval stage of the project, as against information that would be required at a later stage. While not in the original work plan, recent events of September 11, 2001 also necessitated some exploration of issues relating to construction security. Copyright (2002) Commonwealth of Australia

  17. Analysis of a hydrometallurgical route to recover base metals from spent rechargeable batteries by liquid-liquid extraction with Cyanex 272

    Science.gov (United States)

    Mantuano, Danuza Pereira; Dorella, Germano; Elias, Renata Cristina Alves; Mansur, Marcelo Borges

    A hydrometallurgical route is proposed to recover zinc and manganese from spent alkaline batteries in order to separate base metals such as nickel, copper, aluminium, cadmium, lithium and cobalt which constitute the main metallic species of spent NiCd, NiMH and Li-ion rechargeable batteries. The route comprises the following main steps: (1) sorting batteries by type, (2) battery dismantling to separate the spent battery dust from plastic, iron scrap and paper, (3) leaching of the dust with sulphuric acid and (4) metal separation by a liquid-liquid extraction using Cyanex 272 (bis-2,4,4-trimethylpentyl phosphinic acid) as extractant. The metal content of NiCd, NiMH and Li-ion batteries from three distinct manufacturers has been evaluated. A factorial design of experiments was used to investigate the leaching step using operational variables such as temperature, H 2SO 4 concentration, S/L ratio and H 2O 2 concentration. Analysis of metal separation by the liquid-liquid extraction with Cyanex 272 identified a pH 1/2 2.5-3.0 for zinc and aluminium, pH 1/2 4.0-4.5 for manganese, cadmium, copper and cobalt, pH 1/2 6.5 for nickel and pH 1/2 8.0 for lithium. These results indicate that batteries must be previously sorted by type and treated separately. In addition, data fitting to an equilibrium model proposed for the reactive test system by the European Federation of Chemical Engineering (EFChE) have indicated that MR 2(RH) 2 and MR 2 complexes (where M = Zn, Mn, Co, Cd and Cu) co-exist in the organic phase with Cyanex 272 depending on the loading conditions. The route has been found technically viable to separate the main metallic species of all batteries considered in this study.

  18. Dynamic Analysis of Heavy Vehicle Medium Duty Drive Shaft Using Conventional and Composite Material

    Science.gov (United States)

    Kumar, Ashwani; Jain, Rajat; Patil, Pravin P.

    2016-09-01

    The main highlight of this study is structural and modal analysis of single piece drive shaft for selection of material. Drive shaft is used for torque carrying from vehicle transmission to rear wheel differential system. Heavy vehicle medium duty transmission drive shaft was selected as research object. Conventional materials (Steel SM45 C, Stainless Steel) and composite materials (HS carbon epoxy, E Glass Polyester Resin Composite) were selected for the analysis. Single piece composite material drive shaft has advantage over conventional two-piece steel drive shaft. It has higher specific strength, longer life, less weight, high critical speed and higher torque carrying capacity. The main criteria for drive shaft failure are strength and weight. Maximum modal frequency obtained is 919 Hz. Various harmful vibration modes (lateral vibration and torsional vibration) were identified and maximum deflection region was specified. For single-piece drive shaft the natural bending frequency should be higher because it is subjected to torsion and shear stress. Single piece drive shaft was modelled using Solid Edge and Pro-E. Finite Element Analysis was used for structural and modal analysis with actual running boundary condition like frictional support, torque and moment. FEA simulation results were validated with experimental literature results.

  19. The spent fuel fate

    International Nuclear Information System (INIS)

    2001-01-01

    The spent fuel is not a waste. It can be upgrade by a reprocessing which extracts all products able to produce energy. The today situation is presented and economically analyzed and future alternatives are discussed. (A.L.B.)

  20. Spent fuel workshop'2002

    International Nuclear Information System (INIS)

    Poinssot, Ch.

    2002-01-01

    This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO 2 fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO 2 dissolution determined from electrochemical experiments with 238 Pu doped UO 2 M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO 2 studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with α doped UO 2 in Boom clay conditions (K. Lemmens), Studies of the behavior of UO 2 / water interfaces under He 2+ beam (C. Corbel), Alpha and gamma radiolysis effects on UO 2 alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines (M. Kelm), On the potential catalytic behavior of

  1. Spent fuel centralized storage

    International Nuclear Information System (INIS)

    Chometon, P.L.

    1985-01-01

    Nuclear energy producer countries have felt the need to build a centralized spent fuel storage before reprocessing (for example, COGEMA in FRANCE), either in an adjoining plant on an appropriate site, or isolated. More rarely, this storage enables to decide whether to reprocess or to definitely store spent fuel considered as being waste: for example CLAB in Sweden. Our Company is specialized in the design and construction of spent fuel centralized storage plants. Storage generally takes place in a pool in order to facilitate handling operations and retrieving of these fuels, but these operations may also be effected in a dry way, either in concrete structures or in storage casks. With respect to pools, which might currently be the most appropriate and flexible system, several improvements have recently been made in the design of cask reception facilities and spent fuel storage. These improvements are presented, hereafter [fr

  2. Spent fuel centralized storage

    International Nuclear Information System (INIS)

    Baillif, L.; Chometon, P.L.

    1986-01-01

    Nuclear energy producer countries have felt the need to build a centralized spent fuel storage before reprocessing, either in an adjoining plant on an appropriate site, or isolated. More rarely, this storage enables to decide whether to reprocess or to definitely store spent fuel considered as being waste: for example CLAB in Sweden. Our Company SGN is specialized among others in the design and construction of spent fuel centralized storage plants. Storage generally takes place in a pool in order to facilitate handling operations and retrieving of these fuels, but these operations may also be effected in a dry way, either in concrete structures or in storage casks. With respect to pools, which might currently be the most appropriate and flexible system, several improvements have recently been made in the design of cask reception facilities and spent fuel storage. These improvements are presented, hereafter [fr

  3. Rota hidrometalúrgica de recuperação de molibdênio, cobalto, níquel e alumínio de catalisadores gastos de hidrotratamento em meio ácido Hydrometallurgical route to recover molybdenum, nickel, cobalt and aluminum from spent hydrotreating catalysts in acid medium

    Directory of Open Access Journals (Sweden)

    Ivam Macedo Valverde Júnior

    2008-01-01

    Full Text Available This work describes a hydrometallurgical route for processing spent commercial catalysts (CoMo and NiMo/Al2O3. Samples were preoxidized (500 ºC, 5 h in order to eliminate coke and other volatile species present. The calcined solid was dissolved in concentrated H2SO4 and water (1:1 vol/vol at 90 ºC; the insoluble matter was separated from the solution. Molybdenum was recovered by solvent extraction using tertiary amines at pH around 1.8. Cobalt (or nickel was separated by addition of aqueous ammonium oxalate at the above pH. Phosphorus was removed by passing the liquid through a strong anion exchange column. Aluminum was recovered by neutralizing the solution with NaOH. The route presented in this work generates less final aqueous wastes because it is not necessary to use alkaline medium during the metal recovery steps.

  4. Spent fuel management

    International Nuclear Information System (INIS)

    2005-01-01

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

  5. Thermo-mechanical FE-analysis of butt-welding of a Cu-Fe canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    Josefson, B.L.; Karlsson, L.; Lindgren, L.E.; Jonsson, M.

    1992-10-01

    In the Swedish nuclear waste program it has been proposed that spent nuclear fuel shall be placed in composite copper-steel canisters. These canisters will be placed in holes in tunnels located some 500 m underground in a rock storage. The canisters consists of two cylinders of 4850 mm length, one inner cylinder made of steel and one outer cylinder made of copper. The outer diameter of the canister is 880 mm and the wall thickness for each cylinder is 50 mm. At the storage, the steel cylinder, which contains the spent nuclear fuel, is placed inside the copper cylinder. Thereafter, a copper end is butt welded to the copper cylinder using electron beam welding. To obtain penetration through the thickness with this weld method a backing ring is placed at the inside of the copper cylinder. In the present paper, the temperature, strain and stress fields present during welding and after cooling after welding are calculated numerically using the FE-code NIKE-2D. The aim is to use the results of the present calculations to estimate the risk for creep fracture during the subsequent design life. A large strain formulation is employed for the calculation of transient and residual stresses in the canister based on the calculated history of the temperature field present in the canister during the welding process. The contact algorithm available in NIKE-2D is used to detect possible contact between the steel and copper cylinders during the welding. To simplify the numerical calculations and reduce the computational time, rotational symmetry is assumed. For large gap distances between the steel and copper cylinders the residual stress field is calculated to have a shape similar to that observed in butt welded pipes with maximum axial stress values at the yield stress level. For small gap distances the backing ring will come in contact with the steel cylinder which leads to large residual stresses in the backing ring. The maximum accumulated plastic strain in the weld metal and

  6. A multi-attribute utility decision analysis for treatment alternatives for the DOE/SR aluminum-based spent nuclear fuel

    International Nuclear Information System (INIS)

    Davis, Freddie J.; Weiner, Ruth Fleischman; Wheeler, Timothy A.; Sorenson, Ken B.; Kuzio, Kenneth A.

    2000-01-01

    A multi-attribute utility analysis is applied to a decision process to select a treatment method for the management of aluminum-based spent nuclear fuel (Al-SNF) owned by the US Department of Energy (DOE). DOE will receive, treat, and temporarily store Al-SNF, most of which is composed of highly enriched uranium, at its Savannah River Site in South Carolina. DOE intends ultimately to send the treated Al-SNF to a geologic repository for permanent disposal. DOE initially considered ten treatment alternatives for the management of Al-SNF, and has narrowed the choice to two of these: the direct disposal and melt and dilute alternatives. The decision analysis presented in this document focuses on a formal decision process used to evaluate these two remaining alternatives

  7. Shielding analysis of a transport and storage cask for spent BWR fuel applicability of the code SAS4 and discussion of results

    Energy Technology Data Exchange (ETDEWEB)

    Hilbert, F. [Nuclear Cargo and Service GmbH, Hanau (Germany); Morishima, M.; Tamaki, H. [Mitsubishi Heavy Industries, Kobe (Japan)

    2004-07-01

    For the shielding analysis of transport and storage casks for spent fuel the use of computer codes is state of the art. However, in most applications the computer models used for the analysis are simplified to circular geometries to save modelling effort and calculation time. Furthermore, the active zone of the fuel is modelled as homogeneous zone with a uniform average burn-up. In the first part of the present paper it is shown that an exact model is feasible and the effect of the geometrical shape on the dose rates is illustrated. The second part of the paper shows the comparison of the dose rates calculated with 5 different fuel models. Finally, the accuracy of the calculations is discussed.

  8. Compound effects of operating parameters on burnup credit criticality analysis in boiling water reactor spent fuel assemblies

    Directory of Open Access Journals (Sweden)

    Shang-Chien Wu

    2018-02-01

    Full Text Available This study proposes a new method of analyzing the burnup credit in boiling water reactor spent fuel assemblies against various operating parameters. The operating parameters under investigation include fuel temperature, axial burnup profile, axial moderator density profile, and control blade usage. In particular, the effects of variations in one and two operating parameters on the curve of effective multiplication factor (keff versus burnup (B are, respectively, the so-called single and compound effects. All the calculations were performed using SCALE 6.1 together with the Evaluated Nuclear Data Files, part B (ENDF/B-VII238-neutron energy group data library. Furthermore, two geometrical models were established based on the General Electric (GE14 10 × 10 boiling water reactor fuel assembly and the Generic Burnup-Credit (GBC-68 storage cask. The results revealed that the curves of keff versus B, due to single and compound effects, can be approximated using a first degree polynomial of B. However, the reactivity deviation (or changes of keff,Δk in some compound effects was not a summation of the all Δk resulting from the two associated single effects. This phenomenon is undesirable because it may to some extent affect the precise assessment of burnup credit. In this study, a general formula was thus proposed to express the curves of keff versus B for both single and compound effects.

  9. Analysis of dust samples collected from spent nuclear fuel interim storage containers at Hope Creek, Delaware, and Diablo Canyon, California

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Enos, David George [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-07-01

    Potentially corrosive environments may form on the surface of spent nuclear fuel dry storage canisters by deliquescence of deposited dusts. To assess this, samples of dust were collected from in-service dry storage canisters at two near-marine sites, the Hope Creek and Diablo Canyon storage installations, and have been characterized with respect to mineralogy, chemistry, and texture. At both sites, terrestrially-derived silicate minerals, including quartz, feldspars, micas, and clays, comprise the largest fraction of the dust. Also significant at both sites were particles of iron and iron-chromium metal and oxides generated by the manufacturing process. Soluble salt phases were minor component of the Hope Creek dusts, and were compositionally similar to inland salt aerosols, rich in calcium, sulfate, and nitrate. At Diablo Canyon, however, sea-salt aerosols, occurring as aggregates of NaCl and Mg-sulfate, were a major component of the dust samples. The seasalt aerosols commonly occurred as hollow spheres, which may have formed by evaporation of suspended aerosol seawater droplets, possibly while rising through the heated annulus between the canister and the overpack. The differences in salt composition and abundance for the two sites are attributed to differences in proximity to the open ocean and wave action. The Diablo Canyon facility is on the shores of the Pacific Ocean, while the Hope Creek facility is on the shores of the Delaware River, several miles from the open ocean.

  10. Pelletised fuel production from coal tailings and spent mushroom compost - Part II. Economic feasibility based on cost analysis

    International Nuclear Information System (INIS)

    Ryu, Changkook; Khor, Adela; Sharifi, Vida N.; Swithenbank, Jim

    2008-01-01

    Due to the growing market for sustainable energy, in order to increase the quality of the fuels, pellets are being produced from various materials such as wood and other biomass energy crops, and municipal waste. This paper presents the results from an economic feasibility study for pellet production using blends of two residue materials: coal tailings from coal cleaning and spent mushroom compost (SMC) from mushroom production. Key variables such as the mixture composition, raw material haulage and plant scale were considered and the production costs were compared to coal and biomass energy prices. For both wet materials, the moisture content was the critical parameter that influenced the fuel energy costs. The haulage distance of the raw materials was another factor that can pose a high risk. The results showed that the pellet production from the above two materials can be viable when a less energy-intensive drying process is utilised. Potential market outlets and ways to lower the costs are also discussed in this paper. (author)

  11. Measuring the opportunity loss of time spent boarding admitted patients in the emergency department: a multihospital analysis.

    Science.gov (United States)

    Lucas, Raymond; Farley, Heather; Twanmoh, Joseph; Urumov, Andrej; Evans, Bruce; Olsen, Nils

    2009-01-01

    Emergency department (ED) crowding is an international crisis affecting the timeliness and quality of patient care. Boarding of admitted patients in the ED is recognized as a major contributor to ED crowding. The opportunity loss of this time is the benefit or value it could produce if it were used for something else. In crowded EDs, the typical alternative use of this time is to treat patients waiting to be seen. Various ED performance benchmarks related to inpatient boarding have been proposed, but they are not commonly reported and have yet to be evaluated to determine whether they correlate with the opportunity loss of time used for boarding. This study quantified several measures of ED boarding in a variety of hospital settings and looked for correlations between them and the opportunity loss of the time spent on boarding. In particular, average boarding time per admission was found to be easy to measure. Results revealed that it had a near-perfect linear correlation with opportunity loss. The opportunity loss of every 30 minutes of average boarding time equaled the time required to see 3.5 percent of the ED's daily census. For busy hospitals, the opportunity loss allowed sufficient time for staff to be able to see up to 36 additional patients per day. This correlation suggests that average boarding time per admission may be useful in evaluating efforts to reduce ED crowding and improve patient care.

  12. Analysis of alternative transportation methods for radioactive materials shipments including the use of special trains for spent fuel and wastes

    International Nuclear Information System (INIS)

    Smith, D.R.; Luna, R.E.; Taylor, J.M.

    1978-01-01

    Two studies were completed which evaluate the environmental impact of radioactive material transport. The first was a generic study which evaluated all radioactive materials and all transportation modes; the second addressed spent fuel and fuel-cycle wastes shipped by truck, rail and barge. A portion of each of those studies dealing with the change in impact resulting from alternative shipping methods is presented in this paper. Alternatives evaluated in each study were mode shifts, operational constraints, and, in generic case, changes in material properties and package capabilities. Data for the analyses were obtained from a shipper survey and from projections of shipments that would occur in an equilibrium fuel cycle supporting one hundred 1000-MW(e) reactors. Population exposures were deduced from point source radiation formulae using separation distances derived for scenarios appropriate to each shipping mode and to each exposed population group. Fourteen alternatives were investigated for the generic impact case. All showed relatively minor changes in the overall radiological impact. Since the radioactive material transport is estimated to be fewer than 3 latent cancer fatalities (LCF) for each shipment year (compared to some 300,000 yearly cancer fatalities or 5000 LCF's calculated for background radiation using the same radiological effects model), a 15% decrease caused by shifting from passenger air to cargo air is a relatively small effect. Eleven alternatives were considered for the fuel cycle/special train study, but only one produced a reduction in total special train baseline LCF's (.047) that was larger than 5%

  13. On tentative decommissioning cost analysis with specific authentic cost calculations with the application of the Omega code on a case linked to the Intermediate storage facility for spent fuel in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Vasko, Marek; Daniska, Vladimir; Ondra, Frantisek; Bezak, Peter; Kristofova, Kristina; Tatransky, Peter; Zachar, Matej [DECOM Slovakia, spol. s.r.o., J. Bottu 2, SK-917 01 Trnava (Slovakia); Lindskog, Staffan [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    2007-03-15

    The presented report is focused on tentative calculations of basic decommissioning parameters such as costs, manpower and exposure of personnel for activities of older nuclear facility decommissioning in Sweden represented by Intermediate storage facility for spent fuel in Studsvik, by means of calculation code OMEGA. This report continuously follows up two previous projects, which described methodology of cost estimates of decommissioning with an emphasis to derive cost functions for alpha contaminated material and implementation of the advanced decommissioning costing methodology for Intermediate Storage facility for Spent Fuel in Studsvik. The main purpose of the presented study is to demonstrate the trial application of the advanced costing methodology using OMEGA code for Intermediate Storage Facility for Spent Fuel in Studsvik. Basic work packages presented in report are as follows: 1. Analysis and validation input data on Intermediate Storage Facility for Spent Fuel and assemble a database suitable for standardised decommissioning cost calculations including radiological parameters, 2. Proposal of range of decommissioning calculations and define an extent of decommissioning activities, 3. Defining waste management scenarios for particular material waste streams from Intermediate Storage Facility for Spent Fuel, 4. Developing standardised cost calculation structure applied for Intermediate Storage Facility for Spent Fuel decommissioning calculation and 5. Performing tentative decommissioning calculations for Intermediate Storage Facility for Spent Fuel by OMEGA code. Calculated parameters of decommissioning are presented in structure according to Proposed Standardized List of Items for Costing Purposes. All parameters are documented and summed up in both table and graphic forms in text and Annexes. The presented report documents availability and applicability of methodology for evaluation of costs and other parameters of decommissioning in a form implemented

  14. On tentative decommissioning cost analysis with specific authentic cost calculations with the application of the Omega code on a case linked to the Intermediate storage facility for spent fuel in Sweden

    International Nuclear Information System (INIS)

    Vasko, Marek; Daniska, Vladimir; Ondra, Frantisek; Bezak, Peter; Kristofova, Kristina; Tatransky, Peter; Zachar, Matej; Lindskog, Staffan

    2007-03-01

    The presented report is focused on tentative calculations of basic decommissioning parameters such as costs, manpower and exposure of personnel for activities of older nuclear facility decommissioning in Sweden represented by Intermediate storage facility for spent fuel in Studsvik, by means of calculation code OMEGA. This report continuously follows up two previous projects, which described methodology of cost estimates of decommissioning with an emphasis to derive cost functions for alpha contaminated material and implementation of the advanced decommissioning costing methodology for Intermediate Storage facility for Spent Fuel in Studsvik. The main purpose of the presented study is to demonstrate the trial application of the advanced costing methodology using OMEGA code for Intermediate Storage Facility for Spent Fuel in Studsvik. Basic work packages presented in report are as follows: 1. Analysis and validation input data on Intermediate Storage Facility for Spent Fuel and assemble a database suitable for standardised decommissioning cost calculations including radiological parameters, 2. Proposal of range of decommissioning calculations and define an extent of decommissioning activities, 3. Defining waste management scenarios for particular material waste streams from Intermediate Storage Facility for Spent Fuel, 4. Developing standardised cost calculation structure applied for Intermediate Storage Facility for Spent Fuel decommissioning calculation and 5. Performing tentative decommissioning calculations for Intermediate Storage Facility for Spent Fuel by OMEGA code. Calculated parameters of decommissioning are presented in structure according to Proposed Standardized List of Items for Costing Purposes. All parameters are documented and summed up in both table and graphic forms in text and Annexes. The presented report documents availability and applicability of methodology for evaluation of costs and other parameters of decommissioning in a form implemented

  15. Bioleaching of valuable metals from spent lithium-ion mobile phone batteries using Aspergillus niger

    Science.gov (United States)

    Horeh, N. Bahaloo; Mousavi, S. M.; Shojaosadati, S. A.

    2016-07-01

    In this paper, a bio-hydrometallurgical route based on fungal activity of Aspergillus niger was evaluated for the detoxification and recovery of Cu, Li, Mn, Al, Co and Ni metals from spent lithium-ion phone mobile batteries under various conditions (one-step, two-step and spent medium bioleaching). The maximum recovery efficiency of 100% for Cu, 95% for Li, 70% for Mn, 65% for Al, 45% for Co, and 38% for Ni was obtained at a pulp density of 1% in spent medium bioleaching. The HPLC results indicated that citric acid in comparison with other detected organic acids (gluconic, oxalic and malic acid) had an important role in the effectiveness of bioleaching using A. niger. The results of FTIR, XRD and FE-SEM analysis of battery powder before and after bioleaching process confirmed that the fungal activities were quite effective. In addition, bioleaching achieved higher removal efficiency for heavy metals than the chemical leaching. This research demonstrated the great potential of bio-hydrometallurgical route to recover heavy metals from spent lithium-ion mobile phone batteries.

  16. A Meta-Analysis of the Effectiveness of English-Medium Education in Hong Kong

    Science.gov (United States)

    Lo, Yuen Yi; Lo, Eric Siu Chung

    2014-01-01

    To facilitate second language learning, it has become increasingly popular to use a second language as the medium of instruction for content subjects for majority language students. Although numerous research studies have shown the advantages of such kind of programs in North America and Europe, those investigating English as the Medium of…

  17. Spent fuel storage and isolation

    International Nuclear Information System (INIS)

    Bensky, M.S.; Kurzeka, W.J.; Bauer, A.A.; Carr, J.A.; Matthews, S.C.

    1979-02-01

    The principal spent fuel activities conducted within the commercial waste and spent fuel within the Commercial Waste and Spent Fuel Packaging Program are: simulated near-surface (drywell) storage demonstrations at Hanford and the Nevada Test Site; surface (sealed storage cask) and drywell demonstrations at the Nevada Test Site; and spent fuel receiving and packaging facility conceptual design. These investigations are described

  18. HANSF 1.3 Users Manual FAI/98-40-R2 Hanford Spent Nuclear Fuel (SNF) Safety Analysis Model [SEC 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    DUNCAN, D.R.

    1999-10-07

    The HANSF analysis tool is an integrated model considering phenomena inside a multi-canister overpack (MCO) spent nuclear fuel container such as fuel oxidation, convective and radiative heat transfer, and the potential for fission product release. This manual reflects the HANSF version 1.3.2, a revised version of 1.3.1. HANSF 1.3.2 was written to correct minor errors and to allow modeling of condensate flow on the MCO inner surface. HANSF 1.3.2 is intended for use on personal computers such as IBM-compatible machines with Intel processors running under Lahey TI or digital Visual FORTRAN, Version 6.0, but this does not preclude operation in other environments.

  19. Spent Fuel in Chile

    International Nuclear Information System (INIS)

    López Lizana, F.

    2015-01-01

    The government has made a complete and serious study of many different aspects and possible road maps for nuclear electric power with strong emphasis on safety and energy independence. In the study, the chapter of SFM has not been a relevant issue at this early stage due to the fact that it has been left for later implementation stage. This paper deals with the options Chile might consider in managing its Spent Fuel taking into account foreign experience and factors related to safety, economics, public acceptance and possible novel approaches in spent fuel treatment. The country’s distinctiveness and past experience in this area taking into account that Chile has two research reactors which will have an influence in the design of the Spent Fuel option. (author)

  20. Spent fuel pyroprocessing demonstration

    International Nuclear Information System (INIS)

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

    1995-01-01

    A major element of the shutdown of the US liquid metal reactor development program is managing the sodium-bonded spent metallic fuel from the Experimental Breeder Reactor-II to meet US environmental laws. Argonne National Laboratory has refurbished and equipped an existing hot cell facility for treating the spent fuel by a high-temperature electrochemical process commonly called pyroprocessing. Four products will be produced for storage and disposal. Two high-level waste forms will be produced and qualified for disposal of the fission and activation products. Uranium and transuranium alloys will be produced for storage pending a decision by the US Department of Energy on the fate of its plutonium and enriched uranium. Together these activities will demonstrate a unique electrochemical treatment technology for spent nuclear fuel. This technology potentially has significant economic and technical advantages over either conventional reprocessing or direct disposal as a high-level waste option

  1. Spent nuclear fuel disposal liability insurance

    International Nuclear Information System (INIS)

    Martin, D.W.

    1984-01-01

    This thesis examines the social efficiency of nuclear power when the risks of accidental releases of spent fuel radionuclides from a spent fuel disposal facility are considered. The analysis consists of two major parts. First, a theoretical economic model of the use of nuclear power including the risks associated with releases of radionuclides from a disposal facility is developed. Second, the costs of nuclear power, including the risks associated with a radionuclide release, are empirically compared to the costs of fossil fuel-fired generation of electricity. Under the provisions of the Nuclear Waste Policy Act of 1982, the federally owned and operated spent nuclear fuel disposal facility is not required to maintain a reserve fund to cover damages from an accidental radionuclide release. Thus, the risks of a harmful radionuclide release are not included in the spent nuclear fuel disposal fee charged to the electric utilities. Since the electric utilities do not pay the full, social costs of spent fuel disposal, they use nuclear fuel in excess of the social optimum. An insurance mechanism is proposed to internalize the risks associated with spent fueled disposal. Under this proposal, the Federal government is required to insure the disposal facility against any liabilities arising from accidental releases of spent fuel radionuclides

  2. Guidebook on spent fuel storage

    International Nuclear Information System (INIS)

    1984-01-01

    The Guidebook summarizes the experience and information in various areas related to spent fuel storage: technological aspects, the transport of spent fuel, economical, regulatory and institutional aspects, international safeguards, evaluation criteria for the selection of a specific spent fuel storage concept, international cooperation on spent fuel storage. The last part of the Guidebook presents specific problems on the spent fuel storage in the United Kingdom, Sweden, USSR, USA, Federal Republic of Germany and Switzerland

  3. Spent fuel reprocessing options

    International Nuclear Information System (INIS)

    2008-08-01

    The objective of this publication is to provide an update on the latest developments in nuclear reprocessing technologies in the light of new developments on the global nuclear scene. The background information on spent fuel reprocessing is provided in Section One. Substantial global growth of nuclear electricity generation is expected to occur during this century, in response to environmental issues and to assure the sustainability of the electrical energy supply in both industrial and less-developed countries. This growth carries with it an increasing responsibility to ensure that nuclear fuel cycle technologies are used only for peaceful purposes. In Section Two, an overview of the options for spent fuel reprocessing and their level of development are provided. A number of options exist for the treatment of spent fuel. Some, including those that avoid separation of a pure plutonium stream, are at an advanced level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. In Section Three, research and development in support of advanced reprocessing options is described. Next-generation spent fuel reprocessing plants are likely to be based on aqueous extraction processes that can be designed to a country specific set of spent fuel partitioning criteria for recycling of fissile materials to advanced light water reactors or fast spectrum reactors. The physical design of these plants must incorporate effective means for materials accountancy, safeguards and physical protection. Section four deals with issues and challenges related to spent fuel reprocessing. The spent fuel reprocessing options assessment of economics, proliferation resistance, and environmental impact are discussed. The importance of public acceptance for a reprocessing strategy is discussed. A review of modelling tools to support the

  4. 3D distribution of interstellar medium in the Galaxy: Preparation for analysis of Gaia observations

    Energy Technology Data Exchange (ETDEWEB)

    Puspitarini, Lucky, E-mail: rosine.lallement@obspm.fr [GEPI Observatoire de Paris, CNRS, Paris Diderot University, 5 Place Jules Janssen, 92190, Meudon (France); Bosscha Observatory and Department of Astronomy, FMIPA, Institut Teknologi Bandung, Jl. Ganesha 10, Bandung 40132 (Indonesia); Lallement, Rosine, E-mail: rosine.lallement@obspm.fr [GEPI Observatoire de Paris, CNRS, Paris Diderot University, 5 Place Jules Janssen, 92190, Meudon (France)

    2015-09-30

    Accurate and detailed three-dimensional (3D) maps of Galactic interstellar medium (ISM) are still lacking. One way to obtain such 3D descriptions is to record a large set of individual absorption or reddening measurements toward target stars located at various known distances and directions. The inversion of these measurements using a tomographic method can produce spatial distribution of the ISM. Until recently absorption data were very limited and distances to the target stars are still uncertain, but the situation will greatly improve thanks to current and future massive stellar surveys from ground, and to Gaia mission. To prepare absorption data for inversion from a huge number of stellar spectra, automated tools are needed. We have developed various spectral analysis tools adapted to different type of spectra, early- or late- type star. We also have used diffuse interstellar bands (DIBs) to trace IS structures and kinematics. Although we do not know yet their carriers, they can be a promising tool to trace distant interstellar clouds or Galactic arms. We present some examples of the interstellar fitting and show the potentiality of DIBs in tracing the ISM. We will also briefly show and comment the latest 3D map of the local ISM which reveal nearby cloud complexes and cavities.

  5. Analysis of optical scheme for medium-range directed energy laser weapon system

    Science.gov (United States)

    Jabczyński, Jan K.; Kaśków, Mateusz; Gorajek, Łukasz; Kopczyński, Krzysztof

    2017-10-01

    The relations between range of operation and aperture of laser weapon system were investigated, taking into account diffraction and technical limitations as beam quality, accuracy of point tracking, technical quality of optical train, etc. As a result for the medium ranges of 1 - 2 km we restricted the analysis to apertures not wider than 150 mm and the optical system without adaptive optics. To choose the best laser beam shape, the minimization of aperture losses and thermooptical effects inside optics as well as the effective width of laser beam in far field should be taken into account. We have analyzed theoretically such a problem for the group of a few most interesting from that point of view profiles including for reference two limiting cases of Gaussian beam and `top hat' profile. We have found that the most promising is the SuperGaussian profile of index p = 2 for which the surfaces of beam shaper elements can be manufactured in the acceptable cost-effective way and beam quality does not decrease noticeably. Further, we have investigated the thermo-optic effects on the far field parameters of Gaussian and `top hat' beams to determine the influence of absorption in optical elements on beam quality degradation. The simplified formulae were derived for beam quality measures (parameter M2 and Strehl ratio) which enables to estimate the influence of absorption losses on degradation of beam quality.

