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Sample records for improved sfr cores

  1. Heterogeneous recycling in SFR core periphery

    International Nuclear Information System (INIS)

    Varaine, Frederic; Buiron, Laurent; Boucher, Lionel; Chabert, Christine

    2008-01-01

    development, based on the one hand on the solutions offered by the existing fleet (reprocessing, fabrication and NPP) and on the other hand on the solutions offered by the future reactors of fourth generation. The scenario study considers the French nuclear park with a constant nuclear energy demand at 430 TWhe / year. The current nuclear park is replaced between 2020 and 2050 by a mixed nuclear park: 66 % of Generation III EPR reactors and 33% of Generation IV SFR. From 2080 to 2100, the EPR are replaced by SFR. The Plutonium is recycled in the fissile part of the SFR core. The separation of the minor actinides at the reprocessing step starts in 2038. The minor actinides are recycled in the radial blankets of the SFR from 2040 (10% content of MA). Those calculations are performed by the COSI code. The results indicate that the minor actinides inventory can be stabilized with the heterogeneous mode of transmutation using minor actinides in the radial blankets of the SFR. A minor actinides rate around 10% in the radial blankets is sufficient with the condition to involve 100 % of the SFR in the transmutation process. The minor actinides multi-recycling on a depleted uranium oxide matrix in radial blankets of SFR showed good results in terms of transmutation performances. This heterogeneous model allows a massive minor actinides loading while having almost no consequence on the core safety parameters and core fuel management. Two MA enrichment targets have been studied: an ambitious 40% case and a more realistic 10% case. The design of such assembly has to deal with criteria implying multi-physics analysis. The 10% MA content seems a good balance between transmutation performances and back/front end impact (neutrons source, decay heat,..) compared to the 40% content. Investigations, such as dedicated experimental material and fuel irradiation programs, are under process at CEA to set a global vision of an optimized system that can answer all these problems. (authors)

  2. Application of the SPH method in nodal diffusion analyses of SFR cores

    Energy Technology Data Exchange (ETDEWEB)

    Nikitin, Evgeny; Fridman, Emil [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety; Mikityuk, K. [Paul Scherrer Institut, Villigen (Switzerland)

    2016-07-01

    The current study investigated the potential of the SPH method, applied to correct the few-group XS produced by Serpent, to further improve the accuracy of the nodal diffusion solutions. The procedure for the generation of SPH-corrected few-group XS is presented in the paper. The performance of the SPH method was tested on a large oxide SFR core from the OECD/NEA SFR benchmark. The reference SFR core was modeled with the DYN3D and PARCS nodal diffusion codes using the SPH-corrected few-group XS generated by Serpent. The nodal diffusion results obtained with and without SPH correction were compared to the reference full-core Serpent MC solution. It was demonstrated that the application of the SPH method improves the accuracy of the nodal diffusion solutions, particularly for the rodded core state.

  3. Coupled MCNP - SAS-SFR calculations for sodium fast reactor core at steady-state - 15460

    International Nuclear Information System (INIS)

    Ponomarev, A.; Travleev, A.; Pfrang, W.; Sanchez, V.

    2015-01-01

    The prediction of core parameters at steady state is the first step when studying core accident transient behaviour. At this step thermal hydraulics (TH) and core geometry parameters are calculated corresponding to initial operating conditions. In this study we present the coupling of the SAS-SFR code to the Monte-Carlo neutron transport code MCNP at steady state together with application to the European Sodium Fast Reactor (ESFR). The SAS-SFR code employs a multi-channel core representation where each channel represents subassemblies with similar power, thermal-hydraulics and pin mechanics conditions. For every axial node of every channel the individual geometry and material compositions parameters are calculated in accord with power and cooling conditions. This requires supplying the SAS-SFR-code with nodal power values which should be calculated by neutron physics code with given realistic core parameters. In the conventional approach the neutron physics model employs some core averaged TH and geometry data (fuel temperature, coolant density, core axial and radial expansion). In this study we organize a new approach coupling the MCNP neutron physics models and the SAS-SFR models, so that calculations of power can be improved by using distributed core parameters (TH and geometry) taken from SAS-SFR. The MCNP code is capable to describe cores with distributed TH parameters and even to model non-uniform axial expansion of fuel subassemblies. In this way, core TH and geometrical data calculated by SAS-SFR are taken into account accurately in the neutronics model. The coupling implementation is done by data exchange between two codes with help of processing routines managed by driver routine. Currently it is model-specific and realized for the ESFR 'Reference Oxide' core. The Beginning-Of-Life core state is considered with 10 channel representation for fuel subassemblies. For this model several sets of coupled calculations are performed, in which different

  4. Evaluation of core modeling effect on transients for multi-flow zone design of SFR

    International Nuclear Information System (INIS)

    Shin, Andong; Choi, Yong Won

    2016-01-01

    SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

  5. Evaluation of core modeling effect on transients for multi-flow zone design of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Andong; Choi, Yong Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

  6. Integrated CFD investigation of heat transfer enhancement using multi-tray core catcher in SFR

    International Nuclear Information System (INIS)

    Rakhi; Sharma, Anil Kumar; Velusamy, K.

    2017-01-01

    Highlights: • Heat transfer enhancement using multi-tray core catcher for SFR is investigated. • The capability of a single core collector tray is estimated. • Double and triple collector trays with innovative designs is discussed. • Provision of openings in the trays contributed to enhanced natural circulation. - Abstract: To render future SFR more robust and safe, certain BDBE have been considered in the recent years. A Core Disruptive Accident leading to a whole core meltdown scenario has gained the interest of researchers. Various design concepts and safety measures have been suggested and incorporated in design to address such a low probability scenario. A core catcher concept, in particular, has proved to be inevitable as an in-vessel core retention device in SFR for safe retention of core debris arising out after the severe accident. This study aims to analyse the cooling capability of the innovative design concept of core catcher to remove decay heat of degraded core after the accident. First, the capability of single collection tray is established and then the study is extended to two and three collection trays with different design concepts. Transient forms of governing equations of mass, momentum and energy conservations along with k-ε turbulence model are solved by finite volume based CFD solver. Boussinesq approximation is invoked to model buoyancy in sodium. The study shows that a single collection tray is capable of removing up to 20 MW decay heat load in a typical 500 MWe pool type SFR. Further, studies are carried out to improve the natural circulation of sodium around the source, in the lower plenum and to distribute core debris of the whole core to multiple collection trays. It is found that the double and triple collection trays can accommodate decay loads up to 29 MW. Provision of openings in the collection trays has proved to be effective in improving the heat transfer and sodium flow as well as in distributing the core debris to the

  7. Improvement of Steam Generator Reliability for GEN-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong O; Kim Se Yun; Kim, Seok Hoon; Eoh, Jae Hyuk; Lee, Hyeong Yeon; Choi, Byung Seon

    2005-11-15

    The R and D items performed in this study were selected from the R and D task of ' Reliability improvement of Steam Generator' of GEN-IV SFR Component Design and BOP. Since this project deals with one of the most important issues for a GEN-IV SFR system, it needs to enhance the domestic technical backgrounds associated with the corresponding R and D items even for a very short period by 2005. This study provides the R and D results for i) Development of assessment methodology for dissimilar metal weld and ii) Development of multi-dimensional simulation methodology for a SWR event in a SFR steam generator.

  8. Improvement of Steam Generator Reliability for GEN-IV SFR

    International Nuclear Information System (INIS)

    Kim, Seong O; Kim Se Yun; Kim, Seok Hoon; Eoh, Jae Hyuk; Lee, Hyeong Yeon; Choi, Byung Seon

    2005-11-01

    The R and D items performed in this study were selected from the R and D task of ' Reliability improvement of Steam Generator' of GEN-IV SFR Component Design and BOP. Since this project deals with one of the most important issues for a GEN-IV SFR system, it needs to enhance the domestic technical backgrounds associated with the corresponding R and D items even for a very short period by 2005. This study provides the R and D results for i) Development of assessment methodology for dissimilar metal weld and ii) Development of multi-dimensional simulation methodology for a SWR event in a SFR steam generator

  9. Uncertainty Evaluation of Reactivity Coefficients for a large advanced SFR Core Design

    International Nuclear Information System (INIS)

    Khamakhem, Wassim; Rimpault, Gerald

    2008-01-01

    Sodium Cooled Fast Reactors are currently being reshaped in order to meet Generation IV goals on economics, safety and reliability, sustainability and proliferation resistance. Recent studies have led to large SFR cores for a 3600 MWth power plants, cores which exhibit interesting features. The designs have had to balance between competing aspects such as sustainability and safety characteristics. Sustainability in neutronic terms is translated into positive breeding gain and safety into rather low Na void reactivity effects. The studies have been done on two SFR concepts using oxide and carbide fuels. The use of the sensitivity theory in the ERANOS determinist code system has been used. Calculations have been performed with different sodium evaluations: JEF2.2, ERALIB-1 and the most recent JEFF3.1 and ENDF/B-VII in order to make a broad comparison. Values for the Na void reactivity effect exhibit differences as large as 14% when using the different sodium libraries. Uncertainties due to nuclear data on the reactivity coefficients were performed with BOLNA variances-covariances data, the Na Void Effect uncertainties are near to 12% at 1σ. Since, the uncertainties are far beyond the target accuracy for a design achieving high performance, two directions are envisaged: the first one is to perform new differential measurements or in a second attempt use integral experiments to improve effectively the nuclear data set and its uncertainties such as performed in the past with ERALIB1. (authors)

  10. Core Design Concept and Core Structural Material Development for a Prototype SFR

    International Nuclear Information System (INIS)

    Chang, Jinwook

    2013-01-01

    Core design Concept: – Initial core is Uranium metal fueled core, then it will evolve into TRU core; – Tight pressure drop constraint lowers power density; – Trade-off studies with relaxed pressure drop constraint (~0.4MPa) are on-going; – Major feature will be finalized this year. • KAERI is developing advanced cladding for high burnup fuel in Ptototype SFR: – Advanced cladding materials are now developing, which shows superior high temperature mechanical property to the conventional material; – Processing technologies related to tube making process are now developed to enhance high temperature mechanical propertyl – Preliminary HT9 cladding tube was manufactured and out-of pile mechanical properties were evaluated. Advanced cladding tube is now being developed and being prepared for irradiation test

  11. Heat transfer analysis to investigate the core catcher plate assembly in SFR

    International Nuclear Information System (INIS)

    Patil, Swapnil; Sharma, Anil Kumar; Velusamy, K.; Nashine, B.K.; Selvaraj, P.

    2015-01-01

    Severe accident scenario in Sodium Cooled Fast Reactor (SFR) is the major concern for public acceptance. After severe accident, the molten core continuously generates substantial decay heat. However, an in-vessel core catcher plate is provided to remove the decay heat passively. The numerical investigation of pool hydraulics phenomena in sodium pool of typical Indian SFR has been carried out. The debris may form a heap with different angle over the core catcher plate due to molten fuel density and interaction force. Therefore, the debris bed with different heap angle has been analyzed for steady and transient state conditions. The governing equation of fluid flow and heat transfer are solved by finite volume method based solver with the k-ε turbulent model. The time period Δ for which temperature is exceeding above safety limit with different debris heap angle have been established. (author)

  12. Multi-physics and multi-objective design of heterogeneous SFR core: development of an optimization method under uncertainty

    International Nuclear Information System (INIS)

    Ammar, Karim

    2014-01-01

    Since Phenix shutting down in 2010, CEA does not have Sodium Fast Reactor (SFR) in operating condition. According to global energetic challenge and fast reactor abilities, CEA launched a program of industrial demonstrator called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a reactor with electric power capacity equal to 600 MW. Objective of the prototype is, in first to be a response to environmental constraints, in second demonstrates the industrial viability of SFR reactor. The goal is to have a safety level at least equal to 3. generation reactors. ASTRID design integrates Fukushima feedback; Waste reprocessing (with minor actinide transmutation) and it linked industry. Installation safety is the priority. In all cases, no radionuclide should be released into environment. To achieve this objective, it is imperative to predict the impact of uncertainty sources on reactor behaviour. In this context, this thesis aims to develop new optimization methods for SFR cores. The goal is to improve the robustness and reliability of reactors in response to existing uncertainties. We will use ASTRID core as reference to estimate interest of new methods and tools developed. The impact of multi-Physics uncertainties in the calculation of the core performance and the use of optimization methods introduce new problems: How to optimize 'complex' cores (i.e. associated with design spaces of high dimensions with more than 20 variable parameters), taking into account the uncertainties? What is uncertainties behaviour for optimization core compare to reference core? Taking into account uncertainties, optimization core are they still competitive? Optimizations improvements are higher than uncertainty margins? The thesis helps to develop and implement methods necessary to take into account uncertainties in the new generation of simulation tools. Statistical methods to ensure consistency of complex multi-Physics simulation results are also

  13. Core Thermal-Hydraulic Conceptual Design for the Advanced SFR Design Concepts

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Chang, Jin Wook; Yoo, Jae Woon; Song, Hoon; Choi, Sun Rock; Park, Won Seok; Kim, Sang Ji

    2010-01-01

    The Korea Atomic Energy Research Institute (KAERI) has developed the advanced SFR design concepts from 2007 to 2009 under the National longterm Nuclear R and D Program. Two types of core designs, 1,200 MWe breakeven and 600 MWe TRU burner core have been proposed and evaluated whether they meet the design requirements for the Gen IV technology goals of sustainability, safety and reliability, economics, proliferation resistance, and physical protection. In generally, the core thermal hydraulic design is performed during the conceptual design phase to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power of each assembly because the conventional SFR core is composed of hundreds of ducted assemblies with hundreds of fuel rods. In carrying out the thermal and hydraulic design, special attention has to be paid to several performance parameters in order to assure proper performance and safety of fuel and core; the coolant boiling, fuel melting, structural integrity of the components, fuel-cladding eutectic melting, etc. The overall conceptual design procedure for core thermal and hydraulic conceptual design, i.e., flow grouping and peak pin temperature calculations, pressure drop calculations, steady-state and detailed sub-channel analysis is shown Figure 1. In the conceptual design phase, results of core thermal-hydraulic design for advanced design concepts, the core flow grouping, peak pin cladding mid-wall temperature, and pressure drop calculations, are summarized in this study

  14. A Preliminary Design Study of Ultra-Long-Life SFR Cores having Heterogeneous Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Jung, GeonHee; You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    The PWR and CANDU reactors have provided electricity for several decades in our country but they have produced lots of spent fuels and so the safe and efficient disposal of these spent fuels is one of the main issues in nuclear industry. This type ultra-long-life cores are quite efficient in terms of the amount of spent fuel generation per electricity production and they can be used as an interim storage for PWR or CANDU spent fuel over several tens of years if they use the PWR or CANDU spent fuel as the initial fuel. Typically, the previous works have considered radially homogeneous fuel assemblies in which only blanket or driver fuel rods are employed and they considered axially or radially heterogeneous core configurations with the radially homogeneous fuel assemblies. These core configurations result in the propagation of the power distribution which can lead to the significant temperature changes for each fuel assembly over the time. In this work, the radially heterogeneous fuel assemblies are employed in new ultra-long-life SFR (Sodium-cooled Fast Reactor) cores to minimize the propagation of power distribution by allowing the power propagation in the fuel assemblies. In this work, new small ultra-long life SFR cores were designed with heterogeneous fuel assemblies having both blanket and driver fuel rods to minimize the propagation of power distribution over the core by allowing power propagation from driver rods to blanket rods in fuel assemblies. In particular, high fidelity depletion calculation coupled with heterogeneous Monte Carlo neutron transport calculation was performed to assess the neutronic feasibility of the ultralong life cores. The results of the analysis showed that the candidate core has the cycle length of 77 EFPYs, a small burnup reactivity swing of 1590 pcm and acceptably small SVRs both at BOC and EOC.

  15. Numerical investigation on turbulent natural convection in partially connected cylindrical enclosures for analysing SFR safety under core meltdown scenario

    International Nuclear Information System (INIS)

    David, Dijo K.; Mangarjuna Rao, P.; Nashine, B.K.; Selvaraj, P.

    2015-01-01

    Under the unlikely event of severe core meltdown accident in pool type SFR, the molten core materials may rupture the grid plate which supports the fuel subassemblies and it can get relocated in to the lower pool. These debris may eventually settle on the debris collector (i.e., core catcher) installed above the bottom wall of the lower pool. The bed thus formed generates heat due to radioactive decay which has to be passively removed for maintaining the structural integrity of main vessel. By means of natural convection, the heat generated in the debris bed will be transferred to the top pool where the heat sink (i.e., Decay heat exchanger (DHX)) is installed. Heat transfer to the DHX (which is a part of safety grade decay heat removal system) can take place through the opening created in the grid plate which connects the two liquid pools (i.e., the top pool and the lower pool). Heat transfer can also take place through the lateral wall of the lower cylindrical pool to the side pool and eventually to the top pool, and thus to the DHX. This study numerically investigates the effectiveness of heat transfer between lower pool and top pool during PARR by considering them as partially connected cylindrical enclosures. The governing equations have been numerically solved using finite volume method in cylindrical co-ordinates using SIMPLE algorithm. Turbulence has been modeled using k-ω model and the model is validated against benchmark problems of natural convection found in literature. The effect of parameters such as the heat generation rate in the bed and the size of the grid plate opening are evaluated. Also PAHR in SFR pool is modeled using an axi-symmetric model to fund out the influence of grid plate opening on heat removal from core catcher. The results obtained are useful for improving the cooling capability of in-vessel tray type core catcher for handling the whole core meltdown scenarios in SFR. (author)

  16. Improving SFR Economics through Innovations from Thermal Design and Analysis Aspects

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Hongbin Zhang; Vincent Mousseau; Per F. Peterson

    2008-06-01

    Achieving economic competitiveness as compared to LWRs and other Generation IV (Gen-IV) reactors is one of the major requirements for large-scale investment in commercial sodium cooled fast reactor (SFR) power plants. Advances in R&D for advanced SFR fuel and structural materials provide key long-term opportunities to improve SFR economics. In addition, other new opportunities are emerging to further improve SFR economics. This paper provides an overview on potential ideas from the perspective of thermal hydraulics to improve SFR economics. These include a new hybrid loop-pool reactor design to further optimize economics, safety, and reliability of SFRs with more flexibility, a multiple reheat and intercooling helium Brayton cycle to improve plant thermal efficiency and reduce safety related overnight and operation costs, and modern multi-physics thermal analysis methods to reduce analysis uncertainties and associated requirements for over-conservatism in reactor design. This paper reviews advances in all three of these areas and their potential beneficial impacts on SFR economics.

  17. Assessment of SFR reactor safety issues: Part II: Analysis results of ULOF transients imposed on a variety of different innovative core designs with SAS-SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kruessmann, R., E-mail: regina.kruessmann@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Ponomarev, A.; Pfrang, W.; Struwe, D. [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Champigny, J.; Carluec, B. [AREVA, 10, rue J. Récamier, 69456 Lyon Cedex 06 (France); Schmitt, D.; Verwaerde, D. [EDF R& D, 1 avenue du général de Gaulle, 92140 Clamart (France)

    2015-04-15

    Highlights: • Comparison of different core designs for a sodium-cooled fast reactor. • Safety assessment with the code system SAS-SFR. • Unprotected Loss of Flow (ULOF) scenario. • Sodium boiling and core melting cannot be avoided. • A net negative Na void effect provides more grace time prior to local SA destruction. - Abstract: In the framework of cooperation agreements between KIT-INR and AREVA SAS NP as well as between KIT-INR and EDF R&D in the years 2008–2013, the evaluation of severe transient behavior in sodium-cooled fast reactors (SFRs) was investigated. In Part I of this contribution, the efficiency of newly conceived prevention and mitigation measures was investigated for unprotected loss-of-flow (ULOF), unprotected loss-of-heat-sink (ULOHS) and the unprotected transient-overpower (UTOP) transients. In this second part, consequence analyses were performed for the initiation phase of different unprotected loss-of-flow (ULOF) scenarios imposed on a variety of different core design options of SFRs. The code system SAS-SFR was used for this purpose. Results of analyses for cases postulating unavailability of prevention measures as shut-down systems, passive and/or active additional devices show that entering into an energetic power excursion as a consequence of the initiation phase of a ULOF cannot be avoided for those core designs with a cumulative void reactivity feedback larger than zero. However, even for core designs aiming at values of the void reactivity less than zero it is difficult to find system design characteristics which prevent the transient entering into partial core destruction. Further studies of the transient core and system behavior would require codes dedicated to specific aspects of transition phase analyses and of in-vessel material relocation analyses.

  18. Site investigation SFR. Rock type coding, overview geological mapping and identification of rock units and possible deformation zones in drill cores from the construction of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Petersson, Jesper (Vattenfall Power Consultant AB, Stockholm (Sweden)); Curtis, Philip; Bockgaard, Niclas (Golder Associates AB (Sweden)); Mattsson, Haakan (GeoVista AB, Luleaa (Sweden))

    2011-01-15

    This report presents the rock type coding, overview lithological mapping and identification of rock units and possible deformation zones in drill cores from 32 boreholes associated with the construction of SFR. This work can be seen as complementary to single-hole interpretations of other older SFR boreholes earlier reported in /Petersson and Andersson 2010/: KFR04, KFR08, KFR09, KFR13, KFR35, KFR36, KFR54, KFR55, KFR7A, KFR7B and KFR7C. Due to deficiencies in the available material, the necessary activities have deviated somewhat from the established methodologies used during the recent Forsmark site investigations for the final repository for spent nuclear fuel. The aim of the current work has been, wherever possible, to allow the incorporation of all relevant material from older boreholes in the ongoing SFR geological modelling work in spite of the deficiencies. The activities include: - Rock type coding of the original geological mapping according to the nomenclature used during the preceding Forsmark site investigation. As part of the Forsmark site investigation such rock type coding has already been performed on most of the old SFR boreholes if the original geological mapping results were available. This earlier work has been complemented by rock type coding on two further boreholes: KFR01 and KFR02. - Lithological overview mapping, including documentation of (1) rock types, (2) ductile and brittle-ductile deformation and (3) alteration for drill cores from eleven of the boreholes for which no original geological borehole mapping was available (KFR31, KFR32, KFR34, KFR37,KFR38, KFR51, KFR69, KFR70, KFR71, KFR72 and KFR89). - Identification of possible deformation zones and merging of similar rock types into rock units. This follows SKB's established criteria and methodology of the geological Single-hole interpretation (SHI) process wherever possible. Deviations from the standard SHI process are associated with the lack of data, for example BIPS images

  19. Site investigation SFR. Rock type coding, overview geological mapping and identification of rock units and possible deformation zones in drill cores from the construction of SFR

    International Nuclear Information System (INIS)

    Petersson, Jesper; Curtis, Philip; Bockgaard, Niclas; Mattsson, Haakan

    2011-01-01

    This report presents the rock type coding, overview lithological mapping and identification of rock units and possible deformation zones in drill cores from 32 boreholes associated with the construction of SFR. This work can be seen as complementary to single-hole interpretations of other older SFR boreholes earlier reported in /Petersson and Andersson 2010/: KFR04, KFR08, KFR09, KFR13, KFR35, KFR36, KFR54, KFR55, KFR7A, KFR7B and KFR7C. Due to deficiencies in the available material, the necessary activities have deviated somewhat from the established methodologies used during the recent Forsmark site investigations for the final repository for spent nuclear fuel. The aim of the current work has been, wherever possible, to allow the incorporation of all relevant material from older boreholes in the ongoing SFR geological modelling work in spite of the deficiencies. The activities include: - Rock type coding of the original geological mapping according to the nomenclature used during the preceding Forsmark site investigation. As part of the Forsmark site investigation such rock type coding has already been performed on most of the old SFR boreholes if the original geological mapping results were available. This earlier work has been complemented by rock type coding on two further boreholes: KFR01 and KFR02. - Lithological overview mapping, including documentation of (1) rock types, (2) ductile and brittle-ductile deformation and (3) alteration for drill cores from eleven of the boreholes for which no original geological borehole mapping was available (KFR31, KFR32, KFR34, KFR37,KFR38, KFR51, KFR69, KFR70, KFR71, KFR72 and KFR89). - Identification of possible deformation zones and merging of similar rock types into rock units. This follows SKB's established criteria and methodology of the geological Single-hole interpretation (SHI) process wherever possible. Deviations from the standard SHI process are associated with the lack of data, for example BIPS images, or a

  20. Site investigation SFR. Boremap mapping of core drilled borehole KFR106

    Energy Technology Data Exchange (ETDEWEB)

    Winell, Sofia (Geosigma AB (Sweden))

    2010-06-15

    This report presents the result from the Boremap mapping of the core drilled borehole KFR106, drilled from an islet ca 220 m southeast of the pier above SFR. The borehole has a length of 300.13 m, and a bearing and inclination of 195.1 deg and -69.9 deg, respectively. The purpose of the location and orientation of the borehole is to investigate the possible occurrence of gently dipping, water-bearing structures in the area. The geological mapping is based on simultaneous study of drill core and borehole image (BIPS). The two lowermost meters of the drill core was mapped in Boremap without access to complementary BIPS-image. The dominating rock type, which occupies 72% of KFR106, is fine- to medium-grained, metagranite granodiorite (rock code 101057), which is foliated with a medium to strong intensity. Pegmatite to pegmatitic granite (rock code 101061) is the second most common rock type and it occupies 16% of the mapped interval. It is also frequent as smaller rock occurrences (< 1 m) in other rock types throughout the borehole. Subordinate rock types are fine- to medium-grained granite (rock code 111058), felsic to intermediate meta volcanic rock (rock code 103076), fine- to medium-grained metagranitoid (rock code 101051) and amphibolite (rock code 102017). Totally 49% of the rock in KFR106 has been mapped as altered, where muscovitization and oxidation is the two most common. Additional shorter intervals of alterations are in decreasing order of abundance quartz dissolution, epidotization, argillization, albitization, chloritization, laumontization and carbonatization. A total number of 2801 fractures are registered in KFR106. Of these are 1059 open, 1742 sealed and 84 partly open. This result in the following fracture frequencies: 6.0 sealed fractures/m, 3.7 open fractures/m and 0.3 partly open fractures/m. In addition there are 5 narrow brecciated zones, and 20 sealed networks with a total length of 18 m. The most frequent fracture fillings in KFR106 are

  1. Acoustic displacement sensor for harsh environment: application to SFR core support plate monitoring

    International Nuclear Information System (INIS)

    PeRISSE, J.; MACe, J.R.; VOUAGNER, P.

    2013-06-01

    The need for instrumentation able to monitor internal parameters inside reactor vessels during plant operation is getting stronger. Internal mechanical structures important for safety are concerned: for example core support plate, fuel assemblies or primary pumps. Because of very harsh environmental conditions (high temperature, pressure and radiation) and maintenance requirements, sensors are generally located on the outer shell of the vessel with, for example, strain gages, accelerometers, eddy current or US sensors. Then, some complex signal processing calculations must be performed to address internal structure behavior or health analysis but with bias effects (transfer path analysis method for example). This study will show an original displacement sensor based on an acoustic wave guide that can measure small displacement of mechanical structures inside reactor vessels. The application selected in this case is the monitoring of the core support plate for a sodium fast reactor (SFR). The wave guide - a thin tube sealed with pressurized argon gas inside - is installed inside the liquid sodium vessel (temperature between 400 deg. C to 550 deg. C). One extremity is connected to the mechanical structure for control. It includes two acoustic reflectors; such reflectors are dedicated to a calibration procedure to estimate the acoustic wave velocity whatever the temperature profile along the wave guide (velocity is temperature dependent). The opposite extremity of the wave guide is located outside the vessel and includes an emission/reception acoustic transducer. Using acoustic pulse reflectometry method, a plane wave pressure signal propagates inside the tube and reflects from the extremity and acoustic reflectors. The pulse-echo signals are recorded and processed in the frequency domain. Signal processing is performed to estimate the time of flight of pulse reflections patterns along the acoustic path. Then, monitored structure displacement - i.e. movement of the

  2. Multi-objective and multi-physics optimization methodology for SFR core: application to CFV concept

    International Nuclear Information System (INIS)

    Fabbris, Olivier

    2014-01-01

    Nuclear reactor core design is a highly multidisciplinary task where neutronics, thermal-hydraulics, fuel thermo-mechanics and fuel cycle are involved. The problem is moreover multi-objective (several performances) and highly dimensional (several tens of design parameters).As the reference deterministic calculation codes for core characterization require important computing resources, the classical design method is not well suited to investigate and optimize new innovative core concepts. To cope with these difficulties, a new methodology has been developed in this thesis. Our work is based on the development and validation of simplified neutronics and thermal-hydraulics calculation schemes allowing the full characterization of Sodium-cooled Fast Reactor core regarding both neutronics performances and behavior during thermal hydraulic dimensioning transients.The developed methodology uses surrogate models (or meta-models) able to replace the neutronics and thermal-hydraulics calculation chain. Advanced mathematical methods for the design of experiment, building and validation of meta-models allows substituting this calculation chain by regression models with high prediction capabilities.The methodology is applied on a very large design space to a challenging core called CFV (French acronym for low void effect core) with a large gain on the sodium void effect. Global sensitivity analysis leads to identify the significant design parameters on the core design and its behavior during unprotected transient which can lead to severe accidents. Multi-objective optimizations lead to alternative core configurations with significantly improved performances. Validation results demonstrate the relevance of the methodology at the pre-design stage of a Sodium-cooled Fast Reactor core. (author) [fr

  3. Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

    International Nuclear Information System (INIS)

    Sun, K.; Chenu, A.; Mikityuk, K.; Krepel, J.; Chawla, R.

    2012-01-01

    The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

  4. Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Chenu, A. [Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Mikityuk, K.; Krepel, J. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland)

    2012-07-01

    The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

  5. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  6. Korean SFR development program and technical activities for improving economical competitiveness

    International Nuclear Information System (INIS)

    Yoo, Jaewoon

    2013-01-01

    Future Plan: • Construction cost evaluation of PGSFR and commercial SFR; – Component based capital cost evaluation of PGSFR is undergoing and will be completed by the first half of 2014; – Component cost is only based on the experience from that of LWR; • Cost Benefit Analysis of Future Nuclear Energy Mix; – With revised National Energy Plan (as of 2013); – Near-term: Benefit from LWR spent fuel recycling: - In Korean law, Share of Expense for spent fuel disposal is reserved as 0.4M$ per a LWR spent fuel assembly (as of 2003); – Long-term: Competitive power plant to LWR with self sustainable feature; • Revision of commercial SFR conceptual design; – Less constraint in material (fuel, cladding) irradiation experience; – More innovative features as long-term goal

  7. Potential improvements of supercritical recompression CO2 Brayton cycle by mixing other gases for power conversion system of a SFR

    International Nuclear Information System (INIS)

    Jeong, Woo Seok; Lee, Jeong Ik; Jeong, Yong Hoon

    2011-01-01

    Highlights: → S-CO 2 cycle could be enhanced by shifting the critical point of working fluids using gas mixture. → In-house cycle code was developed to analyze supercritical Brayton cycles with gas mixture. → Gas mixture candidates were selected through a screening process: CO 2 mixing with N 2 , O 2 , He, and Ar. → CO 2 -He binary mixture shows the highest cycle efficiency increase. → Lowering the critical temperature and critical pressure of the coolant has a positive effect on the total cycle efficiency. - Abstract: A sodium-cooled fast reactor (SFR) is one of the strongest candidates for the next generation nuclear reactor. However, the conventional design of a SFR concept with an indirect Rankine cycle is subjected to a possible sodium-water reaction. To prevent any hazards from sodium-water reaction, a SFR with the Brayton cycle using Supercritical Carbon dioxide (S-CO 2 ) as the working fluid can be an alternative approach to improve the current SFR design. However, the S-CO 2 Brayton cycle is more sensitive to the critical point of working fluids than other Brayton cycles. This is because compressor work is significantly decreased slightly above the critical point due to high density of CO 2 near the boundary between the supercritical state and the subcritical state. For this reason, the minimum temperature and pressure of cycle are just above the CO 2 critical point. In other words, the critical point acts as a limitation of the lowest operating condition of the cycle. In general, lowering the rejection temperature of a thermodynamic cycle can increase the efficiency. Therefore, changing the critical point of CO 2 can result in an improvement of the total cycle efficiency with the same cycle layout. A small amount of other gases can be added in order to change the critical point of CO 2 . The direction and range of the critical point variation of CO 2 depends on the mixed component and its amount. Several gases that show chemical stability with

  8. On the use of moderating material to enhance the feedback coefficients in SFR cores with high minor actinide content

    International Nuclear Information System (INIS)

    Merk, B.; Weiss, F. P.

    2012-01-01

    The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting sodium cooled fast reactor cores is described. The influence of the moderating material on the neutron spectrum, the power distribution, and the burnup distribution is shown. The consequences of the use of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation is analyzed and the transmutation efficiency is compared. The degradation of the feedback effects due to the insertion of minor actinides and the compensation by the use of moderating materials is discussed. (authors)

  9. On the use of moderating material to enhance the feedback coefficients in SFR cores with high minor actinide content

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B. [Helmholtz-Zentrum Dresden-Rossendorf, Institut fuer Sicherheitsforschung, Postfach 51 01 19, 01314 Dresden (Germany); Weiss, F. P. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH Forschungszentrum, Boltzmannstr. 14, 85748 Garching (Germany)

    2012-07-01

    The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting sodium cooled fast reactor cores is described. The influence of the moderating material on the neutron spectrum, the power distribution, and the burnup distribution is shown. The consequences of the use of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation is analyzed and the transmutation efficiency is compared. The degradation of the feedback effects due to the insertion of minor actinides and the compensation by the use of moderating materials is discussed. (authors)

  10. Improved core monitoring for improved plant operations

    International Nuclear Information System (INIS)

    Mueller, N.P.

    1987-01-01

    Westinghouse has recently installed a core on-line surveillance, monitoring and operations systems (COSMOS), which uses only currently available core and plant data to accurately reconstruct the core average axial and radial power distributions. This information is provided to the operator in an immediately usable, human-engineered format and is accumulated for use in application programs that provide improved core performance predictive tools and a data base for improved fuel management. Dynamic on-line real-time axial and radial core monitoring supports a variety of plant operations to provide a favorable cost/benefit ratio for such a system. Benefits include: (1) relaxation or elimination of certain technical specifications to reduce surveillance and reporting requirements and allow higher availability factors, (2) improved information displays, predictive tools, and control strategies to support more efficient core control and reduce effluent production, and (3) expanded burnup data base for improved fuel management. Such systems can be backfit into operating plants without changing the existing instrumentation and control system and can frequently be implemented on existing plant computer capacity

  11. SFR 1 Vault Database

    International Nuclear Information System (INIS)

    Savage, David; Stenhouse, Mike

    2002-04-01

    SKB is carrying out a safety assessment of the operational SFR 1 repository under the auspices of the 'SAFE' (Safety Assessment of Final Repository for Radioactive Operational Waste) project. SKI in turn, is carrying out its own review of SFR 1. The work presented here is a compilation of physical and chemical data for the SFR 1 repository which will be used in radionuclide transport and assessment calculations by SKI. This compilation has focused on the repository itself (engineered barriers plus near-field rock). Data have been compiled for the following: Physical properties (porosity, hydraulic conductivity, bulk density, effective diffusivity); Sorption of radionuclides (on concrete, sand, bentonite, sand-bentonite, and rock); Radionuclide solubility. In addition, issues affecting gas generation at SFR I have been reviewed and placed in context with research conducted for the SFL 3-5 repository

  12. SFR site investigation. Bedrock Hydrogeochemistry

    International Nuclear Information System (INIS)

    Nilsson, Ann-Chatrin; Tullborg, Eva-Lena; Smellie, John; Gimeno, Maria J.; Gomez, Javier B.; Auque, Luis F.; Sandstroem, Bjoern; Pedersen, Karsten

    2011-11-01

    There are plans that the final repository for low and intermediate level radioactive waste, SFR, located about 150 km north of Stockholm, will be extended. Geoscientific studies to define and characterise a suitable bedrock volume for the extended repository have been carried out from 2007 to 2011, and have included the drilling and evaluation of seven new core drilled and four percussion boreholes. These new data, together with existing data extending back to 1985, have been interpreted and modelled in order to provide the necessary information for safety assessment and repository design. This report presents the final hydrogeochemical site description for the SFR site, and will constitute a background report for the integrated site description (the SFR Site Descriptive Model version 1.0) together with corresponding reports from the geological and hydrogeological disciplines. Most of the hydrogeochemical data from the field investigations consist of major ions and isotopes together with sporadic gas, microbe and measured redox data. Despite the close proximity of the Forsmark site, few data from this source are of relevance because of the shallow nature of the SFR site, the fact that SFR is located beneath the Baltic Sea and also the drawdown/upconing impacts of its construction on the hydrogeochemistry. This artificially imposed dynamic flow system is naturally more prevalent along major deformation fracture zones of higher transmissivity, whilst lower transmissive fractures together with the less transmissive bedrock masses between major deformation zones, still retain some evidence of the natural groundwater mixing patterns established prior to the SFR construction. The groundwaters in the SFR dataset cover a depth down to -250 m.a.s.l. with single sampling locations at -300 and -400 m.a.s.l. and represent a relatively limited salinity range (1,500 to 5,500 mg/L chloride). However, the δ 18 O values show a wide variation (-15.5 to -7.5 per mille V

  13. SFR site investigation. Bedrock Hydrogeochemistry

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Ann-Chatrin [Geosigma AB, Uppsala (Sweden); Tullborg, Eva-Lena [Terralogica AB, Graabo (Sweden); Smellie, John [Conterra AB, Uppsala (Sweden); Gimeno, Maria J.; Gomez, Javier B.; Auque, Luis F. [Univ. of Zaragoza, Zaragoza (Spain); Sandstroem, Bjoern [WSP Sverige AB, Goeteborg (Sweden); Pedersen, Karsten [Micans AB, Moelnlycke (Sweden)

    2011-11-15

    There are plans that the final repository for low and intermediate level radioactive waste, SFR, located about 150 km north of Stockholm, will be extended. Geoscientific studies to define and characterise a suitable bedrock volume for the extended repository have been carried out from 2007 to 2011, and have included the drilling and evaluation of seven new core drilled and four percussion boreholes. These new data, together with existing data extending back to 1985, have been interpreted and modelled in order to provide the necessary information for safety assessment and repository design. This report presents the final hydrogeochemical site description for the SFR site, and will constitute a background report for the integrated site description (the SFR Site Descriptive Model version 1.0) together with corresponding reports from the geological and hydrogeological disciplines. Most of the hydrogeochemical data from the field investigations consist of major ions and isotopes together with sporadic gas, microbe and measured redox data. Despite the close proximity of the Forsmark site, few data from this source are of relevance because of the shallow nature of the SFR site, the fact that SFR is located beneath the Baltic Sea and also the drawdown/upconing impacts of its construction on the hydrogeochemistry. This artificially imposed dynamic flow system is naturally more prevalent along major deformation fracture zones of higher transmissivity, whilst lower transmissive fractures together with the less transmissive bedrock masses between major deformation zones, still retain some evidence of the natural groundwater mixing patterns established prior to the SFR construction. The groundwaters in the SFR dataset cover a depth down to -250 m.a.s.l. with single sampling locations at -300 and -400 m.a.s.l. and represent a relatively limited salinity range (1,500 to 5,500 mg/L chloride). However, the {delta}{sup 18}O values show a wide variation (-15.5 to -7.5 per mille V

  14. Site investigation SFR. Bedrock geology

    International Nuclear Information System (INIS)

    Curtis, Philip; Markstroem, Ingemar; Petersson, Jesper; Triumf, Carl-Axel; Isaksson, Hans; Mattsson, Haakan

    2011-12-01

    geological tunnel mapping and eleven drill cores remapped according to the Boremap system, input to model version 1.0 has included the results from eight new cored boreholes as well as a fuller integration of Forsmark site investigation data, a further more extensive review of the drill core from an additional 32 boreholes associated with the construction of the existing SFR facility and an updated mapping of the lower construction tunnel. The current modelling work has also reviewed the older SFR data and models. While details concerning the earlier zones lying in immediate contact with the existing SFR facility have been changed, the earlier overall position, orientation and number of these deformation zones is maintained. A significant difference concerns their thickness due to the contrasting methodologies used during the different campaigns. In SFR model version 0.1, a single deformation zone model was produced, with a volume corresponding to the regional model volume. The model contained all the deformation zones modelled irrespective of size. Separate local and regional deformation zone models have been produced in SFR model version 1.0, following resolution criteria for the different model volumes. The local model contains zones with a minimum size of 300 m, while the regional model has structures that have a minimum size constraint of 1,000 m trace length at the ground surface. The selection of these size limits is related to the model volume maximum depth (local model -300 masl and regional model -1,000 masl) and the applied methodology that requires the same model resolution throughout the defined model volume (see Section 5.3.1). To assist hydrogeological modelling work, an updated combined model, including all structures from both the regional and local models, has also been delivered. The existing SFR facility and the rock volume directly to the south-east, which is proposed for the new facility extension, lies within a tectonic block that is bounded to the

  15. Site investigation SFR. Bedrock geology

    Energy Technology Data Exchange (ETDEWEB)

    Curtis, Philip; Markstroem, Ingemar (Golder Associates AB (Sweden)); Petersson, Jesper (Vattenfall Power Consultant AB (Sweden)); Triumf, Carl-Axel; Isaksson, Hans; Mattsson, Haakan (GeoVista AB (Sweden))

    2011-12-15

    the geological tunnel mapping and eleven drill cores remapped according to the Boremap system, input to model version 1.0 has included the results from eight new cored boreholes as well as a fuller integration of Forsmark site investigation data, a further more extensive review of the drill core from an additional 32 boreholes associated with the construction of the existing SFR facility and an updated mapping of the lower construction tunnel. The current modelling work has also reviewed the older SFR data and models. While details concerning the earlier zones lying in immediate contact with the existing SFR facility have been changed, the earlier overall position, orientation and number of these deformation zones is maintained. A significant difference concerns their thickness due to the contrasting methodologies used during the different campaigns. In SFR model version 0.1, a single deformation zone model was produced, with a volume corresponding to the regional model volume. The model contained all the deformation zones modelled irrespective of size. Separate local and regional deformation zone models have been produced in SFR model version 1.0, following resolution criteria for the different model volumes. The local model contains zones with a minimum size of 300 m, while the regional model has structures that have a minimum size constraint of 1,000 m trace length at the ground surface. The selection of these size limits is related to the model volume maximum depth (local model -300 masl and regional model -1,000 masl) and the applied methodology that requires the same model resolution throughout the defined model volume (see Section 5.3.1). To assist hydrogeological modelling work, an updated combined model, including all structures from both the regional and local models, has also been delivered. The existing SFR facility and the rock volume directly to the south-east, which is proposed for the new facility extension, lies within a tectonic block that is bounded

  16. Trade-off study on the power capacity of a prototype SFR in Korea

    International Nuclear Information System (INIS)

    Baek, M. H.; Kim, S. J.; Yoo, J.; Bae, I. H.

    2012-01-01

    The major roles of a prototype SFR are to provide irradiation test capability for the fuel and structure materials, and to obtain operational experiences of systems. Due to a compromise between the irradiation capability and construction costs, the power level should be properly determined. In this paper, a trade-off study on the power level of the prototype SFR was performed from a neutronics viewpoint. To select candidate cores, the parametric study of pin diameters was estimated using 20 wt.% uranium fuel. The candidate cores of different power levels, 125 MWt, 250 MWt, 400 MWt, and 500 MWt, were compared with the 1500 MWt reference core. The resulting core performance and economic efficiency indices became insensitive to the power at about 400-500 MWt and sharply deteriorated at about 125-250 MWt with decreasing core sizes. Fuel management scheme, TRU core performance comparing with uranium core, and sodium void reactivity were also evaluated with increasing power levels. It is found that increasing the number of batches showed higher burnup performance and economic efficiency. However, increasing the cycle length showed the trends in lower economic efficiency. Irradiation performance of TRU and enriched TRU cores was improved about 20 % and 50 %, respectively. The maximum sodium void reactivity of 5.2$ was confirmed less than the design limit of 7.5$. As a result, the power capacity of the prototype SFR should not be less than 250 MWt and would be appropriate at ∼ 500 MWt considering the performance and economic efficiency. (authors)

  17. Perspective on the audit calculation for SFR using TRACE code

    Energy Technology Data Exchange (ETDEWEB)

    Shin, An Dong; Choi, Yong Won; Bang, Young Suk; Bae, Moo Hoon; Huh, Byung Gil; Seol, Kwang One [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-10-15

    Korean Sodium Cooled Fast Reactor (SFR) is being developed by KAERI. The Prototype SFR will be a first SFR applied for licensing. KINS started research programs for preparing new concept design licensing recently. Safety analysis for the certain reactor is based on the computational estimation with conservatism and/or uncertainty of modeling. For the audit calculation for sodium cooled fast reactor (SFR), TRACE code is considered as one of analytical tool for SFR since TRACE code have already sodium related properties and models in it and have experience in the liquid metal coolant system area in abroad. Applicability of TRACE code for SFR is prechecked before real audit calculation. In this study, Demonstration Fast Reactor (DFR) 600 steady state conditions is simulated for identification of area of modeling improvements of TRACE code.

  18. SFR Safety Considerations

    International Nuclear Information System (INIS)

    Glatz, Jean-Paul

    2012-01-01

    Objectives of the Safety and Operation Project: • analysis and experiments that support approaches and assess performance of specific safety features, • development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and • valorisation of reactor operation, from experience and testing in operating SFR plants

  19. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K.

    2012-09-15

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  20. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    International Nuclear Information System (INIS)

    Sun, K.

    2012-09-01

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  1. Improvements to core-catchers

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, T C.W.

    1969-07-22

    A core catcher consists of a generally annular holder adapted to be contained within a core barrel with sets of dogs circumferentially disposed in the holder. Each set of dogs consists of at least 2 dogs of different lengths pivotally mounted in the holder to swing inward. The dogs in each set are vertically superimposed. They are of upward decreasing length, with the arc of swing of the vertically adjacent dogs overlapping. (8 claims)

  2. Generation-IV nuclear reactors, SFR concept

    International Nuclear Information System (INIS)

    Dufour, P.

    2010-01-01

    In this presentation author deals with development of sodium-cooled fast reactors and lead-cooled fast reactors. He concluded that: - SFR is a proved concept that has never achieved industrial deployment; - The GEN IV objectives need to reconsider the design of both the core and the reactor design : innovations are being analysed; Future design will benefit from considerable feedback of design, licensing, construction and operation of PX, SPX, etc.

  3. Nuclear data propagation with burnup. Impact on SFR reactivity coefficients

    International Nuclear Information System (INIS)

    Buiron, Laurent; Plisson-Rieunier, Daniele

    2017-01-01

    For the next generation fast reactor design, the Generation IV International Forum (GIF) defined global objectives in terms of safety improvement, sustainability, waste minimization and non-proliferation. Among the possibilities studied at CEA, Sodium cooled Fast Reactor (SFR) are studied as potential industrial tools for next decade's deployment. Many efforts have been made in the last years to obtain advanced industrial core designs that comply with these goals. Concerning safety issues, particular efforts have been made in order to obtain core designs that can be resilient to accidental transients. The 'safety' level of such new designs is often characterized by their 'natural' behavior under unprotected transients such as loss of flow or hypothetical transient over power. Transient analysis needs several accurate neutronic input data such as reactivity coefficient and kinetic parameters. Beside estimation of the level of 'absolute' values, associated uncertainties have also to be evaluated for the whole set of relevant data. These estimations have to be performed for different core state such as end of cycle core for feedback coefficient. This means that uncertainties have to be obtained not only a fixed time but also have to be propagated all through irradiation. To do so, we need to couple Boltzman and Bateman equations at sensitivities level. The coupling process could be done with the help of the perturbation theory which gives adapted framework suited for deterministic calculation codes. This coupling is currently in progress in ERANOS code system. The actual implementation gives access to estimation of sensitivities for both reactivity coefficients and mass balance. After a brief theoretical description of Boltzman/Bateman coupling capabilities in ERANOS, the study presented in this paper focuses on sensitivity and uncertainties estimation for the main feedback coefficients involved in fast reactor transients: the

  4. The nuclide inventory in SFR-1; Nuklidinventariet i SFR-1

    Energy Technology Data Exchange (ETDEWEB)

    Ingemansson, Tor [ALARA Engineering, Skultuna (Sweden)

    2001-10-01

    This report is an account for a project carried out on behalf of the Swedish Radiation Protection Authority (SSI): 'Nuclide inventory in SFR-1' (The Swedish underground disposal facility for low and intermediate level reactor waste). The project comprises the following five sub-projects: 1) Measuring methods for nuclides, difficult to measure, 2) The nuclide inventory in SFR-1, 3) Proposal for nuclide library for SFR-1 and ground disposal, 4) Nuclide library for exemption, and 5) Characterising of the nuclide inventory and documentation for SFL waste. In all five sub-projects long-lived activity, including Cl-36, has been considered.

  5. Improved core protection calculator system algorithm

    International Nuclear Information System (INIS)

    Yoon, Tae Young; Park, Young Ho; In, Wang Kee; Bae, Jong Sik; Baeg, Seung Yeob

    2009-01-01

    Core Protection Calculator System (CPCS) is a digitized core protection system which provides core protection functions based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels which adapted a two out of four trip logic. CPCS algorithm improvement for the newly designed core protection calculator system, RCOPS (Reactor COre Protection System), is described in this paper. New features include the improvement of DNBR algorithm for thermal margin, the addition of pre trip alarm generation for auxiliary trip function, VOPT (Variable Over Power Trip) prevention during RPCS (Reactor Power Cutback System) actuation and the improvement of CEA (Control Element Assembly) signal checking algorithm. To verify the improved CPCS algorithm, CPCS algorithm verification tests, 'Module Test' and 'Unit Test', would be performed on RCOPS single channel facility. It is expected that the improved CPCS algorithm will increase DNBR margin and enhance the plant availability by reducing unnecessary reactor trips

  6. Core-to-core uniformity improvement in multi-core fiber Bragg gratings

    Science.gov (United States)

    Lindley, Emma; Min, Seong-Sik; Leon-Saval, Sergio; Cvetojevic, Nick; Jovanovic, Nemanja; Bland-Hawthorn, Joss; Lawrence, Jon; Gris-Sanchez, Itandehui; Birks, Tim; Haynes, Roger; Haynes, Dionne

    2014-07-01

    Multi-core fiber Bragg gratings (MCFBGs) will be a valuable tool not only in communications but also various astronomical, sensing and industry applications. In this paper we address some of the technical challenges of fabricating effective multi-core gratings by simulating improvements to the writing method. These methods allow a system designed for inscribing single-core fibers to cope with MCFBG fabrication with only minor, passive changes to the writing process. Using a capillary tube that was polished on one side, the field entering the fiber was flattened which improved the coverage and uniformity of all cores.

  7. Uncertainty Evaluation of the SFR Subchannel Thermal-Hydraulic Modeling Using a Hot Channel Factors Analysis

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Cho, Chung Ho; Kim, Sang Ji

    2011-01-01

    In an SFR core analysis, a hot channel factors (HCF) method is most commonly used to evaluate uncertainty. It was employed to the early design such as the CRBRP and IFR. In other ways, the improved thermal design procedure (ITDP) is able to calculate the overall uncertainty based on the Root Sum Square technique and sensitivity analyses of each design parameters. The Monte Carlo method (MCM) is also employed to estimate the uncertainties. In this method, all the input uncertainties are randomly sampled according to their probability density functions and the resulting distribution for the output quantity is analyzed. Since an uncertainty analysis is basically calculated from the temperature distribution in a subassembly, the core thermal-hydraulic modeling greatly affects the resulting uncertainty. At KAERI, the SLTHEN and MATRA-LMR codes have been utilized to analyze the SFR core thermal-hydraulics. The SLTHEN (steady-state LMR core thermal hydraulics analysis code based on the ENERGY model) code is a modified version of the SUPERENERGY2 code, which conducts a multi-assembly, steady state calculation based on a simplified ENERGY model. The detailed subchannel analysis code MATRA-LMR (Multichannel Analyzer for Steady-State and Transients in Rod Arrays for Liquid Metal Reactors), an LMR version of MATRA, was also developed specifically for the SFR core thermal-hydraulic analysis. This paper describes comparative studies for core thermal-hydraulic models. The subchannel analysis and a hot channel factors based uncertainty evaluation system is established to estimate the core thermofluidic uncertainties using the MATRA-LMR code and the results are compared to those of the SLTHEN code

  8. Site investigation SFR. Overview Boremap mapping of drill cores from KFR04, KFR08, KFR09, KFR13, KFR35, KFR36, KFR54, KFR55, KFR7A, KFR7B and KFR7C

    International Nuclear Information System (INIS)

    Petersson, Jesper; Andersson, Ulf B.

    2011-01-01

    This report presents the results from a renewed geological overview mapping of 11 drill cores obtained during the construction of the final repository for low and middle level radioactive operational waste (SFR) during the 80's. Drill cores from KFR04, KFR08, KFR09, KFR13, KFR35, KFR36, KFR54, KFR55, KFR7A, KFR7B and KFR7C, with a total length of 837 m, was selected primarily because of their distinctly crosscutting relationship with inferred deformation zones in the area. The main purpose for this geological mapping is calibration with the original mappings, which in turn aims to facilitate geological single-hole interpretation. The mapping was generally focused on the location and infilling mineralogy of broken and unbroken fractures, as well as crush zones, breccias and sealed networks. Also the overview lithology, alterations and ductile shear zones were documented. All boreholes selected for renewed mapping are located in a ductile, high-strain belt, which defines the northeastern margin of a structurally more homogeneous tectonic lens. The main component of the high-strain belt is felsic to intermediate rocks of inferred volcanic origin. The predominant rock in the selected drill cores is, however, a fine- to finely medium-grained metagranite, which clearly appears to be a high-strain variety of the typically medium-grained metagranite-granodiorite that prevails the tectonic lens. It is obvious that varieties of this high-strain rock previously was inferred to be meta volcanic rocks. Other volumetrically important rock types in the drill cores are pegmatitic granite, finely medium-grained granite and metagranodiorite-tonalite, aplitic metagranite, amphibolites and slightly coarser metagabbros. Virtually all rocks in the borehole have experienced Svecofennian metamorphism under amphibolite facies conditions. Excluding fractures within crush zones and sealed networks, there is a predominance of broken fractures in most of the drill cores. The total fracture

  9. Site investigation SFR. Overview Boremap mapping of drill cores from KFR04, KFR08, KFR09, KFR13, KFR35, KFR36, KFR54, KFR55, KFR7A, KFR7B and KFR7C

    Energy Technology Data Exchange (ETDEWEB)

    Petersson, Jesper; Andersson, Ulf B. (Vattenfall Power Consultant AB, Stockholm (Sweden))

    2011-01-15

    This report presents the results from a renewed geological overview mapping of 11 drill cores obtained during the construction of the final repository for low and middle level radioactive operational waste (SFR) during the 80's. Drill cores from KFR04, KFR08, KFR09, KFR13, KFR35, KFR36, KFR54, KFR55, KFR7A, KFR7B and KFR7C, with a total length of 837 m, was selected primarily because of their distinctly crosscutting relationship with inferred deformation zones in the area. The main purpose for this geological mapping is calibration with the original mappings, which in turn aims to facilitate geological single-hole interpretation. The mapping was generally focused on the location and infilling mineralogy of broken and unbroken fractures, as well as crush zones, breccias and sealed networks. Also the overview lithology, alterations and ductile shear zones were documented. All boreholes selected for renewed mapping are located in a ductile, high-strain belt, which defines the northeastern margin of a structurally more homogeneous tectonic lens. The main component of the high-strain belt is felsic to intermediate rocks of inferred volcanic origin. The predominant rock in the selected drill cores is, however, a fine- to finely medium-grained metagranite, which clearly appears to be a high-strain variety of the typically medium-grained metagranite-granodiorite that prevails the tectonic lens. It is obvious that varieties of this high-strain rock previously was inferred to be meta volcanic rocks. Other volumetrically important rock types in the drill cores are pegmatitic granite, finely medium-grained granite and metagranodiorite-tonalite, aplitic metagranite, amphibolites and slightly coarser metagabbros. Virtually all rocks in the borehole have experienced Svecofennian metamorphism under amphibolite facies conditions. Excluding fractures within crush zones and sealed networks, there is a predominance of broken fractures in most of the drill cores. The total

  10. The progress and efficiency on SFR

    International Nuclear Information System (INIS)

    Li Shisen.

    1985-01-01

    The study comprehends an analysis of construction management at the work site of SFR, the final repository for low and medium level nuclear waste, situated in Forsmark. The period of analysis is 1985. During this year, the most intensive part of the rock excavation work of totaly 430,000 solid cubic metres took place. Many tunnels and big chambers as well as a huge silo were driven. Many drawings and figures are given to show how the project was going on and how the construction efficiency was during the year of 1985. SFR is a highly mechanized underground project. Based on the analysis of the composition of the construction costs in SFR, the work study points out that the machine costs is the biggest part in the construction costs, and that there is a close relation between the construction costs and machine management. Some experience in reducing construction costs and some experience in construction management are introduced in detail. In order to improve the machine management, attention should be paid to increase the utilization ratio of machines. A preliminary study of time utilization ratio of drill jumbo and bolting rig is given. (orig./HP)

  11. Development Perspective of Regulatory Audit Code System for SFR Nuclear Safety Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo Hoon; Lee, Gil Soo; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-05-15

    A sodium-cooled fast reactor (SFR) in Korea is based on the KALIMER-600 concept developed by KAERI. Based on 'Long-term R and D Plan for Future Reactor Systems' which was approved by the Korea Atomic Energy Commission in 2008, the KAERI designer is scheduled to apply the design certification of the prototype SFR in 2017. In order to establish regulatory infrastructure for the licensing of a prototype SFR, KINS has develop the regulatory requirements for the demonstration SFR since 2010, and are scheduled to develop the regulatory audit code systems in regard to core, fuel, and system, etc. since 2012. In this study, the domestic code systems used for core design and safety evaluation of PWRs and the nuclear physics and code system for SFRs were briefly reviewed, and the development perspective of regulatory audit code system for SFR nuclear safety evaluation were derived

  12. Project SAFE. Update of the SFR-1 safety assessment. Phase 1. Appendix A1: Inventory

    International Nuclear Information System (INIS)

    Riggare, P.

    1998-01-01

    One of the aims in the safety assessment of SFR-1 is to estimate the release to the environment. In order to make these calculations there is a need to describe the inventory in greater detail. The new computerised database of waste in SFR-1 gives a good possibility to achieve this. The aim for project SAFE is to make both conservative and realistic radionuclide transport calculations. To achieve this goal there must be two inventories. The conservative inventory is the inventory used in the design of the repository, which in most parts is identical with the limits in the licence for SFR-1. There is a great interest to have good estimates of the volumes of the different waste types. A thorough prognosis should be made in 1999, but until then the latest one from 1995 could be used in the calculations. The total (actual) inventory of nuclides is calculated from the measurements of the easy-to-measure nuclides since, in principle, all hard-to-measure nuclides are calculated by correlation factors to 60 Co and 137 Cs . These factors should be reviewed since there are quite large uncertainties involved. 14 C dominates the individual doses after a few hundred years and the collective dose in the inland-scenario. The amount of the nuclide is uncertain since the correlation factor is very uncertain. The chemical speciation of 14 C is also of interest due to different properties of organic and inorganic carbon. 36 Cl is very hard to measure. Although the authorities in their reviews of the safety reports say that there probably are small doses from chlorine, the inventory should be improved. 59 Ni is a long-lived nuclide that sets a limit to the close-to-the-core metal scrap that can be taken to SFR- 1. There is an ongoing research project to provide a better measuring method for 59 Ni. This should make it possible to improve the knowledge about 59 Ni inventory. The assumption that 90 % of the inventory is collected in the ion-exchange resins should be checked. Actinides

  13. Improving performance on core processes of care.

    Science.gov (United States)

    Austin, John Matthew; Pronovost, Peter J

    2016-06-01

    This article describes the recent literature on using extrinsic and intrinsic motivators to improve performance on core processes of care, highlighting literature that describes general frameworks for quality improvement work. The literature supporting the effectiveness of extrinsic motivators to improve quality is generally positive for public reporting of performance, with mixed results for pay-for-performance. A four-element quality improvement framework developed by The Armstrong Institute at Johns Hopkins Medicine was developed with intrinsic motivation in mind. The clear definition and communication of goals are important for quality improvement work. Training clinicians in improvement science, such as lean sigma, teamwork, or culture change provides clinicians with the skills they need to drive the improvement work. Peer learning communities offer the opportunity for clinicians to engage with each other and offer support in their work. The transparent reporting of performance helps ensure accountability of performance ranging from individual clinicians to governance. Quality improvement work that is led by and engages clinicians offers the opportunity for the work to be both meaningful and sustainable. The literature supports approaching quality improvement work in a systematic way, including the key elements of communication, infrastructure building, training, transparency, and accountability.

  14. Improvements in EBR-2 core depletion calculations

    International Nuclear Information System (INIS)

    Finck, P.J.; Hill, R.N.; Sakamoto, S.

    1991-01-01

    The need for accurate core depletion calculations in Experimental Breeder Reactor No. 2 (EBR-2) is discussed. Because of the unique physics characteristics of EBR-2, it is difficult to obtain accurate and computationally efficient multigroup flux predictions. This paper describes the effect of various conventional and higher order schemes for group constant generation and for flux computations; results indicate that higher-order methods are required, particularly in the outer regions (i.e. the radial blanket). A methodology based on Nodal Equivalence Theory (N.E.T.) is developed which allows retention of the accuracy of a higher order solution with the computational efficiency of a few group nodal diffusion solution. The application of this methodology to three-dimensional EBR-2 flux predictions is demonstrated; this improved methodology allows accurate core depletion calculations at reasonable cost. 13 refs., 4 figs., 3 tabs

  15. Bedrock Hydrogeology - Site investigation SFR

    Energy Technology Data Exchange (ETDEWEB)

    Oehman, Johan [Geosigma AB, Stockholm (Sweden); Bockgaard, Niclas [Golder Assoes AB, Stockholm (Sweden); Follin, Sven [SF GeoLogic AB, Taeby (Sweden)

    2012-06-15

    The Swedish Nuclear Fuel and Waste Management Company (SKB) has conducted site investigations for a planned extension of the existing final repository for short-lived radioactive waste (SFR). This report presents an integrated analysis and interpretation of the historic data from the existing SFR (1980 - 1986), as well as, from the recent investigations for the planned extension of SFR (2008 - 2009). The primary objective is to establish a conceptual hydrogeological model of the bedrock for safety assessment and design analyses. Analyses and interpretations of all (old and new) hydraulic data are analysed with regard to the recently developed geological deformation zone model of the SFR model domain (Curtis et al. 2011). The methodology used by Curtis et al. (2011) has focussed on magnetic anomalies and deformation zone intercepts with ground surface greater than 300 m. In the hydrogeological modelling, however, it has been considered important to also explore the occurrence and characteristics of shallow horizontal to sub-horizontal structures (sheet joints) inside the SFR model domain. Such structures are of considerable importance for the hydrogeology in the uppermost c. 150 m of bedrock in SDM-Site Forsmark; hence the term Shallow Bedrock Aquifer was used to emphasise their hydraulic significance. In this study, the acronym SBA-structure is used for horizontal structures identified in the hydrogeological modelling. In addition to the predominantly steeply dipping geological deformation zones, eight so-called SBA-structures are modelled deterministically in the hydrogeological model. The SBA-structures are envisaged as hydraulically heterogeneous and composed of clusters of minor gently dipping to horizontal fractures rather than extensive single features. A type of structures that is partly included in the definition of the SBA-structures is the Unresolved Possible Deformations Zone (Unresolved PDZ) intercepts identified by Curtis et al. (2011). The Unresolved

  16. Core skills assessment to improve mathematical competency

    Science.gov (United States)

    Carr, Michael; Bowe, Brian; Fhloinn, Eabhnat Ní

    2013-12-01

    Many engineering undergraduates begin third-level education with significant deficiencies in their core mathematical skills. Every year, in the Dublin Institute of Technology, a diagnostic test is given to incoming first-year students, consistently revealing problems in basic mathematics. It is difficult to motivate students to address these problems; instead, they struggle through their degree, carrying a serious handicap of poor core mathematical skills, as confirmed by exploratory testing of final year students. In order to improve these skills, a pilot project was set up in which a 'module' in core mathematics was developed. The course material was basic, but 90% or higher was required to pass. Students were allowed to repeat this module throughout the year by completing an automated examination on WebCT populated by a question bank. Subsequent to the success of this pilot with third-year mechanical engineering students, the project was extended to five different engineering programmes, across three different year-groups. Full results and analysis of this project are presented, including responses to interviews carried out with a selection of the students involved.

  17. Characterization of a sodium-cooled fast reactor in an MHR-SFR synergy for TRU transmutation

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Kim, Yonghee; Venneri, Francesco

    2008-01-01

    In the task of destroying the light water reactor (LWR) transuranics (TRUs), we consider the concept of a synergistic combination of a deep-burn (DB) gas-cooled reactor followed by a sodium-cooled fast reactor (SFR), as an alternative way to the direct feeding of the LWR TRUs to the SFR. In the synergy concept, TRUs from LWR are first deeply incinerated in a graphite-moderated DB-MHR (modular helium reactor) and then the spent fuels of DB-MHR are recycled into the closed-cycle SFR. The DB-MHR core is 100% TRU-loaded and a deep-burning (50-65%) is achieved in a safe manner (as discussed in our previous work). In this analysis, the SFR fuel cycle is closed with a pyro-processing technology to minimize the waste stream to a final repository. Neutronic characteristics of the SFR core in the MHR-SFR synergy have been evaluated from the core physics point of view. Also, we have compared core characteristics of the synergy SFR with those of a stand-alone SFR transuranic burner. For a consistent comparison, the two SFRs are designed to have the same TRU consumption rate of ∼250 kg/GW EFPY that corresponds to the TRU discharge rate from three 600 MW DB-MHRs. The results of our work show that the synergy SFR, fed with TRUs from DB-MHR, has a much smaller burnup reactivity swing, a slightly greater delayed neutron fraction (both positive features) but also a higher sodium void worth and a less negative Doppler coefficients than the conventional SFR, fed with TRUs directly from the LWRs. In addition, several design measures have been considered to reduce the sodium void worth in the synergy SFR core

  18. Improving Core Strength to Prevent Injury

    Science.gov (United States)

    Oliver, Gretchen D.; Adams-Blair, Heather R.

    2010-01-01

    Regardless of the sport or skill, it is essential to have correct biomechanical positioning, or postural control, in order to maximize energy transfer. Correct postural control requires a strong, stable core. A strong and stable core allows one to transfer energy effectively as well as reduce undue stress. An unstable or weak core, on the other…

  19. Site investigation SFR. Hydrogeological modelling of SFR. Model version 0.2

    Energy Technology Data Exchange (ETDEWEB)

    Oehman, Johan (Golder Associates AB (Sweden)); Follin, Sven (SF GeoLogic (Sweden))

    2010-01-15

    The Swedish Nuclear Fuel and Waste Management Company (SKB) has conducted site investigations for a planned extension of the existing final repository for short-lived radioactive waste (SFR). A hydrogeological model is developed in three model versions, which will be used for safety assessment and design analyses. This report presents a data analysis of the currently available hydrogeological data from the ongoing Site Investigation SFR (KFR27, KFR101, KFR102A, KFR102B, KFR103, KFR104, and KFR105). The purpose of this work is to develop a preliminary hydrogeological Discrete Fracture Network model (hydro-DFN) parameterisation that can be applied in regional-scale modelling. During this work, the Geologic model had not yet been updated for the new data set. Therefore, all analyses were made to the rock mass outside Possible Deformation Zones, according to Single Hole Interpretation. Owing to this circumstance, it was decided not to perform a complete hydro-DFN calibration at this stage. Instead focus was re-directed to preparatory test cases and conceptual questions with the aim to provide a sound strategy for developing the hydrogeological model SFR v. 1.0. The presented preliminary hydro-DFN consists of five fracture sets and three depth domains. A statistical/geometrical approach (connectivity analysis /Follin et al. 2005/) was performed to estimate the size (i.e. fracture radius) distribution of fractures that are interpreted as Open in geologic mapping of core data. Transmissivity relations were established based on an assumption of a correlation between the size and evaluated specific capacity of geologic features coupled to inflows measured by the Posiva Flow Log device (PFL-f data). The preliminary hydro-DFN was applied in flow simulations in order to test its performance and to explore the role of PFL-f data. Several insights were gained and a few model technical issues were raised. These are summarised in Table 5-1

  20. Site investigation SFR. Hydrogeological modelling of SFR. Model version 0.2

    International Nuclear Information System (INIS)

    Oehman, Johan; Follin, Sven

    2010-01-01

    The Swedish Nuclear Fuel and Waste Management Company (SKB) has conducted site investigations for a planned extension of the existing final repository for short-lived radioactive waste (SFR). A hydrogeological model is developed in three model versions, which will be used for safety assessment and design analyses. This report presents a data analysis of the currently available hydrogeological data from the ongoing Site Investigation SFR (KFR27, KFR101, KFR102A, KFR102B, KFR103, KFR104, and KFR105). The purpose of this work is to develop a preliminary hydrogeological Discrete Fracture Network model (hydro-DFN) parameterisation that can be applied in regional-scale modelling. During this work, the Geologic model had not yet been updated for the new data set. Therefore, all analyses were made to the rock mass outside Possible Deformation Zones, according to Single Hole Interpretation. Owing to this circumstance, it was decided not to perform a complete hydro-DFN calibration at this stage. Instead focus was re-directed to preparatory test cases and conceptual questions with the aim to provide a sound strategy for developing the hydrogeological model SFR v. 1.0. The presented preliminary hydro-DFN consists of five fracture sets and three depth domains. A statistical/geometrical approach (connectivity analysis /Follin et al. 2005/) was performed to estimate the size (i.e. fracture radius) distribution of fractures that are interpreted as Open in geologic mapping of core data. Transmissivity relations were established based on an assumption of a correlation between the size and evaluated specific capacity of geologic features coupled to inflows measured by the Posiva Flow Log device (PFL-f data). The preliminary hydro-DFN was applied in flow simulations in order to test its performance and to explore the role of PFL-f data. Several insights were gained and a few model technical issues were raised. These are summarised in Table 5-1

  1. A comparison of in-vessel behaviors between SFR and PWR under severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sanggil; Cho, Cheon Hwey [ACT Co., Daejeon (Korea, Republic of); Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This paper aims to provide an easy guide for experts who know well the severe accident phenomenology of Pressurized Water Reactor (PWR) by comparing both reactor design concepts and in vessel behaviors under a postulated severe accident condition. This study only provides a preliminary qualitative comparison based on available literature. The PWR and SFR in-vessel design concepts and their effects under a postulate severe accident are investigated in this paper. Although this work is a preliminary study to compare the in-vessel behaviors for both PWR and SFR, it seems that there is no possibility to lead a significant core damage in the metal fuel SFR concept. In the oxide fuel SFR, there might be a chance to progress to the severe accident initiators such as the energetic reaction, flow blockage and so on.

  2. Current status of SFR development in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Ieda, Yoshiaki; Chikazawa, Yoshitaka [Japan Atomic Energy Agency, Tokyo (Japan). Project Promotion Office; Kotake, Shoji [Japan Atomic Power Company, Tokyo (Japan)

    2012-03-15

    Fast Reactor development experiences and status in Japan are summarized. Even though international SFR circumstances were against in 1980s and 1990s, e.g. CRBRP, SNR-300 and Superphenix terminations, we kept on with our R and D activities steadily aiming at positive development targets in Japan. As results of our efforts, it has shown that our commercialized SFR concept, Japan Sodium-cooled Fast Reactor (JSFR) could meet the targets in the Feasibility Study on Commercialized Fast Reactor Cycle Systems (FS) and the Fast Reactor Cycle Technology Development (FaCT) project. Further, Monju has finally achieved restart in May 2010 after having been shut for almost 15 years. A future plan of Monju is to be determined based on a direction of the national nuclear and energy policies that will be established in 2012. The undergoing FaCT project is pursuing commercialization of fast reactor cycle system around 2050 under cooperation of MEXT (Ministry of Education, Culture, Sports, Science and Technology), METI (Ministry of Economy, Trade and Industry), utilities, venders and JAEA (Japan Atomic Energy Agency). As results of the FaCT Phase I, feasibility of the key technologies for JSFR has been evaluated and the project is waiting for launching the phase II due to the Tohoku large earthquake. It is considered that the nuclear development policy might be affected by the Tohoku large Earthquake/Tsunami in Japan. Nevertheless the significance of nuclear energy will not be changed and thus we will focus on the issues learnt from Fukushima accidents and reflect into the improvement of the safety of Monju and the safety design criteria for the next generation Fast Reactor systems. (orig.)

  3. Current status of SFR development in Japan

    International Nuclear Information System (INIS)

    Ieda, Yoshiaki; Chikazawa, Yoshitaka

    2012-01-01

    Fast Reactor development experiences and status in Japan are summarized. Even though international SFR circumstances were against in 1980s and 1990s, e.g. CRBRP, SNR-300 and Superphenix terminations, we kept on with our R and D activities steadily aiming at positive development targets in Japan. As results of our efforts, it has shown that our commercialized SFR concept, Japan Sodium-cooled Fast Reactor (JSFR) could meet the targets in the Feasibility Study on Commercialized Fast Reactor Cycle Systems (FS) and the Fast Reactor Cycle Technology Development (FaCT) project. Further, Monju has finally achieved restart in May 2010 after having been shut for almost 15 years. A future plan of Monju is to be determined based on a direction of the national nuclear and energy policies that will be established in 2012. The undergoing FaCT project is pursuing commercialization of fast reactor cycle system around 2050 under cooperation of MEXT (Ministry of Education, Culture, Sports, Science and Technology), METI (Ministry of Economy, Trade and Industry), utilities, venders and JAEA (Japan Atomic Energy Agency). As results of the FaCT Phase I, feasibility of the key technologies for JSFR has been evaluated and the project is waiting for launching the phase II due to the Tohoku large earthquake. It is considered that the nuclear development policy might be affected by the Tohoku large Earthquake/Tsunami in Japan. Nevertheless the significance of nuclear energy will not be changed and thus we will focus on the issues learnt from Fukushima accidents and reflect into the improvement of the safety of Monju and the safety design criteria for the next generation Fast Reactor systems. (orig.)

  4. Preliminary Hydrogeochemical Site Description SFR (version 0.2)

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Ann-Chatrin (Geosigma AB, Uppaala (Sweden)); Tullborg, Eva-Lena (Terralogica AB, Graabo (Sweden)); Smellie, John (Conterra AB, Partille (Sweden))

    2010-05-15

    The final repository for low and intermediate level radioactive operational waste, SFR, located about 150 km north of Stockholm, is to undergo a future extension. The present on-going project, scheduled from 2007 to 2011, is to define and characterise a suitable bedrock volume for the extended repository. This will include the drilling and geoscientific evaluation of seven core-drilled and four percussion boreholes as well as subsequent interpretation and modelling based on the obtained results in order to provide the necessary information for safety assessment and repository design. This report presents a preliminary hydrogeochemical site description for the SFR site and should be considered as an early progress report rather than a complete hydrochemical site descriptive model. The completed hydrogeochemical field investigations have yielded chemical data from a total of 12 borehole sections in five boreholes and additional data from the entire length of two open boreholes in connection with hydraulic tests. These data, together with data from a total of 18 early boreholes in the present SFR tunnel system, were used in the interpretation work. The main part of the data consisted of basic groundwater analyses including major ions and isotopes. Some sporadic gas, microbe and measured redox data are available, but these are either not treated in this report, or are only briefly discussed. This was due to time constraints since special care is needed when interpreting few data of varying quality. The groundwaters in the SFR dataset cover a maximum depth down to about .400 masl and represent a relatively limited salinity range (1,500 to 5,500 mg/L chloride). However, the delta18O values show a wide variation (-1.55 to -0.75% V-SMOW) similar to that reported from the Forsmark site investigations. At the SFR, marine indicators such as Mg/Cl, K/Cl and Br/Cl also show relatively large variations considering the limited salinity range. From very few measured Eh values, and

  5. Preliminary Hydrogeochemical Site Description SFR (version 0.2)

    International Nuclear Information System (INIS)

    Nilsson, Ann-Chatrin; Tullborg, Eva-Lena; Smellie, John

    2010-05-01

    The final repository for low and intermediate level radioactive operational waste, SFR, located about 150 km north of Stockholm, is to undergo a future extension. The present on-going project, scheduled from 2007 to 2011, is to define and characterise a suitable bedrock volume for the extended repository. This will include the drilling and geoscientific evaluation of seven core-drilled and four percussion boreholes as well as subsequent interpretation and modelling based on the obtained results in order to provide the necessary information for safety assessment and repository design. This report presents a preliminary hydrogeochemical site description for the SFR site and should be considered as an early progress report rather than a complete hydrochemical site descriptive model. The completed hydrogeochemical field investigations have yielded chemical data from a total of 12 borehole sections in five boreholes and additional data from the entire length of two open boreholes in connection with hydraulic tests. These data, together with data from a total of 18 early boreholes in the present SFR tunnel system, were used in the interpretation work. The main part of the data consisted of basic groundwater analyses including major ions and isotopes. Some sporadic gas, microbe and measured redox data are available, but these are either not treated in this report, or are only briefly discussed. This was due to time constraints since special care is needed when interpreting few data of varying quality. The groundwaters in the SFR dataset cover a maximum depth down to about .400 masl and represent a relatively limited salinity range (1,500 to 5,500 mg/L chloride). However, the δ 18 O values show a wide variation (-1.55 to -0.75% V-SMOW) similar to that reported from the Forsmark site investigations. At the SFR, marine indicators such as Mg/Cl, K/Cl and Br/Cl also show relatively large variations considering the limited salinity range. From very few measured Eh values, and

  6. Improving Deterrence of Hard-Core Cartels

    OpenAIRE

    Mariana Tavares de Araujo

    2010-01-01

    Holding perpetrators accountable and tailoring the optimal mix of sanctions through a combination of administrative and criminal penalties are two core elements of Brazil’s anti-cartel enforcement. Mariana Tavares de Araujo (SDE, Brazil)

  7. Improved sealing for in-core systems

    International Nuclear Information System (INIS)

    Dunford, S.

    1989-01-01

    The in-core instrumentation sealing nozzles designed by Framatome have three mechanical seals in series instead of the one traditional seal, and are pressurized by simply tightening up the nozzle covers. They have been installed from the start on all Framatome PWRs, as well as having been backfitted on Belgium and Yugoslavian units and chosen for the Chinese Qinshan plant. (author)

  8. Risk-Informed Balancing Of Safety, Nonproliferation, And Economics For The SFR

    International Nuclear Information System (INIS)

    Apostolakis, George; Driscoll, Michael; Golay, Michael; Kadak, Andrew; Todreas, Neil; Aldmir, Tunc; Denning, Richard; Lineberry, Michael

    2011-01-01

    , particularly concerning seismic and aircraft impactrelated risks. Most importantly, within the context of the TNF historical SFR safety concerns about energetic core disruptive accidents are seen to be unimportant, but those of rare scenarios mentioned above are seen to be of dominant concern. In terms of proliferation risks the SFR energy system is seen not to be of considerably greater concern than with other nuclear power technologies, providing that highly effective safeguards are employed. We find the economic performance of proposed SFRs likely, due to the problems of using sodium as a coolant, to be inferior to those of LWRs unless they can be credited for services to improve nuclear waste disposal, nuclear fuel utilization and proliferation risk reductions. None of the design innovations investigated offers the promise to reverse this conclusion. The most promising innovation investigated is that of improving the plant's thermodynamic efficiency via use of the supercritical CO 2 (rather than steam Rankine) power conversion system. We were unable to reach conclusions about the economic and proliferation risk implications of competing nuclear fuel processing methods, as available designs are too little developed to justify any such results. Overall, we find the SFR to be a promising alternative to LWRs should the conditions governing the valuation change substantially from current ones.

  9. RISK-INFORMED BALANCING OF SAFETY, NONPROLIFERATION, AND ECONOMICS FOR THE SFR

    Energy Technology Data Exchange (ETDEWEB)

    Apostolakis, George; Driscoll, Michael; Golay, Michael; Kadak, Andrew; Todreas, Neil; Aldmir, Tunc; Denning, Richard; Lineberry, Michael

    2011-10-20

    , particularly concerning seismic and aircraft impactrelated risks. Most importantly, within the context of the TNF historical SFR safety concerns about energetic core disruptive accidents are seen to be unimportant, but those of rare scenarios mentioned above are seen to be of dominant concern. In terms of proliferation risks the SFR energy system is seen not to be of considerably greater concern than with other nuclear power technologies, providing that highly effective safeguards are employed. We find the economic performance of proposed SFRs likely, due to the problems of using sodium as a coolant, to be inferior to those of LWRs unless they can be credited for services to improve nuclear waste disposal, nuclear fuel utilization and proliferation risk reductions. None of the design innovations investigated offers the promise to reverse this conclusion. The most promising innovation investigated is that of improving the plant's thermodynamic efficiency via use of the supercritical CO{sub 2} (rather than steam Rankine) power conversion system. We were unable to reach conclusions about the economic and proliferation risk implications of competing nuclear fuel processing methods, as available designs are too little developed to justify any such results. Overall, we find the SFR to be a promising alternative to LWRs should the conditions governing the valuation change substantially from current ones.

  10. The nuclide inventory in SFR-1

    International Nuclear Information System (INIS)

    Ingemansson, Tor

    2001-10-01

    This report is an account for a project carried out on behalf of the Swedish Radiation Protection Authority (SSI): 'Nuclide inventory in SFR-1' (The Swedish underground disposal facility for low and intermediate level reactor waste). The project comprises the following five sub-projects: 1) Measuring methods for nuclides, difficult to measure, 2) The nuclide inventory in SFR-1, 3) Proposal for nuclide library for SFR-1 and ground disposal, 4) Nuclide library for exemption, and 5) Characterising of the nuclide inventory and documentation for SFL waste. In all five sub-projects long-lived activity, including Cl-36, has been considered

  11. Fast reactor core concepts to improve transmutation efficiency

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

    2015-01-01

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate

  12. Metallic Reactor Fuel Fabrication for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of (1) fuel slug casting, (2) loading and fabrication of the fuel rods, and (3) fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.

  13. Rotary Mode Core Sample System availability improvement

    International Nuclear Information System (INIS)

    Jenkins, W.W.; Bennett, K.L.; Potter, J.D.; Cross, B.T.; Burkes, J.M.; Rogers, A.C.

    1995-01-01

    The Rotary Mode Core Sample System (RMCSS) is used to obtain stratified samples of the waste deposits in single-shell and double-shell waste tanks at the Hanford Site. The samples are used to characterize the waste in support of ongoing and future waste remediation efforts. Four sampling trucks have been developed to obtain these samples. Truck I was the first in operation and is currently being used to obtain samples where the push mode is appropriate (i.e., no rotation of drill). Truck 2 is similar to truck 1, except for added safety features, and is in operation to obtain samples using either a push mode or rotary drill mode. Trucks 3 and 4 are now being fabricated to be essentially identical to truck 2

  14. Improved core electron confinement on JET

    International Nuclear Information System (INIS)

    Litaudon, X.; Baranov, Y.; Voitsekhovitch, I.

    1999-01-01

    Formation of core regions with reduced electron transport is reported in regimes with current profile shaping at JET. The electron heat diffusivity (Χ c ) is reduced down to 0.5 m 2 /s in the region of low magnetic shear with an ICRH power of 1 MW with no indication of a threshold. In the high performance optimised shear regime, obtained in scenarios dominated by ion heating, internal transport barriers on the ion temperature profiles are simultaneously accompanied by a significant reduction of the electron heat diffusivity at two-third of the plasma radius. In this regime, recent results and measurements obtained with the new gas-box divertor configuration are reported together with their transport analyses. The results indicate that Χ c is reduced by one order of magnitude in a spatially localised region. (authors)

  15. Development of basic key technologies for Gen IV SFR

    International Nuclear Information System (INIS)

    Kim, Yeongil; Kim, Sungoh; Choi, Sukgi

    2012-04-01

    The advanced concepts, for the breakeven reactor(1,200MWe) and TRU burner(600MWe), were defined to satisfy the technology goals of Generation IV nuclear systems. Based on the advanced design concepts, a conceptual design of the demonstration SFR has been developed using the available licensing technology. The conceptual core design has been developed for the TRU burner in which an initial core is fueled with less than 20wt% enriched U235, and finally transformed to a self-recycled TRU core. The passive decay heat removal circuit ensuring reactor safety even in case of loss of emergency power has been developed and minimization of a reactor vessel and simplification of reactor internals have been conducted in the conceptual design. For development of advanced technologies, control logics for various power levels and the optimal design concept of heat exchanger applicable to supercritical CO 2 Brayton cycle as an energy conversion system was developed. A novel under-sodium waveguide sensor and a prototype under-sodium inspection system have been developed for under-sodium viewing of in-vessel structures submerged in an opaque liquid sodium. The fabrication technology of fuel slugs using the advanced fuel slug casting system was developed, and U-Zr alloy fuel rods were fabricated and examined. And a HT 9 cladding tube was manufactured using the developed cladding tube fabrication technology. For development of basic technologies, the cross section adjustment code ATCROSS and the MATRA-LMR code with HCFs have been developed to reduce core design uncertainties. The SIE ASME-NH computer program to evaluate high temperature structural design for 60 years design life, and the safety analysis code MARS-LMR with thermal-hydraulic and reactivity feedback models have been developed and validated. In addition, the sodium impurity measurement and control technology, the sodium water reaction event propagation model to predict the sodium leak propagation in a steam generator, and

  16. Improvement of core degradation model in ISAAC

    International Nuclear Information System (INIS)

    Kim, Dong Ha; Kim, See Darl; Park, Soo Yong

    2004-02-01

    If water inventory in the fuel channels depletes and fuel rods are exposed to steam after uncover in the pressure tube, the decay heat generated from fuel rods is transferred to the pressure tube and to the calandria tube by radiation, and finally to the moderator in the calandria tank by conduction. During this process, the cladding will be heated first and ballooned when the fuel gap internal pressure exceeds the primary system pressure. The pressure tube will be also ballooned and will touch the calandria tube, increasing heat transfer rate to the moderator. Although these situation is not desirable, the fuel channel is expected to maintain its integrity as long as the calandria tube is submerged in the moderator, because the decay heat could be removed to the moderator through radiation and conduction. Therefore, loss of coolant and moderator inside and outside the channel may cause severe core damage including horizontal fuel channel sagging and finally loss of channel integrity. The sagged channels contact with the channels located below and lose their heat transfer area to the moderator. As the accident goes further, the disintegrated fuel channels will be heated up and relocated onto the bottom of the calandria tank. If the temperature of these relocated materials is high enough to attack the calandria tank, the calandria tank would fail and molten material would contact with the calandria vault water. Steam explosion and/or rapid steam generation from this interaction may threaten containment integrity. Though a detailed model is required to simulate the severe accident at CANDU plants, complexity of phenomena itself and inner structures as well as lack of experimental data forces to choose a simple but reasonable model as the first step. ISAAC 1.0 was developed to model the basic physicochemical phenomena during the severe accident progression. At present, ISAAC 2.0 is being developed for accident management guide development and strategy evaluation. In

  17. WWER core pattern enhancement using adaptive improved harmony search

    Energy Technology Data Exchange (ETDEWEB)

    Nazari, T. [Nuclear Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of); Aghaie, M., E-mail: M_Aghaie@sbu.ac.ir [Nuclear Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of); Zolfaghari, A.; Minuchehr, A.; Norouzi, A. [Nuclear Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer The classical and improved harmony search algorithms are introduced. Black-Right-Pointing-Pointer The advantage of IHS is demonstrated in Shekel's Foxholes. Black-Right-Pointing-Pointer The CHS and IHS are compared with other Heuristic algorithms. Black-Right-Pointing-Pointer The adaptive improved harmony search is applied for two cases. Black-Right-Pointing-Pointer Two cases of WWER core are optimized in BOC FA pattern. - Abstract: The efficient operation and fuel management of PWRs are of utmost importance. Core performance analysis constitutes an essential phase in core fuel management optimization. Finding an optimum core arrangement for loading of fuel assemblies, FAs, in a nuclear core is a complex problem. In this paper, application of classical harmony search (HS) and adaptive improved harmony search (IHS) in loading pattern (LP) design, for pressurized water reactors, is described. In this analysis, finding the best core pattern, which attains maximum multiplication factor, k{sub eff}, by considering maximum allowable power picking factors (PPF) is the main objective. Therefore a HS based, LP optimization code is prepared and CITATION code which is a neutronic calculation code, applied to obtain effective multiplication factor, neutron fluxes and power density in desired cores. Using adaptive improved harmony search and neutronic code, generated LP optimization code, could be applicable for PWRs core with many numbers of FAs. In this work, at first step, HS and IHS efficiencies are compared with some other heuristic algorithms in Shekel's Foxholes problem and capability of the adaptive improved harmony search is demonstrated. Results show, efficient application of IHS. At second step, two WWER cases are studied and then IHS proffered improved core patterns with regard to mentioned objective functions.

  18. WWER core pattern enhancement using adaptive improved harmony search

    International Nuclear Information System (INIS)

    Nazari, T.; Aghaie, M.; Zolfaghari, A.; Minuchehr, A.; Norouzi, A.

    2013-01-01

    Highlights: ► The classical and improved harmony search algorithms are introduced. ► The advantage of IHS is demonstrated in Shekel's Foxholes. ► The CHS and IHS are compared with other Heuristic algorithms. ► The adaptive improved harmony search is applied for two cases. ► Two cases of WWER core are optimized in BOC FA pattern. - Abstract: The efficient operation and fuel management of PWRs are of utmost importance. Core performance analysis constitutes an essential phase in core fuel management optimization. Finding an optimum core arrangement for loading of fuel assemblies, FAs, in a nuclear core is a complex problem. In this paper, application of classical harmony search (HS) and adaptive improved harmony search (IHS) in loading pattern (LP) design, for pressurized water reactors, is described. In this analysis, finding the best core pattern, which attains maximum multiplication factor, k eff , by considering maximum allowable power picking factors (PPF) is the main objective. Therefore a HS based, LP optimization code is prepared and CITATION code which is a neutronic calculation code, applied to obtain effective multiplication factor, neutron fluxes and power density in desired cores. Using adaptive improved harmony search and neutronic code, generated LP optimization code, could be applicable for PWRs core with many numbers of FAs. In this work, at first step, HS and IHS efficiencies are compared with some other heuristic algorithms in Shekel's Foxholes problem and capability of the adaptive improved harmony search is demonstrated. Results show, efficient application of IHS. At second step, two WWER cases are studied and then IHS proffered improved core patterns with regard to mentioned objective functions.

  19. Dose assessments for SFR 1

    International Nuclear Information System (INIS)

    Bergstroem, Ulla; Avila, Rodolfo; Ekstroem, Per-Anders; Cruz, Idalmis de la

    2008-05-01

    Following a review by the Swedish regulatory authorities of the safety analysis of the SFR 1 disposal facility for low and intermediate level waste, SKB has prepared an updated safety analysis, SAR-08. This report presents estimations of annual doses to the most exposed groups from potential radionuclide releases from the SFR 1 repository for a number of calculation cases, selected using a systematic approach for identifying relevant scenarios for the safety analysis. The dose estimates can be used for demonstrating that the long term safety of the repository is in compliance with the regulatory requirements. In particular, the mean values of the annual doses can be used to estimate the expected risks to the most exposed individuals, which can then be compared with the regulatory risk criteria for human health. The conversion from doses to risks is performed in the main report. For one scenario however, where the effects of an earthquake taking place close to the repository are analysed, risk calculations are presented in this report. In addition, prediction of concentrations of radionuclides in environmental media, such as water and soil, are compared with concentration limits suggested by the Erica-project as a base for estimating potential effects on the environment. The assessment of the impact on non-human biota showed that the potential impact is negligible. Committed collective dose for an integration period of 10,000 years for releases occurring during the first thousand years after closure are also calculated. The collective dose commitment was estimated to be 8 manSv. The dose calculations were carried out for a period of 100,000 years, which was sufficient to observe peak doses in all scenarios considered. Releases to the landscape and to a well were considered. The peaks of the mean annual doses from releases to the landscape are associated with C-14 releases to a future lake around year 5,000 AD. In the case of releases to a well, the peak annual doses

  20. Dose assessments for SFR 1

    Energy Technology Data Exchange (ETDEWEB)

    Bergstroem, Ulla (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)); Avila, Rodolfo; Ekstroem, Per-Anders; Cruz, Idalmis de la (Facilia AB, Bromma (Sweden))

    2008-06-15

    Following a review by the Swedish regulatory authorities of the safety analysis of the SFR 1 disposal facility for low and intermediate level waste, SKB has prepared an updated safety analysis, SAR-08. This report presents estimations of annual doses to the most exposed groups from potential radionuclide releases from the SFR 1 repository for a number of calculation cases, selected using a systematic approach for identifying relevant scenarios for the safety analysis. The dose estimates can be used for demonstrating that the long term safety of the repository is in compliance with the regulatory requirements. In particular, the mean values of the annual doses can be used to estimate the expected risks to the most exposed individuals, which can then be compared with the regulatory risk criteria for human health. The conversion from doses to risks is performed in the main report. For one scenario however, where the effects of an earthquake taking place close to the repository are analysed, risk calculations are presented in this report. In addition, prediction of concentrations of radionuclides in environmental media, such as water and soil, are compared with concentration limits suggested by the Erica-project as a base for estimating potential effects on the environment. The assessment of the impact on non-human biota showed that the potential impact is negligible. Committed collective dose for an integration period of 10,000 years for releases occurring during the first thousand years after closure are also calculated. The collective dose commitment was estimated to be 8 manSv. The dose calculations were carried out for a period of 100,000 years, which was sufficient to observe peak doses in all scenarios considered. Releases to the landscape and to a well were considered. The peaks of the mean annual doses from releases to the landscape are associated with C-14 releases to a future lake around year 5,000 AD. In the case of releases to a well, the peak annual doses

  1. Improving core surgical training in a major trauma centre.

    Science.gov (United States)

    Morris, Daniel L J; Bryson, David J; Ollivere, Ben J; Forward, Daren P

    2016-06-01

    English Major Trauma Centres (MTCs) were established in April 2012. Increased case volume and complexity has influenced trauma and orthopaedic (T&O) core surgical training in these centres. To determine if T&O core surgical training in MTCs meets Joint Committee on Surgical Training (JCST) quality indicators including performance of T&O operative procedures and consultant supervised session attendance. An audit cycle assessing the impact of a weekly departmental core surgical trainee rota. The rota included allocated timetabled sessions that optimised clinical and surgical learning opportunities. Intercollegiate Surgical Curriculum Programme (ISCP) records for T&O core surgical trainees at a single MTC were analysed for 8 months pre and post rota introduction. Outcome measures were electronic surgical logbook evidence of leading T&O operative procedures and consultant validated work-based assessments (WBAs). Nine core surgical trainees completed a 4 month MTC placement pre and post introduction of the core surgical trainee rota. Introduction of core surgical trainee rota significantly increased the mean number of T&O operative procedures led by a core surgical trainee during a 4 month MTC placement from 20.2 to 34.0 (pcore surgical trainee during a 4 month MTC placement was significantly increased (0.3 vs 2.4 [p=0.04]). Those of dynamic hip screw fixation (2.3 vs 3.6) and ankle fracture fixation (0.7 vs 1.6) were not. Introduction of a core surgical trainee rota significantly increased the mean number of consultant validated WBAs completed by a core surgical trainee during a 4 month MTC placement from 1.7 to 6.6 (pcore surgical trainee rota utilising a 'problem-based' model can significantly improve T&O core surgical training in MTCs. Copyright © 2016 Elsevier Ltd. All rights reserved.

  2. Improvement of SSR core design for ABWR-II

    International Nuclear Information System (INIS)

    Moriwaki, Masanao; Aoyama, Motoo; Okada, Hiroyuki; Kitamura, Hideya; Sakurada, Koichi; Tanabe, Akira

    2003-01-01

    In order to enhance the spectral shift effect in the ABWR-II reactor, a novel core design to bring out better performance of spectral shift rods (SSRs) is studied. The SSR is a new type of water rod, in which the water level develops naturally during operation and changes according to the coolant flow rate through the channel. By using the SSR, the average moderator density, which is directly related to core reactivity, can be controlled over a wide range by the core flow rate. In the new SSR core design, two types of SSR bundles, in which settings for the SSR water levels are different, are utilized and loaded according to flow distribution in the core. This two-region SSR core design allows wide variation in the average SSR water level, thus improving fuel economy. Enhancement of SSR function in the two-region SSR core increases the uranium saving factor by about 25%, from the 6% of the conventional uniform SSR core to about 8%. (author)

  3. Improving work control systems: The core team concept

    International Nuclear Information System (INIS)

    Jorgensen, M.D.; Simpson, W.W.

    1996-01-01

    The improved work control system at the Idaho Chemical Processing Plant minimizes review and approval time, maximizes field work time, and maintains full compliance with applicable requirements. The core team method gives ownership and accountability to knowledgeable individuals, and the teams use sophisticated scheduling techniques to improve information sharing and cost control and to establish accurate roll-up master schedules

  4. Improving the calculated core stability by the core nuclear design optimization

    International Nuclear Information System (INIS)

    Partanen, P.

    1995-01-01

    Three different equilibrium core loadings for TVO II reactor have been generated in order to improve the core stability properties at uprated power level. The reactor thermal power is assumed to be uprated from 2160 MW th to 2500 MW th , which moves the operating point after a rapid pump rundown where the core stability has been calculated from 1340 MW th and 3200 kg/s to 1675 MW th and 4000 kg/s. The core has been refuelled with ABB Atom Svea-100 -fuel, which has 3,64% w/o U-235 average enrichment in the highly enriched zone. PHOENIX lattice code has been used to provide the homogenized nuclear constants. POLCA4 static core simulator has been used for core loadings and cycle simulations and RAMONA-3B program for simulating the dynamic response to the disturbance for which the stability behaviour has been evaluated. The core decay ratio has been successfully reduced from 0,83 to 0,55 mainly by reducing the power peaking factors. (orig.) (7 figs., 1 tab.)

  5. Qinshan NPP in-core fuel management improvement

    International Nuclear Information System (INIS)

    Kong Deping; Liao Zejun; Wu Xifeng; Wei Wenbin; Wang Yongming; Li Hua

    2006-01-01

    In the 10-year operation of Qinshan Nuclear Power Plant, the initial designed reloading strategy has been improved step by step based on the operation experiences and the advanced domestic and international fuel management methods. Higher burnup has been achieved and more economic operation gained through the loading pattern improvement and the fuel enrichment increased. The article introduces the in-core fuel management strategy improvement of Qinshan Nuclear Power Plant in its 10-year operation. (authors)

  6. Status of SFR Metal Fuel Development

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Lee, Byoung Oon; Kim, Ki Hwan; Kim, Sung Ho

    2013-01-01

    Conclusion: • Metal fuel recycling in SFR: - Enhanced utilization of uranium resource; - Efficient transmutation of minor actinides; - Inherent passive reactor safety; - Proliferation resistance with pyro-electrochemical fuel recycling. • Demonstration of technical feasibility of recycling TRU metal fuel by 2020: - Remote fuel fabrication; - Irradiation performance up to high burnup

  7. Characterisation of bitumenised waste in SFR 1

    International Nuclear Information System (INIS)

    Pettersson, Michael; Elert, M.

    2001-06-01

    The waste deposited in the Final Repository for Radioactive Operational Waste, SFR, consists in part of waste solidified in bitumen. Bitumen is considered to have favourable chemical and physical properties to act as a fixation material for radioactive waste. However, during interim storage and subsequent disposal bitumen's properties may change. This may influence the stability of the bitumen matrix to retain radionuclides. This report discusses different processes affecting the long-term performance of bitumenised waste, and an evaluation of these properties in waste deposited in SFR 1 is made. The possible effect of a bitumen barrier on the release rate of radionuclides from SFR 1 is assessed. Based on leaching experiments reviewed in this study, it could take some thousand years, possibly more, to release all radionuclides in a 200-litre drum. The results are, however, extrapolated from experiments performed during a short period of time. Long- term deteriorating effects and the effect of a low temperature on the bitumen matrix are not very well documented. The literature focuses principally on bitumenised evaporator concentrate, but the bitumenised waste deposited in SFR 1 consists mainly of ion exchange resins. There are indications that the non-radioactive waste products usually investigated overestimate bitumen's ability to retain waste. Radiolytic effects has been estimated in this work to be negligible for waste categories F.17, F.20 and B.20 deposited in SFR 1, but for categories B.05, B.06 and F.18 the possibility of increased water uptake rate due to radiolysis can not be excluded. A more reasonable assumption is that bitumen will act as an effective barrier for radionuclide release during a time span from some hundreds to thousand of years. Generally, the majority of the inventory of radionuclides in SFR 1 is not solidified in bitumen. By taking the bitumen barrier into account in the modelling of release of radio- nuclides from SFR 1, the total

  8. Characterisation of bitumenised waste in SFR 1

    Energy Technology Data Exchange (ETDEWEB)

    Pettersson, Michael; Elert, M. [Kemakta Konsult AB, Stockholm (Sweden)

    2001-06-01

    The waste deposited in the Final Repository for Radioactive Operational Waste, SFR, consists in part of waste solidified in bitumen. Bitumen is considered to have favourable chemical and physical properties to act as a fixation material for radioactive waste. However, during interim storage and subsequent disposal bitumen's properties may change. This may influence the stability of the bitumen matrix to retain radionuclides. This report discusses different processes affecting the long-term performance of bitumenised waste, and an evaluation of these properties in waste deposited in SFR 1 is made. The possible effect of a bitumen barrier on the release rate of radionuclides from SFR 1 is assessed. Based on leaching experiments reviewed in this study, it could take some thousand years, possibly more, to release all radionuclides in a 200-litre drum. The results are, however, extrapolated from experiments performed during a short period of time. Long- term deteriorating effects and the effect of a low temperature on the bitumen matrix are not very well documented. The literature focuses principally on bitumenised evaporator concentrate, but the bitumenised waste deposited in SFR 1 consists mainly of ion exchange resins. There are indications that the non-radioactive waste products usually investigated overestimate bitumen's ability to retain waste. Radiolytic effects has been estimated in this work to be negligible for waste categories F.17, F.20 and B.20 deposited in SFR 1, but for categories B.05, B.06 and F.18 the possibility of increased water uptake rate due to radiolysis can not be excluded. A more reasonable assumption is that bitumen will act as an effective barrier for radionuclide release during a time span from some hundreds to thousand of years. Generally, the majority of the inventory of radionuclides in SFR 1 is not solidified in bitumen. By taking the bitumen barrier into account in the modelling of release of radio- nuclides from SFR 1, the

  9. Thermal hydraulic analysis of the JMTR improved LEU-core

    Energy Technology Data Exchange (ETDEWEB)

    Tabata, Toshio; Nagao, Yoshiharu; Komukai, Bunsaku; Naka, Michihiro; Fujiki, Kazuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takeda, Takashi [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokai, Ibaraki (Japan)

    2003-01-01

    After the investigation of the new core arrangement for the JMTR reactor in order to enhance the fuel burn-up and consequently extend the operation period, the ''improved LEU core'' that utilized 2 additional fuel elements instead of formerly installed reflector elements, was adopted. This report describes the results of the thermal-hydraulic analysis of the improved LEU core as a part of safety analysis for the licensing. The analysis covers steady state, abnormal operational transients and accidents, which were described in the annexes of the licensing documents as design bases events. Calculation conditions for the computer codes were conservatively determined based on the neutronic analysis results and others. The results of the analysis, that revealed the safety criteria were satisfied on the fuel temperature, DNBR and primary coolant temperature, were used in the licensing. The operation license of the JMTR with the improved LEU core was granted in March 2001, and the reactor operation with new core started in November 2001 as 142nd operation cycle. (author)

  10. Innovative power conversion system for the French SFR prototype, ASTRID

    International Nuclear Information System (INIS)

    Cachon, L.; Biscarrat, C.; Morin, F.; Haubensack, D.; Rigal, E.; Moro, I.; Baque, F.; Madeleine, S.; Rodriguez, G.; Laffont, G.

    2012-01-01

    In the framework of the French Act of 28 June 2006 about nuclear materials and waste management, the prototype ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), foreseen in operation by the 20's, will have to demonstrate not only the minor actinide transmutation capability, but also the progress made in Sodium Fast Reactor (SFR) technology on an industrial scale, by qualifying innovative options. Some of these options still require improvements, especially in the field of operability and safety. In fact, one of the main issues with the standard steam/water Power Conversion System (PCS) of SFR is the fast and energetic chemical reaction between water and sodium, which could occur in steam generators in case of tube failure. To manage the sodium/water reaction, one way consists in minimizing the impact of such event: hence studies are carried out on steam generator design, improvement of the physical knowledge of this phenomenon, development of numerical simulation to predict the reaction onset and consequences, and associated detection improvement. On the other hand, the other way consists in eliminating sodium/water reaction. In this frame, the CEA contribution to the feasibility evaluation of an alternative innovative PCS (replacing steam/water by 180 bar pressurised nitrogen) is focused on the following main topics: - The parametric study leading to nitrogen selection: the thermodynamic cycle efficiency optimisation on Brayton cycles is performed with several gases at different pressures. - The design of innovative compact heat exchangers for the gas loop: here the key points are the nuclear codification associated with inspection capability, the innovative welding process and the thermal-hydraulic and thermal-mechanic optimisations. After a general introduction of the ASTRID project, this paper presents in detail these different feasibility studies being led on the innovative gas PCS for an SFR. (authors)

  11. NPP Krsko core calculations to improve operational safety

    International Nuclear Information System (INIS)

    Ivekovic, I.; Grgic, D.; Nemec, T.

    2007-01-01

    Calculation tools and methodology used to perform independent calculations of cumulative influence of different changes related to fuel and core operation of NPP Krsko were described. Some examples of steady state and transient results are used to illustrate potential improvements to understanding and reviewing plant safety. (author)

  12. Digital imaging improves upright stereotactic core biopsy of mammographic microcalcifications

    International Nuclear Information System (INIS)

    Whitlock, J.P.L.; Evans, A.J.; Burrell, H.C.; Pinder, S.E.; Ellis, I.O.; Blamey, R.W.; Wilson, A.R.M.

    2000-01-01

    AIM: This comparative study was carried out to assess the effect of using digital images compared to conventional film-screen mammography on the accuracy of core biopsy of microcalcifications using upright stereotactic equipment. MATERIALS AND METHODS: The biopsy results from a consecutive series of 104 upright stereotactic 14-gauge core biopsies performed with conventional X-ray (Group A) were compared with 40 biopsies carried out using stereotaxis with digital imaging (Group B). In all cases specimen radiography was performed and analysed for the presence of calcifications. Pathological correlation was then carried out with needle and surgical histology. RESULTS: The use of digital add-on equipment increased the radiographic calcification retrieval rate from 55 to 85% (P < 0.005). The absolute sensitivity of core biopsy in pure ductal carcinoma in situ (DCIS) cases rose from 34 to 69% (P < 0.03), with the complete sensitivity increasing from 52 to 94% (P < 0.005). For DCIS with or without an invasive component the absolute sensitivity rose from 41 to 67% (P = 0.052), while the complete sensitivity was 59% before and 86% after the introduction of digital imaging (P < 0.04). CONCLUSION: Digital equipment improves the performance of upright stereotactic core biopsy of microcalcifications, giving a significantly increased success rate in accurately obtaining calcifications. This leads to an improvement in absolute and complete sensitivity of core biopsy when diagnosing DCIS. Whitlock, J.P.L. (2000)

  13. Bedrock Hydrogeology-Site investigation SFR

    International Nuclear Information System (INIS)

    Oehman, Johan; Bockgaard, Niclas; Follin, Sven

    2012-06-01

    The Swedish Nuclear Fuel and Waste Management Company (SKB) has conducted site investigations for a planned extension of the existing final repository for short-lived radioactive waste (SFR). This report presents an integrated analysis and interpretation of the historic data from the existing SFR (1980 - 1986), as well as, from the recent investigations for the planned extension of SFR (2008 - 2009). The primary objective is to establish a conceptual hydrogeological model of the bedrock for safety assessment and design analyses. Analyses and interpretations of all (old and new) hydraulic data are analysed with regard to the recently developed geological deformation zone model of the SFR model domain (Curtis et al. 2011). The methodology used by Curtis et al. (2011) has focussed on magnetic anomalies and deformation zone intercepts with ground surface greater than 300 m. In the hydrogeological modelling, however, it has been considered important to also explore the occurrence and characteristics of shallow horizontal to sub-horizontal structures (sheet joints) inside the SFR model domain. Such structures are of considerable importance for the hydrogeology in the uppermost c. 150 m of bedrock in SDM-Site Forsmark; hence the term Shallow Bedrock Aquifer was used to emphasise their hydraulic significance. In this study, the acronym SBA-structure is used for horizontal structures identified in the hydrogeological modelling. In addition to the predominantly steeply dipping geological deformation zones, eight so-called SBA-structures are modelled deterministically in the hydrogeological model. The SBA-structures are envisaged as hydraulically heterogeneous and composed of clusters of minor gently dipping to horizontal fractures rather than extensive single features. A type of structures that is partly included in the definition of the SBA-structures is the Unresolved Possible Deformations Zone (Unresolved PDZ) intercepts identified by Curtis et al. (2011). The Unresolved

  14. Core supervision methods and future improvements of the core master/presto system at KKB

    International Nuclear Information System (INIS)

    Lundberg, S.; Wenisch, J.; Teeffelen, W.V.

    2000-01-01

    Kernkraftwerk Brunsbuettel (KKB) is a KWU 806 MW e BWR located at the lower river Elbe, in Germany. The reactor has been in operation since 1976 and is now operating in its 14. cycle. The core supervision at KKB is performed with the ABB CORE MASTER system. This system mainly contains the 3-D simulator PRESTO supplied by Studsvik Scandpower A/S. The core supervision is performed by periodic PRESTO 3-D evaluations of the reactor operation state. The power distribution calculated by PRESTO is adapted with the ABB UPDAT program using the on-line LPRM readings. The thermal margins are based on this adapted power distribution. Related to core supervision, the function of the PRESTO/UPDAT codes is presented. The UPDAT method is working well and is capable of reproducing the true core power distribution. The quality of the 3-D calculation is, however, an important ingredient of the quality of the adapted power distribution. The adaptation method as such is also important for this quality. The data quality of this system during steady state and off-rate states (reactor manoeuvres) are discussed by presenting comparisons between PRESTO and UPDAT thermal margin utilisation from Cycle 13. Recently analysed asymmetries in the UPDAT evaluated MCPR values are also presented and discussed. Improvements in the core supervision such as the introduction of advanced modern nodal methods (PRESTO-2) are presented and an alternative core supervision philosophy is discussed. An ongoing project with the goal to update the data and result presentation interface (GUI) is also presented. (authors)

  15. Modelling of future hydrogeological conditions at SFR

    International Nuclear Information System (INIS)

    Holmen, L.G.; Stigsson, M.

    2001-03-01

    The purpose is to estimate the future groundwater movements at the SFR repository and to produce input to the quantitative safety assessment of the SFR. The future flow pattern of the groundwater is of interest, since components of the waste emplaced in a closed and abandoned repository will dissolve in the groundwater and be transported by the groundwater to the ground surface. The study is based on a system analysis approach. Three-dimensional models were devised of the studied domain. The models include the repository tunnels and the surrounding rock mass with fracture zones. The formal models used for simulation of the groundwater flow are three-dimensional mathematical descriptions of the studied hydraulic system. The studied domain is represented on four scales - regional, local, semi local and detailed - forming four models with different resolutions: regional, local, semi local and detailed models. The local and detailed models include a detailed description of the tunnel system at SFR and of surrounding rock mass and fracture zones. In addition, the detailed model includes description of the different structures that take place inside the deposition tunnels. At the area studied, the shoreline will retreat due to the shore level displacement; this process is included in the models. The studied period starts at 2000 AD and continues until a steady state like situation is reached for the surroundings of the SFR at ca 6000 AD. The models predict that as long as the sea covers the ground above the SFR, the regional groundwater flow as well as the flow in the deposition tunnels are small. However, due to the shore level displacement the shoreline (the sea) will retreat. Because of the retreating shoreline, the general direction of the groundwater flow at SFR will change, from vertical upward to a more horizontal flow; the size of the groundwater flow will be increased as well. The present layout of the SFR includes five deposition tunnels: SILO, BMA, BLA, BTF1

  16. Project SAFE. Update of the SFR-1 safety assessment. Phase 1. Appendix A5: Radionuclide transport

    International Nuclear Information System (INIS)

    Moreno, L.

    1998-01-01

    A critical revision of the previous safety assessments made by SKB on the Final Repository for Radioactive Operational Waste, SFR is presented. The review of the Deepened Safety Assessment is also discussed. Based on this critical revision improvements are suggested. Hydrology, formation of complexes, and long-term behaviour of the barriers are some of the aspects where the safety assessment could be improved

  17. DENSE CORES IN THE PIPE NEBULA: AN IMPROVED CORE MASS FUNCTION

    International Nuclear Information System (INIS)

    Rathborne, J. M.; Lada, C. J.; Muench, A. A.; Alves, J. F.; Kainulainen, J.; Lombardi, M.

    2009-01-01

    In this paper, we derive an improved core mass function (CMF) for the Pipe Nebula from a detailed comparison between measurements of visual extinction and molecular-line emission. We have compiled a refined sample of 201 dense cores toward the Pipe Nebula using a two-dimensional threshold identification algorithm informed by recent simulations of dense core populations. Measurements of radial velocities using complimentary C 18 O (1-0) observations enable us to cull out from this sample those 43 extinction peaks that are either not associated with dense gas or are not physically associated with the Pipe Nebula. Moreover, we use the derived C 18 O central velocities to differentiate between single cores with internal structure and blends of two or more physically distinct cores, superposed along the same line of sight. We then are able to produce a more robust dense core sample for future follow-up studies and a more reliable CMF than was possible previously. We confirm earlier indications that the CMF for the Pipe Nebula departs from a single power-law-like form with a break or knee at M ∼ 2.7 ± 1.3 M sun . Moreover, we also confirm that the CMF exhibits a similar shape to the stellar initial mass function (IMF), but is scaled to higher masses by a factor of ∼4.5. We interpret this difference in scaling to be a measure of the star formation efficiency (22% ± 8%). This supports earlier suggestions that the stellar IMF may originate more or less directly from the CMF.

  18. Fabrication of preliminary fuel rods for SFR

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan

    2012-01-01

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  19. Development of Basic Key Technologies for Gen IV SFR Safety Evaluation

    International Nuclear Information System (INIS)

    Jeong, Hae Yong; Kwon, Young Min; Kim, Tae Woon; Park, Soo Yong; Suk, Soo Dong; Lee, Kwi Lim; Lee, Yong Bum; Chang, Won Pyo; Ha, Kwi Seok; Hahn, Sang Hoon

    2010-07-01

    Safety issues and design requirements on control rod worth were identified through the evaluation of safety design characteristics and the preliminary safety evaluation. This results will be taken into account for the conceptual design studies of the demonstration reactor in the next stage. The Level-1 Pasa has been performed and a quantitative Cdf value was produced for the selected design from the several candidates. The inherent safety characteristics of the selected design were evaluated through the DBE and ATWS analyses. A surrogate material for Tru has been selected which is applicable to the study of liquidus/solidus temperature test for the metallic fuel containing Tru. A methodology for the regression analysis with surrogate material has been developed and valuable data on metal fuel liquidus/solidus temperature have been measured. A simple mechanistic model describing a bending of subassemblies has been formulated based on the foreign test data and existing models. Its applicability has been evaluated for the Phenix design. New criteria of the core damage for the SFR PSA were identified. The list of initiating events, system response event tree, and core response event tree, which constitute a PSA methodology for an SFR, have been introduced. By developing the SFR PIRT, phenomenological model features, which have to be satisfied in a safety code, were defined and the PIRT results were applied to the design of the PDRC test facility. Bases for a safety evaluation methodology for the SFR DBEs have been also prepared. A draft version of the topical report on the code for local fault analysis has been completed. Since 2007, the MARS-LMR code has been developed and assessments for model validation with the test data from EBR-II and Phenix reactor have been continued. The code has been applied to the evaluation of passive safety of a conceptual design of Gen IV SFR

  20. Groundwater chemical changes at SFR in Forsmark

    Energy Technology Data Exchange (ETDEWEB)

    Laaksoharju, Marcus [GeoPoint AB, Sollentuna (Sweden); Gurban, Ioana [3DTerra (Sweden)

    2003-01-01

    The examination of the groundwater sampled at the SFR tunnel system indicated that the groundwater consist mainly of a Na-Cl to Na-Ca-Cl type of water. Most of the samples fall within the Cl range of 2500-5500 mg/l having a neutral pH (6.6-7.7 units). The water is reducing and despite the fact that the tunnel acts like a hydraulic sink constantly withdrawing water out from the rock into the tunnel the groundwater changes are moderate with time. Most of the sampling points in the SFR tunnel system are located under the Sea and M3 calculations indicated that most of the sampling points have a change of water types from an older marine water type affected by glacial melt water to an more modern marine water type such as Baltic Sea water which has been modified by possibly microbial sulphate reduction and ion exchange. Mass balance calculations indicated that the waters seem to be in equilibrium with the fracture filling mineral such as calcite. The quality of the aluminium data made the modelling with the major rock forming aluminium silicates such as feldspars and clay minerals uncertain and was therefore not reported. The conclusion is that the groundwater evolution and patterns at SFR are a result of many factors such as: 1. the changes in hydrogeology related to glaciation/deglaciation and land uplift, 2. repeated Sea/lake water regressions/transgressions 3. the closeness to Baltic Sea resulting in relative small hydrogeological driving forces which could preserve old water types from being flushed out, 4. organic or inorganic alteration of the groundwater caused by microbial processes or in situ water/rock interactions 5. tunnel construction which changed the flow system The modelled present-day groundwater conditions of the SFR site consist of a mixture in varying degrees of different water types. The data indicate that all the groundwater at SFR is strongly affected by Sea water of different origin and ages. The meteoric (0- 1000 B.P) portion is located close

  1. Improvement of JRR-4 core management code system

    International Nuclear Information System (INIS)

    Izumo, H.; Watanabe, S.; Nagatomi, H.; Hori, N.

    2000-01-01

    In the modification of JRR-4, the fuel was changed from 93% high enrichment uranium aluminized fuel to 20% low enriched uranium silicide fuel in conformity with the framework of reduced enrichment program on JAERI research reactors. As changing of this, JRR-4 core management code system which estimates excess reactivity of core, fuel burn-up and so on, was improved too. It had been difficult for users to operate the former code system because its input-output form was text-form. But, in the new code system (COMMAS-JRR), users are able to operate the code system without using difficult text-form input. The estimation results of excess reactivity of JRR-4 LEU fuel core were showed very good agreements with the measured value. It is the strong points of this new code system to be operated simply by using the windows form pictures act on a personal workstation equip with the graphical-user-interface (GUI), and to estimate accurately the specific characteristics of the LEU core. (author)

  2. Safety design approach for JSFR toward the realization of GEN IV SFR

    International Nuclear Information System (INIS)

    Kubo, S.; Yamano, H.; Chikazawa, Y.; Shimakawa, Y.

    2013-01-01

    Conclusion: Safety Design Approach for JSFR: • Based on the safety design criteria for Generation-IV SFR • DECs, Situations practically eliminated and related design measures are identified and selected with due consideration of the safety features of SFR and the lessons learned from the TEPCO’s Fukushima Dai-ichi nuclear power plants accident Safety Design Concept of JSFR: • For failure to shutdown: Passive shutdown capability, Mitigation of core damage (Prevention of severe mechanical energy release, In-Vessel Retention) • For failure to remove heat: Prevention of significant core damage (Natural circulation DHR, Alternative cooling measures) • Containment: Prevention of sever dynamic loads by design measures (IVR, double boundary concept, inertization)

  3. State of the art of CATHARE model for transient safety analysis of ASTRID SFR

    International Nuclear Information System (INIS)

    Lavastre, R.; Conti, A.; Marsault, Ph.; Chenaud, M.S.; Tosello, A.

    2014-01-01

    Within the framework of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), the conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves enhancing the general design in order to : - increase the safety margins for all unprotected-loss-of-flow (ULOF) and unprotected-loss-of-heat-sink (ULOHS) transients, - identify the need for additional safety devices that would complement core natural behavior so that temperature criteria on coolant, core and primary circuit structures can remain under the safety criteria. For this purpose, the use of CATHARE system code has been very important from the early stage of design in order to ensure a feedback for design teams to improve behavior during unprotected transients. Until 2012, CATHARE ULOxx transient calculations have been used mainly to compare different core designs. They contributed to lead to the choice of CFV core (axially heterogeneous core with an upper sodium plenum employed to achieve a negative sodium void reactivity worth). Meanwhile, models for an accurate core description and transients have been developed in CATHARE to improve the calculations towards best estimate calculations for safety analysis. This paper therefore presents these main developments in core modeling achieved for the 2 past years. For instance, we will focus on the way of dealing with fuel assemblies that have to be grouped together in the CATHARE code to form a channel with similar neutronic physics and thermal-hydraulics characteristics. We will also explain the way we deal with heterogeneity of fuel pin to obtain the accurate fuel temperature along the axis and to take into account pellet-cladding gap state. These two points have a great importance on feedback effects linked to the fuel, mainly the Doppler effect. The paper will finally introduce the upcoming improvements that are under development nowadays

  4. Site investigation SFR. Reprocessing of reflection seismic profiles 5b and 8, Forsmark

    Energy Technology Data Exchange (ETDEWEB)

    Juhlin, Christopher; Zhang, Fengjiao (Uppsala Univ., Dept. of Earth Sciences (Sweden))

    2010-12-15

    Reflection seismic profiles 5b and 8 in the northern Forsmark area have been reprocessed with the aim of improving the images in the uppermost 500 metres in the SFR area. The main conclusion is that a new reflection (B10) has been identified that may extend below the SFR site. This reflection was not clearly observed in the previous processing. The reflection strikes approximately N25E and dips at about 35 degrees to the southeast. This orientation is similar to the set B group identified earlier /Juhlin and Palm 2005/. Note that the dip of the reflection is uncertain. On shot gathers it appears to dip at a slightly shallower angle while on the stacked sections it appears to dip at a greater angle. This discrepancy is probably due to the crooked nature of the profiles. However, reflections are clearly observed in shot gathers and its presence below SFR is highly probable. Two new reflections were also identified further north along profile 5b (A11 and A12). These dip to the south-southeast, but would be found at a depth of 1-2 km below SFR if they extend to below the site. There are also signs of a 3rd reflection with similar orientation to the set A group identified earlier, A13, but its existence is very speculative. This reflector would intersect the surface within the SFR area. South of the Singoe deformation zone on profile 5b, another new reflection has been found, N1. The orientation of this reflection is speculative since it is not clearly seen on profile 8. It has been modelled as dipping to the north at about 35 degrees and projects to the surface south of the main SFR area. In addition, the orientation of reflection B7 has been revised as has the lateral extent of A1. Most importantly, A1 is now interpreted not to extend to the surface and not cross the Singoe deformation zone

  5. Improvement of numerical analysis method for FBR core characteristics. 3

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Yamamoto, Toshihisa; Kitada, Takanori; Katagi, Yousuke

    1998-03-01

    As the improvement of numerical analysis method for FBR core characteristics, studies on several topics have been conducted; multiband method, Monte Carlo perturbation and nodal transport method. This report is composed of the following three parts. Part 1: Improvement of Reaction Rate Calculation Method in the Blanket Region Based on the Multiband Method; A method was developed for precise evaluation of the reaction rate distribution in the blanket region using the multiband method. With the 3-band parameters obtained from the ordinary fitting method, major reaction rates such as U-238 capture, U-235 fission, Pu-239 fission and U-238 fission rate distributions were analyzed. Part 2: Improvement of Estimation Method for Reactivity Based on Monte-Carlo Perturbation Theory; Perturbation theory based on Monte-Carlo perturbation theory have been investigated and introduced into the calculational code. The Monte-Carlo perturbation code was applied to MONJU core and the calculational results were compared to the reference. Part 3: Improvement of Nodal Transport Calculation for Hexagonal Geometry; A method to evaluate the intra-subassembly power distribution from the nodal averaged neutron flux and surface fluxes at the node boundaries, was developed based on the transport theory. (J.P.N.)

  6. Application of the Toyota Production System improves core laboratory operations.

    Science.gov (United States)

    Rutledge, Joe; Xu, Min; Simpson, Joanne

    2010-01-01

    To meet the increased clinical demands of our hospital expansion, improve quality, and reduce costs, our tertiary care, pediatric core laboratory used the Toyota Production System lean processing to reorganize our 24-hour, 7 d/wk core laboratory. A 4-month, consultant-driven process removed waste, led to a physical reset of the space to match the work flow, and developed a work cell for our random access analyzers. In addition, visual controls, single piece flow, standard work, and "5S" were instituted. The new design met our goals as reflected by achieving and maintaining improved turnaround time (TAT; mean for creatinine reduced from 54 to 23 minutes) with increased testing volume (20%), monetary savings (4 full-time equivalents), decreased variability in TAT, and better space utilization (25% gain). The project had the unanticipated consequence of eliminating STAT testing because our in-laboratory TAT for routine testing was less than our prior STAT turnaround goal. The viability of this approach is demonstrated by sustained gains and further PDCA (Plan, Do, Check, Act) improvements during the 4 years after completion of the project.

  7. Improving Battery Reactor Core Design Using Optimization Method

    International Nuclear Information System (INIS)

    Son, Hyung M.; Suh, Kune Y.

    2011-01-01

    The Battery Omnibus Reactor Integral System (BORIS) is a small modular fast reactor being designed at Seoul National University to satisfy various energy demands, to maintain inherent safety by liquid-metal coolant lead for natural circulation heat transport, and to improve power conversion efficiency with the Modular Optimal Balance Integral System (MOBIS) using the supercritical carbon dioxide as working fluid. This study is focused on developing the Neutronics Optimized Reactor Analysis (NORA) method that can quickly generate conceptual design of a battery reactor core by means of first principle calculations, which is part of the optimization process for reactor assembly design of BORIS

  8. Minutes of the 2. Meeting of the WPRS / EGRPANS / Sodium Fast Reactor Task Force (SFR)

    International Nuclear Information System (INIS)

    Ivanov, Evgeny; Kereszturi, Andras; Pataki, I.; Tota, A.; Vertes, P.; Kim, Taek K.; Taiwo, T.A.; Kugo, Teruhiko; Lee, Yi Kang; Messaoudi, Nadia; Michel-Sendis, Franco; ); Pascal, Vincent; Buiron, Laurent; Varaine, Frederic; Ponomarev, Alexander

    2012-01-01

    Five organizations (SCK/CEN, KIT, KFKI, CEA, ANL) participated in the Sodium-cooled fast reactor (SFR) Benchmark calculations and all results were collected and compiled by CEA and ANL. The compiled results of the large size cores and medium size cores were presented by V. Pascal (CEA) and T. K. Kim (ANL), respectively. Separately, A. Kereszturi presented his recently updated results. It was observed that there is wide variation in core multiplication factor, kinetics parameters, and reactivity feedback coefficients. In particular, compared to the CEA results, ANL calculated smaller k-eff, Doppler constant, but higher sodium void worth and control rod worth. The core modeling issue (heterogeneous vs. homogeneous) and solution method (diffusion vs. transport) were identified as the potential reasons of these discrepancies, including the minor impacts from the depletion chains and lumped fission product modeling. All participants agreed that additional investigation was needed to identify the reasons of these discrepancies. In addition, V. Pascal presented the informative notes of the reactivity feedback calculations methodology proposed by CEA. This document brings together the 5 presentations (slides) given at this meeting: 1 - SFR Task Force : Core behavior during transient as a function of power size and fuel nature (L. Buiron, V. Pascal, F. Varaine); 2 - Sodium Fast Reactor core Feedback and Transient response (SFRFT) Expert Group: preliminary benchmark results for large cores (L. Buiron, V. Pascal, F. Varaine); 3 - Numerical Benchmark Results for 1000 MWth Sodium-cooled Fast Reactor (T.K. Kim and T.A. Taiwo); 4 - Preliminary results of the WPRS Sodium-Cooled Fast Reactor Benchmark problems (A. Kereszturi, I. Pataki, A. Tota, P. Vertes); 5 - SFR Task Force : proposal for Feedback coefficients estimation methodology (L. Buiron, V.Pascal, F. Varaine)

  9. Review of SFR In-Vessel Radiological Source Term Studies

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Lee, Yong Bum

    2008-10-01

    An effort has been made in this study to search for and review the literatures in public domain on the studies of the phenomena related to the release of radionuclides and aerosols to the reactor containment of the sodium fast reactor (SFR) plants (i.e., in-vessel source term), made in Japan and Europe including France, Germany and UK over the last few decades. Review work is focused on the experimental programs to investigate the phenomena related to determining the source terms, with a brief review on supporting analytical models and computer programs. In this report, the research programs conducted to investigate the CDA (core disruptive accident) bubble behavior in the sodium pool for determining 'primary' or 'instantaneous' source term are first introduced. The studies performed to determine 'delayed source term' are then described, including the various stages of phenomena and processes: fission product (FP) release from fuel , evaporation release from the surface of the pool, iodine mass transfer from fission gas bubble, FP deposition , and aerosol release from core-concrete interaction. The research programs to investigate the release and transport of FPs and aerosols in the reactor containment (i.e., in-containment source term) are not described in this report

  10. Evaluation of alternative fluids for SFR intermediate loops

    International Nuclear Information System (INIS)

    Brissonneau, L.; Simon, N.; Baque, F.

    2009-01-01

    Among the Generation IV systems, Sodium Fast Reactors (SFR) are promising and benefit of considerable technological experience, but improvements are researched on safety approach and capital cost reduction. One of the main drawback to be solved by the standard SFR design is the proper management of the risk of leakage between the intermediate circuit filled with sodium and the energy conversion system using a water Rankine cycle. The limitation of this risk requires notably an early detection of water leakage to prevent a water-sodium reaction. One innovative solution consists in the replacement of the sodium in the secondary loops by an alternative liquid fluid, not or less reactive with water. This alternative fluid might also allow innovative designs, e.g. intermediate heat exchanger and steam generator grouped in the same component. CEA, Areva NP and EdF have joined in a working group in order to evaluate different 'alternative fluids' that might replace sodium. A first selection retained seven fluids on the basis of 'required properties' as large operating range (low melting point, high boiling point ...), fluid cost and availability, acceptable corrosion at SFR working temperature. These are three bismuth alloys, two nitrate salts, one hydroxide melt and sodium with nanoparticles of nickel. Then, it was decided to evaluate these fluids through a multi-criteria analysis in order to quantify advantages and drawbacks of each fluid and to compare them with sodium. Lack of knowledge, impact on materials, design, working conditions and reactor availability should be emphasized by this analysis, in order to provide sound arguments for a research program on one or two promising fluids. A global note is given to each fluid by evaluating them with respect to 'grand criteria', weighted differently according to their importance. The grand criteria are : thermal properties, reactivity with structures, reactivity with other fluids (air, water, sodium), chemistry control

  11. Development of Melting Crucible Materials of Metallic Fuel Slug for SFR

    International Nuclear Information System (INIS)

    Kim, K. H.; Lee, C. T.; Oh, S. J.; Kim, S. K.; Lee, C. B.; Ko, Y. M.; Woo, W. M.