  6. Parametric analysis of the growth of colloidal ZnO nanoparticles synthesized in alcoholic medium

    Energy Technology Data Exchange (ETDEWEB)

    Fonseca, A. S. [National Research Centre for the Working Environment (Denmark); Figueira, P. A.; Pereira, A. S. [Universidade de Aveiro, Departamento de Química—CICECO (Portugal); Santos, R. J. [Universidade do Porto, Laboratory of Separation and Reaction Engineering-Laboratory of Catalysis and Materials (LSRE-LCM), Faculdade de Engenharia (Portugal); Trindade, T. [Universidade de Aveiro, Departamento de Química—CICECO (Portugal); Nunes, M. I., E-mail: isanunes@ua.pt [Universidade de Aveiro, Centre for Environmental and Marine Studies (CESAM), Dep. de Ambiente e Ordenamento (Portugal)

    2017-02-15

    The growth kinetics of nanosized ZnO was studied considering the influence of different parameters (mixing degree, temperature, alcohol chain length, reactant concentration and Zn/OH ratios) on the synthesis reaction and modelling the outputs using typical kinetic growth models, which were then evaluated by means of a sensitivity analysis. The Zn/OH ratio, the temperature and the alcohol chain length were found to be essential parameters to control the growth of ZnO nanoparticles, whereas zinc acetate concentration (for Zn/OH = 0.625) and the stirring during the ageing stage were shown to not have significant influence on the particle size growth. This last operational parameter was for the first time investigated for nanoparticles synthesized in 1-pentanol, and it is of outmost importance for the implementation of continuous industrial processes for mass production of nanosized ZnO and energy savings in the process. Concerning the nanoparticle growth modelling, the results show a different pattern from the more commonly accepted diffusion-limited Ostwald ripening process, i.e. the Lifshitz–Slyozov–Wagner (LSW) model. Indeed, this study shows that oriented attachment occurs during the early stages whereas for the later stages the particle growth is well represented by the LSW model. This conclusion contributes to clarify some controversy found in the literature regarding the kinetic model which better represents the ZnO NPs’ growth in alcoholic medium.

  7. Electric Field Simulations and Analysis for High Voltage High Power Medium Frequency Transformer

    Directory of Open Access Journals (Sweden)

    Pei Huang

    2017-03-01

    Full Text Available The electronic power transformer (EPT raises concerns for its notable size and volume reduction compared with traditional line frequency transformers. Medium frequency transformers (MFTs are important components in high voltage and high power energy conversion systems such as EPTs. High voltage and high power make the reliable insulation design of MFT more difficult. In this paper, the influence of wire type and interleaved winding structure on the electric field distribution of MFT is discussed in detail. The electric field distributions for six kinds of typical non-interleaved windings with different wire types are researched using a 2-D finite element method (FEM. The electric field distributions for one non-interleaved winding and two interleaved windings are also studied using 2-D FEM. Furthermore, the maximum electric field intensities are obtained and compared. The results show that, in this case study, compared with foil conductor, smaller maximum electric field intensity can be achieved using litz wire in secondary winding. Besides, interleaving can increase the maximum electric field intensity when insulation distance is constant. The proposed method of studying the electric field distribution and analysis results are expected to make a contribution to the improvement of electric field distribution in transformers.

  8. 3D distribution of interstellar medium in the Galaxy: Preparation for analysis of Gaia observations

    International Nuclear Information System (INIS)

    Puspitarini, Lucky; Lallement, Rosine

    2015-01-01

    Accurate and detailed three-dimensional (3D) maps of Galactic interstellar medium (ISM) are still lacking. One way to obtain such 3D descriptions is to record a large set of individual absorption or reddening measurements toward target stars located at various known distances and directions. The inversion of these measurements using a tomographic method can produce spatial distribution of the ISM. Until recently absorption data were very limited and distances to the target stars are still uncertain, but the situation will greatly improve thanks to current and future massive stellar surveys from ground, and to Gaia mission. To prepare absorption data for inversion from a huge number of stellar spectra, automated tools are needed. We have developed various spectral analysis tools adapted to different type of spectra, early- or late- type star. We also have used diffuse interstellar bands (DIBs) to trace IS structures and kinematics. Although we do not know yet their carriers, they can be a promising tool to trace distant interstellar clouds or Galactic arms. We present some examples of the interstellar fitting and show the potentiality of DIBs in tracing the ISM. We will also briefly show and comment the latest 3D map of the local ISM which reveal nearby cloud complexes and cavities

  9. Parametric analysis of the thermodynamic properties for a medium with strong interaction between particles

    International Nuclear Information System (INIS)

    Dubovitskii, V.A.; Pavlov, G.A.; Krasnikov, Yu.G.

    1996-01-01

    Thermodynamic analysis of media with strong interparticle (Coulomb) interaction is presented. A method for constructing isotherms is proposed for a medium described by a closed multicomponent thermodynamic model. The method is based on choosing an appropriate nondegenerate frame of reference in the extended space of thermodynamic variables and provides efficient thermodynamic calculations in a wide range of parameters, for an investigation of phase transitions of the first kind, and for determining both the number of phases and coexistence curves. A number of approximate thermodynamic models of hydrogen plasma are discussed. The approximation corresponding to the n5/2 law, in which the effects of particle attraction and repulsion are taken into account qualitatively, is studied. This approximation allows studies of thermodynamic properties of a substance for a wide range of parameters. In this approximation, for hydrogen at a constant temperature, various properties of the degree of ionization are revealed. In addition, the parameters of the second critical point are found under conditions corresponding to the Jovian interior

  10. Milestones for Selection, Characterization, and Analysis of the Performance of a Repository for Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain.

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, Robert P.

    2014-02-01

    This report presents a concise history in tabular form of events leading up to site identification in 1978, site selection in 1987, subsequent characterization, and ongoing analysis through 2008 of the performance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain in southern Nevada. The tabulated events generally occurred in five periods: (1) commitment to mined geologic disposal and identification of sites; (2) site selection and analysis, based on regional geologic characterization through literature and analogous data; (3) feasibility analysis demonstrating calculation procedures and importance of system components, based on rough measures of performance using surface exploration, waste process knowledge, and general laboratory experiments; (4) suitability analysis demonstrating viability of disposal system, based on environment-specific laboratory experiments, in-situ experiments, and underground disposal system characterization; and (5) compliance analysis, based on completed site-specific characterization. Because the relationship is important to understanding the evolution of the Yucca Mountain Project, the tabulation also shows the interaction between four broad categories of political bodies and government agencies/institutions: (a) technical milestones of the implementing institutions, (b) development of the regulatory requirements and related federal policy in laws and court decisions, (c) Presidential and agency directives and decisions, and (d) critiques of the Yucca Mountain Project and pertinent national and world events related to nuclear energy and radioactive waste.

  11. Technical and economic analysis of integrating low-medium temperature solar energy into power plant

    International Nuclear Information System (INIS)

    Wang, Fu; Li, Hailong; Zhao, Jun; Deng, Shuai; Yan, Jinyue

    2016-01-01

    Highlights: • Seven configurations were studied regarding the integration of solar thermal energy. • Economic analysis was conducted on new built plants and retrofitted power plants. • Using solar thermal energy to preheat high pressure feedwater shows the best performance. - Abstract: In order to mitigate CO 2 emission and improve the efficiency of the utilization of solar thermal energy (STE), solar thermal energy is proposed to be integrated into a power plant. In this paper, seven configurations were studied regarding the integration of STE. A 300 MWe subcritical coal-fired plant was selected as the reference, chemical absorption using monoethanolamine solvent was employed for CO 2 ​capture, and parabolic trough collectors and evacuated tube collectors were used for STE collection. Both technical analysis and economic evaluation were conducted. Results show that integrating solar energy with post-combustion CO 2 ​ capture can effectively increase power generation and reduce the electrical efficiency penalty caused by CO 2 capture. Among the different configurations, Config-2 and Config-6, which use medium temperature STE to replace high pressure feedwater without and with CO 2 capture, show the highest net incremental solar efficiency. When building new plants, integrating solar energy can effectively reduce the levelized cost of electricity (LCOE). The lowest LCOE, 99.28 USD/MWh, results from Config-6, with a parabolic trough collector price of 185 USD/m 2 . When retrofitting existing power plants, Config-6 also shows the highest net present value (NPV), while Config-2 has the shortest payback time at a carbon tax of 50 USD/ton CO 2 . In addition, both LCOE and NPV/payback time are clearly affected by the relative solar load fraction, the price of solar thermal collectors and the carbon tax. Comparatively, the carbon tax can affect the configurations with CO 2 capture more clearly than those without CO 2 capture.

  12. Graduate Management Project (GMP) Retrospective Analysis of Promotional Mediums for Tricare Prime in Tricare Region 11

    National Research Council Canada - National Science Library

    Carpenter, Steven

    1997-01-01

    This study provides retrospective market research information about the population who enrolled in TRICARE Prime in TRICARE Region 11 and the advertising mediums used to promote enrollment in the TRICARE Prime program...

  13. Genome-scale metabolic flux analysis of Streptomyces lividans growing on a complex medium.

    Science.gov (United States)

    D'Huys, Pieter-Jan; Lule, Ivan; Vercammen, Dominique; Anné, Jozef; Van Impe, Jan F; Bernaerts, Kristel

    2012-09-15

    Constraint-based metabolic modeling comprises various excellent tools to assess experimentally observed phenotypic behavior of micro-organisms in terms of intracellular metabolic fluxes. In combination with genome-scale metabolic networks, micro-organisms can be investigated in much more detail and under more complex environmental conditions. Although complex media are ubiquitously applied in industrial fermentations and are often a prerequisite for high protein secretion yields, such multi-component conditions are seldom investigated using genome-scale flux analysis. In this paper, a systematic and integrative approach is presented to determine metabolic fluxes in Streptomyces lividans TK24 grown on a nutritious and complex medium. Genome-scale flux balance analysis and randomized sampling of the solution space are combined to extract maximum information from exometabolome profiles. It is shown that biomass maximization cannot predict the observed metabolite production pattern as such. Although this cellular objective commonly applies to batch fermentation data, both input and output constraints are required to reproduce the measured biomass production rate. Rich media hence not necessarily lead to maximum biomass growth. To eventually identify a unique intracellular flux vector, a hierarchical optimization of cellular objectives is adopted. Out of various tested secondary objectives, maximization of the ATP yield per flux unit returns the closest agreement with the maximum frequency in flux histograms. This unique flux estimation is hence considered as a reasonable approximation for the biological fluxes. Flux maps for different growth phases show no active oxidative part of the pentose phosphate pathway, but NADPH generation in the TCA cycle and NADPH transdehydrogenase activity are most important in fulfilling the NADPH balance. Amino acids contribute to biomass growth by augmenting the pool of available amino acids and by boosting the TCA cycle, particularly

  14. Reprocessing of spent plasma

    International Nuclear Information System (INIS)

    Pierini, G.

    1981-01-01

    This invention relates to a process for removing helium and other impurities from a mixture containing deuterium and tritium, a deuterium/tritium mixture when purified in accordance with such a process and, more particularly, to a process for the reprocessing of spent plasma removed from a thermofusion reactor. (U.K.)

  15. Spent fuel counter

    International Nuclear Information System (INIS)

    Drayer, D.D.

    1988-09-01

    In many cases the IAEA must inspect spent fuel shipping casks before they leave facilities. Similarly, inspections may be required at the location where a cask is received and unloaded. In order to reduce the number of inspections required, it would be desirable to develop a system to count spent fuel assemblies as they are loaded or removed from shipping casks. This report discusses several methods which potentially could be used for performing this function. A concept for a Spent Fuel Counter System is proposed which uses a Laser Surveillance System (LASSY), Cerenkov Viewing Device (CVD), and Modular Integrated Video System (MIVS), all coupled together. In the proposed system, LASSY would provide an indication that an object is being placed into or removed from the cask, the CVD would be used to determine if the object has the radiation characteristics of a spent fuel assembly, and the MIVS would record the information. The system may need to be designed so that the operator could determine that it was operating correctly during the loading operations. This would help prevent anomalies from occurring which could only be resolved through reverification measures. Before such a system could be implemented testing would be necessary to determine that the individual components would each work adequately in this application. The issues of reliability, intrusiveness, and cask sealing should also be addressed before a development program is undertaken. 12 refs., 1 fig

  16. Time well spent

    DEFF Research Database (Denmark)

    Fallesen, Peter

    2013-01-01

    Individuals who spent time in foster care as children fare on average worse than non-placed peers in early adult life. Recent research on the effect of foster care placement on early adult life outcomes provides mixed evidence. Some studies suggest negative effects of foster care placement on early...

  17. Importance and performance evaluation tools for small and medium companies: critical analysis of national versus international literature

    Directory of Open Access Journals (Sweden)

    Sandro César Bortoluzzi

    2015-12-01

    Full Text Available The research aims to map the importance and performance evaluation tools for small and medium companies. This descriptive and qualitative study analyzed 33 national articles and 21 international ones. Regarding the importance of performance evaluation for small and medium companies, the literature highlights: (i it increases the success of the network; (ii it is useful for management; (iii it strengthens competitiveness; (iv it consolidates cooperation; and, (v it increases trust among partners. Comparing the national versus international literature on the importance of performance evaluation for small and medium companies, it can be noticed similar and complementary aspects, that is, there is not disagreement between the authors. The authors use tools consolidated in the literature, such as Balanced Scorecard; Benchmarking; Performance Prism and tools proposed specifically to evaluate small and medium networks. The main dimensions evaluated are: (i exchange of information; (ii value management in networks; (Iii level of network maturity; (iv benefits of collaboration; (v social capital; (vi collective efficiency; (vii network life cycle; (viii efficiency and inefficiency of the networks; and, (ix existence and intensity of the relationship between partners. The critical analysis regarding the performance evaluation concept adopted in the present study shows that the tools proposed or implemented to evaluate small and medium business networks have gaps in the process to identify criteria, measure ordinal and cardinally, integrate and generate actions of improvement.

  18. HTGR spent fuel storage study

    International Nuclear Information System (INIS)

    Burgoyne, R.M.; Holder, N.D.

    1979-04-01

    This report documents a study of alternate methods of storing high-temperature gas-cooled reactor (HTGR) spent fuel. General requirements and design considerations are defined for a storage facility integral to a fuel recycle plant. Requirements for stand-alone storage are briefly considered. Three alternate water-cooled storage conceptual designs (plug well, portable well, and monolith) are considered and compared to a previous air-cooled design. A concept using portable storage wells in racks appears to be the most favorable, subject to seismic analysis and economic evaluation verification

  19. Numerical analysis of road pavement thermal deformability, based on Biot viscoelastic model of porous medium

    Directory of Open Access Journals (Sweden)

    Bartlewska-Urban Monika

    2016-03-01

    Full Text Available The following study presents numerical calculations for establishing the impact of temperature changes on the process of distortion of bi-phase medium represented using Biot consolidation equations with Kelvin–Voigt rheological skeleton presented, on the example of thermo-consolidation of a pavement of expressway S17. We analyzed the behavior of the expressway under the action of its own weight, dynamic load caused by traffic and temperature gradient. This paper presents the application of the Biot consolidation model with the Kelvin–Voigt skeleton rheological characteristics and the influence of temperature on the deformation process is taken into account. A three-dimensional model of the medium was created describing the thermal consolidation of a porous medium. The 3D geometrical model of the area under investigation was based on data obtained from the land surveying and soil investigation of a 200 m long section of the expressway and its shoulders.

  20. Surface plasmon enhancement in gold nanoparticles in the presence of an optical gain medium: an analysis

    Energy Technology Data Exchange (ETDEWEB)

    Sathiyamoorthy, K; Sreekanth, K V; Sidharthan, R; Murukeshan, V M [School of Mechanical and Aerospace Engineering, Nanyang Technological University, 50 Nanyang Avenue, Singapore 639798 (Singapore); Xing Bengang, E-mail: mmurukeshan@ntu.edu.sg [Division of Chemistry and Biological Chemistry, School of Physical and Mathematical Sciences, Nanyang Technological University, 21 Nanyang Link, Singapore 637371 (Singapore)

    2011-10-26

    The localized surface plasmon (LSP) enhancement in a gold nanoparticle is demonstrated in this paper. The enhancement of LSP is influenced by both size and the dielectric gain medium surrounding the nanoparticles. The nanoparticle is found to induce plasmonic enhancement of varying degrees depending on its size, and it is inferred that a gold nanoparticle of size 60 nm exhibits the maximum LSP for 532 nm excitation. Singularity due to cancellation of SP loss by an infinite gain medium and LSP enhancement are studied using a pump-probe Rayleigh scattering experiment. Gold nanoparticles of average size 60 nm exhibit the lowest threshold power to observe Rayleigh scattering. Furthermore, compared with the bare nanoparticles, a 12.5 fold enhancement of LSP is observed when the nanoparticle of average size 60 nm is kept in the gain medium.

  1. Analysis of technological innovation strategy for small and medium companies of the aeronautical sector

    Directory of Open Access Journals (Sweden)

    Marcela Barbosa de Moraes

    2010-08-01

    Full Text Available The inherent risk in high-tech activity requires the construction of technological strategies that serve global strategy of the company. The present study aimed to characterize the technological position of small and medium technology-based companies of the aeronautical sector located in the Vale do Paraíba, and to examine whether these companies have a technological strategy formalized and disseminated. The adopted methodology was a descriptive exploratory research, carried out through in-depth individual interviews conducted with the small e medium company’s owners of the cities of Caçapava, São José dos Campos and Taubaté, Brazil. The authors concluded that it is necessary that companies, especially small and medium-sized, adopt a technological innovation strategy integrated with the company’s overall strategy. This will help keeping them competitive within their specificities, and not only in the domestic market, but also in international markets.

  2. Quantitative analysis on the effects of components of MS medium using PIXE technique

    International Nuclear Information System (INIS)

    Li Yao; Li Hua; Shen Daleng; Yao Huiying; Zeng Xianzhou; Xu Weidong; Yang Fujia

    1993-01-01

    The PIXE system with a detective sensitivity of 10 -11 g for elements was used to study the quantitative effects of the components of Murashige and Skoog (M and S) salts. Concentrations of 14 elements in Polygonum tenuicaule cultured in M and S medium with different modifications were analysed. The physiological function of M and S salts and such integral components as KH 2 PO 4 , FeSO 4 , KNO 3 and CaCl 2 were disclosed. Some features of M and S medium and plant absorption mechanism are reported for the first time. (author)

  3. Encapsulating spent nuclear fuel

    International Nuclear Information System (INIS)

    Fleischer, L.R.; Gunasekaran, M.

    1979-01-01

    A system is described for encapsulating spent nuclear fuel discharged from nuclear reactors in the form of rods or multi-rod assemblies. The rods are completely and contiguously enclosed in concrete in which metallic fibres are incorporated to increase thermal conductivity and polymers to decrease fluid permeability. This technique provides the advantage of acceptable long-term stability for storage over the conventional underwater storage method. Examples are given of suitable concrete compositions. (UK)

  4. Study of molybdenum (VI) complexation and precipitation by zirconium (IV) in strongly acid medium. Application to nuclear spent fuel dissolution; Etude de la complexation et de la precipitation du molybdene (VI) par le zirconium (IV) en milieu tres acide. Application a la dissolution du combustible nucleaire irradie

    Energy Technology Data Exchange (ETDEWEB)

    Esbelin, E

    1999-07-01

    These last years the formation of solid deposits has been observed in the dissolution workshops of the La Hague plant. A sample of the solid was withdrawn for expertise: molybdenum and zirconium are the two major components of the solid, identified as zirconium molybdate. This thesis consisted in the approach of the mechanisms in solution liable to induce precipitate formation. After a bibliographical overview on the chemistry of Mo(VI) in highly acidic solution, this system was studied by absorption spectrophotometry in perchloric medium. The implication of two major forms of Mo(VI) in a dimerization equilibrium was confirmed by this way and by {sup 95}Mo NMR. The principal parameters governing this equilibrium were identified. It is thus shown that the molybdenum dimerization reaction is exothermic. Disturbance of the Mo(VI) system in highly acidic solution by Zr(IV) was also studied. In a restricted experimental field, for which 'conventional' exploitation methodologies had to be adapted to the system, a main complex of stoichiometry 1:1 between Mo(VI) and Zr(IV) was found. The precipitation study of Mo(VI) by Zr(IV) under conditions close to those of the dissolution medium of nuclear spent fuel was undertaken. The main parameters which control precipitation kinetics were identified. The results obtained reveal that precipitation is controlled by a single macroscopic process and therefore can be described by a single equation. The solid obtained is composed of only one phase presenting a Mo:Zr non-stoichiometry when compared to the theoretical formula ZrMo{sub 2}O{sub 7}(OH){sub 2},2H{sub 2}O. At last, on the basis of the research results, a descriptive mechanism of the system is proposed in which intervenes a 1:1 intermediate complex, much more soluble than a probable 2:1 precipitation precursor. (author)

  5. Design management and stress analysis of a circular rock tunnel and emplacement holes for storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Kandalaft-Ladkany, N.; Wyman, R.V.

    1992-01-01

    This paper discusses a critical path method (CPM) diagram and logic net which are used for the design cycle of the rock tunnel system for a high level nuclear waste repository. In the analysis the design tunnel is subjected to pre-existing temperature and overburden loads at time of construction. high thermal stresses develop later due to the long term influx of heat from the canisters stored in vertical emplacement holes. Results indicate that thermal stresses reach a critical level for the rock in the vicinity of the canisters which could lead to local collapse of the rock and damage to the canisters

  6. Analysis of radiation doses from operation of postulated commercial spent fuel transportation systems: Analysis of a system containing a monitored retrievable storage facility

    International Nuclear Information System (INIS)

    Smith, R.I.; Daling, P.M.; Faletti, D.W.

    1992-04-01

    This addendum report extends the original study of the estimated radiation doses to the public and to workers resulting from transporting spent nuclear fuel from commercial nuclear power reactor stations through the federal waste management system (FWMS), to a system that contains a monitored retrievable storage (MRS) facility. The system concepts and designs utilized herein are consistent with those used in the original study (circa 1985--1987). Because the FWMS design is still evolving, the results of these analyses may no longer apply to the design for casks and cask handling systems that are currently being considered. Four system scenarios are examined and compared with the reference No-MRS scenario (all spent fuel transported directly from the reactors to the western repository in standard-capacity truck and rail casks). In Scenarios 1 and 2, an MRS facility is located in eastern United States and ships either intact fuel assemblies or consolidated fuel rods and compacted assembly hardware in canisters. In Scenarios 3 and 4, an MRS facility is located in the western United States and ship either intact fuel assemblies or consolidated fuel rods and compacted assembly hardware in canisters

  7. Feasibility of fissile mass assay of spent nuclear fuel using 252Cf-source-driven frequency-analysis

    International Nuclear Information System (INIS)

    Mattingly, J.K.; Valentine, T.E.; Mihalczo, J.T.

    1996-01-01

    The feasibility was evaluated using MCNP-DSP, an analog Monte Carlo transport cod to simulate source-driven measurements. Models of an isolated Westinghouse 17x17 PWR fuel assembly in a 1500-ppM borated water storage pool were used. In the models, the fuel burnup profile was represented using seven axial burnup zones, each with isotopics estimated by the PDQ code. Four different fuel assemblies with average burnups from fresh to 32 GWd/MTU were modeled and analyzed. Analysis of the fuel assemblies was simulated by inducing fission in the fuel using a 252 Cf source adjacent to the assembly and correlating source fissions with the response of a bank of 3 He detectors adjacent to the assembly opposite the source. This analysis was performed at 7 different axial positions on each of the 4 assemblies, and the source-detector cross-spectrum signature was calculated for each of these 28 simulated measurements. The magnitude of the cross-spectrum signature follows a smooth upward trend with increasing fissile material ( 235 U and 239 Pu) content, and the signature is independent of the concentration of spontaneously fissioning isotopes (e.g., 244 Cm) and (α,n) sources. Furthermore, the cross-spectrum signature is highly sensitive to changes in fissile material content. This feasibility study indicated that the signature would increase ∼100% in response to an increase of only 0.1 g/cm 3 of fissile material

  8. Supplement analysis for a container system for the management of DOE spent nuclear fuel located at the INEEL

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-12

    The Council on Environmental Quality (CEQ) regulations for implementing the NEPA, 40 CFR 1502.9 (c), directs federal agencies to prepare a supplement to an environmental impact statement when an agency makes substantial changes in the Proposed Action that are relevant to environmental concerns, or there are significant new circumstances or information relevant to environmental concerns and bearing on the Proposed Action or impacts. When it is unclear whether a supplemental environmental impact statement is required, DOE regulations (10 CFR 1021.314) direct the preparation of a supplement analysis to assist in making that determination. This supplement analysis evaluates the impacts of employing dual-purpose canisters (DPCs) to prepare DOE SNF located at the INEEL for interim onsite storage and transport outside the State of Idaho. Impacts associated with DPC manufacturing, loading and storage of DOE-ID SNF into DPCs, transport of loaded DPCs outside Idaho, and the cumulative impacts are compared with the impacts previously analyzed in the SNF and INEL EIS and the Navy Container System EIS. This SA provides information to determine whether: (1) an existing EIS should be supplemented; (2) a new EIS should be prepared; or (3) no further NEPA documentation is required.

  9. Analysis of the characteristics of a two-fiber sensor for monitorship of a turbid medium

    International Nuclear Information System (INIS)

    Zaccanti, G.; Bruscaglioni, P.; Luzi, G.

    1987-01-01

    Certain characteristics of a two-fiber sensor device, of the type proposed by Papa et al. for sea water turbidity monitorship, are examined. The extension of medium from which most of the received backscattered power originates is investigated, together with possible effects of multiple scattering on the received power

  10. Computational Analysis of a South African Mobile Trailer-Type Medium Sized Tyre Test Rig

    CSIR Research Space (South Africa)

    Sharma, Shikar

    2015-04-01

    Full Text Available To support the South African National Defence Force with their vehicle mobility needs, the CSIR has begun characterising tyres by using a medium, trailer-type, tyre test rig. Two different Pacejka tyre models were generated using two independent...

  11. Analysis of polarized pulse propagation through one-dimensional scattering medium

    Science.gov (United States)

    Zhang, Yong; Yao, Feng-Ju; Xie, Ming; Yi, Hong-Liang

    2017-08-01

    This paper analyzes the polarized light propagation in a one-dimensional scattering medium with the upper surface subjected to an oblique incident short-pulsed laser beam using the natural element method (NEM). The NEM discretization scheme for the transient vector radiative transfer equation (TVRTE) is presented in detail. The accuracy of the natural element method for transient vector radiative transfer in the scattering medium is assessed. Numerical results show that the NEM is accurate, and effective in solving transient polarized radiative problems. We examine a square short-pulsed laser transport firstly in the atmosphere with Mie scattering and then within aerosol scattering medium. We then investigate the transient polarized radiative transfer problem in the atmosphere-ocean system. The time-resolved signals and the polarization state of the Stokes vector are presented and analyzed. It is found that the scattering types of the medium make greatly influence on the transient transportation of the polarized light. Critically, the polarization states of the backward and forward scattered photons show significantly different time varying trends. For the two-layer system with dissimilar refractive index distributions, due to the total-reflection effect, the existence of a Fresnel interface significantly changes the polarization state of the light, and discontinuous distribution features are observed on the interface.

  12. Thermo-mechanical vibration analysis of annular and circular graphene sheet embedded in an elastic medium

    Directory of Open Access Journals (Sweden)

    M. Mohammadi

    Full Text Available In this study, the vibration behavior of annular and circular graphene sheet coupled with temperature change and under in-plane pre-stressed is studied. Influence of the surrounding elastic medium 011 the fundamental frequencies of the single-layered graphene sheets (SLGSs is investigated. Both Winkler-type and Pasternak- type models are employed to simulate the interaction of the graphene sheets with a surrounding elastic medium. By using the nonlocal elasticity theory the governing equation is derived for SLGSs. The closed-form solution for frequency vibration of circular graphene sheets lias been obtained and nonlocal parameter, inplane pre-stressed, the parameters of elastic medium and temperature change appears into arguments of Bessel functions. The results are subsequently compared with valid result reported in the literature and the molecular dynamics (MD results. The effects of the small scale, pre-stressed, mode number, temperature change, elastic medium and boundary conditions on natural frequencies are investigated. The non-dimensional frequency decreases at high temperature case with increasing the temperature change for all boundary conditions. The effect of temperature change 011 the frequency vibration becomes the opposite at high temperature case in compression with the low temperature case. The present research work thus reveals that the nonlocal parameter, boundary conditions and temperature change have significant effects on vibration response of the circular nanoplates. The present results can be used for the design of the next generation of nanodevices that make use of the thermal vibration properties of the graphene.

  13. Statistical analysis of the established salary in small and medium enterprises

    Directory of Open Access Journals (Sweden)

    Yu. S. Pin’kovetskaya

    2017-01-01

    Full Text Available The aim of the study was to analyze the present regularities, specific to the employees’ salaries of aggregates of small and medium enterprises related to the three dimensional categories and located in different regions of Russia. The following tasks were solved: the indexes, characterizing average monthly salary of employees based on the mentioned enterprises were assessed, belonging to different size categories and located in each of the regions; the relations were established between the average monthly salaries of employees of aggregates of small and medium enterprises and the cost of living in all regions of the country.Preliminary results of stopwatch reading of small and medium business activities in 2015 were used as initial data. The research was based on the comparison of indexes for the entrepreneurial sector and the full range of enterprises and organizations.Modeling differentiation of salaries’ values of small and medium enterprises aggregations, as well as its relationship to the values of the subsistence level was based on the development of the density function of normal distribution. The quality of the developed models was checked according to the Kolmogorov-Smirnov, Pearson and Shapiro-Wilk criteria.The obtained results have some theoretical significance, in particular, when conducting research related to the justification of the proposed wage of employees of enterprises different in number, the formation of measures for increasing efficiency of the entrepreneurial sector activity. Density functions of normal distribution given in the paper can be used in the justification of concepts, plans and programs of developing small and medium entrepreneurship in regions and municipalities. The practical importance of research results connected with the possibility of their use by entrepreneurs directly (especially by beginners when assessing the potential of enterprise creation and definition of employees’ proposed salaries. In

  14. Abbreviated report of the critical analysis of the spent nuclear fuel project's activity-based cost estimate

    International Nuclear Information System (INIS)

    Warren, R.N.

    1998-01-01

    In 1997, the SNFP developed a baseline change request (BCR) and submitted it to DOE-RL for approval. The schedule was formally evaluated to have a 19% probability of success [Williams, 1998]. In December 1997, DOE-RL Manager John Wagoner approved the BCR contingent upon a subsequent independent review of the new baseline. The SNFP took several actions during the first quarter of 1998 to prepare for the independent review. The project developed the Estimating Requirements and Implementation Guide [DESH, 1998] and trained cost account managers (CAMS) and other personnel involved in the estimating process in activity-based cost (ABC) estimating techniques. The SNFP then applied ABC estimating techniques to develop the basis for the December Baseline (DB) and documented that basis in Basis of Estimate (BOE) books. These BOEs were provided to DOE in April 1998. DOE commissioned Professional Analysis, Inc. (PAI) to perform a critical analysis (CA) of the DB. PAI's review formally began on April 13. PAI performed the CA, provided three sets of findings to the SNFP contractor, and initiated reconciliation meetings. During the course of PAI's review, DOE directed the SNFP to develop a new baseline with a higher probability of success. The contractor transmitted the new baseline, which is referred to as the High Probability Baseline (HPB), to DOE on April 15, 1998 [Williams, 1998]. The HPB was estimated to approach a 90% confidence level on the start of fuel movement [Williams, 1998]. This high probability resulted in an increased cost and a schedule extension. To implement the new baseline, the contractor initiated 26 BCRs with supporting BOES. PAI's scope was revised on April 28 to add reviewing the HPB and the associated BCRs and BOES

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

    Science.gov (United States)

    Budi Setiawan, M.; Kuntjoro, S.