    2010-01-01

    The fabrication process of metallic fuel for SFR(sodium fast reactor) of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA(minor actinide) elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr-Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized

  12. Preliminary Development of Regulatory PSA Models for SFR

    International Nuclear Information System (INIS)

    Choi, Yong Won; Shin, Andong; Bae, Moohoon; Suh, Namduk; Lee, Yong Suk

    2013-01-01

    Well developed PRA methodology exists for LWR (Light Water Reactor) and PHWR (Pressurized Heavy Water Reactor). Since KAERI is developing a prototype SFR targeting to apply for a license by 2017, KINS needs to have a PRA models to assess the safety of this prototype reactor. The purpose of this study is to develop the regulatory PSA models for the independent verification of the SFR safety. Since the design of the prototype SFR is not mature yet, we have tried to develop the preliminary models based on the design data of KAERI's previous SFR design. In this study, the preliminary initiating events of level 1 internal event for SFR were selected through reviews of existing PRA (LWR, PRISM, ASTRID and KALIMER-600) models. Then, the event tree for each selected initiating event was developed. The regulatory PRA models of SFR developed are preliminary in a sense, because the prototype SFR design is not mature and provided yet. Still it might be utilized for the forthcoming licensing review in assessing the risk of safety issues and the configuration control of the design

  13. Review of SFR Design Safety using Preliminary Regulatory PSA Model

    International Nuclear Information System (INIS)

    Na, Hyun Ju; Lee, Yong Suk; Shin, Andong; Suh, Nam Duk

    2013-01-01

    The major objective of this research is to develop a risk model for regulatory verification of the SFR design, and thereby, make sure that the SFR design is adequate from a risk perspective. In this paper, the development result of preliminary regulatory PSA model of SFR is discussed. In this paper, development and quantification result of preliminary regulatory PSA model of SFR is discussed. It was confirmed that the importance PDRC and ADRC dampers is significant as stated in the result of KAERI PSA model. However, the importance can be changed significantly depending on assumption of CCCG and CCF factor of PDRC and ADRC dampers. SFR (sodium-cooled fast reactor) which is Gen-IV nuclear energy system, is designed to accord with the concept of stability, sustainability and proliferation resistance. KALIMER-600, which is under development in Korea, includes passive safety systems (e. g. passive reactor shutdown, passive residual heat removal, and etc.) as well as active safety systems. Risk analysis from a regulatory perspective is needed to support the regulatory body in its safety and licensing review for SFR (KALIMER-600). Safety issues should be identified in the early design phase in order to prevent the unexpected cost increase and delay of the SFR licensing schedule that may be caused otherwise

  14. Swi5-Sfr1 protein stimulates Rad51-mediated DNA strand exchange reaction through organization of DNA bases in the presynaptic filament.

    KAUST Repository

    Fornander, Louise H

    2013-12-03

    The Swi5-Sfr1 heterodimer protein stimulates the Rad51-promoted DNA strand exchange reaction, a crucial step in homologous recombination. To clarify how this accessory protein acts on the strand exchange reaction, we have analyzed how the structure of the primary reaction intermediate, the Rad51/single-stranded DNA (ssDNA) complex filament formed in the presence of ATP, is affected by Swi5-Sfr1. Using flow linear dichroism spectroscopy, we observe that the nucleobases of the ssDNA are more perpendicularly aligned to the filament axis in the presence of Swi5-Sfr1, whereas the bases are more randomly oriented in the absence of Swi5-Sfr1. When using a modified version of the natural protein where the N-terminal part of Sfr1 is deleted, which has no affinity for DNA but maintained ability to stimulate the strand exchange reaction, we still observe the improved perpendicular DNA base orientation. This indicates that Swi5-Sfr1 exerts its activating effect through interaction with the Rad51 filament mainly and not with the DNA. We propose that the role of a coplanar alignment of nucleobases induced by Swi5-Sfr1 in the presynaptic Rad51/ssDNA complex is to facilitate the critical matching with an invading double-stranded DNA, hence stimulating the strand exchange reaction.

  15. Examining memorandum: Ultimate store for nuclear reactor wastes - SFR-1

    International Nuclear Information System (INIS)

    Bergman, C.; Ericsson, G.; Godaas, T.; Haegg, C.; Johansson, G.

    1988-01-01

    The report constitutes the basis for the position of the National Institute of Radiation Protection as regards permission to operate SFR-1 at Forsmark. The memorandum describes: - existing conditions regarding commissioning SFR-1, - summarily the final safety report from the Swedish Fuel and Waste Management Co, - consultant contributions ordered in connection with the examination, - the judgement of the institute in all questions relevant to radiation protection conditions in SFR-1. The institute has made it's own estimates of the radiation doses the repository could be the source of. It is concluded that the radiation doses from the repository are acceptable and consequently operation permission is recommended. (O.S.)

  16. Project SAFE. Complexing agents in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Fanger, G.; Skagius, K.; Wiborgh, M. [Kemakta Konsult AB, Stockholm (Sweden)

    2001-01-01

    Low- and intermediate level radioactive waste, produced at Swedish nuclear power plants, will be deposited in an underground repository, SFR. Different substances in the waste or in degradation products emanating from the waste, and chemicals added during the building of cementitious barriers in the repository, may exhibit complexing properties. The complexation of radionuclides with such ligands may increase the mobility of the deposited radionuclides as sorption on the cement phases is decreased and solubility increased. This could lead to an increased leaching of the radionuclides from the repository to the geosphere and biosphere. To be able to evaluate the implications for the function and long-term safety of the repository a study has been performed on complexants in SFR. The study is a part of project SAFE (Safety Assessment of Final Repository for operational Radioactive Waste) at the Swedish Nuclear Fuel and Waste Management Co, SKB. Concentrations of complexants were calculated in different waste types in the repository and compared to critical levels above which radionuclide sorption may be affected. The analysis is based on recent research presented in international and national literature sources. The waste in SFR that may act or give rise to substances with complexing properties mainly consists of cellulose materials, including cement additives used in waste conditioning and backfill grout. The radioactive waste also contains chemicals mainly used in decontamination processes at the nuclear power plants, e.g. EDTA, NTA, gluconate, citric acid and oxalic acid. The calculations performed in this report show that the presence of complexants in SFR may lead to a sorption reduction for some radionuclides in certain waste types. This may have to be considered when performing calculations of the radionuclide transport. Concentration calculations of isosaccharinic acid (ISA), using a degradation yield of 0.1 mole/kg cellulose (2%), showed that the limit above

  17. Project SAFE. Complexing agents in SFR

    International Nuclear Information System (INIS)

    Fanger, G.; Skagius, K.; Wiborgh, M.

    2001-01-01

    Low- and intermediate level radioactive waste, produced at Swedish nuclear power plants, will be deposited in an underground repository, SFR. Different substances in the waste or in degradation products emanating from the waste, and chemicals added during the building of cementitious barriers in the repository, may exhibit complexing properties. The complexation of radionuclides with such ligands may increase the mobility of the deposited radionuclides as sorption on the cement phases is decreased and solubility increased. This could lead to an increased leaching of the radionuclides from the repository to the geosphere and biosphere. To be able to evaluate the implications for the function and long-term safety of the repository a study has been performed on complexants in SFR. The study is a part of project SAFE (Safety Assessment of Final Repository for operational Radioactive Waste) at the Swedish Nuclear Fuel and Waste Management Co, SKB. Concentrations of complexants were calculated in different waste types in the repository and compared to critical levels above which radionuclide sorption may be affected. The analysis is based on recent research presented in international and national literature sources. The waste in SFR that may act or give rise to substances with complexing properties mainly consists of cellulose materials, including cement additives used in waste conditioning and backfill grout. The radioactive waste also contains chemicals mainly used in decontamination processes at the nuclear power plants, e.g. EDTA, NTA, gluconate, citric acid and oxalic acid. The calculations performed in this report show that the presence of complexants in SFR may lead to a sorption reduction for some radionuclides in certain waste types. This may have to be considered when performing calculations of the radionuclide transport. Concentration calculations of isosaccharinic acid (ISA), using a degradation yield of 0.1 mole/kg cellulose (2%), showed that the limit above

  18. Tailoring Sandwich Face/Core Interfaces for Improved Damage Tolerance

    DEFF Research Database (Denmark)

    Lundsgaard-Larsen, Christian; Berggreen, Christian; Carlsson, Leif A.

    2010-01-01

    A face/core debond in a sandwich structure may propagate in the interface or kink into either the face or core. It is found that certain modifications of the face/core interface region influence the kinking behavior, which is studied experimentally in the present paper. A sandwich double cantilever....... The transition points where the crack kinks are identified and the influence of four various interface design modifications on the propagation path and fracture resistance are investigated....

  19. Improved Rock Core Sample Break-off, Retention and Ejection System, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed effort advances the design of an innovative core sampling and acquisition system with improved core break-off, retention and ejection features. The...

  20. Improved Rock Core Sample Break-off, Retention and Ejection System, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed effort advances the design of an innovative core sampling and acquisition system with improved core break-off, retention and ejection features. Phase 1...

  1. On-Line Core Thermal-Hydraulic Model Improvement

    International Nuclear Information System (INIS)

    In, Wang Kee; Chun, Tae Hyun; Oh, Dong Seok; Shin, Chang Hwan; Hwang, Dae Hyun; Seo, Kyung Won

    2007-02-01

    The objective of this project is to implement a fast-running 4-channel based code CETOP-D in an advanced reactor core protection calculator system(RCOPS). The part required for the on-line calculation of DNBR were extracted from the source of the CETOP-D code based on analysis of the CETOP-D code. The CETOP-D code was revised to maintain the input and output variables which are the same as in CPC DNBR module. Since the DNBR module performs a complex calculation, it is divided into sub-modules per major calculation step. The functional design requirements for the DNBR module is documented and the values of the database(DB) constants were decided. This project also developed a Fortran module(BEST) of the RCOPS Fortran Simulator and a computer code RCOPS-SDNBR to independently calculate DNBR. A test was also conducted to verify the functional design and DB of thermal-hydraulic model which is necessary to calculate the DNBR on-line in RCOPS. The DNBR margin is expected to increase by 2%-3% once the CETOP-D code is used to calculate the RCOPS DNBR. It should be noted that the final DNBR margin improvement could be determined in the future based on overall uncertainty analysis of the RCOPS

  2. On-Line Core Thermal-Hydraulic Model Improvement

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Chun, Tae Hyun; Oh, Dong Seok; Shin, Chang Hwan; Hwang, Dae Hyun; Seo, Kyung Won

    2007-02-15

    The objective of this project is to implement a fast-running 4-channel based code CETOP-D in an advanced reactor core protection calculator system(RCOPS). The part required for the on-line calculation of DNBR were extracted from the source of the CETOP-D code based on analysis of the CETOP-D code. The CETOP-D code was revised to maintain the input and output variables which are the same as in CPC DNBR module. Since the DNBR module performs a complex calculation, it is divided into sub-modules per major calculation step. The functional design requirements for the DNBR module is documented and the values of the database(DB) constants were decided. This project also developed a Fortran module(BEST) of the RCOPS Fortran Simulator and a computer code RCOPS-SDNBR to independently calculate DNBR. A test was also conducted to verify the functional design and DB of thermal-hydraulic model which is necessary to calculate the DNBR on-line in RCOPS. The DNBR margin is expected to increase by 2%-3% once the CETOP-D code is used to calculate the RCOPS DNBR. It should be noted that the final DNBR margin improvement could be determined in the future based on overall uncertainty analysis of the RCOPS.

  3. Core symptoms of autism improved after vitamin D supplementation.

    Science.gov (United States)

    Jia, Feiyong; Wang, Bing; Shan, Ling; Xu, Zhida; Staal, Wouter G; Du, Lin

    2015-01-01

    Autism spectrum disorder (ASD) is a common neurodevelopmental disorder caused by a complex interaction between genetic and environmental risk factors. Among the environmental factors, vitamin D3 (cholecaliferol) seems to play a significant role in the etiology of ASD because this vitamin is important for brain development. Lower concentrations of vitamin D3 may lead to increased brain size, altered brain shape, and enlarged ventricles, which have been observed in patients with ASD. Vitamin D3 is converted into 25-hydroxyvitamin D3 in the liver. Higher serum concentrations of this steroid may reduce the risk of autism. Importantly, children with ASD are at an increased risk of vitamin D deficiency, possibly due to environmental factors. It has also been suggested that vitamin D3 deficiency may cause ASD symptoms. Here, we report on a 32-month-old boy with ASD and vitamin D3 deficiency. His core symptoms of autism improved significantly after vitamin D3 supplementation. This case suggests that vitamin D3 may play an important role in the etiology of ASD, stressing the importance of clinical assessment of vitamin D3 deficiency and the need for vitamin D3 supplementation in case of deficiency. Copyright © 2015 by the American Academy of Pediatrics.

  4. An Innovative Hybrid Loop-Pool SFR Design and Safety Analysis Methods: Today and Tomorrow

    International Nuclear Information System (INIS)

    Hongbin Zhang; Haihua Zhao; Vincent Mousseau

    2008-01-01

    Investment in commercial sodium cooled fast reactor (SFR) power plants will become possible only if SFRs achieve economic competitiveness as compared to light water reactors and other Generation IV reactors. Toward that end, we have launched efforts to improve the economics and safety of SFRs from the thermal design and safety analyses perspectives at Idaho National Laboratory. From the thermal design perspective, an innovative hybrid loop-pool SFR design has been proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to further improve economics and safety. From the safety analyses perspective, we have initiated an effort to develop a high fidelity reactor system safety code

  5. Detection of cores in fingerprints with improved dimension reduction

    NARCIS (Netherlands)

    Bazen, A.M.; Veldhuis, Raymond N.J.

    In this paper, we present a statistical approach to core detection in fingerprint images that is based on the likelihood ratio, using models of variation of core templates and randomly chosen templates. Additionally, we propose an alternative dimension reduction method. Unlike standard linear

  6. Development of mechanical analysis module for simulation of SFR fuel rod behavior using finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Andong; Jeong, Hyedong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Hyochan; Yang, Yongsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Korean SFR developer decided to adapt metal fuel, current study focused on the metal fuel instead of oxide fuel. The SFR metal fuel has been developed by Korea Atomic Energy Research Institute (KAERI) and many efforts focused on designing and manufacturing the metal fuel. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured during steady-state operation and accident condition within an acceptable range. Whereas the design and evaluation methodologies, code systems and test procedures of a light water reactor fuel are sufficiently established, those of the SFR fuel needs more technical advances. In the view of regulatory point, there are still many challenging issues which are required to secure the safety of fuel and reactors. For this reason, the Korea Institute of Nuclear Safety (KINS) has launched the new project to develop the regulatory technology for SFR system including a fuel area. The ALFUS code was developed by CRIEPI and employs mechanistic model for fission gas release and swelling of fuel slug. In the code system, a finite element method was introduced to analyze the fuel and cladding's mechanical behaviors. The FEAST code is more advanced code system for SFR which adopted mechanistic FGR and swelling model but still use analytical model to simulate fuel and cladding mechanical behavior. Based on the survey of the previous studies, fuel and cladding mechanical model should be improved. Analysis of mechanical behavior for fuel rod is crucial to evaluate overall rod's integrity. In addition, it is because contact between fuel slug and cladding or an over-pressure of rod internal pressure can cause rod failure during steady-state and other operation condition. The most of reference codes have simplified mechanical analysis model, so called 'analytical mode', because the detailed mechanical analysis requires large amount of calculation time and computing power. Even

  7. Development of mechanical analysis module for simulation of SFR fuel rod behavior using finite element method

    International Nuclear Information System (INIS)

    Shin, Andong; Jeong, Hyedong; Suh, Namduk; Kim, Hyochan; Yang, Yongsik

    2014-01-01

    Korean SFR developer decided to adapt metal fuel, current study focused on the metal fuel instead of oxide fuel. The SFR metal fuel has been developed by Korea Atomic Energy Research Institute (KAERI) and many efforts focused on designing and manufacturing the metal fuel. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured during steady-state operation and accident condition within an acceptable range. Whereas the design and evaluation methodologies, code systems and test procedures of a light water reactor fuel are sufficiently established, those of the SFR fuel needs more technical advances. In the view of regulatory point, there are still many challenging issues which are required to secure the safety of fuel and reactors. For this reason, the Korea Institute of Nuclear Safety (KINS) has launched the new project to develop the regulatory technology for SFR system including a fuel area. The ALFUS code was developed by CRIEPI and employs mechanistic model for fission gas release and swelling of fuel slug. In the code system, a finite element method was introduced to analyze the fuel and cladding's mechanical behaviors. The FEAST code is more advanced code system for SFR which adopted mechanistic FGR and swelling model but still use analytical model to simulate fuel and cladding mechanical behavior. Based on the survey of the previous studies, fuel and cladding mechanical model should be improved. Analysis of mechanical behavior for fuel rod is crucial to evaluate overall rod's integrity. In addition, it is because contact between fuel slug and cladding or an over-pressure of rod internal pressure can cause rod failure during steady-state and other operation condition. The most of reference codes have simplified mechanical analysis model, so called 'analytical mode', because the detailed mechanical analysis requires large amount of calculation time and computing power. Even

  8. Design of FCI Experiments to Understand Fuel Out-Pin Phenomena in the SFR

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Park, Seong Dae [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook; Bang, In Cheol [Chungang Univ., Seoul (Korea, Republic of)

    2014-05-15

    It is important to guarantee a passive nuclear safety regarding enhanced negative reactivity by fragmenting the molten fuel. In the SFR, it has a strong point that the negative reactivity is immediately introduced when the metal fuel is melted by the UTOP or ULOP accident. These characteristics of the metal fuel can prevent from progressing in severe accidents such as core disruptive accidents (CDA). As key phenomena in the accidents, fuel-coolant interaction (FCI) phenomena have been studied over the last few decades. Especially, several previous researches focused on instability and fragmentation of a core melt jet in water. However, the studies showed too limited phenomena to fully understand. In the domestic SFR technology development, researches for severe accidents tend to lag behind ones of other countries. Or, South Korea has a very basic level of the research such as literature survey. Recently, the SAS4A code, which was developed at Argonne National Laboratory (ANL) for thermal-hydraulic and neutronic analyses of power and flow transients in liquid-metal-cooled nuclear reactors (LMRs), is still under development to consider for a metal fuel. The other countries carried out basic experiments for molten fuel and coolant interactions. However, in a high temperature condition, methods for analysis of structural interaction between molten fuel and fuel cladding are very limited. The ultimate objective of the study is to evaluate the possibility of recriticality accident induced by fuel-coolant interaction in the SFR adopting metal fuel. It is a key point to analyze the molten-fuel behavior based on the experimental results which show fuel-coolant interaction with the simulant materials. It is necessary to establish the test facility, to build database, and to develop physical models to understand the FCI phenomena in the SFR; molten fuel-coolant interaction as soon as the molten fuel is ejected to the sodium coolant channel and molten fuel-coolant interaction

  9. Operation and Performance of the Supercritical Fluids Reactor (SFR)

    National Research Council Canada - National Science Library

    Hanush, R

    1996-01-01

    The Supercritical Fluids Reactor (SFR) at Sandia National Laboratories, CA has been developed to examine and solve engineering, process, and fundamental chemistry issues regarding the development of supercritical water oxidation (SCWO...

  10. Mechanical modelling of transient- to- failure SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L. E.

    2014-07-01

    The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)

  11. Improvement of Cycle Dependent Core Model for NPP Simulator

    International Nuclear Information System (INIS)

    Song, J. S.; Koo, B. S.; Kim, H. Y. and others

    2003-11-01

    The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations

  12. Improvement of Cycle Dependent Core Model for NPP Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Song, J. S.; Koo, B. S.; Kim, H. Y. and others

    2003-11-15

    The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations.

  13. Weld Joint Design for SFR Metallic Fuel Element Closures

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Kim, Ki Hwan; Yoon, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The sodium-cooled fast reactor (SFR) system is among the six systems selected for Gen-IV promising systems and expected to become available for commercial introduction around 2030. In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the joint designs for endplug welding were investigated. For the irradiation test of SFR metallic fuel element, the TIG welding technique was adopted and the welding joint design was developed based on the welding conditions and parameters established. In order to make SFR metallic fuel elements, the weld joint design was developed based on the TIG welding technique.

  14. Safety analysis SFR 1. Long-term safety

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

    An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

  15. Safety analysis SFR 1. Long-term safety

    International Nuclear Information System (INIS)

    2008-12-01

    An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

  16. Safety analysis SFR 1. Long-term safety

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

    An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the

  17. Review of the Conceptual Design for In-Vessel Fuel Handling Machines in SFR

    International Nuclear Information System (INIS)

    Kim, S. H.; Koo, G. H.

    2012-01-01

    The main in-vessel fuel handling machines in sodium cooled fast reactor(SFR) are composed of the in-vessel transfer machine(IVTM) and the rotating plug. These machines perform the function to handle fuel assemblies inside the reactor core during the refueling time. The IVTM should be able to access all areas above the reactor core and the fuel transfer port which can discharge the fuel assembly by the rotation of the rotating plug. In the 600 MWe demonstration reactor, the conceptual design of the in-vessel fuel handling machines was carried out. As shown in Fig. 1, the invessel fuel handling machines of the demonstration reactor are the double rotating plug type. With reference to the given core configuration of the demonstration reactor, the arrangement design of the rotating plug was carried out by using the developed simulation program. At present, the conceptual design of SFR prototype reactor which has small capacity of about 100 MWe is being started. Thus, it is necessary the economical efficiency and the reliability of the in-vessel fuel handling machines are reviewed according to the reduction of the power capacity. In this study, the preliminary design concepts of the main invessel fuel handling machines according to the fuel handling type are compared. Also, the design characteristics for the driving mechanism of the IVTM in the demonstration reactor and the recovery concept from the malfunction are reviewed

  18. Project Safe. Gas related processes in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno, L. [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Chemical Engineering; Skagius, K.; Soedergren, S.; Wiborgh, M. [Kemakta Konsult AB, Stockholm (Sweden)

    2001-06-01

    The radionuclide release from the SFR repository caused by gas generation was calculated for different scenarios for three repository parts (Silo, BMA and 1BTF). The calculation cases are based on the way the gas escapes from the concrete structures. In the basic cases the gas escapes through the evacuation pipes in the concrete lid of the Silo, through existing gaps between the concrete walls and the lid in BMA, and through the concrete backfill surrounding the waste packages in 1BTF. These cases correspond to the situation that we expect to occur. Another category of cases corresponds to the situation where an initial fracture exists in the concrete structures. The fracture is assumed to exist at the bottom of the respective concrete structure in the Silo and BMA. For 1BTF the initial defect is represented by a fracture transversely crossing the section containing the steel drums with ashes. Other cases were also calculated with the purpose of studying some special situations. For example, the consequences of a silo repository without evacuation pipes and backfill in the interior of BMA. The radionuclide release, for some radionuclides, may be increased by several orders of magnitude when contaminated water is expelled by gas from the interior of the concrete structures. However, the impact on the total doses during the first thousands years after closure of the repository is limited. The total dose is dominated by the release of organic {sup 14}C. Since the radionuclides are released to the coastal area during the first thousand years the dilution is considerable, which results in a very low dose.

  19. Project Safe. Gas related processes in SFR

    International Nuclear Information System (INIS)

    Moreno, L.

    2001-06-01

    The radionuclide release from the SFR repository caused by gas generation was calculated for different scenarios for three repository parts (Silo, BMA and 1BTF). The calculation cases are based on the way the gas escapes from the concrete structures. In the basic cases the gas escapes through the evacuation pipes in the concrete lid of the Silo, through existing gaps between the concrete walls and the lid in BMA, and through the concrete backfill surrounding the waste packages in 1BTF. These cases correspond to the situation that we expect to occur. Another category of cases corresponds to the situation where an initial fracture exists in the concrete structures. The fracture is assumed to exist at the bottom of the respective concrete structure in the Silo and BMA. For 1BTF the initial defect is represented by a fracture transversely crossing the section containing the steel drums with ashes. Other cases were also calculated with the purpose of studying some special situations. For example, the consequences of a silo repository without evacuation pipes and backfill in the interior of BMA. The radionuclide release, for some radionuclides, may be increased by several orders of magnitude when contaminated water is expelled by gas from the interior of the concrete structures. However, the impact on the total doses during the first thousands years after closure of the repository is limited. The total dose is dominated by the release of organic 14 C. Since the radionuclides are released to the coastal area during the first thousand years the dilution is considerable, which results in a very low dose

  20. The terrestrial biosphere in the SFR region

    Energy Technology Data Exchange (ETDEWEB)

    Jerling, L; Isaeus, M [Stockholm Univ. (Sweden). Dept. of Botany; Lanneck, J [Stockholm Univ. (Sweden). Dept. of Physical Geography; Lindborg, T; Schueldt, R [Danish Nature Council, Copenhagen (Denmark)

    2001-03-01

    This report is a part of the SKB project 'SAFE' (Safety Assessment of the Final Repository of Radioactive Operational Waste). The aim of project SAFE is to update the previous safety analysis of SFR-1.SFR-1 is a facility for disposal of low and intermediate level radioactive waste, which is situated in bedrock beneath the Baltic Sea, one km off the coast near the Forsmark nuclear power plant in Northern Uppland. A part of the SAFE-analysis aims at analysing the transport of radionuclides in the ecosystems.To do so one has to build a model that includes a large amount of information concerning the biosphere.The first step is to collect and compile descriptions of the biosphere.This report is a first attempt to characterise the terrestrial environment of the SFR area of Forsmark. In the first part of the report the terrestrial environment, land class distribution and production of the area is described. The primary production in different terrestrial ecosystems is estimated for a model area in the Forsmark region. The estimations are based on the actual land class distribution and the values for the total primary production (d.w. above ground biomass)and the amount carbon produced, presented as g/m{sup 2} for each land class respectively. An important aspect of the biosphere is the vegetation and its development. The future development of vegetation is of interest since production,decomposition and thus storage of organic material, vary strongly among vegetation types and this has strong implications for the transport of radionuclides.Therefore an attempt to describe the development of terrestrial vegetation has been made in the second part. Any prediction of future vegetation is based on knowledge of the past together with premises for the future development.The predictions made, thus, becomes marred with errors enforced by the assumptions and incomplete information of the past. The assumptions made for the predictions in this report are crude and results in a

  1. The terrestrial biosphere in the SFR region

    International Nuclear Information System (INIS)

    Jerling, L.; Isaeus, M.

    2001-03-01

    This report is a part of the SKB project 'SAFE' (Safety Assessment of the Final Repository of Radioactive Operational Waste). The aim of project SAFE is to update the previous safety analysis of SFR-1.SFR-1 is a facility for disposal of low and intermediate level radioactive waste, which is situated in bedrock beneath the Baltic Sea, one km off the coast near the Forsmark nuclear power plant in Northern Uppland. A part of the SAFE-analysis aims at analysing the transport of radionuclides in the ecosystems.To do so one has to build a model that includes a large amount of information concerning the biosphere.The first step is to collect and compile descriptions of the biosphere.This report is a first attempt to characterise the terrestrial environment of the SFR area of Forsmark. In the first part of the report the terrestrial environment, land class distribution and production of the area is described. The primary production in different terrestrial ecosystems is estimated for a model area in the Forsmark region. The estimations are based on the actual land class distribution and the values for the total primary production (d.w. above ground biomass)and the amount carbon produced, presented as g/m 2 for each land class respectively. An important aspect of the biosphere is the vegetation and its development. The future development of vegetation is of interest since production,decomposition and thus storage of organic material, vary strongly among vegetation types and this has strong implications for the transport of radionuclides.Therefore an attempt to describe the development of terrestrial vegetation has been made in the second part. Any prediction of future vegetation is based on knowledge of the past together with premises for the future development.The predictions made, thus, becomes marred with errors enforced by the assumptions and incomplete information of the past. The assumptions made for the predictions in this report are crude and results in a coarse

  2. The terrestrial biosphere in the SFR region

    Energy Technology Data Exchange (ETDEWEB)

    Jerling, L.; Isaeus, M. [Stockholm Univ. (Sweden). Dept. of Botany; Lanneck, J. [Stockholm Univ. (Sweden). Dept. of Physical Geography; Lindborg, T.; Schueldt, R. [Danish Nature Council, Copenhagen (Denmark)

    2001-03-01

    This report is a part of the SKB project 'SAFE' (Safety Assessment of the Final Repository of Radioactive Operational Waste). The aim of project SAFE is to update the previous safety analysis of SFR-1.SFR-1 is a facility for disposal of low and intermediate level radioactive waste, which is situated in bedrock beneath the Baltic Sea, one km off the coast near the Forsmark nuclear power plant in Northern Uppland. A part of the SAFE-analysis aims at analysing the transport of radionuclides in the ecosystems.To do so one has to build a model that includes a large amount of information concerning the biosphere.The first step is to collect and compile descriptions of the biosphere.This report is a first attempt to characterise the terrestrial environment of the SFR area of Forsmark. In the first part of the report the terrestrial environment, land class distribution and production of the area is described. The primary production in different terrestrial ecosystems is estimated for a model area in the Forsmark region. The estimations are based on the actual land class distribution and the values for the total primary production (d.w. above ground biomass)and the amount carbon produced, presented as g/m{sup 2} for each land class respectively. An important aspect of the biosphere is the vegetation and its development. The future development of vegetation is of interest since production,decomposition and thus storage of organic material, vary strongly among vegetation types and this has strong implications for the transport of radionuclides.Therefore an attempt to describe the development of terrestrial vegetation has been made in the second part. Any prediction of future vegetation is based on knowledge of the past together with premises for the future development.The predictions made, thus, becomes marred with errors enforced by the assumptions and incomplete information of the past. The assumptions made for the predictions in this report are crude and results

  3. Tailoring Sandwich Face/Core Interfaces for Improved Damage Tolerance

    DEFF Research Database (Denmark)

    Lundsgaard-Larsen, Christian; Berggreen, Christian; Carlsson, Leif A.

    2010-01-01

    Various modifications of the face/core interface in foam core sandwich specimens are examined in a series of two papers. This paper constitutes part I and describes the finite element analysis of a sandwich test specimen, i.e. a DCB specimen loaded by uneven bending moments (DCB-UBM). Using...... this test almost any mode-mixity between pure mode I and mode II can be obtained. A cohesive zone model of the mixed mode fracture process involving large-scale bridging is developed. Results from the analysis are used in Part II, which describes methods and results of a series of experiments....

  4. An improved one-and-a-half group BWR core simulator for a new-generation core management system

    International Nuclear Information System (INIS)

    Iwamoto, Tatsuya; Yamamoto, Munenari

    2000-01-01

    An improved one-and-a-half group core simulator method for a next-generation BWR core management system is presented. In the improved method, intranodal spectral index (thermal to fast flux ratio) is expanded with analytic solutions to the diffusion equation, and the nodal power density and the interface net current are calculated, taking the intranodal flux shape into consideration. A unique method was developed for assembly heterogeneity correction. Thus eliminating the insufficiencies of the conventional one-and-a-half group method, we can have accurate power distributions as well as local peaking factors for cores having large spectral mismatch between fuel assemblies. The historical effects of spectral mismatch are also considered in both nodal power and local peaking calculations. Although reflectors are not solved explicitly, there is essentially no need for core dependent adjustable parameters, since boundary conditions are derived in the same manner as in the interior nodes. Calculation time for nodal solutions is comparable to that for the conventional method, and is less than 1/10 of a few-group nodal simulator. Verifications of the present method were made by comparing the results with those obtained by heterogeneous fine-mesh multi-group core depletion calculations, and the accuracy was shown to be fairly good. (author)

  5. Endplug Welding Techniques developed for SFR Metallic Fuel Elements

    International Nuclear Information System (INIS)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan

    2013-01-01

    In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established

  6. Endplug Welding Techniques developed for SFR Metallic Fuel Elements

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established.

  7. Project SAFE. Update of the SFR-1 safety assessment. Phase 1

    International Nuclear Information System (INIS)

    Andersson, Johan; Riggare, P.; Skagius, K.

    1998-10-01

    SFR-1 is a facility for disposal of low-level radioactive operational waste from the nuclear power plants in Sweden. Low-level radioactive waste from industry, medicine, and research is also disposed in SFR-1. The facility is situated in bedrock beneath the Baltic Sea, 1 km off the coast near the Forsmark nuclear power plant. SFR-1 was built between the years 1983 and 1988. An assessment of the long-term performance of the facility was included in the vast documentation that was a part of the application for an operational license. The assessment was presented in the form of a final safety report. In the operational licence for SFR-1 it is stated that renewed safety assessments should be carried out at least each ten years. In order to meet this demand SKB has launched a special project, SAFE (Safety Assessment of Final Disposal of Operational Radioactive Waste). The aim of the project is to update the safety analysis and to prepare a safety report that will be presented to the Swedish authorities not later than year 2000. Project SAFE is divided into three phases. The first phase is a prestudy, and the results of the prestudy are given in this report. The aim of the prestudy is to identify issues where additional studies would improve the basis for the updated safety analysis as well as to suggest how these studies should be carried out. The work has been divided into six different topics, namely the inventory, the near field, the far field, the biosphere, radionuclide transport calculations and scenarios. For each topic the former safety reports and regulatory reviews are scrutinised and needs for additional work is identified. The evaluations are given in appendices covering the respective topics. The main report is a summary of the appendices with a more stringent description of the repository system and the processes that are of interest and therefore should be addressed in an updated safety assessment. However, it should be pointed out that one of the

  8. Site investigation SFR. Water-rock interaction and mixing modelling in the SFR

    Energy Technology Data Exchange (ETDEWEB)

    Gimeno, Maria J.; Auque, Luis F.; Gomez, Javier B.; Acero, Patricia (University of Zaragoza (Spain))

    2011-10-15

    During 2008, the Swedish Nuclear Fuel and Waste Management Company (SKB) initiated an investigation programme for a future expansion of the final repository for low and medium level radioactive operational waste, SFR, located about 150 km north of Stockholm. The purpose of the investigations was to define and characterise a bedrock volume large enough to allow further storage of operational waste from existing Swedish nuclear power plants and future waste from the decommissioning and dismantling of nuclear power plant reactors (SKB 2008). Of several alternatives, a selected location was investigated southwest of the present SFR tunnel system. As part of the SFR Site Descriptive Model, the objective of the hydrogeochemical site description is to describe the chemistry, origin and distribution of groundwaters in the bedrock and the hydrogeochemical processes involved in their evolution. Hydrogeochemical information (salinity distribution, groundwater residence time, palaeohydrogeochemical input, etc.) are also of importance to help constrain the hydrogeological descriptive model. The hydrogeochemical modelling work has been performed in three steps, resulting in three model versions (0.1, 0.2 and 1.0). In versions 0.1 and 0.2, explorative analyses using traditional geochemical approaches (trend plots, x-y scatter plots, 3D visualisations, etc.) were performed to describe the data and to provide an early insight and understanding of the site. The final hydrogeochemical site description version 1.0 (Nilsson et al. 2011) includes data from the previous versions, as well as subsequent complementary data from the SFR extension project, and all these data are further evaluated using additional modelling approaches and techniques. In this context, the present report gives a more detailed analysis of the available data for some hydrogeochemical systems and a detailed description of the results of the geochemical and statistical modelling. One of the main aims is to establish

  9. Site investigation SFR. Water-rock interaction and mixing modelling in the SFR

    International Nuclear Information System (INIS)

    Gimeno, Maria J.; Auque, Luis F.; Gomez, Javier B.; Acero, Patricia

    2011-10-01

    During 2008, the Swedish Nuclear Fuel and Waste Management Company (SKB) initiated an investigation programme for a future expansion of the final repository for low and medium level radioactive operational waste, SFR, located about 150 km north of Stockholm. The purpose of the investigations was to define and characterise a bedrock volume large enough to allow further storage of operational waste from existing Swedish nuclear power plants and future waste from the decommissioning and dismantling of nuclear power plant reactors (SKB 2008). Of several alternatives, a selected location was investigated southwest of the present SFR tunnel system. As part of the SFR Site Descriptive Model, the objective of the hydrogeochemical site description is to describe the chemistry, origin and distribution of groundwaters in the bedrock and the hydrogeochemical processes involved in their evolution. Hydrogeochemical information (salinity distribution, groundwater residence time, palaeohydrogeochemical input, etc.) are also of importance to help constrain the hydrogeological descriptive model. The hydrogeochemical modelling work has been performed in three steps, resulting in three model versions (0.1, 0.2 and 1.0). In versions 0.1 and 0.2, explorative analyses using traditional geochemical approaches (trend plots, x-y scatter plots, 3D visualisations, etc.) were performed to describe the data and to provide an early insight and understanding of the site. The final hydrogeochemical site description version 1.0 (Nilsson et al. 2011) includes data from the previous versions, as well as subsequent complementary data from the SFR extension project, and all these data are further evaluated using additional modelling approaches and techniques. In this context, the present report gives a more detailed analysis of the available data for some hydrogeochemical systems and a detailed description of the results of the geochemical and statistical modelling. One of the main aims is to establish

  10. Challenges in mechanical modeling of SFR fuel rod transient behavior

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L. E.

    2013-07-01

    Modeling of SFR fuel rod mechanical behavior under transient conditions entails the development of a creep law to predict cladding viscoplastic strain. In this regard, this work is focused on defining a proper clad creep law structure as the basis to set a suitable model under SFR off-normal conditions as transient overpower and loss of fluid. To do so, a review of in-codes clad creep models has been done by using SAS-SFR, SCANAIR and ASTEC. The proposed creep model has been structured in two parts: viscoplastic behaviour before the failure (primary and secondary creep) and the failure due to viscoplastic collapse (tertiary creep). In order to model the first part, Norton creep law has been proposed as a conservative option. An irradiation hardening factor should be included for best estimate calculations. The recommendation for the second part is to apply a failure criterion based on strain limit or rupture time, which allows achieving conservative results.

  11. Economic Analysis of Pyro-SFR Fuel Cycle

    International Nuclear Information System (INIS)

    Gao, Fanxing; Park, Byungheung; Kwon, Eunha; Ko, Wonil

    2010-01-01

    In this study, based on the material flow the economics of Pyro-SFR has been estimated. These are mainly two methodologies to perform nuclear fuel cycle cost study which is based on the material flow calculations. One is equilibrium model and the other is dynamic model. Equilibrium model focus on the batch study with the assumptions that the whole system is in a steady state and mass flow as well as the electricity production all through the fuel cycle is in equilibrium state, which calculates the electricity production within a certain period and associated material flow with reference to unit cost in order to obtain the cost of electricity. Dynamic model takes the time factor into consideration to simulate the actual cases. Compared with the dynamic analysis model, the outcome of equilibrium model is more theoretical comparisons, especially with regard to the large uncertainty of the development of the pyro-technology evaluated. In this study equilibrium model was built to calculate the material flow on a batch basis. With the unit cost being determined, the cost of each step of fuel cycle could be obtained, so does the FMC. Due to the unavoidable uncertainty with certain unit costs, evaluated cost range and uncertainty study are applied. Sensitivity analysis has also been performed to obtain the breakeven uranium price for Pyro-SFR against PWR-O T. Economics of Pyro-SFR fuel cycle scenario has been calculated and compared by employing equilibrium model. The LFCC were obtained, Pyro-SFR 7.68 mills/kWh. The Uranium price is the dominant driver of LFCC. Pyro-techniques also weight considerably in Pyro-SFR scenario. On consideration of the current unavoidable uncertainties introduced by certain cost data, cost range and triangle techniques were used to perform the uncertainty study which indicates that the gap between Pyro-SFR and PWR-O T fuel cycle scenario is relatively small

  12. Improvements and Revamping of In-Core Instrumentation Systems

    International Nuclear Information System (INIS)

    Garam, Eric De

    1993-01-01

    The results of the improvements done by Fumarate Tia in these domains are really satisfying and it is clear that the problems of leakage existing in the old units on the thermocouples sealing systems and on the seal table of the income instrumentation, cannot exist on the French Units or on the units equipped with Fumarate Tia equipment. At the same time, all the equipment which constitute the Income instrumentation have been improved with the aim of reliability and safety. The equipment used to perform maintenance activities has also been improved to both reduce doses and increase efficiency. The purpose of this paper is to describe the principal improvements and revamping of income instrumentation systems, and to summarize the principal lessons learned from our experience on all designs of PWR. Fumarate Tia, is permanently looking for improving the existing systems of instrumentation with the aim of reduction of the dosimetry during the maintenance services, improvement of the liability and lifetime of the equipment, and of course reduction of the duration of the outages in keeping always the same level of quality

  13. Improvements to the sodium supply system of a nuclear reactor core

    International Nuclear Information System (INIS)

    Chevallier, Rene; Marchais, Christian.

    1981-01-01

    This invention concerns an improvement to the sodium supply system of a nuclear reactor core and, in particular, concerns the area included between the outlet of the primary circulation pumps and the core proper. A simplified structure and a lightening of all this linking area between the circulation pumps and the distribution tank under the core is achieved and this results in a very significant reduction in the risks of deterioration and in a definite increase in the reliability of the reactor. The invention is therefore an improvement to the sodium supply system of the nuclear reactor core vessel with incorporated exchangers, in which the cool sodium, after passing through the primary exchangers, is collected in a ring compartment from whence it is taken up by the pumps and moved to at least one pipe reaching a distribution tank located under the reactor core [fr

  14. Optimized design of an ex-vessel cooling thermosyphon for decay heat removal in SFR

    International Nuclear Information System (INIS)

    Choi, Jae Young; Jeong, Yong Hoon; Song, Sub Lee; Chang, Soon Heung

    2017-01-01

    Passive decay heat removal and sodium fire are two major key issues of nuclear safety in sodium-cooled fast reactor (SFR). Several decay heat removal systems (DHR) were suggested for SFR around the world so far. Those DHRS mainly classified into two concepts: Direct reactor cooling system and ex-vessel cooling system. Direct reactor cooling method represented by PDHRS from PGSFR has disadvantages on its additional in-vessel structure and potential sodium fire risk due to the sodium-filled heat exchanger exposed to air. Contrastively, ex-vessel cooling method represented by RVACS from PRISM has low decay heat removal performance, which cannot be applicable to large scale reactors, generally over 1000 MWth. No passive DHRSs which can solve both side of disadvantages has been suggested yet. The goal of this study was to propose ex-vessel cooling system using two-phase closed thermosyphon to compensate the disadvantages of the past DHRSs. Reference reactor was Innovative SFR (iSFR), a pool-type SFR designed by KAIST and featured by extended core lifetime and increased thermal efficiency. Proposed ex-vessel cooling system consisted of 4 trains of thermosyphons and designed to remove 1% of thermal power with 10% of margin. The scopes of this study were design of proposed passive DHRS, validation of system analysis and optimization of system design. Mercury was selected as working fluid to design ex-vessel thermosyphon in consideration of system geometry, operating temperature and required heat flux. SUS 316 with chrome coated liner was selected as case material to resist against high corrosivity of mercury. Thermosyphon evaporator was covered on the surface of reactor vessel as the geometry of hollow shell filled with mercury. Condenser was consisted of finned tube bundles and was located in isolated water pool, the ultimate heat sink. Operation limits and thermal resistance was estimated to guarantee whether the design was adequate. System analysis was conducted by in

  15. Tools and applications for core design and shielding in fast reactors

    International Nuclear Information System (INIS)

    Rachamin, Reuven

    2013-01-01

    Outline: • Modeling of SFR cores using the Serpent-DYN3D code sequence; • Core shielding assessment for the design of FASTEF-MYRRHA; • Neutron shielding studies on an advanced Molten Salt Fast Reactor (MSFR) design

  16. Preliminary Analysis of the Fuel Bundle Stiffness by ANSYS for SFR

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-15

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

  17. Investigation of the use of thorium in LWRs for improving reactor core performance

    International Nuclear Information System (INIS)

    Lau, Cheuk Wah

    2012-01-01

    Thorium is a fertile material and most of the past research has focused on breeding thorium into fissile material to achieve a more sustainable use of nuclear power. However, the focus in this report is on using thorium to improve reactor core performance. The improvement of reactor core performance is achieved by increasing the thermal margins by homogeneously distributing thorium in the fuel pellets. A proposed uranium-thorium-based fuel assembly is simulated for the Swedish Ringhals-3 PWR core in a realistic demonstration. In order to fully grasp the benefits and drawbacks of the newly proposed uranium-thorium-based fuel, a reload safety evaluation has been performed. For a real core, the Swedish Radiation Safety Authority would require an identical evaluation method to ensure that safety criteria are met during the whole cycle. In this report, only a few key safety parameters, such as isothermal- and Doppler-temperature coefficients of reactivity, pin peak power, boron worth, shutdown margins, and core average beta-effective are presented. The calculations were performed by the two-dimensional transport code CASMO-4E, and the two group three dimensional nodal code SIMULATE-3K from Studsvik Scandpower. The results showed that the uranium-thorium-based fuel assembly improves the thermal margins, both in the pin peak power and the local power (Fq). The improved thermal margins would allow more flexible core loading patterns with less neutron leakage, and could be used in power uprated cores to offer better safety margins

  18. Investigation of the use of thorium in LWRs for improving reactor core performance

    Energy Technology Data Exchange (ETDEWEB)

    Lau, Cheuk Wah

    2012-07-01

    Thorium is a fertile material and most of the past research has focused on breeding thorium into fissile material to achieve a more sustainable use of nuclear power. However, the focus in this report is on using thorium to improve reactor core performance. The improvement of reactor core performance is achieved by increasing the thermal margins by homogeneously distributing thorium in the fuel pellets. A proposed uranium-thorium-based fuel assembly is simulated for the Swedish Ringhals-3 PWR core in a realistic demonstration. In order to fully grasp the benefits and drawbacks of the newly proposed uranium-thorium-based fuel, a reload safety evaluation has been performed. For a real core, the Swedish Radiation Safety Authority would require an identical evaluation method to ensure that safety criteria are met during the whole cycle. In this report, only a few key safety parameters, such as isothermal- and Doppler-temperature coefficients of reactivity, pin peak power, boron worth, shutdown margins, and core average beta-effective are presented. The calculations were performed by the two-dimensional transport code CASMO-4E, and the two group three dimensional nodal code SIMULATE-3K from Studsvik Scandpower. The results showed that the uranium-thorium-based fuel assembly improves the thermal margins, both in the pin peak power and the local power (Fq). The improved thermal margins would allow more flexible core loading patterns with less neutron leakage, and could be used in power uprated cores to offer better safety margins.

  19. Approach to improve the axial power distribution for the application of a core protection system

    International Nuclear Information System (INIS)

    Koo, Bon Seung; Cho, Jin Young; Song, Jae Seung; Lee, Chung Chan

    2008-01-01

    A Core Protection Calculator System (CPCS) is a digital computer based on a safety system for generating trip signals based on a calculation of the Departure from Nucleate Boiling Ratio (DNBR) and the Local Power Density (LPD) by using several on-line measured system parameters including 3-level ex-core detector signals. A few approaches to improve the axial power distribution for the application of a core protection system were performed. For the Yonggwang unit 3 (cycle 1), axial power distributions were synthesized by applying the cubic spline method and compared with the neutronics code results. Several new cubic spline function sets were generated for the drastically distorted axial shapes for a 3-level ex-core detector system. In addition, synthesized axial shapes with a 5-level ex-core detector signals were compared with the conventional 3-level detector results. It demonstrates that the newly generated function sets appear to be better than that of the conventional CPC from the aspect of an axial power synthesis, particularly for the heavily distorted shapes. Moreover, synthesis of an axial power distribution using 5-level ex-core detector signals appears to be better than that of the 3-level ex-core detector signals. From the above results, improvement of the thermal margin is expected because of an uncertainty decreasing a core protection system. (authors)

  20. A study on improving the performance of a research reactor's equilibrium core

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2013-01-01

    Full Text Available Utilizing low enriched uranium silicide fuel (U3Si2-Al of existing uranium density (3.285 g/cm3, different core configurations have been studied in search of an equilibrium core with an improved performance for the Pakistan Research Reactor-1. Furthermore, we have extended our analysis to the performance of higher density silicide fuels with a uranium density of 4.0 and 4.8 U g/cm3. The criterion used in selecting the best performing core was that of “unit flux time cycle length per 235U mass per cycle”. In order to analyze core performance by improving neutron moderation, utilizing higher-density fuel, the effect of the coolant channel width was also studied by reducing the number of plates in the standard/control fuel element. Calculations employing computer codes WIMSD/4 and CITATION were performed. A ten energy group structure for fission neutrons was used for the generation of microscopic cross-sections through WIMSD/4. To search the equilibrium core, two-dimensional core modelling was performed in CITATION. Performance indicators have shown that the higher-density uranium silicide-fuelled core (U density 4.8 g/cm3 without any changes in standard/control fuel elements, comprising of 15 standard and 4 control fuel elements, is the best performing of all analyzed cores.

  1. Electrical Core Transformer for Grid Improvement Incorporating Wire Magnetic Components

    Energy Technology Data Exchange (ETDEWEB)

    Harrie R. Buswell, PhD; Dennis Jacobs, PhD; Steve Meng

    2012-03-26

    The research reported herein adds to the understanding of oil-immersed distribution transformers by exploring and demonstrating potential improvements in efficiency and cost utilizing the unique Buswell approach wherein the unit is redesigned, replacing magnetic sheet with wire allowing for improvements in configuration and increased simplicity in the build process. Exploration of new designs is a critical component in our drive to assure reduction of energy waste, adequate delivery to the citizenry, and the robustness of U.S. manufacturing. By moving that conversation forward, this exploration adds greatly to our base of knowledge and clearly outlines an important avenue for further exploration. This final report shows several advantages of this new transformer type (outlined in a report signed by all of our collaborating partners and included in this document). Although materials development is required to achieve commercial potential, the clear benefits of the technology if that development were a given is established. Exploration of new transformer types and further work on the Buswell design approach is in the best interest of the public, industry, and the United States. Public benefits accrue from design alternatives that reduce the overall use of energy, but it must be acknowledged that new DOE energy efficiency standards have provided some assurance in that regard. Nonetheless the burden of achieving these new standards has been largely shifted to the manufacturers of oil-immersed distribution transformers with cost increasing up to 20% of some units versus 2006 when this investigation was started. Further, rising costs have forced the industry to look closely are far more expensive technologies which may threaten U.S. competitiveness in the distribution transformer market. This concern is coupled with the realization that many units in the nation's grid are beyond their optimal life which suggests that the nation may be headed for an infrastructure

  2. Improvement of Axial Reflector Cross Section Generation Model for PWR Core Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Cheon Bo; Lee, Kyung Hoon; Cho, Jin Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This paper covers the study for improvement of axial reflector XS generation model. In the next section, the improved 1D core model is represented in detail. Reflector XS generated by the improved model is compared to that of the conventional model in the third section. Nuclear design parameters generated by these two XS sets are also covered in that section. Significant of this study is discussed in the last section. Two-step procedure has been regarded as the most practical approach for reactor core designs because it offers core design parameters quite rapidly within acceptable range. Thus this approach is adopted for SMART (System-integrated Modular Advanced Reac- Tor) core design in KAERI with the DeCART2D1.1/ MASTER4.0 (hereafter noted as DeCART2D/ MASTER) code system. Within the framework of the two-step procedure based SMART core design, various researches have been studied to improve the core design reliability and efficiency. One of them is improvement of reflector cross section (XS) generation models. While the conventional FA/reflector two-node model used for most core designs to generate reflector XS cannot consider the actual configuration of fuel rods that intersect at right angles to axial reflectors, the revised model reflects the axial fuel configuration by introducing the radially simplified core model. The significance of the model revision is evaluated by observing HGC generated by DeCART2D, reflector XS, and core design parameters generated by adopting the two models. And it is verified that about 30 ppm CBC error can be reduced and maximum Fq error decreases from about 6 % to 2.5 % by applying the revised model. Error of AO and axial power shapes are also reduced significantly. Therefore it can be concluded that the simplified 1D core model improves the accuracy of the axial reflector XS and leads to the two-step procedure reliability enhancement. Since it is hard for core designs to be free from the two-step approach, it is necessary to find

  3. Visualization Study of Melt Dispersion Behavior for SFR with a Metallic Fuel under Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo Heo; Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Jungang Univ., Seoul (Korea, Republic of)

    2015-05-15

    The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition.

  4. Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ, Seoul (Korea, Republic of)

    2015-10-15

    In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

  5. Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

    International Nuclear Information System (INIS)

    Heo, Hyo; Bang, In Cheol; Jerng, Dong Wook

    2015-01-01

    In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

  6. Transitions to improved core electron heat confinement in JT-II plasmas

    International Nuclear Information System (INIS)

    Estrada, T.; Medina, F.; Ascasibar, E.; Balbin, R.; Castejon, F.; Hidalgo, C.; Lopez-Bruna, D.; Petrov, S.

    2008-01-01

    Transitions to improved core electron heat confinement are triggered by low order rational magnetic surfaces in TJ-II ECH plasmas. Transitions triggered by the rational surface n=4/m=2 show an increase in the ion temperature synchronized with the increase in the electron temperature. SXR measurements demonstrate that, under certain circumstances, the rational surface positioned inside the plasma core region precedes and provides a trigger for the transition. (author)

  7. Remote Core Locking: Migrating Critical-Section Execution to Improve the Performance of Multithreaded Applications

    OpenAIRE

    Lozi , Jean-Pierre; David , Florian; Thomas , Gaël; Lawall , Julia; Muller , Gilles

    2014-01-01

    National audience; The scalability of multithreaded applications on current multicore systems is hampered by the performance of lock algorithms, due to the costs of access contention and cache misses. In this paper, we propose a new lock algorithm, Remote Core Locking (RCL), that aims to improve the performance of critical sections in legacy applications on multicore architectures. The idea of RCL is to replace lock acquisitions by optimized remote procedure calls to a dedicated server core. ...

  8. Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Benoit, Timothy [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hlotke, John Daniel [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2017-07-05

    This document contains an evaluation of the applicability of the current Quality Assurance Standards from the American Society of Mechanical Engineers Standard NQA-1 (NQA-1) criteria and identifies and describes the quality assurance process(es) by which attributes of historical, analytical, and other data associated with sodium-cooled fast reactor [SFR] metallic fuel and/or related reactor fuel designs and constituency will be evaluated. This process is being instituted to facilitate validation of data to the extent that such data may be used to support future licensing efforts associated with advanced reactor designs. The initial data to be evaluated under this program were generated during the US Integral Fast Reactor program between 1984-1994, where the data includes, but is not limited to, research and development data and associated documents, test plans and associated protocols, operations and test data, technical reports, and information associated with past United States Nuclear Regulatory Commission reviews of SFR designs.

  9. The biosphere today and tomorrow in the SFR area

    Energy Technology Data Exchange (ETDEWEB)

    Kautsky, Ulrik (ed.)

    2001-06-01

    This report is a compilation of the work done mainly in the SAFE project for the biosphere from about 14 reports. The SAFE project is the updated safety analysis of SFR-1, the LLW and ILW repository at Forsmark. The aim of the report is to summarize the available information about the present-day biosphere in the area surrounding SFR and to use this information, together with information about the previous development of the biosphere, to predict the future development of the area in a more comparable way than the underlying reports. The data actually used for the models have been taken from the original reports which also justify or validate the data. The report compiles information about climate, oceanography, landscape, sedimentation, shoreline displacement, marine, lake and terrestrial ecosystems.

  10. Radioactive Waste Generation in Pyro-SFR Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Gao, Fanxing; Park, Byung Heung; Ko, Won Il

    2011-01-01

    Which nuclear fuel cycle option to deploy is of great importance in the sustainability of nuclear power. SFR fuel cycle employing pyroprocessing (named as Pyro- SFR Cycle) is one promising fuel cycle option in the near future. Radioactive waste generation is a key criterion in nuclear fuel cycle system analysis, which considerably affects the future development of nuclear power. High population with small territory is one special characteristic of ROK, which makes the waste management pretty important. In this study, particularly the amount of waste generation with regard to the promising advanced fuel cycle option was evaluated, because the difficulty of deploying an underground repository for HLW disposal requires a longer time especially in ROK

  11. The biosphere today and tomorrow in the SFR area

    International Nuclear Information System (INIS)

    Kautsky, Ulrik

    2001-06-01

    This report is a compilation of the work done mainly in the SAFE project for the biosphere from about 14 reports. The SAFE project is the updated safety analysis of SFR-1, the LLW and ILW repository at Forsmark. The aim of the report is to summarize the available information about the present-day biosphere in the area surrounding SFR and to use this information, together with information about the previous development of the biosphere, to predict the future development of the area in a more comparable way than the underlying reports. The data actually used for the models have been taken from the original reports which also justify or validate the data. The report compiles information about climate, oceanography, landscape, sedimentation, shoreline displacement, marine, lake and terrestrial ecosystems

  12. Design Evaluation of UIS and In-vessel Fuel Transfer Machine for a 1200MWe SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Kim, Seok Hoon; Park, Chang Gyu; Lee, Su Yeon

    2008-11-15

    The report describes the structural applicability of the upper internal structure (UIS) and the in-vessel fuel transfer machine for a 1200MWe sodium cooled fast reactor (SFR) of a pool type. In the conceptual design, a two rotating plug type as a refueling system is considered. For the two rotating plug type, the diameters of large and small rotating plugs are determined by 7.2m and 5.6m, respectively. Through the use of an inner fuel transfer machine and the lift change machine with a fixed arm length of 1.10m installed on a small rotating plug, all the core assemblies are accessed within 7mm accuracy. The UIS diameter is determined by 4.7m, which includes the all control drive lines in upper part, the diameter of UIS lower part is restricted by 4.4 m to secure the rotation angle of a refueling machine.

  13. Review of C-14 inventory for the SFR facility

    International Nuclear Information System (INIS)

    Smith, Graham; Merino, Joan; Kerrigan, Emma

    2002-08-01

    The Swedish Radiation Protection Authority (SSI) is currently reviewing SKB's continuing assessment for disposal of radioactive waste to the SFR facility at Forsmark. Among the wastes disposed are reactor operating wastes. Among the relevant radionuclides is C-14, which is relatively difficult to measure and to control because of its mobility. This report documents a review of the C-14 inventory material submitted by SKB for the SFR-facility, to determine its validity and comment on the appropriate assumptions for C-14 content of wastes due to be disposed of to the SFR. The review is based on information provided by SSI as well as other relevant international experience. Conclusions are drawn upon: the chemical form of the C-14 in the waste from BWRs and PWRs; the production rate of C-14 in BWRs and PWRs and quantification of the source term in the IEX waste; the distribution of the C-14 in the IEX waste from BWR between the resins used for treatment of the primary cooling water and the resins used for treatment of the condensate water; quantification of the uncertainties. A suggestion is made that the C-14 inventory could be better developed based upon a mass balance assessment of all the C-14 produced in reactors, and its ultimate fate in effluent and solid wastes, taking account of the reactor specific operational factors identified in the review as relevant to C-14 inventory assessment

  14. Structure of fuel performance audit code for SFR metal fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong Sik; Kim, Hyo Chan [KAERI, Daejeon (Korea, Republic of); Jeong, Hye Dong; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-10-15

    A Sodium Cooled Fast Reactor (SFR) is a promising option to solve the spent fuel problems, but, there are still much technical issues to commercialize a SFR. One of issues is a development of advanced fuel which can solve the safety and the economic issues at the same time. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured. In Korea Institute of Nuclear Safety (KINS), the new project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. To develop the new code system, the code structure design and its requirements need to be studied. Various performance models and code systems are reviewed and their characteristics are analyzed in this paper. Based on this study, the fundamental performance models are deduced and basic code requirements and structure are established.

  15. Review of C-14 inventory for the SFR facility

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Graham; Merino, Joan; Kerrigan, Emma

    2002-08-01

    The Swedish Radiation Protection Authority (SSI) is currently reviewing SKB's continuing assessment for disposal of radioactive waste to the SFR facility at Forsmark. Among the wastes disposed are reactor operating wastes. Among the relevant radionuclides is C-14, which is relatively difficult to measure and to control because of its mobility. This report documents a review of the C-14 inventory material submitted by SKB for the SFR-facility, to determine its validity and comment on the appropriate assumptions for C-14 content of wastes due to be disposed of to the SFR. The review is based on information provided by SSI as well as other relevant international experience. Conclusions are drawn upon: the chemical form of the C-14 in the waste from BWRs and PWRs; the production rate of C-14 in BWRs and PWRs and quantification of the source term in the IEX waste; the distribution of the C-14 in the IEX waste from BWR between the resins used for treatment of the primary cooling water and the resins used for treatment of the condensate water; quantification of the uncertainties. A suggestion is made that the C-14 inventory could be better developed based upon a mass balance assessment of all the C-14 produced in reactors, and its ultimate fate in effluent and solid wastes, taking account of the reactor specific operational factors identified in the review as relevant to C-14 inventory assessment.

  16. Experimental Setup with Transient Behavior of Fuel Cladding of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Sang Hun; Kim, Jun Hwan; Kim, June-Hyung; Ryu, Woo Seog; Park, Sang Gyu; Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Nowadays, in Korea, advanced cladding such as FC92 is developed and its transient behaviors are required for the safety analysis of SFR. Design and safety analyses of sodium-cooled fast reactor (SFR) require understanding fuel pin responses to a wide range of off-normal events. In a loss-of-flow (LOF) or transient over-power (TOP), the temperature of the cladding is rapidly increased above its steady-state service temperature. Transient tests have been performed in sections of fuel pin cladding and a large data base has been established for austenitic stainless steel such as 20% cold-worked 316 SS and ferritic/martensitic steels such as HT9. This paper summarizes the technical status of transient testing facilities and their results. Previous researches showed the transient behaviors of HT9 cladding. For the safety analyses in SFR in Korea, simulated transient tests with newly developed FC92 as well as HT9 cladding are being carried out.

  17. Experimental results on improved JARE deep ice core drill-Experiments in Rikubetsu, Hokkaido in 2002 -

    Directory of Open Access Journals (Sweden)

    Takao Kameda

    2002-07-01

    Full Text Available Deep ice coring to bedrock (3028m in depth at Dome Fuji Station is planned during three successive summer seasons starting from 2003/2004. An improved JARE deep ice core drill (12.2m in length and 3.8m in maximum core length was developed in December 2001 for the ice coring at Dome Fuji. In January/February of 2002,we performed experiments on drill performance using artificial ice blocks in Rikubetsu, Hokkaido. In this paper, we outline the experiment and report the results. It was found through the experiment that an ice core of 3.8m length was smoothly obtained by the improved drill with three screws in the chip chamber and cutting pitch of 5mm/cycle. About 45000 small holes 1.2mm in diameter were made on the surface of the chip chamber. These small holes enabled liquid to circulate between cutters and outside of the drill through the chip chamber in the drill. The dry density of the chips was 440 to 500kg/m^3 and the chip recovery rate during ice coring was 65 to 91%. A check valve installed at the bottom of the chip chamber to prevent outflow of chips from the drill was not tested enough, but more durability is needed for the valve. The newly developed motor system and core catchers of the drill worked perfectly. The average coring speed was 24.5cm/min with cutting pitch of 5mm/cycle. The average power consumption during ice coring was 171W.

  18. SSI and SKI's Review of SKB's Updated Final Safety Report for SFR 1. Review Report

    International Nuclear Information System (INIS)

    2003-10-01

    The Repository for Radioactive Operational Waste (SFR 1) is now the object of a new review by the Swedish Radiation Protection Authority (SSI) and the Swedish Nuclear Power Inspectorate (SKI). One of the stipulations for operating SFR 1 was that a new assessment of the long-term performance and environmental consequences of the repository should be conducted once every 10 years by the licensee, the Swedish Nuclear Fuel and Waste Management Co (SKB). During the time that SFR 1 has been in operation, experience has been gained of operating the facility and new knowledge of long-term performance of SFR 1 has been obtained. New regulations for nuclear facilities have been promulgated since SFR 1 was taken into operation (1988). A review committee comprising employees from SKI and SSI has conducted the review of SSR 2001. This review report has resulted in the committee's evaluation of the safety of SFR 1 and is the basis of the regulatory authorities' decision concerning any amendments to the stipulations for the operation of SFR 1. However, the review has found deficiencies in the follow up of the development of design basis norms since the facility was constructed as well as deficiencies in learning from operating experience. However, the overall evaluation is that the facility is being operated in an acceptable manner from the standpoint of safety. With respect to the long-term performance of the repository, it is a deficiency that SSR 2001 does not describe how compliance with the stipulated radiation protection requirements on optimisation and use of the best available technology (BAT) is achieved during operation. In the opinion of the review committee, issues relating to occupational radiation protection are being handled satisfactorily and the operational releases of radioactive substances are very small. Safety and Radiation Protection after Closure SKB's long-term repository performance assessment contains essential updates and improvements compared with the

  19. Improvement in operating characteristics resulting from the addition of FLIP fuel to a standard TRIGA core

    International Nuclear Information System (INIS)

    Randall, J.D.; Feltz, D.E.; Godsey, T.A.; Schumacher, R.F.

    1974-01-01

    To overcome problems associated with fuel burnup the Nuclear Science Center of Texas A and M University decided to convert from standard TRIGA fuel to FLIP-TRIGA fuel. FLIP fuel, which incorporates erbium as a burnable poison and is enriched to 70 percent in U-235, has a calculated lifetime of 9/MW-years. Due to limited funds a core was designed with a central region of 35 FLIP elements surrounded by 63 standard elements. Calculations indicated that the core excess and neutron fluxes were satisfactory, but no prediction was made of the improvements in core lifetime. The reactivity loss due to burnup for a standard core was measured to be 1.54 cents/MW-day. The addition of 35 FLIP fuel elements has reduced this value to approximately 0.5 cents/MW-day. The incorporation of FLIP fuel has, therefore, increased the lifetime of the core by a factor of three using fuel that is only 20 percent more expensive. The mixed core has other advantages as well. The power coefficient is less, the effect of xenon is less, and the fluxes in experimental facilities are higher. Thus, the mixed core has significant advantages over standard TRIGA fuel. (U.S.)

  20. Improvement of open and semi-open core wall system in tall buildings by closing of the core section in the last story

    Science.gov (United States)

    Kheyroddin, A.; Abdollahzadeh, D.; Mastali, M.

    2014-09-01

    Increasing number of tall buildings in urban population caused development of tall building structures. One of the main lateral load resistant systems is core wall system in high-rise buildings. Core wall system has two important behavioral aspects where the first aspect is related to reduce the lateral displacement by the core bending resistance and the second is governed by increasing of the torsional resistance and core warping of buildings. In this study, the effects of closed section core in the last story have been considered on the behavior of models. Regarding this, all analyses were performed by ETABS 9.2.v software (Wilson and Habibullah). Considering (a) drift and rotation of the core over height of buildings, (b) total and warping stress in the core body, (c) shear in beams due to warping stress, (d) effect of closing last story on period of models in various modes, (e) relative displacement between walls in the core system and (f) site effects in far and near field of fault by UBC97 spectra on base shear coefficient showed that the bimoment in open core is negative in the last quarter of building and it is similar to wall-frame structures. Furthermore, analytical results revealed that closed section core in the last story improves behavior of the last quarter of structure height, since closing of core section in the last story does not have significant effect on reducing base shear value in near and far field of active faults.

  1. Study of the occurrence of organic matter, metals and chemicals in the SFR

    International Nuclear Information System (INIS)

    Sundqvist, J.O.

    2001-03-01

    Low- and intermediate level operational waste from the Swedish nuclear power plants, and the Studsvik facility, is currently placed in a repository, termed SFR-l (final repository for radioactive operational waste) near the Forsmark power plant. Two important components in the waste, which can affect the function of the repository, are organic materials, e.g. cloth and paper, and metals (scrap). The release of radionuclides from the repository may be affected by chemical reactions that involve both organic materials and metals. After sealing the repository, the conditions can be such that complexing agents (e.g. isosaccarinic acid) may form when organic materials degrade. These agents typically increase the mobility of radionuclides. Formation of gas, mainly due to metal corrosion, may affect the barrier system, surrounding the waste, such that the release of radionuclides is enhanced. SKB makes an annual report with a compilation of the waste that has been placed in SFR-l . The compilation contains both the amount of waste placed in the repository during the last year and a compilation of the waste that have been placed since the stall of SFR. Moreover, SKB provides a prognosis of the future situation in SFR-1 every third year. SKI (the Swedish Nuclear Power Inspectorate), is responsible for reviewing this reporting. This study was initiated with the purpose of evaluating the uncertainties in SKB's estimates of the amounts of organic matter, metals and chemicals in the waste in SFR- I. The estimates of the quantities of e.g. cellulose and metals in the waste are based on a method which is utilising what is called normal-containers. The waste is classified into certain waste categories. For each waste category there is a specified, presumed composition, named normal-container. The results of this study suggest that the documentation provided by SKB is lacking in some respects. There are for instance examples of incomplete notification of waste and container types

  2. Isolated core training improves sprint performance in national-level junior swimmers.

    Science.gov (United States)

    Weston, Matthew; Hibbs, Angela E; Thompson, Kevin G; Spears, Iain R

    2015-03-01

    To quantify the effects of a 12-wk isolated core-training program on 50-m front-crawl swim time and measures of core musculature functionally relevant to swimming. Twenty national-level junior swimmers (10 male and 10 female, 16±1 y, 171±5 cm, 63±4 kg) participated in the study. Group allocation (intervention [n=10], control [n=10]) was based on 2 preexisting swim-training groups who were part of the same swimming club but trained in different groups. The intervention group completed the core training, incorporating exercises targeting the lumbopelvic complex and upper region extending to the scapula, 3 times/wk for 12 wk. While the training was performed in addition to the normal pool-based swimming program, the control group maintained their usual pool-based swimming program. The authors made probabilistic magnitude-based inferences about the effect of the core training on 50-m swim time and functionally relevant measures of core function. Compared with the control group, the core-training intervention group had a possibly large beneficial effect on 50-m swim time (-2.0%; 90% confidence interval -3.8 to -0.2%). Moreover, it showed small to moderate improvements on a timed prone-bridge test (9.0%; 2.1-16.4%) and asymmetric straight-arm pull-down test (23.1%; 13.7-33.4%), and there were moderate to large increases in peak EMG activity of core musculature during isolated tests of maximal voluntary contraction. This is the first study to demonstrate a clear beneficial effect of isolated core training on 50-m front-crawl swim performance.

  3. Conceptual Design Study on Electromagnets of Control Rod Drive Mechanism of a SFR

    International Nuclear Information System (INIS)

    Lee, Jaehan; Koo, Gyeonghoi

    2013-01-01

    The prototype SFR has six primary control rod assemblies(CRAs) and three secondary shutdown assemblies. The primary control system is used for power control, burnup compensation and reactor shutdown in response to demands from the plant control or protection systems. This paper describes the design concept of primary control rod drive mechanism shortly, and performs the parametric design studies for the electromagnet device of the drive mechanism to maximize CRA gripping force. The electromagnetic core usually confines and guides the magnetic field. The major parameters influenced on the electromagnetic force are the geometry and arrangement of the electromagnet and armature for a given coil specification. A typical equation calculating the electromagnetic force for a solenoid type is represented in equation. The first one is the increasing of the flux cross section area (Α c , Α g ) in magnetic field connecting of air gap, armature and electromagnets. Secondly, the reducing of the path lengths (l c , l g ) of the armature and electromagnet makes the magnetic flux (Β) resistance to be low. An electromagnet field analyses are performed for the initial design values of the electromagnet device. The gripping force is about 3 times of CRA weight when one coil is power on. The parametric studies on air gap, core sizes configuring of the electromagnet cores are performed to maximize the electromagnetic force

  4. Assessment of the long-term safety for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Greis Dahlberg, Christina; Vahlund, Frederik [Svensk Kaernbraenslehantering AB, Stockholm (Sweden)

    2015-07-01

    During operation and decommissioning of the Swedish nuclear facilities, radioactive waste is generated that must be disposed of. Besides waste from the nuclear facilities, some waste derives from other activities such as industry, research, medical care, etc. Short-lived low- and intermediate-level waste from these activities is disposed of in the final repository for short-lived radioactive waste, SFR, in Forsmark. The facility, which has been in operation since 1988, is owned and operated by Svensk Karnbranslehantering AB, SKB. The existing facility has neither sufficient space nor a license to receive decommissioning waste. SFR must therefore be extended so that shortlived low- and intermediate-level decommissioning waste from the nuclear facilities can also be received. The need for additional capacity has been accentuated by the closure of two reactors in Barseback. These reactors cannot be dismantled until the SFR facility has been extended. The existing repository is built to receive, and after closure serve as a passive repository for, low- and intermediate-level radioactive waste. The disposal rooms are situated in the bedrock beneath the sea floor, covered by about 60 metres of rock. The repository has been designed so that it can be abandoned after closure without requiring further measures to maintain its function. The extension of SFR, is done at the -120 m level immediately adjacent to, and within the same depth range as, the existing facility. The basic function of the existing SFR and of the extended one will be the same. However, a clear difference is the design of the tunnel and the rock vault that are required to permit transport and storage of whole reactor pressure vessels. The application for a license to build this extension includes an assessment of the long-term safety (post-closure safety) of the facility. The safety assessment also contains an updated assessment of the long-term safety of the existing facility. The safety assessment for

  5. Bedrock Hydrogeology - Groundwater flow modelling. Site investigation SFR

    Energy Technology Data Exchange (ETDEWEB)

    Oehman, Johan [Geosigma AB, Uppsala (Sweden); Follin, Sven [SF GeoLogic AB, Taeby (Sweden); Oden, Magnus [SKB, Stockholm (Sweden)

    2013-05-15

    The hydrogeological model developed for the SFR extension project (PSU) consists of 40 geologically modelled deformation zones (DZ) and 8 sub-horizontal structural-hydraulic features, called SBAstructures, not defined in the geological model. However, some of the SBA-structures coincide with what is defined as unresolved possible deformation zones (Unresolved PDZ) in the geological modelling. In addition, the hydrogeological model consists of a stochastic discrete fracture network (DFN) model intended for the less fractured rock mass volumes (fracture domains) between the zones and the SBA-structures, and a stochastic fracture model intended to handle remaining Unresolved PDZs in the geological modelling not modelled as SBA-structures in the hydrogeological modelling. The four structural components of the bedrock in the hydrogeological model, i.e. DZ, SBA, Unresolved PDZ and DFN, are assigned hydraulic properties in the hydrogeological model based on the transmissivities interpreted from single-hole hydraulic tests. The main objective of the present work is to present the characteristics of the hydrogeological model with regard to the needs of the forthcoming safety assessment SR-PSU. In concrete words, simulated data are compared with measured data, i.e. hydraulic heads in boreholes and tunnel inflow to the existing repository (SFR). The calculations suggest that the available data for flow model calibration cannot be used to motivate a substantial adjustment of the initial hydraulic parameterisation (assignment of hydraulic properties) of the hydrogeological model. It is suggested that uncertainties in the hydrogeological model are studied in the safety assessment SR-PSU by means of a large number of calculation cases. These should address hydraulic heterogeneity of deterministic structures (DZ and SBA) and realisations of stochastic fractures/fracture networks (Unresolved PDZ and DFN) within the entire SFR Regional model domain.

  6. Bedrock Hydrogeology-Groundwater flow modelling. Site investigation SFR

    International Nuclear Information System (INIS)

    Oehman, Johan; Follin, Sven; Oden, Magnus

    2013-05-01

    The hydrogeological model developed for the SFR extension project (PSU) consists of 40 geologically modelled deformation zones (DZ) and 8 sub-horizontal structural-hydraulic features, called SBAstructures, not defined in the geological model. However, some of the SBA-structures coincide with what is defined as unresolved possible deformation zones (Unresolved PDZ) in the geological modelling. In addition, the hydrogeological model consists of a stochastic discrete fracture network (DFN) model intended for the less fractured rock mass volumes (fracture domains) between the zones and the SBA-structures, and a stochastic fracture model intended to handle remaining Unresolved PDZs in the geological modelling not modelled as SBA-structures in the hydrogeological modelling. The four structural components of the bedrock in the hydrogeological model, i.e. DZ, SBA, Unresolved PDZ and DFN, are assigned hydraulic properties in the hydrogeological model based on the transmissivities interpreted from single-hole hydraulic tests. The main objective of the present work is to present the characteristics of the hydrogeological model with regard to the needs of the forthcoming safety assessment SR-PSU. In concrete words, simulated data are compared with measured data, i.e. hydraulic heads in boreholes and tunnel inflow to the existing repository (SFR). The calculations suggest that the available data for flow model calibration cannot be used to motivate a substantial adjustment of the initial hydraulic parameterisation (assignment of hydraulic properties) of the hydrogeological model. It is suggested that uncertainties in the hydrogeological model are studied in the safety assessment SR-PSU by means of a large number of calculation cases. These should address hydraulic heterogeneity of deterministic structures (DZ and SBA) and realisations of stochastic fractures/fracture networks (Unresolved PDZ and DFN) within the entire SFR Regional model domain

  7. English translation of three documents relating to the SFR-1

    International Nuclear Information System (INIS)

    1988-01-01

    After approval from the National Institute of Radiation Protection, (the SSI) on April 26th, 1988 the Swedish Nuclear Fuel and Waste Management Company, the SKB, put the Final Repository for Radioactive Waste, the SFR-1 (Forsmark), into operation. This report contains English translations of the Operating Permission issued by SSI and the associated radiation protection instructions. Also included is a translation of chapter 4, the viewpoints and evaluations, of the Assessment Memorandum which was the background material for the Board of the SSI when deciding on the operational permission. (orig./HP)

  8. Experimental Facilities and Plan for a Prototype SFR

    International Nuclear Information System (INIS)

    Chang, Jinwook

    2013-01-01

    Contents of the presentation: 1. STELLA; 2. Under Sodium Viewing; 3. Sodium-CO 2 Interaction Test; Overview of the Sodium Integral Effect Test Loop for Safety Simulation and Assessment (STELLA) Program, the scope of the experiment and the overall characteristics of STELLA-1: Phase 1: STELLA-1, Individual component test; • Performance evaluation of key sodium components; • Heat exchanger design codes V&V. Phase 2: STELLA-2, Integral effect test; • Verification of dynamic plant response after reactor shutdown; • Construction of test DB to support specific design approval for the prototype SFR

  9. A New Streamflow-Routing (SFR1) Package to Simulate Stream-Aquifer Interaction with MODFLOW-2000

    Science.gov (United States)

    Prudic, David E.; Konikow, Leonard F.; Banta, Edward R.

    2004-01-01

    The increasing concern for water and its quality require improved methods to evaluate the interaction between streams and aquifers and the strong influence that streams can have on the flow and transport of contaminants through many aquifers. For this reason, a new Streamflow-Routing (SFR1) Package was written for use with the U.S. Geological Survey's MODFLOW-2000 ground-water flow model. The SFR1 Package is linked to the Lake (LAK3) Package, and both have been integrated with the Ground-Water Transport (GWT) Process of MODFLOW-2000 (MODFLOW-GWT). SFR1 replaces the previous Stream (STR1) Package, with the most important difference being that stream depth is computed at the midpoint of each reach instead of at the beginning of each reach, as was done in the original Stream Package. This approach allows for the addition and subtraction of water from runoff, precipitation, and evapotranspiration within each reach. Because the SFR1 Package computes stream depth differently than that for the original package, a different name was used to distinguish it from the original Stream (STR1) Package. The SFR1 Package has five options for simulating stream depth and four options for computing diversions from a stream. The options for computing stream depth are: a specified value; Manning's equation (using a wide rectangular channel or an eight-point cross section); a power equation; or a table of values that relate flow to depth and width. Each stream segment can have a different option. Outflow from lakes can be computed using the same options. Because the wetted perimeter is computed for the eight-point cross section and width is computed for the power equation and table of values, the streambed conductance term no longer needs to be calculated externally whenever the area of streambed changes as a function of flow. The concentration of solute is computed in a stream network when MODFLOW-GWT is used in conjunction with the SFR1 Package. The concentration of a solute in a

  10. The Swedish final repository for reactor waste (SFR). A summary of the SFR project with special emphasis on the near-field assessments

    International Nuclear Information System (INIS)

    Carlsson, J.

    1988-01-01

    The first phase of the final repository for reactor waste (SFR) is scheduled for operation in April 1988. The construction work is finished and preoperational tests are in progress. Impact on the environment from SFR is analysed in a final safety report. This paper gives a summary of the design and performance of SFR. Assessments, made for the analysises of the long term safety, are given with special emphasis on the near-field. As a conclusion from the analysises, the dose commitment to the most affected individual during the post-closure period, has proved to constitute only an insignificant contribution to the natural radioactive environment of the area

  11. Improvement of composition of core sand and molding sand mixtures for power machine building castings

    International Nuclear Information System (INIS)

    Velikanov, G.F.; Primak, I.N.; Brechko, A.A.

    1982-01-01

    Considered is a problem of development and improvement of mixtures, as well as of antisticking coatings with the given parameters providing production of castings of the necessary quality. Requirements to properties of mixtures and antisticking coatings are formulated proceeding from the conditions of guaranteed production of qualitative steel castings with mass from 0.5 up to 20t and wall thickness from 60 up to 200 mm. Formation of film structure of binding compositions is studied, their marginal contact angle and surface tension are determined. In the result of work carried out on improvement of core sand and molding sand mixtures the labour productivity during the production of core and moldings has been increased in 20-25% in average, the quality has also been improved [ru

  12. Prediction of the Sodium Void Reactivity in the Metal-fueled SFR Using the ENDF/B-VII.0 Library

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Lim, Jae-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The SVR (Sodium Void Reactivity) is one of the most important parameters in SFR (Sodium-cooled Fast Reactor) safety analysis. In this paper, to estimate the error of the SVR in metal-fueled SFR, three physics experiments named as BFS-75-1, BFS-109-2A, and BFS-84-1 were examined using recent cross-section library, ENDF/B-VII.0 and the MCNP code. In the MCNP6 calculation, two million histories/generation with 50 inactive/300 active generations are used with the continuous-energy ENDF/B-VII.0 library. We expect that accuracy of total cross-section of the sodium may play a dominant role in errors of SVRs at core peripheral and sodium plenum regions, whereas accuracy of capture cross-section of the sodium may play a dominant role for the results in errors of SVRs at core central region. In addition, capture cross-sections of the sodium in the ENDF/B-VII.0, the JEFF-3.2, and the JENDL-4.0 libraries show significant differences between each other, while total cross-sections of sodium in three libraries show good agreement.

  13. Improvement of humidity resistance of water soluble core by precipitation method

    Directory of Open Access Journals (Sweden)

    Zhang Long

    2011-05-01

    Full Text Available Water soluble core has been widely used in manufacturing complex metal components with hollow configurations or internal channels; however, the soluble core can absorb water easily from the air at room temperature. To improve the humidity resistance of the water soluble core and optimize the process parameters applied in manufacturing of the water soluble core, a precipitation method and a two-level-three-full factorial central composite design were used, respectively. The properties of the cores treated by the precipitation method were compared with that without any treatment. Through a systematical study by means of both an environmental scanning electron microscope (ESEM and an energy dispersive X-ray (EDX analyzer, the results indicate that the hygroscopicity can be reduced by 20% and the obtained optimal process conditions for three critical control factors affecting the hygroscopicity are 0.2 g·mL-1 calcium chloride concentration, 4% water concentration and 0 min ignition time. The porous surface coated by calcium chloride and the high humidity resistance products generated in the precipitation reaction between calcium chloride and potassium carbonate may contribute to the lower hygroscopicity.

  14. Low Cost, Lightweight Gravity Coring and Improved Epoxy Impregnation Applied to Laminated Maar Sediment in Vietnam

    Directory of Open Access Journals (Sweden)

    Jan P. Schimmelmann

    2018-05-01

    Full Text Available In response to the need for lightweight and affordable sediment coring and high-resolution structural documentation of unconsolidated sediment, we developed economical and fast methods for (i recovering short sediment cores with undisturbed topmost sediment, without the need for a firmly anchored coring platform, and (ii rapid epoxy-impregnation of crayon-shaped subcores in preparation for thin-sectioning, with minimal use of solvents and epoxy resin. The ‘Autonomous Gravity Corer’ (AGC can be carried to remote locations and deployed from an inflatable or makeshift raft. Its utility was tested on modern unconsolidated lacustrine sediment from a ~21 m deep maar lake in Vietnam’s Central Highlands near Pleiku. The sedimentary fabric fidelity of the epoxy-impregnation method was demonstrated for finely laminated artificial flume sediment. Our affordable AGC is attractive not only for work in developing countries, but lends itself broadly for coring in remote regions where challenging logistics prevent the use of heavy coring equipment. The improved epoxy-impregnation technique saves effort and costly chemical reagents, while at the same time preserving the texture of the sediment.

  15. Fracture resistance improvement of polypropylene by joint action of core-shell particles and nucleating agent

    International Nuclear Information System (INIS)

    Yang Gang; Han Liang; Ding Haifeng; Wu Haiyan; Huang Ting; Li Xiaoxi; Wang Yong

    2011-01-01

    Research highlights: →The core-shell particles, which were prepared from melt blending of POE and nano-CaCO 3 , and different nucleating agents (α-form NA or β-form NA) were first introduced into PP to prepare the super toughened PP materials. →NAs control the crystalline structures of PP matrix including the spherulites diameter and the crystal form. →NAs and core-shell particles exhibit apparent joint effect in improving the fracture resistance of PP. - Abstract: As a serial work about the fracture resistance improvement of polypropylene (PP), this work reports the joint effect of core-shell particles and nucleating agent (NA) on the microstructure and fracture resistance of PP. Core-shell particles were prepared through melt blending of ethylene-octene copolymer (POE) and calcium carbonate (CaCO 3 ). Different NA, i.e. α-form NA (P-tert-butylbenzoic acid-Al, MD-NA-28) and β-form NA (aryl amides compound, TMB-5) were introduced into PP matrix to control the crystalline structure. The phase morphology of POE and the distribution of CaCO 3 were characterized by using scanning electron microscope (SEM), and the crystallization behavior of PP matrix were investigated by using differential scanning calorimetry (DSC), wide angle X-ray diffraction (WAXD) and polarization optical microscope (POM). The mechanical properties were obtained through universal tensile measurement and notched Izod impact measurement. Surprisingly, the results show that through addition of so-called core-shell particles and NA simultaneously, the fracture resistance of PP can be dramatically improved.

  16. Recent improvements in on-line core supervision at Loviisa NPP

    International Nuclear Information System (INIS)

    Antila, M.; Kuusisto, J.

    1999-01-01

    On-line core supervision system (RESU) based on monitoring of local fuel limits has been in use at the Loviisa VVER-440 reactors for more than twenty years. Minor modifications were made ten years ago when the computer hardware was upgraded. In April 1998 Loviisa got the licence for 1500 MW power. Power uprating and introduction of new fuel types gave rise to the latest improvements in the core supervision system, which is called RESU-98. In August 1999 the Finnish Safety Authority (STUK) has given approval for RESU-98, which is now in use at the Loviisa NPP. RESU-98 includes essentially the same computer codes, which are used in reload planning. The extensive in-core instrumentation is utilised to adjust the theoretical 3D-power distribution to get a best-estimate results. In this paper a general review of the RESU-98 system is given including instrumentation, methods, core monitoring, predictive functions and validation. Special attention is paid on the recent improvements. (Authors)

  17. Recent improvements in on-line core supervision at Loviisa NPP

    International Nuclear Information System (INIS)

    Antila, M.; Kuusisto, J.

    2000-01-01

    AN on-line core supervision system (RESU) based on monitoring of local fuel limits has been in use at the Loviisa VVER-440 reactors for more than twenty years. Minor modifications were made ten years ago to upgrade the computer hardware. In April 1998 Loviisa obtained a licence for 1500 MW th power. Power up-rating and introduction of new fuel types gave rise to the latest improvements in the core supervision system, which is called RESU-98. In August 1999 the Finnish Safety Authority (STUCK) officially approved RESU-98, which is now in use at the Loviisa NPP. RESU-98 includes essentially the same computer codes, which are used in reload planning. The extensive in-core instrumentation is utilised to adjust the theoretical 3-D power distribution to get a best-estimate result. In this paper a general review of the RESU-98 system is given including instrumentation, methods, core monitoring, predictive functions and validation. Special attention is paid to recent improvements. (author)

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

    International Nuclear Information System (INIS)

    Greenspan, E

    2006-01-01

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

  19. Evaluating core technology capacity based on an improved catastrophe progression method: the case of automotive industry

    Science.gov (United States)

    Zhao, Shijia; Liu, Zongwei; Wang, Yue; Zhao, Fuquan

    2017-01-01

    Subjectivity usually causes large fluctuations in evaluation results. Many scholars attempt to establish new mathematical methods to make evaluation results consistent with actual objective situations. An improved catastrophe progression method (ICPM) is constructed to overcome the defects of the original method. The improved method combines the merits of the principal component analysis' information coherence and the catastrophe progression method's none index weight and has the advantage of highly objective comprehensive evaluation. Through the systematic analysis of the influencing factors of the automotive industry's core technology capacity, the comprehensive evaluation model is established according to the different roles that different indices play in evaluating the overall goal with a hierarchical structure. Moreover, ICPM is developed for evaluating the automotive industry's core technology capacity for the typical seven countries in the world, which demonstrates the effectiveness of the method.

  20. DNA homologous recombination factor SFR1 physically and functionally interacts with estrogen receptor alpha.

    Directory of Open Access Journals (Sweden)

    Yuxin Feng

    Full Text Available Estrogen receptor alpha (ERα, a ligand-dependent transcription factor, mediates the expression of its target genes by interacting with corepressors and coactivators. Since the first cloning of SRC1, more than 280 nuclear receptor cofactors have been identified, which orchestrate target gene transcription. Aberrant activity of ER or its accessory proteins results in a number of diseases including breast cancer. Here we identified SFR1, a protein involved in DNA homologous recombination, as a novel binding partner of ERα. Initially isolated in a yeast two-hybrid screen, the interaction of SFR1 and ERα was confirmed in vivo by immunoprecipitation and mammalian one-hybrid assays. SFR1 co-localized with ERα in the nucleus, potentiated ER's ligand-dependent and ligand-independent transcriptional activity, and occupied the ER binding sites of its target gene promoters. Knockdown of SFR1 diminished ER's transcriptional activity. Manipulating SFR1 expression by knockdown and overexpression revealed a role for SFR1 in ER-dependent and -independent cancer cell proliferation. SFR1 differs from SRC1 by the lack of an intrinsic activation function. Taken together, we propose that SFR1 is a novel transcriptional modulator for ERα and a potential target in breast cancer therapy.

  1. New sol–gel refractory coatings on chemically-bonded sand cores for foundry applications to improve casting surface quality

    DEFF Research Database (Denmark)

    Nwaogu, Ugochukwu Chibuzoh; Poulsen, T.; Stage, R.K.

    2011-01-01

    Foundry refractory coatings protect bonded sand cores and moulds from producing defective castings during the casting process by providing a barrier between the core and the liquid metal. In this study, new sol–gel refractory coating on phenolic urethane cold box (PUCB) core was examined. The coa......Foundry refractory coatings protect bonded sand cores and moulds from producing defective castings during the casting process by providing a barrier between the core and the liquid metal. In this study, new sol–gel refractory coating on phenolic urethane cold box (PUCB) core was examined......–gel coated cores have better surface quality than those from uncoated cores and comparable surface quality with the commercial coatings. Therefore, the new sol–gel coating has a potential application in the foundry industry for improving the surface finish of castings thereby reducing the cost of fettling...

  2. TRIUMF - The Swedish data base system for radioactive waste in SFR

    International Nuclear Information System (INIS)

    Skogsberg, Marie; Andersson, Per-Anders

    2006-01-01

    All short lived LLW/ILW from the operation and maintenance of all Swedish Nuclear Power Plants are disposed in SFR, the Swedish final repository for radioactive operational waste. It is important to save all the information about radioactive waste that is needed now and in the future. To be secure that, we have developed a database system in Sweden called Triumf, consisting information about all the waste packages that are disposed in SFR. The waste producers register data concerning individual waste package during production. Before transport to SFR a data file with all information about the individual waste packages is transferred to Triumf. When transferred, the data are checked against accepted limitations before the waste can be loaded on the ship for transport to SFR. After disposal at SFR the deposition location in the repository is added to the database for each waste package. (author)

  3. Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

    Energy Technology Data Exchange (ETDEWEB)

    Yacout, A. M. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2016-09-16

    The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of data were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.

  4. Improved core-edge tokamak transport simulations with the CORSICA 2 code

    International Nuclear Information System (INIS)

    Tarditi, A.; Cohen, R.H.; Crotinger, J.A.

    1996-01-01

    The CORSICA 2 code models the nonlinear transport between the core and the edge of a tokamak plasma. The code couples a 2D axisymmetric edge/SOL model (UEDGE) to a 1D model for the radial core transport in toroidal flux coordinates (the transport module from the CORSICA 1 code). The core density and temperature profiles are joined to the flux-surface average profiles from the 2D code sufficiently inside the magnetic separatrix, at a flux surface on which the edge profiles are approximately constant. In the present version of the code, the deuterium density and electron and ion temperatures are coupled. The electron density is determined by imposing quasi-neutrality, both in the core and in the edge. The model allows the core-edge coupling of multiple ion densities while retaining a single temperature (corresponding to the equilibration value) for the all ion species. Applications of CORSICA 2 to modeling the DIII-D tokamak are discussed. This work will focus on the simulation of the L-H transition, coupling a single ion species (deuterium) and the two (electron and ion) temperatures. These simulations will employ a new self-consistent model for the L-H transition that is being implemented in the UEDGE code. Applications to the modeling of ITER ignition scenarios are also discussed. This will involve coupling a second density species (the thermal alphas), bringing the total number of coupled variables up to four. Finally, the progress in evolving the magnetic geometry is discussed. Currently, this geometry is calculated by CORSICA's MHD equilibrium module (TEQ) at the beginning of the run and fixed thereafter. However, CORSICA 1 can evolve this geometry quasistatically, and this quasistatic treatment is being extended to include the edge/SOL geometry. Recent improvements for code speed-up are also presented

  5. Boron-bearing Influences of 9Cr-0.5Mo-2W-V-Nb Ferritic/Martensitic Steels for a SFR Fuel Cladding

    International Nuclear Information System (INIS)

    Baek, Jong-Hyuk; Han, Chang-Hee; Kim, Woo-Gon; Kim, Sung-Ho; Lee, Chan-Bock

    2008-01-01

    Currently the principal materials in a SFR (sodium-cooled fast reactor) of Gen-IV nuclear system are considering stainless steels (e.g. austenitic steels and ferritic/martensitic steels) for pressure boundary and structural applications in the primary circuit (cladding, duct, cold and hot leg piping, and pressure vessel). There are sound technical justifications for these material selections, and the adoption of these stainless steels for a wide range of nuclear and non-nuclear applications has generated much industrial technology and experience. However, there are strong incentives to develop advanced materials, especially cladding, for the Gen-IV SFR. The Gen-IV SFR is to have a considerable increase in safety and be economically competitive when compared with the conventional water reactors. To accomplish these objectives, the development of the fuel cladding material should be set forth as a premise because its integrity is directly related to those of the reactor system as well as the fuel in the Gen-IV SFR. Since last year, a R and D program was launched to develop the improved ferritic/martensitic steel for the Gen-IV SFR fuel cladding. Categories of materials considered in the program included 8 - 12% Cr ferritic/ martensitic steels. A strong recommendation was made for the development of a high strength steel equivalent to or superior to ASTM Gr.92 steel to offset the difficulties encountered with commercial available steels of the 8 - 12% Cr group. That is, since fuel cladding in the Gen-IV SFR would operate under higher temperatures than 600 .deg. C, contacting with liquid sodium, and be irradiated by neutrons to as high as 200dpa, the cladding should thus sustain both superior irradiation and temperature stabilities during an operational life. The newly developed advanced steel should overcome the severe drawback; mechanical properties, especially creep, are deteriorated at a higher temperature over 600 .deg. C. In this study, as one of the composition

  6. DANCE, BALANCE AND CORE MUSCLE PERFORMANCE MEASURES ARE IMPROVED FOLLOWING A 9-WEEK CORE STABILIZATION TRAINING PROGRAM AMONG COMPETITIVE COLLEGIATE Dancers.

    Science.gov (United States)

    Watson, Todd; Graning, Jessica; McPherson, Sue; Carter, Elizabeth; Edwards, Joshuah; Melcher, Isaac; Burgess, Taylor

    2017-02-01

    Dance performance requires not only lower extremity muscle strength and endurance, but also sufficient core stabilization during dynamic dance movements. While previous studies have identified a link between core muscle performance and lower extremity injury risk, what has not been determined is if an extended core stabilization training program will improve specific measures of dance performance. This study examined the impact of a nine-week core stabilization program on indices of dance performance, balance measures, and core muscle performance in competitive collegiate dancers. Within-subject repeated measures design. A convenience sample of 24 female collegiate dance team members (age = 19.7 ± 1.1 years, height = 164.3 ± 5.3 cm, weight 60.3 ± 6.2 kg, BMI = 22.5 ± 3.0) participated. The intervention consisted of a supervised and non-supervised core (trunk musculature) exercise training program designed specifically for dance team participants performed three days/week for nine weeks in addition to routine dance practice. Prior to the program implementation and following initial testing, transversus abdominis (TrA) activation training was completed using the abdominal draw-in maneuver (ADIM) including ultrasound imaging (USI) verification and instructor feedback. Paired t tests were conducted regarding the nine-week core stabilization program on dance performance and balance measures (pirouettes, single leg balance in passe' releve position, and star excursion balance test [SEBT]) and on tests of muscle performance. A repeated measures (RM) ANOVA examined four TrA instruction conditions of activation: resting baseline, self-selected activation, immediately following ADIM training and four days after completion of the core stabilization training program. Alpha was set at 0.05 for all analysis. Statistically significant improvements were seen on single leg balance in passe' releve and bilateral anterior reach for the SEBT (both p ≤ 0

  7. On-going activities in the European JASMIN project for the development and validation of ASTEC-Na SFR safety simulation code - 15072

    International Nuclear Information System (INIS)

    Girault, N.; Cloarec, L.; Herranz, L.; Bandini, G.; Perez-Martin, S.; Ammirabile, L.

    2015-01-01

    The 4-year JASMIN collaborative project (Joint Advanced Severe accidents Modelling and Integration for Na-cooled fast reactors), started in Dec.2011 in the frame of the 7. Framework Programme of the European Commission. It aims at developing a new European simulation code, ASTEC-Na, dealing with the primary phase of SFR core disruptive accidents. The development of a new code, based on a robust advanced simulation tool and able to encompass the in-vessel and in-containment phenomena occurring during a severe accident is indeed of utmost interest for advanced and innovative future SFRs for which an enhanced safety level will be required. This code, based on the ASTEC European code system developed by IRSN and GRS for severe accidents in water-cooled reactors, is progressively integrating and capitalizing the state-of-the-art knowledge of SFR accidents through physical model improvement or development of new ones. New models are assessed on in-pile (CABRI, SCARABEE etc...) and out-of pile experiments conducted during the 70's-80's and code-o-code benchmarking with current accident simulation tools for SFRs is also conducted. During the 2 and a half first years of the project, model specifications and developments were conducted and the validation test matrix was built. The first version of ASTEC-Na available in early 2014 already includes a thermal-hydraulics module able to simulate single and two-phase sodium flow conditions, a zero point neutronic model with simple definition of channel and axial dependences of reactivity feedbacks and models derived from SCANAIR IRSN code for simulating fuel pin thermo-mechanical behaviour and fission gas release/retention. Meanwhile, models have been developed in the source term area for in-containment particle generation and particle chemical transformation, but their implementation is still to be done. As a first validation step, the ASTEC-Na calculations were satisfactorily compared to thermal-hydraulics experimental

  8. Sodium-cooled Fast Reactor Cores using Uranium-Free Metallic Fuels for Maximizing TRU Support Ratio

    International Nuclear Information System (INIS)

    You, WuSeung; Hong, Ser Gi

    2014-01-01

    The depleted uranium plays important roles in the SFR burner cores because it substantially contributes to the inherent safety of the core through the negative Doppler coefficient and large delayed neutron. However, the use of depleted uranium as a diluent nuclide leads to a limited value of TRU support ratio due to the generation of TRUs through the breeding. In this paper, we designed sodium cooled fast reactor (SFR) cores having uranium-free fuels 3,4 for maximization of TRU consumption rate. However, the uranium-free fuelled burner cores can be penalized by unacceptably small values of the Doppler coefficient and small delayed neutron fraction. In this work, metallic fuels of TRU-(W or Ni)-Zr are considered to improve the performances of the uranium-free cores. The objective of this work is to consistently compare the neutronic performances of uranium-free sodium cooled fast reactor cores having TRU-Zr metallic fuels added with Ni or W and also to clarify what are the problematic features to be resolved. In this paper, a consistent comparative study of 400MWe sodium cooled burner cores having uranium-based fuels and uranium-free fuels was done to analyze the relative core neutronic features. Also, we proposed a uranium-free metallic fuel based on Nickel. From the results, it is found that tungsten-based uranium-free metallic fuel gives large negative Doppler coefficient due to high resonance of tungsten isotopes but this core has large sodium void worth and small effective delayed neutron fraction while the nickel-based uranium-free metallic fuelled core has less negative Doppler coefficient but smaller sodium void worth and larger effective delayed neutron fraction than the tungsten-based one. On the other hand, the core having TRU-Zr has very high burnup reactivity swing which may be problematic in compensating it using control rods and the least negative Doppler coefficient

  9. Improving core medical training--innovative and feasible ideas to better training.

    Science.gov (United States)

    Tasker, Fiona; Dacombe, Peter; Goddard, Andrew F; Burr, Bill

    2014-12-01

    A recent survey of UK core medical training (CMT) training conducted jointly by the Royal College of Physicians (RCP) and Joint Royal College of Physicians Training Board (JRCPTB) identified that trainees perceived major problems with their training. Service work dominated and compromised training opportunities, and of great concern, almost half the respondents felt that they had not been adequately prepared to take on the role of medical registrar. Importantly, the survey not only gathered CMT trainees' views of their current training, it also asked them for their 'innovative and feasible ways to improve CMT'. This article draws together some of these excellent ideas on how the quality of training and the experience of trainees could be improved. It presents a vision for how CMT trainees, consultant supervisors, training programme directors, clinical directors and managers can work together to implement relevant, feasible and affordable ways to improve training for doctors and deliver the best possible care for patients. © 2014 Royal College of Physicians.

  10. Homogeneous Minor Actinide Transmutation in SFR: Neutronic Uncertainties Propagation with Depletion

    International Nuclear Information System (INIS)

    Buiron, L.; Plisson-Rieunier, D.

    2015-01-01

    In the frame of next generation fast reactor design, the minimisation of nuclear waste production is one of the key objectives for current R and D. Among the possibilities studied at CEA, minor actinides multi-recycling is the most promising industrial way achievable in the near-term. Two main management options are considered: - Multi-recycling in a homogeneous way (minor actinides diluted in the driver fuel). If this solution can help achieving high transmutation rates, the negative impact of minor actinides on safety coefficients allows only a small fraction of the total heavy mass to be loaded in the core (∼ few %). - Multi-recycling in heterogeneous way by means of Minor Actinide Bearing Blanket (MABB) located at the core periphery. This solution offers more flexibility than the previous one, allowing a total minor actinides decoupled management from the core fuel. As the impact on feedback coefficient is small larger initial minor actinide mass can be loaded in this configuration. Starting from a breakeven Sodium Fast Reactor designed jointly by CEA, Areva and EdF teams, the so called SFR V2B, transmutation performances have been studied in frame on the French fleet for both options and various specific isotopic management (all minor actinides, americium only, etc.). Using these results, a sensitivity study has been performed to assess neutronic uncertainties (i.e coming from cross section) on mass balance on the most attractive configurations. This work in based on a new implementation of sensitivity on concentration with depletion in the ERANOS code package. Uncertainties on isotopes masses at the end of irradiation using various variance-covariance is discussed. (authors)

  11. Six weeks of core stability training improves landing kinetics among female capoeira athletes: a pilot study.

    Science.gov (United States)

    Araujo, Simone; Cohen, Daniel; Hayes, Lawrence

    2015-03-29

    Core stability training (CST) has increased in popularity among athletes and the general fitness population despite limited evidence CST programmes alone lead to improved athletic performance. In female athletes, neuromuscular training combining balance training and trunk and hip/pelvis dominant CST is suggested to reduce injury risk, and specifically peak vertical ground reaction forces (vGRF) in a drop jump landing task. However, the isolated effect of trunk dominant core stability training on vGRF during landing in female athletes had not been evaluated. Therefore, the objective of this study was to evaluate landing kinetics during a drop jump test following a CST intervention in female capoeira athletes. After giving their informed written consent, sixteen female capoeira athletes (mean ± SD age, stature, and body mass of 27.3 ± 3.7 years, 165.0 ± 4.0 cm, and 59.7 ± 6.3 kg, respectively) volunteered to participate in the training program which consisted of static and dynamic CST sessions, three times per week for six weeks. The repeated measures T-test revealed participants significantly reduced relative vGRF from pre- to post-intervention for the first (3.40 ± 0.78 vs. 2.85 ± 0.52 N·NBW-1, respectively [pcore stability training improves landing kinetics without improving jump height, and may reduce lower extremity injury risk in female athletes.

  12. Improved methodology for generation of axial flux shapes in digital core protection systems

    International Nuclear Information System (INIS)

    Lee, G.-C.; Baek, W.-P.; Chang, S.H.

    2002-01-01

    An improved method of axial flux shape (AFS) generation for digital core protection systems of pressurized water reactors is presented in this paper using an artificial neural network (ANN) technique - a feedforward network trained by backpropagation. It generates 20-node axial power shapes based on the information from three ex-core detectors. In developing the method, a total of 7173 axial flux shapes are generated from ROCS code simulation for training and testing of the ANN. The ANN trained 200 data predicts the remaining data with the average root mean square error of about 3%. The developed method is also tested with the real plant data measured during normal operation of Yonggwang Unit 4. The RMS errors in the range of 0.9∼2.1% are about twice as accurate as the cubic spline approximation method currently used in the plant. The developed method would contribute to solve the drawback of the current method as it shows reasonable accuracy over wide range of core conditions

  13. Disallowing Same-program Co-schedules to Improve Efficiency in Quad-core Servers

    OpenAIRE

    de Blanche, Andreas; Lundqvist, Thomas

    2017-01-01

    Programs running on different cores in a multicore server are often forced to share resources like off-chip memory, caches, I/O devices, etc. This resource sharing often leads to degraded performance, a slowdown, for the programs that share the resources. A job scheduler can improve performance by co-scheduling programs that use different resources on the same server. The most common approach to solve this co-scheduling problem has been to make job-schedulers resource aware, finding ways to c...

  14. Simulation of nonlinear dynamics of a PWR core by an improved lumped formulation for fuel heat transfer

    International Nuclear Information System (INIS)

    Su, Jian; Cotta, Renato M.

    2000-01-01

    In this work, thermohydraulic behaviour of PWR, during reactivity insertion and partial loss-of-flow, is simulated by using a simplified mathematical model of reactor core and primary coolant. An improved lumped parameter formulation for transient heat conduction in fuel rod is used for core heat transfer modelling. Transient temperature response of fuel, cladding and coolant is analysed. (author)

  15. Basic Design of Experimental Facility for Measuring Pressure Drop of IHX in a SFR

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Yung-Joo; Eoh, Jae-Hyuk; Kim, Hyungmo; Lee, Dong-Won; Jeong, Ji-Young; Lee, Hyeong-Yeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Kyungpook National Univ., Daegu (Korea, Republic of)

    2015-05-15

    The conceptual design of the Prototype gen-IV SFR (PGSFR) with a 150 MWe capacity was commenced in 2012 through the national long-term R and D program by KAERI. Then, PGSFR is now being designed with the defense in depth concept with active, passive and inherent safety features to acquire design approval for PGSFR from the Korean regulatory authority by 2020. PGSFR is a sodium-cooled pool-type fast reactor with all primary components including the primary heat transport system (PHTS) pumps and IHXs are located inside a sodium pool. The heat produced due to fission in the core is transported by primary sodium to secondary sodium in a sodium to sodium intermediate heat exchanger (IHX), which in turn is transferred to water in a steam generator (SG). Basic design of the IHX flow characteristic test facility, WEIPA was conducted based on the three-level scaling methodology in order to preserve the flow characteristics of the IHX in PGSFR. This test facility is intended to measure a high precision pressure drop at the shell-side of the IHX. This paper describes the aspects of the current design features of the IHX in PGSFR, scaling and basic design features of the facility.

  16. Analysis of power ramp rate and minimum power controllability of the MMS model for a plant dynamics analysis of a Prototype SFR

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Kim, Dehee; Joo, Hyungkook; Lee, Taeho

    2014-01-01

    A full plant dynamic model was developed for a prototype SFR using the Modular Modeling System (MMS). It includes the modeling of various subsystems such as the neutronics, primary and intermediate sodium systems of the NSSS, steam and water systems of the BOP, BOP controls, and the supervisory plant controls. The NSSS model is subdivided into component models, such as a Core, IHXs, Pumps, SGs, and the rest of the NSSS loop model. The BOP model is subdivided into a steam subsystem, feedwater subsystem, and preheater subsystem. Plant transient tests were performed to study the operational considerations. It includes varying the power ramp rate and studying the controllability at minimum power. Plant transient tests were performed to study operational considerations by using the MMS model for a prototype SFR. It includes varying the power ramp rate, studying the controllability at the minimum power set point. At a power ramp rate of higher than 2%, the steam temperature has a large deviation from the target. As the power set point decreases, the PHTS hot leg temperature and steam temperature tend to have higher deviations. After further refinement of the MMS model, it can be useful for developing the plant operation logics of the prototype SFR

  17. Integrating Morbidity and Mortality Core Competencies and Quality Improvement in Otolaryngology.

    Science.gov (United States)

    Laury, Adrienne M; Bowe, Sarah N; Lospinoso, Joshua

    2017-02-01

    To date, an otolaryngology-specific morbidity and mortality (M&M) conference has never been reported or evaluated. To propose a novel otolaryngology-specific M&M format and to assess its success using a validated assessment tool. Preintervention and postintervention cohort study spanning 14 months (September 2014 to November 2015), with 32 faculty, residents, and medical students attending the department of otolaryngology M&M conference, conducted at the the San Antonio Uniformed Services Health Education Consortium. A novel quality assurance conference was implemented in the department of otolaryngology at the San Antonio Uniformed Services Health Education Consortium. This conference incorporates patient safety reports, otolaryngology-specific quality metrics, and individual case presentations. The revised format integrates the Accreditation Council for Graduate Medical Education (ACGME) core competencies and Quality Improvement and Patient Safety (QI/PS) system. This format was evaluated by faculty, residents, and medical students every other month for 14 months to assess changes in attitudes regarding the M&M conference as well as changes in presentation quality. Overall, 13 faculty, 12 residents, and 7 medical students completed 232 evaluations. Summary statistics of both resident and faculty attitudes about the success of the M&M format seem to improve over the 14 months between the prequestionnaires and postquestionnaires. General attitudes for both residents and faculty significantly improved from the pretest to posttest (odds ratio, 0.32 per month; 95% CI, 0.29-0.35). In the pretest period, "established presentation format" was considered the most necessary improvement, whereas in the posttest period this changed to "incorporate more QI." For resident presentations evaluated using the situation, background, assessment, and review/recommendations (SBAR) tool, all evaluations, from all participants, improved over time. The M&M conference is an essential

  18. Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

    International Nuclear Information System (INIS)

    Reuscher, J.A.; Pickard, P.S.

    1976-01-01

    The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

  19. MSFR TRU-burning potential and comparison with an SFR

    Energy Technology Data Exchange (ETDEWEB)

    Fiorina, C.; Cammi, A. [Politecnico di Milano: Via La Masa 34, 20136 Milan (Italy); Franceschini, F. [Westinghouse Electric Company LL: 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States); Krepel, J. [Paul Scherrer Institut - PSI WEST, 5234 Villigen (Switzerland)

    2013-07-01

    The objective of this work is to evaluate the Molten Salt Fast Reactor (MSFR) potential benefits in terms of transuranics (TRU) burning through a comparative analysis with a sodium-cooled FR. The comparison is based on TRU- and MA-burning rates, as well as on the in-core evolution of radiotoxicity and decay heat. Solubility issues limit the TRU-burning rate to 1/3 that achievable in traditional low-CR FRs (low-Conversion-Ratio Fast Reactors). The softer spectrum also determines notable radiotoxicity and decay heat of the equilibrium actinide inventory. On the other hand, the liquid fuel suggests the possibility of using a Pu-free feed composed only of Th and MA (Minor Actinides), thus maximizing the MA burning rate. This is generally not possible in traditional low-CR FRs due to safety deterioration and decay heat of reprocessed fuel. In addition, the high specific power and the lack of out-of-core cooling times foster a quick transition toward equilibrium, which improves the MSFR capability to burn an initial fissile loading, and makes the MSFR a promising system for a quick (i.e., in a reactor lifetime) transition from the current U-based fuel cycle to a novel closed Th cycle. (authors)

  20. Fuel elements in the core of the reactor Pegase. Description, successive improvements, actual possibilities

    International Nuclear Information System (INIS)

    Desandre-Navarre, Ch.; Lerouge, B.; Schwartz, J.P.

    1967-01-01

    The core of the research reactor Pegase, in operation at the Cadarache Nuclear Research Centre since 1983, contains fuel elements made from rolled plates of an aluminium-enriched uranium alloy whose characteristics have been changed several times. This report describes the modifications which have been made to these fuel elements with a view both to improving the technical qualities of the reactor and to decreasing its operational costs. Special attention is paid to the neutron aspects of the topic and in particular to the problem of the long-term modification of the reactivity. The 1966 results (30 per cent burn-up associated with only slight movement of the control rods) are particularly satisfying and can probably still be improved in the future. (authors) [fr

  1. Technical specification improvements to containment heat removal and emergency core cooling systems: Final report

    International Nuclear Information System (INIS)

    Sullivan, W.P.; Ha, C.; Pentzien, D.C.; Visweswaran, S.

    1988-07-01

    This report presents the results of an analysis for technical specification improvements to the emergency core cooling systems (ECCS) and containment heat removal systems (EPRI Research Project 2142-3). The objective of this project is to further develop a reliability- and risk-based methodology to provide improvements by considering groups of surveillance test intervals and allowed out-of-service times jointly. This was done for the technical specifications for the ECCS, containment heat removal equipment, and supporting systems of a boiling water reactor plant. The project (1) developed a methodology for optimizing groups of surveillance test intervals and allowed out-of-service times jointly, (2) applied the methodology in a case study of a specific operating plant, Hatch-2, and (3) evaluated benefits of the application. The results of the case study demonstrate that beneficial technical specification improvements can be realized with application of the methodology. By tightening a small group of sensitive surveillance test intervals (STIs) and allowed out-of-service times (AOTs), a larger group of less sensitive STIs and AOTs can be extended resulting in an overall plant operating cost improvement without reducing the plant safety. The reliability- and risk-based methodology and results from this project can be effectively applied for technical specification improvements at other operating plants

  2. ASN’s actions in GEN IV reactors and Sodium Fast Reactors (SFR)

    International Nuclear Information System (INIS)

    Belot, Clotilde

    2013-01-01

    The ASN is involved in 3 actions concerning GEN IV: • Overview of nuclear reactor GEN IV systems; • Specific analysis about transmutation; • Prototype reactor ASTRID (SFR). Furthermore theses actions are in the beginning (no conclusions or results available)

  3. Manufacture and Erection of SFR Components: Feedback from PFBR Experience

    International Nuclear Information System (INIS)

    Chellapandi, P.

    2013-01-01

    Unique Features of SFR Components: • Large diameter thin walled shell and slender structures calling for stringent tolerances posing challenges in manufacturing, handling and erection. • Single side welds are unavoidable at some difficult locations. • In-service inspection is difficult. • Residual stresses should be minimum calling for robust heat treatment strategy. • Minimum number of materials to be used from reliability point of view (but not preferred from economic considerations). • Mainly austenitic stainless steels calling for careful considerations for welding without significant weld repairs and distortions. • Reactor assembly components decide the project time schedule (large manufacturing, assembly and erection time). • Leak tightness is very important in view of resulting sodium leaks. • Limited experience on manufacturing and erection of components. • Design and manufacturing codes still evolvingPFBR Reactor Assembly – Major Lessons: • Grid plate Large number of sleeves, posing difficulty in assembly, hard facing of large diameter plates and heavy flange construction. • Roof slab Large box type structure with many penetrations – complicated manufacturing process, time consuming and difficulty to overcome lamellar tearing problems. • Inclined Fuel Transfer Machine Complex manufacturing processes leading to large time and extensive qualification tests. • Increase of number of primary pipes – essential for enhancing safety. • Integration of components manufactured by different industries took unduly long time

  4. Comparison of Design Concepts for SFR under Development

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Namduk; Choi, Yongwon; Bae, Moohoon; Shin, Andong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    The goal of ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) with a capacity of 600 MWe is to study the technical demonstration that can be scaled up to commercial reactor. It was expected that the success of ASTRID project could eventually lead to operation of industrial reactor around 2040. On 2012, ASTRID designer has submitted the DOrS (Dossier d’Orientations de Sûreté, Safety Orientation Document) for ASTRID to IRSN and IRSN has issued a report after reviewing the DOrS. The report DOrS itself is not available publicly, intellectual property might be the reason, but the review document of IRSN is open to public, so we can understand the basic concept of ASTRID by IRSN report. The DOrS of ASTRID and the TTR for PGSFR have not the same format and also the same purpose, so it is not easy to compare the two design concepts directly. But, still, we think the concepts could be compared in a very general way. Thus, in this paper we have presented the very short comparison results of the two SFR design. Our opinion after first reviewing the TTR is that the PGSFR needs to be designed in a more systematic way. The requirements are coming basically from the previous document used for SMART licensing and do not show prototype reactor specific characters.

  5. Unexpected improvement in core autism spectrum disorder symptoms after long-term treatment with probiotics

    Directory of Open Access Journals (Sweden)

    Enzo Grossi

    2016-08-01

    Full Text Available Objectives: Autism spectrum disorder is a neurodevelopmental condition that typically displays socio-communicative impairment as well as restricted stereotyped interests and activities, in which gastrointestinal disturbances are commonly reported. We report the case of a boy with Autism Spectrum Disorder (ASD diagnosis, severe cognitive disability and celiac disease in which an unexpected improvement of autistic core symptoms was observed after four months of probiotic treatment. Method: The case study refers to a 12 years old boy with ASD and severe cognitive disability attending the Villa Santa Maria Institute in resident care since 2009. Diagnosis of ASDs according to DSM-V criteria was confirmed by ADOS-2 assessment (Autism Diagnostic Observation Schedule. The medication used was VSL#3, a multi-strain mixture of ten probiotics. The treatment lasted 4 weeks followed by a four month follow-up. The rehabilitation program and the diet was maintained stable in the treatment period and in the follow up. ADOS-2 was assessed six times: two times before starting treatment; two times during the treatment and two times after interruption of the treatment. Results: The probiotic treatment reduced the severity of abdominal symptoms as expected but an improvement in Autistic core symptoms was unexpectedly clinically evident already after few weeks from probiotic treatment start. The score of Social Affect domain of ADOS improved changing from 20 to 18 after two months treatment with a further reduction of 1 point in the following two months. The level 17 of severity remained stable in the follow up period. It is well known that ADOS score does not fluctuate spontaneously along time in ASD and is absolutely stable. Conclusions: The appropriate use of probiotics deserves further research, which hopefully will open new avenues in the fight against ASD.

  6. Unexpected improvement in core autism spectrum disorder symptoms after long-term treatment with probiotics.

    Science.gov (United States)

    Grossi, Enzo; Melli, Sara; Dunca, Delia; Terruzzi, Vittorio

    2016-01-01

    Autism spectrum disorder is a neurodevelopmental condition that typically displays socio-communicative impairment as well as restricted stereotyped interests and activities, in which gastrointestinal disturbances are commonly reported. We report the case of a boy with Autism Spectrum Disorder (ASD) diagnosis, severe cognitive disability and celiac disease in which an unexpected improvement of autistic core symptoms was observed after four months of probiotic treatment. The case study refers to a 12 years old boy with ASD and severe cognitive disability attending the Villa Santa Maria Institute in resident care since 2009. Diagnosis of ASDs according to DSM-V criteria was confirmed by ADOS-2 assessment (Autism Diagnostic Observation Schedule). The medication used was VSL#3, a multi-strain mixture of ten probiotics. The treatment lasted 4 weeks followed by a four month follow-up. The rehabilitation program and the diet was maintained stable in the treatment period and in the follow up. ADOS-2 was assessed six times: two times before starting treatment; two times during the treatment and two times after interruption of the treatment. The probiotic treatment reduced the severity of abdominal symptoms as expected but an improvement in Autistic core symptoms was unexpectedly clinically evident already after few weeks from probiotic treatment start. The score of Social Affect domain of ADOS improved changing from 20 to 18 after two months treatment with a further reduction of 1 point in the following two months. The level 17 of severity remained stable in the follow up period. It is well known that ADOS score does not fluctuate spontaneously along time in ASD and is absolutely stable. The appropriate use of probiotics deserves further research, which hopefully will open new avenues in the fight against ASD.

  7. Establishment of Collaboration System for SFR Technology Development between Korea and France

    International Nuclear Information System (INIS)

    Yoo, Jae Woon; Kim, Yeong Il; Choi, Jong Hyeun; Seong, Seung Hwan; Eoh, Jae Hyuk; Jeong, Hae Yong; Hahn, Do Hee; Lee, Yong Bum; Chang, Jin Wook; Lee, Dong Uk

    2010-03-01

    1) Review of the technical status and plane associated with STCs on the SFR R and D · The objective of the study was accomplished by constructing an human network and investigating the status on the following 5 STCs - Analysis of BFS -73 -1 and 75-1 critical experiments using ERANOS. - SFR : Elector-magnetic pumps or mechanical pumps; criteria for selection, description and modeling - Phenix end of life tests - SC-CO2 Brayton cycle : Investigation of sodium-carbon dioxide interactions; potential consequences on reactor operation - Evaluation of lead-bismuth eutectic (LBE) coolant for SFR intermediate loop 2) Holding KAERI-CEA SFR technical meeting on STCs - Final investigation at SFR technical meeting held at CEA Cadarache from Jan. 5 to Jan. 7 in 2010. - Agreement on further action plan for completing the STC - Deduction of future collaboration topics and agreed to submit into the next JCCNE - Agreed to hold next SFR technical meeting in Korea on around October 2010

  8. An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

    Directory of Open Access Journals (Sweden)

    Wuseong You

    2017-12-01

    Full Text Available In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

  9. Improvement of neutronic calculations on a Masurca core using adaptive mesh refinement capabilities

    International Nuclear Information System (INIS)

    Fournier, D.; Archier, P.; Le Tellier, R.; Suteau, C.

    2011-01-01

    The simulation of 3D cores with homogenized assemblies in transport theory remains time and memory consuming for production calculations. With a multigroup discretization for the energy variable and a discrete ordinate method for the angle, a system of about 10"4 coupled hyperbolic transport equations has to be solved. For these equations, we intend to optimize the spatial discretization. In the framework of the SNATCH solver used in this study, the spatial problem is dealt with by using a structured hexahedral mesh and applying a Discontinuous Galerkin Finite Element Method (DGFEM). This paper shows the improvements due to the development of Adaptive Mesh Refinement (AMR) methods. As the SNATCH solver uses a hierarchical polynomial basis, p−refinement is possible but also h−refinement thanks to non conforming capabilities. Besides, as the flux spatial behavior is highly dependent on the energy, we propose to adapt differently the spatial discretization according to the energy group. To avoid dealing with too many meshes, some energy groups are joined and share the same mesh. The different energy-dependent AMR strategies are compared to each other but also with the classical approach of a conforming and highly refined spatial mesh. This comparison is carried out on different quantities such as the multiplication factor, the flux or the current. The gain in time and memory is shown for 2D and 3D benchmarks coming from the ZONA2B experimental core configuration of the MASURCA mock-up at CEA Cadarache. (author)

  10. PGSFR Core Thermal Design Procedure to Evaluate the Safety Margin

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Kim, Sang-Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The Korea Atomic Energy Research Institute (KAERI) has performed a SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal design is to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damage in SFR subassemblies arises from a creep induced failure. The creep limit is evaluated based on the maximum cladding temperature, power, neutron flux, and uncertainties in the design parameters, as shown in Fig. 1. In this work, the core thermal design procedures are compared to verify the present PGSFR methodology based on the nuclear plant design criteria/guidelines and previous SFR thermal design methods. The PGSFR core thermal design procedure is verified based on the nuclear plant design criteria/guidelines and previous methods in LWRs and SFRs. The present method aims to directly evaluate the fuel cladding failure and to assure more safety margin. The 2 uncertainty is similar to 95% one-side tolerance limit of 1.96 in LWRs. The HCFs, ITDP, and MCM reveal similar uncertainty propagation for cladding midwall temperature for typical SFR conditions. The present HCFs are mainly employed from the CRBR except the fuel-related uncertainty such as an incorrect fuel distribution. Preliminary PGSFR specific HCFs will be developed by the end of 2015.

  11. Synthesis of results obtained on sodium components and technology through the Generation IV International Forum SFR Component Design and Balance-of-Plant Project

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Rodriguez, G.; Kisohara, N.; Kim, J. B.; Gerber, A.; Ashurko, Y.; Toyama, S.

    2013-01-01

    Status: The viability of designing SFR components and BOP has been demonstrated with design, construction and operation of previous sodium-cooled reactors. The main objective of this R&D project is related to system performance, or by development on the use of AECS in the BOP that could allow further cost improvements. Objective: To conduct collaborative research and development of components and BOP for the SFR System. The Project has to satisfy the GIF’s criteria of safety, economy, sustainability, proliferation resistance and physical protection. Activities within this Project are addressing experimental and analytical evaluation of advanced ISI&R, LBB assessment, development of AECS with Brayton cycles, advanced SG technologies. Project activities will be based in part on the extensive historical R&D experience with component design and balance of plant for sodium-cooled fast reactors

  12. Greater Biopsy Core Number Is Associated With Improved Biochemical Control in Patients Treated With Permanent Prostate Brachytherapy

    International Nuclear Information System (INIS)

    Bittner, Nathan; Merrick, Gregory S.; Galbreath, Robert W.; Butler, Wayne M.; Adamovich, Edward; Wallner, Kent E.

    2010-01-01

    Purpose: Standard prostate biopsy schemes underestimate Gleason score in a significant percentage of cases. Extended biopsy improves diagnostic accuracy and provides more reliable prognostic information. In this study, we tested the hypothesis that greater biopsy core number should result in improved treatment outcome through better tailoring of therapy. Methods and Materials: From April 1995 to May 2006, 1,613 prostate cancer patients were treated with permanent brachytherapy. Patients were divided into five groups stratified by the number of prostate biopsy cores (≤6, 7-9, 10-12, 13-20, and >20 cores). Biochemical progression-free survival (bPFS), cause-specific survival (CSS), and overall survival (OS) were evaluated as a function of core number. Results: The median patient age was 66 years, and the median preimplant prostate-specific antigen was 6.5 ng/mL. The overall 10-year bPFS, CSS, and OS were 95.6%, 98.3%, and 78.6%, respectively. When bPFS was analyzed as a function of core number, the 10-year bPFS for patients with >20, 13-20, 10-12, 7-9 and ≤6 cores was 100%, 100%, 98.3%, 95.8%, and 93.0% (p < 0.001), respectively. When evaluated by treatment era (1995-2000 vs. 2001-2006), the number of biopsy cores remained a statistically significant predictor of bPFS. On multivariate analysis, the number of biopsy cores was predictive of bPFS but did not predict for CSS or OS. Conclusion: Greater biopsy core number was associated with a statistically significant improvement in bPFS. Comprehensive regional sampling of the prostate may enhance diagnostic accuracy compared to a standard biopsy scheme, resulting in better tailoring of therapy.

  13. InP/ZnSe/ZnS core-multishell quantum dots for improved luminescence efficiency

    Science.gov (United States)

    Greco, Tonino; Ippen, Christian; Wedel, Armin

    2012-04-01

    Semiconductor quantum dots (QDs) exhibit unique optical properties like size-tunable emission color, narrow emission peak, and high luminescence efficiency. QDs are therefore investigated towards their application in light-emitting devices (QLEDs), solar cells, and for bio-imaging purposes. In most cases QDs made from cadmium compounds like CdS, CdSe or CdTe are studied because of their facile and reliable synthesis. However, due to the toxicity of Cd compounds and the corresponding regulation (e.g. RoHS directive in Europe) these materials are not feasible for customer applications. Indium phosphide is considered to be the most promising alternative because of the similar band gap (InP 1.35 eV, CdSe 1.73 eV). InP QDs do not yet reach the quality of CdSe QDs, especially in terms of photoluminescence quantum yield and peak width. Typically, QDs are coated with another semiconductor material of wider band gap, often ZnS, to passivate surface defects and thus improve luminescence efficiency. Concerning CdSe QDs, multishell coatings like CdSe/CdS/ZnS or CdSe/ZnSe/ZnS have been shown to be advantageous due to the improved compatibility of lattice constants. Here we present a method to improve the luminescence efficiency of InP QDs by coating a ZnSe/ZnS multishell instead of a ZnS single shell. ZnSe exhibits an intermediate lattice constant of 5.67 Å between those of InP (5.87 Å) and ZnS (5.41 Å) and thus acts as a wetting layer. As a result, InP/ZnSe/ZnS is introduced as a new core-shell quantum dot material which shows improved photoluminescence quantum yield (up to 75 %) compared to the conventional InP/ZnS system.

  14. Germany: Assessment of the efficiency of a passive safety system for prevention of severe accidents for SFR

    International Nuclear Information System (INIS)

    Bubelis, E.

    2015-01-01

    The aim of the study was the evaluation of severe transient behavior in Sodium-cooled Fast Reactor (SFR) and of the impact of newly conceived inherent mitigation measures (the use of ASD – additional shutdown device). The SFR design taken for the analysis was the SFR(v2b-ST) reactor design, and the system code to be used was selected to be the SIM-SFR code. The transients chosen for evaluation of the efficiency of mitigation measures were the unprotected loss-of-flow (ULOF) and the unprotected loss-of-heat-sink (ULOHS)

  15. What kind of galaxies dominate the cosmic SFR density at z~2?

    Science.gov (United States)

    Perez-Gonzalez, P. G.; Rieke, George; Gonzalez, Anthony; Gallego, Jesus; Guzman, Rafael; Pello, Roser; Egami, Eiichi; Marcillac, D.; Pascual, S.

    2006-08-01

    We propose to obtain near-infrared (JHK-bands) spectroscopy with GEM-S+GNIRS for a sample of 12 galaxies representative of the 3 types of spitzer/MIPS 24 micron detections at 2.0≲z≲2.6: power-law galaxies, star-forming galaxies with prominent 1.6 micron bumps, and Distant Red Galaxies. These sources are located in the Chandra Deep Field South, a unique field for the study of galaxy evolution, given the top quality data available at all wavelengths. Our main goal is to characterize the mid-IR selected galaxy population at this epoch by measuring H(alpha), H(beta), [NII], and [OIII] fluxes and profiles, and combining these observations with the already merged x-ray, ultraviolet, optical, near- and mid-infrared imaging data, to obtain the most reliable estimations of the SFRs, metallicities, stellar and dynamical masses, AGN activity, and extinction properties of the luminous infrared galaxies detected by MIPS, which dominate the SFR density of the Universe at z≳2. Our targets are complementary to others selected in the rest-frame UV/optical at high-z, and they extend the H(alpha) observations of galaxies selected with ISO from z~1 to z~2.6. The work proposed here will help to interpret the results obtained by the spitzer surveys at z≳2, thus substantially improving our understanding of the formation of massive galaxies and their connection to AGN.