    2018-02-01

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

  16. Optimization of Fermentation Medium for the Production of Atrazine Degrading Strain Acinetobacter sp. DNS32 by Statistical Analysis System

    Science.gov (United States)

    Zhang, Ying; Wang, Yang; Wang, Zhi-Gang; Wang, Xi; Guo, Huo-Sheng; Meng, Dong-Fang; Wong, Po-keung

    2012-01-01

    Statistical experimental designs provided by statistical analysis system (SAS) software were applied to optimize the fermentation medium composition for the production of atrazine-degrading Acinetobacter sp. DNS32 in shake-flask cultures. A “Plackett-Burman Design” was employed to evaluate the effects of different components in the medium. The concentrations of corn flour, soybean flour, and K2HPO4 were found to significantly influence Acinetobacter sp. DNS32 production. The steepest ascent method was employed to determine the optimal regions of these three significant factors. Then, these three factors were optimized using central composite design of “response surface methodology.” The optimized fermentation medium composition was composed as follows (g/L): corn flour 39.49, soybean flour 25.64, CaCO3 3, K2HPO4 3.27, MgSO4·7H2O 0.2, and NaCl 0.2. The predicted and verifiable values in the medium with optimized concentration of components in shake flasks experiments were 7.079 × 108 CFU/mL and 7.194 × 108 CFU/mL, respectively. The validated model can precisely predict the growth of atrazine-degraing bacterium, Acinetobacter sp. DNS32. PMID:23093851

  17. Analysis of Dust Samples Collected from an Unused Spent Nuclear Fuel Interim Storage Container at Hope Creek, Delaware.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Enos, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-01

    In July, 2014, the Electric Power Research Institute and industry partners sampled dust on the surface of an unused canister that had been stored in an overpack at the Hope Creek Nuclear Generating Station for approximately one year. The foreign material exclusion (FME) cover that had been on the top of the canister during storage, and a second recently - removed FME cover, were also sampled. This report summarizes the results of analyses of dust samples collected from the unused Hope Creek canister and the FME covers. Both wet and dry samples of the dust/salts were collected, using SaltSmart(TM) sensors and Scotch - Brite(TM) abrasive pads, respectively. The SaltSmart(TM) samples were leached and the leachate analyzed chemically to determine the composition and surface load per unit area of soluble salts present on the canister surface. The dry pad samples were analyzed by X-ray fluorescence and by scanning electron microscopy to determine dust texture and mineralogy; and by leaching and chemical analysis to deter mine soluble salt compositions. The analyses showed that the dominant particles on the canister surface were stainless steel particles, generated during manufacturing of the canister. Sparse environmentally - derived silicates and aluminosilicates were also present. Salt phases were sparse, and consisted of mostly of sulfates with rare nitrates and chlorides. On the FME covers, the dusts were mostly silicates/aluminosilicates; the soluble salts were consistent with those on the canister surface, and were dominantly sulfates. It should be noted that the FME covers were w ashed by rain prior to sampling, which had an unknown effect of the measured salt loads and compositions. Sulfate salts dominated the assemblages on the canister and FME surfaces, and in cluded Ca - SO4 , but also Na - SO4 , K - SO4 , and Na - Al - SO4 . It is likely that these salts were formed by particle - gas conversion reactions, either

  18. Analysis of X-band calibrated sea clutter and small boat reflectivity at medium-to-low grazing angles

    CSIR Research Space (South Africa)

    Herselman, PL

    2008-11-01

    Full Text Available of less than 2 seconds. 5. DETECTABILITY OF SMALL BOATS This section presents the detectability of small boats under- going diff erent manoeuvres using the ALQ detector as an example of the asymptotically optimal class of detectors. 5.1. Overview... in the datasets and subsequent analysis. Copyright © 2008 P. L. Herselman et al. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided...

  19. Feasibility study analysis for multi-function dual energy oven (case study: tapioca crackers small medium enterprise)

    Science.gov (United States)

    Soraya, N. W.; El Hadi, R. M.; Chumaidiyah, E.; Tripiawan, W.

    2017-12-01

    Conventional drying process is constrained by weather (cloudy / rainy), and requires wide drying area, and provides low-quality product. Multi-function dual energy oven is the appropriate technology to solve these problems. The oven uses solar thermal or gas heat for drying various type of products, including tapioca crackers. Investment analysis in technical, operational, and financial aspects show that the multi-function dual energy oven is feasible to be implemented for small medium enterprise (SME) processing tapioca crackers.

  20. Energetic and Exergetic Analysis of Low and Medium Temperature District Heating Network Integration

    DEFF Research Database (Denmark)

    Li, Hongwei; Svendsen, Svend

    In this paper, energetic and exergetic approaches were applied to an exemplary low temperature district heating (LTDH) network with supply/return water temperature at 55oC/25 oC. The small LTDH network is annexed to a large medium temperature district heating (MTDH) network. The LTDH network can ...... will reduce the amount of water supply from the MTDH network and improve the system energy conversion efficiency. Through the simulation, the system energetic and exergetic efficiencies based on the two network integration approaches were calculated and evaluated.......In this paper, energetic and exergetic approaches were applied to an exemplary low temperature district heating (LTDH) network with supply/return water temperature at 55oC/25 oC. The small LTDH network is annexed to a large medium temperature district heating (MTDH) network. The LTDH network can...

  1. A quantitative comparison analysis of diatoms in the lung tissues and the drowning medium as an indicator of drowning.

    Science.gov (United States)

    Zhao, Jian; Ma, Yanbin; Liu, Chao; Wen, Jinfeng; Hu, Sunlin; Shi, He; Zhu, Lingyun

    2016-08-01

    The presence of diatoms in the lung tissues, internal organs and bone marrow is considered as the supportive evidence in the diagnosis of death by drowning. Generally, the diatoms detected in the lung tissues are regarded as insignificant since these diatoms can be detected in the lung tissues of the postmortem immersion bodies. In this study, we analyzed the relationships between the numbers of the diatoms in the lung tissues and the drowning medium. We made a comparison analysis between the diatoms in the lung tissues and the drowning medium using the ratio of diatom numbers in both samples (L/D ratio), utilizing Microwave Digestion - Vacuum Filtration - Automated Scanning Electron Microscopy method. Our data indicate that the L/D ratios in victims of the drowning group were higher than the postmortem immersion group. A higher L/D ratio provides valuable information about the cause of death in drowning victims. Quantitative diatom analysis in the lung tissues, especially combined with the diatom analysis of the drowning medium, provides supportive evidence in determining if a body recovered in water was due to drowning or not. Copyright © 2016 Elsevier Ltd and Faculty of Forensic and Legal Medicine. All rights reserved.

  2. Analysis of Back-to-Back MMC for Medium Voltage Applications under Faulted Condition

    DEFF Research Database (Denmark)

    Bose, Anurag; Martins, Joäo Pedro Rodrigues; Chaudhary, Sanjay K.

    2017-01-01

    This paper analyzes a 10MW medium voltage Back-to-Back (BTB) Modular Multilevel Converter (MMC) without a DC-Link capacitor with halfbridge submodules. It focusses on the system behavior under single-line-to-ground (SLG) fault when there is no capacitor on the DC-Link.The fault current is computed...... to prevent DC overvoltages in the sub-modules during faults....

  3. Spent fuel management fee methodology and computer code user's manual

    International Nuclear Information System (INIS)

    Engel, R.L.; White, M.K.

    1982-01-01

    The methodology and computer model described here were developed to analyze the cash flows for the federal government taking title to and managing spent nuclear fuel. The methodology has been used by the US Department of Energy (DOE) to estimate the spent fuel disposal fee that will provide full cost recovery. Although the methodology was designed to analyze interim storage followed by spent fuel disposal, it could be used to calculate a fee for reprocessing spent fuel and disposing of the waste. The methodology consists of two phases. The first phase estimates government expenditures for spent fuel management. The second phase determines the fees that will result in revenues such that the government attains full cost recovery assuming various revenue collection philosophies. These two phases are discussed in detail in subsequent sections of this report. Each of the two phases constitute a computer module, called SPADE (SPent fuel Analysis and Disposal Economics) and FEAN (FEe ANalysis), respectively

  4. An analysis of long and medium-haul air passenger demand, volume 1

    Science.gov (United States)

    Eriksen, S. E.

    1978-01-01

    A basic model was developed which is a two equation pair econometric system in which air passenger demand and airline level-of-service are the endogenous variables. The model aims to identify the relationship between each of these two variables and its determining factors, and to identify the interaction of demand and level-of-service with each other. The selected variable for the measure of air passenger traffic activity in a given pair market is defined as the number of passengers in a given time that originate in one region and fly to the other region for purposes other than to make a connection to a third region. For medium and long haul markets, the model seems to perform better for larger markets. This is due to a specification problem regarding the route structure variable. In larger markets, a greater percentage of nonlocal passengers are accounted for by this variable. Comparing the estimated fare elasticities of long and medium haul markets, it appears that air transportation demand is more price elastic in longer haul markets. Long haul markets demand will saturate with a fewer number of departures than will demand in medium haul markets.

  5. Spent fuel transportation problems

    International Nuclear Information System (INIS)

    Kondrat'ev, A.N.; Kosarev, Yu.A.; Yulikov, E.A.

    1977-01-01

    In this paper, problems of transportation of nuclear spent fuel to reprocessing plants are discussed. The solutions proposed are directed toward the achievement of the transportation as economic and safe as possible. The increase of the nuclear power plants number in the USSR and the great distances between these plants and the reprocessing plants involve an intensification of the spent fuel transportation. Higher burnup and holdup time reduction cause the necessity of more bulky casks. In this connection, the economic problems become still more important. One of the ways of the problem solution is the development of rational and cheap cask designs. Also, the enforcement in the world of the environmental and personnel health protection requires to increase the transportation reliability and safety. The paper summarizes safe transportation rules with clarifying the following questions: the increase of the transport unit quantity of the spent fuel; rational shipment organization that minimizes vehicle turnover cycle duration; development of the reliable calculation methods to determine strength, thermal conditions and nuclear safety of transport packaging as applied to the vehicles of high capacity; maximum unification of vehicles, calculation methods and documents; and cask testing on models and in pilot scale on specific test rigs to assure that they meet the international safe fuel shipment rules. Besides, some considerations on the choice and use of structural materials for casks are given, and problems of manufacturing such casks from uranium and lead are considered, as well as problems of the development of fireproof shells, control instrumentation, vehicles decontamination, etc. All the problems are considered from the point of view of normal and accidental shipment conditions. Conclusions are presented [ru

  6. Miscibility and oxidation rate of the simulated metallic spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    You, K. S.; Joo, J. S.; Shin, Y. J.; Oh, S. C. [KAERI, Taejon (Korea, Republic of)

    1999-10-01

    The simulated metallic spent fuel was fabricated by using Uranium, Neodymium and Palladium in order to study the miscibility of Neodymium and Palladium with Uranium. For analysis of long-term safty on the metallized spent fuel, the simulated metallic spent fuel was oxidized under pure oxygen environment at 183{approx}250 deg C. From the results, the oxidation rate correlation and activation energy were obtained.

  7. Spent fuel reprocessing method

    International Nuclear Information System (INIS)

    Shoji, Hirokazu; Mizuguchi, Koji; Kobayashi, Tsuguyuki.

    1996-01-01

    Spent oxide fuels containing oxides of uranium and transuranium elements are dismantled and sheared, then oxide fuels are reduced into metals of uranium and transuranium elements in a molten salt with or without mechanical removal of coatings. The reduced metals of uranium and transuranium elements and the molten salts are subjected to phase separation. From the metals of uranium and transuranium elements subjected to phase separation, uranium is separated to a solid cathode and transuranium elements are separated to a cadmium cathode by an electrolytic method. Molten salts deposited together with uranium to the solid cathode, and uranium and transuranium elements deposited to the cadmium cathode are distilled to remove deposited molten salts and cadmium. As a result, TRU oxides (solid) such as UO 2 , Pu 2 in spent fuels can be reduced to U and TRU by a high temperature metallurgical method not using an aqueous solution to separate them in the form of metal from other ingredients, and further, metal fuels can be obtained through an injection molding step depending on the purpose. (N.H.)

  8. Spent fuel interim storage

    International Nuclear Information System (INIS)

    Bilegan, Iosif C.

    2003-01-01

    The official inauguration of the spent fuel interim storage took place on Monday July 28, 2003 at Cernavoda NNP. The inaugural event was attended by local and central public authority representatives, a Canadian Government delegation as well as newsmen from local and central mass media and numerous specialists from Cernavoda NPP compound. Mr Andrei Grigorescu, State Secretary with the Economy and Commerce Ministry, underlined in his talk the importance of this objective for the continuous development of nuclear power in Romania as well as for Romania's complying with the EU practice in this field. Also the excellent collaboration between the Canadian contractor AECL and the Romanian partners Nuclear Montaj, CITON, UTI, General Concret in the accomplishment of this unit at the planned terms and costs. On behalf of Canadian delegation, spoke Minister Don Boudria. He underlined the importance which the Canadian Government affords to the cooperation with Romania aiming at specific objectives in the field of nuclear power such as the Cernavoda NPP Unit 2 and spent fuel interim storage. After traditional cutting of the inaugural ribbon by the two Ministers the festivities continued on the Cernavoda NPP Compound with undersigning the documents regarding the project completion and a press conference

  9. Exergy-topological analysis and optimization of a binary power plant utilizing medium-grade geothermal energy

    International Nuclear Information System (INIS)

    Makhanlall, Deodat; Zhang, Fuzhen; Xu, Ruina; Jiang, Peixue

    2015-01-01

    While it is generally accepted that high efficiency conversion of low- and medium-grade heat to electrical power strongly relies on a suitable combination of thermodynamic cycle and working fluid, the selection of suitable operating parameters may be of even more importance. This study shows that medium-grade geothermal heat-to-power conversion in a well-designed secondary regenerative Rankine cycle can achieve a high degree of thermodynamic perfection and exergy efficiency, without the use of high-cost advance powerfluids. The study is carried out by combining a recently developed exergy-topological analysis scheme with CyclePad ® , an open-source cognitive thermodynamic tool. The powerful sensitivity analysis capability of CyclePad ® is used to determine optimal operating conditions. In addition, application of an extensive exergy-flow diagram, which includes flows in the geothermal production well and cooling cycle, is discussed. A key strategy for improving the thermodynamic efficiency of medium-grade geothermal power plants is also presented

  10. CONSOLIDATED BUDGET REVENUES EVOLUTION ANALYSIS DURING 2012-2014 AND MEDIUM-TERM OUTLOOK

    Directory of Open Access Journals (Sweden)

    CHIRCULESCU MARIA FELICIA

    2015-04-01

    Full Text Available Economic activity now when the need for greater financial resources than existing funds, involves a major responsibility in preparation of budgets at all levels, and hence the structure of each of the general government budget, so that the principles existing in the specialized literature budget, in particular the principle of budgetary balance. Thus this paper wants to identify the structure of consolidated budget revenues and their evolution over the period 2012-2014, as well as highlighting the main medium-term budgetary challenges so as to maintain a concern for reducing the budget deficit.

  11. An Analysis on Development Predicaments of Privately-owned Small & Medium Enterprises and Countermeasures

    Directory of Open Access Journals (Sweden)

    Tan Hanbing

    2015-01-01

    Full Text Available Through field survey on 100 privately-owned small and medium enterprises in Shijiazhuang of He bei province and relevant data collection, the author finds that their development has been severely impeded by such phenomena as financing dilemma, retention difficulty and chaotic management. On the basis of theories of management and social development, reasons for development predicaments are systematically analyzed by means of combining qualitative and quantitative studies. In the meantime, based on relevant economic theories and reference to advanced experience and theories at home and abroad, the author proposes countermeasures and suggestions for development predicament of SMEs, which are relatively suitable for Shijiazhuang City.

  12. Numerical analysis of three-dimensional MHD shock interactions in an inhomogeneous medium

    International Nuclear Information System (INIS)

    Prndergast, M.; Wu, S.T.

    1987-01-01

    Study of the formation and propagation of solar-originated shock waves in heliospheric space has attracted significant attention in the past decade. This attention is important because the propagation of shocks in heliospheric space has been thought of as one of the major physical processes for solar wind and cosmic ray modulations and their subsequent influence on the earth's environment. A version of the two step Lax-Wendroff difference method is used to seek solutions of the unsteady magnetohydrodynamic (MHD) equations for the study of a solar flare generated shock wave propagating through an inhomogeneous medium. 8 references

  13. Modeling and analysis of unsteady axisymmetric squeezing fluid flow through porous medium channel with slip boundary.

    Science.gov (United States)

    Qayyum, Mubashir; Khan, Hamid; Rahim, M Tariq; Ullah, Inayat

    2015-01-01

    The aim of this article is to model and analyze an unsteady axisymmetric flow of non-conducting, Newtonian fluid squeezed between two circular plates passing through porous medium channel with slip boundary condition. A single fourth order nonlinear ordinary differential equation is obtained using similarity transformation. The resulting boundary value problem is solved using Homotopy Perturbation Method (HPM) and fourth order Explicit Runge Kutta Method (RK4). Convergence of HPM solution is verified by obtaining various order approximate solutions along with absolute residuals. Validity of HPM solution is confirmed by comparing analytical and numerical solutions. Furthermore, the effects of various dimensionless parameters on the longitudinal and normal velocity profiles are studied graphically.

  14. ANALYSIS OF FREE TORSIONAL VIBRATION IN CARBON NANOTUBES EMBEDDED IN A VISCOELASTIC MEDIUM

    Directory of Open Access Journals (Sweden)

    Mustafa Arda

    2015-05-01

    Full Text Available Carbon Nanotubes (CNTs have a great potential in many areas like electromechanical systems, medical application, pharmaceutical industry etc. The surrounding physical environment of CNT is very important on torsional vibration behavior of CNT. Damping and elastic effect of medium to the torsional vibration of CNTs are investigated in the present study. Governing equation of motion of nanotube is obtained using Eringen’s Nonlocal Elasticty Theory. The effects of some parameters like nonlocal parameter, stiffness parameter and nanotube length are studied in detail.

  15. Analysis of Medium-Scale Solar Thermal Systems and Their Potential in Lithuania

    OpenAIRE

    Valančius, Rokas; Jurelionis, Andrius; Jonynas, Rolandas; Katinas, Vladislovas; Perednis, Eugenijus

    2015-01-01

    Medium-scale solar hot water systems with a total solar panel area varying from 60 to 166 m 2 have been installed in Lithuania since 2002. However, the performance of these systems varies depending on the type of energy users, equipment and design of the systems, as well as their maintenance. The aim of this paper was to analyse operational SHW systems from the perspective of energy production and economic benefit as well as to outline the differences of their actual performance compared to t...

  16. Metabolic network analysis of Bacillus clausii on minimal and semirich medium using C-13-Labeled glucose

    DEFF Research Database (Denmark)

    Christiansen, Torben; Christensen, Bjarke; Nielsen, Jens

    2002-01-01

    to increase with increasing specific growth rate but at a much lower level than previously reported for Bacillus subtilis. Two futile cycles in the pyruvate metabolism were included in the metabolic network. A substantial flux in the futile cycle involving malic enzyme was estimated, whereas only a very small...... or zero flux through PEP carboxykinase was estimated, indicating that the latter enzyme was not active during growth on glucose. The uptake of the amino acids in a semirich medium containing 15 of the 20 amino acids normally present in proteins was estimated using fully labeled glucose in batch...

  17. Metabolome analysis of Saccharomyces cerevisiae and optimization of culture medium for S-adenosyl-L-methionine production.

    Science.gov (United States)

    Hayakawa, Kenshi; Matsuda, Fumio; Shimizu, Hiroshi

    2016-12-01

    S-Adenosyl-L-methionine (SAM) is a fine chemical used as a nutritional supplement and a prescription drug. It is industrially produced using Saccharomyces cerevisiae owing to its high SAM content. To investigate the optimization of culture medium components for higher SAM production, metabolome analysis was conducted to compare the intracellular metabolite concentrations between Kyokai no. 6 (high SAM-producing) and laboratory yeast S288C (control) under different SAM production conditions. Metabolome analysis and the result of principal component analysis showed that the rate-limiting step for SAM production was ATP supply and the levels of degradation products of adenosine nucleotides were higher in Kyokai 6 strain than in the S288C strain under the L-methionine supplemented condition. Analysis of ATP accumulation showed that the levels of intracellular ATP in the Kyokai 6 strain were also higher compared to those in the S288C strain. Furthermore, as expected from metabolome analysis, the SAM content of Kyokai 6 strain cultivated in the medium without yeast extract increased by 2.5-fold compared to that in the additional condition, by increasing intracellular ATP level with inhibited cell growth. These results suggest that high SAM production is attributed to the enhanced ATP supply with L-methionine condition and high efficiency of intracellular ATP consumption.

  18. ANALYSIS OF LEADERSHIP STYLES IN ŠIBENIK-KNIN COUNTY BASED ON MEDIUM AND LARGE ENTERPRISES

    Directory of Open Access Journals (Sweden)

    Željko Požega

    2012-12-01

    Full Text Available During the last two decades leadership has become the main focus of intense interest and research by scientists and theorists. A large number of leadership models has been developed in order to define and enable the certain level of business flexibility that is crucial to survive in a new business environment which is characterized by frequent market change, growing global competition and technology development together with demographic changes of employees. The purpose of the paper is to determine the dominant leadership style as well as the advantages and disadvantages in medium and large organizations in Šibenik-Knin County. For the purpose of this paper, empirical research was conducted with the goal of defining the leadership styles of top and middle management based on medium and large organizations. The research was based on the leadership model developed and established by Likert Renis which is the most used and recognized model for diagnostic determination of the dominant leadership style. The Likert model is based on six components: leadership, motivation, communication, decision making, goals and control which determine the four leadership styles: extreme authoritative, paternalistic, consultative and participative leadership style.

  19. Classification of spent nuclear fuel (SNF)

    International Nuclear Information System (INIS)

    1990-03-01

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. This document discusses the classification of spent nuclear fuels

  20. Compacting spent fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    A method and apparatus for compacting spent fuel rods comprises transferring the rods from a nuclear fuel rod assembly into a different nuclear fuel rod container having a smaller cross section than the assembly. The individual rods are moved from a fuel assembly and through a transition funnel by movable grippers at opposite ends of the funnel. One movable gripper reciprocates between gripping and release positions in a gap between the fuel assembly and the transition funnel. All of the fuel rods are withdrawn concurrently and are merged towards one another into a tighter array within the transition funnel and emerge as a bundle. A movable and a stationary bundle gripper are provided between the funnel and the storage container to advance the bundle of fuel rods into the container. (author)

  1. Spent fuel transporting vessel

    International Nuclear Information System (INIS)

    Kumagaya, Takeshi.

    1995-01-01

    A large number of annular cooling fins are disposed each at an equal distance on the outer circumferential surface of a vessel main body. An electric power generation module is disposed on the surface of the cooling fins. The electric power generation module comprises a plurality of thermoelectric power generation elements. In each of the thermoelectric generation elements, the inner side thereof in contact with the surface of the cooling fin is at a high temperature while the outer side thereof is at a low temperature nearly equal with an atmospheric temperature. A predetermined amount of electric power is generated by seebeck effect due to the temperature difference. The electric power is always stored in a battery. Accordingly, even if a power generator of a ship should fail and power supply is stopped during transportation of the vessels for spent nuclear fuels, an appropriate amount of electric power can be supplied to a cooling device of the ship. (I.N.)

  2. Spent fuel storage at KURRI

    International Nuclear Information System (INIS)

    Nakagome, Y.; Fujita, Y.; Kanda, K.

    2004-01-01

    The Research Reactor Institute, Kyoto University (KURRI) has more than 200 MTR-type spent fuel elements stored in water pools. The longest pool residence time is 21 years at present. The integrity of spent fuel elements have been confirmed by a visual inspection and a sipping test. The spent fuel elements should be reprocessed in accordance with KURRI's policy. KURRI is now negotiating with a reprocessing plant to make a contract, as considering the consequences in U.S. (author)

  3. Production of single cell proterin from brewery spent grains ...

    African Journals Online (AJOL)

    The production of single cell protein (SCP) by the propagation of the yeast, Saccharomyces cerevisae obtained from the Federal Institute of Industrial Research Oshtxli was studied by using the extract of spent grains obtained from the International Beer and Beverage Industries, Kacltuia, Nigeria as a substrate in a medium ...

  4. Evaluation on the effects of P. ostreatus spent mushroom compost ...

    African Journals Online (AJOL)

    Realization in minimizing production cost for in vitro culture had brought to a study on application of P. ostreatus spent mushroom compost (SMC). Sterile nodal explants was inoculated on different treatments with 15 replicates each. Treatments were MS medium supplemented with different concentrations of SMC (1 and 2 ...

  5. Enablers of the successful implementation of CIM in small and medium enterprises: an empirical analysis

    Science.gov (United States)

    Marri, H. B.; Gunasekaran, Angappa

    2001-10-01

    Small and Medium Enterprises (SMEs) play an important role in the national economy. Since manufacturing has become global, the competition is high among manufacturing/service industries to provide quality goods and services at competitive prices. Computer Integrated Manufacturing (CIM) is a generic term for a group of manufacturing technologies that combine both scope and scale capabilities that are making fundamental changes to manufacturing industry. The enablers of CIM play an important role in the implementation of CIM by SMEs to compete in the global market. Considering the importance of enablers of the successful implementation of CIM, the critical success factors for the implementation of CIM in SMEs have been identified with the help of an empirical study conducted with British SMEs.

  6. An approach to the investment analysis of small and medium hydro-power plants

    International Nuclear Information System (INIS)

    Forouzbakhsh, F.; Hosseini, S.M.H.; Vakilian, M.

    2007-01-01

    Hydro-power plants, as a part of infrastructure projects, play an important role in the economic-social development of countries. Since a large amount of investment is needed for construction of these power plants, which appeared to be an obstacle in these developments, however it is possible to finance these infrastructure plants by assigning these affairs to private sectors by using build operate transfer (BOT) method, which is quite well-known all around the world. This paper reviews the structure of BOT contracts and through an economic evaluation based on different percentage of investments of private sector in providing the expenses of small and medium hydro-power plants (S and M-HPP) (e.g. MHPP in 'Bookan, Iran' and SHPP in 'Nari, Iran'), demonstrates that by increasing the percentage the share of the private sector in the investment, the economic indices B/C and NPV improve substantially

  7. Modeling flow in fractured medium. Uncertainty analysis with stochastic continuum approach

    International Nuclear Information System (INIS)

    Niemi, A.

    1994-01-01

    For modeling groundwater flow in formation-scale fractured media, no general method exists for scaling the highly heterogeneous hydraulic conductivity data to model parameters. The deterministic approach is limited in representing the heterogeneity of a medium and the application of fracture network models has both conceptual and practical limitations as far as site-scale studies are concerned. The study investigates the applicability of stochastic continuum modeling at the scale of data support. No scaling of the field data is involved, and the original variability is preserved throughout the modeling. Contributions of various aspects to the total uncertainty in the modeling prediction can also be determined with this approach. Data from five crystalline rock sites in Finland are analyzed. (107 refs., 63 figs., 7 tabs.)

  8. Analysis of Humid Air Turbine Cycle with Low- or Medium-Temperature Solar Energy

    International Nuclear Information System (INIS)

    Hongbin Zhao, H.; Yue, P.; Cao, L.

    2009-01-01

    A new humid air turbine cycle that uses low- or medium-temperature solar energy as assistant heat source was proposed for increasing the mass flow rate of humid air. Based on the combination of the first and second laws of thermodynamics, this paper described and compared the performances of the conventional and the solar HAT cycles. The effects of some parameters such as pressure ratio, turbine inlet temperature (TIT), and solar collector efficiency on humidity, specific work, cycle's exergy efficiency, and solar energy to electricity efficiency were discussed in detail. Compared with the conventional HAT cycle, because of the increased humid air mass flow rate in the new system, the humidity and the specific work of the new system were increased. Meanwhile, the solar energy to electricity efficiency was greatly improved. Additionally, the exergy losses of components in the system under the given conditions were also studied and analyzed.

  9. Low-frequency asymptotic analysis of seismic reflection from afluid-saturated medium

    Energy Technology Data Exchange (ETDEWEB)

    Silin, D.B.; Korneev, V.A.; Goloshubin, G.M.; Patzek, T.W.

    2004-04-14

    Reflection of a seismic wave from a plane interface betweentwo elastic media does not depend on the frequency. If one of the mediais poroelastic and fluid-saturated, then the reflection becomesfrequency-dependent. This paper presents a low-frequency asymptoticformula for the reflection of seismic plane p-wave from a fluid-saturatedporous medium. The obtained asymptotic scaling of the frequency-dependentcomponent of the reflection coefficient shows that it is asymptoticallyproportional to the square root of the product of the reservoir fluidmobility and the frequency of the signal. The dependence of this scalingon the dynamic Darcy's law relaxation time is investigated as well.Derivation of the main equations of the theory of poroelasticity from thedynamic filtration theory reveals that this relaxation time isproportional to Biot's tortuosity parameter.

  10. Steady flow and heat transfer analysis of third grade fluid with porous medium and heat generation

    Directory of Open Access Journals (Sweden)

    Akinbowale T. Akinshilo

    2017-12-01

    Full Text Available In this study, flow and heat transfer of a non Newtonian third grade fluid with porous medium and internal heat source conveyed through parallel plates held horizontally against each other are investigated. The nonlinear ordinary equations arising due to visco-elastic effects from the mechanics of the fluid are analysed using the adomian decomposition method (ADM adopting Vogel’s temperature dependent model based viscosity. Thermal fluidic parameters effects such as pressure gradient, heat generation parameter and porosity term are examined on the flow and heat transfer. Increasing porosity term shows slight decreasing effect on velocity distribution, as increasing heat generation term demonstrates significant increase on temperature distribution towards the upper plate. Obtained solutions in this paper may be used to advance studies in thin film flow, energy conservation, coal-water mixture, polymer solution and oil recovery application. Also Results from analyses compared against the fourth order Runge kutta numerical solution proves to be in satisfactory agreement.

  11. A Between- and Within-Person Analysis of Parenting And Time Spent in Criminogenic Settings during Adolescence: The Role of Self-Control And Delinquent Attitudes

    Science.gov (United States)

    Janssen, Heleen J.; Bruinsma, Gerben J. N.; Dekovic, Maja; Eichelsheim, Veroni I.

    2018-01-01

    Although spending time in criminogenic settings is increasingly recognized as an explanation for adolescent delinquency, little is known about its determinants. The current study aims to examine the extent to which (change in) self-control and (change in) delinquent attitudes relate to (change in) time spent in criminogenic settings, and the…

  12. Overview of the United States spent nuclear fuel program

    International Nuclear Information System (INIS)

    Hurt, W.L.

    1997-12-01

    As a result of the end of the Cold War, the mission of the US Department of Energy (DOE) has shifted from an emphasis on nuclear weapons development and production to an emphasis on the safe management and disposal of excess nuclear materials including spent nuclear fuel from both production and research reactors. Within the US, there are two groups managing spent nuclear fuel. Commercial nuclear power plants are managing their spent nuclear fuel at the individual reactor sites until the planned repository is opened. All other spent nuclear fuel, including research reactors, university reactors, naval reactors, and legacy material from the Cold War is managed by DOE. DOE's mission is to safely and efficiently manage its spent nuclear fuel and prepare it for disposal. This mission involves correcting existing vulnerabilities in spent fuel storage; moving spent fuel from wet basins to dry storage; processing at-risk spent fuel; and preparing spent fuel in road-ready condition for repository disposal. Most of DOE's spent nuclear fuel is stored in underwater basins (wet storage). Many of these basins are outdated, and spent fuel is to be removed and transferred to more modern basins or to new dry storage facilities. In 1995, DOE completed a complex-wide environmental impact analysis that resulted in spent fuel being sent to one of three principal DOE sites for interim storage (up to 40 years) prior to shipment to a repository. This regionalization by fuel type will allow for economies of scale yet minimize unnecessary transportation. This paper discusses the national SNF program, ultimate disposition of SNF, and the technical challenges that have yet to be resolved, namely, release rate testing, non-destructive assay, alternative treatments, drying, and chemical reactivity

  13. Analysis of the growth of strike-slip faults using effective medium theory

    Energy Technology Data Exchange (ETDEWEB)

    Aydin, A.; Berryman, J.G.

    2009-10-15

    Increases in the dimensions of strike-slip faults including fault length, thickness of fault rock and the surrounding damage zone collectively provide quantitative definition of fault growth and are commonly measured in terms of the maximum fault slip. The field observations indicate that a common mechanism for fault growth in the brittle upper crust is fault lengthening by linkage and coalescence of neighboring fault segments or strands, and fault rock-zone widening into highly fractured inner damage zone via cataclastic deformation. The most important underlying mechanical reason in both cases is prior weakening of the rocks surrounding a fault's core and between neighboring fault segments by faulting-related fractures. In this paper, using field observations together with effective medium models, we analyze the reduction in the effective elastic properties of rock in terms of density of the fault-related brittle fractures and fracture intersection angles controlled primarily by the splay angles. Fracture densities or equivalent fracture spacing values corresponding to the vanishing Young's, shear, and quasi-pure shear moduli were obtained by extrapolation from the calculated range of these parameters. The fracture densities or the equivalent spacing values obtained using this method compare well with the field data measured along scan lines across the faults in the study area. These findings should be helpful for a better understanding of the fracture density/spacing distribution around faults and the transition from discrete fracturing to cataclastic deformation associated with fault growth and the related instabilities.