  16. Site investigation SFR. Fracture mineralogy and geochemistry of borehole sections sampled for groundwater chemistry and Eh. Results from boreholes KFR01, KFR08, KFR10, KFR19, KFR7A and KFR105

    Energy Technology Data Exchange (ETDEWEB)

    Sandstroem, Bjoern (WSP Sverige AB (Sweden)); Tullborg, Eva-Lena (Terralogica AB, Grabo (Sweden))

    2011-01-15

    This report is part of the complementary site investigations for the future expansion of SFR. The report presents the results obtained during a detailed mineralogical and geochemical study of fracture minerals in drill cores from borehole section sampled for groundwater chemistry and where downhole Eh measurements have been performed. The groundwater redox system comprises not only the water, but also the bedrock/fracture mineral system in contact with this water. It is thus important to gain knowledge of the solid phases in contact with the groundwater, i.e. the fracture minerals. The samples studied for mineralogy and geochemistry, here reported, were selected to represent the fracture surfaces in contact with the groundwater in the sampled borehole sections and will give input to the hydrogeochemical model (SFR SDM). The mineralogy was determined using SEM-EDS and XRD and the geochemistry of fracture filling material was analysed by ICP-AES and ICP-QMS. The most common fracture minerals in the samples are mixed layer clay (smectite-illite), illite, chlorite, calcite, quartz, adularia and albite. Other minerals identified in the borehole sections include laumontite, pyrite, barite, chalcopyrite, hematite, Fe-oxyhydroxide, muscovite, REE-carbonate, allanite, biotite, asphaltite, galena, sphalerite, arsenopyrite, uranium phosphate, uranium silicate, Y-Ca silicate, monazite, xenotime, harmotome and fluorite. There are no major differences between the fracture mineralogy of the investigated borehole sections from SFR and the fracture mineralogy of the Forsmark site investigation area. The four fracture mineral generations distinguished within the Forsmark site investigation are also found at SFR. However, some differences have been observed: 1) Barite and uranium minerals are more common in the SFR fractures, 2) clay minerals like mixed layer illite-smectite and illite dominates in contrast to Forsmark where corrensite is by far the most common clay mineral and, 3

  17. Site investigation SFR. Fracture mineralogy and geochemistry of borehole sections sampled for groundwater chemistry and Eh. Results from boreholes KFR01, KFR08, KFR10, KFR19, KFR7A and KFR105

    International Nuclear Information System (INIS)

    Sandstroem, Bjoern; Tullborg, Eva-Lena

    2011-01-01

    This report is part of the complementary site investigations for the future expansion of SFR. The report presents the results obtained during a detailed mineralogical and geochemical study of fracture minerals in drill cores from borehole section sampled for groundwater chemistry and where downhole Eh measurements have been performed. The groundwater redox system comprises not only the water, but also the bedrock/fracture mineral system in contact with this water. It is thus important to gain knowledge of the solid phases in contact with the groundwater, i.e. the fracture minerals. The samples studied for mineralogy and geochemistry, here reported, were selected to represent the fracture surfaces in contact with the groundwater in the sampled borehole sections and will give input to the hydrogeochemical model (SFR SDM). The mineralogy was determined using SEM-EDS and XRD and the geochemistry of fracture filling material was analysed by ICP-AES and ICP-QMS. The most common fracture minerals in the samples are mixed layer clay (smectite-illite), illite, chlorite, calcite, quartz, adularia and albite. Other minerals identified in the borehole sections include laumontite, pyrite, barite, chalcopyrite, hematite, Fe-oxyhydroxide, muscovite, REE-carbonate, allanite, biotite, asphaltite, galena, sphalerite, arsenopyrite, uranium phosphate, uranium silicate, Y-Ca silicate, monazite, xenotime, harmotome and fluorite. There are no major differences between the fracture mineralogy of the investigated borehole sections from SFR and the fracture mineralogy of the Forsmark site investigation area. The four fracture mineral generations distinguished within the Forsmark site investigation are also found at SFR. However, some differences have been observed: 1) Barite and uranium minerals are more common in the SFR fractures, 2) clay minerals like mixed layer illite-smectite and illite dominates in contrast to Forsmark where corrensite is by far the most common clay mineral and, 3

  18. Quality improvement training for core medical and general practice trainees: a pilot study of project participation, completion and journal publication.

    Science.gov (United States)

    McNab, Duncan; McKay, John; Bowie, Paul

    2015-11-01

    Small-scale quality improvement projects are expected to make a significant contribution towards improving the quality of healthcare. Enabling doctors-in-training to design and lead quality improvement projects is important preparation for independent practice. Participation is mandatory in speciality training curricula. However, provision of training and ongoing support in quality improvement methods and practice is variable. We aimed to design and deliver a quality improvement training package to core medical and general practice specialty trainees and evaluate impact in terms of project participation, completion and publication in a healthcare journal. A quality improvement training package was developed and delivered to core medical trainees and general practice specialty trainees in the west of Scotland encompassing a 1-day workshop and mentoring during completion of a quality improvement project over 3 months. A mixed methods evaluation was undertaken and data collected via questionnaire surveys, knowledge assessment, and formative assessment of project proposals, completed quality improvement projects and publication success. Twenty-three participants attended the training day with 20 submitting a project proposal (87%). Ten completed quality improvement projects (43%), eight were judged as satisfactory (35%), and four were submitted and accepted for journal publication (17%). Knowledge and confidence in aspects of quality improvement improved during the pilot, while early feedback on project proposals was valued (85.7%). This small study reports modest success in training core medical trainees and general practice specialty trainees in quality improvement. Many gained knowledge of, confidence in and experience of quality improvement, while journal publication was shown to be possible. The development of educational resources to aid quality improvement project completion and mentoring support is necessary if expectations for quality improvement are to be

  19. Utilization of Encapsulated CaCO_3 in Liquid Core Capsules for Improving Lactic Acid Fermentation

    International Nuclear Information System (INIS)

    Boon-Beng, Lee; Nurul Ainina Zulkifli

    2016-01-01

    Lactic acid bacteria (LAB) have been used for food fermentation due to its fermentative ability to improve and enhance the quality of the end food products. However, the performance of LAB is affected as fermentation time elapsed because the microbial growth is inhibited by its end product, for example lactic acid. In this study, a new approach was introduced to reduce the product inhibition effect using CaCO_3 which is encapsulated in spherical liquid core capsules of diameter 3.5 mm and 3.6 mm produced through extrusion dripping method. The results showed that the pH and lactic acid concentration of LAB fermentation was well maintained by the capsules. The results of the fermentation conducted to control pH and lactic acid concentration using the capsules were better than those of the control set and comparable with that of the free CaCO_3 set. In addition, the viable cell concentration of L. casei shirota was high at the end of fermentation when the fermentation was conducted using the capsules. The results of this study suggested that the capsules have high potential to be applied for pH and lactic acid level control in LAB fermentation for various productions. (author)

  20. Performance Improvement of the Core Protection Calculator System (CPCS) by Introducing Optimal Function Sets

    International Nuclear Information System (INIS)

    Won, Byung Hee; Kim, Kyung O; Kim, Jong Kyung; Kim, Soon Young

    2012-01-01

    The Core Protection Calculator System (CPCS) is an automated device which is adopted to inspect the safety parameters such as Departure from Nuclear Boiling Ratio (DNBR) and Local Power Density (LPD) during normal operation. One function of the CPCS is to predict the axial power distributions using function sets in cubic spline method. Another function of that is to impose penalty when the estimated distribution by the spline method disagrees with embedded data in CPCS (i.e., over 8%). In conventional CPCS, restricted function sets are used to synthesize axial power shape, whereby it occasionally can draw a disagreement between synthesized data and the embedded data. For this reason, the study on improvement for power distributions synthesis in CPCS has been conducted in many countries. In this study, many function sets (more than 18,000 types) differing from the conventional ones were evaluated in each power shape. Matlab code was used for calculating/arranging the numerous cases of function sets. Their synthesis performance was also evaluated through error between conventional data and consequences calculated by new function sets

  1. Development of Basic Key Technologies for Gen IV SFR

    International Nuclear Information System (INIS)

    Han, Do Hee; Kim, Young In; Won, Byung Chool

    2008-11-01

    Technical specifications such as power capacity, type of core, clad alloy, clad barrier material, number of loops, type of SG tube have been evaluated and a optimal design concept has been identified to satisfy the technology goals of Generation IV nuclear systems. The concept for breakeven design is featured by the heat capacity of 1,200 MWe, enrichment-separated core, 2-loop, double-walled SG tube, and a long-life sensor system for in-service inspection

  2. Building the Missing Link between the Common Core and Improved Learning

    Science.gov (United States)

    Rodde, Amy Coe; McHugh, Lija

    2013-01-01

    The Common Core State Standards, adopted by 45 states and the District of Columbia, raise the bar for what students need to learn at each stage of their K-12 education. The goal is to better prepare students for college and careers. The most important thing that education leaders can do to help the Common Core succeed is to support teachers in…

  3. Improvement of Core Performance by Introduction of Moderators in a Blanket Region of Fast Reactors

    Directory of Open Access Journals (Sweden)

    Toshio Wakabayashi

    2013-01-01

    Full Text Available An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety.

  4. Methotrexate carried in lipid core nanoparticles reduces myocardial infarction size and improves cardiac function in rats

    Directory of Open Access Journals (Sweden)

    Maranhão RC

    2017-05-01

    Full Text Available Raul C Maranhão,1,2 Maria C Guido,1 Aline D de Lima,1 Elaine R Tavares,1 Alyne F Marques,1 Marcelo D Tavares de Melo,3 Jose C Nicolau,3 Vera MC Salemi,3 Roberto Kalil-Filho3 1Laboratory of Metabolism and Lipids, 2Faculty of Pharmaceutical Sciences, 3Heart Failure Unit, Clinical Cardiology Division, Heart Institute (InCor, Medical School Hospital, University of São Paulo, São Paulo, Brazil Purpose: Acute myocardial infarction (MI is accompanied by myocardial inflammation, fibrosis, and ventricular remodeling that, when excessive or not properly regulated, may lead to heart failure. Previously, lipid core nanoparticles (LDE used as carriers of the anti-inflammatory drug methotrexate (MTX produced an 80-fold increase in the cell uptake of MTX. LDE-MTX treatment reduced vessel inflammation and atheromatous lesions induced in rabbits by cholesterol feeding. The aim of the study was to investigate the effects of LDE-MTX on rats with MI, compared with commercial MTX treatment.Materials and methods: Thirty-eight Wistar rats underwent left coronary artery ligation and were treated with LDE-MTX, or with MTX (1 mg/kg intraperitoneally, once/week, starting 24 hours after surgery or with LDE without drug (MI-controls. A sham-surgery group (n=12 was also included. Echocardiography was performed 24 hours and 6 weeks after surgery. The animals were euthanized and their hearts were analyzed for morphometry, protein expression, and confocal microscopy.Results: LDE-MTX treatment achieved a 40% improvement in left ventricular (LV systolic function and reduced cardiac dilation and LV mass, as shown by echocardiography. LDE-MTX reduced the infarction size, myocyte hypertrophy and necrosis, number of inflammatory cells, and myocardial fibrosis, as shown by morphometric analysis. LDE-MTX increased antioxidant enzymes; decreased apoptosis, macrophages, reactive oxygen species production; and tissue hypoxia in non-infarcted myocardium. LDE-MTX increased adenosine

  5. The undersea location of the Swedish Final Repository for reactor waste, SFR - human intrusion aspects

    International Nuclear Information System (INIS)

    Eng, T.

    1989-01-01

    The Swedish Final Repository for reactor waste, SFR, is built under the Baltic sea close to the Forsmark nuclear power plant. Sixty metres of rock cover the repository caverns under the seabed. The depth of the Baltic sea is about 5-6 m at this location. A human intrusion scenario that in normal inland locations has shown to be of great importance, is a well that is drilled through or in the close vicinity of the repository. Since the land uplift in the SFR area is about 6 mm/year the undersea location of SFR ensures that no well will be drilled at this location for a considerable time while the area is covered by the Baltic sea

  6. The development of learning materials based on core model to improve students’ learning outcomes in topic of Chemical Bonding

    Science.gov (United States)

    Avianti, R.; Suyatno; Sugiarto, B.

    2018-04-01

    This study aims to create an appropriate learning material based on CORE (Connecting, Organizing, Reflecting, Extending) model to improve students’ learning achievement in Chemical Bonding Topic. This study used 4-D models as research design and one group pretest-posttest as design of the material treatment. The subject of the study was teaching materials based on CORE model, conducted on 30 students of Science class grade 10. The collecting data process involved some techniques such as validation, observation, test, and questionnaire. The findings were that: (1) all the contents were valid, (2) the practicality and the effectiveness of all the contents were good. The conclusion of this research was that the CORE model is appropriate to improve students’ learning outcomes for studying Chemical Bonding.

  7. Further optimization of the M1 PAM VU0453595: Discovery of novel heterobicyclic core motifs with improved CNS penetration.

    Science.gov (United States)

    Panarese, Joseph D; Cho, Hykeyung P; Adams, Jeffrey J; Nance, Kellie D; Garcia-Barrantes, Pedro M; Chang, Sichen; Morrison, Ryan D; Blobaum, Anna L; Niswender, Colleen M; Stauffer, Shaun R; Conn, P Jeffrey; Lindsley, Craig W

    2016-08-01

    This Letter describes the continued chemical optimization of the VU0453595 series of M1 positive allosteric modulators (PAMs). By surveying alternative 5,6- and 6,6-heterobicylic cores for the 6,7-dihydro-5H-pyrrolo[3,4-b]pyridine-5-one core of VU453595, we found new cores that engendered not only comparable or improved M1 PAM potency, but significantly improved CNS distribution (Kps 0.3-3.1). Moreover, this campaign provided fundamentally distinct M1 PAM chemotypes, greatly expanding the available structural diversity for this valuable CNS target, devoid of hydrogen-bond donors. Copyright © 2016 Elsevier Ltd. All rights reserved.

  8. Construction of in-situ creep strain test facility for the SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Heo, Hyeong Min; Kim, Jun Hwan; Kim, Sung Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, in-situ laser inspection creep test machine was developed for the measuring the creep strain of SFR fuel cladding materials. Ferritic-martensitic steels are being considered as an attractive candidate material for a fuel cladding of a SFR due to their low expansion coefficients, high thermal conductivities and excellent irradiation resistances to a void swelling. HT9 steel (12CrMoVW) is initially developed as a material for power plants in Europe in the 1960. This steel has experienced to expose up to 200dpa in FFTE and EBR-II. Ferritic-Martensitic steel's maximum creep strength in existence is 180Mpa for 106 hour 600 .deg., but HT9 steel is 60Mpa. Because SFR is difficult to secure in developing and applying materials, HT9 steel has accumulated validated data and is suitable for SFR component. And also, because of its superior dimensional stability against fast neutron irradiation, Ferritic-martensitic steel of 9Cr and 12Cr steels, such as HT9 and FC92(12Cr-2W) are preferable to utilize in the fuel cladding of an SFR in KAERI. The pressurized thermal creep test of HT9 and FC92 claddings are being conducted in KAERI, but the change of creep strain in cladding is not easy to measure during the creep test due to its pressurized and closed conditions. In this paper, in-situ laser inspection pressurized creep test machine developed for SFR fuel cladding specimens is described. Moreover, the creep strain rate of HT9 at 650 .deg. C was examined from the in-situ laser inspection pressurized creep test machine.

  9. X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    In KINS (Korea Institute of Nuclear Safety), to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable.

  10. X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel

    International Nuclear Information System (INIS)

    Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk

    2016-01-01

    In KINS (Korea Institute of Nuclear Safety), to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable

  11. Use of TRIGA flip fuel for improved in-core irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Whittemore, W L [General Atomic Co., San Diego, CA (United States)

    1974-07-01

    Use of standard TRIGA fuel (20% enriched uranium) in a reactor provides a suitable facility for in-core irradiations. However, large numbers of in-core samples irradiated for long periods (many months) can be handled more economically with a TRIGA loaded with FLIP fuel. As an example, ten or more in-core thermionic devices (each worth 50 to 80 cents with respect to a water-filled position) were irradiated in the Mark III TRIGA at General Atomic Company for 18 months with only a modest change in excess reactivity due to core burnup. A core loading of FLIP fuel has been added to the General Atomic Mark F reactor in order to provide numerous in-core irradiation sites for the production of radioisotopes. Since the worth of a 500-gram sample of a molybdenum compound (used for the production of {sup 99}Mo) is about 25 to 50 cents with respect to a water-filled position, use of a FLIP- TRIGA core will permit the irradiation of more than 5 kilograms of a molybdenum compound. A procedure is under development for the production of {sup 99}Mo with relatively high specific activity. Several techniques to concentrate {sup 99}Mo have been tested experimentally. The results will be reported. (author)

  12. A motional Stark effect diagnostic analysis routine for improved resolution of iota in the core of the large helical device.

    Science.gov (United States)

    Dobbins, T J; Ida, K; Suzuki, C; Yoshinuma, M; Kobayashi, T; Suzuki, Y; Yoshida, M

    2017-09-01

    A new Motional Stark Effect (MSE) analysis routine has been developed for improved spatial resolution in the core of the Large Helical Device (LHD). The routine was developed to reduce the dependency of the analysis on the Pfirsch-Schlüter (PS) current in the core. The technique used the change in the polarization angle as a function of flux in order to find the value of diota/dflux at each measurement location. By integrating inwards from the edge, the iota profile can be recovered from this method. This reduces the results' dependency on the PS current because the effect of the PS current on the MSE measurement is almost constant as a function of flux in the core; therefore, the uncertainty in the PS current has a minimal effect on the calculation of the iota profile. In addition, the VMEC database was remapped from flux into r/a space by interpolating in mode space in order to improve the database core resolution. These changes resulted in a much smoother iota profile, conforming more to the physics expectations of standard discharge scenarios in the core of the LHD.

  13. Generation IV SFR Nuclear Reactors: Under Sodium Robotics for ASTRID

    International Nuclear Information System (INIS)

    Jouan-de-Kervenoael, T.; Rey, F.; Baque, F.

    2013-06-01

    For non-removable components of the future ASTRID prototype, repair operations will be performed in a gas environment. If the faulty area is located under the sodium free level, the gas-tight system will have to contain the inspection and repair tools and to protect them from the surrounding liquid sodium. Concerning repair tools, the unique laser tool has been selected for future SFRs: the repair scenario for in-sodium structures will first involve removing the sodium (after bulk draining), then machining and finally welding. Concerning conventional tools (brush or gas blower for sodium removal, milling machine for machining and TIG for welding for which its feasibility was demonstrated in the 1990's) are still considered as a back-up solution. The maintenance of future ASTRID nuclear reactor prototype (inspection, repair) will be performed during shut down periods with some robotic carriers which have to be introduced within the main vessel, in primary 200 deg. C sodium coolant with argon gas cover. Inspection campaigns will be 20 days long. These robots (or carriers) will allow bringing inspection and repairing tools up to concerned components and structures. The needed degrees of freedom associated to these operations will be assumed either directly by the carrier itself or by specifics lower end carrier device for accurate local positioning. Several carriers will be designed, well adapted to specific needs: type of maintenance operation and location of inspection and repair sites. Feedback experience was gained during Superphenix SFR operation with the MIR robot which allowed to successfully make the Non Destructive Examination of main vessel welding joints, the carrier being outside bulk sodium. Operating conditions for the ASTRID robots will be harder from those of the MIR robot: temperature ranging from 180 deg. C to 200 deg. C, radiation dose ranging from 105 to 106 Gy, but mainly direct immersion within liquid sodium coolant. At the design phase of

  14. Cladding defects in hollow core fibers for surface mode suppression and improved birefringence

    DEFF Research Database (Denmark)

    Michieletto, Mattia; Lyngso, J. K.; Lægsgaard, Jesper

    2014-01-01

    We demonstrate a novel polarization maintaining hollow-core photonic bandgap fiber geometry that reduces the impact of surface modes on fiber transmission. The cladding structure is modified with a row of partially collapsed holes to strip away unwanted surface modes. A theoretical investigation...... of the surface mode stripping is presented and compared to the measured performance of four 7-cells core fibers that were drawn with different collapse ratio of the defects. The varying pressure along the defect row in the cladding during drawing introduces an ellipticity of the core. This, combined...... with the presence of antiresonant features on the core wall, makes the fibers birefringent, with excellent polarization maintaining properties. (C) 2014 Optical Society of America...

  15. Influence Factors and Improvement Recommendations for Core Competency of Township Enterprises

    OpenAIRE

    Zhang, Chengjun

    2014-01-01

    Core competency of township enterprises may be influenced from the property right, technology, scale operation, financial management and talent. In view of these influence factors, township enterprises should conduct technological innovation, bring into full play functions of talents, promote corporate culture of township enterprises, attach great importance to development of core products and innovation of relevant systems, and establish market information platform for township enterprises.

  16. Study of risk reduction by improving operation of reactor core isolation cooling system

    International Nuclear Information System (INIS)

    Watanabe, Yamato; Tazai, Ayuko; Yamagishi, Shohei; Muramatsu, Ken; Muta, Hitoshi

    2014-01-01

    The Fukushima Daiichi nuclear power plant fell into a station blackout (SBO) due to the earthquake and tsunami in which most of the core cooling systems were disabled. In the units 2 and 3, water injection to the core was performed only by water injection system with turbine driven pumps. In particular, it is inferred from observed plant parameters that the reactor core isolation cooling system (RCIC) continued its operation much longer than it was originally expected (8 hours). Since the preparation of safety measures did not work, the reactor core damaged. With a view to reduce risk of station blackout events in a BWR by accident management, this study investigated the efficacy of operation procedures that takes advantage of RCIC which can be operated with only equipment inside reactor building and does not require an AC power source. The efficacy was assessed in this study by two steps. The first step is a thermal hydraulic analysis with the RETRAN3D code to estimate the potential extension of duration of core cooling by RCIC and the second step is the estimation of time required for recovery of off-site power from experiences at nuclear power stations under the 3.11 earthquake. This study showed that it is possible to implement more reliable measures for accident termination and to greatly reduce the risk of SBO by the installation of accident management measures with use of RCIC for extension of core cooling under SBO conditions. (author)

  17. Project SAFE. Update of the SFR-1 safety assessment. Phase 1. Appendix A4: Far-field

    International Nuclear Information System (INIS)

    Follin, S.; Andersson, Johan; Holmen, J.; Axelsson, C.L.

    1998-01-01

    This appendix has identified potential needs for updated hydrogeological modelling of the SFR in connection to the planned update of the performance assessment of the SFR within the framework of the SAFE-project. The objectives of such updated modelling should be to present a credible representation of the hydrogeological system, to explore effects of seals and repository extensions and to provide input to the release and transport calculations of the assessment. The last objective has led to the conclusion that an important focus of the modelling should be to determine the flow through the vaults under different conditions as this flow appear to be a very important quantity in the radionuclide release calculations. The suggested modelling should use relevant data and apply modern modelling tools and techniques, but should be geared towards the objectives. For this reasons it is suggested to apply a set of complementary and sometimes nested approaches, where each model approach is set up in order to address a specific set of questions. Answering these questions would motivate simplifications made in subsequent steps of the modelling. To the extent possible the models should be compared with existing data on flow and Baltic water breakthrough. However, in making such comparisons the accuracy of the measurements and the precision of the models need to be considered. A one-to-one match cannot be expected. It appears that careful geochemical evaluation of the site would only be necessary if more credit is placed on migration in the geosphere. If such an evaluation is considered it should be co-ordinated with the regional groundwater modelling. The issue of gas production should be reconsidered in a scenario and process analysis of SFR. However, given the strong conclusions already made it appears that gas migration in the rock will still remain as a minor issue. The major assumptions going into the analysis of the near-field in the final safety report and the deepened

  18. Short-term variations in core surface flow resolved from an improved method of calculating observatory monthly means

    DEFF Research Database (Denmark)

    Olsen, Nils; Whaler, K. A.; Finlay, Chris

    2014-01-01

    Monthly means of the magnetic field measurements taken by ground observatories are a useful data source for studying temporal changes of the core magnetic field and the underlying core flow. However, the usual way of calculating monthly means as the arithmetic mean of all days (geomagnetic quiet...... as well as disturbed) and all local times (day and night) may result in contributions from external (magnetospheric and ionospheric) origin in the (ordinary, omm) monthly means. Such contamination makes monthly means less favourable for core studies. We calculated revised monthly means (rmm......), and their uncertainties, from observatory hourly means using robust means and after removal of external field predictions, using an improved method for characterising the magnetospheric ring current. The utility of the new method for calculating observatory monthly means is demonstrated by inverting their first...

  19. [Improvement of sensitivity in the second generation HCV core antigen assay by a novel concentration method using polyethylene glycol (PEG)].

    Science.gov (United States)

    Higashimoto, Makiko; Takahashi, Masahiko; Jokyu, Ritsuko; Syundou, Hiromi; Saito, Hidetsugu

    2007-11-01

    A HCV core antigen (Ag) detection assay system, Lumipulse Ortho HCV Ag has been developed and is commercially available in Japan with a lower detection level limit of 50 fmol/l, which is equivalent to 20 KIU/ml in PCR quantitative assay. HCV core Ag assay has an advantage of broader dynamic range compared with PCR assay, however the sensitivity is lower than PCR. We developed a novel HCV core Ag concentration method using polyethylene glycol (PEG), which can improve the sensitivity five times better than the original assay. The reproducibility was examined by consecutive five-time measurement of HCV patients serum, in which the results of HCV core Ag original and concentrated method were 56.8 +/- 8.1 fmol/l (mean +/- SD), CV 14.2% and 322.9 +/- 45.5 fmol/l CV 14.0%, respectively. The assay results of HCV negative samples in original HCV core Ag were all 0.1 fmol/l and the results were same even in the concentration method. The results of concentration method were 5.7 times higher than original assay, which was almost equal to theoretical rate as expected. The assay results of serially diluted samples were also as same as expected data in both original and concentration assay. We confirmed that the sensitivity of HCV core Ag concentration method had almost as same sensitivity as PCR high range assay in the competitive assay study using the serially monitored samples of five HCV patients during interferon therapy. A novel concentration method using PEG in HCV core Ag assay system seems to be useful for assessing and monitoring interferon treatment for HCV.

  20. Development of Ultrasonic Visual Inspection Program for In-Vessel Structures of SFR

    International Nuclear Information System (INIS)

    Joo, Y. S.; Park, C. G.; Lee, J. H.

    2009-02-01

    As the liquid sodium of a sodium-cooled fast reactor (SFR) is opaque to light, a conventional visual inspection is unavailable for the evaluation of the in-vessel structures under a sodium level. ASME Section XI Division 3 provides rules and guidelines for an in-service inspection (ISI) and testing of the components of SFR. For the ISI of in-vessel structures, the ASME code specifies visual examinations. An ultrasonic wave should be applied for an under-sodium visual inspection of the in-vessel structures. The plate-type waveguide sensor has been developed and the feasibility of the waveguide sensor technique has been successfully demonstrated for an ultrasonic visual inspection of the in-vessel structures of SFR. In this study, the C-scan image mapping program (Under-Sodium MultiView) is developed to apply this waveguide sensor technology to an under-sodium visual inspection of in-vessel structures in SFR by using a LabVIEW graphical programming language. The Under-Sodium MultiVIEW program has the functions of a double rotating scanner motion control, a high power pulser receiver control, a image mapping and a signal processing. The performance of Under-Sodium MultiVIEW program was verified by a C-scanning test

  1. Evaluation of Spent Fuel Recycling Scenario using Pyro-SFR related System

    International Nuclear Information System (INIS)

    Lee, Yong Kyo; Kim, Sang Ji; Kim, Young Jin

    2014-01-01

    It is needed to validate whether the recycling scenario connecting pyro-processing and sodium-cooled fast reactor(SFR) is promising or not. The latest technologies of pyro-processing are applied to SFR and the recycling scenario is evaluated through the SFR's performance analysis. The analyzed SFR is KALIMER-600 TRU burner which purpose is to transmute transuranics (TRU). National policy of CANDU SF management has not been decided yet. However, the stored quantity of this SF is large enough not to be neglected. So this study includes additionally the recycling scenario of CANDU SF. Adopting the mass ratio of TRU and RE recovered in pyro-processing is 4 to 1 on PWR SF recycling, the sodium void reactivity is higher than design basis of metal fuel. So the current pyro-processing technology is may not be acceptable. If pyro-processing technology of CANDU SF is assumed to be the same as PWR's case, CANDU recycling scenario is acceptable. Transmutation performance is worse than PWR's, while the sodium void reactivity is within design limit

  2. Establishment of Experimental Apparatus and Mechanical Test for SFR Metallic Fuel

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Lee, Chong Tak; Oh, Seok Jin; Ko, Young Mo; Kim, Ki Hwan; Woo, Yoon Myung; Lee, Chan Bock

    2010-12-01

    U-Zr binary alloys and U-Zr-Ce ternary alloys as SFR surrogate metallic fuels were fabricated by a casting process. Tensile tests were performed to evaluate the mechanical properties of the fuels. As a results, the mechanical properties such as yield strength, ultimate tensile strength, and elongation were measured. In this report, these experimental results are presented

  3. Improvements in Sand Mold/Core Technology: Effects on Casting Finish

    Energy Technology Data Exchange (ETDEWEB)

    Prof. John J. Lannutti; Prof. Carroll E. Mobley

    2005-08-30

    In this study, the development and impact of density gradients on metal castings were investigated using sand molds/cores from both industry and from in-house production. In spite of the size of the castings market, almost no quantitative information about density variation within the molds/cores themselves is available. In particular, a predictive understanding of how structure and binder content/chemistry/mixing contribute to the final surface finish of these products does not exist. In this program we attempted to bridge this gap by working directly with domestic companies in examining the issues of surface finish and thermal reclamation costs resulting from the use of sand molds/cores. We show that these can be substantially reduced by the development of an in-depth understanding of density variations that correlate to surface finish. Our experimental tools and our experience with them made us uniquely qualified to achieve technical progress.

  4. Data for calibration and validation of numerical models at SFR Nuclear Waste Repository

    International Nuclear Information System (INIS)

    Axelsson, Carl-Lennart

    1997-12-01

    The final repository for low and intermediate radioactive waste, SFR, is located below the Baltic, offshore of the nuclear power plant at Forsmark. The current operating permit for SKB stipulates that the safety assessment is updated at least every ten year. In response, SKB has started the SAFE project which aims at submitting a complete revised safety analysis before or during the year 2000. The current report is part of the far-field analyses in SAFE and provides information that can be used in a revised hydrogeological modelling of the facility. Information have been collected mainly during the construction phase of SFR, 1983 - 88, and the operation phase from 1988. The specific information that is available for the construction phase is: pressure responses in different bore holes when pumping in one bore hole, groundwater pressure in sections of bore holes, inflow to different parts of the SFR, and groundwater chemistry and isotope analyses in sections of bore holes. During the operation phase, the following information is available: ground-water pressure in sections of bore holes, inflow to different parts of the SFR facility, and groundwater chemistry and isotope analyses in sections of bore holes. The important issues in the groundwater modelling for the performance assessment study of SFR is the amount of water that flows through the storage caverns and the silo together with the possible retention and adsorption in the rock mass, i.e. the flow paths from the repository parts. Thus, the most important type of information is the inflow measurements made in different parts of SFR. The groundwater chemistry may be used to understand the flow pattern and mixing of water with various origin such as fresh groundwater, saline rock/fracture groundwater and Baltic Sea water, especially to predict breakthrough time for the Baltic Sea water at different bore hole sections in fracture zones. The report discusses especially the availability and evolution of inflow and

  5. Q-profile evolution and improved core electron confinement in the full current drive operation on Tore Supra

    International Nuclear Information System (INIS)

    Litaudon, X.; Peysson, Y.; Aniel, T.; Huysmans, G.; Imbeaux, F.; Joffrin, E.; Lasalle, J.; Lotte, Ph.; Schunke, B.; Segui, J.; Tresset, G.; Zabiego, M.

    2000-12-01

    The formation of a core region with improved electron confinement is reported in the recent full current drive operation of Tore Supra where the plasma current is sustained with the Lower Hybrid, LH, wave. Current profile evolution and thermal electron transport coefficients are directly assessed using the data of the new fast electron Bremsstrahlung tomography that provides the most accurate determination of the LH current and power deposition profiles. The spontaneous rise of the core electron temperature observed a few seconds after the application of the LH power is ascribed to a bifurcation towards a state of reduced electron transport. The role of the magnetic shear is invoked to partly stabilize the anomalous electron turbulence. The electron temperature transition occurs when the q-profile evolves towards a non-inductive state with a non-monotonic shape i.e. when the magnetic shear is reduced close to zero in the plasma core. The improved core confinement phase is often terminated by a sudden MHD activity when the minimum q approaches two. (authors)

  6. Low and intermediate level waste in SFR-1. Reference Waste Inventory 2007

    Energy Technology Data Exchange (ETDEWEB)

    Almkvist, Lisa (Vattenfall Power Consultant AB, Stockholm (SE)); Gordon, Anna (Swedish Nuclear Fuel and Waste Management Co., Stockholm (SE))

    2007-11-15

    The objective with this report is to describe all the waste and the waste package that is expected to be deposited in SFR 1 at the time of closure. The report will form the basis for the release calculation in the safety analysis for SFR 1. Three different scenarios are explored in this report; the waste inventory is based on an estimated operational lifetime of the Swedish nuclear power plants of 50 and 60 years and that closure of the SFR 1 repository will take place in 2040 or 2050 respectively. The third scenario is where the repository is full (one part where the activity adds up to 1016 Bq and one part where the repository is considered full regarding volume). In the report, data about geometries, weights, materials, chemicals and radionuclide are given. No chemotoxic material has been identified in the waste. The inventory is estimated using the Prosit-interface which extracts information from the Triumf database. The inventory is based on so called 'waste types' and the waste types' 'reference waste package'. The reference waste package combined with a prognosis of the number of waste packages to be delivered to SFR 1 gives the final waste inventory for SFR 1. All reference waste packages are thoroughly described in the appendices of this report. The reference waste packages are as far as possible based on actual experiences and measurements. The radionuclide inventory is also based on actual measurements. The inventory is based on measurements of 60Co and 137Cs in waste packages and on measurements of 239Pu and 240Pu in reactor water. Other nuclides in the inventory are calculated with correlation factors

  7. Low and intermediate level waste in SFR-1. Reference Waste Inventory 2007

    International Nuclear Information System (INIS)

    Almkvist, Lisa; Gordon, Ann

    2007-11-01

    The objective with this report is to describe all the waste and the waste package that is expected to be deposited in SFR 1 at the time of closure. The report will form the basis for the release calculation in the safety analysis for SFR 1. Three different scenarios are explored in this report; the waste inventory is based on an estimated operational lifetime of the Swedish nuclear power plants of 50 and 60 years and that closure of the SFR 1 repository will take place in 2040 or 2050 respectively. The third scenario is where the repository is full (one part where the activity adds up to 1016 Bq and one part where the repository is considered full regarding volume). In the report, data about geometries, weights, materials, chemicals and radionuclide are given. No chemotoxic material has been identified in the waste. The inventory is estimated using the Prosit-interface which extracts information from the Triumf database. The inventory is based on so called 'waste types' and the waste types' 'reference waste package'. The reference waste package combined with a prognosis of the number of waste packages to be delivered to SFR 1 gives the final waste inventory for SFR 1. All reference waste packages are thoroughly described in the appendices of this report. The reference waste packages are as far as possible based on actual experiences and measurements. The radionuclide inventory is also based on actual measurements. The inventory is based on measurements of 60 Co and 137 Cs in waste packages and on measurements of 239 Pu and 240 Pu in reactor water. Other nuclides in the inventory are calculated with correlation factors

  8. Recriticality, a Key Phenomenon to Investigate in Core Disruptive Accident Scenarios of Current and Future Fast Reactor Designs

    International Nuclear Information System (INIS)

    Maschek, W.; Rineiski, A.; Flad, M.; Kriventsev, V.; Gabrielli, F.; Morita, K.

    2012-01-01

    Final comments and conclusions: • Modern plants, should have performed better under Fukushima type event. • In future fast reactor systems significantly higher active and passive safety features are installed, which should cope with events like Fukushima. • One important lesson: put a focus on rare initiators, accident routes and consequences that are neither expected nor have been observed, events that are categorized under ‘black swans’. • Importance of severe accident research demonstrated - both analytically and experimentally for assessing and interpreting accident scenarios and developments. Precondition for developing preventive & mitigative safety measures. Passive safety measures are in the focus of advanced design options and must work under conditions of multiple loads and aggravating events. • Fast reactor systems behavior as the SFR under severe accident conditions: – In fast spectrum systems as the SFR the core is not in its neutronically most reactive configuration and SFRs may be loaded with MAs for waste management; – Recriticalities have a high probability because of the higher enrichment levels; – Short time scales have to be envisioned for core melt-down; – Decay heat levels might be significantly higher, if MA bearing fuel is involved. • Improve design by measures for prevention and/or mitigation of recriticalities; – High reliability of simulations required for proof; • Assessment of fuel relocated on peripheral structures; • Preventive/mitigating measures should not replace containment measures

  9. Core ADHD Symptom Improvement with Atomoxetine versus Methylphenidate: A Direct Comparison Meta-Analysis

    Science.gov (United States)

    Hazell, Philip L.; Kohn, Michael R.; Dickson, Ruth; Walton, Richard J.; Granger, Renee E.; van Wyk, Gregory W.

    2011-01-01

    Objective: Previous studies comparing atomoxetine and methylphenidate to treat ADHD symptoms have been equivocal. This noninferiority meta-analysis compared core ADHD symptom response between atomoxetine and methylphenidate in children and adolescents. Method: Selection criteria included randomized, controlled design; duration 6 weeks; and…

  10. Improved Fabrication of Ceramic Matrix Composite/Foam Core Integrated Structures

    Science.gov (United States)

    Hurwitz, Frances I.

    2009-01-01

    The use of hybridized carbon/silicon carbide (C/SiC) fabric to reinforce ceramic matrix composite face sheets and the integration of such face sheets with a foam core creates a sandwich structure capable of withstanding high-heatflux environments (150 W/cm2) in which the core provides a temperature drop of 1,000 C between the surface and the back face without cracking or delamination of the structure. The composite face sheet exhibits a bilinear response, which results from the SiC matrix not being cracked on fabrication. In addition, the structure exhibits damage tolerance under impact with projectiles, showing no penetration to the back face sheet. These attributes make the composite ideal for leading edge structures and control surfaces in aerospace vehicles, as well as for acreage thermal protection systems and in high-temperature, lightweight stiffened structures. By tailoring the coefficient of thermal expansion (CTE) of a carbon fiber containing ceramic matrix composite (CMC) face sheet to match that of a ceramic foam core, the face sheet and the core can be integrally fabricated without any delamination. Carbon and SiC are woven together in the reinforcing fabric. Integral densification of the CMC and the foam core is accomplished with chemical vapor deposition, eliminating the need for bond-line adhesive. This means there is no need to separately fabricate the core and the face sheet, or to bond the two elements together, risking edge delamination during use. Fibers of two or more types are woven together on a loom. The carbon and ceramic fibers are pulled into the same pick location during the weaving process. Tow spacing may be varied to accommodate the increased volume of the combined fiber tows while maintaining a target fiber volume fraction in the composite. Foam pore size, strut thickness, and ratio of face sheet to core thickness can be used to tailor thermal and mechanical properties. The anticipated CTE for the hybridized composite is managed by

  11. IMPROVEMENT OF STRATEGIC MANIPULATED FEDERAL PROPERTY THE EXAMPLE NON-CORE ASSETS OF JSC «CENTER OF NUCLEAR INDUSTRY NONCORE ASSETS» STATE CORPORATION «ROSATOM»

    OpenAIRE

    Ilya I. Rodin

    2015-01-01

    The article describes the main measures to improve the management of assets, federally-owned or private of public corporations - an inventory of the property, the recognition of non-core assets, the organization of decision-making systems, the sale of non-core assets at market value. The article provides the rationale for the creation within the large state-owned corporations specialized management companies responsible for the restructuring of non-core assets and improve management of the pr...

  12. Conceptual Design on Primary Control Rod Drive Mechanism of a Prototype Gen-IV SFR

    International Nuclear Information System (INIS)

    Lee, Jae Han; Koo, Gyeong Hoi

    2013-01-01

    This paper describes the key concept of the drive mechanism, and suggests a required motor power and reducer gears to meet the functional design requirements, and a seismic response analysis of CRDM housing is performed to check its structural integrity. An AC servo motor is selected as a CRA driving power because it uses permanent magnets and is brushless type while DC motor needs a brush and a coil rotates. The control shim motor size is constrained by a housing diameter of 250mm. The driving system has several design requirements. To calculate the motor power, the drive shaft torque is needed. One part of the drive shaft has a lead screw, driving by a ball-nut. The ball screw driver torque (Tr) is calculated by some equations as follow; A servo motor with a nominal power of 100W, a nominal torque of 0.32 N-m (max. 0.48N-m) is selected considering a safety margin. Its diameter is about 50mm. The fast drive-in motor needs a strong power to insert enforcedly the stuck CRA into core within a required time. The motor sizes are calculated by the same procedure. The diameters are in the range of 80mm to 110mm by the insertion time (10 ∼ 24 seconds). The prototype Gen-IV SFR (sodium-cooled Fast Reactor) is of 150MWe capacity. The reactor has six primary control rod assemblies(CRAs). The primary control rod is used for power control, burn-up compensation and reactor shutdown in response to demands from the plant control or protection systems. The control rod drive mechanism (CRDM) consists of the drive motor assembly, the driveline, and its housing. The driveline consists of three concentric members of a drive shaft, a tension tube, and a position indicator rod, and it connects the drive motor assembly to the CRA. Main issue is that these many driving parts shall be enclosed within a limited housing diameter because the available pitch of CRDMs is limited by 300mm

  13. Conceptual Design on Primary Control Rod Drive Mechanism of a Prototype Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Koo, Gyeong Hoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    This paper describes the key concept of the drive mechanism, and suggests a required motor power and reducer gears to meet the functional design requirements, and a seismic response analysis of CRDM housing is performed to check its structural integrity. An AC servo motor is selected as a CRA driving power because it uses permanent magnets and is brushless type while DC motor needs a brush and a coil rotates. The control shim motor size is constrained by a housing diameter of 250mm. The driving system has several design requirements. To calculate the motor power, the drive shaft torque is needed. One part of the drive shaft has a lead screw, driving by a ball-nut. The ball screw driver torque (Tr) is calculated by some equations as follow; A servo motor with a nominal power of 100W, a nominal torque of 0.32 N-m (max. 0.48N-m) is selected considering a safety margin. Its diameter is about 50mm. The fast drive-in motor needs a strong power to insert enforcedly the stuck CRA into core within a required time. The motor sizes are calculated by the same procedure. The diameters are in the range of 80mm to 110mm by the insertion time (10 ∼ 24 seconds). The prototype Gen-IV SFR (sodium-cooled Fast Reactor) is of 150MWe capacity. The reactor has six primary control rod assemblies(CRAs). The primary control rod is used for power control, burn-up compensation and reactor shutdown in response to demands from the plant control or protection systems. The control rod drive mechanism (CRDM) consists of the drive motor assembly, the driveline, and its housing. The driveline consists of three concentric members of a drive shaft, a tension tube, and a position indicator rod, and it connects the drive motor assembly to the CRA. Main issue is that these many driving parts shall be enclosed within a limited housing diameter because the available pitch of CRDMs is limited by 300mm.

  14. Transitions to improved core electron heat confinement triggered by low order rational magnetic surfaces in the stellarator TJ-II

    International Nuclear Information System (INIS)

    Estrada, T.; Medina, F.; Lopez-Bruna, D.; AscasIbar, E.; BalbIn, R.; Cappa, A.; Castejon, F.; Eguilior, S.; Fernandez, A.; Guasp, J.; Hidalgo, C.; Petrov, S.

    2007-01-01

    Transitions to improved core electron heat confinement are triggered by low order rational magnetic surfaces in TJ-II electron cyclotron heated (ECH) plasmas. Experiments are performed changing the magnetic shear around the rational surface n = 3/m = 2 to study its influence on the transition; ECH power modulation is used to look at transport properties. The improvement in the electron heat confinement shows no obvious dependence on the magnetic shear. Transitions triggered by the rational surface n = 4/m = 2 show, in addition, an increase in the ion temperature synchronized with the increase in the electron temperature. Ion temperature changes had not been previously observed either in TJ-II or in any other helical device. SXR measurements demonstrate that, under certain circumstances, the rational surface positioned inside the plasma core region precedes and provides a trigger for the transition

  15. Surrogates based multi-criteria predesign methodology of Sodium-cooled Fast Reactor cores – Application to CFV-like cores

    Energy Technology Data Exchange (ETDEWEB)

    Fabbris, Olivier [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Dardour, Saied, E-mail: saied.dardour@cea.fr [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Blaise, Patrick [CEA DEN/DER/SPEX, 13108 Saint-Paul-Lez-Durance (France); Ferrasse, Jean-Henry [Aix-Marseille Université, CNRS, ECM, M2P2 UMR 7340, 13451 Marseille (France); Saez, Manuel [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France)

    2016-08-15

    Highlights: • We developed an ERANOS calculation scheme to evaluate the neutronics of CFV cores. • We used this scheme to simulate a number if cores within a predefined study space. • Simulation results were used to build surrogate models describing CFV neutronics. • These models were used to carry on global sensitivity analyses. • The methodology helped identify the most important core design parameters. - Abstract: The Sodium-cooled Fast Reactor (SFR) core predesign process is commonly realized on the basis of expert advices and local parametric studies. As such, in-deep knowledge of physical phenomena avoids an important number of expensive simulations. However, the study space is explored only partially. To ease the computational burden metamodels, or surrogate models, can be used, to quickly evaluate the performances of a wide set of different cores, individually defined by a set of parameters (pellet diameter, fissile height…), in the study space. This paper presents the development of a simplified neutronics ERANOS reference core calculation scheme that is then implemented in the construction of the Design of Experiment (DOE) database. The surrogate models for SFR CFV-like cores performances are developed, biases and uncertainties are quantified against the CFV-v1 version. Global Sensitivity Analysis also allowed highlighting antagonist performances for the design and to propose two alternative core configurations. A broadened application of the method with an optimization of a CFV-like core is also detailed. The Pareto front of the seven selected performance parameters has been studied using eleven surrogate models, based on Artificial Neural Network (ANN). The optimization demonstrates that the CFV-v1, designed using Best Estimate codes, under given performance constraints, is Pareto optimal: no other configuration is highlighted from the Multi-Objective Optimization (MOO) study. Further MOO analysis, including a specific study on impact of new

  16. Surrogates based multi-criteria predesign methodology of Sodium-cooled Fast Reactor cores – Application to CFV-like cores

    International Nuclear Information System (INIS)

    Fabbris, Olivier; Dardour, Saied; Blaise, Patrick; Ferrasse, Jean-Henry; Saez, Manuel

    2016-01-01

    Highlights: • We developed an ERANOS calculation scheme to evaluate the neutronics of CFV cores. • We used this scheme to simulate a number if cores within a predefined study space. • Simulation results were used to build surrogate models describing CFV neutronics. • These models were used to carry on global sensitivity analyses. • The methodology helped identify the most important core design parameters. - Abstract: The Sodium-cooled Fast Reactor (SFR) core predesign process is commonly realized on the basis of expert advices and local parametric studies. As such, in-deep knowledge of physical phenomena avoids an important number of expensive simulations. However, the study space is explored only partially. To ease the computational burden metamodels, or surrogate models, can be used, to quickly evaluate the performances of a wide set of different cores, individually defined by a set of parameters (pellet diameter, fissile height…), in the study space. This paper presents the development of a simplified neutronics ERANOS reference core calculation scheme that is then implemented in the construction of the Design of Experiment (DOE) database. The surrogate models for SFR CFV-like cores performances are developed, biases and uncertainties are quantified against the CFV-v1 version. Global Sensitivity Analysis also allowed highlighting antagonist performances for the design and to propose two alternative core configurations. A broadened application of the method with an optimization of a CFV-like core is also detailed. The Pareto front of the seven selected performance parameters has been studied using eleven surrogate models, based on Artificial Neural Network (ANN). The optimization demonstrates that the CFV-v1, designed using Best Estimate codes, under given performance constraints, is Pareto optimal: no other configuration is highlighted from the Multi-Objective Optimization (MOO) study. Further MOO analysis, including a specific study on impact of new

  17. Structural improvement of unliganded simian immunodeficiency virus gp120 core by normal-mode-based X-ray crystallographic refinement

    International Nuclear Information System (INIS)

    Chen, Xiaorui; Lu, Mingyang; Poon, Billy K.; Wang, Qinghua; Ma, Jianpeng

    2009-01-01

    The structural model of the unliganded and fully glycosylated simian immunodeficiency virus gp120 core determined to 4.0 Å resolution was substantially improved using a recently developed normal-mode-based anisotropic B-factor refinement method. The envelope protein gp120/gp41 of simian and human immunodeficiency viruses plays a critical role in viral entry into host cells. However, the extraordinarily high structural flexibility and heavy glycosylation of the protein have presented enormous difficulties in the pursuit of high-resolution structural investigation of some of its conformational states. An unliganded and fully glycosylated gp120 core structure was recently determined to 4.0 Å resolution. The rather low data-to-parameter ratio limited refinement efforts in the original structure determination. In this work, refinement of this gp120 core structure was carried out using a normal-mode-based refinement method that has been shown in previous studies to be effective in improving models of a supramolecular complex at 3.42 Å resolution and of a membrane protein at 3.2 Å resolution. By using only the first four nonzero lowest-frequency normal modes to construct the anisotropic thermal parameters, combined with manual adjustments and standard positional refinement using REFMAC5, the structural model of the gp120 core was significantly improved in many aspects, including substantial decreases in R factors, better fitting of several flexible regions in electron-density maps, the addition of five new sugar rings at four glycan chains and an excellent correlation of the B-factor distribution with known structural flexibility. These results further underscore the effectiveness of this normal-mode-based method in improving models of protein and nonprotein components in low-resolution X-ray structures

  18. Site description of the SFR area at Forsmark at completion of the site investigation phase. SDM-PSU Forsmark

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-05-15

    The site descriptive model (SDM) presented in this report is an integrated model for bedrock geology, rock mechanics, bedrock hydrogeology and bedrock hydrogeochemistry of the site investigated in the SFR extension project (PSU). A description of the surface system is also included in the report. However, the surface system is not integrated with the other disciplines as new data regarding the surface system will not be available until after the completion of SDM-PSU. It is noted that SDM-PSU does not include all disciplines handled in SDM-Site Forsmark (SKB 2008b), the focus is to produce a site description that meets the needs of the SFR extension project. The overall objective of the SFR extension project is to have the application for the extension ready by 2013. This report presents an integrated site model incorporating the historic data acquired from the investigations for and construction of the existing SFR facility (1980-1986), as well as from the recent investigations for the planned extension of SFR (2008-2009). It also provides a summary of the abundant underlying data and the discipline-specific models that support the integrated site model. The description relies heavily on background reports concerning detailed data analyses and modelling in the different disciplines. It is noteworthy that the investigations conducted during the SFR extension project were guided by the choice of site prior to the investigations, which was based on the experience gained during the construction of the existing SFR facility.

  19. Site description of the SFR area at Forsmark at completion of the site investigation phase. SDM-PSU Forsmark

    International Nuclear Information System (INIS)

    2013-05-01

    The site descriptive model (SDM) presented in this report is an integrated model for bedrock geology, rock mechanics, bedrock hydrogeology and bedrock hydrogeochemistry of the site investigated in the SFR extension project (PSU). A description of the surface system is also included in the report. However, the surface system is not integrated with the other disciplines as new data regarding the surface system will not be available until after the completion of SDM-PSU. It is noted that SDM-PSU does not include all disciplines handled in SDM-Site Forsmark (SKB 2008b), the focus is to produce a site description that meets the needs of the SFR extension project. The overall objective of the SFR extension project is to have the application for the extension ready by 2013. This report presents an integrated site model incorporating the historic data acquired from the investigations for and construction of the existing SFR facility (1980-1986), as well as from the recent investigations for the planned extension of SFR (2008-2009). It also provides a summary of the abundant underlying data and the discipline-specific models that support the integrated site model. The description relies heavily on background reports concerning detailed data analyses and modelling in the different disciplines. It is noteworthy that the investigations conducted during the SFR extension project were guided by the choice of site prior to the investigations, which was based on the experience gained during the construction of the existing SFR facility

  20. Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: simulations and experimental validations [ANIMMA--2015-IO-392

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, C.; Filliatre, P.; Izarra, G. de [CEA, DEN, Cadarache, Reactor Studies Department, 13108 Saint-Paul-lez-Durance (France); Elter, Zs. [CEA, DEN, Cadarache, Reactor Studies Department, 13108 Saint-Paul-lez-Durance (France); Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE-412 96 Goeteborg (Sweden); Verma, V. [CEA, DEN, Cadarache, Reactor Studies Department, 13108 Saint-Paul-lez-Durance (France); Uppsala University, Division of Applied Nuclear Physics, Box 516, SE-75120 Uppsala (Sweden); Hamrita, H.; Bakkali, M. [CEA, DRT, LIST, Metrology, Instrumentation and Information Department, Saclay, 91191 Gif-sur-Yvette (France); Chapoutier, N.; Scholer, A.C.; Verrier, D. [AREVA NP, 10 rue Juliette Recamier F-69456 Lyon (France); Hellesen, C.; Jacobsson, S. [Uppsala University, Division of Applied Nuclear Physics, Box 516, SE-75120 Uppsala (Sweden); Pazsit, I. [Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE-412 96 Goeteborg (Sweden); Cantonnet, B.; Nappe, J.C. [PHOTONIS France, Nuclear Instrumentation, 19100 Brive-la-Gaillarde (France); Molinie, P.; Dessante, P.; Hanna, R.; Kirkpatrick, M.; Odic, E. [Supelec, Energy Department, 3 rue Joliot-Curie, 91191 Gif-sur-Yvette (France)

    2015-07-01

    France has a long experience of about 50 years in designing, building and operating sodium-cooled fast reactors (SFR) such as RAPSODIE, PHENIX and SUPER PHENIX. Fast reactors feature the double capability of reducing nuclear waste and saving nuclear energy resources by burning actinides. Since this reactor type is one of those selected by the Generation IV International Forum, the French government asked, in the year 2006, CEA, namely the French Alternative Energies and Atomic Energy Commission, to lead the development of an innovative GEN-IV nuclear- fission power demonstrator. The major objective is to improve the safety and availability of an SFR. The neutron flux monitoring (NFM) system of any reactor must, in any situation, permit both reactivity control and power level monitoring from startup to full power. It also has to monitor possible changes in neutron flux distribution within the core region in order to prevent any local melting accident. The neutron detectors will have to be installed inside the reactor vessel because locations outside the vessel will suffer from severe disadvantages; radially the neutron shield that is also contained in the reactor vessel will cause unacceptable losses in neutron flux; below the core the presence of a core-catcher prevents from inserting neutron guides; and above the core the distance is too large to obtain decent neutron signals outside the vessel. Another important point is to limit the number of detectors placed in the vessel in order to alleviate their installation into the vessel. In this paper, we show that the architecture of the NFM system will rely on high-temperature fission chambers (HTFC) featuring wide-range flux monitoring capability. The definition of such a system is presented and the justifications of technological options are brought with the use of simulation and experimental results. Firstly, neutron-transport calculations allow us to propose two in-vessel regions, namely the above-core and under-core

  1. Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

    International Nuclear Information System (INIS)

    Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

    2016-01-01

    One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

  2. A study of ECC water bypass reduction technology for an improvement of core cooling capability

    International Nuclear Information System (INIS)

    Song, C. G.; Kwon, T. S.; Yun, B. J.

    2006-02-01

    The research for the reduction of ECC water bypass fraction was mainly performed to develop the flow mechanisms that ECC water can penetrate more effectively into a lower downcomer. Evaluation were carried out for the effect of major parameters on the bypass of ECC water in the downcomer with DVI. The following various physical models were derived for the reduction of the bypass fraction of ECC water: models of changing DVI injection angle, models of rearranging relative angles of DVI nozzles, model of grooved-and-subchannel type core barrel, model of dual core barrel. CFD analysis and MARS design verification were performed for the derived models as a first step performance estimation. Through out air-water verification experiments, quantitative evaluation were performed for each model, and three most efficient models were suggested. Examination were carried out for the requirement of structural modification and the change in structural integrity due to the adoption of one of the schemes

  3. An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

    International Nuclear Information System (INIS)

    Clark, J.S.; Walton, J.T.; Mcguire, M.L.

    1992-07-01

    Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines. 11 refs

  4. Uncertainties in estimating the {sup 90}Sr and actinides inventory in SFR 1; Osaekerheter vid uppskattning av Sr-90 och aktinidinventariet i SFR 1

    Energy Technology Data Exchange (ETDEWEB)

    Ingemansson, Tor [ALARA Engineering AB, Skultuna (Sweden)

    2000-04-01

    SFR-1 is a facility for disposal of low and intermediate level radioactive waste. The uncertainty in estimation of the activity accumulated in different cleaning filters, originating in the Swedish BWR-, PWR-reactors and CLAB - the Central interim storage facility for spent nuclear fuel - has been analyzed to be 10 - 14%, depending on the methods used for measuring the activity at the power plants. Other waste or scrap contribute with approx. 1.5% of the total amount of actinides and {sup 90}Sr. The uncertainty in this fraction is about 20%. The uncertainties are surprisingly small, and explain the good agreement between estimates made with different methods.

  5. Comparative Studies of Core Thermal Hydraulic Design Methods for the Prototype Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Lim, Jae Yong; Kim, Sang Ji

    2013-01-01

    In this work, various core thermal-hydraulic design methods, which have arisen during the development of a prototype SFR, are compared to establish a proper design procedure. Comparative studies have been performed to determine the appropriate design method for the prototype SFR. The results show that the minimization method show a lower cladding midwall temperature than the fixed outlet temperature methods and superior thermal safety margin with the same coolant flow. The Korea Atomic energy Research Institute (KAERI) has performed a conceptual SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal-hydraulic design is used to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damages in SFR subassemblies are arisen from a creep induced failure. The creep limit is evaluated based on both the maximum cladding temperature and the uncertainties of the design parameters. Therefore, the core thermalhydraulic design method, which eventually determines the cladding temperature, is highly important to assure a safe and reliable operation of the reactor systems

  6. Complete Au@ZnO core-shell nanoparticles with enhanced plasmonic absorption enabling significantly improved photocatalysis

    Science.gov (United States)

    Sun, Yiqiang; Sun, Yugang; Zhang, Tao; Chen, Guozhu; Zhang, Fengshou; Liu, Dilong; Cai, Weiping; Li, Yue; Yang, Xianfeng; Li, Cuncheng

    2016-05-01

    Nanostructured ZnO exhibits high chemical stability and unique optical properties, representing a promising candidate among photocatalysts in the field of environmental remediation and solar energy conversion. However, ZnO only absorbs the UV light, which accounts for less than 5% of total solar irradiation, significantly limiting its applications. In this article, we report a facile and efficient approach to overcome the poor wettability between ZnO and Au by carefully modulating the surface charge density on Au nanoparticles (NPs), enabling rapid synthesis of Au@ZnO core-shell NPs at room temperature. The resulting Au@ZnO core-shell NPs exhibit a significantly enhanced plasmonic absorption in the visible range due to the Au NP cores. They also show a significantly improved photocatalytic performance in comparison with their single-component counterparts, i.e., the Au NPs and ZnO NPs. Moreover, the high catalytic activity of the as-synthesized Au@ZnO core-shell NPs can be maintained even after many cycles of photocatalytic reaction. Our results shed light on the fact that the Au@ZnO core-shell NPs represent a promising class of candidates for applications in plasmonics, surface-enhanced spectroscopy, light harvest devices, solar energy conversion, and degradation of organic pollutants.Nanostructured ZnO exhibits high chemical stability and unique optical properties, representing a promising candidate among photocatalysts in the field of environmental remediation and solar energy conversion. However, ZnO only absorbs the UV light, which accounts for less than 5% of total solar irradiation, significantly limiting its applications. In this article, we report a facile and efficient approach to overcome the poor wettability between ZnO and Au by carefully modulating the surface charge density on Au nanoparticles (NPs), enabling rapid synthesis of Au@ZnO core-shell NPs at room temperature. The resulting Au@ZnO core-shell NPs exhibit a significantly enhanced plasmonic

  7. Predicting core losses and efficiency of SRM in continuous current mode of operation using improved analytical technique

    International Nuclear Information System (INIS)

    Parsapour, Amir; Dehkordi, Behzad Mirzaeian; Moallem, Mehdi

    2015-01-01

    In applications in which the high torque per ampere at low speed and rated power at high speed are required, the continuous current method is the best solution. However, there is no report on calculating the core loss of SRM in continuous current mode of operation. Efficiency and iron loss calculation which are complex tasks in case of conventional mode of operation is even more involved in continuous current mode of operation. In this paper, the Switched Reluctance Motor (SRM) is modeled using finite element method and core loss and copper loss of SRM in discontinuous and continuous current modes of operation are calculated using improved analytical techniques to include the minor loop losses in continuous current mode of operation. Motor efficiency versus speed in both operation modes is obtained and compared. - Highlights: • Continuous current method for Switched Reluctance Motor (SRM) is explained. • An improved analytical technique is presented for SRM core loss calculation. • SRM losses in discontinuous and continuous current operation modes are presented. • Effect of mutual inductances on SRM performance is investigated

  8. Predicting core losses and efficiency of SRM in continuous current mode of operation using improved analytical technique

    Energy Technology Data Exchange (ETDEWEB)

    Parsapour, Amir, E-mail: amirparsapour@gmail.com [Department of Electrical Engineering, University of Isfahan, Isfahan (Iran, Islamic Republic of); Dehkordi, Behzad Mirzaeian, E-mail: mirzaeian@eng.ui.ac.ir [Department of Electrical Engineering, University of Isfahan, Isfahan (Iran, Islamic Republic of); Moallem, Mehdi, E-mail: moallem@cc.iut.ac.ir [Department of Electrical Engineering, Isfahan University of Technology, Isfahan (Iran, Islamic Republic of)

    2015-03-15

    In applications in which the high torque per ampere at low speed and rated power at high speed are required, the continuous current method is the best solution. However, there is no report on calculating the core loss of SRM in continuous current mode of operation. Efficiency and iron loss calculation which are complex tasks in case of conventional mode of operation is even more involved in continuous current mode of operation. In this paper, the Switched Reluctance Motor (SRM) is modeled using finite element method and core loss and copper loss of SRM in discontinuous and continuous current modes of operation are calculated using improved analytical techniques to include the minor loop losses in continuous current mode of operation. Motor efficiency versus speed in both operation modes is obtained and compared. - Highlights: • Continuous current method for Switched Reluctance Motor (SRM) is explained. • An improved analytical technique is presented for SRM core loss calculation. • SRM losses in discontinuous and continuous current operation modes are presented. • Effect of mutual inductances on SRM performance is investigated.

  9. Status of the Astrid core at the end of the pre-conceptual design phase 1

    International Nuclear Information System (INIS)

    Chenaud, Ms.; Devictor, N.; Mignot, G.; Varaine, F.; Venard, C.; Martin, L.; Phelip, M.; Lorenzo, D.; Serre, F.; Bertrand, F.; Alpy, N.; Le-Flem, M.; Gavoille, P.; Lavastre, R.; Richard, P.; Verrier, D.; Schmitt, D.

    2013-01-01

    Within the framework of the ASTRID project, core design studies are being conducted by the CEA with support from AREVA and EDF. The pre-conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves limiting the consequences of 1) a hypothetical control rod withdrawal accident (by minimizing the core reactivity loss during the irradiation cycle), and 2) an hypothetical loss-of-flow accident (by reducing the sodium void worth). Two types of cores are being studied for the ASTRID project. The first is based on a 'large pin/small spacing wire' concept derived from the SFR V2b, while the other is based on an innovative CFV design. A distinctive feature of the CFV core is its negative sodium void worth. In 2011, the evaluation of a preliminary version (v1) of this CFV core for ASTRID underlined its potential capacity to improve the prevention of severe accidents. An improved version of the ASTRID CFV core (v2) was proposed in 2012 to comply with all the control rod withdrawal criteria, while increasing safety margins for all unprotected-loss-of-flow (ULOF) transients and improving the general design. This paper describes the CFV v2 design options and reports on the progress of the studies at the end of pre-conceptual design phase 1 concerning: - Core performance, - Intrinsic behavior during unprotected transients, - Simulation of severe accident scenarios, - Qualification requirements. The paper also specifies the open options for the materials, sub-assemblies, absorbers, and core monitoring that will continue to be studied during the conceptual design phase. (authors)

  10. Evaluation of improved light water reactor core designs. Final progress report, September 1979. LWRCD-20

    International Nuclear Information System (INIS)

    1979-01-01

    The work conducted under this research project has developed information which supports in all respects the U.S. position evolved under the NASAP/INFCE programs with respect to the near and intermediate term potential for ore conservation in LWRs on the once-through fuel cycle. Moreover, in the even longer term, it has been confirmed that contention by Edlund and others that tight-pitch Pu/UO 2 PWR cores can achieve conversion ratios which may allow these reactors to provide a competitive energy source far into the ore-scarce post-2000 era

  11. Minutes of the kick-off Meeting of the WPRS / EGRPANS / Sodium Fast Reactor Task Force (SFR)

    International Nuclear Information System (INIS)

    Buiron, Laurent; Stauff, Nicolas; Varaine, Frederic; Blanchet, D.; Stauff, N.; Ivanov, Evgeny; Michel-Sendis, Franco; ); Mikityuk, Konstantin; Pelloni, Sandro; Ponomarev, Alexander; Kim, Taek K.; Taiwo, Temitope; Kereszturi, Andras; Van den Eynde, Gert; Kotiluoto, Petri; Juutilainen, Pauli; Lepp Anen, Jaakko

    2011-01-01

    calculations could become an optional branch of the benchmark. F. Varaine agreed also to prepare and provide an informational note detailing the methodology used at CEA for feedback coefficient calculations. A similar note could then be prepared by ANL on the methodology they use. It was requested that the Task Force should meet again at the next WPRS meeting in 2012, in a parallel session for a full day. Secretariat agreed to set up a dedicated web site working area for the Task Force where documents can be shared. This document brings together the 3 presentations (slides) given at this meeting: 1 - Fresh look from the back-end [what kinds of reactor parameters we need for the transient modeling (E. Ivanov); 2 - Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force - Review of Mandate and Planning of SFR Benchmark (F. Varaine); 3 - Sodium Fast Reactor Task Force (L. Buiron, D. Blanchet, N. Stauff)

  12. Design and Performance Evaluation of a Combined DHX unit for SFR Design Application

    International Nuclear Information System (INIS)

    Eoh, Jaehyuk; Kim, Dehee; Park, Chang-Gyu; Jeong, Ji-Young

    2015-01-01

    Based on a higher operating temperature with excellent thermal conductivity and larger thermal inertia of liquid sodium coolant, the SFR system has employed passive safety systems to ensure reliable decay heat removal (DHR) and consequential plant safety enhancement. Although a passive type DHR system has many advantages over an active one, designing a well coordinated passive system is usually more difficult than designing an effective active system. This is mainly because a cooling flow control is made directly by the system designer in an active system, while it is determined automatically by an intricate balance between the flow head loss and natural circulation head generation obtained from the density difference through the whole thermal flow system. To this end, securing a sufficient natural-circulation flow becomes one of the primary challenges for designing a reliable and successful Dh system in passive. In a current pool-type Sf design, an internal cooling flow path from the hot sodium pool to the cold pool is somewhat ambiguous owing to the split flow ratio formed in parallel paths between the intermediate heat exchangers (IHXs) and decay heat exchangers (DHXs), which results in a large uncertainty in the DHX shell-side flowrate and corresponding heat transfer to the DHR sodium loops. To improve passive the DHR performance, we proposed a new design concept with a simplified flow path from the hot pool to the cold pool through a unified flow path serially passing the DHX and IHX units. The present study aims at introducing the innovative design concept of the combined IHX-DHX unit and evaluating its design features in view of the heat transfer capability. From a comparison of the CHX performance designed by a one-dimensional approach with that made by a CFD analysis, it was quantitatively obtained that the difference in heat transfer rate is about 5.7%. It was also found that unexpected bypass flow in the shell-side CHX unit gave rise to a discrepancy

  13. Project SFR 1 SAR-08. Update of priority of FEPs from Project SAFE

    International Nuclear Information System (INIS)

    Gordon, Anna; Loefgren, Martin; Lindgren, Maria

    2008-03-01

    SFR 1 is a repository for final disposal of low and intermediate level radioactive waste produced at Swedish nuclear power plants, as well as at Swedish industrial, research, and medical treatment facilities. The repository obtained operational license in March 1988. The aim of Project SFR 1 SAR-08 is to perform an updated safety analysis, according to requirements in the regulations. A major difference between this and previous safety analyses is that repository safety should be demonstrated for 100,000 years after repository closure. This should be compared with the time frame of the safety assessment in Project SAFE that was 10,000 years. Due to the extended time frame, permafrost and glaciation have to be considered in the reference evolution of Project SFR 1 SAR-08. Other rationales for the update are recent input from the authorities concerning SAFE documents and the SFR 1 repository, as well as new data concerning the SFR 1 inventory. This report describes the outcome of revisiting the qualitative FEP (Features, Events and Processes) analysis carried out within Project SAFE for the SFR 1 repository. Each and every interaction definition, as defined in SAFE, has been examined with the aim at assuring that the SAFE interaction matrices are also applicable for SAR-08. It was found that this is generally the case, but seven new interactions were defined in order to make the interaction matrices more applicable for SAR-08. The priority of all interactions assigned priority 1 and many interactions assigned priority 2 in SAFE have been carefully examined. The examination has been made in the context of the general initial and boundary conditions that should also form the basis for the SAR-08 main scenario and less probable scenarios. In 48 cases, the priority of the interaction needed upgrading, compared to in SAFE. In a majority of these cases, the upgrade is due to the extended time frame of the safety assessment, from 10,000 years in SAFE to 100,000 years in SAR

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-15

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

  15. Guide on Project Web Access of SFR R and D and Technology Monitoring System

    International Nuclear Information System (INIS)

    Lee, Dong Uk; Won, Byung Chool; Lee, Yong Bum; Kim, Young In; Hahn, Do Hee

    2008-09-01

    The SFR R and D and technology monitoring system based on the MS enterprise project management is developed for systematic effective management of 'Development of Basic Key Technologies for Gen IV SFR' project which was performed under the Mid- and Long-term Nuclear R and D Program sponsored by the Ministry of Education, Science and Technology. This system is a tool for project management based on web access. Therefore this manual is a detailed guide for Project Web Access(PWA). Section 1 describes the common guide for using of system functions such as project server 2007 client connection setting, additional outlook function setting etc. The section 2 describes the guide for system administrator. It is described the guide for project management in section 3, 4

  16. Future extension of the Swedish repository for low and intermediate level waste (SFR)

    International Nuclear Information System (INIS)

    Carlsson, Jan

    2006-01-01

    The existing Swedish repository for low and intermediate level waste (SFR) is licensed for disposal of short-lived waste originated from operation and maintenance of Swedish nuclear power plants. The repository is foreseen to be extended to accommodate short-lived waste from the future decommissioning of the Nuclear Power Plants. Long-lived waste from operation, maintenance and eventually decommissioning will be stored some years before disposal in a geological repository. This repository can be build either as a further extension of the SFR facility or as a separate repository. This paper discusses the strategy of a step-wise extended repository where the extensions are performed during operation of the existing parts of the repository. It describes the process for licensing new parts of the repository (and re-license of the existing parts). (author)

  17. Project SFR 1 SAR-08. Update of priority of FEPs from Project SAFE

    Energy Technology Data Exchange (ETDEWEB)

    Gordon, Anna (Swedish Nuclear Fuel and Waste Management Co., Stockholm (SE)); Loefgren, Martin; Lindgren, Maria (Kemakta Konsult AB, Stockholm (SE)) (eds.)

    2008-03-15

    SFR 1 is a repository for final disposal of low and intermediate level radioactive waste produced at Swedish nuclear power plants, as well as at Swedish industrial, research, and medical treatment facilities. The repository obtained operational license in March 1988. The aim of Project SFR 1 SAR-08 is to perform an updated safety analysis, according to requirements in the regulations. A major difference between this and previous safety analyses is that repository safety should be demonstrated for 100,000 years after repository closure. This should be compared with the time frame of the safety assessment in Project SAFE that was 10,000 years. Due to the extended time frame, permafrost and glaciation have to be considered in the reference evolution of Project SFR 1 SAR-08. Other rationales for the update are recent input from the authorities concerning SAFE documents and the SFR 1 repository, as well as new data concerning the SFR 1 inventory. This report describes the outcome of revisiting the qualitative FEP (Features, Events and Processes) analysis carried out within Project SAFE for the SFR 1 repository. Each and every interaction definition, as defined in SAFE, has been examined with the aim at assuring that the SAFE interaction matrices are also applicable for SAR-08. It was found that this is generally the case, but seven new interactions were defined in order to make the interaction matrices more applicable for SAR-08. The priority of all interactions assigned priority 1 and many interactions assigned priority 2 in SAFE have been carefully examined. The examination has been made in the context of the general initial and boundary conditions that should also form the basis for the SAR-08 main scenario and less probable scenarios. In 48 cases, the priority of the interaction needed upgrading, compared to in SAFE. In a majority of these cases, the upgrade is due to the extended time frame of the safety assessment, from 10,000 years in SAFE to 100,000 years in SAR

  18. Development of SFR Research and Integration Management System (S-RIMS)

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Chang, Jin Wook; Kim, Young Gyun; Kim, Yeong Il

    2011-01-01

    Up to the present, the management of research and development (R and D) for a sodium cooled fast reactor (SFR) could be individually performed on each project without an organic relationship. However, a more systemic and effective integrated management of a project is required because the research and development environment is currently changing. Thus, we developed a Research and Integration Management System for SFR (S-RIMS) based on the enterprise project management (EPM) solution. The functional goals of the S-RIMS are as follows: 1. Provide data that show the progress and status of a project 2. Manage the design process and R and D products 3. Share the consistent design data between sub-projects

  19. Operational experience from SFR - Final repository for low- and intermediate level waste in Sweden

    International Nuclear Information System (INIS)

    Skogsberg, Marie; Ingvarsson, Roger

    2006-01-01

    SFR, the Swedish Final Repository for Radioactive Waste, has been in operation since April 1988. It was designed for short lived LLW/ILW from the operation and maintenance of all Swedish Nuclear Power Plants. The first stage was constructed for 63 000 m 3 which was assumed to give a margin and flexibility for the preliminary operational period. Today this volume represents the whole prediction of operational waste. Until the end of 2005 SFR has received 30 930 m 3 waste. In average it has been 2-3 derivations per year at the repository. The most derivations happened in the years 1993-1995, and that was also the years when the repository received the most volume of waste. The most of the derivations those years was related to the waste packages. The dose rate to the personal has always been very low in the latest years the collective dose has been under 0,1 mmanSv/year. (author)

  20. Guide on Project Web Access of SFR R and D and Technology Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Uk; Won, Byung Chool; Lee, Yong Bum; Kim, Young In; Hahn, Do Hee

    2008-09-15

    The SFR R and D and technology monitoring system based on the MS enterprise project management is developed for systematic effective management of 'Development of Basic Key Technologies for Gen IV SFR' project which was performed under the Mid- and Long-term Nuclear R and D Program sponsored by the Ministry of Education, Science and Technology. This system is a tool for project management based on web access. Therefore this manual is a detailed guide for Project Web Access(PWA). Section 1 describes the common guide for using of system functions such as project server 2007 client connection setting, additional outlook function setting etc. The section 2 describes the guide for system administrator. It is described the guide for project management in section 3, 4.

  1. Project SFR 1 SAR-08. Update of priority of FEPs from Project SAFE

    Energy Technology Data Exchange (ETDEWEB)

    Gordon, Anna [Swedish Nuclear Fuel and Waste Management Co., Stockholm (SE); Loefgren, Martin; Lindgren, Maria [Kemakta Konsult AB, Stockholm (SE); eds.

    2008-03-15

    SFR 1 is a repository for final disposal of low and intermediate level radioactive waste produced at Swedish nuclear power plants, as well as at Swedish industrial, research, and medical treatment facilities. The repository obtained operational license in March 1988. The aim of Project SFR 1 SAR-08 is to perform an updated safety analysis, according to requirements in the regulations. A major difference between this and previous safety analyses is that repository safety should be demonstrated for 100,000 years after repository closure. This should be compared with the time frame of the safety assessment in Project SAFE that was 10,000 years. Due to the extended time frame, permafrost and glaciation have to be considered in the reference evolution of Project SFR 1 SAR-08. Other rationales for the update are recent input from the authorities concerning SAFE documents and the SFR 1 repository, as well as new data concerning the SFR 1 inventory. This report describes the outcome of revisiting the qualitative FEP (Features, Events and Processes) analysis carried out within Project SAFE for the SFR 1 repository. Each and every interaction definition, as defined in SAFE, has been examined with the aim at assuring that the SAFE interaction matrices are also applicable for SAR-08. It was found that this is generally the case, but seven new interactions were defined in order to make the interaction matrices more applicable for SAR-08. The priority of all interactions assigned priority 1 and many interactions assigned priority 2 in SAFE have been carefully examined. The examination has been made in the context of the general initial and boundary conditions that should also form the basis for the SAR-08 main scenario and less probable scenarios. In 48 cases, the priority of the interaction needed upgrading, compared to in SAFE. In a majority of these cases, the upgrade is due to the extended time frame of the safety assessment, from 10,000 years in SAFE to 100,000 years in SAR

  2. A carbon budget for the aquatic ecosystem above SFR in Oeregrundsgrepen

    International Nuclear Information System (INIS)

    Kumblad, L

    1999-07-01

    The potential hazards of radionuclide release to humans and the environment is regularly evaluated in safety assessments of SFR, the final repository for radioactive operational waste. SFR handles, since 1988, low and intermediate level nuclear waste from Swedish nuclear power plants, medical care attendance, industries and research laboratories and is located in the bedrock 50 meters under the seabed of Oeregrundsgrepen in the southern Bothnian Sea. This report presents a description of the aquatic ecosystem and a carbon budget for the area above SFR with the aim to include ecosystem dynamics in the present safety assessment of the repository (SAFE). The carbon budget will support SAFE by facilitating evaluations of transport and fate of radionuclides, primarily 14 C, in case of a release from the repository and describe the ecosystem structure and function. Furthermore, 14 C is the dose-dominant radionuclide in the repository which most likely will follow the general carbon flow in the ecosystem if there should be a release. The carbon budget was based on biomass and flow of carbon between thirteen functional groups (including POC and DOC) in the ecosystem above SFR and the results indicates that the organisms are self-sufficient on carbon and that the area exports carbon corresponding to approximately 50% of the annual primary production. The largest organic carbon pool is DOC (one and a half time larger than the total biomass) and the major functional organism groups are the macrophytes (37% of the total biomass), benthic macrofauna (36%), and the microphytes (11%). The soft bottom and phytobenthic communities seem to have important roles in the ecosystem since these communities comprise the main part of the living carbon in the studied area

  3. Security-by-design approach of the KALIMER 600 SFR plant

    International Nuclear Information System (INIS)

    So, Dong Sup; Lee, Yong Bum

    2012-01-01

    Security measures as well as safety and safeguards measures should be incorporated and addressed early in the design process to enhance the cost effectiveness of a PPS (Physical Protection System). Safety, security, operations, and safeguards design teams and regulators need to be flexible and perform 'trade studies' on the available options. In this paper, SBD (Security by Design) measures in the design phase of the KALIMER 600 SFR (Sodium Cooled Reactor) plant are identified and discussed qualitatively

  4. A carbon budget for the aquatic ecosystem above SFR in Oeregrundsgrepen

    Energy Technology Data Exchange (ETDEWEB)

    Kumblad, L [Stockholm Univ. (Sweden). Dept. of Systems Ecology

    1999-07-01

    The potential hazards of radionuclide release to humans and the environment is regularly evaluated in safety assessments of SFR, the final repository for radioactive operational waste. SFR handles, since 1988, low and intermediate level nuclear waste from Swedish nuclear power plants, medical care attendance, industries and research laboratories and is located in the bedrock 50 meters under the seabed of Oeregrundsgrepen in the southern Bothnian Sea. This report presents a description of the aquatic ecosystem and a carbon budget for the area above SFR with the aim to include ecosystem dynamics in the present safety assessment of the repository (SAFE). The carbon budget will support SAFE by facilitating evaluations of transport and fate of radionuclides, primarily {sup 14}C, in case of a release from the repository and describe the ecosystem structure and function. Furthermore, {sup 14}C is the dose-dominant radionuclide in the repository which most likely will follow the general carbon flow in the ecosystem if there should be a release. The carbon budget was based on biomass and flow of carbon between thirteen functional groups (including POC and DOC) in the ecosystem above SFR and the results indicates that the organisms are self-sufficient on carbon and that the area exports carbon corresponding to approximately 50% of the annual primary production. The largest organic carbon pool is DOC (one and a half time larger than the total biomass) and the major functional organism groups are the macrophytes (37% of the total biomass), benthic macrofauna (36%), and the microphytes (11%). The soft bottom and phytobenthic communities seem to have important roles in the ecosystem since these communities comprise the main part of the living carbon in the studied area.

  5. Low and intermediate level waste in SFR-1. Reference waste inventory

    International Nuclear Information System (INIS)

    Riggare, P.; Johansson, Claes

    2001-06-01

    The objective with this report is to describe all the waste and the waste package that is expected to be deposited in SFR-1 at the time of closure. This report is a part of the SAFE project (Safety Assessment of Final Repository for Radioactive Operational Waste), i.e. the renewed safety assessment of SFR-1. The accounted waste inventory has been used as input to the release calculation that has been performed in the SAFE project. The waste inventory is based on an estimated operational lifetime of the Swedish nuclear power plants of 40 years and that closure of the SFR repository will happen in 2030. In the report, data about geometries, weights, materials, chemicals and radionuclide are given. No chemo toxic material has been identified in the waste. The inventory is based on so called waste types and the waste types reference waste package. The reference waste package combined with a prognosis of the number of waste packages to the year 2030 gives the final waste inventory for SFR-1. All reference waste packages are thoroughly described in the appendices of this report. The reference waste packages are as far as possible based on actual experiences and measurements. The radionuclide inventory is also based on actual measurements. The inventory is based on measurements of 60 Co and 137 Cs in waste packages and on measurements 239 Pu and 240 Pu in reactor water. Other nuclides in the inventory are calculated with correlation factors. In the SAFE project's prerequisites it was said that one realistic and one conservative (pessimistic) inventory should be produced. The conservative one should then be used for the release calculations. In this report one realistic and one conservative radionuclide inventory is presented. The conservative one adds up to 10 16 Bq. Regarding materials there is only one inventory given since it is not certain what is a conservative assumption

  6. Feasibility Study on Two-phase Thermosiphon for External Vessel Cooling Application of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Young; Song, Sub Lee; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    This study shows that ex-vessel cooling by two-phase thermosiphon is feasible for large size of SFR. The result presents that further studies to increase heat transfer on condenser-air and gap is necessary and the experiment should be conducted for the validation. Also, the heat loss through evaporator during normal operation, corrosion, consideration of organic fluid to exclude the poison of mercury should be studied. As the necessity of sodium fast reactor in order to reduce spent fuel, the development of designing sodium fast reactor becomes an issue. Even though there is PDRC and RVACS for the decay heat removal (DHR) system, each system has disadvantage of sodium fire and low performance, respectively. Therefore, to increase the safety of SFR, the new passive safety system design is needed without sodium fire and high performance, which can applied for large SFR. The DHR system using two-phase thermosiphon for external vessel cooling application is suggested in this paper. The proposed design have advantage that there is no structure in reactor vessel, which means no system modification and no sodium fire with perfect isolation. Also, it provide the method to mitigate sodium fire in case of sodium leakage from reactor vessel.

  7. Measurement and flow visualization research of thermal hydraulic characteristics for the SFR reactor Vessel

    International Nuclear Information System (INIS)

    Cha, J. E.; Kim, S. O.; Choi, H. L.; Kim, H. B.; Kim, H. W.; Lee, S. H.

    2012-01-01

    In this report, the thermal hydraulic and flow visualization experiment was described for the KALIMER-600 water-scaled model. In order to investigate a thermal hydraulic characteristics for the SFR KALIMER-600, which has been conceptually designed in the KAERI, a water-scaled 1/10 reactor vessel model was designed and prepared through the scaling analysis during three-years research. In this research, SFR Photos system, which has inherently very complicated the internal structures, was fabricated with a transparent vessel. It was shown that a serious of thermal hydraulic test was conducted within a short period if modeled with water than sodium. Natural circulation test was successfully performed with the modeled heater assembly and heat exchanger system coupled with cooling system. The water-scaled RSV experimental facility made in this research could be used to study the USA development for the future SFR system and utilized to analyze the flow characteristics before changing a main internal part of Photos system. It could also be used to test a pool-inspection study and a sensor selection study before large scale sodium experiment. The PCV system prepared in this research could be utilized to test other TSH experiment and temperature field measurement

  8. Can Technology Improve Large Class Learning? The Case of an Upper-Division Business Core Class

    Science.gov (United States)

    Stanley, Denise

    2013-01-01

    Larger classes are often associated with lower student achievement. The author tested the hypothesis that the introduction of personal response systems significantly improves scores in a 250-seat classroom, through the channels of improved attendance and engagement. She focused on how continuous participation with the technology could change…

  9. The Mediator subunit SFR6/MED16 controls defence gene expression mediated by salicylic acid and jasmonate responsive pathways.

    Science.gov (United States)

    Wathugala, Deepthi L; Hemsley, Piers A; Moffat, Caroline S; Cremelie, Pieter; Knight, Marc R; Knight, Heather

    2012-07-01

    • Arabidopsis SENSITIVE TO FREEZING6 (SFR6) controls cold- and drought-inducible gene expression and freezing- and osmotic-stress tolerance. Its identification as a component of the MEDIATOR transcriptional co-activator complex led us to address its involvement in other transcriptional responses. • Gene expression responses to Pseudomonas syringae, ultraviolet-C (UV-C) irradiation, salicylic acid (SA) and jasmonic acid (JA) were investigated in three sfr6 mutant alleles by quantitative real-time PCR and susceptibility to UV-C irradiation and Pseudomonas infection were assessed. • sfr6 mutants were more susceptible to both Pseudomonas syringae infection and UV-C irradiation. They exhibited correspondingly weaker PR (pathogenesis-related) gene expression than wild-type Arabidopsis following these treatments or after direct application of SA, involved in response to both UV-C and Pseudomonas infection. Other genes, however, were induced normally in the mutants by these treatments. sfr6 mutants were severely defective in expression of plant defensin genes in response to JA; ectopic expression of defensin genes was provoked in wild-type but not sfr6 by overexpression of ERF5. • SFR6/MED16 controls both SA- and JA-mediated defence gene expression and is necessary for tolerance of Pseudomonas syringae infection and UV-C irradiation. It is not, however, a universal regulator of stress gene transcription and is likely to mediate transcriptional activation of specific regulons only. © 2012 The Authors. New Phytologist © 2012 New Phytologist Trust.

  10. Core-shell nanoparticles optical sensors - Rational design of zinc ions fluorescent nanoprobes of improved analytical performance

    Science.gov (United States)

    Woźnica, Emilia; Gasik, Joanna; Kłucińska, Katarzyna; Kisiel, Anna; Maksymiuk, Krzysztof; Michalska, Agata

    2017-10-01

    In this work the effect of affinity of an analyte to a receptor on the response of nanostructural fluorimetric probes is discussed. Core-shell nanoparticles sensors are prepared that benefit from the properties of the phases involved leading to improved analytical performance. The optical transduction system chosen is independent of pH, thus the change of sample pH can be used to control the analyte - receptor affinity through the "conditional" binding constant prevailing within the lipophilic phase. It is shown that by affecting the "conditional" binding constant the performance of the sensor can be fine-tuned. As expected, increase in "conditional" affinity of the ligand embedded in the lipophilic phase to the analyte results in higher sensitivity over narrow concentration range - bulk reaction and sigmoidal shape response of emission intensity vs. logarithm of concentration changes. To induce a linear dependence of emission intensity vs. logarithm of analyte concentration covering a broad concentration range, a spatial confinement of the reaction zone is proposed, and application of core-shell nanostructures. The core material, polypyrrole nanospheres, is effectively not permeable for the analyte - ligand complex, thus the reaction is limited to the outer shell layer of the polymer prepared from poly(maleic anhydride-alt-1-octadecene). For herein introduced system a linear dependence of emission intensity vs. logarithm of Zn2+ concentration was obtained within the range from 10-7 to 10-1 M.

  11. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    Hong, S. G.; Kim, Y. H.; Venneri, F.

    2010-01-01

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  12. The SFR-M∗ main sequence archetypal star-formation history and analytical models

    Science.gov (United States)

    Ciesla, L.; Elbaz, D.; Fensch, J.

    2017-12-01

    The star-formation history (SFH) of galaxies is a key assumption to derive their physical properties and can lead to strong biases. In this work, we derive the SFH of main sequence (MS) galaxies and show how the peak SFH of a galaxy depends on its seed mass at, for example, z = 5. This seed mass reflects the galaxy's underlying dark matter (DM) halo environment. We show that, following the MS, galaxies undergo a drastic slow down of their stellar mass growth after reaching the peak of their SFH. According to abundance matching, these masses correspond to hot and massive DM halos which state could result in less efficient gas inflows on the galaxies and thus could be the origin of limited stellar mass growth. As a result, we show that galaxies, still on the MS, can enter the passive region of the UVJ diagram while still forming stars. The best fit to the MS SFH is provided by a right skew peak function for which we provide parameters depending on the seed mass of the galaxy. The ability of the classical analytical SFHs to retrieve the star-formation rate (SFR) of galaxies from spectral energy distribution (SED) fitting is studied. Due to mathematical limitations, the exponentially declining and delayed SFH struggle to model high SFR, which starts to be problematic at z > 2. The exponentially rising and log-normal SFHs exhibit the opposite behavior with the ability to reach very high SFR, and thus model starburst galaxies, but they are not able to model low values such as those expected at low redshift for massive galaxies. By simulating galaxies SED from the MS SFH, we show that these four analytical forms recover the SFR of MS galaxies with an error dependent on the model and the redshift. They are, however, sensitive enough to probe small variations of SFR within the MS, with an error ranging from 5 to 40% depending on the SFH assumption and redshift; but all the four fail to recover the SFR of rapidly quenched galaxies. However, these SFHs lead to an artificial

  13. The development and application of bioinformatics core competencies to improve bioinformatics training and education.

    Science.gov (United States)

    Mulder, Nicola; Schwartz, Russell; Brazas, Michelle D; Brooksbank, Cath; Gaeta, Bruno; Morgan, Sarah L; Pauley, Mark A; Rosenwald, Anne; Rustici, Gabriella; Sierk, Michael; Warnow, Tandy; Welch, Lonnie

    2018-02-01

    Bioinformatics is recognized as part of the essential knowledge base of numerous career paths in biomedical research and healthcare. However, there is little agreement in the field over what that knowledge entails or how best to provide it. These disagreements are compounded by the wide range of populations in need of bioinformatics training, with divergent prior backgrounds and intended application areas. The Curriculum Task Force of the International Society of Computational Biology (ISCB) Education Committee has sought to provide a framework for training needs and curricula in terms of a set of bioinformatics core competencies that cut across many user personas and training programs. The initial competencies developed based on surveys of employers and training programs have since been refined through a multiyear process of community engagement. This report describes the current status of the competencies and presents a series of use cases illustrating how they are being applied in diverse training contexts. These use cases are intended to demonstrate how others can make use of the competencies and engage in the process of their continuing refinement and application. The report concludes with a consideration of remaining challenges and future plans.

  14. The development and application of bioinformatics core competencies to improve bioinformatics training and education

    Science.gov (United States)

    Brooksbank, Cath; Morgan, Sarah L.; Rosenwald, Anne; Warnow, Tandy; Welch, Lonnie

    2018-01-01

    Bioinformatics is recognized as part of the essential knowledge base of numerous career paths in biomedical research and healthcare. However, there is little agreement in the field over what that knowledge entails or how best to provide it. These disagreements are compounded by the wide range of populations in need of bioinformatics training, with divergent prior backgrounds and intended application areas. The Curriculum Task Force of the International Society of Computational Biology (ISCB) Education Committee has sought to provide a framework for training needs and curricula in terms of a set of bioinformatics core competencies that cut across many user personas and training programs. The initial competencies developed based on surveys of employers and training programs have since been refined through a multiyear process of community engagement. This report describes the current status of the competencies and presents a series of use cases illustrating how they are being applied in diverse training contexts. These use cases are intended to demonstrate how others can make use of the competencies and engage in the process of their continuing refinement and application. The report concludes with a consideration of remaining challenges and future plans. PMID:29390004

  15. Improvement of molten core-concrete interaction model of the debris spreading analysis model in the SAMPSON code - 15193

    International Nuclear Information System (INIS)

    Hidaka, M.; Fujii, T.; Sakai, T.

    2015-01-01

    A debris spreading analysis (DSA) module has been developed and improved. The module is used in the severe accident analysis code SAMPSON and it has models for 3-dimensional natural convection with simultaneous spreading, melting and solidification. The existing analysis method of the quasi-3D boundary transportation to simulate downward concrete erosion for evaluation of molten-core concrete interaction (MCCI) was improved to full-3D to solve, for instance, debris lateral erosion under concrete floors at the bottom of the sump pit. In the advanced MCCI model, buffer cells were defined in order to solve numerical problems in case of trammel formation. Mass, momentum, and the advection term of energy between the debris melt cells and the buffer cells are solved. On the other hand, only the heat transfer and thermal conduction are solved between the debris melt cells and the structure cells, and the crust cells and the structure cells. As a preliminary analysis, a validation calculation was performed for erosion that occurred in the core-concrete interaction (CCI-2) test in the OECD/MCCI program. Comparison between the calculation and the CCI-2 test results showed the analysis has the ability to simulate debris lateral erosion under concrete floors. (authors)

  16. Synthesis and characterization of mesoporous ZSM-5 core-shell particles for improved catalytic properties

    DEFF Research Database (Denmark)

    Kustova, Marina; Holm, Martin Spangsberg; Christensen, Claus H.

    2008-01-01

    samples were tested in the MTG reaction, and the results showed that both the shell-coated and the desilicated zeolites are significantly more resistant to coke formation. These results are ascribed to the effect of the removal of structural defects rather than to an improvement of the diffusion......HZSM-5 is a unique catalyst for the conversion of methanol, dimethyl ether and other oxygenates into gasoline. During this process, catalyst deactivation by coking requires frequent regeneration and the improvement of catalyst life time is one of the challenges in catalyst development...

  17. Mapping of reed in shallow bays. SFR-Site Forsmark

    Energy Technology Data Exchange (ETDEWEB)

    Stroemgren, Maarten; Lindgren, Fredrik (Umeaa Univ. (Sweden))

    2011-03-15

    The regolith-lake development model (RLDM) describes the development of shallow bays to lakes and the infilling of lakes in the Forsmark area during an interglacial. The sensitivity analysis has shown the need for an update of the infill procedure in the RLDM. Data from the mapping of reed in shallow bays in the Forsmark area will be used to improve the infill procedure of an updated RLDM. The field work was performed in August 26-31, 2010. The mapping of reed was done in 124 points. In these points, coordinates and water depth were mapped using an echo sounder and a DGPS. Quaternary deposits and the thickness of soft sediments were mapped using an earth probe. Measurement points were delivered in ESRI shape format with coordinates in RT90 2.5 gon W and altitudes in the RHB70 system for storage in SKB's GIS data base

  18. Status of SFR Codes and Methods QA Implementation

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J. [Argonne National Lab. (ANL), Argonne, IL (United States); Briggs, Laural L. [Argonne National Lab. (ANL), Argonne, IL (United States); Fanning, Thomas H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-01-31

    This report details development of the SAS4A/SASSYS-1 SQA Program and describes the initial stages of Program implementation planning. The provisional Program structure, which is largely focused on the establishment of compliant SQA documentation, is outlined in detail, and Program compliance with the appropriate SQA requirements is highlighted. Additional program activities, such as improvements to testing methods and Program surveillance, are also described in this report. Given that the programmatic resources currently granted to development of the SAS4A/SASSYS-1 SQA Program framework are not sufficient to adequately address all SQA requirements (e.g. NQA-1, NUREG/BR-0167, etc.), this report also provides an overview of the gaps that remain the SQA program, and highlights recommendations on a path forward to resolution of these issues. One key finding of this effort is the identification of the need for an SQA program sustainable over multiple years within DOE annual R&D funding constraints.

  19. The Dependence of Convective Core Overshooting on Stellar Mass: Additional Binary Systems and Improved Calibration

    Science.gov (United States)

    Claret, Antonio; Torres, Guillermo

    2018-06-01

    Many current stellar evolution models assume some dependence of the strength of convective core overshooting on mass for stars more massive than 1.1–1.2 M ⊙, but the adopted shapes for that relation have remained somewhat arbitrary for lack of strong observational constraints. In previous work, we compared stellar evolution models to well-measured eclipsing binaries to show that, when overshooting is implemented as a diffusive process, the fitted free parameter f ov rises sharply up to about 2 M ⊙, and remains largely constant thereafter. Here, we analyze a new sample of eight binaries selected to be in the critical mass range below 2 M ⊙ where f ov is changing the most, nearly doubling the number of individual stars in this regime. This interval is important because the precise way in which f ov changes determines the shape of isochrones in the turnoff region of ∼1–5 Gyr clusters, and can thus affect their inferred ages. It also has a significant influence on estimates of stellar properties for exoplanet hosts, on stellar population synthesis, and on the detailed modeling of interior stellar structures, including the calculation of oscillation frequencies that are observable with asteroseismic techniques. We find that the derived f ov values for our new sample are consistent with the trend defined by our earlier determinations, and strengthen the relation. This provides an opportunity for future series of models to test the new prescription, grounded on observations, against independent observations that may constrain overshooting in a different way.

  20. Core lifter

    Energy Technology Data Exchange (ETDEWEB)

    Pavlov, N G; Edel' man, Ya A

    1981-02-15

    A core lifter is suggested which contains a housing, core-clamping elements installed in the housing depressions in the form of semirings with projections on the outer surface restricting the rotation of the semirings in the housing depressions. In order to improve the strength and reliability of the core lifter, the semirings have a variable transverse section formed from the outside by the surface of the rotation body of the inner arc of the semiring aroung the rotation axis and from the inner a cylindrical surface which is concentric to the outer arc of the semiring. The core-clamping elements made in this manner have the possibility of freely rotating in the housing depressions under their own weight and from contact with the core sample. These semirings do not have weakened sections, have sufficient strength, are inserted into the limited ring section of the housing of the core lifter without reduction in its through opening and this improve the reliability of the core lifter in operation.

  1. IMPROVEMENT OF STRATEGIC MANIPULATED FEDERAL PROPERTY THE EXAMPLE NON-CORE ASSETS OF JSC «CENTER OF NUCLEAR INDUSTRY NONCORE ASSETS» STATE CORPORATION «ROSATOM»

    Directory of Open Access Journals (Sweden)

    Ilya I. Rodin

    2015-01-01

    Full Text Available The article describes the main measures to improve the management of assets, federally-owned or private of public corporations - an inventory of the property, the recognition of non-core assets, the organization of decision-making systems, the sale of non-core assets at market value. The article provides the rationale for the creation within the large state-owned corporations specialized management companies responsible for the restructuring of non-core assets and improve management of the property. Also calculated the cost-effectiveness of the proposed measures on the example of the State Atomic Energy Corporation «Rosatom».

  2. Acoustic Leak Detection Requirements for a SFR Steam Generator Protection

    International Nuclear Information System (INIS)

    Kim, Tae-Joon; Jeong, Ji-Young; Oeh, Jae-Hyuk; Kim, Jong-Man; Kim, Byung-Ho; Yughay, Valery S.

    2008-01-01

    A large volume of fast reactor research has been executed in Russia, Japan, France, India and the United Kingdom. At present, an unique fast reactor named BN- 600 is operating in Russia. Also, the operation of research reactors such as Phenix (France), JOYO (Japan), BOR-60 (Russia) and FBTR (India) proceeds. The last project to be completed was the reactor Monju (Japan) which is now stopped. In addition activities for the development of fast reactors are being conducted in China, India, and South Korea. Fast reactors are a choice for the subsequent nuclear power generation in Korea, and their increased safety is one of the basic requirements. The basis for a tightening of the requirements on safety is the emergencies in NPPs in Russia, USA, France, Japan and other countries. These emergencies testify that the existing monitoring systems do not fully provide a well-timed detection of the distresses arising in a NPP, because of a poor sensitivity and response, thus the necessity for a better diagnostic system is obvious. In accordance with the USA GNEP initiative in Obninsk, Russia, 2007 the main efforts should be directed toward a sodium-water steam generator safety increase due to improvement of the hydrogen monitoring system and the acoustic leak detection system

  3. Development of oxide dispersion strengthened steels for FBR core application. 2. Morphology improvement by martensite transformation

    International Nuclear Information System (INIS)

    Ukai, Shigeharu; Nishida, Toshio; Yoshitake, Tunemitsu; Okuda, Takanari

    1998-01-01

    Previously manufactured oxide dispersion strengthened (ODS) ferritic steel cladding tubes had inferior internal creep rupture strength in the circumferential hoop direction. This unexpected feature of ODS cladding tubes was substantially ascribed to the needle-like grain structure aligned with the forming direction. In this study, the grain morphology was controlled by using the martensite transformation in ODS martensitic steels to produce an equi-axial grain structure. A major improvement in the strength anisotropy was successfully achieved. The most effective yttria addition was about 1 mass% in improving the strength of the ODS martensitic steels. A simple addition of titanium was particularly effective in increasing the strength level of the ODS martensitic steels to that of ODS ferritic steels. (author)

  4. Improvement of the properties of MnZn ferrite power cores through improvements on the microstructure of the compacts

    International Nuclear Information System (INIS)

    Kogias, G.; Tsakaloudi, V.; Van der Valk, P.; Zaspalis, V.

    2012-01-01

    In the present work freeze drying and wet-pressing technologies are applied and evaluated in the manufacturing process of functional ceramics such as MnZn power ferrites. In particular, the implementations of freeze drying instead of spray drying and the implementation of wet pressing instead of dry uniaxial pressing are investigated. It appeared that at high frequencies there is almost 25% power loss reduction by the implementation of freeze drying instead of spray drying. At low frequencies there is almost 23% power loss reduction by the implementation of wet pressing instead of dry pressing. By introducing wet pressing technology, MnZn ferrite materials exhibiting power losses of 210 mW cm -3 (100 kHz, 200 mT and 100 deg. C) could be synthesized. This is one of the lowest power loss values reported in the scientific or patented literature. - Highlights: → New drying and pressing technology concepts are applied in the manufacturing process of MnZn power ferrites. → Implementation of freeze drying instead of spray drying investigated. → Implementation of wet pressing instead of dry pressing also investigated. → Freeze drying and wet pressing improve the magnetic properties of magnetic ceramics.

  5. Development of Core Heat Removal Objective Provision Trees for Sodium-Cooled Fast Reactor Defense-in-Depth Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Huichang; Kang, Bongsuk; Lee, Youngho [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    Based on the definition of Defense-in-Depth levels and safety functions for KALIMER sodium-cooled fast reactor, suggested in the reference and, OPTs for level 1, 2, and 3 defense-in-depth and core heat removal safety function, were developed and suggested in this paper. The purpose of this OPT is first to assure the defensein-depth design during the licensing of Sodium-Cooled Fast Reactors (SFR), but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of PGSFR (Prototype Gen-IV SFR) is not mature yet, the OPT is developed for KALIMER design. Developed OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defensein-depth evaluation frame for the regulatory reviews for the licensing process. At this moment, the identified provisions have both aspects as requirements and design features already adopted in KALIMER design. In the next stage of this study, derived provisions to be adopted will be compared with the actual design features and findings can be suggested as recommendations for the safety improvement.

  6. A study on design improvement of the emergency core cooling system for a nuclear ship reactor

    International Nuclear Information System (INIS)

    Naruko, Yoshinori

    1982-01-01

    ECCS performances are predicted for the N.S. ''MUTSU'' Reactor using new computing techniques. The actual performances are found to be inadequate with regard to the total injection flow rate and the infection head of the High Pressure Injection System (HPIS). The actual Safety Injection (SI) Signal is also shown to be inoperative in a gas phase LOCA. Because of this fact, the ECCS has been improved by replacing the existing HPIS with two large high pressure pumps and by adding ''Reactor Pressure Low-Low'' and ''Containment Pressure High'' signals. This report deals with a numerical study of the dynamical behavior of the N.S. MUTSU Reactor in the postulated small break LOCA, which has been calculated to verify the design improvement by evaluating the new ECCS performances on the basis of RELAP4/MOD6/SUS and TOODEE2/SUS codes. In conclusion, the improved system performs satisfactorily in the whole range of gas phase breaks and in the small break range of liquid phase LOCA. (author)

  7. Evolution of near-field physico-chemical characteristics of the SFR repository

    International Nuclear Information System (INIS)

    Savage, D.; Stenhouse, M.; Benbow, S.

    2000-08-01

    The evaluation of the post-closure performance of the SFR repository needs to consider time dependent evolution of the repository environment. Time-dependent reaction of near-field barriers (cement, steel, bentonite) with saturating groundwater will lead to the development of hyper alkaline repository pore fluids, chemically reducing conditions, and ultimately, the generation of gas through anaerobic corrosion of metals. Cement and concrete will act as chemical conditioning agents to minimise metal corrosion and ultimately, maximise radioelement sorption. The chemical and physical evolution of cement and concrete through reaction with ambient groundwater will thus affect sorption processes through changes in pH, complexing ligands, and solid surface properties. It is desirable that these changes be incorporated into the safety assessment. The sorption behaviour of radionuclides in cementitious systems has been reviewed in detail. The available evidence from experimental work carried out on the influence of organic materials on the sorption behaviour of radionuclides, indicates that most organic degradation products will not affect sorption significantly at the concentrations expected in a cementitious repository. The notable exception to this conclusion involves the degradation products of cellulose and, in particular, polycarboxylic acids represented by iso-saccharinic acid (ISA). Results using ISA indicate a significant reduction in sorption of Pu, by several orders of magnitude, for an ISA concentration of about 10 -3 M. More recent data indicate that the negative effect is not as great, though still significant. Therefore, some scoping calculations are advisable to determine how realistic an ISA concentration of about 10 -3 M would be for the SFR repository and to estimate concentrations of other relevant organic compounds, in particular EDTA, for comparison. Scoping calculations relevant to the longevity of hyper alkaline pore fluid conditions at SFR have been

  8. Evolution of near-field physico-chemical characteristics of the SFR repository

    Energy Technology Data Exchange (ETDEWEB)

    Savage, D [Quintessa Ltd., Nottingham (United Kingdom); Stenhouse, M [Monitor Scientific LLC, Denver, CO (United States); Benbow, S [Quintessa Ltd., Henley-on-Thames (United Kingdom)

    2000-08-01

    The evaluation of the post-closure performance of the SFR repository needs to consider time dependent evolution of the repository environment. Time-dependent reaction of near-field barriers (cement, steel, bentonite) with saturating groundwater will lead to the development of hyper alkaline repository pore fluids, chemically reducing conditions, and ultimately, the generation of gas through anaerobic corrosion of metals. Cement and concrete will act as chemical conditioning agents to minimise metal corrosion and ultimately, maximise radioelement sorption. The chemical and physical evolution of cement and concrete through reaction with ambient groundwater will thus affect sorption processes through changes in pH, complexing ligands, and solid surface properties. It is desirable that these changes be incorporated into the safety assessment. The sorption behaviour of radionuclides in cementitious systems has been reviewed in detail. The available evidence from experimental work carried out on the influence of organic materials on the sorption behaviour of radionuclides, indicates that most organic degradation products will not affect sorption significantly at the concentrations expected in a cementitious repository. The notable exception to this conclusion involves the degradation products of cellulose and, in particular, polycarboxylic acids represented by iso-saccharinic acid (ISA). Results using ISA indicate a significant reduction in sorption of Pu, by several orders of magnitude, for an ISA concentration of about 10{sup -3} M. More recent data indicate that the negative effect is not as great, though still significant. Therefore, some scoping calculations are advisable to determine how realistic an ISA concentration of about 10{sup -3} M would be for the SFR repository and to estimate concentrations of other relevant organic compounds, in particular EDTA, for comparison. Scoping calculations relevant to the longevity of hyper alkaline pore fluid conditions at SFR

  9. IASCC susceptibility and IT'S improvement of austenitic stainless steels for core internals of PWR

    International Nuclear Information System (INIS)

    Yonezawa, T.; Arioka, K.; Kanasaki, S.; Fujimoto, K.; Otsuka, E.; Urata, S.; Mizuta, H.

    1998-01-01

    In this study, the cause of inter-granular cracking in BFBs was determined from the viewpoint of the metallurgical field based upon the experimental data and information published by EdF and the direction of the development on the IASCC resistance improvement of BFB materials are investigated. From the post-irradiation examinations and basic out of pile tests for 316CW and 347 stainless, the inter-granular cracking in BFBs seems to be caused by the PWSCC due to change in the grain boundary chemical composition to low chromium, high nickel and silicon under irradiation (IASCC), but not caused by SCC due to residual oxygen in the bolt stagnant region, alkaline SCC due to the dry and wet phenomenon or now pH SCC due to oxygen concentration cell. From the PWSCC test using simulated materials of grain boundary chemical composition after irradiation, increasing of chromium, purifying and precipitating the coherent grain boundary carbides must be very effective at improving the IASCC resistance. (authors)

  10. Improved Solar-Driven Photocatalytic Performance of Highly Crystalline Hydrogenated TiO2 Nanofibers with Core-Shell Structure

    Science.gov (United States)

    Wu, Ming-Chung; Chen, Ching-Hsiang; Huang, Wei-Kang; Hsiao, Kai-Chi; Lin, Ting-Han; Chan, Shun-Hsiang; Wu, Po-Yeh; Lu, Chun-Fu; Chang, Yin-Hsuan; Lin, Tz-Feng; Hsu, Kai-Hsiang; Hsu, Jen-Fu; Lee, Kun-Mu; Shyue, Jing-Jong; Kordás, Krisztián; Su, Wei-Fang

    2017-01-01

    Hydrogenated titanium dioxide has attracted intensive research interests in pollutant removal applications due to its high photocatalytic activity. Herein, we demonstrate hydrogenated TiO2 nanofibers (H:TiO2 NFs) with a core-shell structure prepared by the hydrothermal synthesis and subsequent heat treatment in hydrogen flow. H:TiO2 NFs has excellent solar light absorption and photogenerated charge formation behavior as confirmed by optical absorbance, photo-Kelvin force probe microscopy and photoinduced charge carrier dynamics analyses. Photodegradation of various organic dyes such as methyl orange, rhodamine 6G and brilliant green is shown to take place with significantly higher rates on our novel catalyst than on pristine TiO2 nanofibers and commercial nanoparticle based photocatalytic materials, which is attributed to surface defects (oxygen vacancy and Ti3+ interstitial defect) on the hydrogen treated surface. We propose three properties/mechanisms responsible for the enhanced photocatalytic activity, which are: (1) improved absorbance allowing for increased exciton generation, (2) highly crystalline anatase TiO2 that promotes fast charge transport rate, and (3) decreased charge recombination caused by the nanoscopic Schottky junctions at the interface of pristine core and hydrogenated shell thus promoting long-life surface charges. The developed H:TiO2 NFs can be helpful for future high performance photocatalysts in environmental applications.

  11. Experimental validation of thermal design of top shield for a pool type SFR

    International Nuclear Information System (INIS)

    Aithal, Sriramachandra; Babu, V. Rajan; Balasubramaniyan, V.; Velusamy, K.; Chellapandi, P.

    2016-01-01

    Highlights: • Overall thermal design of top shield in a SFR is experimentally verified. • Air jet cooling is effective in ensuring the temperatures limits for top shield. • Convection patterns in narrow annulus are in line with published CFD results. • Wire mesh insulation ensures gradual thermal gradient at top portion of main vessel. • Under loss of cooling scenario, sufficient time is available for corrective action. - Abstract: An Integrated Top Shield Test Facility towards validation of thermal design of top shield for a pool type SFR has been conceived, constructed & commissioned. Detailed experiments were performed in this experimental facility having full-scale features. Steady state temperature distribution within the facility is measured for various heater plate temperatures in addition to simulating different operating states of the reactor. Following are the important observations (i) jet cooling system is effective in regulating the roof slab bottom plate temperature and thermal gradient across roof slab simulating normal operation of reactor, (ii) wire mesh insulation provided in roof slab-main vessel annulus is effective in obtaining gradual thermal gradient along main vessel top portion and inhibiting the setting up of cellular convection within annulus and (iii) cellular convection with four distinct convective cells sets in the annular gap between roof slab and small rotatable plug measuring ∼ϕ4 m in diameter & gap width varying from 16 mm to 30 mm. Repeatability of results is also ensured during all the above tests. The results presented in this paper is expected to provide reference data for validation of thermal hydraulic models in addition to serving as design validation of jet cooling system for pool type SFR.

  12. Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor (SFR)

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock

    2013-01-01

    To achieve manufacturing technology of the fuel cladding tube in order to keep pace with the predetermined schedule in developing SFR fuel, KAERI has launched in developing fuel cladding tube in cooperation with a domestic steelmaking company. After fabricating medium-sized 1.1 ton HT9 ingot, followed by the multiple processes of hot and cold working, preliminary samples of HT9 seamless cladding tube having 7.4mm in outer diameter, 0.56mm in thickness, and 3m in length were fabricated. The objective of this study is to summarize the brief development status of the HT9 cladding tubes. Mechanical properties like axial tension, biaxial burst, pressurized creep and sodium compatibility of the cladding tubes were carried out to set up the performance evaluation technology to test the prototype FMS cladding tube which is going to be manufactured in next stage. As a part of developing fuel cladding for the Sodium-cooled Fast Reactor (SFR), preliminary HT9 cladding tube was fabricated in cooperation with a domestic steelmaking company. Microstructure as well as mechanical tests like axial tensile test, biaxial burst test, and pressurized creep test of the fuel cladding were carried out. Performance of the domestic HT9 tube was revealed to be similar in the previously fabricated foreign HT9 tube. Further prototype FMS cladding tube is going to be manufactured in next year based on this experience. Various test items like mechanical test, sodium compatibility test, microstructural analysis, basic property, cladding performance under transient situation, and performance under ion and neutron irradiation are going be performed in the future to set up the relevant technology for the licensing of the SFR cladding tube

  13. Fabrication of U-10wt.%Zr Fuel slug for SFR by Injection Casting

    International Nuclear Information System (INIS)

    Kim, Jong Hwan; Song, Hoon; Kim, Hyung Tae; Ko, Young Mo; Kim, Ki Hwan; Lee, Chan B.

    2013-01-01

    The fabrication technology of metal fuel has been developed by various methods such as rolling, swaging, wire drawing, and co-extrusion, but each of these methods had process limitations requiring an additional subsequent process, and needing the fabrication equipment is complex, which is not favorable for remote use. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, in the early 1950s, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II (EBR-II) driver and the Fast Flux Test Facility (FFTF) fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, vacuum injection casting suitable for remote operation has been developed to fabricate metallic fuel for an SFR. Vacuum injection casting technique was developed to fabricate metallic fuel for an SFR. The appearance of the fabricated U-10wt.%Zr fuel was generally sound and the internal integrity was found to be satisfactory through gamma-ray radiography. Minimum fuel losses after casting relative to the initial charge amount of U-10wt.%Zr fuel slugs met the proposed goal of less than 0.1% fuel losses during fabrication. Modifications of the current facility system and advanced casting techniques are underway to produce higher quality fuel slugs

  14. Fabrication of U-10wt.%Zr Fuel slug for SFR by Injection Casting

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Hwan; Song, Hoon; Kim, Hyung Tae; Ko, Young Mo; Kim, Ki Hwan; Lee, Chan B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The fabrication technology of metal fuel has been developed by various methods such as rolling, swaging, wire drawing, and co-extrusion, but each of these methods had process limitations requiring an additional subsequent process, and needing the fabrication equipment is complex, which is not favorable for remote use. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, in the early 1950s, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II (EBR-II) driver and the Fast Flux Test Facility (FFTF) fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, vacuum injection casting suitable for remote operation has been developed to fabricate metallic fuel for an SFR. Vacuum injection casting technique was developed to fabricate metallic fuel for an SFR. The appearance of the fabricated U-10wt.%Zr fuel was generally sound and the internal integrity was found to be satisfactory through gamma-ray radiography. Minimum fuel losses after casting relative to the initial charge amount of U-10wt.%Zr fuel slugs met the proposed goal of less than 0.1% fuel losses during fabrication. Modifications of the current facility system and advanced casting techniques are underway to produce higher quality fuel slugs.

  15. Short-term variations in core surface flow resolved from an improved method of calculating observatory monthly means

    Science.gov (United States)

    Olsen, Nils; Whaler, Kathryn A.; Finlay, Christopher C.

    2014-05-01

    Monthly means of the magnetic field measurements taken by ground observatories are a useful data source for studying temporal changes of the core magnetic field and the underlying core flow. However, the usual way of calculating monthly means as the arithmetic mean of all days (geomagnetic quiet as well as disturbed) and all local times (day and night) may result in contributions from external (magnetospheric and ionospheric) origin in the (ordinary, omm) monthly means. Such contamination makes monthly means less favourable for core studies. We calculated revised monthly means (rmm), and their uncertainties, from observatory hourly means using robust means and after removal of external field predictions, using an improved method for characterising the magnetospheric ring current. The utility of the new method for calculating observatory monthly means is demonstrated by inverting their first differences for core surface advective flows. The flow is assumed steady over three consecutive months to ensure uniqueness; the effects of more rapid changes should be attenuated by the weakly conducting mantle. Observatory data are inverted directly for a regularised core flow, rather than deriving it from a secular variation spherical harmonic model. The main field is specified by the CHAOS-4 model. Data from up to 128 observatories between 1997 and 2013 were used to calculate 185 flow models from the omm and rmm, for each possible set of three consecutive months. The full 3x3 (non-diagonal) data covariance matrix was used, and two-norm (least squares) minimisation performed. We are able to fit the data to the target (weighted) misfit of 1, for both omm and rmm inversions, provided we incorporate the full data covariance matrix, and produce consistent, plausible flows. Fits are better for rmm flows. The flows exhibit noticeable changes over timescales of a few months. However, they follow rapid excursions in the omm that we suspect result from external field contamination

  16. SKB's Project SAFE for the SFR 1 Repository. A Review by Consultants to SKI

    International Nuclear Information System (INIS)

    Chapman, N.A.; Maul, P.R.; Robinson, P.C.; Savage, D.

    2002-06-01

    The SFR 1 repository used for final disposal of low- and intermediate level radioactive waste produced by the Swedish nuclear power programme, industry, medicine and research. In 1992 it was granted a full-scale operating permit following additional reporting on long-term safety aspects by SKB, including the first in-depth safety assessment in 1991. It was stipulated as part of the full-scale licence for SFR 1 that a revised safety assessment should be carried out by SKB at least every ten years during the continued operation of the facility. The first 10-year SKB re-evaluation, called 'Project SAFE', was submitted to the regulators in 2001. The review of Project SAFE presented in this report is the culmination of several years' work with SKI including: 1. The extension and application of SKI's 'systems' approach to set up a description of the SFR 1 repository using Process Influence Diagrams (PIDs). 2. Participation in the development of a flexible Performance Assessment (PA) software tool (the AMBER code) that enables time-dependent analyses to be made of system behaviour. 3. Use of the PID database to explore, from first principles, issues that are likely to be important in the safety performance of SFR 1 and thereby to identify topics to be explored by PA modelling. 4. Peer review of the main SKB Project SAFE supporting documentation to evaluate quality, completeness and the implications of the results. 5. An independent PA exercise, using the AMBER code. 6. A review of an English translation of Section 5 of SKB's Project SAFE Final Safety Report. The present report covers only items 3 to 6, and a separate report provides a more detailed description of item 5. As a result of this review, the key issues that the regulatory authorities will need to address when reviewing SKB's safety case for SFR 1 have been identified as: 1. There is no clear statement of SKB's overall safety concept for SFR 1. It is therefore difficult to judge the results of the PA against

  17. SFR Safety Consideration in Light of Fukushima Dai-ichi Accident

    International Nuclear Information System (INIS)

    Yamaguchi, Akira

    2013-01-01

    SFR Considerations: Fukushima Dai-ichi Accident: • Combined LORL and LOHS type initiated from SBO; • High pressure water-steam cooling system: – Depressurization - Not needed; – Ultimate heat sink - Robust (NC to atmosphere); – Continuous injection - Not needed (large sensible heat capacity). • Severe accident management: – RPV failure resulted in depressurization - Elevated temperature; – Heat sink to atmosphere - Freeing risk, sodium fire risk; – Mobile power supply - External resource may not be needed; – Seawater injection with fire engines - Sodium injection not needed; • Containment performance and accessibility: – Containment - Large containment volume and low pressure system; – Explosives - Sodium fire and hydrogen explosion

  18. R&D Challenges for SFR Design and Safety Analysis – Opportunities for International Cooperations

    International Nuclear Information System (INIS)

    Devictor, Nicolas

    2013-01-01

    Examples of R&D challenges related to safety have been presented. For any domain, R&D activities includes modelling, codes development and their V&V process, with the support of experimental programs. The success in the R&D will help the safety case and the acceptability of SFR. Some of these activities are relevant for international cooperation especially benchmarks and sharing of experimental facilities. This last point could take benefit of recent catalogues experimental facilities (already operational or in project), for example from the TAREF Task Force of OECD/NEA and the European project ADRIANA

  19. Improving the electrode performance of Ge through Ge@C core-shell nanoparticles and graphene networks.

    Science.gov (United States)

    Xue, Ding-Jiang; Xin, Sen; Yan, Yang; Jiang, Ke-Cheng; Yin, Ya-Xia; Guo, Yu-Guo; Wan, Li-Jun

    2012-02-08

    Germanium is a promising high-capacity anode material for lithium ion batteries, but it usually exhibits poor cycling stability because of its huge volume variation during the lithium uptake and release process. A double protection strategy to improve the electrode performance of Ge through the use of Ge@C core-shell nanostructures and reduced graphene oxide (RGO) networks has been developed. The as-synthesized Ge@C/RGO nanocomposite showed excellent cycling performance and rate capability in comparison with Ge@C nanoparticles when used as an anode material for Li ion batteries, which can be attributed to the electronically conductive and elastic RGO networks in addition to the carbon shells and small particle sizes of Ge. The strategy is simple yet very effective, and because of its versatility, it may be extended to other high-capacity electrode materials with large volume variations and low electrical conductivities.

  20. Design of cladding rods-assisted depressed-core few-mode fibers with improved modal spacing

    Science.gov (United States)

    Han, Jiawei; Zhang, Jie

    2018-03-01

    This paper investigates the design details of cladding rods-assisted (CRA) depressed-core (DC) few-mode fibers (FMFs) that feature more equally spaced linearly polarized (LP) modal effective indices, suitable for high-spatial-density weakly-coupled mode-division multiplexing systems. The influences of the index profile of cladding rods on LP mode-resolved effective index, bending sensitivity, and effective area Aeff, are numerically described. Based on the design considerations of LP modal Aeff-dependent spatial efficiency and LP modal bending loss-dependent robustness, the small LP21-LP02 and LP22-LP03 modal spacing limitations, encountered in state-of-the-art weakly-coupled step-index FMFs, have been substantially improved by at least 25%. In addition, the proposed CRA DC FMFs also show sufficiently large effective areas (in excess of 110 μm2) for all guided LP modes, which are expected to exhibit good nonlinear performance.

  1. Iterative image reconstruction algorithms in coronary CT angiography improve the detection of lipid-core plaque - a comparison with histology

    Energy Technology Data Exchange (ETDEWEB)

    Puchner, Stefan B. [Massachusetts General Hospital, Harvard Medical School, Cardiac MR PET CT Program, Department of Radiology, Boston, MA (United States); Medical University of Vienna, Department of Biomedical Imaging and Image-Guided Therapy, Vienna (Austria); Ferencik, Maros [Massachusetts General Hospital, Harvard Medical School, Cardiac MR PET CT Program, Department of Radiology, Boston, MA (United States); Harvard Medical School, Division of Cardiology, Massachusetts General Hospital, Boston, MA (United States); Maurovich-Horvat, Pal [Massachusetts General Hospital, Harvard Medical School, Cardiac MR PET CT Program, Department of Radiology, Boston, MA (United States); Semmelweis University, MTA-SE Lenduelet Cardiovascular Imaging Research Group, Heart and Vascular Center, Budapest (Hungary); Nakano, Masataka; Otsuka, Fumiyuki; Virmani, Renu [CV Path Institute Inc., Gaithersburg, MD (United States); Kauczor, Hans-Ulrich [University Hospital Heidelberg, Ruprecht-Karls-University of Heidelberg, Department of Diagnostic and Interventional Radiology, Heidelberg (Germany); Hoffmann, Udo [Massachusetts General Hospital, Harvard Medical School, Cardiac MR PET CT Program, Department of Radiology, Boston, MA (United States); Schlett, Christopher L. [Massachusetts General Hospital, Harvard Medical School, Cardiac MR PET CT Program, Department of Radiology, Boston, MA (United States); University Hospital Heidelberg, Ruprecht-Karls-University of Heidelberg, Department of Diagnostic and Interventional Radiology, Heidelberg (Germany)

    2015-01-15

    To evaluate whether iterative reconstruction algorithms improve the diagnostic accuracy of coronary CT angiography (CCTA) for detection of lipid-core plaque (LCP) compared to histology. CCTA and histological data were acquired from three ex vivo hearts. CCTA images were reconstructed using filtered back projection (FBP), adaptive-statistical (ASIR) and model-based (MBIR) iterative algorithms. Vessel cross-sections were co-registered between FBP/ASIR/MBIR and histology. Plaque area <60 HU was semiautomatically quantified in CCTA. LCP was defined by histology as fibroatheroma with a large lipid/necrotic core. Area under the curve (AUC) was derived from logistic regression analysis as a measure of diagnostic accuracy. Overall, 173 CCTA triplets (FBP/ASIR/MBIR) were co-registered with histology. LCP was present in 26 cross-sections. Average measured plaque area <60 HU was significantly larger in LCP compared to non-LCP cross-sections (mm{sup 2}: 5.78 ± 2.29 vs. 3.39 ± 1.68 FBP; 5.92 ± 1.87 vs. 3.43 ± 1.62 ASIR; 6.40 ± 1.55 vs. 3.49 ± 1.50 MBIR; all p < 0.0001). AUC for detecting LCP was 0.803/0.850/0.903 for FBP/ASIR/MBIR and was significantly higher for MBIR compared to FBP (p = 0.01). MBIR increased sensitivity for detection of LCP by CCTA. Plaque area <60 HU in CCTA was associated with LCP in histology regardless of the reconstruction algorithm. However, MBIR demonstrated higher accuracy for detecting LCP, which may improve vulnerable plaque detection by CCTA. (orig.)

  2. GALAXY STRUCTURE AND MODE OF STAR FORMATION IN THE SFR-MASS PLANE FROM z {approx} 2.5 TO z {approx} 0.1

    Energy Technology Data Exchange (ETDEWEB)

    Wuyts, Stijn; Foerster Schreiber, Natascha M.; Magnelli, Benjamin; Genzel, Reinhard; Lutz, Dieter; Berta, Stefano; Gracia-Carpio, Javier; Nordon, Raanan [Max-Planck-Institut fuer extraterrestrische Physik, Giessenbachstrasse, D-85748 Garching (Germany); Van der Wel, Arjen [Max-Planck-Institut fuer Astronomie, Koenigstuhl 17, D-69117 Heidelberg (Germany); Guo, Yicheng [Astronomy Department, University of Massachusetts, 710 N. Pleasant Street, Amherst, MA 01003 (United States); Aussel, Herve; Le Floc' h, Emeric [Laboratoire AIM, CEA/DSM-CNRS-Universite Paris Diderot, IRFU/Service d' Astrophysique, Bat. 709, CEA-Saclay, F-91191 Gif-sur-Yvette Cedex (France); Barro, Guillermo; Kocevski, Dale D.; McGrath, Elizabeth J. [UCO/Lick Observatory, Department of Astronomy and Astrophysics, University of California, Santa Cruz, CA 95064 (United States); Cava, Antonio [Departamento de Astrofisica, Facultad de CC. Fisicas, Universidad Complutense de Madrid, E-28040 Madrid (Spain); Hathi, Nimish P. [Observatories of the Carnegie Institution of Washington, Pasadena, CA 91101 (United States); Huang, Kuang-Han [Johns Hopkins University, 3400 North Charles Street, Baltimore, MD 21218 (United States); Koekemoer, Anton M. [Space Telescope Science Institute, 3700 San Martin Drive, Baltimore, MD 21218 (United States); Lee, Kyoung-Soo [Yale Center for Astronomy and Astrophysics, Department of Physics, Yale University, New Haven, CT 06520 (United States); and others

    2011-12-01

    We analyze the dependence of galaxy structure (size and Sersic index) and mode of star formation ({Sigma}{sub SFR} and SFR{sub IR}/SFR{sub UV}) on the position of galaxies in the star formation rate (SFR) versus mass diagram. Our sample comprises roughly 640,000 galaxies at z {approx} 0.1, 130,000 galaxies at z {approx} 1, and 36,000 galaxies at z {approx} 2. Structural measurements for all but the z {approx} 0.1 galaxies are based on Hubble Space Telescope imaging, and SFRs are derived using a Herschel-calibrated ladder of SFR indicators. We find that a correlation between the structure and stellar population of galaxies (i.e., a 'Hubble sequence') is already in place since at least z {approx} 2.5. At all epochs, typical star-forming galaxies on the main sequence are well approximated by exponential disks, while the profiles of quiescent galaxies are better described by de Vaucouleurs profiles. In the upper envelope of the main sequence, the relation between the SFR and Sersic index reverses, suggesting a rapid buildup of the central mass concentration in these starbursting outliers. We observe quiescent, moderately and highly star-forming systems to co-exist over an order of magnitude or more in stellar mass. At each mass and redshift, galaxies on the main sequence have the largest size. The rate of size growth correlates with specific SFR, and so does {Sigma}{sub SFR} at each redshift. A simple model using an empirically determined star formation law and metallicity scaling, in combination with an assumed geometry for dust and stars, is able to relate the observed {Sigma}{sub SFR} and SFR{sub IR}/SFR{sub UV}, provided a more patchy dust geometry is assumed for high-redshift galaxies.

  3. GALAXY STRUCTURE AND MODE OF STAR FORMATION IN THE SFR-MASS PLANE FROM z ∼ 2.5 TO z ∼ 0.1

    International Nuclear Information System (INIS)

    Wuyts, Stijn; Förster Schreiber, Natascha M.; Magnelli, Benjamin; Genzel, Reinhard; Lutz, Dieter; Berta, Stefano; Graciá-Carpio, Javier; Nordon, Raanan; Van der Wel, Arjen; Guo, Yicheng; Aussel, Hervé; Le Floc'h, Emeric; Barro, Guillermo; Kocevski, Dale D.; McGrath, Elizabeth J.; Cava, Antonio; Hathi, Nimish P.; Huang, Kuang-Han; Koekemoer, Anton M.; Lee, Kyoung-Soo

    2011-01-01

    We analyze the dependence of galaxy structure (size and Sérsic index) and mode of star formation (Σ SFR and SFR IR /SFR UV ) on the position of galaxies in the star formation rate (SFR) versus mass diagram. Our sample comprises roughly 640,000 galaxies at z ∼ 0.1, 130,000 galaxies at z ∼ 1, and 36,000 galaxies at z ∼ 2. Structural measurements for all but the z ∼ 0.1 galaxies are based on Hubble Space Telescope imaging, and SFRs are derived using a Herschel-calibrated ladder of SFR indicators. We find that a correlation between the structure and stellar population of galaxies (i.e., a 'Hubble sequence') is already in place since at least z ∼ 2.5. At all epochs, typical star-forming galaxies on the main sequence are well approximated by exponential disks, while the profiles of quiescent galaxies are better described by de Vaucouleurs profiles. In the upper envelope of the main sequence, the relation between the SFR and Sérsic index reverses, suggesting a rapid buildup of the central mass concentration in these starbursting outliers. We observe quiescent, moderately and highly star-forming systems to co-exist over an order of magnitude or more in stellar mass. At each mass and redshift, galaxies on the main sequence have the largest size. The rate of size growth correlates with specific SFR, and so does Σ SFR at each redshift. A simple model using an empirically determined star formation law and metallicity scaling, in combination with an assumed geometry for dust and stars, is able to relate the observed Σ SFR and SFR IR /SFR UV , provided a more patchy dust geometry is assumed for high-redshift galaxies.