  14. An approach to the investment analysis of small and medium hydro-power plants

    Energy Technology Data Exchange (ETDEWEB)

    Forouzbakhsh, F. [University of Tehran (Iran). Faculty of Engineering; Hosseini, S.M.H. [Islamic Azad University, Tehran (Iran). Faculty of Engineering; Vakilian, M. [Sharif Institute of Technology, Tehran (Iran). Faculty of Electrical Engineering

    2007-02-15

    Hydro-power plants, as a part of infrastructure projects, play an important role in the economic-social development of countries. Since a large amount of investment is needed for construction of these power plants, which appeared to be an obstacle in these developments, however it is possible to finance these infrastructure plants by assigning these affairs to private sectors by using build operate transfer (BOT) method, which is quite well-known all around the world. This paper reviews the structure of BOT contracts and through an economic evaluation based on different percentage of investments of private sector in providing the expenses of small and medium hydro-power plants (S and M-HPP) (e.g. MHPP in ''Bookan, Iran'' and SHPP in ''Nari, Iran''), demonstrates that by increasing the percentage the share of the private sector in the investment, the economic indices B/C and NPV improve substantially. (author)

  15. Cross correlation analysis of medium energy gamma rays for the northern hemisphere

    International Nuclear Information System (INIS)

    Long, J.; Zanrosso, E.; Zych, A.D.; White, R.S.

    1982-01-01

    Data obtained with the UCR gamma telescope have been analyzed using the cross-correlation method. The observations extended over 37.5 hr from 0930 UT, 30 Sept. to 2300 UT, 1 oct. 1978 at 32deg N. Lat. (Palestine, Texas). The Crab Nebula- Anticenter region was observed on consecutive days. The telescope's wide field-of-view permitted the search for a number of other medium energy (1-30 MeV) source candidates. As the telescope swept the sky, the count rates for fixed celestial directions were correlated with the expected response as a function of time and telescope geometry. Similar correlations were carried out for sources measured in the laboratory and computer-simulated sources. In the correlation method the time independence and azimuthal symmetry of the atmospheric and cosmic diffuse backgrounds provide zero correlation. In contrast, a celestial source produces an asymmetric response with respect to the azimuthal direction which varies predictably in time to give a positive correlation. Preliminary correlation skymaps of the Anticenter region are presented and their statistical significance discussed. An energy spectrum obtained from the ''correlated counts'' is compared with measurements by other methods

  16. Analysis of the Strength on the Rotor Punching Sheet of Nuclear Reactor Cooling Medium Driving Motor

    Directory of Open Access Journals (Sweden)

    GE Bao-jun

    2017-02-01

    Full Text Available A strong stress is withstood by the rotor punching sheet during the running of nuclear reactor cooling medium driving motor. In order to study the strength on the rotor punching sheet and the influential factor of its stress,the rotor of driving motor was the research object, the three-dimensional rotor model of driving motor is established by the finite element method to obtain the Mires equivalent stress nephogram and check the rotor’s strength with setting parameters and constraints. According to different rotor speeds,the different average temperatures of rotor punching sheet and shaft and the different static magnitude of interference between rotor punching sheet and shaft,the research about how the contact pressure of matching surface between rotor punching sheet and shaft and the Mires equivalent stress are impacted is carried on. The results show that the maximum Miser equivalent stress value of rotor punching sheet emerges on the axial vents,the stress value is beyond the tensile limit of the materialand. The greater the static magnitude of interference and the smaller temperature difference of rotor punching sheet and shaft lead to the greater interface compressive stress of rotor punching sheet and shaft and the greater maximum Mires equivalent stress value of rotor punching sheet. The higher the rotor speed lead to the smaller interface compressive stress of rotor punching sheet and shaft and the greater equivalent stress value of rotor punching sheet.

  17. Spent fuel management in Japan

    International Nuclear Information System (INIS)

    Mineo, H.; Nomura, Y.; Sakamoto, K.

    1998-01-01

    In Japan 52 commercial nuclear power units are now operated, and the total power generation capacity is about 45 GWe. The cumulative amount of spent fuel arising is about 13,500 tU as of March 1997. Spent fuel is reprocessed, and recovered nuclear materials are to be recycled in LWRs and FBRs. In February 1997 short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, backend measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away from reactor sites, considering the increasing amount of spent fuel arising. Research and development on spent fuel storage has been carried out, particularly on dry storage technology. Fundamental studies are also conducted to implement the burnup credit into the criticality safety design of storage and transportation casks. Rokkasho reprocessing plant is being constructed towards its commencement in 2003, and Pu utilization in LWRs will be started in 1999. Research and development of future recycling technology are also continued for the establishment of nuclear fuel cycle based on FBRs and LWRs. (author)

  18. Total quality in spent fuel pool reracking

    International Nuclear Information System (INIS)

    Cranston, J.S.; Bradbury, R.B.; Cacciapouti, R.J.

    1993-01-01

    The nuclear utility environment is one of strict cost control under prescriptive regulations and increasing public scrutiny. This paper presents the results of A Total Quality approach, by a dedicated team, that addresses the need for increased on-site spent fuel storage in this environment. Innovations to spent fuel pool reracking, driven by utilities' specific technical needs and shrinking budgets, have resulted in both product improvements and lower prices. A Total Quality approach to the entire turnkey project is taken, thereby creating synergism and process efficiency in each of the major phases of the project: design and analysis, licensing, fabrication, installation and disposal. Specific technical advances and the proven quality of the team members minimizes risk to the utility and its shareholders and provides a complete, cost effective service. Proper evaluation of spent fuel storage methods and vendors requires a full understanding of currently available customer driven initiatives that reduce cost while improving quality. In all phases of a spent fuel reracking project, from new rack design and analysis through old rack disposal, the integration of diverse experts, at all levels and throughout all phases of a reracking project, better serves utility needs. This Total Quality environment in conjunction with many technical improvements results in a higher quality product at a lower cost

  19. Milestones for Selection, Characterization, and Analysis of the Performance of a Repository for Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain.

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, Robert P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-02-01

    This report presents a concise history in tabular form of events leading up to site identification in 1978, site selection in 1987, subsequent characterization, and ongoing analysis through 2009 of the performance of a repository for spent nuclear fuel and high - level radioactive waste at Yucca Mountain in southern Nevada. The tabulated events generally occurred in five periods: (1) commitment to mined geologic disposal and identification of sites; (2) site selection and analysis, based on regional geologic characterization through literature and analogous data; (3) feasibility analysis demonstrating calculation procedures and importance of system components, based on rough measures of performance using surface exploration, waste process knowledge, and general laboratory experiments; (4) suitability analysis demonstrating viability of disposal system, based on environment - specific laboratory experiments, in - situ experiments, and underground disposal system characterization; and (5) compliance analysis, based on completed site - specific characterization . The current sixth period beyond 2010 represents a new effort to set waste management policy in the United States. Because the relationship is important to understanding the evolution of the Yucca Mountain Project , the tabulation also shows the interaction between the policy realm and technical realm using four broad categories of events : (a) Regulatory requirements and related federal policy in laws and court decisions, (c) Presidential and agency directives, (c) technical milestones of implementing institutions, and (d) critiques of the Yucca Mountain Project and pertinent national and world events related to nuclear energy and radioactive waste. Preface The historical progression of technical milestones for the Yucca Mountain Project was originally developed for 10 journal articles in a special issue of Reliability Engineering System Safety on the performance assessment for the Yucca Mountain license

  20. WWER spent fuel storage

    International Nuclear Information System (INIS)

    Bower, C.C.; Lettington, C.

    1994-01-01

    Selection criteria for PAKS NPP dry storage system are outlined. They include the following: fuel temperature in storage; sub-criticality assurance (avoidance of criticality for fuel in the unirradiated condition without having to take credit for burn-up); assurance of decay heat removal; dose uptake to the operators and public; protection of environment; volume of waste produced during operation and decommissioning; physical protection of stored irradiated fuel assemblies; IAEA safeguards assurance; storage system versus final disposal route; cost of construction and extent of technology transfer to Hungarian industry. Several available systems are evaluated against these criteria, and as a result the GEC ALSTHOM Modular Vault Dry Store (MVDS) system has been selected. The MVDS is a passively cooled dry storage facility. Its most important technical, safety, licensing and technology transfer characteristics are outlined. On the basis of the experience gained some key questions and considerations related to the East European perspective in the field of spent fuel storage are discussed. 8 figs

  1. Containing method for spent fuel and spent fuel containing vessel

    International Nuclear Information System (INIS)

    Maekawa, Hiromichi; Hanada, Yoshine.

    1996-01-01

    Upon containing spent fuels, a metal vessel main body and a support spacer having fuel containing holes are provided. The support spacer is disposed in the inside of the metal vessel main body, and spent fuel assemblies are loaded in the fuel containing holes. Then, a lid is welded at the opening of the metal vessel main body to provide a sealing state. In this state, heat released from the spent fuel assemblies is transferred to the wall of the metal vessel main body via the support spacer. Since the support spacer has a greater heat conductivity than gases, heat of the spent fuel assemblies tends to be released to the outside, thereby capable of removing heat of the spent fuel assemblies effectively. In addition, since the surfaces of the spent fuel assemblies are in contact with the inner surface of the fuel containing holes of the support spacer, impact-resistance and earthquake-resistance are ensured, and radiation from the spent fuel assemblies is decayed by passing through the layer of the support spacer. (T.M.)

  2. Quantitative considerations in medium energy ion scattering depth profiling analysis of nanolayers

    Energy Technology Data Exchange (ETDEWEB)

    Zalm, P.C.; Bailey, P. [International Institute for Accelerator Applications, University of Huddersfield, Queensgate, Huddersfield HD1 3DH (United Kingdom); Reading, M.A. [Physics and Materials Research Centre, University of Salford, Salford M5 4WT (United Kingdom); Rossall, A.K. [International Institute for Accelerator Applications, University of Huddersfield, Queensgate, Huddersfield HD1 3DH (United Kingdom); Berg, J.A. van den, E-mail: j.vandenberg@hud.ac.uk [International Institute for Accelerator Applications, University of Huddersfield, Queensgate, Huddersfield HD1 3DH (United Kingdom)

    2016-11-15

    The high depth resolution capability of medium energy ion scattering (MEIS) is becoming increasingly relevant to the characterisation of nanolayers in e.g. microelectronics. In this paper we examine the attainable quantitative accuracy of MEIS depth profiling. Transparent but reliable analytical calculations are used to illustrate what can ultimately be achieved for dilute impurities in a silicon matrix and the significant element-dependence of the depth scale, for instance, is illustrated this way. Furthermore, the signal intensity-to-concentration conversion and its dependence on the depth of scattering is addressed. Notably, deviations from the Rutherford scattering cross section due to screening effects resulting in a non-coulombic interaction potential and the reduction of the yield owing to neutralization of the exiting, backscattered H{sup +} and He{sup +} projectiles are evaluated. The former mainly affects the scattering off heavy target atoms while the latter is most severe for scattering off light target atoms and can be less accurately predicted. However, a pragmatic approach employing an extensive data set of measured ion fractions for both H{sup +} and He{sup +} ions scattered off a range of surfaces, allows its parameterization. This has enabled the combination of both effects, which provides essential information regarding the yield dependence both on the projectile energy and the mass of the scattering atom. Although, absolute quantification, especially when using He{sup +}, may not always be achievable, relative quantification in which the sum of all species in a layer adds up to 100%, is generally possible. This conclusion is supported by the provision of some examples of MEIS derived depth profiles of nanolayers. Finally, the relative benefits of either using H{sup +} or He{sup +} ions are briefly considered.

  3. Analysis of industrial markets for low and medium Btu coal gasification. [Forecasting

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-30

    Low- and medium-Btu gases (LBG and MBG) can be produced from coal with a variety of 13 existing and 25 emerging processes. Historical experience and previous studies indicate a large potential market for LBG and MBG coal gasification in the manufacturing industries for fuel and feedstocks. However, present use in the US is limited, and industry has not been making substantial moves to invest in the technology. Near-term (1979-1985) market activity for LBG and MBG is highly uncertain and is complicated by a myriad of pressures on industry for energy-related investments. To assist in planning its program to accelerate the commercialization of LBG and MBG, the Department of Energy (DOE) contracted with Booz, Allen and Hamilton to characterize and forecast the 1985 industrial market for LBG and MBG coal gasification. The study draws five major conclusions: (1) There is a large technically feasible market potential in industry for commercially available equipment - exceeding 3 quadrillion Btu per year. (2) Early adopters will be principally steel, chemical, and brick companies in described areas. (3) With no additional Federal initiatives, industry commitments to LBG and MBG will increase only moderately. (4) The major barriers to further market penetration are lack of economic advantage, absence of significant operating experience in the US, uncertainty on government environmental policy, and limited credible engineering data for retrofitting industrial plants. (5) Within the context of generally accepted energy supply and price forecasts, selected government action can be a principal factor in accelerating market penetration. Each major conclusion is discussed briefly and key implications for DOE planning are identified.

  4. Disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    1979-12-01

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste

  5. Disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste.

  6. Performance analysis of electricity generation by the medium temperature geothermal resources: Velika Ciglena case study

    International Nuclear Information System (INIS)

    Rašković, Predrag; Guzović, Zvonimir; Cvetković, Svetislav

    2013-01-01

    During the last decade, a design of an energy efficient and cost effective geothermal plant represents a significant and on-going technical challenge in all the Western Balkan countries. In the Republic of Croatia, the geothermal field Velika Ciglena is identified as one of the most valuable geothermal heat sources and probably the location where the first geothermal plant in the Western Balkan area will be built. The purpose of this work is the conceptual design and performance analysis of the binary plants–the one which operates under the Organic Rankine Cycle (ORC) and the other under Kalina (KLN) cycle–which can be used for geothermal energy utilization in Velika Ciglena. A conceptual plant design is performed by the equation-oriented modelling approach and supported by the two steady-state spreadsheet simulators. The performance analysis of all design solutions is conducted through energy and exergy analysis, and by the estimated total cost of operating units in the plant. The results of the analysis indicate that the plant design based on the ORC cycle has a higher thermodynamic efficiency and lower cost of equipment, and consequently, it is more suitable for the future geothermal plant in Velika Ciglena. - Highlights: ► Paper presents the analysis of binary geothermal plant for the utilization of recourses in Velika Ciglena field (Croatia). ► Thermodynamic and economical parameters of both cycles are calculated by the spreadsheet simulation software. ► The results of performance analysis indicate the advantage of electricity production based on ORC cycle

  7. Analysis of successful rate factors for small and medium enterprises in furniture manufacturing sector in Klaten Regency - Central Java, Indonesia

    Science.gov (United States)

    Budhi Utomo, R.; Lasminiasih; Prajaka, S.

    2018-03-01

    Small and Medium Enterprises (SMEs) are business activities that can expand the level of employment rate and provide economic services to the wider community and can play a role in the process of equalizing and improving people’s income, stimulating economic growth as well as realizing national stabilities. The aim of this study is to identify the factors of the success rate for Small and Medium Enterprises (SMEs) in furniture manufacturing sector in Klaten regency, Central Java, Indonesia. The method employed in this study was descriptive qualitative by also employing quantitative analysis of which the data were collected through observations, interviews and by administering questionnaires. The results seemed to indicate that the furniture business in Klaten is still experiencing difficulties in managing its various aspects of business, namely in terms of marketing (either directly or indirectly or by making the best use media of technology) and managing capital. All this time, the SMEs in furniture manufacturing sector in Klaten have been utilizing a very simple system in producing tables, chairs, wardrobes and any other furniture products which are then distributed to be sold by larger furniture companies. This condition makes the SMEs unable to be independent in running their business.

  8. Analysis of the neutrons dispersion in a semi-infinite medium based in transport theory and the Monte Carlo method

    International Nuclear Information System (INIS)

    Arreola V, G.; Vazquez R, R.; Guzman A, J. R.

    2012-10-01

    In this work a comparative analysis of the results for the neutrons dispersion in a not multiplicative semi-infinite medium is presented. One of the frontiers of this medium is located in the origin of coordinates, where a neutrons source in beam form, i.e., μο=1 is also. The neutrons dispersion is studied on the statistical method of Monte Carlo and through the unidimensional transport theory and for an energy group. The application of transport theory gives a semi-analytic solution for this problem while the statistical solution for the flow was obtained applying the MCNPX code. The dispersion in light water and heavy water was studied. A first remarkable result is that both methods locate the maximum of the neutrons distribution to less than two mean free trajectories of transport for heavy water, while for the light water is less than ten mean free trajectories of transport; the differences between both methods is major for the light water case. A second remarkable result is that the tendency of both distributions is similar in small mean free trajectories, while in big mean free trajectories the transport theory spreads to an asymptote value and the solution in base statistical method spreads to zero. The existence of a neutron current of low energy and toward the source is demonstrated, in contrary sense to the neutron current of high energy coming from the own source. (Author)

  9. Assessment of medium-term cardiovascular disease risk after Japan's 2011 Fukushima Daiichi nuclear accident: a retrospective analysis.

    Science.gov (United States)

    Toda, Haruka; Nomura, Shuhei; Gilmour, Stuart; Tsubokura, Masaharu; Oikawa, Tomoyoshi; Lee, Kiwon; Kiyabu, Grace Y; Shibuya, Kenji

    2017-12-22

    To assess the medium-term indirect impact of the 2011 Fukushima Daiichi nuclear accident on cardiovascular disease (CVD) risks and to identify whether risk factors for CVD changed after the accident. Residents aged 40 years and over participating in annual public health check-ups from 2009 to 2012, administered by Minamisoma city, located about 10 to 40 km from the Fukushima Daiichi nuclear plant. The sex-specific Framingham CVD risk score was considered as the outcome measure and was compared before (2009-2010) and after the accident (2011-2012). A multivariate regression analysis was employed to evaluate risk factors for CVD. Data from 563 individuals (60.2% women) aged 40 to 74 years who participated in the check-ups throughout the study period was analysed. After adjusting for covariates, no statistically significant change was identified in the CVD risk score postaccident in both sexes, which may suggest no obvious medium-term health impact of the Fukushima nuclear accident on CVD risk. The risk factors for CVD and their magnitude and direction (positive/negative) did not change after the accident. There was no obvious increase in CVD risks in Minamisoma city, which may indicate successful management of health risks associated with CVD in the study sample. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2017. All rights reserved. No commercial use is permitted unless otherwise expressly granted.

  10. Survey of wet and dry spent fuel storage

    International Nuclear Information System (INIS)

    1999-07-01

    Spent fuel storage is one of the important stages in the nuclear fuel cycle and stands among the most vital challenges for countries operating nuclear power plants. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for exchanging information and for coordinating and encouraging closer co-operation among Member States. Spent fuel management is recognized as a high priority IAEA activity. In 1997, the annual spent fuel arising from all types of power reactors worldwide amounted to about 10,500 tonnes heavy metal (t HM). The total amount of spent fuel accumulated worldwide at the end of 1997 was about 200,000 t HM of which about 130,000 t HM of spent fuel is presently being stored in at-reactor (AR) or away-from-reactor (AFR) storage facilities awaiting either reprocessing or final disposal and 70,000 t HM has been reprocessed. Projections indicate that the cumulative amount generated by 2010 may surpass 340,000 t HM and by the year 2015 395,000 t HM. Part of the spent fuel will be reprocessed and some countries took the option to dispose their spent fuel in a repository. Most countries with nuclear programmes are using the deferral of a decision approach, a 'wait and see' strategy with interim storage, which provides the ability to monitor the storage continuously and to retrieve the spent fuel later for either direct disposal or reprocessing. Some countries use different approaches for different types of fuel. Today the worldwide reprocessing capacity is only a fraction of the total spent fuel arising and since no final repository has yet been constructed, there will be an increasing demand for interim storage. The present survey contains information on the basic storage technologies and facility types, experience with wet and dry storage of spent fuel and international experience in spent fuel transport. The main aim is to provide spent fuel

  11. XMM-Newton Science Analysis System (SAS): medium and long term strategy

    Science.gov (United States)

    Gabriel, C.

    2017-10-01

    The Science Analysis System (SAS) is so far the only analysis software able to process the data taken with all the scientific instruments on board the XMM-Newton satellite. It also represents the core of the Pipeline Processing System (PPS), the system used to reduce all XMM-Newton observations for providing final calibrated products to the observers, and to populate the XMM-Newton Scientific Archive (XSA), as the source of data for the whole community. While maintaining and extending SAS, recognised as a fundamental component of optimising the mission's science output, we are in the process of simplifying it, modernising its code, and maximising its longevity for post-mission times.

  12. Local interstellar medium

    International Nuclear Information System (INIS)

    Crutcher, R.M.; and Radio Astronomy Laboratory, University of California, Berkeley)

    1982-01-01

    Analysis of the velocities of optical interstellar lines shows that the Sun is immersed in a coherently moving local interstellar medium whose velocity vector agrees with that of the interstellar wind observed through backscatter of solar H Lyα and He lambda584 photons. The local interstellar medium consists of both cool clouds and warm intercloud medium gas, has a mass of perhaps approx.30 M/sub sun/, does not have severe depletion of trace elements from the gas phase, and appears to be material which has been shocked and accelerated by stellar winds and supernovae associated with the Sco-Oph OB association

  13. Analysis of Heat Release from Gain Medium of Chemical Oxygen Iodine Laser

    Science.gov (United States)

    Takeuchi, Noriyuki; Sugimoto, Daichi; Tei, Kazuyoku; Nanri, Kenzo; Fujioka, Tomoo

    2005-02-01

    Heat release into the operating gas of a chemical oxygen iodine laser is analyzed on the basis of stagnation and cavity pressures. The energy of excited oxygen molecules is released as heat in this device through pooling reactions, iodine dissociation, and the interactions of these processes with water vapor. The proposed estimation method is applied to the analysis of subsonic and transonic iodine injection schemes to examine energy loss during iodine dissociation. The results also provide the number n of excited oxygen molecules consumed in each iodine dissociation. The values of n were estimated to be n≥ 10 and n≥ 6 in the subsonic and transonic injection schemes, respectively.

  14. A model for transient analysis of a multiple-medium confinement filter system

    International Nuclear Information System (INIS)

    Hyder, M.L.; Ellison, P.G.; Leonard, M.T.; Louie, D.L.Y.; Donbroski, E.L.; Wagner, K.C.

    1990-01-01

    A computational model is described that calculates the transient behavior of aerosol and vapor (adsorption) filter compartments such as those used in the Savannah River Site (SRS) production reactor confinement system. The principal application of the model is in the analysis of confinement response to hypothetical severe (core melt) accidents. Under these conditions, aerosol and radio-iodine deposition on filter compartments may be substantial. Attendant filter degradation mechanisms are modeled. Sample calculations are included to illustrate model performance. 6 refs., 14 figs., 1 tab

  15. Spent fuel management in Spain

    International Nuclear Information System (INIS)

    Gonzalez, J.L.

    2002-01-01

    The spent fuel management strategy in Spain is presented. The strategy includes temporary solutions and plans for final disposal. The need for R and D including partitioning and transmutation, as well as the financial constraints are also addressed. (author)

  16. Intermodal transportation of spent fuel

    International Nuclear Information System (INIS)

    Elder, H.K.

    1983-09-01

    Concepts for transportation of spent fuel in rail casks from nuclear power plant sites with no rail service are under consideration by the US Department of Energy in the Commercial Spent Fuel Management program at the Pacific Northwest Laboratory. This report identifies and evaluates three alternative systems for intermodal transfer of spent fuel: heavy-haul truck to rail, barge to rail, and barge to heavy-haul truck. This report concludes that, with some modifications and provisions for new equipment, existing rail and marine systems can provide a transportation base for the intermodal transfer of spent fuel to federal interim storage facilities. Some needed land transportation support and loading and unloading equipment does not currently exist. There are insufficient shipping casks available at this time, but the industrial capability to meet projected needs appears adequate

  17. Performance Analysis of a Medium Frequency Offshore Grid for Identification Of Vessels Sailing on High Density Maritime European Routes

    Directory of Open Access Journals (Sweden)

    López Alfonso

    2017-12-01

    Full Text Available The paper analyses the performance of an Automatic Vessel Identification System on Medium Frequency (AVISOMEF, which works with the Grid Method (GM on high density maritime European routes using real data and uniformly distributed data. Compared to other systems, AVISOMEF is a novelty, as it is not a satellite system, nor is it limited by a given coverage distance, in contrast to the Automatic Identification System (AIS, though in exceptional circumstances it leans towards it. To perform the analysis, special simulation software was developed. Moreover, a number of maritime routes along with their traffic density data were selected for the study. For each route, two simulations were performed, the first of which based on the uniform traffic distribution along the route, while the second one made use of real AIS data positioning of vessels sailing on the selected routes. The obtained results for both simulations made the basis for formulating conclusions regarding the capacity of selected routes to support AVISOMEF.

  18. 2D and 3D parameter images for analysis of contrast medium enhancement based on dynamic CT and MR

    International Nuclear Information System (INIS)

    Beier, J.; Stroszczynski, C.; Oellinger, H.; Felix, R.; Buege, T.; Fleck, E.

    1998-01-01

    Aim: For dynamic contrast medium (CM) studies, parameter images exploit specific features of the time/intensity curve (TIC) of each pixel and represent these values in a new image. Existing concepts of two-dimensional CM analysis are extended for three-dimensional applications using adequate computer graphic visualization. Results: In first-pass analyses, TMIP and TG allowed the simultaneous or separted presentation of different temporal phases of the CM bolus. Correlation images emphasized regions with similarities to given TIC patterns. Three-dimensional computer graphic techniques enabled (1) anatomical/function mapping of original image and CM accumulation and (2) fused display of both spatial CM enhancement and color-encoded time of TIC peak in one common image. Conclusions: The quantifiction of presence, magnitude, and time-of-peak of CM accumulation in local image regions supports the assessment of vascularization and of ischemic or necrotic areas. (orig./AJ) [de

  19. Energy Storage Characteristic Analysis of Voltage Sags Compensation for UPQC Based on MMC for Medium Voltage Distribution System

    Directory of Open Access Journals (Sweden)

    Yongchun Yang

    2018-04-01

    Full Text Available The modular multilevel converter (MMC, as a new type of voltage source converter, is increasingly used because it is a distributed storage system. There are many advantages of using the topological structure of the MMC on a unified power quality controller (UPQC, and voltage sag mitigation is an important use of the MMC energy storage system for the power quality compensation process. In this paper, based on the analysis of the topology of the MMC, the essence of energy conversion in a UPQC of voltage sag compensation is analyzed; then, the energy storage characteristics are calculated and analyzed to determine the performance index of voltage sag compensation; in addition, the simulation method is used to verify the voltage sag compensation characteristics of the UPQC; finally, an industrial prototype of the UPQC based on an MMC for 10 kV of medium voltage distribution network has been developed, and the basic functions of UPQC have been tested.

  20. HAM on MHD Convective Flow of a Third grade Fluid through Porous Medium during Wire Coating Analysis with Hall effects

    Science.gov (United States)

    Reddy, B. Siva Kumar; Surya Narayana Rao, K. V.; Bhuvana Vijaya, R.

    2017-08-01

    In this study, wire coating is performed using MHD convective flow of third grade fluid through porous medium taking Hall current into account. The governing equations are first modelled and then solved analytically by utilizing the Homotopy analysis method (HAM). The convergence of the series solution is established. The effect of pertinent parameters on the velocity field and temperature profile is shown with the help of graphs. It is observed that the velocity profiles increase as the value of visco-elastic third grade fluid parameter β increase and decrease as the Hartmann number M and permeability parameter K increase. It is also observed that the temperature profiles increases as the Brinkman number Br, permeability parameter K, magnetic parameter M and third grade fluid parameter β increase.

  1. Analysis of the quality of service of small and medium size hotel companies in Bucaramanga and its metropolitan area

    Directory of Open Access Journals (Sweden)

    Carolina Monsalve Castro

    2015-12-01

    Full Text Available In recent years, the hotel industry has been facing an increment of travelers willing to enjoy better conditions in tourist destinations. The demanding changes in tourist preferences, and the increasing range of offers that they find to meet their needs, has made the hotel industry to focus its attention to service. Therefore, in this document, the principles that impact the evaluation of a quality service are analyzed. This research takes a sample of 384 small and medium size hotel companies’ guests. A Likert type questionnaire is applied. Its evaluation is conducted through SPSS software through bivariate analysis. In the results, influential aspects were identified such as: encouraging customer loyalty, special offers, innovation, premises and equipment, human talent training, among others.

  2. A Semiotic Analysis of Visual Elements Particular to the Medium of Comics in Alan Ford by Max Bunker and Magnus

    Directory of Open Access Journals (Sweden)

    Hrvoje Gržina

    2014-12-01

    Full Text Available This paper presents the results of a semiotic analysis of visual elements characteristic for the medium of comic books applied to the first seventy-five issues of the Croatian edition of Alan Ford. After a description of the cultural and historical framework, it analyzes individual signs in comics and different elements specific for expression in comic books in Western culture with the aim of exploring which of these signs are present in Alan Ford, and to what extent. The results show that the analyzed comic book is deeply rooted in the visual and literary Western tradition, and that it contains virtually all the characteristic elements of representation in comic books. However, the paper also concludes that certain iconic elements of the vocabulary of comics – i.e. onomatopoeic neologisms – are to a certain extent specific and typical only for Alan Ford.

  3. Study of discharges produced by surface waves under medium and high pressure: application to chemical analysis

    International Nuclear Information System (INIS)

    Laye epouse Granier, Agnes

    1986-01-01

    This report deals with the study of microwave discharges produced in argon gas by surface waves in the 20-760 Torr pressure range. Application to chemical analysis by emission optical spectroscopy is also investigated. First of all we study the propagation of a surface wave in a bounded plasma in which the effective collision frequency for momentum transfer ν is higher than the excitation one. The axial electron density profile is determined from two diagnostic techniques, i.e., phase variations of the wave field and Stark broadening of H β line. Then we deduce the discharge characteristics ν, θ (maintaining power of an electron-ion pair) and E eff (effective electric field for discharge sustaining) from the electron density profile. Then an energy balance of the discharge is developed. It explains the change of operating conditions in the 20-50 Torr range. At low pressure the discharge is governed by ambipolar diffusion whereas at high pressure, the electrons are mainly lost by volume recombination of Ar 2 + . Finally, we report on chemical analysis experiment of gases (optimum sensibility in found near 100 Torr) and of metallic solutions sprayed by a graphite oven. Performances of such a design and ICP plasma torches are compared. (author) [fr

  4. Cytogenetic analysis of peripheral blood lymphocytes after arteriography (exposure to x-rays and contrast medium)

    International Nuclear Information System (INIS)

    Popova, L.; Hadjidekova, V.; Karadjov, G.; Agova, S.; Traskov, D.; Hadjidekov, V.

    2005-01-01

    Backgrounds. The purpose of our study is to investigate the cytogenetic analysis findings in peripheral blood lymphocytes of 29 patients who had undergone diagnostic radiography. Methods. Peripheral blood samples were taken from 22 patients submitted to renal arteriography and 7 patients submitted to cerebral arteriography (17 male and 12 female, aged between 13-68 years). Cytogenetic analyses of peripheral lymphocytes were performed before the procedure, immediately after and 24 hours later. The entrance skin dose obtained during the whole diagnostic X-ray exposure was measured by thermoluminescent dosimeters and varied between 0.03-0.30 Gy. Both low and high osmolarity contrast media were used. Chromosomal aberrations and micronuclei frequency were used as biomarkers of genotoxicity. Results. The estimated frequency of chromosomal aberrations and micronuclei in the peripheral blood lymphocytes of patients after arteriography examination was significantly higher than the level before the diagnostic exposure. The mean frequency of cells with chromosomal aberrations was nearly double after examination and proved to be constant in the analysis after 24 hours. Conclusions. Radiological diagnostic procedures involving iodinated contrast media as arteriography may cause a significant increase in cytogenetic damage in peripheral blood lymphocytes. (author)

  5. Stakeholder Analysis as a Medium to Aid Change in Information System Reengineering Projects

    Directory of Open Access Journals (Sweden)

    Jean Davison

    2004-04-01

    Full Text Available The importance of involving stakeholders within a change process is well recognised, and successfully managed change is equally important. Information systems development and redesign is a form of change activity involving people and social issues and therefore resistance to change may occur. A stakeholder identification and analysis (SIA technique has been developed as an enhancement to PISO® (Process Improvement for Strategic Objectives, a method that engages the users of a system in the problem solving and reengineering of their own work-based problem areas. The SIA technique aids the identification and analysis of system stakeholders, and helps view the projected outcome of system changes and their effect on relevant stakeholders with attention being given to change resistance to ensure smooth negotiation and achieve consensus. A case study is presented here describing the successful implementation of a direct appointment booking system for patients within the National Health Service in the UK, utilising the SIA technique, which resulted in a feeling of empowerment and ownership of the change of those involved.

  6. A Smoothing Technique for the Multifractal Analysis of a Medium Voltage Feeders Electric Current

    Science.gov (United States)

    de Santis, Enrico; Sadeghian, Alireza; Rizzi, Antonello

    2017-12-01

    The current paper presents a data-driven detrending technique allowing to smooth complex sinusoidal trends from a real-world electric load time series before applying the Detrended Multifractal Fluctuation Analysis (MFDFA). The algorithm we call Smoothed Sort and Cut Fourier Detrending (SSC-FD) is based on a suitable smoothing of high power periodicities operating directly in the Fourier spectrum through a polynomial fitting technique of the DFT. The main aim consists of disambiguating the characteristic slow varying periodicities, that can impair the MFDFA analysis, from the residual signal in order to study its correlation properties. The algorithm performances are evaluated on a simple benchmark test consisting of a persistent series where the Hurst exponent is known, with superimposed ten sinusoidal harmonics. Moreover, the behavior of the algorithm parameters is assessed computing the MFDFA on the well-known sunspot data, whose correlation characteristics are reported in literature. In both cases, the SSC-FD method eliminates the apparent crossover induced by the synthetic and natural periodicities. Results are compared with some existing detrending methods within the MFDFA paradigm. Finally, a study of the multifractal characteristics of the electric load time series detrendended by the SSC-FD algorithm is provided, showing a strong persistent behavior and an appreciable amplitude of the multifractal spectrum that allows to conclude that the series at hand has multifractal characteristics.

  7. IAEA spent fuel storage glossary

    International Nuclear Information System (INIS)

    1985-10-01

    The aim of this glossary is to provide a basis for improved international understanding of terms used in the important area of spent fuel storage technology. The glossary is the product of an IAEA Consultant Group with valuable input from a substantial list of reviewers. The glossary emphasizes fuel storage relevant to power reactors, but is also widely applicable to research reactors. The intention is to define terms from current technologies. Terms are limited to those directly related to spent fuel storage

  8. Spent-fuel-storage alternatives

    International Nuclear Information System (INIS)

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed

  9. A review and analysis of European industrial experience in handling LWR [light water reactor] spent fuel and vitrified high-level waste

    International Nuclear Information System (INIS)

    Blomeke, J.O.

    1988-06-01

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performances of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States. 79 refs., 71 figs., 10 tabs

  10. Analysis of eddy current induced in track on medium-low speed maglev train

    Science.gov (United States)

    Li, Guanchun; Jia, Zhen; He, Guang; Li, Jie

    2017-06-01

    Electromagnetic levitation (EMS) maglev train relies on the attraction between the electromagnets and rails which are mounted on the train to achieve suspension. During the movement, the magnetic field generated by the electromagnet will induce the eddy current in the orbit and the eddy current will weaken the suspended magnetic field. Which leads to the attenuation of the levitation force, the increases of suspension current and the degradation the suspension performance. In this paper, the influence of eddy current on the air gap magnetic field is solved by theoretical analysis, and the correction coefficient of air gap magnetic field is fitted according to the finite element data. The levitation force and current are calculated by the modified formula, and the velocity curves of the levitation force and current are obtained. The results show that the eddy current effect increases the load power by 61.9% in the case of heavy loads.

  11. Variational analysis for simulating free-surface flows in a porous medium

    Directory of Open Access Journals (Sweden)

    Shabbir Ahmed

    2003-01-01

    is used to obtain a discrete form of equations for a two-dimensional domain. The matrix characteristics and the stability criteria have been investigated to develop a stable numerical algorithm for solving the governing equation. A computer programme has been written to solve a symmetric positive definite system obtained from the variational finite element analysis. The system of equations is solved using the conjugate gradient method. The solution generates time-varying hydraulic heads in the subsurface. The interfacing free surface between the unsaturated and saturated zones in the variably saturated domain is located, based on the computed hydraulic heads. Example problems are investigated. The finite element solutions are compared with the exact solutions for the example problems. The numerical characteristics of the finite element solution method are also investigated using the example problems.

  12. Analysis of Strategic Management in Small and Medium Business Organizations in Graphic Industry

    Directory of Open Access Journals (Sweden)

    Ilija Jolevski

    2013-06-01

    Full Text Available Graphic industry as a field of work and a field of study offers ability for economic growth of every country or region. The increase in competition and the development of the market method of prediction, increase the need for business organizations to continuously observe their own position. Faced with a variable, turbulent and complex environment, the business organizations adopt strategic management, and modern managers, especially those in top management, have to possess knowledge and skills of the concepts and application of strategic management.In this paper the case method, as a method used in strategic management, will be applied as a general approach of strategic analysis in creating strategy map for businesses in graphic industry.

  13. Thermodynamic and economic analysis and optimization of power cycles for a medium temperature geothermal resource

    International Nuclear Information System (INIS)

    Coskun, Ahmet; Bolatturk, Ali; Kanoglu, Mehmet

    2014-01-01

    Highlights: • We conduct the thermodynamic and economic analysis of various geothermal power cycles. • The optimization process was performed to minimize the exergy losses. • Kalina cycle is a new technology compared to flash and binary cycles. • It is shown that Kalina cycle presents a viable choice for both thermodynamically and economically. - Abstract: Geothermal power generation technologies are well established and there are numerous power plants operating worldwide. Turkey is rich in geothermal resources while most resources are not exploited for power production. In this study, we consider geothermal resources in Kutahya–Simav region having geothermal water at a temperature suitable for power generation. The study is aimed to yield the method of the most effective use of the geothermal resource and a rational thermodynamic and economic comparison of various cycles for a given resource. The cycles considered include double-flash, binary, combined flash/binary, and Kalina cycle. The selected cycles are optimized for the turbine inlet pressure that would generate maximum power output and energy and exergy efficiencies. The distribution of exergy in plant components and processes are shown using tables. Maximum first law efficiencies vary between 6.9% and 10.6% while the second law efficiencies vary between 38.5% and 59.3% depending on the cycle considered. The maximum power output, the first law, and the second law efficiencies are obtained for Kalina cycle followed by combined cycle and binary cycle. An economic analysis of four cycles considered indicates that the cost of producing a unit amount of electricity is 0.0116 $/kW h for double flash and Kalina cycles, 0.0165 $/kW h for combined cycle and 0.0202 $/kW h for binary cycle. Consequently, the payback period is 5.8 years for double flash and Kalina cycles while it is 8.3 years for combined cycle and 9 years for binary cycle

  14. Medium Term Analysis of Technical and Allocative Efficiency in Romanian Farms Using FADN Dataset

    Directory of Open Access Journals (Sweden)

    Nicola GALLUZZO

    2017-05-01

    Full Text Available The Farm Accountancy Data Network is an annual survey proposed by the European Union in order to estimate the impact of the Common Agricultural Policy on farmers. Lots of scholars have investigated the technical, economical and allocative efficiency using a non parametric approach such as the Data Envelopment Analysis (DEA in Romanian farms throughout the Farm Accountancy Data Network dataset pointing out poor levels of technical efficiency, which were lower than the average European value. The purpose of this study was to assess using DEA approach technical, economic and allocative efficiency in Romanian farms part of the FADN dataset over six year time from 2007 to 2012. Findings pointed out an increase of technical efficiency compared to previous studies, as a consequence of a significant turn over of a younger high skill and qualified farmers generation. Poor land capital, in terms of utilized agricultural areas, connected to an increase of new technologies, was the downside of Romanian farms and this implied that the National Rural Development Plan should  have taken into account financial subsides in order to implement agricultural areas scattered in Romanian rural space.

  15. Kinetic Analysis of the Anodic Carbon Oxidation Mechanism in a Molten Carbonate Medium

    International Nuclear Information System (INIS)

    Allen, Jessica A.; Tulloch, John; Wibberley, Louis; Donne, Scott W.

    2014-01-01

    The oxidation mechanism for carbon in a carbonate melt was modelled using an electrochemical kinetic approach. Through the Butler-Volmer equation for electrode kinetics, a series of expressions was derived assuming each step of the proposed carbon oxidation mechanism is in turn the rate determining step (RDS). Through the derived expressions the transfer coefficient and Tafel slope were calculated for each possible RDS of the proposed mechanism and these were compared with real data collected on carbon based electrodes including graphite and coal. It was established that the RDS of the electrochemical oxidation process is dependent on both the carbon type and the potential region of oxidation. The simplified kinetic analysis suggested that the RDS in the main oxidation region is likely to be the first or second electron transfer on a graphite electrode surface, which occurs following initial adsorption of an oxygen anion to an active carbon site. This is contrary to previous suggestions that adsorption of the second anion to the carbon surface will be rate determining. It was further shown that use of a coal based carbon introduces a change in mechanism with an additional reaction region where a different mechanism is proposed to be operating

  16. Entropy generation analysis for viscoelastic MHD flow over a stretching sheet embedded in a porous medium

    Directory of Open Access Journals (Sweden)

    S. Baag

    2017-12-01

    Full Text Available In this paper it is intended to analyse entropy generation by applying second law of thermodynamics to magnetohydrodynamic flow, heat and mass transfer of an electrically conducting viscoelastic liquid (Walters B′ past on a stretching surface, taking into account the effects of Joule dissipation, viscous dissipation and Darcy dissipation, and internal heat generation. The boundary layer equations are solved analytically by using Kummer’s function. The entropy generation has been computed considering Darcy dissipation besides viscous and Joule dissipation. Results for some special cases of the present analysis are in good agreement with the existing literature. Increase in viscoelastic and magnetic parameter reduces the velocity. Increase in elastic parameter causes a greater retardation in the velocity. Presence of porous matrix enhances temperature whereas increase in Prandtl number decreases the temperature. One striking result of the present study is that Darcy dissipation favours higher level entropy generation in all the cases except the flow of liquid with low thermal diffusivity assuming the process to be irreversible.

  17. Extraction and isotopic analysis of medium molecular weight hydrocarbons from Murchison using supercritical carbon dioxide

    Science.gov (United States)

    Gilmour, Iain; Pillinger, Colin

    1993-03-01

    The large variety of organic compounds present in carbonaceous chondrites poses particular problems in their analysis not the least of which is terrestrial contamination. Conventional analytical approaches employ simple chromatographic techniques to fractionate the extractable compounds into broad classes of similar chemical structure. However, the use of organic solvents and their subsequent removal by evaporation results in the depletion or loss of semi-volatile compounds as well as requiring considerable preparative work to assure solvent purity. Supercritical fluids have been shown to provide a powerful alternative to conventional liquid organic solvents used for analytical extractions. A sample of Murchison from the Field Museum was analyzed. Two interior fragments were used; the first (2.85 g) was crushed in an agate pestel and mortar to a grain size of ca. 50-100 micron, the second (1.80 g) was broken into chips 3-8 mm in size. Each sample was loaded into a stainless steel bomb and placed in the extraction chamber of an Isco supercritical fluid extractor maintained at 35 C. High purity (99.9995 percent) carbon dioxide was used and was pressurized using an Isco syringe pump. The samples were extracted dynamically by flowing CO2 under pressure through the bomb and venting via a 50 micron fused filica capillary into 5 mls of hexane used as a collection solvent. The hexane was maintained at a temperature of 0.5 C. A series of extractions were done on each sample using CO2 of increasing density. The principal components extracted in each fraction are summarized.

  18. Transportation of spent nuclear fuels

    International Nuclear Information System (INIS)

    Meguro, Toshiichi

    1976-01-01

    The spent nuclear fuel taken out of reactors is cooled in the cooling pool in each power station for a definite time, then transported to a reprocessing plant. At present, there is no reprocessing plant in Japan, therefore the spent nuclear fuel is shipped abroad. In this paper, the experiences and the present situation in Japan are described on the transport of the spent nuclear fuel from light water reactors, centering around the works in Tsuruga Power Station, Japan Atomic Power Co. The spent nuclear fuel in Tsuruga Power Station was first transported in Apr. 1973, and since then, about 36 tons were shipped to Britain by 5 times of transport. The reprocessing plant in Japan is expected to start operation in Apr. 1977, accordingly the spent nuclear fuel used for the trial will be transported in Japan in the latter half of this year. Among the permission and approval required for the transport of spent nuclear fuel, the acquisition of the certificate for transport casks and the approval of land and sea transports are main tasks. The relevant laws are the law concerning the regulations of nuclear raw material, nuclear fuel and reactors and the law concerning the safety of ships. The casks used in Tsuruga Power Station and EXL III type, and the charging of spent nuclear fuel, the decontamination of the casks, the leak test, land transport with a self-running vehicle, loading on board an exclusive carrier and sea transport are briefly explained. The casks and the ship for domestic transport are being prepared. (Kato, I.)

  19. Robotic cleaning of a spent fuel pool

    International Nuclear Information System (INIS)

    Roman, H.T.; Marian, F.A.; Silverman, E.B.; Barkley, V.P.

    1987-01-01

    Spent fuel pools at nuclear power plants are not cleaned routinely, other than by purifying the water that they contain. Yet, debris can collect on the bottom of a pool and should be removed prior to fuel transfer. At Public Service Electric and Gas Company's Hope Creek Nuclear Power Plant, a submersible mobile robot - ARD Corporation's SCAVENGER - was used to clean the bottom of the spent fuel pool prior to initial fuel loading. The robotic device was operated remotely (as opposed to autonomously) with a simple forward/reverse control, and it cleaned 70-80% of the pool bottom. This paper reports that a simple cost-benefit analysis shows that the robotic device would be less expensive, on a per mission basis, than other cleaning alternatives, especially if it were used for other similar cleaning operations throughout the plant

  20. Forced-Vibration Analysis of a Coupled System of SLGSs by Visco- Pasternak Medium Subjected to a Moving Nano-particle

    Directory of Open Access Journals (Sweden)

    A. Ghorbanpour-Arani

    2013-06-01

    Full Text Available In this study, forced-vibration analysis of a coupled system of single layered graphene sheets (SLGSs subjected to the moving nano-particle is carried out based on nonlocal elasticity theory of orthotropic plate. Two SLGSs are coupled with elastic medium which is simulated by Pasternak and Visco-Pasternak models. Using Hamilton’s principle, governing differential equations of motion are derived and solved analytically. The effects of small scale, aspect ratio, velocity of nano-particle, time parameter, mechanical properties of graphene sheets, Visco-elastic medium on the maximum dynamic responses of each SLGSs are studied. Results indicate that, if the medium (elastic or visco-elastic medium of coupled system becomes more rigid, the maximum dynamic displacements of both SLGSs will be closer together.

  1. Spent fuel management in Japan - Facts and prospects

    International Nuclear Information System (INIS)

    Nagano, K.

    2002-01-01

    This paper discusses recent developments and future issues related to spent fuel management in Japan. With increasing pressure of spent fuel discharge from the power plants in operation and, in contrast, uncertainties in their processing and management services, spent fuel storage in short and medium terms has been receiving the highest priority in nuclear policy discussions in Japan. While small-scale interim storage devices, as well as capacity expansion (re-racking, etc.) and shared uses of existing devices, are introduced at number of power stations, large scale AFR (away from reactor) 'Storage of Recycle Fuel Resources' is expected to come in a medium and long-run. Commercial operation of 'Storage of Recycle Fuel Resources' is allowed its way, as the bill of amendment to the law for regulation of nuclear power reactors and other nuclear-related activities has passed in the Diet. In the meantime, the Atomic Energy Commission has launched working group discussions for revision of 'The Long-term Program of Research, Development and Utilization of Nuclear Energy' to be completed in 2000. This revision is hoped to set up a stage of national debate of nuclear policy, which might lead to fill conceptual gaps between bodies promoting nuclear development and general public. The author's attempt to illustrate the role of storage in spent fuel management is also presented from a theoretical point of view. (author)

  2. SWOT ANALYSIS MICRO SMALL MEDIUM ENTREPRISE (MSME GEULIS CRAFT UMBRELLA TO SUCCESS IN LOCAL ECONOMIC DEVELOPMENT RESOURCES DISTRICT TASIKMALAYA

    Directory of Open Access Journals (Sweden)

    K. Dianta A. Sebayang

    2012-03-01

    Full Text Available Development of local economic resources, a new trend in the effort to increase the income of the community and the region. Local factors that determine both in terms of natural resources (raw materials and human resources (labor. This paper attempts to present how small and medium enterprises "Kerajinan Payung Geulis" try to improve economic development based on the development of local economic resources in Tasikmalaya. This study aims to illustrate the potential of entrepreneurs that include the competence and commitment of entrepreneurs in small business business, and to illustrate the strength of business / competitive position, business profile and entrepreneur influenced by environmental condition of external and internal environment, seen from the positive and negative side. The research was conducted on umbrella industry of handicraft business in Tasikmalaya. The method used in this research is descriptive analysis by using SWOT analysis. The results show many problems encountered and very complex, such as: low quality of human resources, limited business capital, low access to markets, access to financial institutions / banks are absent, administrative procedures ignorance, sustainability and limited capacity production; Coupled with the business climate is not conducive to the development of SMEs and entrepreneurship.

  3. 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

  4. Container for spent fuel assembly

    International Nuclear Information System (INIS)

    Sawai, Takeshi.

    1996-01-01

    The container of the present invention comprises a container main body having a body portion which can contain spent fuel assemblies and a lid, and heat pipes having an evaporation portion disposed along the outer surface of the spent fuel assemblies to be contained and a condensation portion exposed to the outside of the container main body. Further, the heat pipe is formed spirally at the evaporation portions so as to surround the outer circumference of the spent fuel assemblies, branched into a plurality of portions at the condensation portion, each of the branched portion of the condensation portion being exposed to the outside of the container main body, and is tightly in contact with the periphery of the slit portions disposed to the container main body. Then, since released after heat is transferred to the outside of the container main body from the evaporation portion of the heat pipe along the outer surface of the spent fuel assemblies by way of the condensation portion of the heat pipes exposed to the outside of the container main body, the efficiency of the heat transfer is extremely improved to enhance the effect of removing heat of spent fuel assemblies. Further, cooling effect is enhanced by the spiral form of the evaporation portion and the branched condensation portion. (N.H.)

  5. Spent Nuclear Fuel project, project management plan

    International Nuclear Information System (INIS)

    Fuquay, B.J.

    1995-01-01

    The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

  6. On the participation in medium- and long-distance travel: a decomposition analysis for the UK and the Netherlands

    NARCIS (Netherlands)

    Limtanakool, N.; Dijst, M.J.; Schwanen, T.

    2006-01-01

    Social and economic benefits have accrued from medium- and long-distance travel, but at the expense of the environment. Since the travel behaviour literature tends to concentrate on shortdistance trips or trips within daily urban systems, a better understanding of the factors shaping medium- and

  7. Spent Pot Lining Characterization Framework

    Science.gov (United States)

    Ospina, Gustavo; Hassan, Mohamed I.

    2017-09-01

    Spent pot lining (SPL) management represents a major concern for aluminum smelters. There are two key elements for spent pot lining management: recycling and safe storage. Spent pot lining waste can potentially have beneficial uses in co-firing in cement plants. Also, safe storage of SPL is of utmost importance. Gas generation of SPL reaction with water and ignition sensitivity must be studied. However, determining the feasibility of SPL co-firing and developing the required procedures for safe storage rely on determining experimentally all the necessary SPL properties along with the appropriate test methods, recognized by emissions standards and fire safety design codes. The applicable regulations and relevant SPL properties for this purpose are presented along with the corresponding test methods.

  8. Spent fuel management in Japan

    International Nuclear Information System (INIS)

    Shirahashi, K.; Maeda, M.; Nakai, T.

    1996-01-01

    Japan has scarce energy resources and depends on foreign resources for 84% of its energy needs. Therefore, Japan has made efforts to utilize nuclear power as a key energy source since mid-1950's. Today, the nuclear energy produced from 49 nuclear power plants is responsible for about 31% of Japan's total electricity supply. The cumulative amount of spent fuel generated as of March 1995 was about 11,600 Mg U. Japan's policy of spent fuel management is to reprocess spent nuclear fuel and recycle recovered plutonium and uranium as nuclear fuel. The Tokai reprocessing plant continues stable operation keeping the annual treatment capacity or around 90 Mg U. A commercial reprocessing plant is under construction at Rokkasho, northern part of Japan. Although FBR is the principal reactor to use plutonium, LWR will be a major power source for some time and recycling of the fuel in LWRs will be prompted. (author). 3 figs

  9. Spent fuel management: Current status and prospects 1993

    International Nuclear Information System (INIS)

    1994-02-01

    Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and it is still one of the most vital problems common to all countries with nuclear reactors. It begins with the discharge of spent fuel from a power or a research reactor and ends with its ultimate disposition, either by direct disposal or by reprocessing of the spent fuel. Two options exist at present - an open, once-through cycle with direct disposal of the spent fuel and a closed cycle with reprocessing of the spent fuel and recycling of plutonium and uranium in new mixed oxide fuels. The selection of a spent fuel strategy is a complex procedure in which many factors have to be weighed, including political, economic and safeguards issues as well as protection of the environment. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for the exchange of information and to co-ordinate and to encourage closer co-operation among Member States in certain research an development activities that are of common interest. Refs, figs and tabs

  10. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

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

    1997-12-01

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

  11. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

  12. Spent-fuel-storage alternatives

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

  13. Spent fuel management in Spain

    International Nuclear Information System (INIS)

    Gago, J.A.; Gravalos, J.M.

    1996-01-01

    There are presently nine Light Water Reactors in operation, representing around a 34% of the overall electricity production. In the early years, a small amount of spent fuel was sent to be reprocessed, although this policy was cancelled in favor of the open cycle option. A state owned company, ENRESA, was created in 1984, which was given the mandate to manage all kinds of radioactive wastes generated in the country. Under the present scenario, a rough overall amount of 7000 tU of spent fuel will be produced during the lifetime of the plants, which will go into final disposal. (author)

  14. EXPERIMENTAL ANALYSIS AND ISHIKAWA DIAGRAM FOR BURN ON EFFECT ON MANGANESE SILICON ALLOY MEDIUM CARBON STEEL SHAFT

    Directory of Open Access Journals (Sweden)

    AsmamawTegegne

    2013-12-01

    Full Text Available Burn on/metal penetration is one of the surface defects of metal castings in general and steel castings in particular. A research on the effect of burn on the six ton medium carbon steel shaft for making a roller of cold rolled steel sheet produced at one of the metals industry was carried out. The shaft was cast using sand casting by pouring through riser/feeding head step by step (with time interval of pouring. As it was required to use foam casting method for better surface finish and dimensional accuracy of the cast, the pattern was prepared from polystyrene and embedded by silica sand. Physical observations, photographic analysis, visual inspection, measurement of depth of penetration and fish bone diagram were used as method of results analysis. The shaft produced has strongly affected by sand sintering (burn on/metal penetration. Many reasons may be the case for these defects, however analysis results showed that the use of poorly designed gating system led to turbulence flow, uncontrollable high temperature fused the silica sand and liquid polystyrene penetrated the poorly reclaimed and rammed sand mold as a result of which eroded sand has penetrated the liquid metal deeply and reacted with it, consequently after solidification and finishing the required 240mm diameter of the shaft has reduced un evenly to 133mm minimum and 229mm maximum mm that end in the rejection of the shaft from the product since it is below the required standard for the designed application. In addition, it was not possible to remove the adhered sand by grinding. Thus burn on is included in mechanical type burn on.

  15. Integration of a solar thermal system in a medium-sized brewery using pinch analysis: Methodology and case study

    International Nuclear Information System (INIS)

    Eiholzer, Tobias; Olsen, Donald; Hoffmann, Sebastian; Sturm, Barbara; Wellig, Beat

    2017-01-01

    Highlights: • Methodologies to reduce energy consumption in batch processes are presented. • Pinch analysis is used to improve energy efficiency. • Integration potential for solar heat is presented on a Scottish brewery case study. • Governmental support is important in a company’s investment in renewables. - Abstract: In the food industry a major portion of thermal energy is required for low temperature applications (below 100 °C). As a consequence, there is a significant potential to substitute fossil fuels by the use of solar heat. This paper presents a methodology that first uses pinch analysis to optimize a medium-sized Scottish brewery from a direct heat recovery perspective followed by the integration of a solar thermal system. Both the time average model and time slice model were used to determine direct and indirect heat recovery potentials. In a second stage, an optimization of a chosen integration point was conducted to assess the viability of the resulting design concept. The economic analysis includes an assessment of the impact of restrictions in the UK government’s Renewable Heat Incentive program. It was determined that since solar thermal systems are financially supported up to an installed capacity of 200 kW, solar heat can only account for a maximum of 7.7% of the heat demand based on the investigated brewery. However, if there was no limitation in capacity, from an economic point of view, the solar fraction could almost be doubled drawing into question the need for the restriction. Nevertheless, a CO 2 saving potential of approximately 38 tons per year in conjunction with a payback period of 6.4 years was determined.

  16. Development of advanced spent fuel management process

    International Nuclear Information System (INIS)

    Shin, Young Joon; Cho, S. H.; You, G. S.

    2001-04-01

    Currently, the economic advantage of any known approach to the back end fuel cycle of a nuclear power reactor has not been well established. Thus the long term storage of the spent fuel in a safe manner is one of the important issues to be resolved in countries where the nuclear power has a relatively heavy weight in power production of that country. At KAERI, as a solution to this particular issue midterm storage of the spent fuel, an alternative approach has been developed. This approach includes the decladding and pulverization process of the spent PWR fuel rod, the reducing process from the uranium oxide to a metallic uranium powder using Li metal in a LiCl salt, the continuous casting process of the reduced metal, and the recovery process of Li from mixed salts by the electrolysis. We conducted the laboratory scale tests of each processes for the technical feasibility and determination for the operational conditions for this approach. Also, we performed the theoretical safety analysis and conducted integral tests for the equipment integration through the Mock-up facility with non-radioactive samples. There were no major issues in the approach, however, material incompatibility of the alkaline metal and oxide in a salt at a high temperature and the reactor that contains the salt became a show stopper of the process. Also the difficulty of the clear separation of the salt with metals reduced from the oxide became a major issue

  17. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

    Park, B. S.; Yoon, J. S.; Hong, H. D. (and others)

    2007-02-15

    In this research, the remote handling technology was developed for the ACP application. The ACP gives a possible solution to reduce the rapidly cumulative amount of spent fuels generated from the nuclear power plants in Korea. The remote technologies developed in this work are a slitting device, a voloxidizer, a modified telescopic servo manipulator and a digital mock-up. A slitting device was developed to declad the spent fuel rod-cuts and collect the spent fuel UO{sub 2} pellets. A voloxidizer was developed to convert the spent fuel UO{sub 2} pellets obtained from the slitting process in to U{sub 3}O{sub 8} powder. Experiments were performed to test the capabilities and remote operation of the developed slitting device and voloxidizer by using simulated rod-cuts and fuel in the ACP hot cell. A telescopic servo manipulator was redesigned and manufactured improving the structure of the prototype. This servo manipulator was installed in the ACP hot cell, and the target module for maintenance of the process equipment was selected. The optimal procedures for remote operation were made through the maintenance tests by using the servo manipulator. The ACP digital mockup in a virtual environment was established to secure a reliability and safety of remote operation and maintenance. The simulation for the remote operation and maintenance was implemented and the operability was analyzed. A digital mockup about the preliminary conceptual design of an enginnering-scale ACP was established, and an analysis about a scale of facility and remote handling was accomplished. The real-time diagnostic technique was developed to detect the possible fault accidents of the slitting device. An assessment of radiation effect for various sensors was also conducted in the radiation environment.

  18. Effects of medium-chain triglycerides on weight loss and body composition: a meta-analysis of randomized controlled trials.

    Science.gov (United States)

    Mumme, Karen; Stonehouse, Welma

    2015-02-01

    Medium-chain triglycerides (MCTs) may result in negative energy balance and weight loss through increased energy expenditure and lipid oxidation. However, results from human intervention studies investigating the weight reducing potential of MCTs, have been mixed. To conduct a systematic review and meta-analysis of randomized controlled trials comparing the effects of MCTs, specifically C8:0 and C10:0, to long-chain triglycerides (LCTs) on weight loss and body composition in adults. Changes in blood lipid levels were secondary outcomes. Randomized controlled trials >3 weeks' duration conducted in healthy adults were identified searching Web of Knowledge, Discover, PubMed, Scopus, New Zealand Science, and Cochrane CENTRAL until March 2014 with no language restriction. Identified trials were assessed for bias. Mean differences were pooled and analyzed using inverse variance models with fixed effects. Heterogeneity between studies was calculated using I(2) statistic. An I(2)>50% or Pweight (-0.51 kg [95% CI-0.80 to -0.23 kg]; Preductions in body weight and composition without adversely affecting lipid profiles. However, further research is required by independent research groups using large, well-designed studies to confirm the efficacy of MCT and to determine the dosage needed for the management of a healthy body weight and composition. Copyright © 2015 Academy of Nutrition and Dietetics. Published by Elsevier Inc. All rights reserved.

  19. FANAC - a shape analysis program for resonance parameter extraction from neutron capture data for light and medium-weight nuclei

    International Nuclear Information System (INIS)

    Froehner, F.H.

    1977-11-01

    A least-squares shape analysis program is described which is used at the Karlsruhe Nuclear Research Center for the extraction of resonance parameters from high-resolution capture data. The FORTRAN program was written for light to medium-weight or near-magic target nuclei whose cross sections are characterized on one hand by broad s-wave levels with negligible Doppler broadening but pronounced multi-level interference, on the other hand by narrow p-, d- ... wave resonances with negligible multi-level interference but pronounced Doppler broadening. Accordingly the Reich-Moore multi-level formalism without Doppler broadening is used for s-wave levels, and a single-level description with Doppler braodening for p-, d- ... wave levels. Calculated capture yields are resolution broadened. Multiple-collision events are simulated by Monte Carlo techniques. Up to five different time-of-flight capture data sets can be fitted simultaneously for samples containing up to ten isotopes. Input and output examples are given and a FORTRAN list is appended. (orig.)

  20. In Vitro Propagation of Pink Lapacho: Response Surface Methodology and Factorial Analysis for Optimisation of Medium Components

    Directory of Open Access Journals (Sweden)

    Ezequiel Enrique Larraburu

    2012-01-01

    Full Text Available Handroanthus impetiginosus, pink lapacho, is a timber, ornamental, and medicinal tree. Experiments on the in vitro propagation of H. impetiginosus were conducted using nodal segments cultivated in both Murashige and Skoog salts with Gamborg vitamins (MSG and Woody Plant Medium (WPM with different concentrations of 6-benzylaminopurine (BA and indole butyric acid (IBA. Morphogenic responses were differentially affected by salt compositions and their interactions with plant growth regulators in each micropropagation stage. According to response surface analysis, the optimum multiplication rate with 1 μM IBA ranged from 16.7 to 21.3 μM BA in WPM, and the inhibitors of endogenous auxins could increase multiplication rates. A pulse with 50 μM IBA in 1/2 MSG produced 83% rooting with 3.2 roots per shoots and higher fresh and dry weights of shoots and roots. In the acclimatisation stage, 50% of plants survived after 1 year. This methodology optimised the culture media for the in vitro propagation of the H. impetiginosus clonal pool and could be applied to related species, several of which are categorised as vulnerable on the International Union for the Conservation of Nature Red List.

  1. A computational fluid dynamics analysis on stratified scavenging system of medium capacity two-stroke internal combustion engines

    Directory of Open Access Journals (Sweden)

    Pitta Srinivasa Rao

    2008-01-01

    Full Text Available The main objective of the present work is to make a computational study of stratified scavenging system in two-stroke medium capacity engines to reduce or to curb the emissions from the two-stroke engines. The 3-D flows within the cylinder are simulated using computational fluid dynamics and the code Fluent 6. Flow structures in the transfer ports and the exhaust port are predicted without the stratification and with the stratification, and are well predicted. The total pressure and velocity map from computation provided comprehensive information on the scavenging and stratification phenomenon. Analysis is carried out for the transfer ports flow and the extra port in the transfer port along with the exhaust port when the piston is moving from the top dead center to the bottom dead center, as the ports are closed, half open, three forth open, and full port opening. An unstructured cell is adopted for meshing the geometry created in CATIA software. Flow is simulated by solving governing equations namely conservation of mass momentum and energy using SIMPLE algorithm. Turbulence is modeled by high Reynolds number version k-e model. Experimental measurements are made for validating the numerical prediction. Good agreement is observed between predicted result and experimental data; that the stratification had significantly reduced the emissions and fuel economy is achieved.

  2. A Manufacturing Cost and Supply Chain Analysis of SiC Power Electronics Applicable to Medium-Voltage Motor Drives

    Energy Technology Data Exchange (ETDEWEB)

    Horowitz, Kelsey [National Renewable Energy Lab. (NREL), Golden, CO (United States); Remo, Timothy [National Renewable Energy Lab. (NREL), Golden, CO (United States); Reese, Samantha [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2017-03-24

    Wide bandgap (WBG) semiconductor devices are increasingly being considered for use in certain power electronics applications, where they can improve efficiency, performance, footprint, and, potentially, total system cost compared to systems using traditional silicon (Si) devices. Silicon carbide (SiC) devices in particular -- which are currently more mature than other WBG devices -- are poised for growth in the coming years. Today, the manufacturing of SiC wafers is concentrated in the United States, and chip production is split roughly equally between the United States, Japan, and Europe. Established contract manufacturers located throughout Asia typically carry out manufacturing of WBG power modules. We seek to understand how global manufacturing of SiC components may evolve over time by illustrating the regional cost drivers along the supply chain and providing an overview of other factors that influence where manufacturing is sited. We conduct this analysis for a particular case study where SiC devices are used in a medium-voltage motor drive.

  3. Diffusion of Innovation - The Adoption of Electronic Commerce by Small and Medium Enterprises (SMES- A Comparative Analysis

    Directory of Open Access Journals (Sweden)

    Wayne Pease

    2005-11-01

    Full Text Available This paper explores the issues that influence the diffusion of innovation as it relates to the adoption of e-commerce by Small and Medium Enterprises (SMEs. It seeks to identify factors facilitating and inhibiting such adoption across contexts – regional, small city and large city. This analysis is cross cultural so the impact of differing economic and cultural issues also will be identified in this research. Whilst it is generally accepted that the strategic use of information technology (IT is vital in the marketplace, the rate of such uptake between small and large businesses varies. This research seeks to identify the reasons for this variation. It is critical to understand such factors so that steps can be taken to redress inequity of uptake that might be identified. The paper endeavours to explore factors that are needed to facilitate and encourage IT adoption and so positively influence user acceptance and use of IT innovations in SMEs. Reasons for such uptake as well as strategic approach to the adoption of e-commerce, and variations regarding same also will be considered. The paper examines existing theory as it pertains to the diffusion of innovation acknowledging the perspective of regional and urban SMEs in various cultural contexts. Empirical investigation exploring this diffusion, the rate of and approach to the uptake by SMEs is planned using a case study methodology

  4. HPLC analysis of midodrine and desglymidodrine in culture medium: evaluation of static and shaken conditions on the biotransformation by fungi.

    Science.gov (United States)

    Barth, Thiago; Aleu, Josefina; Pupo, Mônica Tallarico; Bonato, Pierina Sueli; Collado, Isidro G

    2013-01-01

    A high-performance liquid chromatography (HPLC) method is presented for the simultaneous determination of midodrine and desglymidodrine (DMAE) in Czapek-Dox culture medium, to be used in biotransformation studies by fungi. The HPLC analysis was conducted using a Lichrospher 100 RP18 column, acetonitrile-40 mmol/L formic acid solution (60:40, v/v) as mobile phase, and ultraviolet detection at 290 nm. The sample preparation was conducted by liquid-liquid extraction using ethyl acetate as extractor solvent. The method was linear over the concentration range of 0.4-40.0 µg/mL for midodrine (r ≥ 0.9997) and DMAE (r ≥ 0.9998). Within-day and between-day precision and accuracy were evaluated by relative standard deviations (≤ 8.2%) and relative errors (-7.3 to 7.4%), respectively. The validated method was used to assess midodrine biotransformation by the fungi Papulaspora immersa Hotson SS13, Botrytis cinerea UCA 992 and Botrytis cinerea 2100 under static and shaken conditions. Under shaken conditions, the biotransformation of midodrine to DMAE was more efficient for all studied fungi, especially for the fungus Botrytis cinerea 2100, which converted 42.2% of midodrine to DMAE.

  5. An Exploratory Energy Analysis of Electrochromic Windows in Small and Medium Office Buildings - Simulated Results Using EnergyPlus

    Energy Technology Data Exchange (ETDEWEB)

    Belzer, David B.

    2010-08-01

    The Department of Energy’s (DOE) Building Technologies Program (BTP) has had an active research program in supporting the development of electrochromic (EC) windows. Electrochromic glazings used in these windows have the capability of varying the transmittance of light and heat in response to an applied voltage. This dynamic property allows these windows to reduce lighting, cooling, and heating energy in buildings where they are employed. The exploratory analysis described in this report examined three different variants of EC glazings, characterized by the amount of visible light and solar heat gain (as measured by the solar heat gain coefficients [SHGC] in their “clear” or transparent states). For these EC glazings, the dynamic range of the SHGC’s between their “dark” (or tinted) state and the clear state were: (0.22 - 0.70, termed “high” SHGC); (0.16 - 0.39, termed “low” SHGC); and (0.13 - 0.19; termed “very low” SHGC). These glazings are compared to conventional (static) glazing that meets the ASHRAE Standard 90.1-2004 energy standard for five different locations in the U.S. All analysis used the EnergyPlus building energy simulation program for modeling EC windows and alternative control strategies. The simulations were conducted for a small and a medium office building, where engineering specifications were taken from the set of Commercial Building Benchmark building models developed by BTP. On the basis of these simulations, total source-level savings in these buildings were estimated to range between 2 to 7%, depending on the amount of window area and building location.

  6. Spent nuclear fuel transport problems

    International Nuclear Information System (INIS)

    Kondrat'ev, A.N.; Kosarev, Yu.A.; Yulikov, E.I.

    1977-01-01

    The paper considers the problems of shipping spent fuel from nuclear power stations to reprocessing plants and also the principal ways of solving these problems with a view to achieving maximum economy and safety in transport. The increase in the number of nuclear power plants in the USSR will entail an intensification of spent-fuel shipments. Higher burnup and the need to reduce cooling time call for heavier and more complex shipping containers. The problem of shipping spent fuel should be tackled comprehensively, bearing in mind the requirements of safety and economy. One solution to these problems is to develop rational and cheap designs of such containers. In addition, the world-wide trend towards more thorough protection of the environment against pollution and of the health of the population requires the devotion of constant attention to improving the reliability and safety of shipments. The paper considers the prospects for nuclear power development in the USSR and in other member countries of the CMEA (1976-1980), the composition and design of some Soviet packaging assemblies, the appropriate cooling time for spent fuel from thermal reactor power stations, procedures for reducing fuel-shipping costs, some methodological problems of container calculation and design, and finally problems of testing and checking containers on test rigs. (author)

  7. Characteristics of spent nuclear fuel

    International Nuclear Information System (INIS)

    Notz, K.J.

    1988-04-01

    The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the spent fuels and other wastes that will, or may, eventually be disposed of in a geological repository. The two major sources of these materials are commercial light-water reactor (LWR) spent fuel and immobilized high-level waste (HLW). Other wastes that may require long-term isolation include non-LWR spent fuels and miscellaneous sources such as activated metals. This report deals with spent fuels, but for completeness, the other sources are described briefly. Detailed characterizations are required for all of these potential repository wastes. These characteristics include physical, chemical, and radiological properties. The latter must take into account decay as a function of time. In addition, the present inventories and projected quantities of the various wastes are needed. This information has been assembled in a Characteristics Data Base which provides data in four formats: hard copy standard reports, menu-driven personal computer (PC) data bases, program-level PC data bases, and mainframe computer files. 5 refs., 3 figs., 4 tabs

  8. Worldwide spent fuel transportation logistics

    International Nuclear Information System (INIS)

    Best, R.E.; Garrison, R.F.

    1978-01-01

    This paper presents an overview of the worldwide transportation requirements for spent fuel. Included are estimates of numbers and types of shipments by mode and cask type for 1985 and the year 2000. In addition, projected capital and transportation costs are presented. For the year 1977 and prior years inclusive, there is a cumulative worldwide requirement for approximately 300 MTU of spent fuel storage at away-from-reactor (AFR) facilities. The cumulative requirements for years through 1985 are projected to be nearly 10,000 MTU, and for the years through 2000 the requirements are conservatively expected to exceed 60,000 MTU. These AFR requirements may be related directly to spent fuel transportation requirements. In total nearly 77,000 total cask shipments of spent fuel will be required between 1977 and 2000. These shipments will include truck, rail, and intermodal moves with many ocean and coastal water shipments. A limited number of shipments by air may also occur. The US fraction of these is expected to include 39,000 truck shipments and 14,000 rail shipments. European shipments to regional facilities are expected to be primarily by rail or water mode and are projected to account for 16,000 moves. Pacific basin shipments will account for 4500 moves. The remaining are from other regions. Over 400 casks will be needed to meet the transportation demands. Capital investment is expected to reach $800,000,000 in 1977 dollars. Cumulative transport costs will be a staggering $4.4 billion dollars

  9. Analysis of reasons for decline of bioleaching efficiency of spent Zn-Mn batteries at high pulp densities and exploration measure for improving performance.

    Science.gov (United States)

    Xin, Baoping; Jiang, Wenfeng; Li, Xin; Zhang, Kai; Liu, Changhao; Wang, Renqing; Wang, Yutao

    2012-05-01

    The reasons for decline of bioleaching efficiency of Zn and Mn from spent batteries at high pulp densities were analyzed; the measures for improving bioleaching efficiency were investigated. The results showed that extraction efficiency of Zn dropped from 100% at 1% of pulp density to 29.9% at 8% of pulp density, with Mn from 94% to only 2.5%. It was almost the linear reduction of the activity of the sulfur-oxidizing bacteria with increase of pulp density that witnessed declined bioleaching efficiency of Zn; it was the complete inactivation of the iron-oxidizing bacteria at 2% of pulp density or higher that witnessed declined bioleaching dose of Mn. By means of reducing initial pH value of leaching media, increasing concentration of energy matters and exogenous acid adjustment of media during bioleaching, the maximum extraction efficiency of almost 100% for Zn and 89% for Mn at 4% of pulp density was attained, respectively. Copyright © 2012 Elsevier Ltd. All rights reserved.

  10. ATR Spent Fuel Options Study

    International Nuclear Information System (INIS)

    Connolly, Michael James; Bean, Thomas E.; Brower, Jeffrey O.; Luke, Dale E.; Patterson, M. W.; Robb, Alan K.; Sindelar, Robert; Smith, Rebecca E.; Tonc, Vincent F.; Tripp, Julia L.; Winston, Philip L.

    2017-01-01

    The Advanced Test Reactor (ATR) is a materials and fuels test nuclear reactor that performs irradiation services for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Naval Reactors, the National Nuclear Security Administration (NNSA), and other research programs. ATR achieved initial criticality in 1967 and is expected to operate in support of needed missions until the year 2050 or beyond. It is anticipated that ATR will generate approximately 105 spent nuclear fuel (SNF) elements per year through the year 2050. Idaho National Laboratory (INL) currently stores 2,008 ATR SNF elements in dry storage, 976 in wet storage, and expects to have 1,000 elements in wet storage before January 2017. A capability gap exists at INL for long-term (greater than the year 2050) management, in compliance with the Idaho Settlement Agreement (ISA), of ATR SNF until a monitored retrievable geological repository is open. INL has significant wet and dry storage capabilities that are owned by the DOE Office of Environmental Management (EM) and operated and managed by Fluor Idaho, which include the Idaho Nuclear Technology and Engineering Center's (INTEC's) CPP-666, CPP-749, and CPP-603. In addition, INL has other capabilities owned by DOE-NE and operated and managed by Battelle Energy Alliance, LLC (BEA), which are located at the Materials and Fuel Complex (MFC). Additional storage capabilities are located on the INL Site at the Naval Reactors Facility (NRF). Current INL SNF management planning, as defined in the Fluor Idaho contract, shows INTEC dry fuel storage, which is currently used for ATR SNF, will be nearly full after transfer of an additional 1,000 ATR SNF from wet storage. DOE-NE tasked BEA with identifying and analyzing options that have the potential to fulfill this capability gap. BEA assembled a team comprised of SNF management experts from Fluor Idaho, Savannah River Site (SRS), INL/BEA, and the MITRE Corp with an objective of developing and

  11. ATR Spent Fuel Options Study

    Energy Technology Data Exchange (ETDEWEB)

    Connolly, Michael James [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bean, Thomas E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Brower, Jeffrey O. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Luke, Dale E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Patterson, M. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robb, Alan K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sindelar, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Rebecca E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tonc, Vincent F. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tripp, Julia L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    The Advanced Test Reactor (ATR) is a materials and fuels test nuclear reactor that performs irradiation services for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Naval Reactors, the National Nuclear Security Administration (NNSA), and other research programs. ATR achieved initial criticality in 1967 and is expected to operate in support of needed missions until the year 2050 or beyond. It is anticipated that ATR will generate approximately 105 spent nuclear fuel (SNF) elements per year through the year 2050. Idaho National Laboratory (INL) currently stores 2,008 ATR SNF elements in dry storage, 976 in wet storage, and expects to have 1,000 elements in wet storage before January 2017. A capability gap exists at INL for long-term (greater than the year 2050) management, in compliance with the Idaho Settlement Agreement (ISA), of ATR SNF until a monitored retrievable geological repository is open. INL has significant wet and dry storage capabilities that are owned by the DOE Office of Environmental Management (EM) and operated and managed by Fluor Idaho, which include the Idaho Nuclear Technology and Engineering Center’s (INTEC’s) CPP-666, CPP-749, and CPP-603. In addition, INL has other capabilities owned by DOE-NE and operated and managed by Battelle Energy Alliance, LLC (BEA), which are located at the Materials and Fuel Complex (MFC). Additional storage capabilities are located on the INL Site at the Naval Reactors Facility (NRF). Current INL SNF management planning, as defined in the Fluor Idaho contract, shows INTEC dry fuel storage, which is currently used for ATR SNF, will be nearly full after transfer of an additional 1,000 ATR SNF from wet storage. DOE-NE tasked BEA with identifying and analyzing options that have the potential to fulfill this capability gap. BEA assembled a team comprised of SNF management experts from Fluor Idaho, Savannah River Site (SRS), INL/BEA, and the MITRE Corp with an objective of developing and analyzing

  12. Collective Beer Brand Identity: A Semiotic Analysis of the Websites Representing Small and Medium Enterprises in the Brewing Industry of Western PA

    Science.gov (United States)

    Cincotta, Dominic

    2014-01-01

    This research studies how brand identities, individually and communally, as read through websites are created among small and medium-sized enterprise breweries in western Pennsylvania. Content analysis through the frame of Kress and van Leeuwen was used as the basis for the codebook that reads each brand identity for the researcher. The…

  13. Review and analysis of potential safety impacts of and regulatory barriers to fuel efficiency technologies and alternative fuels in medium- and heavy-duty vehicles

    Science.gov (United States)

    2015-06-01

    This report summarizes a safety analysis of medium- and heavy-duty vehicles (MD/HDVs) equipped with fuel efficiency (FE) technologies and/or using alternative fuels (natural gas-CNG and LNG, propane, biodiesel and power train electrification). The st...

  14. System architecture and operational analysis of medium displacement unmanned surface vehicle sea hunter as a surface warfare component of distributed lethality

    Science.gov (United States)

    2017-06-01

    NAVAL POSTGRADUATE SCHOOL MONTEREY, CALIFORNIA THESIS Approved for public release. Distribution is unlimited. SYSTEM ARCHITECTURE ...TITLE AND SUBTITLE SYSTEM ARCHITECTURE AND OPERATIONAL ANALYSIS OF MEDIUM DISPLACEMENT UNMANNED SURFACE VEHICLE SEA HUNTER AS A SURFACE WARFARE...traceability, requirements and capabilities while determining the architecture framework in accordance with the Department of Defense Architectural

  15. Radioactive Source Specification of Bushehr’s VVER-1000 Spent Fuels

    OpenAIRE

    Rezaeian, Mahdi; Kamali, Jamshid

    2016-01-01

    Due to high radioactivity and significant content of medium- and long-lived radionuclides, different operations with spent nuclear fuels (e.g., handling, transportation, and storage) shall be accompanied by suitable radiation protections. On the other hand, determination of radioactive source specification is the initial step for any radiation protection design. In this study, radioactive source specification of the spent fuels of Bushehr nuclear power plant, which is a VVER-1000 type pressur...

  16. Flow, Transport, and Reaction in Porous Media: Percolation Scaling, Critical-Path Analysis, and Effective Medium Approximation

    Science.gov (United States)

    Hunt, Allen G.; Sahimi, Muhammad

    2017-12-01

    We describe the most important developments in the application of three theoretical tools to modeling of the morphology of porous media and flow and transport processes in them. One tool is percolation theory. Although it was over 40 years ago that the possibility of using percolation theory to describe flow and transport processes in porous media was first raised, new models and concepts, as well as new variants of the original percolation model are still being developed for various applications to flow phenomena in porous media. The other two approaches, closely related to percolation theory, are the critical-path analysis, which is applicable when porous media are highly heterogeneous, and the effective medium approximation—poor man's percolation—that provide a simple and, under certain conditions, quantitatively correct description of transport in porous media in which percolation-type disorder is relevant. Applications to topics in geosciences include predictions of the hydraulic conductivity and air permeability, solute and gas diffusion that are particularly important in ecohydrological applications and land-surface interactions, and multiphase flow in porous media, as well as non-Gaussian solute transport, and flow morphologies associated with imbibition into unsaturated fractures. We describe new applications of percolation theory of solute transport to chemical weathering and soil formation, geomorphology, and elemental cycling through the terrestrial Earth surface. Wherever quantitatively accurate predictions of such quantities are relevant, so are the techniques presented here. Whenever possible, the theoretical predictions are compared with the relevant experimental data. In practically all the cases, the agreement between the theoretical predictions and the data is excellent. Also discussed are possible future directions in the application of such concepts to many other phenomena in geosciences.

  17. Characterizing nuclear and mitochondrial DNA in spent embryo culture media: genetic contamination identified.

    Science.gov (United States)

    Hammond, Elizabeth R; McGillivray, Brent C; Wicker, Sophie M; Peek, John C; Shelling, Andrew N; Stone, Peter; Chamley, Larry W; Cree, Lynsey M

    2017-01-01

    To characterize nuclear and mitochondrial DNA (mtDNA) in spent culture media from normally developing blastocysts to determine whether it could be used for noninvasive genetic assessment. Prospective embryo cohort study. Academic center and private in vitro fertilization (IVF) clinic. Seventy patients undergoing intracytoplasmic sperm injection (ICSI) and 227 blastocysts. Culture media assessment, artificial blastocoele fluid collapse and DNA analysis using digital polymerase chain reaction (dPCR), long-range PCR, quantitative PCR (qPCR), and DNA fingerprinting. Presence of nuclear and mtDNA in three different commercial culture media from Vitrolife and Irvine Scientific, spent embryo media assessment at the cleavage and blastocyst stages of development, and analysis of the internal media controls for each patient that had been exposed to identical conditions as embryo media but did not come into contact with embryos. Higher levels of nuclear and mtDNA were observed in the culture media that had been exposed to embryos compared with the internal media controls. Nuclear DNA (∼4 copies) and mtDNA (∼600 copies) could be detected in spent media, and the levels increased at the blastocyst stage. No increase in DNA was detected after artificial blastocoele fluid collapse. Mixed sex chromosome DNA was detected. This originated from contamination in the culture media and from maternal (cumulus) cells. Due to the limited amount of template, the presence of embryonic nuclear DNA could not be confirmed by DNA fingerprinting analysis. Currently DNA from culture media cannot be used for genetic assessment because embryo-associated structures release DNA into the culture medium and the DNA is of mixed origin. Copyright © 2016 American Society for Reproductive Medicine. Published by Elsevier Inc. All rights reserved.

  18. Proteome analysis of Aspergillus niger: Lactate added in starch-containing medium can increase production of the mycotoxin fumonisin B2 by modifying acetyl-CoA metabolism

    DEFF Research Database (Denmark)

    Sørensen, Louise Marie; Lametsch, Rene; Andersen, Mikael Rørdam

    2009-01-01

    and regulation of this metabolite. Proteome analysis was used with the purpose to reveal how fumonisin B2 production by A. niger is influenced by starch and lactate in the medium. Results Fumonisin B2 production by A. niger was significantly increased when lactate and starch were combined in the medium....... Production of a few other A. niger secondary metabolites was affected similarly by lactate and starch (fumonisin B4, orlandin, desmethylkotanin and pyranonigrin A), while production of others was not (ochratoxin A, ochratoxin alpha, malformin A, malformin C, kotanin, aurasperone B and tensidol B......). The proteome of A. niger was clearly different during growth on media containing 3% starch, 3% starch + 3% lactate or 3% lactate. The identity of 59 spots was obtained, mainly those showing higher or lower expression levels on medium with starch and lactate. Many of them were enzymes in primary metabolism...

  19. Optimization of time and location dependent spent nuclear fuel storage capacity

    International Nuclear Information System (INIS)

    Macek, V.

    1977-01-01

    A linear spent fuel storage model is developed to identify cost-effective spent nuclear fuel storage strategies. The purpose of this model is to provide guidelines for the implementation of the optimal time-dependent spent fuel storage capacity expansion in view of the current economic and regulatory environment which has resulted in phase-out of the closed nuclear fuel cycle. Management alternatives of the spent fuel storage backlog, which is created by mismatch between spent fuel generation rate and spent fuel disposition capability, are represented by aggregate decision variables which describe the time dependent on-reactor-site and off-site spent fuel storage capacity additions, and the amount of spent fuel transferred to off-site storage facilities. Principal constraints of the model assure determination of cost optimal spent fuel storage expansion strategies, while spent fuel storage requirements are met at all times. A detailed physical and economic analysis of the essential components of the spent fuel storage problem, which precedes the model development, assures its realism. The effects of technological limitations on the on-site spent fuel storage expansion and timing of reinitiation of the spent fuel reprocessing on optimal spent fuel storage capacity expansion are investigated. The principal results of the study indicate that (a) expansion of storage capacity beyond that of currently planned facilities is necessary, and (b) economics of the post-reactor fuel cycle is extremely sensitive to the timing of reinitiation of spent fuel reprocessing. Postponement of reprocessing beyond mid-1982 may result in net negative economic liability of the back end of the nuclear fuel cycle

  20. Analysis on the entrance surface dose and contrast medium dose at computed tomography and angiography in cardiovascular examination

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Young Hyun [Dept. of Cardiovascular Center, Yeocheon Jeonnam Hospital, Yeosu (Korea, Republic of); Han, Jae Bok; Choi, Nam Gil; Song, Jong Nam [Dept. of Radiological Science, Dongshin University, Naju (Korea, Republic of)

    2016-12-15

    This study aimed to identify dose reduction measures by retrospectively analyzing the entrance surface dose at computed tomography and angiography in cardiovascular examination and to contribute the patients with renal impairmend and a high probability of side effects to determine the inspection's direction by measuring the contrast usages actually to active actions for the dose by actually measuring the contrast medium dose. The CTDIvol value and air kerma value, which are the entrance surface doses of the two examinations, and the contrast medium dose depending on the number of slides were compared and analyzed. This study was conducted in 21 subjects (11 males; 10 females) who underwent Cardiac Computed Tomographic Angiography (CCTA) and Coronary Angiography (CAG) in this hospital during the period from May 2014 to May 2016. The subject's age was 48-85 years old (mean 65±10 years old), and the weight was 37.6~83.3 kg (mean 63±6 kg). Dose reduction could be expected in the cardiovascular examination using CCTA rather than in the examination using CAG. In terms of contrast medium dose, CAG used a smaller dose than CCTA. In particular, as the number of slides increases at CAG, the contrast medium dose increases. Therefore, in order to reduce the contrast medium dose, the number of slides suitable for the scan range must be selected.

  1. Antifeedant activity of xanthohumol and supercritical carbon dioxide extract of spent hops against stored product pests.

    Science.gov (United States)

    Jackowski, J; Hurej, M; Rój, E; Popłoński, J; Kośny, L; Huszcza, E

    2015-08-01

    Xanthohumol, a prenylated flavonoid from hops, and a supercritical carbon dioxide extract of spent hops were studied for their antifeedant activity against stored product insect pests: Sitophilus granarius L., Tribolium confusum Duv. and Trogoderma granarium Everts. Xanthohumol exhibited medium deterrent activity against the adults of S. granarius L. and larvae of T. confusum Duv. The spent hops extract was more active than xanthohumol towards the adults of T. confusum Duv. The potential application of the crude spent hops extract as a feeding deterrent against the stored product pests is proposed.

  2. Metals removal from spent salts

    Science.gov (United States)

    Hsu, Peter C.; Von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Brummond, William A.; Adamson, Martyn G.

    2002-01-01

    A method and apparatus for removing metal contaminants from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents may be added to precipitate the metal oxide and/or the metal as either metal oxide, metal hydroxide, or as a salt. The precipitated materials are filtered, dried and packaged for disposal as waste or can be immobilized as ceramic pellets. More than about 90% of the metals and mineral residues (ashes) present are removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be spray-dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration ion exchange column, which yields salt solutions that contain less than 1.0 ppm of contaminants.

  3. Actinide removal from spent salts

    Science.gov (United States)

    Hsu, Peter C.; von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Adamson, Martyn G.

    2002-01-01

    A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration salt solutions that contain less than 0.1 ppm of thorium or uranium.

  4. TMI-2 spent fuel shipping

    International Nuclear Information System (INIS)

    Quinn, G.J.; Burton, H.M.

    1985-01-01

    TMI-2 failed fuel will be shipped to the Idaho National Engineering Laboratory for use in the DOE Core Examination Program. The fuel debris will be loaded into three types of canisters during defueling and dry loaded into a spent fuel shipping cask. The cask design accommodates seven canisters per cask and has two separate containment vessels with ''leaktight'' seals. Shipments are expectd to begin in early 1986

  5. Spent nuclear fuel in Bulgaria

    International Nuclear Information System (INIS)

    Peev, P.; Kalimanov, N.

    1999-01-01

    The development of the nuclear energy sector in Bulgaria is characterized by two major stages. The first stage consisted of providing a scientific basis for the programme for development of the nuclear energy sector in the country and was completed with the construction of an experimental water-water reactor. At present, spent nuclear fuel from this reactor is placed in a water filled storage facility and will be transported back to Russia. The second stage consisted of the construction of the 6 NPP units at the Kozloduy site. The spent nuclear fuel from the six units is stored in at reactor pools and in an additional on-site storage facility which is nearly full. In order to engage the government of the country with the on-site storage problems, the new management of the National Electric Company elaborated a policy on nuclear fuel cycle and radioactive waste management. The underlying policy is de facto the selection of the 'deferred decision' option for its spent fuel management. (author)

  6. Reuse of Hydrotreating Spent Catalyst

    International Nuclear Information System (INIS)

    Habib, A.M.; Menoufy, M.F.; Amhed, S.H.

    2004-01-01

    All hydro treating catalysts used in petroleum refining processes gradually lose activity through coking, poisoning by metal, sulfur or halides or lose surface area from sintering at high process temperatures. Waste hydrotreating catalyst, which have been used in re-refining of waste lube oil at Alexandria Petroleum Company (after 5 years lifetime) compared with the same fresh catalyst were used in the present work. Studies are conducted on partial extraction of the active metals of spent catalyst (Mo and Ni) using three leaching solvents,4% oxidized oxalic acid, 10% aqueous sodium hydroxide and 10% citric acid. The leaching experiments are conducting on the de coked extrude [un crushed] spent catalyst samples. These steps are carried out in order to rejuvenate the spent catalyst to be reused in other reactions. The results indicated that 4% oxidized oxalic acid leaching solution gave total metal removal 45.6 for de coked catalyst samples while NaOH gave 35% and citric acid gave 31.9 % The oxidized leaching agent was the most efficient leaching solvent to facilitate the metal removal, and the rejuvenated catalyst was characterized by the unchanged crystalline phase The rejuvenated catalyst was applied for hydrodesulfurization (HDS) of vacuum gas oil as a feedstock, under different hydrogen pressure 20-80 bar in order to compare its HDS activity

  7. Spent Fuel Working Group Report

    International Nuclear Information System (INIS)

    O'Toole, T.

    1993-11-01

    The Department of Energy is storing large amounts of spent nuclear fuel and other reactor irradiated nuclear materials (herein referred to as RINM). In the past, the Department reprocessed RINM to recover plutonium, tritium, and other isotopes. However, the Department has ceased or is phasing out reprocessing operations. As a consequence, Department facilities designed, constructed, and operated to store RINM for relatively short periods of time now store RINM, pending decisions on the disposition of these materials. The extended use of the facilities, combined with their known degradation and that of their stored materials, has led to uncertainties about safety. To ensure that extended storage is safe (i.e., that protection exists for workers, the public, and the environment), the conditions of these storage facilities had to be assessed. The compelling need for such an assessment led to the Secretary's initiative on spent fuel, which is the subject of this report. This report comprises three volumes: Volume I; Summary Results of the Spent Fuel Working Group Evaluation; Volume II, Working Group Assessment Team Reports and Protocol; Volume III; Operating Contractor Site Team Reports. This volume presents the overall results of the Working Group's Evaluation. The group assessed 66 facilities spread across 11 sites. It identified: (1) facilities that should be considered for priority attention. (2) programmatic issues to be considered in decision making about interim storage plans and (3) specific vulnerabilities for some of these facilities

  8. Advanced three-dimensional thermal modeling of a baseline spent fuel repository

    International Nuclear Information System (INIS)

    Altenbach, T.J.; Lowry, W.E.

    1980-01-01

    A three-dimensional thermal analysis using finite difference techniques was performed to determine the near-field response of a baseline spent fuel repository in a deep geologic salt medium. A baseline design incorporates previous thermal modeling experience and OWI recommendations for areal thermal loading in specifying the waste form properties, package details, and emplacement configuration. The base case in this thermal analysis considers one 10-year old PWR spent fuel assembly emplaced to yield a 36 kW/acre (8.9 W/m 2 ) loading. A unit cell model in an infinite array is used to simplify the problem and provide upper-bound temperatures. Boundary conditions are imposed which allow simulations to 1000 years. Variations studied include a comparison of ventilated and unventilated storage room conditions, emplacement packages with and without air gaps surrounding the canister, and room cool-down scenarios with ventilation following an unventilated state for retrieval purposes. It was found that at this low-power level, ventilating the emplacement room has an immediate cooling influence on the canister and effectively maintains the emplacement room floor near the temperature of the ventilating air

  9. A global sensitivity analysis of two-phase flow between fractured crystalline rock and bentonite with application to spent nuclear fuel disposal.

    Science.gov (United States)

    Dessirier, Benoît; Frampton, Andrew; Jarsjö, Jerker

    2015-11-01

    Geological disposal of spent nuclear fuel in deep crystalline rock is investigated as a possible long term solution in Sweden and Finland. The fuel rods would be cased in copper canisters and deposited in vertical holes in the floor of deep underground tunnels, embedded within an engineered bentonite buffer. Recent experiments at the Äspö Hard Rock Laboratory (Sweden) showed that the high suction of unsaturated bentonite causes a de-saturation of the adjacent rock at the time of installation, which was also independently predicted in model experiments. Remaining air can affect the flow patterns and alter bio-geochemical conditions, influencing for instance the transport of radionuclides in the case of canister failure. However, thus far, observations and model realizations are limited in number and do not capture the conceivable range and combination of parameter values and boundary conditions that are relevant for the thousands of deposition holes envisioned in an operational final repository. In order to decrease this knowledge gap, we introduce here a formalized, systematic and fully integrated approach to study the combined impact of multiple factors on air saturation and dissolution predictions, investigating the impact of variability in parameter values, geometry and boundary conditions on bentonite buffer saturation times and on occurrences of rock de-saturation. Results showed that four parameters consistently appear in the top six influential factors for all considered output (target) variables: the position of the fracture intersecting the deposition hole, the background rock permeability, the suction representing the relative humidity in the open tunnel and the far field pressure value. The combined influence of these compared to the other parameters increases as one targets a larger fraction of the buffer reaching near-saturation. Strong interaction effects were found, which means that some parameter combinations yielded results (e.g., time to

  10. Nevada Nuclear Waste Storage Investigations Project: thermal analysis of spent fuel disposal in vertical emplacement boreholes in a welded tuff repository

    International Nuclear Information System (INIS)

    St John, C.M.

    1985-11-01

    Two- and three-dimensional heat transfer analyses were conducted to determine temperatures in the vicinity of a waste canister and an emplacement drift. The effect of emplacement of canisters containing spent fuel in vertical boreholes was simulated for the cases of an emplacement drift either fully ventilated or sealed immediately after canister emplacement. PORFLOW and THERM3D respectively solve the two- and three-dimensional forms of the diffusion equation. In the unventilated case, the effect of radiation was approximated by defining an equivalent radiation thermal conductivity. A simple code, TEMP3D, based on the closed form solutions for constant and decaying heat sources, was also used. Calculations indicate that the temperature at the canister borehole wall will peak at about 215 0 C if the drift is ventilated and about 240 0 C if it is unventilated. The peak temperature occurs sooner in the ventilated case; after 3 to 4 yr versus 9 yr. For a point 1 m from the wall of the emplacement borehole, the corresponding peak temperatures are 150 0 C for the ventilated case and 185 0 C for the unventilated case and occur at about 5 and 17 yr. We assumed that the effect of drift ventilation would be to maintain a uniform temperature of 30 0 C at the drift perimeter. If the drift is unventilated the wall rock temperature peaks some 75 to 100 yr after waste emplacement; reaching about 125 0 C at the mid-height of the drift wall. Comparisons between the results of the three-dimensional analyses performed using TEMP3D and THERM3D indicated that the simpler modeling technique provided a good estimate of temperatures in the immediate vicinity of the canister for both the ventilated and unventilated cases. Comparisons of the results of two- and three-dimensional analyses performed using the PORFLOW and THERM3D codes indicated that the two-dimensional approximation is excellent, except in the immediate vicinity of the canister

  11. Hydrogeological approach to the regional analysis of low flow in medium and small streams of the hilly and mountainous areas of Serbia

    Directory of Open Access Journals (Sweden)

    Nikić Zoran

    2006-01-01

    Full Text Available During the long rainless spells of the dry season, flows in medium and small streams get reduced to what is generally known as "low flow". For ungauged streams, the controlling "low flows" are determined using the regional analysis method. In the presently described exploration, the method applied was based on the assumption that dry weather discharges in medium and small rivers depended on the hydrogeological conditions. The controlling effect of hydrogeology on the natural low flow in medium and small streams of the hilly and mountainous part of Serbia was analyzed applying the theory of multiple linear regression. The thirty-day minimum mean 80 and 95 per cent exceedance flows were taken for dependent variables, and quantified hydrogeological elements as independent variables. The analysis covered streams that had small or medium size catchment areas. The treated example encompassed sixty-one gauged catchments. The resulting regional relations for the thirty day minimum mean 80 and 95 per cent exceedance flows are presented in this paper. The quality of the established relation was controlled by relevant statistic tests.

  12. Predicting In vitro Culture Medium Macro-Nutrients Composition for Pear Rootstocks Using Regression Analysis and Neural Network Models.

    Science.gov (United States)

    Jamshidi, S; Yadollahi, A; Ahmadi, H; Arab, M M; Eftekhari, M

    2016-01-01

    Two modeling techniques [artificial neural network-genetic algorithm (ANN-GA) and stepwise regression analysis] were used to predict the effect of medium macro-nutrients on in vitro performance of pear rootstocks (OHF and Pyrodwarf). The ANN-GA described associations between investigating eight macronutrients (NO[Formula: see text], NH[Formula: see text], Ca(2+), K(+), Mg(2+), PO[Formula: see text], SO[Formula: see text], and Cl(-)) and explant growth parameters [proliferation rate (PR), shoot length (SL), shoot tip necrosis (STN), chlorosis (Chl), and vitrification (Vitri)]. ANN-GA revealed a substantially higher accuracy of prediction than for regression models. According to the ANN-GA results, among the input variables concentrations (mM), NH[Formula: see text] (301.7), and NO[Formula: see text], NH[Formula: see text] (64), SO[Formula: see text] (54.1), K(+) (40.4), and NO[Formula: see text] (35.1) in OHF and Ca(2+) (23.7), NH[Formula: see text] (10.7), NO[Formula: see text] (9.1), NH[Formula: see text] (317.6), and NH[Formula: see text] (79.6) in Pyrodwarf had the highest values of VSR in data set, respectively, for PR, SL, STN, Chl, and Vitri. The ANN-GA showed that media containing (mM) 62.5 NO[Formula: see text], 5.7 NH[Formula: see text], 2.7 Ca(2+), 31.5 K(+), 3.3 Mg(2+), 2.6 PO[Formula: see text], 5.6 SO[Formula: see text], and 3.5 Cl(-) could lead to optimal PR for OHF and optimal PR for Pyrodwarf may be obtained with media containing 25.6 NO[Formula: see text], 13.1 NH[Formula: see text], 5.5 Ca(2+), 35.7 K(+), 1.5 Mg(2+), 2.1 PO[Formula: see text], 3.6 SO[Formula: see text], and 3 Cl(-).

  13. Spent mushroom substrate biochar as a potential amendment in pig manure and rice straw composting processes.

    Science.gov (United States)

    Chang, Ken-Lin; Chen, Xi-Mei; Sun, Jian; Liu, Jing-Yong; Sun, Shui-Yu; Yang, Zuo-Yi; Wang, Yin

    2017-07-01

    Spent mushroom substrate (SMS) is a bulky waste byproduct of commercial mushroom production, which can cause serious environmental problems and, therefore, poses a significant barrier to future expansion of the mushroom industry. In the present study, we explored the use of SMS as a biochar to improve the quality of bio-fertilizer. Specifically, we performed a series of experiments using composting reactors to investigate the effects of SMS biochar on the physio-chemical properties of bio-fertilizer. Biochar was derived from dry SMS pyrolysed at 500°C and mixed with pig manure and rice straw. Results from this study demonstrate that the addition of biochar significantly reduced electrical conductivity and loss of organic matter in compost material. Nutrient analysis revealed that the SMS-derived biochar is rich in fertilizer nutrients such as P, K, Na, and N. All of these findings suggest that SMS biochar could be an excellent medium for compost.

  14. Near surface spent fuel storage: environmental issues

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  15. Overview of spent fuel management and problems

    International Nuclear Information System (INIS)

    Ritchie, I.G.; Ernst, P.C.

    1998-01-01

    Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel worldwide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialized and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. Some projections of spent fuel inventories to the year 2006 are presented and discussed. (author)

  16. Considerations for a national program on spent fuel management

    International Nuclear Information System (INIS)

    Lopez-Perez, B.; Melches-Serrano, C.

    1980-01-01

    The spent fuel discharged from the two LWR's that are in operation (Zorita, 160 MW PWR, and Santa Maria de Garona, 460 MW BWR) is being reprocessed under contracts with BNFL; these contracts will expire in the next few years. The fuel discharged from Vandelos (50 MW GCR) is being reprocessed by Cogema under a long-term contract. No new reprocessing contracts for LWR's in operation, under construction, or planned have been signed or are being considered for the near future. The plutonium and the residual uranium contained in LWR spent fuel are considered important potential energy resources. They are especially valuable for countries such as Spain, which is short of energy resources, and they might be used in the future in fast breeder or thermal reactors. This is the reason that, until reprocessing is justified and appropriate solutions to make reprocessing available are developed, Spain has decided to build the appropriate capacity for the temporary storage of spent fuel. The capacity is being achieved, on short term, by the extension of AR storage capacity. It is being achieved, at medium or longer term, by the construction of centralized AFR facilities to serve all Spanish nuclear power plants. Spanish utilities are undertaking the expansion of reactor storage capacities, using densified racks, to increment capacity to at least 8 to 10 reloads, in addition to full core discharge capacity. Spain has the time and the financial and technical resources to implement a national solution for spent fuel storage. Financial strategy, technology choice, and licensing considerations are under examination in order to make a decision for medium- and long-term storage alternatives

  17. 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

  18. Spent fuel management fee methodology and computer code user's manual.

    Energy Technology Data Exchange (ETDEWEB)

    Engel, R.L.; White, M.K.

    1982-01-01

    The methodology and computer model described here were developed to analyze the cash flows for the federal government taking title to and managing spent nuclear fuel. The methodology has been used by the US Department of Energy (DOE) to estimate the spent fuel disposal fee that will provide full cost recovery. Although the methodology was designed to analyze interim storage followed by spent fuel disposal, it could be used to calculate a fee for reprocessing spent fuel and disposing of the waste. The methodology consists of two phases. The first phase estimates government expenditures for spent fuel management. The second phase determines the fees that will result in revenues such that the government attains full cost recovery assuming various revenue collection philosophies. These two phases are discussed in detail in subsequent sections of this report. Each of the two phases constitute a computer module, called SPADE (SPent fuel Analysis and Disposal Economics) and FEAN (FEe ANalysis), respectively.

  19. Evaluation of the perception and application of social responsibility practices in micro, small and medium companies in Barranquilla. An analysis from the theory of Stakeholders

    OpenAIRE

    Guillén León; Hamadys L. Benavides Gutiérrez; José María Castán Farrero

    2017-01-01

    The purpose of this paper is to evaluate the degree of comprehension and enforcement of social responsibility (SR) practices in micro, small and medium companies in Barranquilla (Colombia), based on the Stakeholders theory. Using an exploratory factor analysis on 779 companies it was found that the variables with a stronger explanatory influence for socially responsible performance are employees, environment, and community. By contrast, corporate management, value chain, and government/public...

  20. Electrochemical studies on spent fuel corrosion processes

    International Nuclear Information System (INIS)

    Pablo, J. de; Casas, I.; Clarens, F.; Gimenez, J.; Rovira, M.

    2003-01-01

    This presentations is mainly based on the electrochemical studies carried out by the Canadian team and the research group of the Berlin University. Electrochemical studies allow to study separately both the anodic reaction which corresponds-sources on UO 2 -electrodes response is one of to the UO 2 dissolution and the cathodic reaction that is the reduction of the oxidants. By using intensity current-potential plots a mechanisms of UO 2 corrosion has been established. At-300 mV (vs SCE), irreversible oxidation of UO 2 takes place and dissolution begins. In the absence of complexing agents like carbonate, an oxidised layer is formed at 100 mV a stoichiometry close to UO 2 . In carbonate medium, the oxidized layer is not formed because the U(VI) formed is rapidly dissolved. Results in terms of dissolution rates obtained by electrochemical measurements are similar to the ones obtained in dissolution experiments by using flow through reactors and similar kinetic laws are obtained. The effect of external α and γ-sources on UO 2 -electrodes response is one of the few available data on the effects of radiolysis on the UO 2 dissolution rate and can offer a complementary knowledge to the spent fuel and α-doped pellets dissolution experiments. (Author)

  1. Integrated risk assessment for spent fuel transportation using developed software

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  2. Integrated risk assessment for spent fuel transportation using developed software

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  3. The solidification of spent resin

    International Nuclear Information System (INIS)

    Shiao, S. J.; Tsai, C. M.; Shyu, Y. H.

    1991-01-01

    A quasi-steady apparatus was applied to measure the thermal conductivity of solids ranging in size for 0.3 to 200 L, and temperature distributions in the solids were recorded during the curing, and theoretical equation for conduction in a cylindrical form with uniform energy generation was established to define the thermal state of reaction. The heat of reaction calculated from the theoretical equation with experimental values for the maximum temperature and thermal conductivity agrees very well with the data reported. The relationships among heat of reaction and amount of curing agent, retardant, loading of spent resin, and water were established

  4. Analysis of combined conduction and radiation heat transfer in presence of participating medium by the development of hybrid method

    International Nuclear Information System (INIS)

    Mahapatra, S.K.; Dandapat, B.K.; Sarkar, A.

    2006-01-01

    The current study addresses the mathematical modeling aspects of coupled conductive and radiative heat transfer in the presence of absorbing, emitting and isotropic scattering gray medium within two-dimensional square enclosure. A blended method where the concepts of modified differential approximation employed by combining discrete ordinate method and spherical harmonics method, has been developed for modeling the radiative transport equation. The gray participating medium is bounded by isothermal walls of two-dimensional enclosure which are considered to be opaque, diffuse and gray. The effect of various influencing parameters i.e., radiation-conduction parameter, surface emissivity, single scattering albedo and optical thickness has been illustrated. The adaptability of the present method has also been addressed

  5. Three dimensional vibration and bending analysis of carbon nanotubes embedded in elastic medium based on theory of elasticity

    Directory of Open Access Journals (Sweden)

    M. Shaban

    Full Text Available This paper studies free vibration and bending behavior of singlewalled carbon nanotubes (SWCNTs embedded on elastic medium based on three-dimensional theory of elasticity. To accounting the size effect of carbon nanotubes, non-local theory is adopted to shell model. The nonlocal parameter is incorporated into all constitutive equations in three dimensions. The surrounding medium is modeled as two-parameter elastic foundation. By using Fourier series expansion in axial and circumferential direction, the set of coupled governing equations are reduced to the ordinary differential equations in thickness direction. Then, the state-space method as an efficient and accurate method is used to solve the resulting equations analytically. Comprehensive parametric studies are carried out to show the influences of the nonlocal parameter, radial and shear elastic stiffness, thickness-to-radius ratio and radiusto-length ratio.

  6. Spectral element boundary integral method with periodic layered medium dyadic Green's function for multiscale nano-optical scattering analysis.

    Science.gov (United States)

    Niu, Jun; Ren, Yi; Liu, Qing Huo

    2017-10-02

    In this work, we propose a numerical solver combining the spectral element - boundary integral (SEBI) method with the periodic layered medium dyadic Green's function. The periodic layered medium dyadic Green's function is formulated under matrix representation. The surface integral equations (SIEs) are then implemented as the radiation boundary condition to truncate the top and bottom computation domain. After describing the interior computation domain with the vector wave equations, and treating the lateral boundaries with Bloch periodic boundary conditions, the whole computation domains are discretized with mixed-order Gauss- Lobatto-Legendre basis functions in the SEBI method. This method avoids the discretization of the top and bottom layered media, so it can be much more efficient than conventional methods. Numerical results validate the proposed solver with fast convergence throughout the whole computation domain and good performance for typical multiscale nano-optical applications.

  7. Stated preference analysis for new public transport in a medium-sized asian city: A case study in Malang, Indonesia

    OpenAIRE

    坂東, 徹; Bando, Tetsu; 福田, 大輔; FUKUDA, DAISUKE

    2015-01-01

    This study analyzes citizens' travel choice behavior in a medium-sized Southeast Asian city to observe their intention to use new public transport. We selected Malang in Indonesia as a case study. A travel behavior and intention survey including stated preference questions was conducted with university students. The results of the latent-class model of commute mode choices show that respondents could be divided into “cost and delay time” and “travel and access time” oriented classes. Responde...

  8. Innovation through Coopetition: An analysis of small- and medium-sized trust companies operating in the Liechtenstein financial centre

    Directory of Open Access Journals (Sweden)

    Sascha Kraus

    2018-02-01

    Full Text Available Coopetition has received increasing attention in the academic literature. Prior research has examined the benefits and risks of coopetition as well as its potential impact on innovation in many different contexts, including large companies and manufacturing industries. Surprisingly, despite the omnipresence of small- and- medium-sized enterprises (SMEs and the growing relevance of service industries, coopetition in these contexts has not yet been widely explored. This study seeks to broaden the present understanding of coopetition by finding an answer to the research question “How do small- and medium-sized trust companies apply coopetition in the Liechtenstein trust industry and how can this strategy facilitate innovation?” As such, the presented work investigates the application of coopetition by small- and medium-sized trust companies operating in the Liechtenstein financial centre. The qualitative expert interviews with major actors in the Liechtenstein trust industry reveal that coopetition is a frequently applied business strategy among Liechtenstein trust companies, members of the Liechtenstein financial centre and international competitors. The trustees’ conservative attitude, however, is found to be a typical barrier to coopetition, since it induces trustees to give priority to the protection of their own business. Nevertheless, coopeting partners recognise their ability to derive crucial benefits from their cooperative interactions with rival organisations in terms of possibilities to share resources, costs and know-how. Moreover, coopetition enables coopetitors to innovate their current business models.

  9. Development of finite element code for the analysis of coupled thermo-hydro-mechanical behaviors of saturated-unsaturated medium

    International Nuclear Information System (INIS)

    Ohnishi, Y.; Shibata, H.; Kobayashi, A.

    1985-01-01

    A model is presented which describes fully coupled thermo-hydro-mechanical behavior of porous geologic medium. The mathematical formulation for the model utilizes the Biot theory for the consolidation and the energy balance equation. The medium is in the condition of saturated-unsaturated flow, then the free surfaces are taken into consideration in the model. The model, incorporated in a finite element numerical procedure, was implemented in a two-dimensional computer code. The code was developed under the assumptions that the medium is poro-elastic and in plane strain condition; water in the ground does not change its phase; heat is transferred by conductive and convective flow. Analytical solutions pertaining to consolidation theory for soils and rocks, thermoelasticity for solids and hydrothermal convection theory provided verification of stress and fluid flow couplings, respectively in the coupled model. Several types of problems are analyzed. The one is a study of some of the effects of completely coupled thermo-hydro-mechanical behavior on the response of a saturated-unsaturated porous rock containing a buried heat source. Excavation of an underground opening which has radioactive wastes at elevated temperatures is modeled and analyzed. The results shows that the coupling phenomena can be estimated at some degree by the numerical procedure. The computer code has a powerful ability to analyze of the repository the complex nature of the repository

  10. Production, Purification, and Characterization of a Major Penicillium glabrum Xylanase Using Brewer's Spent Grain as Substrate

    Directory of Open Access Journals (Sweden)

    Adriana Knob

    2013-01-01

    Full Text Available In recent decades, xylanases have been used in many processing industries. This study describes the xylanase production by Penicillium glabrum using brewer's spent grain as substrate. Additionally, this is the first work that reports the purification and characterization of a xylanase using this agroindustrial waste. Optimal production was obtained when P. glabrum was grown in liquid medium in pH 5.5, at 25 °C, under stationary condition for six days. The xylanase from P. glabrum was purified to homogeneity by a rapid and inexpensive procedure, using ammonium sulfate fractionation and molecular exclusion chromatography. SDS-PAGE analysis revealed one band with estimated molecular mass of 18.36 kDa. The optimum activity was observed at 60 °C, in pH 3.0. The enzyme was very stable at 50 °C, and high pH stability was verified from pH 2.5 to 5.0. The ion Mn2+ and the reducing agents β-mercaptoethanol and DTT enhanced xylanase activity, while the ions Hg2+, Zn2+, and Cu2+ as well as the detergent SDS were strong inhibitors of the enzyme. The use of brewer's spent grain as substrate for xylanase production cannot only add value and decrease the amount of this waste but also reduce the xylanase production cost.

  11. Spent Fuel NDA Research Path for the Sweden Encapsulation-Repository

    Energy Technology Data Exchange (ETDEWEB)

    Tobin, Stephen J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Swedish Nuclear Fuel and Waste Management Company (Sweden); Trellue, Holly R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Liljenfeldt, Henrik [Swedish Nuclear Fuel and Waste Management Company (Sweden)

    2015-01-22

    This set of slides provides a description of research performed to date on spent fuel NDA: Next Generation Safeguards Initiative Spent Fuel Project, and NDA analysis and research planned for CLINK. The general purpose is strengthening the technical toolkit of safeguard inspectors. Data mining is being applied to determine the optimal mathematical structure to match the complexity of spent fuel NDA signals and to enable a range of quantities to be estimated.

  12. Fissile Content Assay of Spent Fuel Using LSDS System

    International Nuclear Information System (INIS)

    Jeon, Ju Young; Lee, Yong Deok; Park, Chang Je

    2016-01-01

    About 1.5 % fissile materials still exist in the spent fuel. Therefore, for reutilization of fissile materials in spent fuel at SFR, resource material is produced through the pyro process. Fissile material contents in the resource material must be analyzed before fabricating SFR fuel for reactor safety and economics. The new technology for an isotopic fissile material content assay is under development at KAERI using a lead slowing down spectrometer (LSDS). LSDS is very sensitive to distinguish fission signals from each fissile isotope in spent and recycled fuel. In an assay of fissile content of spent fuel and recycled fuel, an intense radiation background gives limits the direct analysis of fissile materials. However, LSDS is not influenced by such a radiation background in a fissile assay. Based on the decided LSDS geometry set up, a self shielding parameter was calculated at the fuel assay zone by introducing spent fuel or pyro produced nuclear material. When nuclear material is inserted into the assay area, the spent fuel assembly or pyro recycled fuel material perturbs the spatial distribution of slowing down neutrons in lead and the prompt fast fission neutrons produced by fissile materials are also perturbed. The self shielding factor is interpreted as how much of the absorption is created inside the fuel area when it is in the lead. The self shielding effect provides a non-linear property in the isotopic fissile assay. When the self shielding is severe, the assay system becomes more complex and needs a special parameter to treat this non linear effect. Additionally, an assay of isotopic fissile content will contribute to an accuracy improvement of the burn-up code and increase the transparency and credibility for spent fuel storage and usage, as internationally increasing demand. The fissile contents result came out almost exactly with relative error ∼ 2% in case of Pu239, Pu241 for two different plutonium contents. In this study, meaningful results were

  13. Sealed can of spent fuel

    International Nuclear Information System (INIS)

    Suzuki, Yasuyuki.

    1976-01-01

    Object: To provide a seal plug cover with a gripping portion fitted to a canning machine and a gripping portion fitted to a gripper of the same configuration as a fuel body for handling the fuel body so as to facilitate the handling work. Structure: A sealed can comprises a vessel and a seal plug cover, said cover being substantially in the form of a bottomed cylinder, which is slipped on the vessel and air-tightly secured by a fastening bolt between it and a flange. The spent fuel body is received into the vessel together with coolant during the step of canning operation. Said seal plug cover has two gripping portions, one for opening and closing the plug cover of the canning machine as an exclusive use member, the other being in the form of a hook-shaped peripheral groove, whereby the gripping portions may be effectively used using the same gripper when the spent fuel body is transported while being received in the sealed can or when the fuel body is removed from the sealed can. (Kawakami, Y.)

  14. 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)

  15. Proteome analysis of Aspergillus niger: Lactate added in starch-containing medium can increase production of the mycotoxin fumonisin B2 by modifying acetyl-CoA metabolism

    Directory of Open Access Journals (Sweden)

    Andersen Mikael R

    2009-12-01

    Full Text Available Abstract Background Aspergillus niger is a filamentous fungus found in the environment, on foods and feeds and is used as host for production of organic acids, enzymes and proteins. The mycotoxin fumonisin B2 was recently found to be produced by A. niger and hence very little is known about production and regulation of this metabolite. Proteome analysis was used with the purpose to reveal how fumonisin B2 production by A. niger is influenced by starch and lactate in the medium. Results Fumonisin B2 production by A. niger was significantly increased when lactate and starch were combined in the medium. Production of a few other A. niger secondary metabolites was affected similarly by lactate and starch (fumonisin B4, orlandin, desmethylkotanin and pyranonigrin A, while production of others was not (ochratoxin A, ochratoxin alpha, malformin A, malformin C, kotanin, aurasperone B and tensidol B. The proteome of A. niger was clearly different during growth on media containing 3% starch, 3% starch + 3% lactate or 3% lactate. The identity of 59 spots was obtained, mainly those showing higher or lower expression levels on medium with starch and lactate. Many of them were enzymes in primary metabolism and other processes that affect the intracellular level of acetyl-CoA or NADPH. This included enzymes in the pentose phosphate pathway, pyruvate metabolism, the tricarboxylic acid cycle, ammonium assimilation, fatty acid biosynthesis and oxidative stress protection. Conclusions Lactate added in a medium containing nitrate and starch can increase fumonisin B2 production by A. niger as well as production of some other secondary metabolites. Changes in the balance of intracellular metabolites towards a higher level of carbon passing through acetyl-CoA and a high capacity to regenerate NADPH during growth on medium with starch and lactate were found to be the likely cause of this effect. The results lead to the hypothesis that fumonisin production by A. niger

  16. Small-medium sized nuclear coal and gas power plant. A probabilistic analysis of their financial performances and influence of CO2 cost

    International Nuclear Information System (INIS)

    Locatelli, Giorgio; Mancini, Mauro

    2010-01-01

    Nations or regions with limited electrical grid and restricted financial resources are a suitable market for small medium power plants with a size of 300-400 MWe. The literature presents several comparisons about the economics of large power plants (of about 1000 MWe); however there are not probabilistic analysis regarding the economics of small medium power plants. This paper fills this gap comparing, with a Monte Carlo evaluation, the economical and financial performances of a nuclear reactor, a coal fired power plant and a combined cycle gas turbine (CCGT) of 335 MWe. The paper aims also to investigate the effect of the carbon tax and electrical energy price on the economics of these plants. The analysis show as, without any carbon tax, the coal plant has the lowest levelised unit electricity cost (LUEC) and the highest net present value (NPV). Introducing the carbon tax the rank changes: depending on its amount the first and the nuclear after becomes the plant with lower LUEC and highest NPV. Therefore, the uncertainty in the carbon tax cost increases the risk of investing in a coal plant above the level of the new small medium reactor. (author)

  17. Determinants of small and medium sized fast growing enterprises in central and eastern Europe: a panel data analysis

    Directory of Open Access Journals (Sweden)

    Miroslav Mateev

    2010-09-01

    Full Text Available The purpose of this paper is to explore the main determinants of growth in small and medium sized enterprises (SMEs in central and eastern Europe. The important role played by SMEs in the economic development of central and eastern European (CEE countries has attracted the recent attention of academics and policymakers but remains relatively unexplored. Empirical research has suggested that firm growth is determined not only by the traditional characteristics of size and age but also by other firm-specific factors such as indebtedness, internal financing, future growth opportunities, process and product innovation, and organisational changes. Although growth in manufacturing and service SMEs in transition economies is well explained by the traditional firm characteristics of size and age, there is no empirical evidence concerning what other specific factors may be associated with SME growth and performance in these countries. Using a panel dataset of 560 fast growing small and medium enterprises from six transition economies we find that firm size when measured by firm total assets can explain to a large extent the growth in SMEs in these countries. When size is proxied by a firm’s number of employees the observed effect is marginal. Firm specific characteristics such as leverage, current liquidity, future growth opportunities, internally generated funds, and factor productivity are found to be important factors in determining a firm’s growth and performance. Age and ownership do not seem to be able to explain firm growth. The results of our empirical study have also some policy implications: we argue that governments in transition economies need to pay an increased attention to small and medium sized enterprises and try to create a business environment that will be beneficial for SME development.

  18. Analysis of Cattaneo-Christov heat and mass fluxes in the squeezed flow embedded in porous medium with variable mass diffusivity

    Directory of Open Access Journals (Sweden)

    M. Farooq

    Full Text Available This research article investigates the squeezing flow of Newtonian fluid with variable viscosity over a stretchable sheet inserted in Darcy porous medium. Cattaneo-Christov double diffusion models are implemented to scrutinize the characteristics of heat and mass transfer via variable thermal conductivity and variable mass diffusivity. These models are the modification of conventional laws of Fourier’s and Fick’s via thermal and solutal relaxation times respectively. The homotopy analysis Method (HAM is being utilized to provide the solution of highly nonlinear system of coupled partial differential equations after converted into dimensionless governing equations. The behavior of flow parameters on velocity, concentration, and temperature distributions are sketched and analyzed physically. The result indicates that both concentration and temperature distributions decay for higher solutal and thermal relaxation parameters respectively. Keywords: Squeezing flow, Porous medium, Variable viscosity, Cattaneo-Christov heat and mass flux models, Variable thermal conductivity, Variable mass diffusivity

  19. Cross Talk Analysis on Multiple Coupled Transmission Lines; (The calculation of transfer functions on multiple coupled tansmission lines in an inhomogeneous dielectric medium)

    DEFF Research Database (Denmark)

    Dalby, Arne Brejning

    1994-01-01

    A flow graph relating voltages and the forward and reflected propagation modes (¿ TEM) on multiple coupled transmission lines in an inhomogeneous dielectric medium is presented. This flow graph directy gives the different transfer functions, including S-parameters, in matrix form needed to calcul......A flow graph relating voltages and the forward and reflected propagation modes (¿ TEM) on multiple coupled transmission lines in an inhomogeneous dielectric medium is presented. This flow graph directy gives the different transfer functions, including S-parameters, in matrix form needed...... to calculate crosstalk on the lines. An 8 bit databus is analysed in the frequency-and time domain. This analysis shows, as expected, that crosstalk can be a problem in connection with high speed logic circuits. The same databus is also analysed using the quasi-propagation-mode method proposed by Dalby [1...

  20. ANALYSIS OF THE COMPETITIVE MANAGEMENT OF THE SMALL AND MEDIUM COMMERCIAL COMPANIES OF ESMERALDAS, REPUBLIC OF THE ECUADOR

    Directory of Open Access Journals (Sweden)

    Manuel Ruvin Quiñónez-Cabeza

    2016-01-01

    Full Text Available This paper analyzes the competitive administration of the small and medium companies of the smallest trade of Emeralds, Republic of the Ecuador. The different companies related with activities are studied that have discharge it demands in the local population. In this sense, an empiric study has been developed through 344 surveys and personal interviews carried out proprietors, administrators and workers. Finally, the most outstanding revelations in this empiric work are commented, standing out those important barriers that limit the competitiveness of the studied companies. 

  1. Spent fuel heatup following loss of water during storage

    International Nuclear Information System (INIS)

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

    1979-03-01

    An analysis of spent fuel heatup following a hypothetical accident involving drainage of the storage pool is presented. Computations based upon a new computer code called SFUEL have been performed to assess the effect of decay time, fuel element design, storage rack design, packing density, room ventilation, drainage level, and other variables on the heatup characteristics of the spent fuel and to predict the conditions under which clad failure will occur. Possible storage pool design modifications and/or onsite emergency action have also been considered

  2. Spent fuel. Dissolution and oxidation

    International Nuclear Information System (INIS)

    Grambow, B.

    1989-03-01

    Data from studies of the low temperature air oxidation of spent fuel were retrieved in order to provide a basis for comparison between the mechanism of oxidation in air and corrosion in water. U 3 O 7 is formed by diffusion of oxygen into the UO 2 lattice. A diffusion coefficient of oxygen in the fuel matric was calculated for 25 degree C to be in the range of 10 -23 to 10 -25 m 2 /s. The initial rates of U release from spent fuel and from UO 2 appear to be similar. The lowest rates (at 25 degree c >10 -4 g/(m 2 d)) were observed under reducing conditions. Under oxidizing conditions the rates depend mainly of the nature and concentraion of the oxidant and/or on corbonate. In contact with air, typical initial rates at room temperature were in the range between 0.001 and 0.1 g/(m 2 d). A study of apparent U solubility under oxidizing conditions was performed and it was suggested that the controlling factor is the redox potential at the UO 2 surface rather than the E h of the bulk solution. Electrochemical arguments were used to predict that at saturation, the surface potential will eventually reach a value given by the boundaries at either the U 3 O 7 /U 3 O 8 or the U 3 O 7 /schoepite stability field, and a comparison with spent fuel leach data showed that the solution concentration of uranium is close to the calculated U solubility at the U 3 O 7 /U 3 O 8 boundary. The difference in the cumulative Sr and U release was calculated from data from Studsvik laboratory. The results reveal that the rate of Sr release decreases with the square root of time under U-saturated conditions. This time dependence may be rationalized either by grain boundary diffusion or by diffusion into the fuel matrix. Hence, there seems to be a possibility of an agreement between the Sr release data, structural information and data for oxygen diffusion in UO 2 . (G.B.)

  3. Status of Proposed Repository for Latin-American Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ferrada, J.J.

    2004-10-04

    This report compiles preliminary information that supports the premise that a repository is needed in Latin America and analyzes the nuclear situation (mainly in Argentina and Brazil) in terms of nuclear capabilities, inventories, and regional spent-fuel repositories. The report is based on several sources and summarizes (1) the nuclear capabilities in Latin America and establishes the framework for the need of a permanent repository, (2) the International Atomic Energy Agency (IAEA) approach for a regional spent-fuel repository and describes the support that international institutions are lending to this issue, (3) the current situation in Argentina in order to analyze the Argentinean willingness to find a location for a deep geological repository, and (4) the issues involved in selecting a location for the repository and identifies a potential location. This report then draws conclusions based on an analysis of this information. The focus of this report is mainly on spent fuel and does not elaborate on other radiological waste sources.

  4. Development of nuclear spent fuel Maritime transportation scenario

    International Nuclear Information System (INIS)

    Yoo, Min; Kang, Hyun Gook

    2014-01-01

    Spent fuel transportation of South Korea is to be conducted through near sea because it is able to ship a large amount of the spent fuel far from the public comparing to overland transportation. The maritime transportation is expected to be increased and its risk has to be assessed. For the risk assessment, this study utilizes the probabilistic safety assessment (PSA) method and the notions of the combined event. Risk assessment of maritime transportation of spent fuel is not well developed in comparison with overland transportation. For the assessment, first, the transportation scenario should be developed and categorized. Categories are assorted into the locations, release aspects and exposure aspects. This study deals with accident that happens on voyage and concentrated on ship-ship collision. The collision accident scenario is generated with event tree analysis. The scenario will be exploited for the maritime transportation risk model which includes consequence and accident probability

  5. Handling of final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    1978-01-01

    In this report the various facilities incorporated in the proposed handling chain for spent fuel from the power stations to the final repository are discribed. Thus the geological conditions which are essential for a final repository is discussed as well as the buffer and canister materials and how they contribute towards a long-term isolation of the spent fuel. Furthermore one chapter deals with leaching of the deposited fuel in the event that the canister is penetrated as well as the transport mechanisms which determine the migration of the radioactive substances through the buffer material. The dispersal processes in the geosphere and the biosphere are also described together with the transfer mechanisms to the ecological systems as well as radiation doses. Finally a summary is given of the safety analysis of the proposed method for the handling and final storage of the spent fuel. (E.R.)

  6. Screening of Candida boidinii from Chemlal spent olive ...

    African Journals Online (AJOL)

    The morphological, biochemical characterization and 18S rDNA gene analysis of the selected strain, confirms that it is Candida boidinii KF156789. The production of lipase and biomass were carried out in liquid and solid (spent olive) media. In submerged fermentation, it seemed that the production of enzyme reached its ...

  7. Amino Acid Profile of Biodegraded Brewers Spent Grains (BSG ...

    African Journals Online (AJOL)

    The amino acids profiles of biodegraded brewers spent grains (BSG) were determined. The analysis revealed the presence of 17 amino acids including the major amino acids (cysteine, lysine and methionine) required in poultry nutrition. The concentrations of the amino acids however varied with the microbial species used ...

  8. Spent fuel shipping cask accident evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Fields, S.R.

    1975-12-01

    Mathematical models have been developed to simulate the dynamic behavior, following a hypothetical accident and fire, of typical casks designed for the rail shipment of spent fuel from nuclear reactors, and to determine the extent of radioactive releases under postulated conditions. The casks modeled were the IF-300, designed by the General Electric Company for the shipment of spent LWR fuel, and a cask designed by the Aerojet Manufacturing Company for the shipment of spent LMFBR fuel.

  9. Spent fuel management: reprocessing or storage

    International Nuclear Information System (INIS)

    Lima Soares, M.L. de; Oliveira Lopes, M.J. de

    1986-01-01

    A review of the spent fuel management concepts generally adopted in several countries is presented, including an analysis of the brazilian situation. The alternatives are the reprocessing, the interim storage and the final disposal in a repository after appropriate conditioning. The commercial operating reprocessing facilities in the Western World are located in France and in the United Kingdom. In the USA the anti-reprocessing policy from 1977 changed in 1981, when the government supported the resumption of commercial reprocessing and designated the private sector as responsible for providing these services. Small scale facilities are operating in India, Italy, Japan and West Germany. Pilot plants for LWR fuel are being planned by Spain, Pakistan and Argentina. (Author) [pt

  10. Spent fuel management: reprocessing or storage

    International Nuclear Information System (INIS)

    Lima Soares, M.L. de; Oliveira Lopes, M.J. de.

    1986-01-01

    A review of the spent fuel management concepts generally adopted in several countries is presented, including an analysis of the brazilian situation. The alternatives are the reprocessing, the interim storage and the final disposal in a repository after appropriate conditioning. The commercial operating reprocessing facilities in the Western World are located in France and in the United Kingdom. In the USA the anti-reprocessing policy from 1977 changed in 1981, when the Government supported the resumption of commercial reprocessing and designated the private sector as responsible for providing these services. Small scale facilities are operating in India, Italy, Japan and West Germany. Pilot plant for LWR fuel are being planned by Spain, Pakistan and Argentina. (Author) [pt

  11. Spent fuels transportation coming from Australia

    International Nuclear Information System (INIS)

    2002-01-01

    Maritime transportation of spent fuels from Australia to France fits into the contract between COGEMA and ANSTO, signed in 1999. This document proposes nine information cards in this domain: HIFAR a key tool of the nuclear, scientific and technological australian program; a presentation of the ANSTO Australian Nuclear Science and Technology Organization; the HIFAR spent fuel management problem; the COGEMA expertise in favor of the research reactor spent fuel; the spent fuel reprocessing at La Hague; the transports management; the transport safety (2 cards); the regulatory framework of the transports. (A.L.B.)

  12. Thermal model of spent fuel transport cask

    International Nuclear Information System (INIS)

    Ahmed, E.E.M.; Rahman, F.A.; Sultan, G.F.; Khalil, E.E.

    1996-01-01

    The investigation provides a theoretical model to represent the thermal behaviour of the spent fuel elements when transported in a dry shipping cask under normal transport conditions. The heat transfer process in the spent fuel elements and within the cask are modeled which include the radiant heat transfer within the cask and the heat transfer by thermal conduction within the spent fuel element. The model considers the net radiant method for radiant heat transfer process from the inner most heated element to the surrounding spent elements. The heat conduction through fuel interior, fuel-clad interface and on clad surface are also presented. (author) 6 figs., 9 refs

  13. A durable and dependable solution for RTR spent fuel management

    International Nuclear Information System (INIS)

    Thomasson, J.

    1999-01-01

    RTR Operators need efficient and cost-effective services for the management of their spent fuel and this, for the full lifetime of their facility. Thanks to the integration of transport, reprocessing and conditioning services, COGEMA provides a cogent solution, with the utmost respect for safety and preservation of the environment, for the short, medium and long terms. As demonstrated in this paper, this option offers the only durable and dependable solution for the RTR spent fuel management, leading to a conditioning for the final residues directly suitable for final disposal. The main advantage of such an option is obviously the significant reduction in terms of volume and radiotoxicity of the ultimate waste when compared to direct disposal of spent fuels. The efficiency of such a solution has been proven, some RTR operators having already trusted COGEMA for the management of their aluminide fuel. With its commitment in R and D activities for the development of a high performance and reprocessable LEU fuels, COGEMA will be able to propose a solution for all types of fuels, HEU and LEU

  14. Application of the Maxwell-Wagner-Hanai effective medium theory to the analysis of the interfacial polarization relaxations in conducting composite films

    International Nuclear Information System (INIS)

    Adohi, B J-P; Bouanga, C Vanga; Fatyeyeva, K; Tabellout, M

    2009-01-01

    A new approach to explain the interfacial polarization phenomenon in conducting composite films is proposed. HCl-doped poly(ethylene terephthalate) (PET) and polyamide-6 (PA-6) matrices with embedded polyaniline (PANI) particles as filler were investigated and analysed, combining dielectric spectroscopy and AFM electrical images with the effective medium theory analysis. Up to three relaxation peaks attributed to the interfacial polarization phenomena were detected in the studied frequency range (0.1 Hz-1 MHz). The AFM electrical images revealed that the doped PA-6/PANI composite can be modelled as a single-type particle medium and the PET/PANI one as a two-type particle medium. A simple dielectric loss expression was derived from the Maxwell-Wagner-Hanai mixture equation and was applied to the experimental data to identify the interfaces involved in each of the relaxation peaks. The parameter values (permittivity, conductivity, volume fraction of the PANI particles) were found to agree well with the measured one, hence validating the models.

  15. Statistical analysis of the occurrence of medium-scale traveling ionospheric disturbances over Brazilian low latitudes using OI 630.0 nm emission all-sky images

    Science.gov (United States)

    Candido, C. M. N.; Pimenta, A. A.; Bittencourt, J. A.; Becker-Guedes, F.

    2008-09-01

    In this work we report a statistical analysis of the occurrence frequency of medium-scale traveling ionospheric disturbances (MSTIDs) observed over Cachoeira Paulista (22.7°S, 45.0°W, -13.2° mag lat), Brazil. The optical signatures of the low-latitude MSTIDs in the southern hemisphere observed in the OI 630.0 nm emission images can be a single dark band structure or alternating dark/light bands aligned in the northeast-southwest direction and propagating towards northwest. Because this feature these events were also referred as thermospheric dark band structures. The statistical study is based on 28 events of MSTIDs observed during seven years of optical data, obtained during low, medium, and high solar activities, for geomagnetically quiet nights. We find that the occurrence frequency of the MSTIDs presents a maximum during low solar activity, decreasing during medium solar activity with no occurrences during high solar activity. Also, the occurrence rates are greater near the June-solstice months.

  16. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix D, Part B: Naval spent nuclear fuel management

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This volume contains the following attachments: transportation of Naval spent nuclear fuel; description of Naval spent nuclear receipt and handling at the Expended Core Facility at the Idaho National Engineering Laboratory; comparison of storage in new water pools versus dry container storage; description of storage of Naval spent nuclear fuel at servicing locations; description of receipt, handling, and examination of Naval spent nuclear fuel at alternate DOE facilities; analysis of normal operations and accident conditions; and comparison of the Naval spent nuclear fuel storage environmental assessment and this environmental impact statement.

  17. Situation and perspective of spent fuel management in Spain

    International Nuclear Information System (INIS)

    Lopez Garcia, A.

    2009-01-01

    Between september 2006 and september 2007, the Foundation for Energy Studies carried out the study Radioactive Waste Management: Situation, Analysis and Perspectives. This study focuses specially on spent fuel and high level radioactive waste management. The different aspects covered in this study are as follows: -Description of the different applicable technologies -Analysis and comparison of the different options of spent fuel management, including the strategic and economic aspects. - Situation, strategies and forecasts in the main countries. -Analysis of the situation and alternatives for the spent fuel management in spain. Although the report focuses principally pn the technological and environmental aspects other issues related with the management of these materials were considered, such as the strategic, economic and institutional aspects as well as the social acceptation. In answer to the request of the SNE publication, the article enclosed is a summary of one of the chapters of this study, and more particularly the one dedicated to the situation of spent fuel and high level radioactive waste management in Spain. (Author)

  18. Size-dependent thermal stability analysis of graded piezomagnetic nanoplates on elastic medium subjected to various thermal environments

    Science.gov (United States)

    Ebrahimi, Farzad; Barati, Mohammad Reza

    2016-10-01

    This paper investigates the thermal stability of magneto-electro-thermo-elastic functionally graded (METE-FG) nanoplates based on the nonlocal theory and a refined plate model. The METE-FG nanoplate is subjected to the external electric potential, magnetic potential and different temperature rises. Interaction of elastic medium with the METE-FG nanoplate is modeled via Winkler-Pasternak foundation model. The governing equations are derived by using the Hamilton principle and solved by using an analytical method to determine the critical buckling temperatures. To verify the validity of the developed model, the results of the present work are compared with those available in the literature. A detailed parametric study is conducted to study the influences of the nonlocal parameter, foundation parameters, temperature rise, external electric and magnetic potentials on the size-dependent thermal buckling characteristics of METE-FG nanoplates.

  19. Phase Change of Water in a Hygroscopic Porous Medium. Phenomenological Relation and Experimental Analysis for Water in Soil

    Science.gov (United States)

    Bénet, Jean-Claude; Lozano, Anne-Laure; Cherblanc, Fabien; Cousin, Bruno

    2009-06-01

    The phenomenological relation of non-equilibrium liquid-gas phase change in a porous medium is described at the macroscopic level using the difference in chemical potentials between the liquid and its vapor. The experiments conducted consisted in lowering the partial pressure of water vapor in the pores of a hygroscopic soil and analyzing the return to equilibrium by two measurements: the macroscopic temperature and the partial pressure of vapor. The central hypothesis of the study is that the characteristic time associated with thermal equilibrium is much lower than the characteristic time associated with mass transfers. From these measurements, it is possible to determine the relation that links phase change rate to the logarithm of the ratio of partial vapor pressure divided by the equilibrium pressure (RH). The representation of this relation according to RH reveals two regimes in the return to equilibrium. The characteristics of these regimes are analyzed according to water content, temperature, and total gas phase pressure.

  20. Dynamic analysis of reactive oxygen nitrogen species in plasma-activated culture medium by UV absorption spectroscopy

    Science.gov (United States)

    Brubaker, Timothy R.; Ishikawa, Kenji; Takeda, Keigo; Oh, Jun-Seok; Kondo, Hiroki; Hashizume, Hiroshi; Tanaka, Hiromasa; Knecht, Sean D.; Bilén, Sven G.; Hori, Masaru

    2017-12-01

    The liquid-phase chemical kinetics of a cell culture basal medium during treatment by an argon-fed, non-equilibrium atmospheric-pressure plasma source were investigated using real-time ultraviolet absorption spectroscopy and colorimetric assays. Depth- and time-resolved NO2- and NO3- concentrations were strongly inhomogeneous and primarily driven by convection during and after plasma-liquid interactions. H2O2 concentrations determined from deconvolved optical depth spectra were found to compensate for the optical depth spectra of excluded reactive species and changes in dissolved gas content. Plasma-activated media remained weakly basic due to NaHCO3 buffering, preventing the H+-catalyzed decomposition of NO2- seen in acidic plasma-activated water. An initial increase in pH may indicate CO2 sparging. Furthermore, the pH-dependency of UV optical depth spectra illustrated the need for pH compensation in the fitting of optical depth data.

  1. Environmental management in small and medium-sized companies: an analysis from the perspective of the theory of planned behavior.

    Science.gov (United States)

    Sánchez-Medina, Agustín J; Romero-Quintero, Leonardo; Sosa-Cabrera, Silvia

    2014-01-01

    In the business context, concern for the environment began to develop when pressure from the public administration and environmental awareness groups raised the specific requirements for companies. The Theory of Planned Behavior considers that people's conduct is determined by the intention of carrying out a certain behavior. Thus, the individual's intent is determined by three factors related to the desired outcome of the behavior: the Personal Attitude toward the Results, the Perceived Social Norms, and the Perceived Behavioral Control over the action. Therefore, the objectives of this paper are to clarify the attitudes of the managers of Canarian small and medium-sized companies about taking environmental measures, and try to demonstrate whether there is a relationship between the proposed factors and the intention to take these measures.

  2. An Analysis of Accountancy Firms Perspectives of the Need to Apply IFRS in Small and Medium Enterprises

    Directory of Open Access Journals (Sweden)

    Fernando Lins Alves

    2013-12-01

    Full Text Available This paper examines the services provided by accountancy professionals to small and medium enterprises (SME. More specifically, the purpose is to understand the level of utilisation of international standards to these organisations “IFRS for SME” in Brazil in accordance to the national standards (NBC T 19.41. Overall, 32 interviews were carried out in accountancy organisations in the city of Recife-PE. The study focuses on the knowledge, utilisation and applicability of the International Standards to SME. The results show that there is a lack of utilisation of the standards and that some accountancy professionals do not have knowledge regarding the changes produced by the convergence process. Moreover, the study reveals that the majority of accountancy professionals consider unnecessary to adopt the “IFRS for SME”.

  3. [Cultivation strategy and path analysis on big brand Chinese medicine for small and medium-sized enterprises].

    Science.gov (United States)

    Wang, Yong-Yan; Yang, Hong-Jun

    2014-03-01

    Small and medium-sized enterprises (SMEs) are important components in Chinese medicine industry. However, the lack of big brand is becoming an urgent problem which is critical to the survival of SMEs. This article discusses the concept and traits of Chinese medicine of big brand, from clinical, scientific and market value three aspects. Guided by market value, highlighting clinical value, aiming at the scientific value improvement of big brand cultivation, we put forward the key points in cultivation, aiming at obtaining branded Chinese medicine with widely recognized efficacy, good quality control system and mechanism well explained and meanwhile which can bring innovation improvement to theory of Chinese medicine. According to the characters of SMEs, we hold a view that to build multidisciplinary research union could be considered as basic path, and then, from top-level design, skill upgrading and application three stages to probe the implementation strategy.

  4. A Case Analysis on the Adequacy of Work-Life Balance Practices in UK Small- and Medium-Sized Enterprises

    Directory of Open Access Journals (Sweden)

    Babatunde Akanji

    2017-09-01

    Full Text Available Objective: The purpose of this study is to examine whether work-life balance (WLB practices are satisfactorily provided in UK small and medium-sized enterprises (SMEs and the impact of the availability of these work-life policies on turnover intentions. A review of extant literature reveals scarce knowledge in this area of research and this study presents a rudimentary effort to fill this gap. Research Design & Methods: Using qualitative design, the data set comprised of in-depth interviews with thirty-six employees working in small and medium-sized UK convenience stores and supermarkets with less than ninety employees. Findings: Informal nature of human resource management policies emerged from the findings as one of the constraining forces impeding work-life agendas in SMEs and causing low staff retention in UK. Although other themes were found to contribute to retention challenges, however, these additional reasons were not independent, but all considered integrated. Implications & Recommendations: Consequently, the practical implication of devising ways to overcome WLB and retention deficiencies in this context were also explored. Contribution & Value Added: The originality of this work lies in studying the importance of WLB practices to some of these grass root businesses that make up a large proportion of the economy in the UK. As the limitation of this study is that it is wholly qualitative in nature, it is suggested that future research should rely on quantitative designs that provides more internally valid tests via computational techniques.

  5. [Analysis of use of personal protective equipment among rural-to-urban migrant workers in small and medium enterprises in Zhongshan and Shenzhen, China].

    Science.gov (United States)

    Zeng, Zhi; Lu, Liming; Rao, Zhanhong; Han, Lu; Shi, Jingrong; Ling, Li

    2014-04-01

    To investigate the current supply and use of personal protective equipment (PPE) among rural-to-urban migrant workers in small and medium enterprises (SMEs) in Zhongshan and Shenzhen, China and the influential factors for the use of PPE, and to provide a basis for better occupational health services and ensuring the health of migrant workers. Multi-stage sampling was used to select 856 migrant workers from 27 SMEs in Zhongshan and Shenzhen, and face-to-face questionnaire survey was conducted in these subjects. Statistical analysis was performed by one-way analysis of variance, chi-square test, and logistic regression. Of all migrant workers, 38.67%were supplied with free PPE by the factory, and this rate varied across industries (furniture industry: 45.81%; electronic industry: 31.46%) and SMEs (medium enterprises: 42.13%; small enterprises: 39.20%; micro enterprises: 22.16%); 22.43% insisted on the use of PPE. The logistic regression analysis showed that factors associated with the use of PPE included sex, age, awareness of occupational health knowledge, and the size of enterprise. The rates of supply and use of PPE among migrant workers are low. The larger the enterprise, the better the supply of PPE. Male gender, being elder, and high occupational health knowledge score were favorable factors for the use of PPE, while small enterprise size was the unfavorable factor for the use of PPE.

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  7. Research reactor spent fuel in Ukraine

    International Nuclear Information System (INIS)

    Trofimenko, A.P.

    1996-01-01

    This paper describes the research reactors in Ukraine, their spent fuel facilities and spent fuel management problems. Nuclear sciences, technology and industry are highly developed in Ukraine. There are 5 NPPs in the country with 14 operating reactors which have total power capacity of 12,800 MW

  8. Prospects of spent management in Spain

    International Nuclear Information System (INIS)

    Melches, C.; Ramirez, E.; Selgas, F.; Cabanilles, P.A.; Lopez Perez, B.; Uriarte, A.

    1978-01-01

    The purpose of this paper is to outline the forecast on spent fuel management in Spain, taking into account the international developments produced during the last years and specially on LWR fuels. This forecast is based on the following actions: increase of the storage capacity in the reactors: construction of an independent spent fuel storage installation (ISFSI) and a fuel reprocessing pilot plant. (author)

  9. Spent fuel storage for ISER plant

    International Nuclear Information System (INIS)

    Nakajima, Takasuke; Kimura, Yuzi

    1987-01-01

    ISER is an intrinsically safe reactor basing its safety only on physical laws, and uses a steel reactor vessel in order to be economical. For such a new type reactor, it is essentially important to be accepted by the society by showing that the reactor is more profitable than conventional reactors to the public in both technical and economic viewpoint. It is also important that the reactor raises no serious problem in the total fuel cycle. Reprocessing seems one of the major worldwide fuel cycle issues. Spent fuel storage is also one of the key technologies for fuel cycle back end. Various systems for ISER spent fuel storages are examined in the present report. Spent fuel specifications of ISER are similar to those of LWR and therefore, most of LWR spent fuel technologies are basically applicable to ISER spent fuel. Design requirements and examples of storage facilities are also discussed. Dry storage seems to be preferable for the relatively long cooling time spent fuel like ISER's one from economical viewpoint. Vault storage will possibly be the most advantageous for large storage capacity. Another point for discussion is the location and international collaboration for spent fuel storages: ISER expected to be a worldwide energy source and therefore, international spent fuel management seems to be fairly attractive way for an energy recipient country. (Nogami, K.)

  10. Spent Nuclear Fuel (SNF) Project Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    LEROY, P.G.

    2000-11-03

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

  11. Spent Nuclear Fuel (SNF) Project Execution Plan

    International Nuclear Information System (INIS)

    LEROY, P.G.

    2000-01-01

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities

  12. What does time spent on searching indicate?

    DEFF Research Database (Denmark)

    Borlund, Pia; Dreier, Sabine; Byström, Katriina

    2012-01-01

    In this paper, we report a comparative study on what users’ time spent on searching for information is an indication of. Time spent is commonly interpreted as an implicit measure of interest, but might indeed describe other circumstances of the information retrieval (IR) interaction. This phenome......In this paper, we report a comparative study on what users’ time spent on searching for information is an indication of. Time spent is commonly interpreted as an implicit measure of interest, but might indeed describe other circumstances of the information retrieval (IR) interaction....... This phenomenon of time spent is interesting from an IR evaluation point of view with reference to how time spent is to be interpreted. A comparison of time spent between a semi-lab interactive IR (IIR) study using simulated work task situations and a naturalistic IIR study is presented. The findings...... of this comparison are further related to a study on information searching and seeking in the real work environment that provides a resonance board for the reported IIR studies. The main conclusion is that time spent searching depends not only on interest, but also on circumstances such as prior knowledge...

  13. Investigation of nuclide importance to functional requirements related to transport and long-term storage of LWR spent fuel

    International Nuclear Information System (INIS)

    Broadhead, B.L.; DeHart, M.D.; Ryman, J.C.; Tang, J.S.; Parks, C.V.

    1995-06-01

    This study investigates the relative importances of the various actinide, fission-product, and light-element isotopes associated with LWR spent fuel with respect to five analysis areas: criticality safety (absorption fractions), shielding (dose rate fractions), curies (fractional curies levels), decay heat (fraction of total watts), and radiological toxicity (fraction of potential committed effective dose equivalent). These rankings are presented for up to six different burnup/enrichment scenarios and at decay times from 2 to 100,000 years. Ranking plots for each of these analysis areas are given in an Appendix for completeness, as well as summary tables in the main body of the report. Summary rankings are presented in terms of high (greater than 10% contribution to the total), medium (between 1% and 10% contribution), and low (less than 1% contribution) for both short- and long-term cooling. When compared with the expected measurement accuracies, these rankings show that most of the important isotopes can be characterized sufficiently for the purpose of radionuclide generation/depletion code validation in each of the analysis areas. Because the main focus of this work is on the relative importances of isotopes associated with L at sign spent fuel, some conclusions may not be applicable to similar areas such as high-level waste (HLW) and nonfuel-bearing components (NFBC)

  14. Corrosion of research reactor aluminium clad spent fuel in water

    International Nuclear Information System (INIS)

    2009-12-01

    reactor aluminium clad spent fuel. These corrosion activities were quite similar to those carried out in the CRP. Eight Member States participated in Phase-II of the CRP and five Member States in the Regional Project RLA/4/018. Two of the countries participating in the regional project were also participants in the CRP. This report documents the work performed in the IAEA Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II) and in the IAEA's Technical Cooperation Regional Project for Latin America (RLA/4/018) entitled Management of Spent Fuel from Research Reactors. The key activity of both, the CRP and the Regional Project, consisted of the exposure of standard racks of corrosion coupons in the spent fuel pools of the participating research reactor laboratories and the evaluation of the coupons after predetermined exposure times, along with periodic monitoring of the storage water and evaluation of sediments settling in the spent fuel pools. The report includes: a description of the standard corrosion racks, experimental protocols, test procedures and water quality monitoring; the specific contributions by each of the participating laboratories; a compilation of all experimental results obtained and the analysis and discussion of the results, along with conclusions

  15. 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

  16. Spent fuel management in India

    International Nuclear Information System (INIS)

    Balu, K.

    1998-01-01

    From Indian point of view, the spent fuel management by the reprocessing and plutonium recycle option is considered to be a superior and an inevitable option. The nuclear energy programme in Indian envisages three stages of implementation involving installation of thermal reactors in the first phase followed by recycling of plutonium from reprocessed fuel in fast breeder reactors and in the third phase utilization of its large thorium reserves in reactor system based on U-233-Th cycle. The Indian programme for Waste Management envisages disposal of low and intermediate level radioactive waste in near surface disposal facilities and deep geological disposal for high level and alpha bearing wastes. A Waste Immobilization Plant (WHIP), employing metallic melter for HLW vitrification is operational at Tarapur. Two more WIPs are being set up at Kalpakkam and Tarapur. A Solid waste Storage Surveillance Facility (SSSF) is also set up for interim storage of vitrified HLW. Site investigations are in progress for selecting site for ultimate disposal in igneous rock formations. R and D works is taken up on partitioning of HLW. Solvent extraction and extraction chromatographic studies are in progress. Presently emphasis is on separation of heat generating short lived nuclides like strontium and alpha emitters. (author)

  17. Analysis of heat transfer for unsteady MHD free convection flow of rotating Jeffrey nanofluid saturated in a porous medium

    Science.gov (United States)

    Mohd Zin, Nor Athirah; Khan, Ilyas; Shafie, Sharidan; Alshomrani, Ali Saleh

    In this article, the influence of thermal radiation on unsteady magnetohydrodynamics (MHD) free convection flow of rotating Jeffrey nanofluid passing through a porous medium is studied. The silver nanoparticles (AgNPs) are dispersed in the Kerosene Oil (KO) which is chosen as conventional base fluid. Appropriate dimensionless variables are used and the system of equations is transformed into dimensionless form. The resulting problem is solved using the Laplace transform technique. The impact of pertinent parameters including volume fraction φ , material parameters of Jeffrey fluid λ1 , λ , rotation parameter r , Hartmann number Ha , permeability parameter K , Grashof number Gr , Prandtl number Pr , radiation parameter Rd and dimensionless time t on velocity and temperature profiles are presented graphically with comprehensive discussions. It is observed that, the rotation parameter, due to the Coriolis force, tends to decrease the primary velocity but reverse effect is observed in the secondary velocity. It is also observed that, the Lorentz force retards the fluid flow for both primary and secondary velocities. The expressions for skin friction and Nusselt number are also evaluated for different values of emerging parameters. A comparative study with the existing published work is provided in order to verify the present results. An excellent agreement is found.

  18. Numerical analysis of 3D micropolar nanofluid flow induced by an exponentially stretching surface embedded in a porous medium

    Science.gov (United States)

    Subhani, M.; Nadeem, S.

    2017-10-01

    The present article is devoted to probe the behavior of a three-dimensional micropolar nanofluid over an exponentially stretching surface in a porous medium. The mathematical model is constructed in the form of partial differential equations using the boundary layer approach. Then by employing similarity transformations, the modelled partial differential equations are transformed to ordinary differential equations. The solution of subsequent ODEs is derived by utilizing the BVP-4C technique alongside the shooting scheme. The graphical illustrations are presented to interpret the salient features of pertinent physical parameters on the concerned profiles for different kinds of nanoparticles, namely copper, titania and alumina with water as the base fluid. For a better understanding of the fluid flow, the numerical variation in the local skin friction coefficients, Cfx and Cfy , and local Nusselt number is analyzed through tables. We can see, from the present study, that the omission of porous matrix enhances the flow of the fluid. Microrotation has a decreasing impact on the skin friction whereas it increases the rate of the heat transfer of the nanofluid.

  19. Analysis of heat transfer for unsteady MHD free convection flow of rotating Jeffrey nanofluid saturated in a porous medium

    Directory of Open Access Journals (Sweden)

    Nor Athirah Mohd Zin

    Full Text Available In this article, the influence of thermal radiation on unsteady magnetohydrodynamics (MHD free convection flow of rotating Jeffrey nanofluid passing through a porous medium is studied. The silver nanoparticles (AgNPs are dispersed in the Kerosene Oil (KO which is chosen as conventional base fluid. Appropriate dimensionless variables are used and the system of equations is transformed into dimensionless form. The resulting problem is solved using the Laplace transform technique. The impact of pertinent parameters including volume fraction φ, material parameters of Jeffrey fluid λ1, λ, rotation parameter r, Hartmann number Ha, permeability parameter K, Grashof number Gr, Prandtl number Pr, radiation parameter Rd and dimensionless time t on velocity and temperature profiles are presented graphically with comprehensive discussions. It is observed that, the rotation parameter, due to the Coriolis force, tends to decrease the primary velocity but reverse effect is observed in the secondary velocity. It is also observed that, the Lorentz force retards the fluid flow for both primary and secondary velocities. The expressions for skin friction and Nusselt number are also evaluated for different values of emerging parameters. A comparative study with the existing published work is provided in order to verify the present results. An excellent agreement is found. Keywords: Jeffrey nanofluid, AgNPs, MHD and Porosity, Rotating flow, Laplace transform technique

  20. Lie group analysis and numerical solutions for non-Newtonian nanofluid flow in a porous medium with internal heat generation

    International Nuclear Information System (INIS)

    Uddin, Md Jashim; Yusoff, N H Md; Ismail, Ahamd Izani; Anwar Bég, O

    2013-01-01

    A mathematical model is presented and analysed for steady two-dimensional non-isothermal boundary layer flow from a heated horizontal surface which is embedded in a porous medium saturated with a non-Newtonian power-law nanofluid. It is assumed that the wall temperature and nanoparticle volume fraction vary nonlinearly with the axial distance. By applying appropriate group transformations, the governing transport equations are reduced to a system of coupled, nonlinear ordinary differential equations with associated boundary conditions. The reduced equations are then solved numerically using the Runge–Kutta–Fehlberg fourth–fifth-order numerical method with Maple 13 software. The effects of several thermophysical parameters including rheological power-law index, non-isothermal index, Lewis number, Brownian motion parameter, thermophoresis parameter, buoyancy ratio and internal heat generation/absorption parameter on the non-dimensional velocity, temperature, nanoparticle volume fraction (concentration) and also on the friction factor, heat and mass transfer rates are investigated. A comparison of the present results with the existing published results shows excellent agreement, verifying the accuracy of the present numerical code. The study finds applications in nano biopolymeric manufacturing processes and also thermal enhancement of energy systems employing rheological working fluids. (paper)