  4. Development of electromagnetic acoustic transducer (EMAT) phased arrays for SFR inspection

    Energy Technology Data Exchange (ETDEWEB)

    Le Bourdais, Florian; Marchand, Benoît [CEA LIST, Centre de Saclay F-91191 Gif-sur-Yvette (France)

    2014-02-18

    A long-standing problem for Sodium cooled Fast Reactor (SFR) instrumentation is the development of efficient under-sodium visualization systems adapted to the hot and opaque sodium environment. Electromagnetic Acoustic Transducers (EMAT) are potential candidates for a new generation of Ultrasonic Testing (UT) probes well-suited for SFR inspection that can overcome drawbacks of classical piezoelectric probes in sodium environment. Based on the use of new CIVA simulation tools, we have designed and optimized an advanced EMAT probe for under-sodium visualization. This has led to the development of a fully functional L-wave EMAT sensing system composed of 8 elements and a casing withstanding 200° C sodium inspection. Laboratory experiments demonstrated the probe's ability to sweep an ultrasonic beam to an angle of 15 degrees. Testing in a specialized sodium facility has shown that it was possible to obtain pulse-echo signals from a target under several different angles from a fixed position.

  5. Model summary report for the safety assessment SFR 1 SAR-08

    Energy Technology Data Exchange (ETDEWEB)

    2008-03-15

    This document is the model summary report for the safety assessment SFR 1 SAR-08. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SFR1 SAR-08, a number of different computer codes are used. In order to better understand how these codes are related an Assessment Model Flowchart, AMF, has been produced within the project. From the AMF, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A number of different computer codes are used in the assessment of which some are commercial while others are developed for assessment projects. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met

  6. Model summary report for the safety assessment SFR 1 SAR-08

    International Nuclear Information System (INIS)

    2008-03-01

    This document is the model summary report for the safety assessment SFR 1 SAR-08. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SFR1 SAR-08, a number of different computer codes are used. In order to better understand how these codes are related an Assessment Model Flowchart, AMF, has been produced within the project. From the AMF, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A number of different computer codes are used in the assessment of which some are commercial while others are developed for assessment projects. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met

  7. SDSS IV MaNGA - sSFR profiles and the slow quenching of discs in green valley galaxies

    Science.gov (United States)

    Belfiore, Francesco; Maiolino, Roberto; Bundy, Kevin; Masters, Karen; Bershady, Matthew; Oyarzún, Grecco; Lin, Lihwai; Cano-Diaz, Mariana; Wake, David; Spindler, Ashley; Thomas, Daniel; Brownstein, Joel R.; Drory, Niv; Yan, Renbin

    2018-03-01

    We study radial profiles in Hα equivalent width and specific star formation rate (sSFR) derived from spatially-resolved SDSS-IV MaNGA spectroscopy to gain insight on the physical mechanisms that suppress star formation and determine a galaxy's location in the SFR-M_\\star diagram. Even within the star-forming `main sequence', the measured sSFR decreases with stellar mass, both in an integrated and spatially-resolved sense. Flat sSFR radial profiles are observed for log(M_\\star / M_⊙ ) history. Our primary focus is the green valley, constituted by galaxies lying below the star formation main sequence, but not fully passive. In the green valley we find sSFR profiles that are suppressed with respect to star-forming galaxies of the same mass at all galactocentric distances out to 2 effective radii. The responsible quenching mechanism therefore appears to affect the entire galaxy, not simply an expanding central region. The majority of green valley galaxies of log(M_\\star / M_⊙ ) > 10.0 are classified spectroscopically as central low-ionisation emission-line regions (cLIERs). Despite displaying a higher central stellar mass concentration, the sSFR suppression observed in cLIER galaxies is not simply due to the larger mass of the bulge. Drawing a comparison sample of star forming galaxies with the same M_\\star and Σ _{1 kpc} (the mass surface density within 1 kpc), we show that a high Σ _{1 kpc} is not a sufficient condition for determining central quiescence.

  8. Iterative image reconstruction algorithms in coronary CT angiography improve the detection of lipid-core plaque - a comparison with histology

    International Nuclear Information System (INIS)

    Puchner, Stefan B.; Ferencik, Maros; Maurovich-Horvat, Pal; Nakano, Masataka; Otsuka, Fumiyuki; Virmani, Renu; Kauczor, Hans-Ulrich; Hoffmann, Udo; Schlett, Christopher L.

    2015-01-01

    To evaluate whether iterative reconstruction algorithms improve the diagnostic accuracy of coronary CT angiography (CCTA) for detection of lipid-core plaque (LCP) compared to histology. CCTA and histological data were acquired from three ex vivo hearts. CCTA images were reconstructed using filtered back projection (FBP), adaptive-statistical (ASIR) and model-based (MBIR) iterative algorithms. Vessel cross-sections were co-registered between FBP/ASIR/MBIR and histology. Plaque area 2 : 5.78 ± 2.29 vs. 3.39 ± 1.68 FBP; 5.92 ± 1.87 vs. 3.43 ± 1.62 ASIR; 6.40 ± 1.55 vs. 3.49 ± 1.50 MBIR; all p < 0.0001). AUC for detecting LCP was 0.803/0.850/0.903 for FBP/ASIR/MBIR and was significantly higher for MBIR compared to FBP (p = 0.01). MBIR increased sensitivity for detection of LCP by CCTA. Plaque area <60 HU in CCTA was associated with LCP in histology regardless of the reconstruction algorithm. However, MBIR demonstrated higher accuracy for detecting LCP, which may improve vulnerable plaque detection by CCTA. (orig.)

  9. Improved Design Concept for ensuring the Passive Decay Heat Removal Performance of an SFR

    International Nuclear Information System (INIS)

    Eoh, Jae Hyuk; Lee, Tae Ho; Han, Ji Woong; Kim, Seong O

    2011-01-01

    In order to enhance the operational reliability of a purely passive decay heat removal system in KALIMER, which is named as PDRC, three design options to prevent a sodium freezing in an intermediate decay heat removal circuit were proposed, and their feasibilities was quantitatively evaluated. For all the options, more specific design considerations were made to confirm their feasibility to properly materialize their concepts in a practical system design procedure, and the general definitions for a purely passive concept and its design features have been discussed. A numerical study to evaluate the coastdown flow effect of the primary pump was performed to figure out the early stage DHR capability inside reactor pool during a loss of normal heat sink accident. The thermal-hydraulic calculations have been made by using the COMMIX-1AR/P code, and it was found that the initiation of heat removal by DHX could be accelerated by the increase of the coastdown time but it needs a large-sized flywheel. For the demonstration of the innovative concept, a large scale sodium thermal-hydraulic test facility is currently being designed. It is very difficult to reproduce both a hydrodynamic and a thermodynamic similarity to the prototype plant if the thermal driving head is determined by structure-to-fluid heat transfer under natural circulation flow. Hence the similitude requirements for the sodium thermal-hydraulic test facility employing natural convection heat transfer were developed, and the preliminary design data of the test facility by implementing proper scaling methodologies was produced. The design restrictions imposed on the test facility and the scaling distortions of the design data to the full-scale system were also discussed

  10. Long-time stability following freezing and thawing of concrete and bentonite in deposition of low- and intermediate-level radioactive waste in SFR 1

    International Nuclear Information System (INIS)

    Emborg, Mats; Jonasson, Jan-Erik; Knutsson, Sven

    2007-10-01

    This document describes the effect of freezing on the concrete and bentonite barriers in SFR 1. The document constitutes one of the references describing the degradation of barriers in a long-time perspective and is used in the safety analysis SFR 1 SAR-08

  11. Effects of Ta addition on the Microstructural and Mechanical Properties of 9Cr-0.5Mo-2W F/M Steels for a SFR Fuel Cladding

    International Nuclear Information System (INIS)

    Baek, Jong-Hyuk; Han, Chang-Hee; Kim, Tae-Kyu; Kim, Sung-Ho; Lee, Chan-Bock

    2007-01-01

    Today twenty fission reactors provide about 40% of the domestic electricity supply. The world-wide distribution of some nuclear reactors will be aging and will need replacement and enhancement to both keep pace with and to take up a large share of the growing world-wide electricity demand. A new generation (Gen IV) of nuclear plant concepts has become the focus of international advanced reactor activity. Gen IV nuclear systems embodies greater improvements and innovative advances in technology over earlier ones. The Gen IV systems are to have a considerable increase in safety and be economically competitive when compared with the existed commercial reactors. In particular, the systems should produce a significantly reduced volume of nuclear wastes. From this point of view, sodium-cooled Fast Reactor (SFR) is strongly considered as a future nuclear energy system in Korea

  12. Exploration of Important Issues for the Safety of SFR 1 using Performance Assessment Calculations

    International Nuclear Information System (INIS)

    Maul, P.R.; Robinson, P.C.

    2002-06-01

    SKB has produced a revised safety case for the SFR 1 disposal facility for low and intermediate level radioactive wastes at Forsmark: project SAFE. This assessment includes a Performance Assessment (PA) for the long term post-closure safety of the facility. SKI has a responsibility to scrutinise SKB's safety case that is shared with SSI. Quintessa has undertaken a review of SKB's case for the long term safety of SFR 1 to assist SKI's evaluation of SAFE, and this is given in SKI-R--02-61, henceforth referred to as the Quintessa Review. The current report describes the independent PA calculations that provided an input to that review. Since 1999 SKI has been developing a PA capability for SFR 1 using the AMBER software. Two key features of the approach taken have been: To represent the whole system in a single model; and To allow the time-dependency of all key features, events and processes to be represented. These capabilities allow a better understanding of the key features of the system to be obtained for different future evolutions (scenarios). This report presents a summary of the work undertaken to provide SKI with a PA capability for SFR 1 and the calculations undertaken with it. Calculations have been undertaken for radionuclides transported in groundwater and gas, but not for direct intrusion by humans into the wastes. It should be emphasised that the purpose of the Performance Assessment calculations described in this report is not to provide an alternative assessment of potential radiological impacts to that produced by SKB. The aim is to use the models that have been developed to investigate the important features of the system and to help SKI scrutinise the case put to them by SKB. The PA calculations that have been undertaken are by no means comprehensive, and various issues could be investigated further if required. The key issues that have been identified can be summarised as follows: 1. The SFR 1 system has a number of different timescales that can

  13. Thermodynamic analysis and preliminary design of closed Brayton cycle using nitrogen as working fluid and coupled to small modular Sodium-cooled fast reactor (SM-SFR)

    International Nuclear Information System (INIS)

    Olumayegun, Olumide; Wang, Meihong; Kelsall, Greg

    2017-01-01

    Highlights: • Nitrogen closed Brayton cycle for small modular sodium-cooled fast reactor studied. • Thermodynamic modelling and analysis of closed Brayton cycle performed. • Two-shaft configuration proposed and performance compared to single shaft. • Preliminary design of heat exchangers and turbomachinery carried out. - Abstract: Sodium-cooled fast reactor (SFR) is considered the most promising of the Generation IV reactors for their near-term demonstration of power generation. Small modular SFRs (SM-SFRs) have less investment risk, can be deployed more quickly, are easier to operate and are more flexible in comparison to large nuclear reactor. Currently, SFRs use the proven Rankine steam cycle as the power conversion system. However, a key challenge is to prevent dangerous sodium-water reaction that could happen in SFR coupled to steam cycle. Nitrogen gas is inert and does not react with sodium. Hence, intercooled closed Brayton cycle (CBC) using nitrogen as working fluid and with a single shaft configuration has been one common power conversion system option for possible near-term demonstration of SFR. In this work, a new two shaft nitrogen CBC with parallel turbines was proposed to further simplify the design of the turbomachinery and reduce turbomachinery size without compromising the cycle efficiency. Furthermore, thermodynamic performance analysis and preliminary design of components were carried out in comparison with a reference single shaft nitrogen cycle. Mathematical models in Matlab were developed for steady state thermodynamic analysis of the cycles and for preliminary design of the heat exchangers, turbines and compressors. Studies were performed to investigate the impact of the recuperator minimum terminal temperature difference (TTD) on the overall cycle efficiency and recuperator size. The effect of turbomachinery efficiencies on the overall cycle efficiency was examined. The results showed that the cycle efficiency of the proposed

  14. The SSI and SKI review of the updated Final Safety Report for SFR 1 issued by SKB. Review report; SSI:s och SKI:s granskning av SKB:s uppdaterade Slutlig Saekerhetsrapport foer SFR 1. Granskningsrapport

    Energy Technology Data Exchange (ETDEWEB)

    Avila, Rodolfo; Jensen, Mikael; Larsson, Carl-Magnus; Lund, Ingemar; Loefgren, Tomas; Moberg, Leif; Norden, Maria; Wiebert, Anders [Swedish Radiation Protection Authority, Stockholm (Sweden); Berglund, Thomas; Dverstorp, Bjoern; Hedberg, Bengt; Kautsky, Fritz; Lilja, Christina; Simic, Eva; Stroemberg, Bo; Sundstroem, Benny; Toverud, Oeivind; Wingefors, Stig; Zika, Helmuth [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    2003-11-01

    The repository for operational radioactive wastes in Sweden, SFR1, has been the object for a new safety assessment study by SKB (The Swedish Nuclear Fuel and Waste Management Co.). The findings of the review group will form the basis for decisions by the authorities on the provisions for the future operation of the repository.

  15. "Score the Core" Web-based pathologist training tool improves the accuracy of breast cancer IHC4 scoring.

    Science.gov (United States)

    Engelberg, Jesse A; Retallack, Hanna; Balassanian, Ronald; Dowsett, Mitchell; Zabaglo, Lila; Ram, Arishneel A; Apple, Sophia K; Bishop, John W; Borowsky, Alexander D; Carpenter, Philip M; Chen, Yunn-Yi; Datnow, Brian; Elson, Sarah; Hasteh, Farnaz; Lin, Fritz; Moatamed, Neda A; Zhang, Yanhong; Cardiff, Robert D

    2015-11-01

    Hormone receptor status is an integral component of decision-making in breast cancer management. IHC4 score is an algorithm that combines hormone receptor, HER2, and Ki-67 status to provide a semiquantitative prognostic score for breast cancer. High accuracy and low interobserver variance are important to ensure the score is accurately calculated; however, few previous efforts have been made to measure or decrease interobserver variance. We developed a Web-based training tool, called "Score the Core" (STC) using tissue microarrays to train pathologists to visually score estrogen receptor (using the 300-point H score), progesterone receptor (percent positive), and Ki-67 (percent positive). STC used a reference score calculated from a reproducible manual counting method. Pathologists in the Athena Breast Health Network and pathology residents at associated institutions completed the exercise. By using STC, pathologists improved their estrogen receptor H score and progesterone receptor and Ki-67 proportion assessment and demonstrated a good correlation between pathologist and reference scores. In addition, we collected information about pathologist performance that allowed us to compare individual pathologists and measures of agreement. Pathologists' assessment of the proportion of positive cells was closer to the reference than their assessment of the relative intensity of positive cells. Careful training and assessment should be used to ensure the accuracy of breast biomarkers. This is particularly important as breast cancer diagnostics become increasingly quantitative and reproducible. Our training tool is a novel approach for pathologist training that can serve as an important component of ongoing quality assessment and can improve the accuracy of breast cancer prognostic biomarkers. Copyright © 2015 Elsevier Inc. All rights reserved.

  16. Using a Core Vocabulary Intervention to Improve Communication of Students Who Use Augmentative and Alternative Communication (AAC)

    Science.gov (United States)

    Riccelli-Sherman, Angela

    2017-01-01

    This study measured the impact of core vocabulary selection and core vocabulary use on overall communication effectiveness and literacy. In this study, 30 kindergarten special education students, both male and female, who were enrolled in the Developmental Kindergarten program (a self-contained special education classroom) and Inclusive…

  17. Incorporating trnH-psbA to the core DNA barcodes improves significantly species discrimination within southern African Combretaceae

    Directory of Open Access Journals (Sweden)

    Jephris Gere

    2013-12-01

    Full Text Available Recent studies indicate that the discriminatory power of the core DNA barcodes (rbcLa + matK for land plants may have been overestimated since their performance have been tested only on few closely related species. In this study we focused mainly on how the addition of complementary barcodes (nrITS and trnH-psbA to the core barcodes will affect the performance of the core barcodes in discriminating closely related species from family to section levels. In general, we found that the core barcodes performed poorly compared to the various combinations tested. Using multiple criteria, we finally advocated for the use of the core + trnH-psbA as potential DNA barcode for the family Combretaceae at least in southern Africa. Our results also indicate that the success of DNA barcoding in discriminating closely related species may be related to evolutionary and possibly the biogeographic histories of the taxonomic group tested.

  18. Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, Luis E., E-mail: luisen.herranz@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Garcia, Monica, E-mail: monica.gmartin@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Morandi, Sonia, E-mail: sonia.morandi@rse-web.it [Nuclear and Industrial Plant Safety Team, Power Generation System Department, RSE, via Rubattino 54, 20134 Milano (Italy)

    2013-12-15

    Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have

  19. Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

    International Nuclear Information System (INIS)

    Herranz, Luis E.; Garcia, Monica; Morandi, Sonia

    2013-01-01

    Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have been adopted so that

  20. Design characteristics of pantograph type in vessel fuel handling system in SFR

    International Nuclear Information System (INIS)

    Kim, S. H.; Koo, G. H.

    2012-01-01

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied

  1. Design characteristics of pantograph type in vessel fuel handling system in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. H.; Koo, G. H. [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.

  2. Effect of a Background Noise on the Acoustic Leak Detection Methodology for a SFR Steam Generator

    International Nuclear Information System (INIS)

    Kim, Tae-Joon; Jeong, Ji-Young; Kim, Jong-Man

    2007-01-01

    The protection of a water/steam leak into a sodium in the SFR SG at an early phase of a leak origin depends on a fast response and sensitivity of a leak detection system not to a response against the several kinds of noises. The subject in this study is to introduce a detection performance by using our developed acoustic leak detection methodology discriminated by a backpropagation neural network according to a preprocessing of the 1/6 Octave band analysis or 1/12 Octave band analysis and the x n method defined by us. It was used for the acoustic signals generated from the simulation works which are the noises of an artificial background such as a scratching, a hammering on a steel structure and so on. In a previous study, we showed that the performance of a LabVIEW tool embedded with the developed acoustic leak detection methodology detected the SWR leak signals

  3. The evolution of the dust temperatures of galaxies in the SFR-M∗ plane up to z ∼ 2

    Science.gov (United States)

    Magnelli, B.; Lutz, D.; Saintonge, A.; Berta, S.; Santini, P.; Symeonidis, M.; Altieri, B.; Andreani, P.; Aussel, H.; Béthermin, M.; Bock, J.; Bongiovanni, A.; Cepa, J.; Cimatti, A.; Conley, A.; Daddi, E.; Elbaz, D.; Förster Schreiber, N. M.; Genzel, R.; Ivison, R. J.; Le Floc'h, E.; Magdis, G.; Maiolino, R.; Nordon, R.; Oliver, S. J.; Page, M.; Pérez García, A.; Poglitsch, A.; Popesso, P.; Pozzi, F.; Riguccini, L.; Rodighiero, G.; Rosario, D.; Roseboom, I.; Sanchez-Portal, M.; Scott, D.; Sturm, E.; Tacconi, L. J.; Valtchanov, I.; Wang, L.; Wuyts, S.

    2014-01-01

    We study the evolution of the dust temperature of galaxies in the SFR- M∗ plane up to z ~ 2 using far-infrared and submillimetre observations from the Herschel Space Observatory taken as part of the PACS Evolutionary Probe (PEP) and Herschel Multi-tiered Extragalactic Survey (HerMES) guaranteed time key programmes. Starting from a sample of galaxies with reliable star-formation rates (SFRs), stellar masses (M∗) and redshift estimates, we grid the SFR- M∗parameter space in several redshift ranges and estimate the mean dust temperature (Tdust) of each SFR-M∗ - z bin. Dust temperatures are inferred using the stacked far-infrared flux densities (100-500 μm) of our SFR-M∗ - z bins. At all redshifts, the dust temperature of galaxies smoothly increases with rest-frame infrared luminosities (LIR), specific SFRs (SSFR; i.e., SFR/M∗), and distances with respect to the main sequence (MS) of the SFR- M∗ plane (i.e., Δlog (SSFR)MS = log [SSFR(galaxy)/SSFRMS(M∗,z)]). The Tdust - SSFR and Tdust - Δlog (SSFR)MS correlations are statistically much more significant than the Tdust - LIR one. While the slopes of these three correlations are redshift-independent, their normalisations evolve smoothly from z = 0 and z ~ 2. We convert these results into a recipe to derive Tdust from SFR, M∗ and z, valid out to z ~ 2 and for the stellar mass and SFR range covered by our stacking analysis. The existence of a strong Tdust - Δlog (SSFR)MS correlation provides us with several pieces of information on the dust and gas content of galaxies. Firstly, the slope of the Tdust - Δlog (SSFR)MS correlation can be explained by the increase in the star-formation efficiency (SFE; SFR/Mgas) with Δlog (SSFR)MS as found locally by molecular gas studies. Secondly, at fixed Δlog (SSFR)MS, the constant dust temperature observed in galaxies probing wide ranges in SFR and M∗ can be explained by an increase or decrease in the number of star-forming regions with comparable SFE enclosed in

  4. Implementation of project Safe in Amber. Verification study for SFR 1 SAR-08

    Energy Technology Data Exchange (ETDEWEB)

    Thomson, Gavin; Herben, Martin; Lloyd, Pam; Rose, Danny; Smith, Chris; Barraclough, Ian (Enviros Consulting Ltd (GB))

    2008-03-15

    This report documents an exercise in which AMBER has been used to represent the models used in Project SAFE, a safety assessment undertaken on SFR 1. (AMBER is a flexible, graphical-user-interface based tool that allows users to build their own dynamic compartmental models to represent the migration, degradation and fate of contaminants in an environmental system. AMBER allows the user to assess routine, accidental and long-term contaminant release.) AMBER has been used to undertake assessment calculations on all of the disposal system, including all disposal tunnels and the Silo, the geosphere and several biosphere modules. The near-field conceptual models were implemented with minimal changes to the approach undertaken previously in Project SAFE. Model complexity varied significantly between individual disposal facilities increasing significantly from the BLA to the BTF and BMA tunnels and Silo. Radionuclide transport through the fractured granite geosphere was approximated using a compartment model approach in AMBER. Several biosphere models were implemented in AMBER including reasonable biosphere development, which considered the evolution of the Forsmark area from coastal to lacustrine to agricultural environments in response to land uplift. Parameters were sampled from distributions and simulations were run for 1,000 realisations. In undertaking the comparison of AMBER with the various codes and calculation tools used in Project SAFE it was necessary to undertake a detailed analysis of the modelling approach previously adopted, with particular focus given to the near-field models. As a result some discrepancies in the implementation of the models and documentation were noted. The exercise demonstrates that AMBER is fully capable of representing the features of the SFR 1 disposal system in a safety assessment suitable for SAR-08

  5. Implementation of project Safe in Amber. Verification study for SFR 1 SAR-08

    International Nuclear Information System (INIS)

    Thomson, Gavin; Herben, Martin; Lloyd, Pam; Rose, Danny; Smith, Chris; Barra clough, Ian

    2008-03-01

    This report documents an exercise in which AMBER has been used to represent the models used in Project SAFE, a safety assessment undertaken on SFR 1. (AMBER is a flexible, graphical-user-interface based tool that allows users to build their own dynamic compartmental models to represent the migration, degradation and fate of contaminants in an environmental system. AMBER allows the user to assess routine, accidental and long-term contaminant release.) AMBER has been used to undertake assessment calculations on all of the disposal system, including all disposal tunnels and the Silo, the geosphere and several biosphere modules. The near-field conceptual models were implemented with minimal changes to the approach undertaken previously in Project SAFE. Model complexity varied significantly between individual disposal facilities increasing significantly from the BLA to the BTF and BMA tunnels and Silo. Radionuclide transport through the fractured granite geosphere was approximated using a compartment model approach in AMBER. Several biosphere models were implemented in AMBER including reasonable biosphere development, which considered the evolution of the Forsmark area from coastal to lacustrine to agricultural environments in response to land uplift. Parameters were sampled from distributions and simulations were run for 1,000 realisations. In undertaking the comparison of AMBER with the various codes and calculation tools used in Project SAFE it was necessary to undertake a detailed analysis of the modelling approach previously adopted, with particular focus given to the near-field models. As a result some discrepancies in the implementation of the models and documentation were noted. The exercise demonstrates that AMBER is fully capable of representing the features of the SFR 1 disposal system in a safety assessment suitable for SAR-08

  6. On the effect of different placing ZrH moderator material on the performance of a SFR core

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B. [Helmholtz-Zentrum Dresden-Rossendorf, Institut fuer Sicherheitsforschung, Postfach 51 01 19, 01314 Dresden (Germany); Weiss, F. P. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH Forschungszentrum, Boltzmannstr. 14, 85748 Garching (Germany)

    2012-07-01

    This study describes the development of a sodium fast reactor fuel assembly design with reduced void reactivity coefficient, achieved through the use of the ZrH moderating material. In the study the sodium void effect, as well as the major feedback coefficients are analyzed. Besides the feedback coefficients, the influence on the operational parameters like neutron flux distribution, power distribution, and burnup distribution is investigated for the different possibilities of arranging the moderating material in the fuel assembly. Additionally, the fuel cycle parameters - breeding and minor actinide production - are analyzed. For a first evaluation of the behavior during transients the influence of temperature changes in the ZrH is studied. (authors)

  7. Does core stability exercise improve lumbopelvic stability (through endurance tests) more than general exercise in chronic low back pain? A quasi-randomized controlled trial.

    Science.gov (United States)

    Shamsi, Mohammad Bagher; Rezaei, Mandana; Zamanlou, Mehdi; Sadeghi, Mehdi; Pourahmadi, Mohammad Reza

    2016-01-01

    The aim was to compare core stability and general exercises (GEs) in chronic low back pain (LBP) patients based on lumbopelvic stability (LPS) assessment through three endurance core stability tests. There is a controversy about preference of core stability exercise (CSE) over other types of exercise for chronic LBP. Studies which have compared these exercises used other outcomes than those related to LPS. As it is claimed that CSE enhances back stability, endurance tests for LPS were used. A 16-session CSE program and a GE program with the same duration were conducted for two groups of participants. Frequency of interventions for both groups was three times a week. Forty-three people (aged 18-60 years) with chronic non-specific LBP were alternately allocated to core stability (n = 22) or GE group (n = 21) when admitted. The primary outcomes were three endurance core stability tests including: (1) trunk flexor; (2) trunk extensor; and (3) side bridge tests. Secondary outcomes were disability and pain. Measurements were taken at baseline and the end of the intervention. After the intervention, test times increased and disability and pain decreased within groups. There was no significant difference between two groups in increasing test times (p = 0.23 to p = 0.36) or decreasing disability (p = 0.16) and pain (p = 0.73). CSE is not more effective than GE for improving endurance core stability tests and reducing disability and pain in chronic non-specific LBP patients.

  8. Site investigation SFR. Fracture mineralogy including identification of uranium phases and hydrochemical characterisation of groundwater in borehole KFR106

    International Nuclear Information System (INIS)

    Sandstroem, Bjoern; Nilsson, Kersti; Tullborg, Eva-Lena

    2011-12-01

    This report presents the fracture mineralogy and hydrochemistry of borehole KFR106. The most abundant fracture minerals in the examined drill core samples are clay minerals, calcite, quartz and adularia; chlorite is also common but is mostly altered and found interlayered with corrensite. The most common clay mineral is a mixed layer clay consisting of illite-smectite. Pyrite, galena, chalcopyrite, barite (-celestine) and hematite are also commonly found in the fractures, but usually in trace amounts. Other minerals identified in the examined fractures are U-phosphate, pitchblende, U(Ca)-silicate, asphaltite, biotite, monazite, fluorite, titanite, sericite, xenotime, rutile and (Ca, REEs)-carbonate. Uranium has been introduced, mobilised and reprecipitated during at least four different episodes: 1) Originally, during emplacement of U-rich pegmatites, probably as uraninite. 2) At a second event, uranium was mobilised under brittle conditions during formation of breccia/cataclasite. Uraninite was altered to pitchblende and partly coffinitised. Mobilised uranium precipitated as pitchblende closely associated with hematite and chlorite in cataclasite and fracture sealings prior to 1,000 Ma. 3) During the Palaeozoic U was remobilised and precipitated as U-phosphate on open fracture surfaces. 4) An amorphous U-silicate has also been found in open fractures; the age of this precipitation is not known but it is inferred to be Palaeozoic or younger. Groundwater was sampled in two sections in borehole KFR106 with pumping sequences of about 6 days for each section. The samples from sections KFR106:1 and KFR106:2 (260-300 m and 143-259 m borehole length, i.e. -261 and -187 m.a.s.l. mid elevation of the section, respectively) were taken in November 2009 and yielded groundwater chemistry data in accordance with SKB chemistry class 3 and 5. In section KFR106:1 and KFR106:2, the chloride contents were 850 and 1,150 mg/L and the drilling water content 6 and 4%, respectively

  9. Site investigation SFR. Fracture mineralogy including identification of uranium phases and hydrochemical characterisation of groundwater in borehole KFR106

    Energy Technology Data Exchange (ETDEWEB)

    Sandstroem, Bjoern [WSP Sverige AB, Goeteborg (Sweden); Nilsson, Kersti [Geosigma AB, Uppsala (Sweden); Tullborg, Eva-Lena [Terralogica AB, Graabo (Sweden)

    2011-12-15

    This report presents the fracture mineralogy and hydrochemistry of borehole KFR106. The most abundant fracture minerals in the examined drill core samples are clay minerals, calcite, quartz and adularia; chlorite is also common but is mostly altered and found interlayered with corrensite. The most common clay mineral is a mixed layer clay consisting of illite-smectite. Pyrite, galena, chalcopyrite, barite (-celestine) and hematite are also commonly found in the fractures, but usually in trace amounts. Other minerals identified in the examined fractures are U-phosphate, pitchblende, U(Ca)-silicate, asphaltite, biotite, monazite, fluorite, titanite, sericite, xenotime, rutile and (Ca, REEs)-carbonate. Uranium has been introduced, mobilised and reprecipitated during at least four different episodes: 1) Originally, during emplacement of U-rich pegmatites, probably as uraninite. 2) At a second event, uranium was mobilised under brittle conditions during formation of breccia/cataclasite. Uraninite was altered to pitchblende and partly coffinitised. Mobilised uranium precipitated as pitchblende closely associated with hematite and chlorite in cataclasite and fracture sealings prior to 1,000 Ma. 3) During the Palaeozoic U was remobilised and precipitated as U-phosphate on open fracture surfaces. 4) An amorphous U-silicate has also been found in open fractures; the age of this precipitation is not known but it is inferred to be Palaeozoic or younger. Groundwater was sampled in two sections in borehole KFR106 with pumping sequences of about 6 days for each section. The samples from sections KFR106:1 and KFR106:2 (260-300 m and 143-259 m borehole length, i.e. -261 and -187 m.a.s.l. mid elevation of the section, respectively) were taken in November 2009 and yielded groundwater chemistry data in accordance with SKB chemistry class 3 and 5. In section KFR106:1 and KFR106:2, the chloride contents were 850 and 1,150 mg/L and the drilling water content 6 and 4%, respectively

  10. Improvement of the Cubic Spline Function Sets for a Synthesis of the Axial Power Distribution of a Core Protection System

    International Nuclear Information System (INIS)

    Koo, Bon-Seung; Lee, Chung-Chan; Zee, Sung-Quun

    2006-01-01

    Online digital core protection system(SCOPS) for a system-integrated modular reactor is being developed as a part of a plant protection system at KAERI. SCOPS calculates the minimum CHFR and maximum LPD based on several online measured system parameters including 3-level ex-core detector signals. In conventional ABB-CE digital power plants, cubic spline synthesis technique has been used in online calculations of the core axial power distributions using ex-core detector signals once every 1 second in CPC. In CPC, pre-determined cubic spline function sets are used depending on the characteristics of the ex-core detector responses. But this method shows an unnegligible power distribution error for the extremely skewed axial shapes by using restrictive function sets. Therefore, this paper describes the cubic spline method for the synthesis of an axial power distribution and it generates several new cubic spline function sets for the application of the core protection system, especially for the severely distorted power shapes needed reactor type

  11. A key to improved ion core confinement in the JET tokamak: ion stiffness mitigation due to combined plasma rotation and low magnetic shear.

    Science.gov (United States)

    Mantica, P; Angioni, C; Challis, C; Colyer, G; Frassinetti, L; Hawkes, N; Johnson, T; Tsalas, M; deVries, P C; Weiland, J; Baiocchi, B; Beurskens, M N A; Figueiredo, A C A; Giroud, C; Hobirk, J; Joffrin, E; Lerche, E; Naulin, V; Peeters, A G; Salmi, A; Sozzi, C; Strintzi, D; Staebler, G; Tala, T; Van Eester, D; Versloot, T

    2011-09-23

    New transport experiments on JET indicate that ion stiffness mitigation in the core of a rotating plasma, as described by Mantica et al. [Phys. Rev. Lett. 102, 175002 (2009)] results from the combined effect of high rotational shear and low magnetic shear. The observations have important implications for the understanding of improved ion core confinement in advanced tokamak scenarios. Simulations using quasilinear fluid and gyrofluid models show features of stiffness mitigation, while nonlinear gyrokinetic simulations do not. The JET experiments indicate that advanced tokamak scenarios in future devices will require sufficient rotational shear and the capability of q profile manipulation.

  12. KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code

    International Nuclear Information System (INIS)

    Kim, Young Gyun; Kim, Young Il

    2006-12-01

    Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006

  13. Dithranol-loaded lipid-core nanocapsules improve the photostability and reduce the in vitro irritation potential of this drug

    Energy Technology Data Exchange (ETDEWEB)

    Savian, Ana L. [Programa de Pós-Graduação em Ciências Farmacêuticas, Universidade Federal de Santa Maria, Av. Roraima, 1000, Santa Maria, RS 97105-900 (Brazil); Rodrigues, Daiane [Curso de Farmácia, Centro de Ciências da Saúde, Universidade Federal de Santa Maria, Av. Roraima, 1000, Santa Maria, RS 97105-900 (Brazil); Weber, Julia; Ribeiro, Roseane F. [Programa de Pós-Graduação em Ciências Farmacêuticas, Universidade Federal de Santa Maria, Av. Roraima, 1000, Santa Maria, RS 97105-900 (Brazil); Motta, Mariana H. [Curso de Farmácia, Centro de Ciências da Saúde, Universidade Federal de Santa Maria, Av. Roraima, 1000, Santa Maria, RS 97105-900 (Brazil); Schaffazick, Scheila R.; Adams, Andréa I.H. [Programa de Pós-Graduação em Ciências Farmacêuticas, Universidade Federal de Santa Maria, Av. Roraima, 1000, Santa Maria, RS 97105-900 (Brazil); Andrade, Diego F. de; Beck, Ruy C.R. [Programa de Pós-Graduação em Ciências Farmacêuticas, Universidade Federal do Rio Grande do Sul, Av. Ipiranga, 2752, Porto Alegre, RS 90610-000 (Brazil); and others

    2015-01-01

    Dithranol is a very effective drug for the topical treatment of psoriasis. However, it has some adverse effects such as irritation and stain in the skin that make its application and patient adherence to treatment difficult. The aims of this work were to prepare and characterize dithranol-loaded nanocapsules as well as to evaluate the photostability and the irritation potential of these nanocarriers. Lipid-core nanocapsules containing dithranol (0.5 mg/mL) were prepared by interfacial deposition of preformed polymer. EDTA (0.05%) or ascorbic acid (0.02%) was used as antioxidants. After preparation, dithranol-loaded lipid-core nanocapsules showed satisfactory characteristics: drug content close to the theoretical concentration, encapsulation efficiency of about 100%, nanometric mean size (230–250 nm), polydispersity index below 0.25, negative zeta potential, and pH values from 4.3 to 5.6. In the photodegradation study against UVA light, we observed a higher stability of the dithranol-loaded lipid-core nanocapsules comparing to the solution containing the free drug (half-life times around 4 and 1 h for the dithranol-loaded lipid-core nanocapsules and free drug solution containing EDTA, respectively; half-life times around 17 and 7 h for the dithranol-loaded lipid-core nanocapsules and free drug solution containing ascorbic acid, respectively). Irritation test by HET-CAM method was conducted to evaluate the safety of the formulations. From the results it was found that the nanoencapsulation of the drug decreased its toxicity compared to the effects observed for the free drug. - Highlights: • Strategy to prepare lipid-core nanocapsules containing dithranol • Evaluation of the nanoencapsulation effect on the photostability and irritation • Evaluation of the in vitro release of dithranol-loaded lipid-core nanocapsules.

  14. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  15. Low void effect (CFV) core concept flexibility: from self-breeder to burner core - 15091

    International Nuclear Information System (INIS)

    Buiron, L.; Dujcikova, L.

    2015-01-01

    In the frame of the French strategy on sustainable nuclear energy, several scenarios consider fuel cycle transition toward a plutonium multi-recycling strategy in sodium cooled fast reactor (SFR). Basically, most of these scenarios consider the deployment of a 60 GWe SFR fleet in 2 steps to renew the French PWR fleet. As scenarios do investigate long term deployment configurations, some of them require tools for nuclear phase-out studies. Instead of designing new reactors, the adopted strategy does focus on adaptation of existing ones into burner configurations. This is what was done in the frame of the EFR project at the end of the 90's using the CAPRA approach (French acronym for Enhance Plutonium Consumption in Fast Reactor). The EFR burner configuration was obtained by inserting neutronic penalties inside the core (absorber material and/or diluent subassembly). Starting from the preliminary industrial image of a SFR 3600 MWth core based on Low Sodium Void concept (CFV in French), a 'CAPRA-like' approach has been studied. As the CFV self-breeding is ensured by fertile blankets, a first modification consisted in the substitution of the corresponding depleted uranium by 'inert' or absorber material leading to a 'natural burner' core with only small impacts on flux distribution. The next step forward CAPRA configuration was the substitution of 1/3 of the fuel pins by 'dummy' pins (MgO pellets). The small spectrum shift due to MgO material insertion leads to an increase Doppler constant which exceeds the value of the reference case. As the core sodium void worth value is conserved, the CFV CAPRA core 'safety' potential is quite similar to the one of the reference core. Fuel thermo-mechanical requirements are met by both nominal core power and fuel time residence reduction. However, these reduction factors are lower than those obtained for EFR core. The management of the enhanced reactivity swing is discussed

  16. Development of Core Design Technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Il; Hong, S. G.; Jang, J. W. (and others)

    2007-06-15

    This report describes the contents of core design technology and computer code system development performed during 2005 and 2006 on the objects of nuclear proliferation resistant core and nuclear fuel basic key technology development security. Also, it is including the future application plans for the results and the developed methodology, important information and the materials acquired in this period. Two core designs with single enrichment were considered for the KALIMER-600 during the first year : 1) the first core uses the non-fuel rods such as B4C, ZrH1.8, and dummy rods, 2) the core using different cladding thickness for each core region (inner, middle, and outer cores) without non-fuel rods to flatten the power distribution. In particular, the latter design was intended to simplify the fuel assembly design by eliminating the heterogeneity. It was found that the proposed design satisfy all of the Gen IV SFR design goals on the cycle length longer than 18 EFPM, fuel discharge burnup larger than 80GWd/t, sodium void worth, conversion ratio, reactivity burnup swing and so on. For this object reactor, the structure integrity outside of reactor is confirmed for the radiation exposure during the plant life according to the result of shielding design and evaluation. The transmutation capability and the core characteristics of sodium cooled fast reactor was also evaluated according to the change of MA amount. The reactivity coefficients for the BN-600 reactor with MA fueled are calculated and the results are compared and evaluated with other participants results. Even though the discrepancies between the results of participants are somewhat large but the K-CORE results are close to the average within a standard deviation. To have the capability of 3-dimensional core dynamic analysis such as analyzing power distribution and reactivity variations according to the asymmetric insertion/withdrawal of control rods, the calculation module for core dynamic parameters was

  17. An approach of SFR safety study through the most penalizing sodium void reactivity - 105

    International Nuclear Information System (INIS)

    Tiberi, V.; Ivanov, E.; Pignet, S.

    2010-01-01

    Sodium void reactivity effects can affect the plant safety significantly during accidental transients. Accordingly, they have to be accurately investigated for any new sodium cooled fast reactor concept, even if a fuel with a melting point lower than the sodium boiling temperature is adopted. Thus all new reactor concepts should be compared to each - others adopting the value of the maximal possible sodium void reactivity as a discrimination parameter. However, taking into account that the sodium void worth is spatially depended, it is not evident which volume could be voided in order to obtain the maximal reactivity increase. Typically the complete active core voiding (zones initially loaded with 235 U or 239 Pu) is taken into account. This paper summarizes the extensive work carried-out in the IRSN to investigate the sodium-void reactivity spatial profiles of a fast sodium-cooled reactor core in the aim of defining a methodology to search for the area where the void contribution to the reactivity is strictly positive. Perturbation theory design approach available in the ERANOS 2.1 code has been adopted to evaluate the 'area of positive void worth'. To do that, the code has been previously validated against experimental based benchmarks (IRPhEP) and reference calculations. The evaluation of the absolute values of reactivity profiles can be improved later-on adopting more sophisticated methodologies to perform more accurate calculations of the sample with the voided area determined adopting the rough procedure described here. It has been demonstrated that even the non-reference way of ERANOS calculations could be used to provide the basis for different core concepts inter-comparison. (authors)

  18. Improved microbial growth inhibition activity of bio-surfactant induced Ag–TiO{sub 2} core shell nanoparticles

    Energy Technology Data Exchange (ETDEWEB)

    Nithyadevi, D. [Department of Nanoscience and Technology, Bharathiar University, Coimbatore 641 046 (India); Kumar, P. Suresh [Thin Film and Nanomaterials Laboratory, Department of Physics, Bharathiar University, Coimbatore 641 046 (India); Mangalaraj, D., E-mail: dmraj800@yahoo.com [Department of Nanoscience and Technology, Bharathiar University, Coimbatore 641 046 (India); Ponpandian, N.; Viswanathan, C. [Department of Nanoscience and Technology, Bharathiar University, Coimbatore 641 046 (India); Meena, P. [Department of Physics, PSGR Krishnammal college for women, Coimbatore 641 004 (India)

    2015-02-01

    Graphical abstract: - Highlights: • TiO{sub 2} nanoparticles were synthesized by hydrolysis process and Ag nanoparticles were prepared by using hydrazine reduction method. • Ag–TiO{sub 2} core shell nanoparticles were synthesized by reverse micelle method. • Coatings of TiO{sub 2} shell leads to decrease the usage of silver particles and also it reduces the release of silver ions from the matrix. • Optimum ratio of TiO{sub 2} particles: Ag atoms are needed for better antibacterial activity. • Sodium alginate (Bio-copolymer) induced core shell nanoparticles results 100% cell growth inhibition toward Staphylococcus aureus. - Abstract: Surfactant induced silver–titanium dioxide core shell nanoparticles within the size range of 10–50 nm were applied in the antibacterial agent to inhibit the growth of bacterial cells. The single crystalline silver was located in the core part of the composite powder and the titanium dioxide components were uniformly distributed in the shell part. HRTEM and XRD results indicated that silver was completely covered by titanium dioxide and its crystal structure was not affected after being coated by titanium dioxide. The effect of silver–titanium dioxide nanoparticles in the inhibition of bacterial cell growth was studied by means of disk diffusion method. The inhibition zone results reveal that sodium alginate induced silver–titanium dioxide nanoparticles exhibit 100% more antibacterial activity than that with cetyltrimethylbromide or without surfactant. UV–vis spectroscopic analysis showed a large concentration of silver was rapidly released into phosphate buffer solution (PBS) within a period of 1 day, with a much smaller concentration being released after this 1-day period. It was concluded that sodium alginate induced silver–titanium dioxide core shell nanoparticles could enhance long term cell growth inhibition in comparison with cetyltrimethylbromide or without surfactant. The surfactant mediated core shell

  19. Core stability exercise is as effective as task-oriented motor training in improving motor proficiency in children with developmental coordination disorder: a randomized controlled pilot study.

    Science.gov (United States)

    Au, Mei K; Chan, Wai M; Lee, Lin; Chen, Tracy Mk; Chau, Rosanna Mw; Pang, Marco Yc

    2014-10-01

    To compare the effectiveness of a core stability program with a task-oriented motor training program in improving motor proficiency in children with developmental coordination disorder (DCD). Randomized controlled pilot trial. Outpatient unit in a hospital. Twenty-two children diagnosed with DCD aged 6-9 years were randomly allocated to the core stability program or the task-oriented motor program. Both groups underwent their respective face-to-face training session once per week for eight consecutive weeks. They were also instructed to carry out home exercises on a daily basis during the intervention period. Short Form of the Bruininks-Oseretsky Test of Motor Proficiency (Second Edition) and Sensory Organization Test at pre- and post-intervention. Intention-to-treat analysis revealed no significant between-group difference in the change of motor proficiency standard score (P=0.717), and composite equilibrium score derived from the Sensory Organization Test (P=0.100). Further analysis showed significant improvement in motor proficiency in both the core stability (mean change (SD)=6.3(5.4); p=0.008) and task-oriented training groups (mean change(SD)=5.1(4.0); P=0.007). The composite equilibrium score was significantly increased in the task-oriented training group (mean change (SD)=6.0(5.5); P=0.009), but not in the core stability group (mean change(SD) =0.0(9.6); P=0.812). In the task-oriented training group, compliance with the home program was positively correlated with change in motor proficiency (ρ=0.680, P=0.030) and composite equilibrium score (ρ=0.638, P=0.047). The core stability exercise program is as effective as task-oriented training in improving motor proficiency among children with DCD. © The Author(s) 2014.

  20. Oxidation driven ZnS Core-ZnO shell photocatalysts under controlled oxygen atmosphere for improved photocatalytic solar water splitting

    Science.gov (United States)

    Bak, Daegil; Kim, Jung Hyeun

    2018-06-01

    Zinc type photocatalysts attract great attentions in solar hydrogen production due to their easy availability and benign environmental characteristics. Spherical ZnS particles are synthesized with a facile hydrothermal method, and they are further used as core materials to introduce ZnO shell layer surrounding the core part by partial oxidation under controlled oxygen contents. The resulting ZnS core-ZnO shell photocatalysts represent the heterostructural type II band alignment. The existence of oxide layer also influences on proton adsorption power with an aid of strong base cites derived from highly electronegative oxygen atoms in ZnO shell layer. Photocatalytic water splitting reaction is performed to evaluate catalyst efficiency under standard one sun condition, and the highest hydrogen evolution rate (1665 μmolg-1h-1) is achieved from the sample oxidized at 16.2 kPa oxygen pressure. This highest hydrogen production rate is achieved in cooperation with increased light absorption and promoted charge separations. Photoluminescence analysis reveals that the improved visible light response is obtained after thermal oxidation process due to the oxygen vacancy states in the ZnO shell layer. Therefore, overall photocatalytic efficiency in solar hydrogen production is enhanced by improved charge separations, crystallinity, and visible light responses from the ZnS core-ZnO shell structures induced by thermal oxidation.

  1. Models for dose assessments. Models adapted to the SFR-area, Sweden

    International Nuclear Information System (INIS)

    Karlsson, Sara; Bergstroem, U.; Meili, M.

    2001-10-01

    This report presents a model system created to be used to predict dose rates to the most exposed individuals in case of a long-term release of radionuclides from the Final repository for radioactive operational waste (SFR) in Forsmark, Sweden. The system accounts for an underground point source emitting radionuclides to the biosphere,their transport and distribution in various ecosystem types, their uptake by various biota, and calculation of doses to man from a multitude of exposure pathways. Long-term aspects include the consideration of processes at geological time scales, such as land uplift and conversion of marine sediments to arable land. Model parameters are whenever possible based on local conditions and recent literature. The system has been used for simulations based on geospheric releases varying over time of a mixture of radionuclides. Here, the models have been subjected to various test scenarios, including different radionuclide entry points and sorption properties. Radionuclides released from SFR are efficiently diluted to low concentrations in the water of the coastal model. A large fraction of the nuclides leave the Model Area quickly as a consequence of the rapid water turnover. The total amount of radionuclides in water is about the same independent of particle affinity (K d ), and at most some percents of the amounts retained in the sediments for some time. The latter is also true for the lake model when releases of radionuclides to the water is simulated. The role of sediments as a radionuclide source seems of minor importance in lakes at least for long-term radiation doses. Modelling the sediments as a source of radionuclides most of the 'low K d radionuclides' will leave the lake whereas the 'high K d nuclides' are still present within the deeper sediments after 1 000 years. The amount of 'low K d radionuclides' present in the water and on suspended matter are about 8x10 -5 of the initial inventory in the sediments. For 'high K d nuclides

  2. Experimental results of passive vibro-acoustic leak detection in SFR steam generator mock-up

    International Nuclear Information System (INIS)

    Moriot, J.; Gastaldi, O.; Maxit, L.; Guyader, J-L.; Perisse, J.; Migot, B.

    2013-06-01

    Regarding to GEN 4 context, it is necessary to fulfil the high safety standards for sodium fast reactors (SFR), particularly against water-sodium reaction which may occur in the steam generator units (SGU) in case of leak. This reaction can cause severe damages in the component in a short time. Detecting such a leak by visual in-sodium inspection is impossible because of sodium opacity. Hydrogen detection is then used but the time response of this method can be high in certain operating conditions. Active and passive acoustic leak detection methods were studied before SUPERPHENIX plant shutdown in 1997 to detect a water-into-sodium leak with a short time response. In the context of the new R and D studies for SFR, an innovative passive vibro-acoustic method is developed in the framework of a Ph.D. thesis to match with GEN 4 safety requirements. The method consists in assuming that a small leak emits spherical acoustic waves in a broadband frequency domain, which propagate in the liquid sodium and excite the SGU cylindrical shell. These spatially coherent waves are supposed to be buried by a spatially incoherent background noise. The radial velocities of the shell is measured by an array of accelerometers positioned on the external envelop of the SGU and a beam forming treatment is applied to increase the signal-to-noise ratio (SNR) and to detect and localize the acoustic source. Previous numerical experiments were achieved and promising results were obtained. In this paper, experimental results of the proposed passive vibro-acoustic leak detection are presented. The experiment consists in a cylindrical water-filled steel pipe representing a model of SGU shell without tube bundle. A hydro-phone emitting an acoustic signal is used to simulate an acoustic monopole. Spatially uncorrelated noise or water-flow induced shell vibrations are considered as the background noise. The beam-forming method is applied to vibration signals measured by a linear array of

  3. Four-Strand Core Suture Improves Flexor Tendon Repair Compared to Two-Strand Technique in a Rabbit Model

    Directory of Open Access Journals (Sweden)

    Alice Wichelhaus

    2016-01-01

    Full Text Available Introduction. This study was designed to investigate the influence of the amount of suture material on the formation of peritendinous adhesions of intrasynovial flexor tendon repairs. Materials and Methods. In 14 rabbits, the flexor tendons of the third and the fourth digit of the right hind leg were cut and repaired using a 2- or 4-strand core suture technique. The repaired tendons were harvested after three and eight weeks. The range of motion of the affected toes was measured and the tendons were processed histologically. The distance between the transected tendon ends, the changes in the peritendinous space, and cellular and extracellular inflammatory reaction were quantified by different staining. Results. A 4-strand core suture resulted in significantly less gap formation. The 2-strand core suture showed a tendency to less adhesion formation. Doubling of the intratendinous suture material was accompanied by an initial increase in leukocyte infiltration and showed a greater amount of formation of myofibroblasts. From the third to the eighth week after flexor tendon repair, both the cellular and the extracellular inflammation decreased significantly. Conclusion. A 4-strand core suture repair leads to a significantly better tendon healing process with less diastasis between the sutured tendon ends despite initially pronounced inflammatory response.

  4. Examining the Effects of a National League for Nursing Core Competencies Workshop as an Intervention to Improve Nurse Faculty Practice

    Science.gov (United States)

    VanBever Wilson, Robin R.

    2010-01-01

    Due to the complex challenges facing schools of nursing, a research study was implemented to introduce nurse faculty at one small rural northeastern Tennessee school of nursing to the NLN "Core Competencies for Nurse Educators". Utilizing Kalb's Nurse Faculty Self-Evaluation Tool as a pre- and post-intervention test, 30 nurse faculty…

  5. Accurate pan-specific prediction of peptide-MHC class II binding affinity with improved binding core identification

    DEFF Research Database (Denmark)

    Andreatta, Massimo; Karosiene, Edita; Rasmussen, Michael

    2015-01-01

    with known binding registers, the new method NetMHCIIpan-3.1 significantly outperformed the earlier 3.0 version. We illustrate the impact of accurate binding core identification for the interpretation of T cell cross-reactivity using tetramer double staining with a CMV epitope and its variants mapped...

  6. Formulation and evaluation of gas release scenarios for the silo in Swedish Final Repository for Radioactive Waste (SFR)

    International Nuclear Information System (INIS)

    Carlsson, J.; Moreno, L.

    1992-01-01

    The Swedish Final Repository for Radioactive Waste (SFR) has been in operation since 1988 and is located in the crystalline rock, 60 m below the Baltic Sea. In the licensing procedure for the SFR the safety assessment has been complemented with a detailed scenario analysis of the performance of the repository. The scenarios include the influence on radionuclide release by gas formation and gas transport processes in the silo. The overall conclusion is that the release of most radionuclides from the silo is only marginally affected by the formation and release of gas, even for scenarios considering unexpected events. The largest effects were found for short-lived radionuclides and radionuclides that have no or low sorption ability. Except for very extreme scenarios for the silo the overall impact from repository on the environment is by far dominated by the release of radionuclides from the rock vaults. 10 refs., 6 figs

  7. Swi5-Sfr1 protein stimulates Rad51-mediated DNA strand exchange reaction through organization of DNA bases in the presynaptic filament.

    KAUST Repository

    Fornander, Louise H; Renodon-Corniè re, Axelle; Kuwabara, Naoyuki; Ito, Kentaro; Tsutsui, Yasuhiro; Shimizu, Toshiyuki; Iwasaki, Hiroshi; Nordé n, Bengt; Takahashi, Masayuki

    2013-01-01

    The Swi5-Sfr1 heterodimer protein stimulates the Rad51-promoted DNA strand exchange reaction, a crucial step in homologous recombination. To clarify how this accessory protein acts on the strand exchange reaction, we have analyzed how the structure

  8. Development and Applicability Demonstration of a Remote Inspection Module for Inspection of Reactor Internals in an SFR

    International Nuclear Information System (INIS)

    Kim, Hoewoong; Joo, Youngsang; Park, Changgyu; Kim, Jongbum; Bae, Jinho

    2014-01-01

    Since liquid sodium is optically opaque, the ultrasonic inspection technique has been mainly employed for inspection of reactor internals in a Sodium-cooled Fast Reactor (SFR). Until now, two types of ultrasonic sensors have been mainly developed; immersion and waveguide sensors. An immersion sensor can provide a high-resolution image, but it may have problems in terms of reliability and life time because the sensor is exposed to high temperature during inspection. On the other hand, a waveguide sensor can maintain its performance during long-term inspection in high temperature because it installs an ultrasonic transducer in a cold region even though such a high-frequency ultrasonic wave cannot be used owing to the long propagation distance [4-6]. In this work, a remote inspection module employing four 10 m long waveguide sensors was newly developed and several performance tests were carried out in water to demonstrate the applicability of the developed remote inspection module to inspection of reactor internals in an SFR. In this work, a remote inspection module for inspection of reactor internals in an SFR was newly developed. The developed remote inspection module employs four 10 m long waveguide sensors for multiple inspection applications: a horizontal beam waveguide sensor for ranging inspection, two vertical beam waveguide sensors for viewing inspection and a 45 .deg. angle beam waveguide sensor for identification inspection. Several performance tests such as ranging, viewing and identification inspections were carried out for simulated nuclear fuel assembly specimens in water, and the applicability of the developed remote inspection module to inspection of reactor internals in an SFR was demonstrated

  9. Development and Applicability Demonstration of a Remote Inspection Module for Inspection of Reactor Internals in an SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hoewoong; Joo, Youngsang; Park, Changgyu; Kim, Jongbum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Bae, Jinho [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Since liquid sodium is optically opaque, the ultrasonic inspection technique has been mainly employed for inspection of reactor internals in a Sodium-cooled Fast Reactor (SFR). Until now, two types of ultrasonic sensors have been mainly developed; immersion and waveguide sensors. An immersion sensor can provide a high-resolution image, but it may have problems in terms of reliability and life time because the sensor is exposed to high temperature during inspection. On the other hand, a waveguide sensor can maintain its performance during long-term inspection in high temperature because it installs an ultrasonic transducer in a cold region even though such a high-frequency ultrasonic wave cannot be used owing to the long propagation distance [4-6]. In this work, a remote inspection module employing four 10 m long waveguide sensors was newly developed and several performance tests were carried out in water to demonstrate the applicability of the developed remote inspection module to inspection of reactor internals in an SFR. In this work, a remote inspection module for inspection of reactor internals in an SFR was newly developed. The developed remote inspection module employs four 10 m long waveguide sensors for multiple inspection applications: a horizontal beam waveguide sensor for ranging inspection, two vertical beam waveguide sensors for viewing inspection and a 45 .deg. angle beam waveguide sensor for identification inspection. Several performance tests such as ranging, viewing and identification inspections were carried out for simulated nuclear fuel assembly specimens in water, and the applicability of the developed remote inspection module to inspection of reactor internals in an SFR was demonstrated.

  10. Characterization and modelling of the thermodynamic behavior of SFR fuel under irradiation

    International Nuclear Information System (INIS)

    Pham-Thi, Tam-Ngoc

    2014-01-01

    For a burn-up higher than 7 at%, the volatile FP like Cs, I and Te or metallic (Mo) are partially released from the fuel pellet in order to form a layer of compounds between the outer surface of the fuel and the inner surface of the stainless cladding. This layer is called the JOG, french acronym for Joint-Oxyde-Gaine. My subject is focused on two topics: the thermodynamic study of the (Cs-I-Te-Mo-O) system and the migration of those FP towards the gap to form the JOG. The thermodynamic study was the first step of my work. On the basis of critical literature survey, the following systems have been optimized by the CALPHAD method: Cs-Te, Cs-I and Cs-Mo-O. In parallel, an experimental study is undertaken in order to validate our CALPHAD modelling of the Cs-Te system. In a second step, the thermodynamic data coming from the CALPHAD modelling have been introduced into the database that we use with the thermochemical computation code ANGE (CEA code derived from the SOLGASMIX software) in order to calculate the chemical composition of the irradiated fuel versus burn-up and temperature. In a third and last step, the thermochemical computation code ANGE (Advanced Numeric Gibbs Energy minimizer) has been coupled with the fuel performance code GERMINAL V2, which simulates the thermo-mechanical behavior of SFR fuel. (author) [fr

  11. Project SAFE. Modelling of long-term concrete degradation processes in the Swedish SFR repository

    Energy Technology Data Exchange (ETDEWEB)

    Hoeglund, L.O. [Kemakta Konsult AB, Stockholm (Sweden)

    2001-04-01

    This study concerns the leaching of concrete barriers, in particular the silo construction, in the Swedish SFR repository for low and intermediate level radioactive waste. A conceptual model for the leaching of concrete in a saline groundwater has been proposed based on the increased understanding achieved from research studies presented in the literature. The conceptual model has been used to set up a numerical model for the complex chemical interactions between the cement minerals of the concrete with the groundwater. The calculations show that various chemical reactions are expected to occur in the concrete over time. Different cases have been calculated. The results show that the chemical conditions in the concrete barriers will maintain alkaline for long time. In the most exposed parts of the concrete a high degree of leaching can be expected during the considered 10,000 years, whereas only for the most unfavourable assumptions (initially fractured concrete with groundwater flow-through) the inner parts of the concrete will be degraded to any significant degree.

  12. Experimental and Numerical Analysis of S-CO2 Critical Flow for SFR Recovery System Design

    International Nuclear Information System (INIS)

    Kim, Min Seok; Jung, Hwa-Young; Ahn, Yoonhan; Lee, Jekyoung; Lee, Jeong Ik

    2016-01-01

    This paper presents both numerical and experimental studies of the critical flow of S-CO 2 while special attention is given to the turbo-machinery seal design. A computational critical flow model is described first. The experiments were conducted to validate the critical flow model. Various conditions have been tested to study the flow characteristic and provide validation data for the model. The comparison of numerical and experimental results of S-CO 2 critical flow will be presented. In order to eliminate SWR, a concept of coupling the Supercritical CO 2 (S-CO 2 ) cycle with SFR has been proposed. It is known that for a closed system controlling the inventory is important for stable operation and achieving high efficiency. Since the S-CO 2 power cycle is a highly pressurized system, certain amount of leakage flow is inevitable in the rotating turbo-machinery via seals. To simulate the CO 2 leak flow in a turbo-machinery with higher accuracy in the future, the real gas effect and friction factor will be considered for the CO 2 critical flow model. Moreover, experimentally obtained temperature data were somewhat different from the numerically obtained temperature due to the insufficient insulation and large thermal inertia of the CO 2 critical flow facility. Insulation in connecting pipes and the low-pressure tank will be added and additional tests will be conducted

  13. Project SAFE. Update of the SFR-1 safety assessment. Phase 1. Appendix A2: Scenarios

    International Nuclear Information System (INIS)

    Skagius, K.; Wiborgh, M.

    1998-01-01

    This appendix gives a short description of the scenario methodology adopted in the previous safety assessment of SFR. Since then new methodologies for developing structured descriptions of how processes and interactions between processes affect the evolution of a repository system. Two such methods are briefly described. These methods are very similar, but they differ in the way the system is graphically structured. One of the methods is based on Process Influence Diagrams, PID, and the other on Interaction matrices. It is proposed that the method based on Interaction matrices is used for the scenario work in project SAFE. The main reason for this is that the method already has been applied by SKB, which means that it will be possible to use already existing procedures and documentation systems. The proposed procedure involves the development of Interaction matrices for a defined Reference scenario and the use of these matrices to illustrate the effect of different Scenario initiating FEPs. The proposed procedure is described in this appendix

  14. Source terms; isolation and radiological consequences of carbon-14 waste in the Swedish SFR repository

    International Nuclear Information System (INIS)

    Hesboel, R.; Puigdomenech, I.; Evans, S.

    1990-01-01

    The source term, isolation capacity, and long-term radiological exposure of 14 C from the Swedish underground repository for low and intermediate level waste (SFR) is assessed. The prospective amount of 14 C in the repository is assumed to be 5 TBq. Spent ion exchange resins will be the dominant source of 14 C. The pore water in the concrete repository is expected to maintain a pH of >10.5 for a period of at least 10 6 y. The cement matrix of the repository will retain most of the 14 CO 3 2- initially present. Bacterial production of CO 2 and CH 4 from degradation of ion-exchange resins and bitumen may contribute to 14 C release to the biosphere. However, CH 4 contributes only to a small extent to the overall carbon loss from freshwater ecosystems. The individual doses to local and regional individuals peaked with 5x10 -3 and regional individuals peaked with 5x10 -3 and 8x10 -4 μSv y -1 respectively at about 2.4x10 4 years. A total leakage of 8.4 GBq of 14 C from the repository will cause a total collective dose commitment of 1.1 manSv or 130 manSv TBq -1 . (authors)

  15. Project SAFE. Modelling of long-term concrete degradation processes in the Swedish SFR repository

    International Nuclear Information System (INIS)

    Hoeglund, L.O.

    2001-04-01

    This study concerns the leaching of concrete barriers, in particular the silo construction, in the Swedish SFR repository for low and intermediate level radioactive waste. A conceptual model for the leaching of concrete in a saline groundwater has been proposed based on the increased understanding achieved from research studies presented in the literature. The conceptual model has been used to set up a numerical model for the complex chemical interactions between the cement minerals of the concrete with the groundwater. The calculations show that various chemical reactions are expected to occur in the concrete over time. Different cases have been calculated. The results show that the chemical conditions in the concrete barriers will maintain alkaline for long time. In the most exposed parts of the concrete a high degree of leaching can be expected during the considered 10,000 years, whereas only for the most unfavourable assumptions (initially fractured concrete with groundwater flow-through) the inner parts of the concrete will be degraded to any significant degree

  16. Development of the On-line Acoustic Leak Detection Tool for the SFR Steam Generator Protection

    International Nuclear Information System (INIS)

    Kim, Tae-Joon; Jeong, Ji-Young; Kim, Jong-Man; Kim, Byung-Ho; Kim, Seong-O

    2007-01-01

    The successful detection of a water/steam into a sodium leak in the SFR SG (steam generator) at an early phase of a leak origin depends on the fast response and sensitivity of a leak detection system. This intention of an acoustic leak detection system is stipulated by a key impossibility of a fast detecting of an intermediate leak by the present nominal systems such as the hydrogen meter. Subject of this study is to introduce the detection performance of an on-line acoustic leak detection tool discriminated by a back-propagation neural network with a preprocessing of the 1/m Octave band analysis, and to introduce the status of an on-line development being developed with the acoustic leak detection tool(S/W) in KAERI. For a performance test, it was used with the acoustic signals for a sodium-water reaction from the injected steam into water experiments in KAERI, the acoustic signals injected from the water into the sodium obtained in IPPE, and the background noise of the PFR superheater

  17. Improved intact soil-core carbon determination applying regression shrinkage and variable selection techniques to complete spectrum laser-induced breakdown spectroscopy (LIBS).

    Science.gov (United States)

    Bricklemyer, Ross S; Brown, David J; Turk, Philip J; Clegg, Sam M

    2013-10-01

    Laser-induced breakdown spectroscopy (LIBS) provides a potential method for rapid, in situ soil C measurement. In previous research on the application of LIBS to intact soil cores, we hypothesized that ultraviolet (UV) spectrum LIBS (200-300 nm) might not provide sufficient elemental information to reliably discriminate between soil organic C (SOC) and inorganic C (IC). In this study, using a custom complete spectrum (245-925 nm) core-scanning LIBS instrument, we analyzed 60 intact soil cores from six wheat fields. Predictive multi-response partial least squares (PLS2) models using full and reduced spectrum LIBS were compared for directly determining soil total C (TC), IC, and SOC. Two regression shrinkage and variable selection approaches, the least absolute shrinkage and selection operator (LASSO) and sparse multivariate regression with covariance estimation (MRCE), were tested for soil C predictions and the identification of wavelengths important for soil C prediction. Using complete spectrum LIBS for PLS2 modeling reduced the calibration standard error of prediction (SEP) 15 and 19% for TC and IC, respectively, compared to UV spectrum LIBS. The LASSO and MRCE approaches provided significantly improved calibration accuracy and reduced SEP 32-55% over UV spectrum PLS2 models. We conclude that (1) complete spectrum LIBS is superior to UV spectrum LIBS for predicting soil C for intact soil cores without pretreatment; (2) LASSO and MRCE approaches provide improved calibration prediction accuracy over PLS2 but require additional testing with increased soil and target analyte diversity; and (3) measurement errors associated with analyzing intact cores (e.g., sample density and surface roughness) require further study and quantification.

  18. Volatile Compound Profiling by HS-SPME/GC-MS-FID of a Core Olive Cultivar Collection as a Tool for Aroma Improvement of Virgin Olive Oil

    Directory of Open Access Journals (Sweden)

    Lourdes García-Vico

    2017-01-01

    Full Text Available Virgin olive oil (VOO is the only food product requiring official sensory analysis to be classified in commercial categories, in which the evaluation of the aroma plays a very important role. The selection of parents, with the aim of obtaining new cultivars with improved oil aroma, is of paramount importance in olive breeding programs. We have assessed the volatile fraction by headspace-solid-phase microextraction/gas chromatography-mass spectrometry-flame ionization detection (HS-SPME/GC-MS-FID and the deduced aroma properties of VOO from a core set of olive cultivars (Core-36 which possesses most of the genetic diversity found in the World Olive Germplasm Collection (IFAPA Alameda del Obispo located in Cordoba, Spain. The VOO volatile fractions of Core-36 cultivars display a high level of variability. It is mostly made of compounds produced from polyunsaturated fatty acids through the lipoxygenase pathway, which confirms to be a general characteristic of the olive species (Olea europaea L.. The main group of volatile compounds in the oils was six straight-chain carbon compounds derived from linolenic acid, some of them being the main contributors to the aroma of the olive oils according to their odor activity values (OAV. The high level of variability found for the volatile fraction of the oils from Core-36 and, therefore, for the aroma odor notes, suggest that this core set may be a very useful tool for the choice of optimal parents in olive breeding programs in order to raise new cultivars with improved VOO aroma.

  19. Volatile Compound Profiling by HS-SPME/GC-MS-FID of a Core Olive Cultivar Collection as a Tool for Aroma Improvement of Virgin Olive Oil.

    Science.gov (United States)

    García-Vico, Lourdes; Belaj, Angjelina; Sánchez-Ortiz, Araceli; Martínez-Rivas, José M; Pérez, Ana G; Sanz, Carlos

    2017-01-14

    Virgin olive oil (VOO) is the only food product requiring official sensory analysis to be classified in commercial categories, in which the evaluation of the aroma plays a very important role. The selection of parents, with the aim of obtaining new cultivars with improved oil aroma, is of paramount importance in olive breeding programs. We have assessed the volatile fraction by headspace-solid-phase microextraction/gas chromatography-mass spectrometry-flame ionization detection (HS-SPME/GC-MS-FID) and the deduced aroma properties of VOO from a core set of olive cultivars (Core-36) which possesses most of the genetic diversity found in the World Olive Germplasm Collection (IFAPA Alameda del Obispo) located in Cordoba, Spain. The VOO volatile fractions of Core-36 cultivars display a high level of variability. It is mostly made of compounds produced from polyunsaturated fatty acids through the lipoxygenase pathway, which confirms to be a general characteristic of the olive species ( Olea europaea L.). The main group of volatile compounds in the oils was six straight-chain carbon compounds derived from linolenic acid, some of them being the main contributors to the aroma of the olive oils according to their odor activity values (OAV). The high level of variability found for the volatile fraction of the oils from Core-36 and, therefore, for the aroma odor notes, suggest that this core set may be a very useful tool for the choice of optimal parents in olive breeding programs in order to raise new cultivars with improved VOO aroma.

  20. Core Hunter 3: flexible core subset selection.

    Science.gov (United States)

    De Beukelaer, Herman; Davenport, Guy F; Fack, Veerle

    2018-05-31

    Core collections provide genebank curators and plant breeders a way to reduce size of their collections and populations, while minimizing impact on genetic diversity and allele frequency. Many methods have been proposed to generate core collections, often using distance metrics to quantify the similarity of two accessions, based on genetic marker data or phenotypic traits. Core Hunter is a multi-purpose core subset selection tool that uses local search algorithms to generate subsets relying on one or more metrics, including several distance metrics and allelic richness. In version 3 of Core Hunter (CH3) we have incorporated two new, improved methods for summarizing distances to quantify diversity or representativeness of the core collection. A comparison of CH3 and Core Hunter 2 (CH2) showed that these new metrics can be effectively optimized with less complex algorithms, as compared to those used in CH2. CH3 is more effective at maximizing the improved diversity metric than CH2, still ensures a high average and minimum distance, and is faster for large datasets. Using CH3, a simple stochastic hill-climber is able to find highly diverse core collections, and the more advanced parallel tempering algorithm further increases the quality of the core and further reduces variability across independent samples. We also evaluate the ability of CH3 to simultaneously maximize diversity, and either representativeness or allelic richness, and compare the results with those of the GDOpt and SimEli methods. CH3 can sample equally representative cores as GDOpt, which was specifically designed for this purpose, and is able to construct cores that are simultaneously more diverse, and either are more representative or have higher allelic richness, than those obtained by SimEli. In version 3, Core Hunter has been updated to include two new core subset selection metrics that construct cores for representativeness or diversity, with improved performance. It combines and outperforms the

  1. Use of Added Sugars Instead of Total Sugars May Improve the Capacity of the Health Star Rating System to Discriminate between Core and Discretionary Foods.

    Science.gov (United States)

    Menday, Hannah; Neal, Bruce; Wu, Jason H Y; Crino, Michelle; Baines, Surinder; Petersen, Kristina S

    2017-12-01

    The Australian Government has introduced a voluntary front-of-package labeling system that includes total sugar in the calculation. Our aim was to determine the effect of substituting added sugars for total sugars when calculating Health Star Ratings (HSR) and identify whether use of added sugars improves the capacity to distinguish between core and discretionary food products. This study included packaged food and beverage products available in Australian supermarkets (n=3,610). The product categories included in the analyses were breakfast cereals (n=513), fruit (n=571), milk (n=309), non-alcoholic beverages (n=1,040), vegetables (n=787), and yogurt (n=390). Added sugar values were estimated for each product using a validated method. HSRs were then estimated for every product according to the established method using total sugar, and then by substituting added sugar for total sugar. The scoring system was not modified when added sugar was used in place of total sugar in the HSR calculation. Products were classified as core or discretionary based on the Australian Dietary Guidelines. To investigate whether use of added sugar in the HSR algorithm improved the distinction between core and discretionary products as defined by the Australian Dietary Guidelines, the proportion of core products that received an HSR of ≥3.5 stars and the proportion of discretionary products that received an HSR of added sugars were determined. There were 2,263 core and 1,347 discretionary foods; 1,684 of 3,610 (47%) products contained added sugar (median 8.4 g/100 g, interquartile range=5.0 to 12.2 g). When the HSR was calculated with added sugar instead of total sugar, an additional 166 (7.3%) core products received an HSR of ≥3.5 stars and 103 (7.6%) discretionary products received a rating of ≥3.5 stars. The odds of correctly identifying a product as core vs discretionary were increased by 61% (odds ratio 1.61, 95% CI 1.26 to 2.06; Padded compared to total sugars. In the six

  2. Standardization of the methodology used for fuel pressure drop evaluation to improve hydraulic calculation of heterogeneous cores

    International Nuclear Information System (INIS)

    Le Borgne, E.; Mattei, A.

    1994-01-01

    Continuous searching for safer and more efficient fuel, and diversification of fuel supply have as a consequence a possible change in the characteristics of the fuel assemblies used in nuclear reactors. By partially refueling cores with new assemblies, nuclear power plant operators are confronted with the problem of heterogeneous cores. The complexity of the problem increases as products diversify in isotopic concentration, types of alloy, size and shape of structure components. This document will focus strictly on the differences in hydraulic resistance related to the modifications in grid structures having no effect on DNB correlations. Although this is an extremely simplified approach to the problem, establishing data to evaluate the hydraulic compatibility between two different assemblies can be difficult, and if not controlled closely, can lead to false conclusions that may affect the operation and safety of the reactor. (authors). 2 figs

  3. Conceptual study of axial offset fluctuations upon stepwise power changes in a thorium–plutonium core to improve load-following conditions

    International Nuclear Information System (INIS)

    Lau, Cheuk Wah; Dykin, Victor; Nylén, Henrik; Björk, Klara Insulander; Sandberg, Urban

    2014-01-01

    Highlights: • Thorium–plutonium mixed oxide to improve nuclear reactors load-following capability. • SIMULATE-3 was the main calculation tool. • The Ringhals-3 PWR unit in Sweden was used as a reference. • Lower xenon poisoning and shorter reactor dead time. - Abstract: The increased share of renewable energy, such as wind and solar power, will increase the demand for load-following power sources, and nuclear reactors could be one option. However, during rapid load-following events, traditional UOX cores could be restricted by the volatile oscillation of the power distribution. Therefore, a conceptual study on stability properties of Th-MOX PWR concerning axial offset power excursion during load-following events are investigated and discussed. The study is performed in SIMULATE-3 for a realistic PWR core (Ringhals-3) at the end of cycle, where the largest amplitude of the axial offset oscillations is expected. It is shown that the Th-MOX core possesses much better stability characteristics and shorter reactor dead time compared with a traditional UOX core, and the main reasons are the lower sensitivity to perturbations in the neutron spectrum, lower xenon poisoning and lower thermal neutron flux

  4. Development of an optimization technique of CETOP-D inlet flow factor for reactor core thermal margin improvement

    International Nuclear Information System (INIS)

    Hong, Sung Duk; Im, Jong Sun; Yoo, Yun Jong; Kwon, Jung Taek; Park, Jong Ryool

    1995-01-01

    The recent ABB/CE(Asea Brown Boveri Combustion Engineering) type pressurized water reactors have the on-line monitoring system, i.e., the COLSS(core operating limit supervisory system), to prevent the specified acceptable fuel design limits from being violated during normal operation and anticipated operational occurrences. One of the main functions of COLSS is the on-line monitoring of the DNB(departure from nucleate boiling) overpower margin by calculating the MDNBR(minimum DNB ratio) for the measured operating condition at every second. The CETOP-D model, used in the MDNBR calculation of COLSS, is benchmarked conservatively against the TORC model using an inlet flow factor of hot assembly in CETOP-D as an adjustment factor for TORC. In this study, a technique to optimize the CETOP-D inlet flow factor has been developed by eliminating the excessive conservatism in the ABB/CE's. A correlation is introduced to account for the actual variation of the CETOP-D inlet flow factor within the core operating limits. This technique was applied to the core operating range of the Yonggwang Units 3 and 4 Cycle 1, which results in the increase of 2% in the DNB overpower margin at the normal operating condition, compared with that from the ABB/CE method. 7 figs., 2 tabs., 10 refs. (Author)

  5. Can concurrent core biopsy and fine needle aspiration biopsy improve the false negative rate of sonographically detectable breast lesions?

    Directory of Open Access Journals (Sweden)

    Chang Tsai-Wang

    2010-07-01

    Full Text Available Abstract Background The aims of this study were to determine the accuracy of concurrent core needle biopsy (CNB and fine needle aspiration biopsy (FNAB for breast lesions and to estimate the false-negative rate using the two methods combined. Methods Over a seven-year period, 2053 patients with sonographically detectable breast lesions underwent concurrent ultrasound-guided CNB and FNAB. The sonographic and histopathological findings were classified into four categories: benign, indeterminate, suspicious, and malignant. The histopathological findings were compared with the definitive excision pathology results. Patients with benign core biopsies underwent a detailed review to determine the false-negative rate. The correlations between the ultrasonography, FNAB, and CNB were determined. Results Eight hundred eighty patients were diagnosed with malignant disease, and of these, 23 (2.5% diagnoses were found to be false-negative after core biopsy. After an intensive review of discordant FNAB results, the final false-negative rate was reduced to 1.1% (p-value = 0.025. The kappa coefficients for correlations between methods were 0.304 (p-value p-value p-value Conclusions Concurrent CNB and FNAB under ultrasound guidance can provide accurate preoperative diagnosis of breast lesions and provide important information for appropriate treatment. Identification of discordant results using careful radiological-histopathological correlation can reduce the false-negative rate.

  6. A Review of Models for Dose Assessment Employed by SKB in the Renewed Safety Assessment for SFR 1

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, George [Imperial College of Science Technology and Medicine (United Kingdom)

    2002-09-01

    This document provides a critical review, on behalf of SSI, of the models employed by the Swedish Nuclear Fuel and Waste Management Co (SKB) for dose assessment in the renewed safety assessment for the final repository for radioactive operational waste (SFR 1) in Forsmark, Sweden. The main objective of the review is to examine the models used by SKB for radiological dose assessment in a series of evolving biotopes in the vicinity of the Forsmark repository within a time frame beginning in 3000 AD and extending beyond 7500 AD. Five biosphere models (for coasts, lakes, agriculture, mires and wells) are described in Report TR-01-04. The principal consideration of the review is to determine whether these models are fit for the purpose of dose evaluation over the time frames involved and in the evolving sequence of biotopes specified. As well as providing general observations and comments on the modelling approach taken, six specific questions are addressed, as follows. Are the assumptions underlying the models justifiable? Are all reasonably foreseeable environmental processes considered? Has parameter uncertainty been sufficiently and reasonably addressed? Have sufficient models been used to address all reasonably foreseeable biotopes? Are the transitions between biotopes modelled adequately (specifically, are initial conditions for developing biotopes adequately specified by calculations for subsiding biotopes)? Have all critical radionuclides been identified? It is concluded that, in general, the assumptions underlying most of the models are justifiable. The exceptions are a) the rather simplistic approach taken in the Coastal Model and b) the lack of consideration of wild foods and age-dependence when calculating exposures of humans to radionuclides via dietary pathways. Most foreseeable processes appear to have been accounted for within the constraints of the models used, although it is recommended that attention be paid to future climate states when considering

  7. Proteomics Core

    Data.gov (United States)

    Federal Laboratory Consortium — Proteomics Core is the central resource for mass spectrometry based proteomics within the NHLBI. The Core staff help collaborators design proteomics experiments in a...

  8. Investigation of the Stress Intensity Limits of ASME Section III Div.5 for Structure Design Criteria of SFR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Jin-Yup; Kim, Hyung-Kyu; Cheon, Jin-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    These affect the mechanical design of the fuel assembly components. And thus, appropriate structural design criteria should also be chosen to incorporate the specific design conditions of the SFR fuel assemblies. Among them, the temperature is one of the most crucial conditions to be concerned because the sodium coolant temperature is normally more than 500ºC which is much higher than that of the LWR (< 350ºC). This implies that a thermal creep should be significantly considered in the SFR fuel assembly mechanical design. In addition to the high temperature condition, an irradiation swelling is also an important behavior that the SFR fuel assembly material should accommodate. To incorporate the temperature and irradiation impacts, the material of the fuel assembly components is presently determined to be made of HT-9, the ferriticmartensitic steel. In this paper, the ASME Sec. III Div. 5 (referred to as ‘Div. 5’ hereinafter), which was developed for a ‘high temperature reactor’, is considered as one of the structural design criteria for the mechanical design of SFR fuel assemblies. In this paper, the stress intensity limits, S{sub m} and S{sub t} of HT-9 were built for the structural criteria of an SFR fuel assembly. S{sub m} is obtained from the ultimate strength. As for S{sub t}, it is more complicated because of its dependency of time duration in addition to temperature. Following the definition of S{sub mt}, the method in the ASME Sec. III Div. 1, Subsec. NH was consulted. We found that the Sm is adopted as S{sub mt} under the temperature about 470ºC which is relatively low temperature range and over 470ºC with relatively short time duration as 1000 hours. And the S{sub t} is adopted as Smt at over 470ºC and long time duration over 34800 hours, and over 520ºC and 10{sup 4} hours too. And at over 570ºC and 1000 hours, and at over 630ºC and 100 hours, S{sub t} is also adopted for S{sub mt}.

  9. System studies in PA: Development of process influence diagram (PID) for SFR-1 repository near-field + far-field

    International Nuclear Information System (INIS)

    Stenhouse, M.J.; Miller, W.M.; Chapman, N.A.

    2001-05-01

    Scenario development is a key component of the performance assessment (PA) process for radioactive waste disposal, the primary objective being to ensure that all relevant factors associated with the future evolution of the repository system are properly considered in PA. As part of scenario development, a list of features, events and processes (FEPs) are identified and assembled, representing the Process System, with interactions/influences between FEPs incorporated in a Process Influence Diagram (PID). This report documents the technical work conducted between 1997 and the end of 1999 under the Systems Studies Project. The overall objective of this project has been the construction of a PID for the SFR-1 repository (final repository for reactor waste), this PID being the first stage in the identification of scenarios to describe future evolution of this repository. The PIDs discussed in this report have been created using two software applications: existing commercial software (Business Modeller, Infotool AB. Stockholm, Sweden) and, more recently, a newly developed software tool SPARTA (Enviros QuantiSci, Henley, U.K.). Although the focus of this report is on the application of SPARTA to PID development, it is important to document the work carried out prior to SPARTA being available, in order to provide a complete record of the entire SFR-1 PID development effort as well as preserving the context of the multi-year project. Following a description of the different disposal sections of the SFR-1 and the various near-field barriers, the sequential development (i.e. near-field of Silo, BMA, BLA, BTF sections; far-field; integrated near-field + far-field) of the PID for SFR-1 repository system using Business Modeller is described. Owing to the complexity of the repository, in terms of number of both different disposal sections (Silo, BLA, BMA, BTF) and barriers associated with each section, the two-dimensional (2D) PID created for SFR-1 using Business Modeller is

  10. System studies in PA: Development of process influence diagram (PID) for SFR-1 repository near-field + far-field

    Energy Technology Data Exchange (ETDEWEB)

    Stenhouse, M.J. [Monitor Scientific, LLC, Denver, CO (United States); Miller, W.M.; Chapman, N.A. [QuantiSci Ltd., Melton Mowbray (United Kingdom)

    2001-05-01

    Scenario development is a key component of the performance assessment (PA) process for radioactive waste disposal, the primary objective being to ensure that all relevant factors associated with the future evolution of the repository system are properly considered in PA. As part of scenario development, a list of features, events and processes (FEPs) are identified and assembled, representing the Process System, with interactions/influences between FEPs incorporated in a Process Influence Diagram (PID). This report documents the technical work conducted between 1997 and the end of 1999 under the Systems Studies Project. The overall objective of this project has been the construction of a PID for the SFR-1 repository (final repository for reactor waste), this PID being the first stage in the identification of scenarios to describe future evolution of this repository. The PIDs discussed in this report have been created using two software applications: existing commercial software (Business Modeller, Infotool AB. Stockholm, Sweden) and, more recently, a newly developed software tool SPARTA (Enviros QuantiSci, Henley, U.K.). Although the focus of this report is on the application of SPARTA to PID development, it is important to document the work carried out prior to SPARTA being available, in order to provide a complete record of the entire SFR-1 PID development effort as well as preserving the context of the multi-year project. Following a description of the different disposal sections of the SFR-1 and the various near-field barriers, the sequential development (i.e. near-field of Silo, BMA, BLA, BTF sections; far-field; integrated near-field + far-field) of the PID for SFR-1 repository system using Business Modeller is described. Owing to the complexity of the repository, in terms of number of both different disposal sections (Silo, BLA, BMA, BTF) and barriers associated with each section, the two-dimensional (2D) PID created for SFR-1 using Business Modeller is

  11. Summary final report: Contract between the Japan atomic power company and the U.S. Department of Energy Improvement of core safety - study on GEM (III)

    International Nuclear Information System (INIS)

    Burke, T.M.; Lucoff, D.M.

    1997-01-01

    This report provides a summary of activities associated with the technical exchange between representatives of the Japan Atomic Power Company (JAPC) and the United States Department of Energy (DOE) regarding the development and testing of Gas Expansion Modules (GEM) at the Fast Flux Test Facility (FFTF). Issuance of this report completes the scope of work defined in the original contract between JAPC and DOE titled ''Study on Improvement of Core Safety - Study on GEM (III).'' Negotiations related to potential modification of the contract are in progress. Under the proposed contract modification, DOE would provide an additional report documenting FFTF pump start tests with GEMs and answer additional JAPC questions related to core safety with and without GEMs

  12. Summary final report: Contract between the Japan atomic power company and the U.S. Department of Energy Improvement of core safety - study on GEM (III)

    Energy Technology Data Exchange (ETDEWEB)

    Burke, T.M.; Lucoff, D.M.

    1997-03-18

    This report provides a summary of activities associated with the technical exchange between representatives of the Japan Atomic Power Company (JAPC) and the United States Department of Energy (DOE) regarding the development and testing of Gas Expansion Modules (GEM) at the Fast Flux Test Facility (FFTF). Issuance of this report completes the scope of work defined in the original contract between JAPC and DOE titled ''Study on Improvement of Core Safety - Study on GEM (III).'' Negotiations related to potential modification of the contract are in progress. Under the proposed contract modification, DOE would provide an additional report documenting FFTF pump start tests with GEMs and answer additional JAPC questions related to core safety with and without GEMs.

  13. Models for dose assessments. Models adapted to the SFR-area, Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Karlsson, Sara; Bergstroem, U.; Meili, M. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    2001-10-01

    This report presents a model system created to be used to predict dose rates to the most exposed individuals in case of a long-term release of radionuclides from the Final repository for radioactive operational waste (SFR) in Forsmark, Sweden. The system accounts for an underground point source emitting radionuclides to the biosphere,their transport and distribution in various ecosystem types, their uptake by various biota, and calculation of doses to man from a multitude of exposure pathways. Long-term aspects include the consideration of processes at geological time scales, such as land uplift and conversion of marine sediments to arable land. Model parameters are whenever possible based on local conditions and recent literature. The system has been used for simulations based on geospheric releases varying over time of a mixture of radionuclides. Here, the models have been subjected to various test scenarios, including different radionuclide entry points and sorption properties. Radionuclides released from SFR are efficiently diluted to low concentrations in the water of the coastal model. A large fraction of the nuclides leave the Model Area quickly as a consequence of the rapid water turnover. The total amount of radionuclides in water is about the same independent of particle affinity (K{sub d} ), and at most some percents of the amounts retained in the sediments for some time. The latter is also true for the lake model when releases of radionuclides to the water is simulated. The role of sediments as a radionuclide source seems of minor importance in lakes at least for long-term radiation doses. Modelling the sediments as a source of radionuclides most of the 'low K{sub d} radionuclides' will leave the lake whereas the 'high K{sub d} nuclides' are still present within the deeper sediments after 1 000 years. The amount of 'low K{sub d} radionuclides' present in the water and on suspended matter are about 8x10{sup -5} of the

  14. Generation IV SFR Nuclear Reactors: Under-Sodium Repair for ASTRID

    International Nuclear Information System (INIS)

    Baque, F.; Chagnot, C.; Bruguiere, L.; Augem, J.M.; Delalande, V.; Sibilo, J.

    2013-06-01

    For non-removable components of the future ASTRID prototype, repair operations will be performed in a gas environment. If the faulty area is located under the sodium free level, the gas-tight system will have to contain the inspection and repair tools and to protect them from the surrounding liquid sodium. Concerning repair tools, the unique laser tool has been selected for future SFRs: the repair scenario for in-sodium structures will first involve removing the sodium (after bulk draining), then machining and finally welding. Concerning conventional tools (brush or gas blower for sodium removal, milling machine for machining and TIG for welding for which its feasibility was demonstrated in the 1990's) are still considered as a back-up solution. In-pile examination or repair requires robotic carriers. These carriers have to be compatible with the sodium environment: either in the cover-gas plenum or in gas after sodium draining, or even under liquid sodium. This R and D programme has been divided into nine parts in order to provide an overall design of the required robotic carriers and to develop technological solutions for their components: detailed definition for SFR carrier needs (access to internal structures, possible defects to be detected/repaired), definition and specifications of carrier architecture (depending on inspection and repair scenarios), in-sodium leak-tightness of carrier components, carrier material compatibility with sodium, temperature resistance (200 deg. C), irradiation resistance (depending on the location of the main vessel), gas-tight bell for operations under liquid sodium, carrier positioning control in liquid sodium, development, validation and qualification of technological solutions, for future SFRs, and worldwide benchmark regarding the previous areas of investigation. (authors)

  15. Site investigation SFR. Boremap mapping of percussion drilled borehole HFR106

    Energy Technology Data Exchange (ETDEWEB)

    Winell, Sofia (Geosigma AB (Sweden))

    2010-06-15

    This report presents the result from the Boremap mapping of the percussion drilled borehole HFR106, which is drilled from an islet located ca 220 m southeast of the pier above SFR. The purpose of the location and orientation of the borehole is to investigate the possible occurrence of gently dipping, water-bearing structures in the area. HFR106 has a length of 190.4 m and oriented 269.4 deg/-60.9 deg. The mapping is based on the borehole image (BIPS), investigation of drill cuttings and generalized, as well as more detailed geophysical logs. The dominating rock type, which occupies 68% of HFR106, is fine- to medium-grained, pinkish grey metagranite-granodiorite (rock code 101057) mapped as foliated with a medium to strong intensity. Pegmatite to pegmatitic granite (rock code 101061) occupies 29% of the borehole. Subordinate rock types are felsic to intermediate meta volcanic rock (rock code 103076) and fine- to medium-grained granite (rock code 111058). Rock occurrences (rock types < 1 m in length) occupy about 16% of the mapped interval, of which half is veins, dykes and unspecified occurrences of pegmatite and pegmatitic granite. Only 5.5% of HFR106 is inferred to be altered, mainly oxidation in two intervals with an increased fracture frequency. A total number of 845 fractures are registered in HFR106. Of these are 64 interpreted as open with a certain aperture, 230 open with a possible aperture, and 551 sealed. This result in the following fracture frequencies: 1.6 open fractures/m and 3.0 sealed fractures/m. Three fracture sets of open and sealed fractures with the orientations 290 deg/70 deg, 150 deg/85 deg and 200 deg/85 deg can be distinguished in HFR106. The fracture frequency is generally higher in the second half of the borehole, and particularly in the interval 176-187.4 m.

  16. Site investigation SFR. Hydro Monitoring Program. Report for May 2008-August 2009

    Energy Technology Data Exchange (ETDEWEB)

    Nyberg, Goeran; Wass, Eva [Geosigma AB, Uppsala (Sweden)

    2009-10-15

    SKB conducts bedrock investigations for a future extension of the final repository for low- and medium-level radioactive waste (SFR) at Forsmark within the Oesthammar municipality. As a part of this investigation, hydrogeological monitoring is performed. The objectives of the groundwater monitoring are, in a short-term perspective, to measure pressure responses during drilling, pumping and interference tests and also, in a long-term perspective, to create time series in order to increase the knowledge of the hydraulic conditions. Data presented in this report are collected during the period of May 2008 until August 2009 and include groundwater levels in surface boreholes and groundwater pressure in boreholes situated in the tunnel. The data collecting system in HMS (Hydro Monitoring System) consists of a measurement station (computer) that communicates with and collects data from a number of data loggers. The computer is connected to the SKB Ethernet LAN. All data are collected by means of different transducers connected to different types of data loggers: Minitroll, LevelTroll and Datataker. In order to calibrate registrations from the data loggers, manual levelling of all surface borehole sections is made, normally once every month. The logger data are converted to water levels using calibration constants. All collected data are quality checked once every three or four months. During this work, obviously erroneous data are omitted and calibration constants are corrected so that the monitored data comply with the manual levelling. At these occasions the status of the equipment is also checked and service might be initiated. Diagrams of groundwater levels and groundwater pressure for the period of May 2008-August 2009 (one data point per section and twenty-four hours) are presented in Appendix 2. The original data are stored in the primary data base Sicada. The data in this data base may then be used for further analysis. There are no nonconformities with respect

  17. Site investigation SFR. Hydro Monitoring Program. Report for September 2009 - August 2010

    Energy Technology Data Exchange (ETDEWEB)

    Nyberg, Goeran; Wass, Eva [Geosigma AB, Uppsala (Sweden)

    2010-11-15

    SKB is conducting bedrock investigations for a future extension of the final repository for low- and medium-level radioactive waste (SFR) at Forsmark within the Oesthammar municipality. As a part of this investigation, hydrogeological monitoring is performed. The objectives of the groundwater monitoring are, in a short-term perspective, to measure pressure responses during drilling, pumping and interference tests and also, in a long-term perspective, to obtain time series in order to increase the knowledge of the hydraulic conditions. Data presented in this report are collected during the period of September 2009 until August 2010 and include groundwater levels in surface boreholes and groundwater pressure in boreholes situated in the tunnel. The data collecting system in HMS (Hydro Monitoring System) consists of a measurement station (computer) that communicates with and collects data from a number of data loggers. The computer is connected to the SKB Ethernet LAN. All data are collected by means of different types of transducers connected to different types of data loggers: Minitroll, LevelTroll and Datataker. In order to calibrate registrations from the data loggers, manual levelling of the groundwater table of all surface borehole sections is made, usually once every month. The logger data are converted to water levels using calibration constants. All collected data are quality checked, generally once every four months. During this work, obviously erroneous data are omitted and calibration constants are corrected so that the monitored data comply with the manual levelling. At these occasions the status of the equipment is also checked and service might be initiated. Diagrams of groundwater levels and groundwater pressure for the period of September 2009 - August 2010 (one data point per section and 24 hours) are presented in Appendix 2. The original data are stored in the primary data base Sicada. The data in this data base may then be used for further analysis

  18. Methodology for a thermal analysis of a proposed SFR transport cask with the thermal code SYRTHES

    International Nuclear Information System (INIS)

    Peniguel, C.; Rupp, I.; Schneider, J. P.

    2010-01-01

    Fast reactors with liquid metal coolant have received a renewed interest owing to the need of a more efficient usage of the primary uranium resources, and they are one of the proposal for the next Generation IV. In the framework of the 2006 French law on sustainable management of radioactive materials and waste, an evaluation of the industrial perspectives of minor actinides transmutation advantages and drawbacks in Generation IV fast spectrum reactors system is requested for 2012. The CEA is in charge of studying the global problem, but on some aspects, EDF is interested to do its own exploratory studies. Among other points, transport is seen as important for the nuclear industry, to link points of production and treatment. Nuclear fuel is generally transported in thick walled rail or truck casks. These packages are designed to provide confinement, shielding and criticality protection during normal and severe transport conditions. Heat generated within the fuel (and a contribution of solar heating) makes the package becoming quite hot, but one must demonstrate that the cladding temperature does not exceed a long term temperature limit during normal transport. This paper presents a thermal study done on a package in which 9 SFR assemblies are included. Each of them is of hexagonal shape and contains 271 fuel pins. The approach followed for these calculations is to rely on an explicit representation of all pins. For these calculations a 2D analysis is performed thanks to the thermal code SYRTHES. Conduction is solved thanks to a finite element method, while thermal radiation is handled through a radiosity approach. The main aim of this paper is to present a possible numerical methodology to handle the thermal problem. (authors)

  19. Site investigation SFR. Boremap mapping of percussion drilled borehole HFR106

    International Nuclear Information System (INIS)

    Winell, Sofia

    2010-06-01

    This report presents the result from the Boremap mapping of the percussion drilled borehole HFR106, which is drilled from an islet located ca 220 m southeast of the pier above SFR. The purpose of the location and orientation of the borehole is to investigate the possible occurrence of gently dipping, water-bearing structures in the area. HFR106 has a length of 190.4 m and oriented 269.4 deg/-60.9 deg. The mapping is based on the borehole image (BIPS), investigation of drill cuttings and generalized, as well as more detailed geophysical logs. The dominating rock type, which occupies 68% of HFR106, is fine- to medium-grained, pinkish grey metagranite-granodiorite (rock code 101057) mapped as foliated with a medium to strong intensity. Pegmatite to pegmatitic granite (rock code 101061) occupies 29% of the borehole. Subordinate rock types are felsic to intermediate meta volcanic rock (rock code 103076) and fine- to medium-grained granite (rock code 111058). Rock occurrences (rock types < 1 m in length) occupy about 16% of the mapped interval, of which half is veins, dykes and unspecified occurrences of pegmatite and pegmatitic granite. Only 5.5% of HFR106 is inferred to be altered, mainly oxidation in two intervals with an increased fracture frequency. A total number of 845 fractures are registered in HFR106. Of these are 64 interpreted as open with a certain aperture, 230 open with a possible aperture, and 551 sealed. This result in the following fracture frequencies: 1.6 open fractures/m and 3.0 sealed fractures/m. Three fracture sets of open and sealed fractures with the orientations 290 deg/70 deg, 150 deg/85 deg and 200 deg/85 deg can be distinguished in HFR106. The fracture frequency is generally higher in the second half of the borehole, and particularly in the interval 176-187.4 m

  20. Supercritical CO2 Brayton Cycle Energy Conversion System Coupled with SFR

    International Nuclear Information System (INIS)

    Cha, Jae Eun; Kim, S. O.; Seong, S. H.; Eoh, J. H.; Lee, T. H.; Choi, S. K.; Han, J. W.; Bae, S. W.

    2008-12-01

    This report contains the description of the S-CO 2 Brayton cycle coupled to KALIMER-600 as an alternative energy conversion system. For a system development, a computer code was developed to calculate heat balance of normal operation condition. Based on the computer code, the S-CO 2 Brayton cycle energy conversion system was constructed for the KALIMER-600. Computer codes were developed to analysis for the S-CO 2 turbomachinery. Based on the design codes, the design parameters were prepared to configure the KALIMER-600 S-CO 2 turbomachinery models. A one-dimensional analysis computer code was developed to evaluate the performance of the previous PCHE heat exchangers and a design data for the typical type PCHE was produced. In parallel with the PCHE-type heat exchanger design, an airfoil shape fin PCHE heat exchanger was newly designed. The new design concept was evaluated by three-dimensional CFD analyses. Possible control schemes for power control in the KALIMER-600 S-CO 2 Brayton cycle were investigated by using the MARS code. The MMS-LMR code was also developed to analyze the transient phenomena in a SFR with a supercritical CO 2 Brayton cycle to develop the control logic. Simple power reduction and recovery event was selected and analyzed for the transient calculation. For the evaluation of Na-CO 2 boundary failure event, a computer was developed to simulate the complex thermodynamic behaviors coupled with the chemical reaction between liquid sodium and CO 2 gas. The long term behavior of a Na-CO 2 boundary failure event and its consequences which lead to a system pressure transient were evaluated

  1. Uniform Luminous Perovskite Nanofibers with Color-Tunability and Improved Stability Prepared by One-Step Core/Shell Electrospinning.

    Science.gov (United States)

    Tsai, Ping-Chun; Chen, Jung-Yao; Ercan, Ender; Chueh, Chu-Chen; Tung, Shih-Huang; Chen, Wen-Chang

    2018-04-30

    A one-step core/shell electrospinning technique is exploited to fabricate uniform luminous perovskite-based nanofibers, wherein the perovskite and the polymer are respectively employed in the core and the outer shell. Such a coaxial electrospinning technique enables the in situ formation of perovskite nanocrystals, exempting the needs of presynthesis of perovskite quantum dots or post-treatments. It is demonstrated that not only the luminous electrospun nanofibers can possess color-tunability by simply tuning the perovskite composition, but also the grain size of the formed perovskite nanocrystals is largely affected by the perovskite precursor stoichiometry and the polymer solution concentration. Consequently, the optimized perovskite electrospun nanofiber yields a high photoluminescence quantum yield of 30.9%, significantly surpassing the value of its thin-film counterpart. Moreover, owing to the hydrophobic characteristic of shell polymer, the prepared perovskite nanofiber is endowed with a high resistance to air and water. Its photoluminescence intensity remains constant while stored under ambient environment with a relative humidity of 85% over a month and retains intensity higher than 50% of its initial intensity while immersed in water for 48 h. More intriguingly, a white light-emitting perovskite-based nanofiber is successfully fabricated by pairing the orange light-emitting compositional perovskite with a blue light-emitting conjugated polymer. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  2. An investigation of sodium–CO{sub 2} interaction byproduct cleaning agent for SFR coupled with S-CO{sub 2} Brayton cycle

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hwa-Young, E-mail: jhy0523@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Division of SFR NSSS System Design, Korea Atomic Energy Research Institute, 111, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Wi, Myung-Hwan, E-mail: mhwi@kaeri.re.kr [Division of SFR NSSS System Design, Korea Atomic Energy Research Institute, 111, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Ahn, Hong Joo, E-mail: ahjoo@kaeri.re.kr [Division of Nuclear Chemistry Research, Korea Atomic Energy Research Institute, 111, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2016-02-15

    Highlights: • Study on cleaning agent was conducted to remove Na–CO{sub 2} interaction byproducts. • Screening criteria to select candidate substances as cleaning agents were suggested. • The mixtures of Na{sub 2}CO{sub 3} with NaBrO{sub 3}, NaClO{sub 3}, or NaBF{sub 4} were thermally analyzed with the TG/DTA studies. • Three candidate substances decomposed before 600 °C and did not react with Na{sub 2}CO{sub 3}. - Abstract: One of the promising future nuclear energy systems, the Sodium-cooled Fast Reactor (SFR) has been actively developed internationally. Recently, to improve safety and economics of a SFR further, coupling supercritical CO{sub 2} power cycle was suggested. However, there can be a chemical reaction between sodium and CO{sub 2} at high temperature (more than 400 °C) when the pressure boundary fails in a sodium–CO{sub 2} heat exchanger. To ensure the performance of such a system, it is important to employ a cleaning agent to recover the system back to normal condition after the reaction. When sodium and CO{sub 2} react, solid and gaseous reaction products such as sodium carbonate (Na{sub 2}CO{sub 3}) and carbon monoxide (CO) appear. Since most of solid reaction products are hard and can deteriorate system performance, quick removal of solid reaction products becomes very important for economic performance of the system. Thus, the authors propose the conceptual method to remove the byproducts with a chemical reaction at high temperature. The chemical reaction will take place between the reaction byproducts and a cleaning agent while the cleaning agent is inert with sodium. Thus, various sodium-based compounds were first investigated and three candidate substances satisfying several criteria were selected; sodium bromate (NaBrO{sub 3}), sodium chlorate (NaClO{sub 3}), and sodium tetrafluoroborate (NaBF{sub 4}). The selected substances were thermally analyzed with the TG/DTA studies. Unfortunately, it was revealed that all candidate

  3. Improving the {sup 210}Pb-chronology of Pb deposition in peat cores from Chao de Lamoso (NW Spain)

    Energy Technology Data Exchange (ETDEWEB)

    Olid, Carolina, E-mail: carolina.olid@emg.umu.se [Department of Ecology and Environmental Science, Umeå University, SE-901 87 Umeå (Sweden); Departament de Física, Universitat Autònoma de Barcelona, E-08193 Bellaterra (Spain); Garcia-Orellana, Jordi, E-mail: jordi.garcia@uab.cat [Departament de Física, Universitat Autònoma de Barcelona, E-08193 Bellaterra (Spain); Institut de Ciència i Tecnologia Ambientals (ICTA), Universitat Autònoma de Barcelona, E-08193 Bellaterra (Spain); Masqué, Pere, E-mail: pere.masque@uab.cat [Departament de Física, Universitat Autònoma de Barcelona, E-08193 Bellaterra (Spain); Institut de Ciència i Tecnologia Ambientals (ICTA), Universitat Autònoma de Barcelona, E-08193 Bellaterra (Spain); Cortizas, Antonio Martínez, E-mail: antonio.martinez.cortizas@usc.es [Departamento de Edafoloxía e Química Agrícola, Universidade de Santiago de Compostela, E-15782 Santiago de Compostela (Spain); and others

    2013-01-15

    The natural radionuclide {sup 210}Pb is commonly used to establish accurate and precise chronologies for the recent (past 100–150 years) layers of peat deposits. The most widely used {sup 210}Pb-dating model, Constant Rate of Supply (CRS), was applied using data from three peat cores from Chao de Lamoso, an ombrotrophic mire in Galicia (NW Spain). On the basis of the CRS-chronologies, maximum Pb concentrations and enrichment factors (EFs) occurred in the 1960s and late 1970s, consistent with the historical use of Pb. However, maximum Pb fluxes were dated in the 1940s and the late 1960s, 10 to 20 years earlier. Principal component analysis (PCA) showed that, although the {sup 210}Pb distribution was mainly (74%) controlled by radioactive decay, about 20% of the {sup 210}Pb flux variability was associated with atmospheric metal pollution, suggesting an extra {sup 210}Pb supply source and thus invalidating the main assumption of the CRS model. When the CRS-ages were recalculated after correcting for the extra input from the {sup 210}Pb inventory of the uppermost peat layers of each core, Pb flux variations were consistent with the historical atmospheric Pb deposition. Our results not only show the robustness of the CRS model to establish accurate chronologies of recent peat deposits but also provide evidence that there are confounding factors that might influence the calculation of reliable peat accumulation rates (and thus also element accumulation rates/fluxes). This study emphasizes the need to verify the hypotheses of {sup 210}Pb-dating models and the usefulness of a full geochemical interpretation of peat bog records. - Highlights: ► Peat cores collected in NW Spain were used to reconstruct recent Pb deposition. ► Applicability of {sup 210}Pb-dating models (CRS) in bogs is discussed based on PCA results. ► Results showed that ∼ 20% of the {sup 210}Pb flux was related to anthropogenic metal pollution. ► Geochemical analysis of bogs is useful to

  4. Evaluation of design variants for improved inherent regulation of advanced small modular reactors - 15325

    International Nuclear Information System (INIS)

    Vilim, R.B.; Passerini, S.

    2015-01-01

    This paper examines design variants that can improve inherent regulation in Advanced Small Modular Reactors (ASMR). It looks at the nature of unprotected upsets and then develops appropriate design measures to ensure that no upset can override a capability for safe self-regulation. This work adopts a reference sodium fast reactor (SFR) design to serve as a baseline for operational and safety performance and for comparison with variants on this design. The effect of design measures on plant stability is then examined. It is found that compared to full-power operation, the stability margin is reduced under islanded-operation. Islanded-operation is more likely for an ASMR deployed in a small regional electric grid with high penetration of renewable energy sources. The stability of core power production is a function of the inlet temperature coefficient, coolant transport times, and temperature-front attenuation in heat exchangers. The interaction of these phenomena with the control system is described

  5. An improved method for quantitatively measuring the sequences of total organic carbon and black carbon in marine sediment cores

    Science.gov (United States)

    Xu, Xiaoming; Zhu, Qing; Zhou, Qianzhi; Liu, Jinzhong; Yuan, Jianping; Wang, Jianghai

    2018-01-01

    Understanding global carbon cycle is critical to uncover the mechanisms of global warming and remediate its adverse effects on human activities. Organic carbon in marine sediments is an indispensable part of the global carbon reservoir in global carbon cycling. Evaluating such a reservoir calls for quantitative studies of marine carbon burial, which closely depend on quantifying total organic carbon and black carbon in marine sediment cores and subsequently on obtaining their high-resolution temporal sequences. However, the conventional methods for detecting the contents of total organic carbon or black carbon cannot resolve the following specific difficulties, i.e., (1) a very limited amount of each subsample versus the diverse analytical items, (2) a low and fluctuating recovery rate of total organic carbon or black carbon versus the reproducibility of carbon data, and (3) a large number of subsamples versus the rapid batch measurements. In this work, (i) adopting the customized disposable ceramic crucibles with the microporecontrolled ability, (ii) developing self-made or customized facilities for the procedures of acidification and chemothermal oxidization, and (iii) optimizing procedures and carbon-sulfur analyzer, we have built a novel Wang-Xu-Yuan method (the WXY method) for measuring the contents of total organic carbon or black carbon in marine sediment cores, which includes the procedures of pretreatment, weighing, acidification, chemothermal oxidation and quantification; and can fully meet the requirements of establishing their highresolution temporal sequences, whatever in the recovery, experimental efficiency, accuracy and reliability of the measurements, and homogeneity of samples. In particular, the usage of disposable ceramic crucibles leads to evidently simplify the experimental scenario, which further results in the very high recovery rates for total organic carbon and black carbon. This new technique may provide a significant support for

  6. Improvement on electrochemical performance by partial replacement of Ru@Pt core-shell nanocatalyst by temperature modification

    International Nuclear Information System (INIS)

    Chang, Chih-Juei; Lin, Liang-You; Tseng, Fan-Gang

    2014-01-01

    In this paper, the homemade open-loop reduction system (OLRS), and redox transmetalation method were utilized to produce the core-shell Ru (ruthenium)/Pt (platinum) catalysts on the carbon cloth (CC) for direct methanol fuel cell (DMFC) application. By adjusting pH value and heating to proper temperature of the ionized reduction environment, Pt 4+ can be first converted into Pt 2+ to allow partial Ru replacement with Pt by redox transmetalation and produce Ru@Pt core-shell nanostructures[1]. And we change the reduction temperature to see how it affects the efficiency of the DMFC. The scanning electron microscopic (SEM) top-view micrographs showing that the apparent Ru@Pt nanoparticles successfully deposited on both the inner and outer surfaces of the hydrophilically-treated CC. At high SEM magnification, the small size and high-density distribution of the Ru@Pt nanoparticles were clearly observed on the hydrophilically-treated CC, and much more Pt@Ru catalyst deposit on the CC surface with the sample of 80 °C. The electrosorption charges of hydrogen ion (Q H ) and the peak current density (I P ) of the samples in the cyclic voltammetry (CV) curves. The magnitude of peak current density is positive correlation to the temperature. However, the CO tolerance, indicated that the better CO tolerance contributed to the less Pt replace on Ru cluster, which allow the Ru oxidizing CO to CO 2 efficiently, is negative correlation-- to the temperature. The sample of 50 °C shows the better combination catalyst efficiency between the CO tolerance and the electrochemical performance

  7. Improving safety margin of LWRs by rethinking the emergency core cooling system criteria and safety system capacity

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Bokyung, E-mail: bkkim2@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

    2016-10-15

    Highlights: • Zircaloy embrittlement criteria can increase to 1370 °C for CP-ECR lower than 13%. • The draft ECCS criteria of U.S. NRC allow less than 5% in power margin. • The Japanese fracture-based criteria allow around 5% in power margin. • Increasing SIT inventory is effective in assuring safety margin for power uprates. - Abstract: This study investigates the engineering compatibility between emergency core cooling system criteria and safety water injection systems, in the pursuit of safety margin increase of light water reactors. This study proposes an acceptable temperature increase to 1370 °C as long as equivalent cladding reacted calculated by the Cathcart–Pawel equation is below 13%, after an extensive literature review. The influence of different ECCS criteria on the safety margin during large break loss of coolant accident is investigated for OPR-1000 by the system code MARS-KS, implemented with the KINS-REM method. The fracture-based emergency core cooling system (ECCS) criteria proposed in this study are shown to enable power margins up to 10%. In the meantime, the draft U.S. NRC’s embrittlement criteria (burnup-sensitive) and Japanese fracture-based criteria are shown to allow less than 5%, and around 5% of power margins, respectively. Increasing safety injection tank (SIT) water inventory is the key, yet convenient, way of assuring safety margin for power increase. More than 20% increase in the SIT water inventory is required to allow 15% power margins, for the U.S. NRC’s burnup-dependent embrittlement criteria. Controlling SIT water inventory would be a useful option that could allow the industrial desire to pursue power margins even under the recent atmosphere of imposing stricter ECCS criteria for the considerable burnup effects.

  8. Developing a set of consensus indicators to support maternity service quality improvement: using Core Outcome Set methodology including a Delphi process.

    Science.gov (United States)

    Bunch, K J; Allin, B; Jolly, M; Hardie, T; Knight, M

    2018-05-16

    To develop a core metric set to monitor the quality of maternity care. Delphi process followed by a face-to-face consensus meeting. English maternity units. Three representative expert panels: service designers, providers and users. Maternity care metrics judged important by participants. Participants were asked to complete a two-phase Delphi process, scoring metrics from existing local maternity dashboards. A consensus meeting discussed the results and re-scored the metrics. In all, 125 distinct metrics across six domains were identified from existing dashboards. Following the consensus meeting, 14 metrics met the inclusion criteria for the final core set: smoking rate at booking; rate of birth without intervention; caesarean section delivery rate in Robson group 1 women; caesarean section delivery rate in Robson group 2 women; caesarean section delivery rate in Robson group 5 women; third- and fourth-degree tear rate among women delivering vaginally; rate of postpartum haemorrhage of ≥1500 ml; rate of successful vaginal birth after a single previous caesarean section; smoking rate at delivery; proportion of babies born at term with an Apgar score improvement. Achieving consensus on core metrics for monitoring the quality of maternity care. © 2018 The Authors. BJOG: An International Journal of Obstetrics and Gynaecology published by John Wiley & Sons Ltd on behalf of Royal College of Obstetricians and Gynaecologists.

  9. Long-term follow-up of a randomized controlled trial on additional core stability exercises training for improving dynamic sitting balance and trunk control in stroke patients.

    Science.gov (United States)

    Cabanas-Valdés, Rosa; Bagur-Calafat, Caritat; Girabent-Farrés, Montserrat; Caballero-Gómez, Fernanda Mª; du Port de Pontcharra-Serra, Helena; German-Romero, Ana; Urrútia, Gerard

    2017-11-01

    Analyse the effect of core stability exercises in addition to conventional physiotherapy training three months after the intervention ended. A randomized controlled trial. Outpatient services. Seventy-nine stroke survivors. In the intervention period, both groups underwent conventional physiotherapy performed five days/week for five weeks, and in addition the experimental group performed core stability exercises for 15 minutes/day. Afterwards, during a three-month follow-up period, both groups underwent usual care that could eventually include conventional physiotherapy or physical exercise but not in a controlled condition. Primary outcome was trunk control and dynamic sitting balance assessed by the Spanish-Version of Trunk Impairment Scale 2.0 and Function in Sitting Test. Secondary outcomes were standing balance and gait evaluated by the Berg Balance Scale, Tinetti Test, Brunel Balance Assessment, Spanish-Version of Postural Assessment Scale for Stroke and activities of daily living using the Barthel Index. A total of 68 subjects out of 79 completed the three-month follow-up period. The mean difference (SD) between groups was 0.78 (1.51) points ( p = 0.003) for total score on the Spanish-Version of Trunk Impairment Scale 2.0, 2.52 (6.46) points ( p = 0.009) for Function in Sitting Test, dynamic standing balance was 3.30 (9.21) points ( p= 0.009) on the Berg Balance Scale, gait was 0.82 (1.88) points ( p = 0.002) by Brunel Balance Assessment (stepping), and 1.11 (2.94) points ( p = 0.044) by Tinetti Test (gait), all in favour of core stability exercises. Core stability exercises plus conventional physiotherapy have a positive long-term effect on improving dynamic sitting and standing balance and gait in post-stroke patients.

  10. Largely improved the low temperature toughness of acrylonitrile-styrene-acrylate (ASA) resin: Fabricated a core-shell structure of two elastomers through the differences of interfacial tensions

    Science.gov (United States)

    Mao, Zepeng; Zhang, Jun

    2018-06-01

    The phase morphology of two elastomers (i.e., chlorinated polyethylene (CPE) and polybutadiene rubber (BR)) were devised to be a core-shell structure in acrylonitrile-styrene-acrylate (ASA) resin matrix, via the interfacial tension differences of polymer pairs. Selective extraction test and scanning electron microscopy (SEM) were utilized to verify this special phase morphology. The results demonstrated that the core-shell structure, BR core and CPE shell, significantly contributed to improve the low temperature toughness of ASA/CPE/BR ternary blends, which may be because the nonpolar BR core was segregated from polar ASA by the CPE shell. The CPE shell served dual functions: Not only did it play compatibilizing effect in the interface between BR and ASA matrix, but it also toughened the blends at 25 and 0 °C. The blends of ASA/CPE/BR (100/27/3, w/w/w) and ASA/CPE/BR (100/22/8, w/w/w) showed the peak impact strengths at about 28 and 9 kJ/m2 at 0 and -30 °C, respectively, which were higher than both that of ASA/CPE/BR (100/30/0, w/w/w) and ASA/CPE/BR (100/0/30, w/w/w). Moreover, the impact strength of ternary blends at room temperature kept at 40 kJ/m2 when BR content was lower than 10 phr. Other characterizations including contact angle measurement, dynamic mechanical thermal analysis (DMTA), morphology of impact-fractured surfaces, tensile properties, flexural properties, and Fourier transform infrared spectroscopy (FTIR) were measured as well.

  11. Performance Estimation of Supercritical CO2 Cycle for the PG-SFR application with Heat Sink Temperature Variation

    International Nuclear Information System (INIS)

    Ahn, Yoonhan; Cho, Seong Kuk; Lee, Jeong Ik

    2015-01-01

    The heat sink temperature conditions are referred from the annual database of sea water temperature in East sea. When the heat sink temperature increases, the compressor inlet temperature can be influenced and the sudden power decrease can happen due to the large water pumping power. When designing the water pump, the pumping margin should be considered as well. As a part of Prototype Generation IV Sodium-cooled Fast Reactor (PG-SFR) development, the Supercritical CO 2 cycle (S-CO 2 ) is considered as one of the promising candidate that can potentially replace the steam Rankine cycle. S-CO 2 cycle can achieve distinctively high efficiency compared to other Brayton cycles and even competitive performance to the steam Rankine cycle under the mild turbine inlet temperature region. Previous studies explored the optimum size of the S-CO 2 cycle considering component designs including turbomachinery, heat exchangers and pipes. Based on the preliminary design, the thermal efficiency is 31.5% when CO 2 is sufficiently cooled to the design temperature. However, the S-CO 2 compressor performance is highly influenced by the inlet temperature and the compressor inlet temperature can be changed when the heat sink temperature, in this case sea water temperature varies. To estimate the S-CO 2 cycle performance of PG-SFR in the various regions, a Quasi-static system analysis code for S-CO 2 cycle is developed by the KAIST research team. A S-CO 2 cycle for PG-SFR is designed and assessed for off-design performance with the heat sink temperature variation

  12. A neutronics study for improving the safety and performance parameters of a 3600 MWth Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Sun, Kaichao; Krepel, Jiri; Mikityuk, Konstantin; Chawla, Rakesh

    2013-01-01

    Highlights: ► The potential for neutronics design optimization is assessed for a large SFR core. ► Both beginning-of-life and equilibrium fuel cycle conditions are considered. ► The sodium void effect is decomposed via a neutron balance based methodology. ► The optimized core options adopt an appropriate sodium plenum design to reduce the void effect. ► The introduction of moderator pins is considered for enhancing the Doppler effect. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many performance advantages, but has one dominating neutronics drawback – a positive sodium void reactivity. The starting point for the present study is an SFR core design considered in the Collaborative Project on the European Sodium-cooled Fast Reactor (CP-ESFR). The aim is to analyze, for this reference core, four safety and performance parameters from the viewpoint of four different optimization options, and to propose possible optimized core designs. In doing so, the study focuses not only on the beginning-of-life state of the core, but also on the beginning of equilibrium closed fuel cycle. The four studied optimization options are: (a) introducing an upper sodium plenum and boron layer, (b) varying the core height-to-diameter (H/D) ratio, (c) introducing moderator pins into the fuel assembly, and (d) modifying the initial plutonium content. The sensitivity of the void reactivity, Doppler constant, nominal reactivity and breeding gain has been evaluated. In particular, the void reactivity, which is the most crucial safety parameter for the SFR, has been decomposed into its reaction-wise, isotope-wise and energy-group-wise components using a methodology based on the neutron balance equation. Extended voiding in the upper sodium plenum region – in conjunction with the effect of a boron layer introduced above the plenum – is found to be particularly effective in the void effect reduction while, at the same time

  13. Next generation self-shielded flux cored electrode with improved toughness for off shore oil well platform structures

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Daya; Soltis, Patrick; Narayanan, Badri; Quintana, Marie; Fox, Jeff [The Lincoln Electric Company (United States)

    2005-07-01

    Self-shielded flux cored arc welding electrodes (FCAW-S) are ideal for outdoor applications, particularly open fabrication yards where high winds are a possibility. Development work was carried out on a FCAW-S electrode for welding 70 and 80 ksi yield strength base materials with a required minimum average Charpy V-Notch (CVN) absorbed energy value of 35 ft-lb at -40 deg F in the weld metal. The effect of Al, Mg, Ti, and Zr on CVN toughness was evaluated by running a Design of Experiments approach to systematically vary the levels of these components in the electrode fill and, in turn, the weld metal. These electrodes were used to weld simulated pipe joints. Over the range of compositions tested, 0.05% Ti in the weld metal was found to be optimum for CVN toughness. Ti also had a beneficial effect on the usable voltage range. Simulated offshore joints were welded to evaluate the effect of base metal dilution, heat input, and welding procedure on the toughness of weld metal. CVN toughness was again measured at -40 deg F on samples taken from the root and the cap pass regions. The root pass impact toughness showed strong dependence on the base metal dilution and the heat input used to weld the root and fill passes. (author)

  14. New Hybrid Multiple Attribute Decision-Making Model for Improving Competence Sets: Enhancing a Company’s Core Competitiveness

    Directory of Open Access Journals (Sweden)

    Kuan-Wei Huang

    2016-02-01

    Full Text Available A company’s core competitiveness depends on the strategic allocation of its human resources in alignment with employee capabilities. Competency models can identify the range of capabilities at a company’s disposal, and this information can be used to develop internal or external education training policies for sustainable development. Such models can ensure the importation of a strategic orientation reflecting the growth of its employee competence set and enhancing human resource sustainably. This approach ensures that the most appropriate people are assigned to the most appropriate positions. In this study, we proposed a new hybrid multiple attributed decision-making model by using the Decision-making trial and Evaluation Laboratory Technique (DEMATEL to construct an influential network relation map (INRM and determined the influential weights by using the basic concept of the analytic network process (called DEMATEL-based ANP, DANP; the influential weights were then adopted with a modified Vise Kriterijumska Optimizacija I Kompromisno Resenje (VIKOR method. A simple forecasting technique as an iteration function was also proposed. The proposed model was effective. We expect that the proposed model can facilitate making timely revisions, reflecting the growth of employee competence sets, reducing the performance gap toward the aspiration level, and ensuring the sustainability of a company.

  15. Development of an improved wearable device for core body temperature monitoring based on the dual heat flux principle.

    Science.gov (United States)

    Feng, Jingjie; Zhou, Congcong; He, Cheng; Li, Yuan; Ye, Xuesong

    2017-04-01

    In this paper, a miniaturized wearable core body temperature (CBT) monitoring system based on the dual heat flux (DHF) principle was developed. By interspersing calcium carbonate powder in PolyDimethylsiloxane (PDMS), a reformative heat transfer medium was produced to reduce the thermal equilibrium time. Besides, a least mean square (LMS) algorithm based active noise cancellation (ANC) method was adopted to diminish the impact of ambient temperature fluctuations. Theoretical analyses, finite element simulation, experiments on a hot plate and human volunteers were performed. The results showed that the proposed system had the advantages of small size, reduced initial time (~23.5 min), and good immunity to fluctuations of the air temperature. For the range of 37-41 °C on the hot plate, the error compared with a Fluke high accuracy thermometer was 0.08  ±  0.20 °C. In the human experiments, the measured temperature in the rest trial (34 subjects) had a difference of 0.13  ±  0.22 °C compared with sublingual temperature, while a significant increase of 1.36  ±  0.44 °C from rest to jogging was found in the exercise trial (30 subjects). This system has the potential for reliable continuous CBT measurement in rest and can reflect CBT variations during exercise.

  16. Dual-shell hollow polyaniline/sulfur-core/polyaniline composites improving the capacity and cycle performance of lithium–sulfur batteries

    Energy Technology Data Exchange (ETDEWEB)

    An, Yanling; Wei, Pan; Fan, Meiqiang, E-mail: fanmeiqiang@126.com; Chen, Da; Chen, Haichao; Ju, QiangJian; Tian, Guanglei; Shu, Kangying

    2016-07-01

    Highlights: • A dual core-shell hPANI/S/PANI composite was prepared in situ synthesis. • Cycle performance of the hPANI/S/PANI composite was enhanced. • The improvement was due to fine sulfur particles wrapped by two PANI films. • Some positive effects were elaborated. - Abstract: In this study, a dual-shell hollow polyaniline/sulfur-core/polyaniline (hPANI/S/PANI) composite was prepared by successively depositing PANI, S, and PANI on the surface of a template silicon sphere. The electrochemical properties of this composite were evaluated using a lithium plate as an anode in lithium/sulfur cells. The hPANI/S/PANI composite showed a discharge capacity of 572.2 mAh g{sup −1} after 214 cycles at 0.1 C, and the Coulombic efficiency was above 87% in the whole charge/discharge cycle. The improved cycle property of the hPANI/S/PANI composite can be ascribed to the fine sulfur particles homogeneously deposited on the PANI surface and sprawled inside the two PANI layers during the charge/discharge cycle. This behavior stabilized the nanostructure of sulfur and enhanced its conductivity.

  17. Improving communication skill training in patient centered medical practice for enhancing rational use of laboratory tests: The core of bioinformation for leveraging stakeholder engagement in regulatory science.

    Science.gov (United States)

    Moura, Josemar de Almeida; Costa, Bruna Carvalho; de Faria, Rosa Malena Delbone; Soares, Taciana Figueiredo; Moura, Eliane Perlatto; Chiappelli, Francesco

    2013-01-01

    Requests for laboratory tests are among the most relevant additional tools used by physicians as part of patient's health problemsolving. However, the overestimation of complementary investigation may be linked to less reflective medical practice as a consequence of a poor physician-patient communication, and may impair patient-centered care. This scenario is likely to result from reduced consultation time, and a clinical model focused on the disease. We propose a new medical intervention program that specifically targets improving the patient-centered communication of laboratory tests results, the core of bioinformation in health care. Expectations are that medical students training in communication skills significantly improve physicians-patient relationship, reduce inappropriate use of laboratorial tests, and raise stakeholder engagement.

  18. The CORE Service Improvement Programme for mental health crisis resolution teams: study protocol for a cluster-randomised controlled trial.

    Science.gov (United States)

    Lloyd-Evans, Brynmor; Fullarton, Kate; Lamb, Danielle; Johnston, Elaine; Onyett, Steve; Osborn, David; Ambler, Gareth; Marston, Louise; Hunter, Rachael; Mason, Oliver; Henderson, Claire; Goater, Nicky; Sullivan, Sarah A; Kelly, Kathleen; Gray, Richard; Nolan, Fiona; Pilling, Stephen; Bond, Gary; Johnson, Sonia

    2016-03-22

    As an alternative to hospital admission, crisis resolution teams (CRTs) provide intensive home treatment to people experiencing mental health crises. Trial evidence supports the effectiveness of the CRT model, but research suggests that the anticipated reductions in inpatient admissions and increased user satisfaction with acute care have been less than hoped for following the scaling up of CRTs nationally in England, as mandated by the National Health Service (NHS) Plan in 2000. The organisation and service delivery of the CRTs vary substantially. This may reflect the lack of a fully specified CRT model and the resources to enhance team model fidelity and to improve service quality. We will evaluate the impact of a CRT service improvement programme over a 1-year period on the service users' experiences of care, service use, staff well-being, and team model fidelity. Twenty-five CRTs from eight NHS Trusts across England will be recruited to this cluster-randomised trial: 15 CRTs will be randomised to receive the service improvement programme over a 1-year period, and ten CRTs will not receive the programme. Data will be collected from 15 service users and all clinical staff from each participating CRT at baseline and at the end of the intervention. Service use data will be collected from the services' electronic records systems for two 6-month periods: the period preceding and the period during months 7-12 of the intervention. The study's primary outcome is service user satisfaction with CRT care, measured using a client satisfaction questionnaire. Secondary outcomes include the following: perceived continuity of care, hospital admission rates and bed use, rates of readmission to acute care following CRT support, staff morale, job satisfaction, and general health. The adherence of the services to a model of best practice will be assessed at baseline and follow-up. Outcomes will be compared between the intervention and control teams, adjusting for baseline

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-17

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

  20. Establishment of the international collaboration and licensing preparation planning for the specific design of a prototype SFR

    International Nuclear Information System (INIS)

    Kim, Y. G.; Joo, H. K.; Cho, C. H.; Yoo, J. W.; Lee, D. U.; Ahn, K. S.; Hwang, Y. S.

    2013-05-01

    The conceptual design of prototype of Gen IV SFR (PGSFR) will be early determined through the review of the international experts. After this, the technology demonstration plan and validation of fuel design will be determined in more detail. The project will be accomplished efficiently by introducing the proven technology already validated from the international collaboration. The conceptual design and its requirements of PGSFR will be reviewed by ANL, who has a lot of design experiences in the metal fueled SFR development. The collaboration with ANL has been done through Work For Others (WFO) contract, and the MOU was signed between SFRA and Terra Power(USA), and SFRA and IGCAR. The licensing issues raised during PFBR and FBTR licensing in India will be discussed and reflected into the PGSFR design by inviting the high level expert from India, for example Dr. Chetal in IGCAR. The specific design, technology validation plan and fuel development plan will be established in more detail through the annual International Technical Review Meeting (ITRM) and experimental facilities available from the international institute and companies, which will be the basis for shortening the project period and to reduce the development cost

  1. Replaceable LMFBR core components

    International Nuclear Information System (INIS)

    Evans, E.A.; Cunningham, G.W.

    1976-01-01

    Much progress has been made in understanding material and component performance in the high temperature, fast neutron environment of the LMFBR. Current data have provided strong assurance that the initial core component lifetime objectives of FFTF and CRBR can be met. At the same time, this knowledge translates directly into the need for improved core designs that utilize improved materials and advanced fuels required to meet objectives of low doubling times and extended core component lifetimes. An industrial base for the manufacture of quality core components has been developed in the US, and all procurements for the first two core equivalents for FFTF will be completed this year. However, the problem of fabricating recycled plutonium while dramatically reducing fabrication costs, minimizing personnel exposure, and protecting public health and safety must be addressed

  2. Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor