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Sample records for burned core analysis

  1. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  2. Neutronic Analysis of Advanced SFR Burner Cores using Deep-Burn PWR Spent Fuel TRU Feed

    International Nuclear Information System (INIS)

    In this work, an advanced sodium-cooled fast TRU (Transuranics) burner core using deep-burn TRU feed composition discharged from small LWR cores was neutronically analyzed to show the effects of deeply burned TRU feed composition on the performances of sodium-cooled fast burner core. We consider a nuclear park that is comprised of the commercial PWRs, small PWRs of 100MWe for TRU deep burning using FCM (Fully Ceramic Micro-encapsulated) fuels and advanced sodium-cooled fast burners for their synergistic combination for effective TRU burning. In the small PWR core having long cycle length of 4.0 EFPYs, deep burning of TRU up to 35% is achieved with FCM fuel pins whose TRISO particle fuels contain TRUs in their central kernel. In this paper, we analyzed the performances of the advanced SFR burner cores using TRU feeds discharged from the small long cycle PWR deep-burn cores. Also, we analyzed the effect of cooling time for the TRU feeds on the SFR burner core. The results showed that the TRU feed composition from FCM fuel pins of the small long cycle PWR core can be effectively used into the advanced SFR burner core by significantly reducing the burnup reactivity swing which reduces smaller number of control rod assemblies to satisfy all the conditions for the self controllability than the TRU feed composition discharged from the typical PWR cores

  3. Thermal-Fluid and Safety Analysis of the TRU Deep-Burn MHR Core

    International Nuclear Information System (INIS)

    The DB-MHR (Deep Burn-Modular Helium Reactor) concept was proposed by GA to achieve a very high burnup of the LWR TRU fuel. To increase the TRU discharge burnup, the original GT-MHR of GA was modified for the DB-MHR core: a 5-fuel-ring configuration was adopted instead of the original 3-fueling concept. This paper describes the GAMMA+ thermal-fluid analysis of the 600MWth DB-MHR system at the steady state and the transient condition of LPCC (Low Pressure Conduction Cooling) event. The objective of this study is to characterize the DB-MHR core in terms of the fuel temperature during the nominal and LPCC conditions

  4. PWR rod ejection accident: uncertainty analysis on a high burn-up core configuration

    Energy Technology Data Exchange (ETDEWEB)

    Le Pallec, J.C.; Studer, E.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Service d' Etudes de Reacteurs et de Modelisation Avancee (DEN/SERMA), 91 - Gif sur Yvette (France)

    2003-07-01

    With the increasing of the discharge burn-up assembly, the rod ejection accident (REA) methodology based on the analyse of the hot spot from a decoupling methods of calculation does not allow to ensure the respect of safety criteria. The main reason is that the irradiated fuel certainly less solicited thermally is in the other hand more sensitive to a transient due to a rod ejection. Thus, the hot spot is not necessarily the sensitive point of the core. In the framework of high burn-up configurations, a new methodology tends to replace the former. It characterizes by the use of a best-estimate 3-dimensional modelling: coupling of the thermal hydraulics and neutronics, taking in account fuel properties depending on irradiation. To ensure the conservatism of the modelling response, this new approach has to be followed by an uncertainties analysis. Inputs from the benchmark RIA TMI-1 conducted by IRSN (France), NRC (United State of America) and KI (Russian) are used to perform a first analysis. The response of the modelling is the enthalpy deposited in an assembly. The analysis is based on the Design of Experiments (DoE) that permits to measure the weight of the main parameters and their interactions on the response. These last cannot be disregarded because they represent up to 20% of the penalizing uncertainty. This study shows that the main fuel modifications due to irradiation (radial power distribution, thermal properties degradation) have to be taken into account in a realistic thermal modelling during a strong transient.

  5. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

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

  6. Multiphysics Model Development and the Core Analysis for In Situ Breeding and Burning Reactor

    Directory of Open Access Journals (Sweden)

    Shengyi Si

    2013-01-01

    Full Text Available The in situ breeding and burning reactor (ISBBR, which makes use of the outstanding breeding capability of metallic pellet and the excellent irradiation-resistant performance of SiCf/SiC ceramic composites cladding, can approach the design purpose of ultralong cycle and ultrahigh burnup and maintain stable radial power distribution during the cycle life without refueling and shuffling. Since the characteristics of the fuel pellet and cladding are different from the traditional fuel rod of ceramic pellet and metallic cladding, the multiphysics behaviors in ISBBR are also quite different. A computer code, named TANG, to model the specific multiphysics behaviors in ISBBR has been developed. The primary calculation results provided by TANG demonstrate that ISBBR has an excellent comprehensive performance of GEN-IV and a great development potential.

  7. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

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

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope

    2011-10-01

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

  9. High-burn up 10 x 10 100%MOX ABWR core physics analysis with APOLLO2.8 and TRIPOLI-4.5 codes

    Energy Technology Data Exchange (ETDEWEB)

    Blaise, Patrick, E-mail: patrick.blaise@cea.f [Centre de Cadarache, DEN-CAD/DER/SPRC - building 230, F-13108 Saint Paul-Lez-Durance (France); Huot, Nicolas [Centre de Saclay, DEN-DANS/DM2S/SERMA - building 470, F-91191 Gif-sur-Yvette (France); Thiollay, Nicolas [Centre de Cadarache, DEN-CAD/DER/SPEX - building 238, F-13108 Saint Paul-Lez-Durance (France); Fougeras, Philippe; Santamarina, Alain [Centre de Cadarache, DEN-CAD/DER/SPRC - building 230, F-13108 Saint Paul-Lez-Durance (France)

    2010-07-15

    Within the frame of several extensive experimental core physics programs led between 1996 and 2008 between CEA and Japan Nuclear Energy Safety Organization (JNES), the FUBILA experiment has been conducted in the French EOLE Facility between 2005 and 2006 to obtain valuable data for the validation of core analysis methods related to full MOX advanced BWR and high-burn up BWR cores. During this experimental campaign, a particular FUBILA 10 x 10 Advanced BWR configuration devoted to the validation of high-burn up 100%MOX BWR bundles was built. It is characterized by an assembly average total Pu enrichment of 10.6 wt.% and in-channel void of 40%, representative of hot full power conditions at core mid-plane and average discharge burnup of 65 GWd/t. This paper details the validation work led on the TRIPOLI-4.5 Continuous Energy Monte Carlo code and APOLLO2.8/CEA2005V4 deterministic code package for the interpretation of this 10 x 10 high-burn up configuration. The APOLLO2.8/CEA2005V4 package relies on the deterministic lattice transport code APOLLO2.8 based on the Method of Characteristics (MOC), and its new CEA2005v4 multigroup library based on the latest JEFF-3.1.1 nuclear data file, processed also for the TRIPOLI-4.5 code. The results obtained on critical mass and radial pin-by-pin power distributions are presented. For critical mass, the calculation-to-experiment C-E on the k{sub eff} spreads from 300 pcm for TRIPOLI to 600 pcm for APOLLO2.8 in its Optimized BWR Scheme (OBS) in 26 groups. For pin-by-pin radial power distributions, all codes give acceptable results, with maximum discrepancies on C/E - 1 of the order of 3-4% for very heterogeneous bundles where P{sub max}/P{sub min} reaches 4, 2. These values are within 2 standard deviations of the experimental uncertainty. Those results demonstrate the capability of both codes and schemes to accurately predict Advanced High burnup 100%-MOX BWR key-neutron parameters.

  10. Enhanced minor actinide burning core for closed fuel cycle

    International Nuclear Information System (INIS)

    This paper presents core concepts enhancing TRU burning or MA transmutation in sodium cooled reactor satisfying the void reactivity requirements. In this study, two concepts of transmutation system are considered; in the first system TRUs are burned only by ARR whose target is maximizing TRU burning. The second is a system that Pu is burned by LWR and ARR, Am is transmuted by ARR whose target is maximizing Am transmutation. Therefore some innovative and challenging technologies have been examined under the safety requirements; MA burning fuel with 50% TRU fraction, moderator pin, fuel of high Am fraction, and Am blanket. According to the detailed calculation of high TRU contained oxide core with moderator pins of 12% arranged driver fuel assemblies, the TRU conversion ratio decreases to 0.33 and the TRU burning capability is improved to 67 kg/TWeh. Deploying Am blanket which is oxide fuel with Am 50% and U 50%, the total of Am transmutation capability of oxide fueled core becomes 69 kg/TWeh. (author)

  11. Persistent spectral hole burning in oxygen-evolving photosystem II cores from cyanobacteria and higher plants

    International Nuclear Information System (INIS)

    Persistent spectral hole burning was performed on active photosystem II (PSII) cores from spinach and synechocystis 6803, each containing ∼32 chlorophyll a molecules per core. Hole-burning action spectra are presented. The data appear inconsistent with a mechanism involving non-photochemical hole burning, as has previously been observed in isolated PSII protein-pigment fragments. A photochemical hole-burning mechanism involving charge separation in P680 accounts for the features of the spectra presented

  12. Fast reactor calculational route for Pu burning core design

    International Nuclear Information System (INIS)

    This document provides a description of a calculational route, used in the Reactor Physics Research Section for sensitivity studies and initial design optimization calculations for fast reactor cores. The main purpose in producing this document was to provide a description of and user guides to the calculational methods, in English, as an aid to any future user of the calculational route who is (like the author) handicapped by a lack of literacy in Japanese. The document also provides for all users a compilation of information on the various parts of the calculational route, all in a single reference. In using the calculational route (to model Pu burning reactors) the author identified a number of areas where an improvement in the modelling of the standard calculational route was warranted. The document includes comments on and explanations of the modelling assumptions in the various calculations. Practical information on the use of the calculational route and the computer systems is also given. (J.P.N.)

  13. Determination of the fuel element burn-up for mixed TRIGA core by measurement and calculation with new TRIGLAV code

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, T.; Ravnik, M.; Persic, A. (J.Stefan Institute, Ljubljana (Slovenia))

    1999-12-15

    Results of fuel element burn-up determination by measurement and calculation are given. Fuel element burn-up was calculated with two different programs TRIGLAV and TRIGAC using different models. New TRIGLAV code is based on cylindrical, two-dimensional geometry with four group diffusion approximation. TRIGAC program uses one-dimensional cylindrical geometry with twogroup diffusion approximation. Fuel element burn-up was measured with reactivity method. In this paper comparison and analysis of these three methods is presented. Results calculated with TRIGLAV show considerably better alignment with measured values than results calculated with TRIGAC. Some two-dimensional effects in fuel element burn-up can be observed, for instance smaller standard fuel element burn-up in mixed core rings and control rod influence on nearby fuel elements. (orig.)

  14. Verification of Burned Core Modeling Method for Monte Carlo Simulation of HANARO

    International Nuclear Information System (INIS)

    The reactor core has been managed well by the HANARO core management system called HANAFMS. The heterogeneity of the irradiation device and core made the neutronic analysis difficult and sometimes doubtable. To overcome the deficiency, MCNP was utilized in neutron transport calculation of the HANARO. For the most part, a MCNP model with the assumption that all fuels are filled with fresh fuel assembly showed acceptable analysis results for a design of experimental devices and facilities. However, it sometimes revealed insufficient results in the design, which requires good accuracy like neutron transmutation doping (NTD), because it didn't consider the flux variation induced by depletion of the fuel. In this study, a depleted-core modeling method previously proposed was applied to build burned core model of HANARO and verified through a comparison of the calculated result from the depleted-core model and that from an experiment. The modeling method to establish a depleted-core model for the Monte Carlo simulation was verified by comparing the neutron flux distribution obtained by the zirconium activation method and the reaction rate of 30Si(n, γ) 31Si obtained by a resistivity measurement method. As a result, the reaction rate of 30Si(n, γ) 31Si also agreed well with about 3% difference. It was therefore concluded that the modeling method and resulting depleted-core model developed in this study can be a very reliable tool for the design of the planned experimental facility and a prediction of its performance in HANARO

  15. Verification of Burned Core Modeling Method for Monte Carlo Simulation of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dongkeun; Kim, Myongseop [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The reactor core has been managed well by the HANARO core management system called HANAFMS. The heterogeneity of the irradiation device and core made the neutronic analysis difficult and sometimes doubtable. To overcome the deficiency, MCNP was utilized in neutron transport calculation of the HANARO. For the most part, a MCNP model with the assumption that all fuels are filled with fresh fuel assembly showed acceptable analysis results for a design of experimental devices and facilities. However, it sometimes revealed insufficient results in the design, which requires good accuracy like neutron transmutation doping (NTD), because it didn't consider the flux variation induced by depletion of the fuel. In this study, a depleted-core modeling method previously proposed was applied to build burned core model of HANARO and verified through a comparison of the calculated result from the depleted-core model and that from an experiment. The modeling method to establish a depleted-core model for the Monte Carlo simulation was verified by comparing the neutron flux distribution obtained by the zirconium activation method and the reaction rate of {sup 30}Si(n, γ) {sup 31}Si obtained by a resistivity measurement method. As a result, the reaction rate of {sup 30}Si(n, γ) {sup 31}Si also agreed well with about 3% difference. It was therefore concluded that the modeling method and resulting depleted-core model developed in this study can be a very reliable tool for the design of the planned experimental facility and a prediction of its performance in HANARO.

  16. PWR degraded core analysis

    International Nuclear Information System (INIS)

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  17. Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

    International Nuclear Information System (INIS)

    Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package MTRPC system, using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTRPC Package, Empirical Formula For Fuel Burn-Up.

  18. Burns

    Science.gov (United States)

    ... Chemical burns Burns can be the result of: House and industrial fires Car accidents Playing with matches ... hairs Burned lips and mouth Coughing Difficulty breathing Dark, black-stained mucus Voice changes Wheezing

  19. COREBN: A core burn-up calculation module for SRAC2006

    International Nuclear Information System (INIS)

    COREBN is an auxiliary code of the SRAC system for multi-dimensional core burn-up calculation based on the diffusion theory and interpolation of macroscopic cross-sections tabulated to local parameters such as burn-up degree, moderator temperature and so on. The macroscopic cross-sections are prepared by cell burn-up calculations with the collision probability method of SRAC. SRAC and COREBN have wide applicability for various types of cell and core geometries. They have been used mainly for the purpose of core burn-up management of research reactors in Japan Atomic Energy Agency. The report is a revision of the users manual for the latest version of COREBN served with the SRAC released in 2006. (author)

  20. Second law analysis of convective droplet burning

    International Nuclear Information System (INIS)

    In this paper the entropy generation due to burning particles in a gaseous stream is considered and the contribution to it compared. A second law analysis is undertaken in order to minimize the entropy generation and therefore the lost available work. The optimum flow conditions from this thermodynamically advantageous perspective are determined for a burning droplet at low Reynolds number and an optimum transfer number obtained. The transfer number so obtained depends directly on the square of the relative velocity, and inversely on the net enthalpy rise due to burning and the ratio of ambient to flame temperature. In realistic flows, where the transfer number and net heat release are fixed, these quantities are related to the relative velocity and ambient to flame temperature ratio in order to operate at optimum conditions. The square of the relative velocity in such flows is a small fraction of the net heat release so that, to operate at optimum thermodynamic conditions, it is determined that the droplet Reynolds number must be large suggesting a large droplet size and low gas velocity. Considerations pertaining to engineering practice are also considered and it is concluded that within constraints practice is consistent with the implications of the second law analysis

  1. Simplified models for pebble-bed HTR core burn-up calculations with Monteburns2.0©

    International Nuclear Information System (INIS)

    Highlights: ► PBMR-400 annular core is very difficult to simulate in a reliable way. ► Nuclide evolutions given by different lattice models can differ significantly. ► To split fixed lattice models into two axial zones does not affect results significantly. ► We can choose a (simplified) core model on the basis of the analysis aim. ► Monteburns gives by survey burn-up calculations reasonable nuclide evolution trends. - Abstract: This paper aims at comparing some simplified models to simulate irradiation cycles of Pu fuelled pebble-bed reactors with Monteburns2.0© code. As a reference core, the PBMR-400 (proposed in the framework of the EU PUMA project, where this kind of core fuelled by a Pu and Pu–Np fuel has been studied) was taken into account. Pebble-bed High Temperature Reactor (HTR) cores consist of hundreds of thousands pebbles arranged stochastically in a cylindrical or annular space and each pebble is a single fuel element, and it is able to reach ultra-high burn-ups, i.e. up to 750 GWd/tHM (for Pu-based fuels). Additionally, pebble-bed cores are characterised by a continuous recirculation of pebbles from the top to the bottom of the core. Modelling accurately with current computer codes such an arrangement, in order to predict the behaviour of the core itself, is a very difficult task and any depletion code specifically devoted to pebble-bed burn-up calculation is not available at the moment. Because of limitations of the most common current MCNP-based depletion codes as well as huge calculation times, simplified models have to be implemented. After an analysis of the literature available on pebble-bed models for criticality and burn-up calculations, a preliminary assessment of the impact of different kind of simplified models for a Pu-Np fuelled Pebble-Bed Modular Reactor (PBMR), proposed in the framework of the EU PUMA project, is shown, particularly as far as burn-up prediction with Monteburns2.0© code is concerned.

  2. In-core fuel management amd attainable fuel burn-up in TRIGA

    International Nuclear Information System (INIS)

    The principles of in-core fuel management in research reactors, and especially in TRIGA, are discussed. Calculations made to determine the attainable fuel burn-up values of various fuel element types in the Otaniemi TRIGA Mark II reactor are described and the results obtained are given. Recommendations are given of how to perform the in-core fuel management to achieve good fuel utilization. The results obtained indicate that burn-up values of up to 5 and 2.5 MWd/element can be achieved for the 8 wt-% U Al clad and the 8.5 wt-% U SS clad elements, respectively. (author)

  3. Advanced sodium cooled reactor cores having thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    In this paper, a design concept of 400 MWe sodium cooled fast reactor (SFR) cores having thorium blankets for effective burning of TRU (Transuranics) from LWR spent fuel is described. Specifically, we considered two recycling options of thorium blankets : 1) no recycling and 2) fully recycling. The thorium blankets are loaded in the axially central regions of the core regions and their axial heights are adjusted so as to increase TRU burning rate and to reduce burnup reactivity swing. Also, we analyzed the performances of the cores having different fuel management batch sizes and different recycling options for the searched core configuration. The results show that the axial thorium blankets with no recycling option can be effectively used to increase TRU burning rate with a significant reduction of burnup reactivity swing in comparison with typical SFR burner cores having no blankets while the recycling of thorium blanket degrades TRU burning rate and burnup reactivity swing but it leads to a reduction of sodium void worth and more negative Doppler coefficient. (author)

  4. Optimization studies for the prism alternative oxide core, and its response to the actinide burning strategy

    International Nuclear Information System (INIS)

    The PRISM advanced liquid metal reactor is designed by General Electric in a reference solution equipped by a metal fuelled core. An alternative oxide core is studied by General Electric and ENEA in the frame of a collaboration existing since 1989. This paper deals with the ENEA contribution on the oxide solution, aimed at the core optimization both from safety parameters and fuel cycle economy points of view. Moreover, synthetic information about ENEA evaluations about the minor actinide burning capability of the PRISM oxide core are given. (author)

  5. Preliminary design study of a board type radial fuel shuffling sodium cooled breed and burn reactor core

    International Nuclear Information System (INIS)

    Highlights: • A 1500MWt radial fuel shuffling sodium cooled breed and burn reactor core was designed. • The board type radial fuel shuffling strategy was applied and demonstrated. • Influences of the fuel height and core radius were investigated. - Abstract: In this paper, a preliminary board type radial fuel shuffling sodium cooled breed and burn reactor core is designed. In the current design, a number of breeding subassemblies are arranged in the center core to ensure enough breeding. A self-developed MCNP-ORIGEN coupled system with the ENDF/B-VI data library is applied to perform neutronics and burn-up calculations. For a 2.0 m radius and 2.5 m height core, the results demonstrate the feasibility of the board type radial fuel shuffling strategy. Breeding mainly occurs in the breeding subassemblies during the first 6 fuel cycles as they are moved to the burning/breeding region. The core will become asymptotically stable after about 24 years. The discharged burn-up of most subassemblies is about 15.0–30.0%. The influences of the core size on the major core parameters, such as initial keff, steady keff, maximum power density, peak burn-up and burn-up ratio between breeding and ignition subassemblies are calculated and investigated. The results indicate that the initial keff increases with fuel height and core radius and finally reaches stability; the steady keff increases with fuel height and core radius, then reaches peak value and finally decreases; the maximum power density, the peak burn-up and the burn-up ratio between breeding and ignition subassemblies decrease with the increase of fuel height and core radius; if core radius is less than 1.875 m, they increase sharply with the decrease of core radius

  6. Burns

    Science.gov (United States)

    ... touching the stove This list is not all-inclusive. You can also burn your airways if you ... extinguishers in key locations at home, work, and school. Remove electrical cords from floors and keep them ...

  7. TMI-2 core debris analysis

    International Nuclear Information System (INIS)

    One of the ongoing examination tasks for the damaged TMI-2 reactor is analysis of samples of debris obtained from the debris bed presently at the top of the core. This paper summarizes the results reported in the TMI-2 Core Debris Grab Sample Examination and Analysis Report, which will be available early in 1986. The sampling and analysis procedures are presented, and information is provided on the key results as they relate to the present core condition, peak temperatures during the transient, temperature history, chemical interactions, and core relocation. The results are then summarized

  8. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    OpenAIRE

    Muhammad Atta; Iqbal Masood; Mahmood Tayyab

    2011-01-01

    The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determin...

  9. Plasma-wall interaction data needs critical to a Burning Core Experiment (BCX)

    International Nuclear Information System (INIS)

    The Division of Development and Technology has sponsored a four day US-Japan workshop ''Plasma-Wall Interaction Data Needs Critical to a Burning Core Experiment (BCX)'', held at Sandia National Laboratories, Livermore, California on June 24 to 27, 1985. The workshop, which brought together fifty scientists and engineers from the United States, Japan, Germany, and Canada, considered the plasma-material interaction and high heat flux (PMI/HHF) issues for the next generation of magnetic fusion energy devices, the Burning Core Experiment (BCX). Materials options were ranked, and a strategy for future PMI/HHF research was formulated. The foundation for international collaboration and coordination of this research was also established. This volume contains the last three of the five technical sessions. The first of the three is on plasma materials interaction issues, the second is on research facilities and the third is from smaller working group meetings on graphite, beryllium, advanced materials and future collaborations

  10. Full MOX BWR core physics experiment. Experimental and analysis results of 9x9 reference core

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization (JNES) has been conducting an experimental program that is aimed to obtain a comprehensive data base for validation of core analysis methods applied to the full MOX ABWR and also for high burn up MOX fuel expected in the future. As a part of this program, JNES has been performing a MOX core physics experimental program, FUBILA, with collaboration of a French Consortium (CEA and COGEMA). The experiments has been designed to obtain the core physics data of the operating conditions of the full MOX BWR cores consisting of high burn up BWR MOX assemblies. The experiments started from January 2005 and completed in September 2006 at the EOLE critical facility of the CEA Cadarache center in France. Theoretical analysis of the experimental data has been also carried out with a continuous energy Monte Carlo calculation and a deterministic method with major nuclear data libraries. This report presents the outline of the FUBILA program, the measured data of the critical mass and the power distribution of a 9x9 Ref core, the first experimental core of the seven cores of FUBILA, and the core analysis by the continuous energy Monte Carlo calculation code MVP. (author)

  11. Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

    International Nuclear Information System (INIS)

    Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void

  12. Design and performance of 2D and 3D-shuffled breed-and-burn cores

    International Nuclear Information System (INIS)

    Highlights: • We report an in-depth study of breed-and-burn (B&B) reactor cores. • Principles for fuel, fuel assembly and core design of B&B systems were studied. • A concept of three-dimensional shuffling of fuel was introduced. • 3D-shuffling enables B&B operation with radiation damage below 350 DPA. - Abstract: The primary objective of this work is to find design approaches that will enable 3D fuel shuffling in stationary breed-and-burn (B&B) cores and to quantify the attainable reduction in peak DPA and change in additional performance characteristics going from conventional 2D to 3D fuel shuffling strategies. An additional objective is to establish the tradeoff between the minimum required DPA (displacements per atom) and average required burnup (fuel utilization) for B&B cores spanning a core power range from 1250 to 3500 MWth. It is found possible to design a B&B core fuelled with depleted uranium to have a peak radiation damage at or below 350 DPA when using 3D-shuffling. Relative to conventional 2D-shuffling, 3D-shuffling offers between 30% and 40% reduction in the peak DPA along with up to 30% increase in the average discharge burnup and, hence, in the depleted uranium utilization as well as significant increase in the core average and specific power density. Per DPA, the 3D shuffling option offers up to 60% higher uranium utilization. Even though 350 DPA is above the 200 DPA peak radiation damage HT9 steels were exposed to so far, it is below the 400 DPA advanced structural materials are expected to tolerate

  13. A neutronic feasibility study on the deep-burning of TRU in a commercial LWR core

    International Nuclear Information System (INIS)

    In this paper, the neutronic feasibility of deep burning of transuranics (TRU) nuclides in a commercial LWR reactor is investigated. New fuel assemblies having both TRU FCM (Fully Ceramic Micro-encapsulated) fuel pins and conventional UO2 pins are designed for 'deep burning' of the Transuranics without violating the safety-related parameters. In the FCM fuel pins, TRU TRISO particles and burnable poison BISO particle are distributed through a dense SiC matrix and formed into pellets and fuel pins. The combination of TRU FCM fuel pins and regular UO2 fuel pins are made into new fuel assemblies and loaded into the reload cores of the YongGwang unit 3 cycle 6 and its following cycles. The neutronic physics performance of the new fuel assemblies and the reload cores are analyzed in detail. The results show that the reload cores can be designed without significant changes in the safety-related parameters, and the equilibrium cycle core has self recycling capability for Transuranics and a high TRU destruction rate of ∼60%. (author)

  14. Sodium-cooled Fast Reactor Core Designs for the TRU burning with Thorium blanket

    International Nuclear Information System (INIS)

    In this study, the SFR(Sodium-cooled Fast Reactor) burner cores are designed with thorium blanket to have smaller burnup reactivity swing but higher TRU burning capability than the typical SFR burner cores using the TRU-U-10Zr fuel. Furthermore, we expect the SFR burner cores using thorium blanket have smaller coolant void reactivity because of the fact that the η-value increases much less with energy for 233U than for 239Pu and 232Th is less fissile than 238U. From the results, it is found that use of the thorium blanket both in inner and outer cores gives several desirable features such as the reduction of sodium void worth, small burnup reactivity swing but less negative Doppler coefficient and reduced control rod worth and that the use of thorium blanket only in the inner core gives much smaller sodium void worth but larger burnup reactivity swing than the cores using thorium blanket both in the inner and outer cores

  15. Analysis of Alaskan burn severity patterns using remotely sensed data

    Science.gov (United States)

    Duffy, P.A.; Epting, J.; Graham, J.M.; Rupp, T.S.; McGuire, A.D.

    2007-01-01

    Wildland fire is the dominant large-scale disturbance mechanism in the Alaskan boreal forest, and it strongly influences forest structure and function. In this research, patterns of burn severity in the Alaskan boreal forest are characterised using 24 fires. First, the relationship between burn severity and area burned is quantified using a linear regression. Second, the spatial correlation of burn severity as a function of topography is modelled using a variogram analysis. Finally, the relationship between vegetation type and spatial patterns of burn severity is quantified using linear models where variograms account for spatial correlation. These results show that: 1) average burn severity increases with the natural logarithm of the area of the wildfire, 2) burn severity is more variable in topographically complex landscapes than in flat landscapes, and 3) there is a significant relationship between burn severity and vegetation type in flat landscapes but not in topographically complex landscapes. These results strengthen the argument that differential flammability of vegetation exists in some boreal landscapes of Alaska. Additionally, these results suggest that through feedbacks between vegetation and burn severity, the distribution of forest vegetation through time is likely more stable in flat terrain than it is in areas with more complex topography. ?? IAWF 2007.

  16. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    JNES has been developing a technical data base used in reviewing validation of core analysis methods of LWRs in the occasions: (1) confirming core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical data base, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The data base will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects (1) measurements of Doppler reactivity in experimental MOX core simulating LWR cores, (2) measurement of isotopic compositions of fission product nuclides on high-burn up BWR UO2 fuels and (3) neutronics analysis of the experimental data that has been obtained in the international joint programs such as FUBILA and REBUS. (author)

  17. Cost analysis of a major burn.

    Science.gov (United States)

    Lofts, J A

    1991-11-27

    A retrospective review was undertaken of 26 patients admitted to Middlemore Hospital between January 1986 to July 1989 with burns totalling more than 30% of total body surface area. An attempt was made to estimate the total cost of successful inpatient management of a major burn using known and assumed values. The new schedule of interboard hospital charges was also employed for greater accuracy. The 20 survivors had a mean initial hospital stay of 68.7 days at a cost of between $37,077 and $40,702 (1989 values) and $46,069 (1991 values). This latter figure equates to an average cost of $647 per patient per day or, alternatively, $927 per % burn. Suggestions to reduce costs and improve treatment include: earlier excision and grafting; the establishment of a regional skin bank and keratinocyte culture facility to aid wound closure; and guidelines on antibiotic prescribing. PMID:1745459

  18. Thermal oscillations during carbon burning in an electron-degenerate stellar core

    International Nuclear Information System (INIS)

    Carbon-burning reactions in an electron-degenerate core of mass approaching the Chandrasekhar limit initiate a sequence of thermal oscillations. Prominent roles in driving convection near the center are played both by exothermic nuclear reactions and by the rearrangement of electrons in the Fermi sea that accompanies electron capture on 23Na. Convection provides the energy for driving electrons inward from the outer edge of the core, where they are created by the electron decay of 23Ne, to the center. During the high-temperature phase of an oscillation, the neutrinos emitted from regions near the center of the convective core as a consequence of electron capture are an order of magnitude more energetic than antineutrinos emitted from the edge of the core. During the rising temperature phase of an oscillation, the heating produced by exothermic nuclear reactions and by the electron Fermi sea in response to transfers of electrons to and from the sea is greater than the cooling due to Urca-neutrino losses. Rising temperatures cause an increase in the heating rate. The increased flux of energy forces an increase in the size of the convective core and the Urca-neutrino (neutrinloss rate also increases. Eventually the loss rate overtakes the heating rate temperatures in the convective core drop, and the core shrinks in size. The neutrino loss rate does not, at first, drop as rapidly as does the heating rate. Ultimately, however, the neutrino-loss rate drops below the heating rate, and a new heating phase sets in. Once central density reaches the threshold for electron capture on 21Ne, the convective zone breaks into two parts: a new, smaller core, in which 21Ne and 21F play the same role as did 23Na and 23Ne in the old core; and a convective shell whose edges are limited by the thresholds for electron captures on 23Na and21Ne. The shell continues to oscillate thermally

  19. Analysis of antibiotic consumption in burn patients.

    Science.gov (United States)

    Soleymanzadeh-Moghadam, Somayeh; Azimi, Leila; Amani, Laleh; Rastegar Lari, Aida; Alinejad, Faranak; Rastegar Lari, Abdolaziz

    2015-01-01

    Infection control is very important in burn care units, because burn wound infection is one of the main causes of morbidity and mortality among burn patients. Thus, the appropriate prescription of antibiotics can be helpful, but unreasonable prescription can have detrimental consequences, including greater expenses to patients and community alike. The aim of this study was to determine the effect of antibiotic therapy on the emergence of antibiotic-resistant bacteria. 525 strains of Pseudomonas aeruginosa, Acinetobacter baumannii and Staphylococcus aureus were isolated from 335 hospitalized burn patients. Antibiotic susceptibility tests were performed after identification the strains. The records of patients were audited to find the antibiotic used. The results indicated that P. aeruginosa is the most prevalent Gram-negative bacteria. Further, it showed a relation between abuse of antibiotics and emergence of antibiotic resistance. Control of resistance to antibiotics by appropriate prescription practices not only facilitates prevention of infection caused by multi-drug resistant (MDR) microorganisms, but it can also decrease the cost of treatment. PMID:26124986

  20. Analysis of antibiotic consumption in burn patients

    Directory of Open Access Journals (Sweden)

    Soleymanzadeh-Moghadam, Somayeh

    2015-06-01

    Full Text Available Infection control is very important in burn care units, because burn wound infection is one of the main causes of morbidity and mortality among burn patients. Thus, the appropriate prescription of antibiotics can be helpful, but unreasonable prescription can have detrimental consequences, including greater expenses to patients and community alike. The aim of this study was to determine the effect of antibiotic therapy on the emergence of antibiotic-resistant bacteria. 525 strains of and were isolated from 335 hospitalized burn patients. Antibiotic susceptibility tests were performed after identification the strains. The records of patients were audited to find the antibiotic used.The results indicated that is the most prevalent Gram-negative bacteria. Further, it showed a relation between abuse of antibiotics and emergence of antibiotic resistance. Control of resistance to antibiotics by appropriate prescription practices not only facilitates prevention of infection caused by multi-drug resistant (MDR microorganisms, but it can also decrease the cost of treatment.

  1. TRR-1/M1 Core Analysis with MVP

    International Nuclear Information System (INIS)

    Full text: Since early 1990s, the in-core fuel management of TRR-1/M1 has been performed by TRIGAP. This code was specifically developed for reactor physics calculations of the TRIGA-type reactor. However, because of its limitations in geometrical and cross sectional options, the attempt of using other techniques/codes are provoked. Nowadays, the choice of using the Monte Carlo method to perform core analysis becomes more satisfaction with acceptable computational time. The MVP is one of the codes that utilize the Monte Carlo method with continuous-energy library. It is able to explicitly model the problem in 3-D geometry. It also has a burn-up calculation feature called MVP-BURN. The aim of the current work is to apply the MVP code for TRR-1/M1 core analysis. In this paper, the MVP code was verified with the experiment results for the fresh core and some burn-up cores. The calculated-eigenvalue results agree well with the experimental data within an acceptable range of statistical error

  2. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  3. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease. (author)

  4. Research on burn control of core plasma with the transport code

    International Nuclear Information System (INIS)

    For the fusion reactors or experimental devices, one will be required to control several plasma parameters, like the fusion power, the heat flux, the neutron flux, the beta-value and so on. To control these parameters, many diagnostics and actuators are needed, but the diagnostics and actuators available in DEMO/commercial reactors are limited because of the high heat or neutron flux. For these reasons, to realize the fusion reactors, the construction of the reactor control logic is required. We are developing the burn control logic in the core plasma with a 1.5D transport code, and discussing on the relationship between control parameters and actuators. To demonstrate the feasibility of the core plasma control, we have demonstrated the simultaneous control of the fusion power and the safety factor profile with the gas-puff and NBI. (author)

  5. Plasma-wall interaction data needs critical to a Burning Core Experiment (BCX)

    International Nuclear Information System (INIS)

    The Division of Development and Technology has sponsored a four day US-Japan workshop ''Plasma-Wall Interaction Data Needs Critical to a Burning Core Experiment (BCX)'', held at Sandia National Laboratories, Livermore, California on June 24 to 27, 1985. The workshop, which brought together fifty scientists and engineers from the United States, Japan, Germany, and Canada, considered the plasma-material interaction and high heat flux (PMI/HHF) issues for the next generation of magnetic fusion energy devices, the Burning Core Experiment (BCX). Materials options were ranked, and a strategy for future PMI/HHF research was formulated. The foundation for international collaboration and coordination of this research was also established. This volume contains the first two of the five technical sessions. The first one being the BCX overview, the second on the BCX candidate materials. The remaining three sessions in volume two are on the plasma materials interaction issues, research facilities and small working group meeting on graphite, beryllium, advanced materials and future collaborations

  6. Plasma-wall interaction data needs critical to a Burning Core Experiment (BCX)

    Energy Technology Data Exchange (ETDEWEB)

    1985-11-01

    The Division of Development and Technology has sponsored a four day US-Japan workshop ''Plasma-Wall Interaction Data Needs Critical to a Burning Core Experiment (BCX)'', held at Sandia National Laboratories, Livermore, California on June 24 to 27, 1985. The workshop, which brought together fifty scientists and engineers from the United States, Japan, Germany, and Canada, considered the plasma-material interaction and high heat flux (PMI/HHF) issues for the next generation of magnetic fusion energy devices, the Burning Core Experiment (BCX). Materials options were ranked, and a strategy for future PMI/HHF research was formulated. The foundation for international collaboration and coordination of this research was also established. This volume contains the last three of the five technical sessions. The first of the three is on plasma materials interaction issues, the second is on research facilities and the third is from smaller working group meetings on graphite, beryllium, advanced materials and future collaborations.

  7. The micro-analysis of human burned bones: some remarks

    OpenAIRE

    Gonçalves, David

    2012-01-01

    The interdisciplinary research of burned bones is focused in this paper by presenting and discussing some methods that can assist the bioanthropologist in the analysis of this kind of remains. In particular, some techniques based on the histological structure of bone and on its molecular composition allow new ways of identifying burned human bone and of determining some aspects of the biological and ontological profile of an individual. A brief summary of those techniques is thus here...

  8. Application of integral ex-core and differential in-core neutron measurements for adjustment of fuel burn-up distributions in VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Borodkin, P. G.; Borodkin, G. I.; Khrennikov, N. N. [SEC NRS, Moscow, (Russian Federation)

    2009-07-01

    The paper deals with calculational and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Time-integrated neutron source distributions used for DORT calculations were prepared by two different approaches based on a) calculated fuel burn-up (standard routine procedure) and b) in-core measurements by means of SPD (self-powered detectors) and TC (thermocouples) in a new approach. Taking into account that fuel burn-up distributions in operating VVER may be evaluated now by analytical methods (calculations) it is only needed to develop new approaches for testing and correcting calculational evaluations. Results presented in this paper allow to consider a reverse task of alternative estimation of fuel burn-up distributions. The approach proposed is based on adjustment (fitting) of time-integrated neutron source distributions, and hence fuel burn-up patterns in some part of reactor core, on the base of ex-core neutron leakage measurement, neutron-physical calculation and in-core SPD and TC measurement data. (authors)

  9. The treatment of mixing in core helium burning models: I. Implications for asteroseismology

    CERN Document Server

    Constantino, Thomas; Christensen-Dalsgaard, Joergen; Lattanzio, John C; Stello, Dennis

    2015-01-01

    The detection of mixed oscillation modes offers a unique insight into the internal structure of core helium burning (CHeB) stars. The stellar structure during CHeB is very uncertain because the growth of the convective core, and/or the development of a semiconvection zone, is critically dependent on the treatment of convective boundaries. In this study we calculate a suite of stellar structure models and their non-radial pulsations to investigate why the predicted asymptotic g-mode $\\ell = 1$ period spacing $\\Delta\\Pi_1$ is systematically lower than is inferred from Kepler field stars. We find that only models with large convective cores, such as those calculated with our newly proposed "maximal-overshoot" scheme, can match the average $\\Delta\\Pi_1$ reported. However, we also find another possible solution that is related to the method used to determine $\\Delta\\Pi_1$: mode trapping can raise the observationally inferred $\\Delta\\Pi_1$ well above its true value. Even after accounting for these two proposed reso...

  10. Safety evaluation of LWR-Deep Burn core with 16x16 TRU FCM fuel

    International Nuclear Information System (INIS)

    The concept of destroying Transuranics (TRU) obtained from recycled fuel in a one-time irradiation (Deep Burn) using a commercial LWR was proposed by US Nuclear and Korea Atomic Energy Research Institute (KAERI). The LWR-DB concept adopts a fully ceramic micro-encapsulated (FCM) TRU fuel in order to achieve the required high burn-up and at the same time meet required safety by its fission product retentive and refractory features. In the FCM fuel TRISO (Tristructural-isotropic) particles contain the TRU fuel in a central kernel surrounded by multiple coatings. The TRISO particles are compacted in SiC matrix to make a TRU FCM pellet. TRU pellets are then inserted into Zr-alloy cladding to make a TRU rod which is compatible and can be used with existing LWR fuel assembly. A TRU fuel assembly for the LWR-DB consists of both TRU rods and conventional UO2 rods. The high thermal-conductivity of the SiC matrix also adds to the safety of FCM fuel. In this study, preliminary safety assessment of a proposed LWR-DB core is carried out using the MARS3.1 code. Reference plant is YongGwang unit 3, a System80-type PWR consisting of 177 16x16 fuel assemblies. Accident scenarios assessed are LOFA, a limiting transient for DNB and LBLOCA, a limiting transient for peak cladding temperature. (author)

  11. Analysis of Parallel Burn Without Crossfeed TSTO RLV Architectures and Comparison to Parallel Burn With Crossfeed and Series Burn Architectures

    Science.gov (United States)

    Smith, Garrett; Phillips, Alan

    2002-01-01

    There are currently three dominant TSTO class architectures. These are Series Burn (SB), Parallel Burn with crossfeed (PBw/cf), and Parallel Burn without crossfeed (PBncf). The goal of this study was to determine what factors uniquely affect PBncf architectures, how each of these factors interact, and to determine from a performance perspective whether a PBncf vehicle could be competitive with a PBw/cf or SB vehicle using equivalent technology and assumptions. In all cases, performance was evaluated on a relative basis for a fixed payload and mission by comparing gross and dry vehicle masses of a closed vehicle. Propellant combinations studied were LOX: LH2 propelled orbiter and booster (HH) and LOX: Kerosene booster with LOX: LH2 orbiter (KH). The study conclusions were: 1) a PBncf orbiter should be throttled as deeply as possible after launch until the staging point. 2) a detailed structural model is essential to accurate architecture analysis and evaluation. 3) a PBncf TSTO architecture is feasible for systems that stage at mach 7. 3a) HH architectures can achieve a mass growth relative to PBw/cf of ratio and to the position of the orbiter required to align the nozzle heights at liftoff. 5 ) thrust to weight ratios of 1.3 at liftoff and between 1.0 and 0.9 when staging at mach 7 appear to be close to ideal for PBncf vehicles. 6) performance for all vehicles studied is better when staged at mach 7 instead of mach 5. The study showed that a Series Burn architecture has the lowest gross mass for HH cases, and has the lowest dry mass for KH cases. The potential disadvantages of SB are the required use of an air-start for the orbiter engines and potential CG control issues. A Parallel Burn with crossfeed architecture solves both these problems, but the mechanics of a large bipropellant crossfeed system pose significant technical difficulties. Parallel Burn without crossfeed vehicles start both booster and orbiter engines on the ground and thus avoid both the risk of

  12. The treatment of mixing in core helium burning models -- II. Constraints from cluster star counts

    CERN Document Server

    Constantino, Thomas; Lattanzio, John C; van Duijneveldt, Adam

    2015-01-01

    The treatment of convective boundaries during core helium burning is a fundamental problem in stellar evolution calculations. In Paper~I we showed that new asteroseismic observations of these stars imply they have either very large convective cores or semiconvection/partially mixed zones that trap g-modes. We probe this mixing by inferring the relative lifetimes of asymptotic giant branch (AGB) and horizontal branch (HB) from $R_2$, the observed ratio of these stars in recent HST photometry of 48 Galactic globular clusters. Our new determinations of $R_2$ are more self-consistent than those of previous studies and our overall calculation of $R_2 = 0.117 \\pm 0.005$ is the most statistically robust now available. We also establish that the luminosity difference between the HB and the AGB clump is $\\Delta \\log{L}_\\text{HB}^\\text{AGB} = 0.455 \\pm 0.012$. Our results accord with earlier findings that standard models predict a lower $R_2$ than is observed. We demonstrate that the dominant sources of uncertainty in ...

  13. A neutronic study on advanced sodium cooled fast reactor cores with thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    Highlights: • SFR burner core configurations are explored and analyzed for effective use of thorium blankets. • Thorium blankets can significantly improve SFR burner core performances. • No recycling or partial recycling of Th blankets with multi-batches is very effective. - Abstract: In this paper, new design concepts of sodium cooled fast reactor (SFR) cores having thorium blanket are suggested for pursuing effective burning of TRU (transuranics) nuclides from LWR spent fuels and their neutronic performances are analyzed. Several core configurations having different arrangements of thorium blankets are explored to improve the core performances and safety-related parameters including sodium void worth which is one of main concerns on safety of SFR cores. Specifically, axial and radial thorium blankets are considered for two type cores. The first one is the typical annular type cores having two different fuel regions where axial thorium blankets are placed in the axially central regions while the second one is the single fuel region cores having central non-fuel region where the axial blanket and radial blankets are considered. Also, the effects of the recycling options and fuel management schemes of the used thorium blanket on the core performances are analyzed. The core performance analyses show that thorium blankets with no recycling option and multi-batch fuel management schemes are very effective to improve the core performances including burnup reactivity swing, sodium void worth and TRU consumption rate

  14. Effects of conversion ratio change on the core performances in medium to large TRU burning reactors

    International Nuclear Information System (INIS)

    Conceptual fast reactor core designs with sodium coolant are developed at 1,500, 3,000 and 4,500 MWt which are configured to transmute recycled transuranics (TRU) elements with external feeds consisting of LWR spent fuel. Even at each pre-determined power level, the performance parameters, reactivity coefficients and their implications on the safety analysis can be different when the target TRU conversion ratio changes. In order to address this aspect of design, a study on TRU conversion ratio change was performed. The results indicate that it is feasible to design a TRU burner core to accommodate a wide range of conversion ratios by employing different fuel cladding thicknesses. The TRU consumption rate is found to be proportional to the core power without any significant deterioration in the core performance at higher power levels. A low conversion ratio core has an increased TRU consumption rate and much faster burnup reactivity loss, which calls for appropriate means for reactivity compensation. As for the reactivity coefficients related with the conversion ratio change, the core with a low conversion ratio has a less negative Doppler coefficient, a more negative axial expansion coefficient, a more negative control rod worth per rod, a more negative radial expansion coefficient, a less positive sodium density coefficient and a less positive sodium void worth. A slight decrease in the delayed neutron fraction is also noted, reflecting the fertile U-238 fraction reduction. (author)

  15. Design of the core of a breed/burn fast reactor with the deterministic code KANEXT

    International Nuclear Information System (INIS)

    The breeding fast reactors are interesting because they generate more plutonium than they consume, however, the fuel has to be reprocessed for the generated plutonium is used in another reactor. In a breed/burn reactor (BBR) the plutonium is generated and used -in situ- inside the same reactor, reducing this way costs and the proliferation possibility. In this work, the core of a BBR was designed; cooled by sodium that consists of 210 active assemblies and 7 spaces for control rods, each assembly consists of 169 pines. The design differs from other BBR it includes a blanket in the reactor center. The above-mentioned was to take advantage of the fact by geometry that the population of fast and epithermal neutrons will be high in the area, due to the fissions in adjacent fissile areas. Favorable results were obtained, although not definitive with exchange scheme of spent fuel. Efforts should be made in the future to homogenize the power generation within the reactor and replace the spent assemblies more efficiently. (Author)

  16. Core analysis of the first cycle of Chashma nuclear power plant

    International Nuclear Information System (INIS)

    The up coming 300 MWe CHASHMA NPP will provide the opportunity to study the burn-up behavior of the fuel. Our experience is limited to the in-core fuel management studies when fuel burn-up remains within the design limits. The initial core is loaded in three regions with fuel of three different enrichments 2.4 w/o, 2.67 w/o and 3.0 w/o. It is intended to study the enhanced fuel burn-up vis-a-vis the expected cost benefit in due course of time. The core of the Chashma nuclear power plant is that of a typical PWR NPP of 300 MWe capacity. It has 121 fuel assemblies and all of them have identical external dimensions and hydraulic characteristics. The core height is 290 cm and equivalent diameter is 248·6 cm. The core is cooled and moderated by H2O and surrounded by a stainless steel baffle. Each fuel assembly consists of 15x15 rod array and the assembly pitch is 20·03 cm The average discharge burn-up is 30,000 MWd/MTU. Core analysis was carried out for the first cycle at hot full power (HFP). Two dimensional calculations were performed for burn-up analysis including core multiplication, flux distribution, burn-up length, isotopic inventory, peaking factor and critical boron concentration to achieve the economical fuel management within the constraints imposed by safe reactor operation. Calculations indicate that expected burn-up of the first cycle is 13479 MWd/MTU equivalent to 485 EFPD, with 25 ppm of boron is still in the system, which is very near to the design value. Similarly assembly power distribution, pin by pin power distribution and reactivity coefficients, calculated are within the acceptable limits. Efforts are on to improve further these calculations. (author)

  17. 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)

  18. Use of an Esophageal Heat Exchanger to Maintain Core Temperature during Burn Excisions and to Attenuate Pyrexia on the Burns Intensive Care Unit

    Directory of Open Access Journals (Sweden)

    David Williams

    2016-01-01

    Full Text Available Introduction. Burns patients are vulnerable to hyperthermia due to sepsis and SIRS and to hypothermia due to heat loss during excision surgery. Both states are associated with increased morbidity and mortality. We describe the first use of a novel esophageal heat exchange device in combination with a heater/cooler unit to manage perioperative hypothermia and postoperative pyrexia. Material and Methods. The device was used in three patients with full thickness burns of 51%, 49%, and 45% body surface area to reduce perioperative hypothermia during surgeries of >6 h duration and subsequently to control hyperthermia in one of the patients who developed pyrexia of 40°C on the 22nd postoperative day due to E. coli/Candida septicaemia which was unresponsive to conventional cooling strategies. Results. Perioperative core temperature was maintained at 37°C for all three patients, and it was possible to reduce ambient temperature to 26°C to increase comfort levels for the operating team. The core temperature of the pyrexial patient was reduced to 38.5°C within 2.5 h of instituting the device and maintained around this value thereafter. Conclusion. The device was easy to use with no adverse incidents and helped maintain normothermia in all cases.

  19. Use of an Esophageal Heat Exchanger to Maintain Core Temperature during Burn Excisions and to Attenuate Pyrexia on the Burns Intensive Care Unit.

    Science.gov (United States)

    Williams, David; Leslie, Gordon; Kyriazis, Dimitrios; O'Donovan, Benjamin; Bowes, Joanne; Dingley, John

    2016-01-01

    Introduction. Burns patients are vulnerable to hyperthermia due to sepsis and SIRS and to hypothermia due to heat loss during excision surgery. Both states are associated with increased morbidity and mortality. We describe the first use of a novel esophageal heat exchange device in combination with a heater/cooler unit to manage perioperative hypothermia and postoperative pyrexia. Material and Methods. The device was used in three patients with full thickness burns of 51%, 49%, and 45% body surface area to reduce perioperative hypothermia during surgeries of >6 h duration and subsequently to control hyperthermia in one of the patients who developed pyrexia of 40°C on the 22nd postoperative day due to E. coli/Candida septicaemia which was unresponsive to conventional cooling strategies. Results. Perioperative core temperature was maintained at 37°C for all three patients, and it was possible to reduce ambient temperature to 26°C to increase comfort levels for the operating team. The core temperature of the pyrexial patient was reduced to 38.5°C within 2.5 h of instituting the device and maintained around this value thereafter. Conclusion. The device was easy to use with no adverse incidents and helped maintain normothermia in all cases. PMID:27018074

  20. Performance and reactivity coefficient analysis of large TRU burning fast reactors

    International Nuclear Information System (INIS)

    In pursuit of size of economy, a core design study on higher powered TRU burning fast reactor has been performed. The selected core power levels subject to the core design development were 1500, 3000, and 4500 MWt. During the core design development, a set of the same design ground rules were applied simultaneously regardless of the core power levels, which enables comparisons in a consistent basis. In addition to the different core power levels, a variation in the TRU conversion ratio was also exercised at each power level. As a result, there became available a set of nine different core performance parameters and reactivity coefficients. The trending analysis for the developed cores having similar TRU conversion ratios showed that the core performance at a higher power level is as good as those of a lower powered one and a TRU transmutation rate is proportional to the core power level. As for the TRU loading, 1500 MWt is shown to be the lowest power level above which the relative neutron leakage compared with the neutron production becomes saturated. This means that a core power lower than 1500 MWt has a TRU loading penalty due to more neutron leakage. As the power level increases at a similar TRU conversion ratio, notable trends are as follows: a similar Doppler coefficient, a less negative axial expansion coefficient and a more negative radial expansion coefficient. Therefore, the power level dependent total reactivity change is negligible once the sodium void worth and the TRU enrichment are kept constant. Upon the conversion ratio change at each power level, a core with a reduced conversion ratio has a less negative Doppler coefficient, a more negative axial expansion coefficient, a more negative radial radial expansion coefficient and a less positive sodium density coefficient. As an effort to synthesize all the reactivity coefficients dependent on the core power level and TRU conversion ratio, a set of simple equations describing reactivity coefficient

  1. DEUTERIUM BURNING IN MASSIVE GIANT PLANETS AND LOW-MASS BROWN DWARFS FORMED BY CORE-NUCLEATED ACCRETION

    Energy Technology Data Exchange (ETDEWEB)

    Bodenheimer, Peter [UCO/Lick Observatory, Department of Astronomy and Astrophysics, University of California, Santa Cruz, CA 95064 (United States); D' Angelo, Gennaro; Lissauer, Jack J. [Space Science and Astrobiology Division, NASA-Ames Research Center, Moffett Field, CA 94035 (United States); Fortney, Jonathan J. [Department of Astronomy and Astrophysics, University of California, Santa Cruz, CA 95064 (United States); Saumon, Didier, E-mail: peter@ucolick.org, E-mail: gennaro.dangelo@nasa.gov, E-mail: Jack.J.Lissauer@nasa.gov, E-mail: jfortney@ucolick.org, E-mail: dsaumon@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2013-06-20

    Using detailed numerical simulations, we study the formation of bodies near the deuterium-burning limit according to the core-nucleated giant planet accretion scenario. The objects, with heavy-element cores in the range 5-30 M{sub Circled-Plus }, are assumed to accrete gas up to final masses of 10-15 Jupiter masses (M{sub Jup}). After the formation process, which lasts 1-5 Myr and which ends with a ''cold-start'', low-entropy configuration, the bodies evolve at constant mass up to an age of several Gyr. Deuterium burning via proton capture is included in the calculation, and we determined the mass, M{sub 50}, above which more than 50% of the initial deuterium is burned. This often-quoted borderline between giant planets and brown dwarfs is found to depend only slightly on parameters, such as core mass, stellar mass, formation location, solid surface density in the protoplanetary disk, disk viscosity, and dust opacity. The values for M{sub 50} fall in the range 11.6-13.6 M{sub Jup}, in agreement with previous determinations that do not take the formation process into account. For a given opacity law during the formation process, objects with higher core masses form more quickly. The result is higher entropy in the envelope at the completion of accretion, yielding lower values of M{sub 50}. For masses above M{sub 50}, during the deuterium-burning phase, objects expand and increase in luminosity by one to three orders of magnitude. Evolutionary tracks in the luminosity versus time diagram are compared with the observed position of the companion to Beta Pictoris.

  2. DEUTERIUM BURNING IN MASSIVE GIANT PLANETS AND LOW-MASS BROWN DWARFS FORMED BY CORE-NUCLEATED ACCRETION

    International Nuclear Information System (INIS)

    Using detailed numerical simulations, we study the formation of bodies near the deuterium-burning limit according to the core-nucleated giant planet accretion scenario. The objects, with heavy-element cores in the range 5-30 M⊕, are assumed to accrete gas up to final masses of 10-15 Jupiter masses (MJup). After the formation process, which lasts 1-5 Myr and which ends with a ''cold-start'', low-entropy configuration, the bodies evolve at constant mass up to an age of several Gyr. Deuterium burning via proton capture is included in the calculation, and we determined the mass, M50, above which more than 50% of the initial deuterium is burned. This often-quoted borderline between giant planets and brown dwarfs is found to depend only slightly on parameters, such as core mass, stellar mass, formation location, solid surface density in the protoplanetary disk, disk viscosity, and dust opacity. The values for M50 fall in the range 11.6-13.6 MJup, in agreement with previous determinations that do not take the formation process into account. For a given opacity law during the formation process, objects with higher core masses form more quickly. The result is higher entropy in the envelope at the completion of accretion, yielding lower values of M50. For masses above M50, during the deuterium-burning phase, objects expand and increase in luminosity by one to three orders of magnitude. Evolutionary tracks in the luminosity versus time diagram are compared with the observed position of the companion to Beta Pictoris.

  3. Deuterium Burning in Massive Giant Planets and Low-Mass Brown Dwarfs formed by Core-Nucleated Accretion

    CERN Document Server

    Bodenheimer, Peter; Lissauer, Jack J; Fortney, Jonathan J; Saumon, Didier

    2013-01-01

    Formation of bodies near the deuterium-burning limit is considered by detailed numerical simulations according to the core-nucleated giant planet accretion scenario. The objects, with heavy-element cores in the range 5-30 Mearth, are assumed to accrete gas up to final masses of 10-15 Jupiter masses (Mjup). After the formation process, which lasts 1-5 Myr and which ends with a 'cold-start', low-entropy configuration, the bodies evolve at constant mass up to an age of several Gyr. Deuterium burning via proton capture is included in the calculation, and we determined the mass, M50, above which more than 50% of the initial deuterium is burned. This often-quoted borderline between giant planets and brown dwarfs is found to depend only slightly on parameters, such as core mass, stellar mass, formation location, solid surface density in the protoplanetary disk, disk viscosity, and dust opacity. The values for M50 fall in the range 11.6-13.6 Mjup, in agreement with previous determinations that do not take the formati...

  4. Burn severity estimation using GeoEye imagery, object-based image analysis (OBIA), and Composite Burn Index (CBI) measurements

    Science.gov (United States)

    Dragozi, E.; Gitas, Ioannis Z.; Stavrakoudis, Dimitris G.; Minakou, C.

    2015-06-01

    Forest fires greatly influence the stability and functions of the forest ecosystems. The ever increasing need for accurate and detailed information regarding post-fire effects (burn severity) has led to several studies on the matter. In this study the combined use of Very High Resolution (VHR) satellite data (GeoEye), Objectbased image analysis (OBIA) and Composite Burn Index (CBI) measurements in estimating burn severity, at two different time points (2011 and 2012) is assessed. The accuracy of the produced maps was assessed and changes in burn severity between the two dates were detected using the post classification comparison approach. It was found that the produced burn severity map for 2011 was approximately 10% more accurate than that of 2012. This was mainly attributed to the increased heterogeneity of the study area in the second year, which led to an increased number of mixed class objects and consequently made it more difficult to spectrally discriminate between the severity classes. Following the post-classification analysis, the severity class changes were mainly attributed to the trees' ability to survive severe fire damage and sprout new leaves. Moreover, the results of the study suggest that when classifying CBI-based burn severity using VHR imagery it would be preferable to use images captured soon after the fire.

  5. A Core design study on the fuel displacement options for an effective transition between breakeven and TRU burning fast reactors

    International Nuclear Information System (INIS)

    A core design study to convert a breakeven core into a TRU burner is performed for a 600 MWe rated fast reactor. No change in the core and subassembly layouts is assumed, which only allows geometry variations within the fuel rods. Investigated alternatives are to use variable cladding thicknesses, smearing fraction adjustments and annular fuel rod concepts with a central liner of a variable diameter consisting of void, Zr, B4C, Al, W, etc. The variable cladding thickness concept could not be employed due to too high a clad inner wall temperature. A smearing fraction adjustment below a typical fraction of 75% leads to a moderate TRU burning and reduced sodium void worth, but to a relatively high burnup swing. Placing a central non-fuel rod with the fuel arranged in an annular ring affects the core performance and reactivity coefficients, depending on whether it is a moderator or an absorber. In general, candidate materials of high atomic numbers contribute to large positive sodium void worths, but enhanced negative expansion effects. Among the light elements, vanadium reveals a favourable performance with a comparable TRU burning and a reduced sodium void worth, suggesting this material can be regarded as a solid substitute for sodium. (authors)

  6. Structural safety analysis of HTGR core supports

    Energy Technology Data Exchange (ETDEWEB)

    Ju, F.; Bennett, J.G.; Anderson, C.A.

    1977-01-01

    In the current design of the High Temperature Gas-Cooled Reactor (HTGR), the core is made up of stacked columns of graphite fuel blocks. Structural support for the core takes the form of graphite columns beneath the core together with lateral springs, which position and restrain the core from contact with the sides of the reactor containment vessel. Each individual support column carries the dead load of several fuel columns together with the equivalent load caused by the coolant pressure drop through the core. The adequacy of the support structure to provide torsional stability of the core for both static and seismic loadings as well as long term stability of the core support structure itself is discussed. Analysis for long term stability of the core support columns involves consideration of eccentric loading (caused by damaged spherical seats) and imperfections in the form of surface cracks. Nonlinear graphite behavior must also be taken into consideration. For predictions of the core torsional seismic response, the core was represented as a right circular cylinder supported on elastic posts; the lateral support was represented by a single torsional spring. Energy losses from friction and material hysteresis were represented by viscous dampers. The coupled equations for vertical and rotational motions were integrated numerically and dynamic core response was computed fromtorsional acceleration time-histories obtained by differentiating a horizontal accelerogram and dividing by the shear wave speed for hard and soft soil conditions.

  7. Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope

    2012-07-01

    High temperature reactors (HTRs) have been under consideration for production of electricity, process heat, and for destruction of transuranics for decades. As part of the transmutation analysis efforts within the Fuel Cycle Research and Development (FCR&D) campaign, a need was identified for detailed discharge isotopics from HTRs for use in the VISION code. A conventional HTR using enriched uranium in UCO fuel was modeled having discharge burnup of 120 GWd/MTiHM. Also, a deep burn HTR (DB-HTR) was modeled burning transuranic (TRU)-only TRU-O2 fuel to a discharge burnup of 648 GWd/MTiHM. For each of these cases, unit cell depletion calculations were performed with SCALE/TRITON. Unit cells were used to perform this analysis using SCALE 6.1. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were first set by using Serpent calculations to match a spectral index between unit cell and whole core domains. In the case of the DB-HTR, the unit cell which was arrived at in this way conserved the ratio of fuel to moderator found in a single block of fuel. In the conventional HTR case, a larger moderator-to-fuel ratio than that of a single block was needed to simulate the whole core spectrum. Discharge isotopics (for 500 nuclides) and one-group cross-sections (for 1022 nuclides) were delivered to the transmutation analysis team. This report provides documentation for these calculations. In addition to the discharge isotopics, one-group cross-sections were provided for the full list of 1022 nuclides tracked in the transmutation library.

  8. Development of spectral history methods for pin-by-pin core analysis method using three-dimensional direct response matrix

    International Nuclear Information System (INIS)

    Spectral history methods for pin-by-pin core analysis method using the three-dimensional direct response matrix have been developed. The direct response matrix is formalized by four sub-response matrices in order to respond to a core eigenvalue k and thus can be recomposed at each outer iteration in the core analysis. For core analysis, it is necessary to take into account the burn-up effect related to spectral history. One of the methods is to evaluate the nodal burn-up spectrum obtained using the out-going neutron current. The other is to correct the fuel rod neutron production rates obtained the pin-by-pin correction. These spectral history methods were tested in a heterogeneous system. The test results show that the neutron multiplication factor error can be reduced by half during burn-up, the nodal neutron production rates errors can be reduced by 30% or more. The root-mean-square differences between the relative fuel rod neutron production rate distributions can be reduced within 1.1% error. This means that these methods can accurately reflect the effects of intra- and inter-assembly heterogeneities during burn-up and can be used for core analysis. Core analysis with the DRM method was carried out for an ABWR quarter core and it was found that both thermal power and coolant-flow distributions were smoothly converged. (authors)

  9. The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

  10. Verification of JUPITER standard analysis method for upgrading Joyo MK-III core design and management

    International Nuclear Information System (INIS)

    The verification of calculated core characteristics of the Joyo MK-III core using the JUPITER fast reactor standard analysis method was conducted by comparing with the measured values through the core physics tests. The purpose is to upgrade the core performance to increase the driver fuel burn-up and to increase the excess reactivity necessary for conducting various irradiation tests in the core region. Most of the C/Es are within 5% of unity. Through the comparisons, the calculation accuracy of the JUPITER standard analysis method for a small size sodium cooled fast reactor with a hard neutron spectrum like Joyo was clarified. As a result of this study, more irradiation tests can be performed with appropriate safety margin and the efficient core and fuel management can be achieved to save the number of refueling. (author)

  11. A method for analysis of vanillic acid in polar ice cores

    Science.gov (United States)

    Grieman, M. M.; Greaves, J.; Saltzman, E. S.

    2015-02-01

    Biomass burning generates a wide range of organic compounds that are transported via aerosols to the polar ice sheets. Vanillic acid is a product of conifer lignin combustion, which has previously been observed in laboratory and ambient biomass burning aerosols. In this study a method was developed for analysis of vanillic acid in melted polar ice core samples. Vanillic acid was chromatographically separated using reversed-phase liquid chromatography (HPLC) and detected using electrospray ionization-triple quadrupole mass spectrometry (ESI-MS/MS). Using a 100 μL injection loop and analysis time of 4 min, we obtained a detection limit of 77 ppt (parts per trillion by mass) and an analytical precision of ±10%. Measurements of vanillic acid in Arctic ice core samples from the Siberian Akademii Nauk core are shown as an example application of the method.

  12. Systems Analysis of a Compact Next Step Burning Plasma Experiment

    Energy Technology Data Exchange (ETDEWEB)

    S.C. Jardin; C.E. Kessel; D. Meade; C. Neumeyer

    2002-02-06

    A new burning plasma systems code (BPSC) has been developed for analysis of a next step compact burning plasma experiment with copper-alloy magnet technology. We consider two classes of configurations: Type A, with the toroidal field (TF) coils and ohmic heating (OH) coils unlinked, and Type B, with the TF and OH coils linked. We obtain curves of the minimizing major radius as a function of aspect ratio R(A) for each configuration type for typical parameters. These curves represent, to first order, cost minimizing curves, assuming that device cost is a function of major radius. The Type B curves always lie below the Type A curves for the same physics parameters, indicating that they lead to a more compact design. This follows from that fact that a high fraction of the inner region, r < R-a, contains electrical conductor material. However, the fact that the Type A OH and TF magnets are not linked presents fewer engineering challenges and should lead to a more reliable design. Both the Type A and Type B curves have a minimum in major radius R at a minimizing aspect ratio A typically above 2.8 and at high values of magnetic field B above 10 T. The minimizing A occurs at larger values for longer pulse and higher performance devices. The larger A and higher B design points also have the feature that the ratio of the discharge time to the current redistribution time is largest so that steady-state operation can be more realistically prototyped. A sensitivity study is presented for the baseline Type A configuration showing the dependence of the results on the parameters held fixed for the minimization study.

  13. Overview on Hydrate Coring, Handling and Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jon Burger; Deepak Gupta; Patrick Jacobs; John Shillinglaw

    2003-06-30

    Gas hydrates are crystalline, ice-like compounds of gas and water molecules that are formed under certain thermodynamic conditions. Hydrate deposits occur naturally within ocean sediments just below the sea floor at temperatures and pressures existing below about 500 meters water depth. Gas hydrate is also stable in conjunction with the permafrost in the Arctic. Most marine gas hydrate is formed of microbially generated gas. It binds huge amounts of methane into the sediments. Worldwide, gas hydrate is estimated to hold about 1016 kg of organic carbon in the form of methane (Kvenvolden et al., 1993). Gas hydrate is one of the fossil fuel resources that is yet untapped, but may play a major role in meeting the energy challenge of this century. In June 2002, Westport Technology Center was requested by the Department of Energy (DOE) to prepare a ''Best Practices Manual on Gas Hydrate Coring, Handling and Analysis'' under Award No. DE-FC26-02NT41327. The scope of the task was specifically targeted for coring sediments with hydrates in Alaska, the Gulf of Mexico (GOM) and from the present Ocean Drilling Program (ODP) drillship. The specific subjects under this scope were defined in 3 stages as follows: Stage 1: Collect information on coring sediments with hydrates, core handling, core preservation, sample transportation, analysis of the core, and long term preservation. Stage 2: Provide copies of the first draft to a list of experts and stakeholders designated by DOE. Stage 3: Produce a second draft of the manual with benefit of input from external review for delivery. The manual provides an overview of existing information available in the published literature and reports on coring, analysis, preservation and transport of gas hydrates for laboratory analysis as of June 2003. The manual was delivered as draft version 3 to the DOE Project Manager for distribution in July 2003. This Final Report is provided for records purposes.

  14. Design study on BN-600 hybrid core. 1. Evaluation of core neutronic and thermalhydraulic characteristics by Japanese analysis methods

    International Nuclear Information System (INIS)

    A program of disposition of Russian weapon-grade plutonium by containing the plutonium in vibropacked MOX fuel subassemblies and burning them in the BN-600 hybrid reactor core has been progressed. The relevant design works on the BN-600 hybrid core were carried out under the contract between Japan Nuclear Cycle Development Institute (JNC) and OKB Mechanical Engineering (OKBM), Russian public enterprise, and completed. JNC obtained a series of design technical reports. Japanese analysis methods were adopted to evaluate neutronic and thermal-hydraulic characteristics of the BN-600 hybrid core, based on the design technical data described in the obtained reports. The evaluation results of the key core performances, such as maximum linear heat rate, core-averaged discharge burnup, sodium void reactivity, capability of disposition of weapon-grade plutonium and, and reactivity control balance, were found to satisfy the design criteria and/or targets provided by Russia, and meet the Russian rule. The results of this study showed that the core and fuel specifications determined by Russia can be considered reasonable and proper in terms of neutronic and thermal-hydraulic design, and that the Japanese analysis methods are expected to contribute to increasing reliability of the Russian design works. (author)

  15. BWR core stability analysis with RETRAN

    International Nuclear Information System (INIS)

    A RETRAN model was developed for determining the stability of boiling water reactor. This model was benchmarked against plant data from stability tests conducted during plant operations. The stability analysis with RETRAN is demonstrated using best estimate RETRAN input data representative of the NSSS. All of the important neutronic and thermal hydraulic feedback mechanisms are taken into account through the modeling of the reactor vessel, recirculation loops and core neutronics. The analysis was performed with the RETRAN02 MOD003 computer code. The transient is initialized by a small step decrease in the steam dome pressure. The core exit (upper plenum) pressure and core power transient responses to this perturbation are transformed into frequency data and a system transfer function is then obtained. The system transfer function is fitted to a second order equation from which the decay ratio and natural frequency can be determined

  16. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Directory of Open Access Journals (Sweden)

    Borodkin Pavel

    2016-01-01

    Full Text Available Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  17. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Science.gov (United States)

    Borodkin, Pavel; Borodkin, Gennady; Khrennikov, Nikolay

    2016-02-01

    Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  18. Variants of WWER-440 in-core fuel management with low burned assemblies reuse

    International Nuclear Information System (INIS)

    Not concerning technical and regulatory problems of the unburned fuel transition from stopped to working WWER-440 reactors in the paper are considered only ideas of the methodical approach to mixed core loadings formation with purpose to reach the residual fuel design burnup at preservation standard operational reactor parameters. With this purpose a common model is developed on the basis of a steady state mode of reloading with three cycles of use for fuel assemblies with enrichment of 3.6%. So at the end of each cycle 1/3 assemblies are removed from reactor core and each assembly with enrichment of 3.6% works 3 cycles. In the presented model it is supposed also to use fuel assemblies with enrichment of 2.4% working 2 cycles and with enrichment of 1.6% - operated one cycle. In the model it is supposed also that for this time of operation the assemblies of the corresponding enrichment reach the design burnup

  19. A transition cycle strategy to enhance minor actinide burning potential in the pan-shape LMR core

    International Nuclear Information System (INIS)

    This study summarizes the neutronic performances and fuel cycle behaviors of the pan-shape transuranic (TRU) burner core from the initial core through the end of a core life. The cycle-by-cycle evolution of isotopic compositions and neutronics characteristics are compared with those calculated from the analysis of an assumed equilibrium cycle. The amount of burnt TRU per cycle after Cycle 8 turned out to be comparable to that of the equilibrium cycle, while the isotopic compositions and the resulting neutronics performances up to about Cycle 20 have shown considerable deviations from those of the equilibrium cycle. The reference core in this analysis has been designed to meet a target sodium void reactivity at the end of the equilibrium cycle by reducing the active core height. Since the core isotopic loading approaches that of the equilibrium cycle after many cycles of operation, significant margins to the target sodium void reactivity are noted in the early cycles. This finding has led to the loading of concentrated minor actinides (MA) relative to the Pu isotopes in the first three cycles. Thereafter, they are homogeneously self-recycled with the external feed TRU makeup composed of typical LWR discharge TRU compositions. The transition cycle analysis with the higher MA loading reveals that the total MA consumed through 50 cycles of operation is 1.89 times larger than the case for the constant external feed makeup TRU with a typical LWR discharge compositions, without exceeding the sodium void reactivity observed in the equilibrium cycle

  20. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    OpenAIRE

    Borodkin Pavel; Borodkin Gennady; Khrennikov Nikolay

    2016-01-01

    Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new...

  1. CFD Analysis of Core Bypass Phenomena

    International Nuclear Information System (INIS)

    The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the one-twelfth grid can be set as a symmetry boundary

  2. CFD Analysis of Core Bypass Phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Richard W. Johnson; Hiroyuki Sato; Richard R. Schultz

    2010-03-01

    The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the sector grid can be set as a symmetry boundary

  3. CFD Analysis of Core Bypass Phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Richard W. Johnson; Hiroyuki Sato; Richard R. Schultz

    2009-11-01

    The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the one-twelfth grid can be set as a symmetry boundary

  4. SEDIMENT CORE SAMPLING AND ANALYSIS OF KAW LAKE

    Directory of Open Access Journals (Sweden)

    Dejene Alemayehu

    2014-01-01

    Full Text Available The Kaw Nation and Black land Research Center in July 2012 conducted a sediment core sampling from Kaw Lake. Kaw Lake is a reservoir constructed in 1976 by the Army Corps of Engineers for the purpose of water supply and recreation. It is located 11 miles east of Ponca City, Kay County, Oklahoma. This reservoir covers approximately 17,040 acres (69 km2 and is also known to be the seventh largest lake in Oklahoma by surface area. This lake holds 428,600 acre feet (528,700, 00 m3 of water which is said to be the ninth largest lake in Oklahoma. The lake is fed by the Arkansas River that flows from Colorado, through kansas and into Kaw Lake. The Arkansas River flows through Kaw Lake shared by several small creeks and empties into the Mississippi River. The purpose of the study was to determine the rate of sediment accumulation and examine the level of nutrient and heavy metals accumulation or deposition at the bottom of the lake. Four core samples from different parts of the lake at different depth were sampled for analysis. Each core sample was sectioned into 20 cm, dried and ground into homogenous powder. Samples from each section were tested for organic carbon content and heavy metals. Organic carbon content was verified by burning through a muffle furnace, while the remaining core samples were digested into a solution and ran through an Atomic Absorption Spectrophotometer (AAS to evaluate the concentration of heavy metals. Particle size analyses were also determined. Results were organized by depth, concentration, particle size distribution and bulk density. Data showed phosphorous and some heavy metals concentrations at core 3 and 4 were higher than core 1 and 2. Phosphorous concentration at four sediment core sites ranged from 350 mg kg-1 to 550 mg kg-1. Whereas Aluminum concentration was 40,000 mg kg-1 to 70,000 mg kg-1, Barium 280 mg kg-1 to 420 mg kg-1

  5. Power excursion analysis for high burnup cores

    International Nuclear Information System (INIS)

    A study was undertaken of power excursions in high burnup cores. There were three objectives in this study. One was to identify boiling water reactor (BWR) and pressurized water reactor (PWR) transients in which there is significant energy deposition in the fuel. Another was to analyze the response of BWRs to the rod drop accident (RDA) and other transients in which there is a power excursion. The last objective was to investigate the sources of uncertainty in the RDA analysis. In a boiling water reactor, the events identified as having significant energy deposition in the fuel were a rod drop accident, a recirculation flow control failure, and the overpressure events; in a pressurized water reactor, they were a rod ejection accident and boron dilution events. The RDA analysis was done with RAMONA-4B, a computer code that models the space- dependent neutron kinetics throughout the core along with the thermal hydraulics in the core, vessel, and steamline. The results showed that the calculated maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important uncertainties in each of these categories are discussed in the report

  6. Evaluation of core compositions for use in breed and burn reactors and limited-separations fuel cycles

    International Nuclear Information System (INIS)

    Highlights: ► Calculated minimum burnup and irradiation damage for B and B reactor compositions. ► Computed doubling time of fuel cycles using B and B reactors and no chemical separations. ► Determined sensitivity of doubling time to using melt refining vs. direct reuse. ► Examined tradeoff between power density and neutronics for different coolants. - Abstract: Previously developed methods for analyzing breed-and-burn (B and B) reactors are applied to a wide range of core compositions. The compositions studied include different fuel types, steel and silicon carbide structure, and sodium, lead/lead bismuth eutectic (LBE), and gas coolants. These compositions are evaluated for use in “minimum burnup” B and B reactors in which it is assumed that blocks comprising the core can be shuffled in all three dimensions to flatten out non-uniformities in burnup. The two figures of merit evaluated are the minimum irradiation damage requirement and reactor fleet doubling time. To minimize irradiation damage, gas coolants perform best, followed by lead/LBE then sodium. High uranium-content metal fuel outperforms compound fuels, and different types of steel are similar and perform slightly better than silicon carbide. Once-through irradiation damage requirements can be surprisingly modest in minimum burnup B and B reactors, with a wide range of compositions viable at irradiation damage levels 50% higher than existing materials data. Doubling times were calculated for a reactor fleet consisting of B and B reactors operating in a limited-separations fuel cycle; i.e., a fuel cycle with no chemical separation of actinides. The effects of different cooling times and removal of fission products using a melt refining process are evaluated. To minimize doubling time, sodium cooled compositions perform best because they are able to achieve core power densities several times larger than compositions using other coolants. A hypothetical sodium-cooled core composition with high

  7. Design study on BN-600 hybrid core. 2. Evaluation of fuel integrity and core neutronic characteristics by Japanese analysis methods

    International Nuclear Information System (INIS)

    A program of disposal of Russian surplus weapon-grade plutonium by containing the plutonium in vibropacked MOX fuel subassemblies and burning them in the BN-600 hybrid reactor core has been progressed. The relevant design works on the BN-600 hybrid core have been carried out under the contract between Japan Nuclear Cycle Development Institute (JNC) and OKB Mechanical Engineering (OKBM), Russian public enterprise. JNC obtained a series of design technical reports. Japanese analysis methods were adopted to evaluate fuel integrity in the design basis transients and neutronic characteristics of the BN-600 hybrid core, based on the design technical data described in the obtained reports. The evaluation results of the key performances, such as maximum cladding and fuel temperatures, coolant (sodium) void reactivity, reactivity coefficient, were found to satisfy the design criteria and/or target provided by Russia, and meet the Russian rule. The results of this study showed that the core and fuel specifications determined by Russia can be considered reasonable and proper from the viewpoint of safety and neutronic designs, and that the Japanese analysis methods are expected to contribute to increasing reliability of the Russian design works. (author)

  8. The chemical composition of White Dwarfs as a test of convective efficiency during core He-burning

    CERN Document Server

    Straniero, O; Imbriani, G; Piersanti, L; Straniero, Oscar; Dominguez, Inma; Imbriani, Gianluca; Piersanti, Luciano

    2003-01-01

    Pulsating white dwarfs provide constraints to the evolution of progenitor stars. We revise He-burning stellar models, with particular attention to core convection and to its connection with the nuclear reactions powering energy generation and chemical evolution Theoretical results are compared to the available measurements for the variable white dwarf GD 358, which indicate a rather large abundance of central oxygen. We show that the attempt to constrain the relevant nuclear reaction rate by means of the white dwarf composition is faced with a large degree of uncertainty related to evaluating the efficiency of convection-induced mixing.By combining the uncertainty of the convection theory with the error on the relevant reaction rate we derive that the present theoretical prediction for the central oxygen mass fraction in white dwarfs varies between 0.3 and 0.9. Unlike previous claims, we find that models taking into account semiconvection and a moderate C12(alpha,gamma)O16 reaction rate are able to account fo...

  9. Albumin in Burn Shock Resuscitation: A Meta-Analysis of Controlled Clinical Studies

    OpenAIRE

    Navickis, Roberta J.; Greenhalgh, David G; Wilkes, Mahlon M.

    2016-01-01

    Critical appraisal of outcomes after burn shock resuscitation with albumin has previously been restricted to small relatively old randomized trials, some with high risk of bias. Extensive recent data from nonrandomized studies assessing the use of albumin can potentially reduce bias and add precision. The objective of this meta-analysis was to determine the effect of burn shock resuscitation with albumin on mortality and morbidity in adult patients. Randomized and nonrandomized controlled cli...

  10. Overview of the South American biomass burning analysis (SAMBBA) field experiment

    Science.gov (United States)

    Morgan, W. T.; Allan, J. D.; Flynn, M.; Darbyshire, E.; Hodgson, A.; Johnson, B. T.; Haywood, J. M.; Freitas, S.; Longo, K.; Artaxo, P.; Coe, H.

    2013-05-01

    Biomass burning represents one of the largest sources of particulate matter to the atmosphere, which results in a significant perturbation to the Earth's radiative balance coupled with serious negative impacts on public health. Globally, biomass burning aerosols are thought to exert a small warming effect of 0.03 Wm-2, however the uncertainty is 4 times greater than the central estimate. On regional scales, the impact is substantially greater, particularly in areas such as the Amazon Basin where large, intense and frequent burning occurs on an annual basis for several months (usually from August-October). Furthermore, a growing number of people live within the Amazon region, which means that they are subject to the deleterious effects on their health from exposure to substantial volumes of polluted air. Initial results from the South American Biomass Burning Analysis (SAMBBA) field experiment, which took place during September and October 2012 over Brazil, are presented here. A suite of instrumentation was flown on-board the UK Facility for Airborne Atmospheric Measurement (FAAM) BAe-146 research aircraft and was supported by ground based measurements, with extensive measurements made in Porto Velho, Rondonia. The aircraft sampled a range of conditions with sampling of fresh biomass burning plumes, regional haze and elevated biomass burning layers within the free troposphere. The physical, chemical and optical properties of the aerosols across the region will be characterized in order to establish the impact of biomass burning on regional air quality, weather and climate.

  11. Core dimensions of recovery: a psychometric analysis.

    Science.gov (United States)

    Gordon, Sarah E; Ellis, Pete M; Siegert, Richard J; Walkey, Frank H

    2014-07-01

    Core recovery dimensions lie between the large general factor of recovery and its underlying components. Identifying these could enhance recovery frameworks, practice and research. In contrast to existing conceptually based taxonomies, we sought to empirically identify the core dimensions of recovery through further psychometric analysis of a robust eleven factor (sub-scale) consumer recovery outcome measure, My Voice, My Life. We subjected the sub-scale scores of 504 consumers to further principal components analyses, beginning with a single unrotated factor and progressing through two to nine factors with varimax rotation. We found the five-factor solution to provide an orderly intermediate configuration with the eleven recovery factors having either aligned and/or disengaged through the process to result in the following core dimensions: (1) Belonging and relating (encompassing the individual factors of spirituality, culture, and relationships); (2) Being and doing (encompassing the individual factors of physical health, day-to-day life, and quality of life); (3) Thinking and feeling (encompassing the individual factors of recovery, mental health, and hope and empowerment); (4) Resources (which maintained its independence); and (5) Satisfaction with Services (which also maintained its independence). We compare this empirical configuration with conceptually based taxonomies. PMID:23588506

  12. Computation system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals

  13. Maximization of burning and/or transmutation (B/T) capacity in coupled spectrum reactor (CSR) by fuel and core adjustment

    International Nuclear Information System (INIS)

    A conceptual design of burning and/or transmutation (B/T) reactor, based on a modified conventional 1150 MWe-PWR system, consisted of two core regions for thermal and fast neutrons, respectively, was proposed herein for the treatments of minor actinides (MA). In the outer region 237Np, 241Am, and 243Am burned by thermal neutrons, while in the inner region 244Cm was burned mainly by fast neutrons. The geometry of B/T fuel in the outer region was left the same with that of PWR, while in the inner region the B/T fuel was arranged in a tight-lattice geometry that allowed a higher fuel to coolant volume ratio. The maximization of B/T capacity in CSR were done by, first, increasing the radius of the inner region. Second, reducing the coolant to fuel volume ratio, and third, choosing a suitable B/T fuel type. The result of the calculations showed that the equilibrium of main isotopes in CSR can be achieved after about 5 recycle stages. This study also showed that the CSR can burn and transmute up to 808 kg of MA in a single reactor core effectively and safely. (author)

  14. Global analysis of the persistence of the spectral signal associated with burned areas

    Science.gov (United States)

    Melchiorre, A.; Boschetti, L.

    2015-12-01

    Systematic global burned area maps at coarse spatial resolution (350 m - 1 km) have been produced in the past two decades from several Earth Observation (EO) systems (including MODIS, Spot-VGT, AVHRR, MERIS), and have been extensively used in a variety of applications related to emissions estimation, fire ecology, and vegetation monitoring (Mouillot et al. 2014). There is however a strong need for moderate to high resolution (10-30 m) global burned area maps, in order to improve emission estimations, in particular on heterogeneous landscapes and for local scale air quality applications, for fire management and environmental restoration, and in support of carbon accounting (Hyer and Reid 2009; Mouillot et al. 2014; Randerson et al. 2012). Fires causes a non-permanent land cover change: the ash and charcoal left by the fire can be visible for a period ranging from a few weeks in savannas and grasslands ecosystems, to over a year in forest ecosystems (Roy et al. 2010). This poses a major challenge for designing a global burned area mapping system from moderate resolution (10-30 m) EO data, due to the low revisit time frequency of the satellites (Boschetti et al. 2015). As a consequence, a quantitative assessment of the permanence of the spectral signature of burned areas at global scale is a necessary step to assess the feasibility of global burned area mapping with moderate resolution sensors. This study presents a global analysis of the post-fire reflectance of burned areas, using the MODIS MCD45A1 global burned area product to identify the location and timing of burning, and the MO(Y)D09 global surface reflectance product to retrieve the time series of reflectance values after the fire. The result is a spatially explicit map of persistence of burned area signal, which is then summarized by landcover type, and by fire zone using the subcontinental regions defined by Giglio et al. (2006).

  15. Burns in diabetic patients

    OpenAIRE

    Maghsoudi, Hemmat; Aghamohammadzadeh, Naser; Khalili, Nasim

    2008-01-01

    CONTEXT AND AIMS: Diabetic burn patients comprise a significant population in burn centers. The purpose of this study was to determine the demographic characteristics of diabetic burn patients. MATERIALS AND METHODS: Prospective data were collected on 94 diabetic burn patients between March 20, 2000 and March 20, 2006. Of 3062 burns patients, 94 (3.1%) had diabetes; these patients were compared with 2968 nondiabetic patients with burns. Statistical analysis was performed using the statistical...

  16. Geologic analysis of Devonian Shale cores

    Energy Technology Data Exchange (ETDEWEB)

    None

    1982-02-01

    Cleveland Cliffs Iron Company was awarded a DOE contract in December 1977 for field retrieval and laboratory analysis of cores from the Devonian shales of the following eleven states: Michigan, Illinois, Indiana, Ohio, New York, Pennsylvania, West Virginia, Maryland, Kentucky, Tennessee and Virginia. The purpose of this project is to explore these areas to determine the amount of natural gas being produced from the Devonian shales. The physical properties testing of the rock specimens were performed under subcontract at Michigan Technological University (MTU). The study also included LANDSAT information, geochemical research, structural sedimentary and tectonic data. Following the introduction, and background of the project this report covers the following: field retrieval procedures; laboratory procedures; geologic analysis (by state); references and appendices. (ATT)

  17. Research on Mechanism of Paper Burning by Thermogravimetric Analysis

    Institute of Scientific and Technical Information of China (English)

    QIN Da; HAN Xingzhou; WANG Xiaoguang; QI Fengliang; WANG Zijie; GUO Zihan; HAO Hongguang

    2015-01-01

    The examination of charred document is a challenge and usually requires a careful application of certain scientific techniques due to its unstable property. To address this issue, the mechanism of paper burning was studied in this paper. Here thermal-gravimetry (TG) was applied to investigate five kinds of paper, along with their TG and derivative thermogravimetric curve (DTG) observed at different atmospheric conditions. The results showed that the shape of curves, albeit similar, varied with the physical and chemical composition of paper. In the burning process, dehydration and de-polymerization are the two main pathways for cellulose, the major ingredient of paper. The heating rate indicated little influence on the curves while the sort of atmosphere worked strongly. The reason is due to the lack of tar oxidation when nitrogen used as the atmospheric environment. At moderate temperature, de-polymerization prevails and the tar can be observed. With temperature increasing, the tar and cellulose are further decomposed, leading to products of high boiling-point. According to the results, the charred document can be classified as one of the dehydrated, tarred, charred and ashed. Except the ashed stage, the other three can be handled and the writing whereon can be deciphered. The results exposed hereof may provide a fundamental for examining and deciphering charred document.

  18. PWR Core 2 Project accident analysis

    International Nuclear Information System (INIS)

    The various operations required for receipt, handling, defueling and storage of spent Shippingport PWR Core 2 fuel assemblies have been evaluated to determine the potential accidents and their consequences. These operations will introduce approximately 16,500 kilograms of depleted natural uranium (as UO2), 139 kilograms of plutonium, 2.8 megacuries of mixed fission products, and 14 kilograms of Zircaloy-4 (cladding and hardware) into the 221-T Canyon Building. Event sequences for potential accidents that were considered included (1) leaking fuel assemblies, (2) fire and explosion, (3) loss of coolant or cooling capability, (4) dropped and/or damaged fuel assemblies, and extrinsic occurrences such as loss of services, missile impact, and natural occurrences (e.g., earthquake, tornado). Accident frequencies were determined by formal analysis to be very low. Accident consequences are greatly mitigated by the safety and containment features designed into the fuel modules and shipping cask, the long cooling time since reactor discharge, and the redundant safety features designed into the facilities, equipment, and operating procedures for the PWR Core 2 Project. Possible hazards associated with the handling of these fuels have been considered and adequate safeguards and storage constraints identified. The operations of M-160 cask unloading and module storage will not involve identifiable risks as great or significantly greater than those for comparable licensed nuclear facilities, nor will hazards or risks be significantly different from comparable past 221-T Plant programs. Therefore, it is concluded that the operations required for receipt, handling, and defueling of the M-160 cask and for the storage and surveillance of the PWR Core 2 fuel assemblies at the 221-T Canyon Building can be performed without undue risk to the safety of the involved personnel, the public, the environment or the facility

  19. Core Backbone Convergence Mechanisms and Microloops Analysis

    Directory of Open Access Journals (Sweden)

    Abdelali Ala

    2012-07-01

    Full Text Available In this article we study approaches that can be used to minimise the convergence time, we also make a focus on microloops phenomenon, analysis and means to mitigate them. The convergence time reflects the time required by a network to react to a failure of a link or a router failure itself. When all nodes (routers have updated their respective routing and forwarding databases, we can say the network has converged. This study will help in building real-time and resilient network infrastructure, the goal is to make any evenement in the core network, as transparent as possible to any sensitive and real-time flows. This study is also, a deepening of earlier works presented in [10] and [11].

  20. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO2 and MOX fuel rods, (3) analysis of isotopic composition data for UO2 and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  1. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    JNES has been developing a technical data base used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical data base, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The data base will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects (1) analysis of the measurement data of Doppler reactivity in experimental MOX core simulating LWR cores, (2) measurements of isotopic compositions of fission product nuclides on high-burnup BWR UO2 fuels and the analysis of the measurement data, and (3) neutronics analysis of the experimental data that has been obtained in the international joint programs such as FUBILA and REBUS. (author)

  2. The Clinical Analysis of 47 Patients with Burning Mouth Syndrome

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    目的:研究灼口综合征的临床特征及精神因素分析.方法:对47例灼口综合征患者的临床资料进行分析并用HAD记分方法对其精神因素进行评估.结果:BMS患者中女性患病率高,男女比率1:6.5,舌部发病最为多见,平均发病年龄为53岁,BMS患者均有不同程度的精神因素的影响.%Objective:To analyse the clinical nature and psychiatric assessment.Methods:47 patients with burning mouth syndrome (BMS) were investigated.The clinical nature including distribution of orafacial sites affected by BMS and psychiatric assessment by the way of the Hospital Anxiety and Depression (HAD) scale were determined.Result:BMS affected females much more frequently than males, with a ratio of about 6.5∶1.The mean age of patients was around 53 years old. The most often affected site was the tongue and anxiety was a more important psychological factor in BMS than depression.

  3. Multi-Core Processor Memory Contention Benchmark Analysis Case Study

    Science.gov (United States)

    Simon, Tyler; McGalliard, James

    2009-01-01

    Multi-core processors dominate current mainframe, server, and high performance computing (HPC) systems. This paper provides synthetic kernel and natural benchmark results from an HPC system at the NASA Goddard Space Flight Center that illustrate the performance impacts of multi-core (dual- and quad-core) vs. single core processor systems. Analysis of processor design, application source code, and synthetic and natural test results all indicate that multi-core processors can suffer from significant memory subsystem contention compared to similar single-core processors.

  4. BN-600 full MOX core benchmark analysis

    International Nuclear Information System (INIS)

    As a follow-up of the BN-600 hybrid core benchmark, a full MOX core benchmark was performed within the framework of the IAEA co-ordinated research project. Discrepancies between the values of main reactivity coefficients obtained by the participants for the BN-600 full MOX core benchmark appear to be larger than those in the previous hybrid core benchmarks on traditional core configurations. This arises due to uncertainties in the proper modelling of the axial sodium plenum above the core. It was recognized that the sodium density coefficient strongly depends on the core model configuration of interest (hybrid core vs. fully MOX fuelled core with sodium plenum above the core) in conjunction with the calculation method (diffusion vs. transport theory). The effects of the discrepancies revealed between the participants results on the ULOF and UTOP transient behaviours of the BN-600 full MOX core were investigated in simplified transient analyses. Generally the diffusion approximation predicts more benign consequences for the ULOF accident but more hazardous ones for the UTOP accident when compared with the transport theory results. The heterogeneity effect does not have any significant effect on the simulation of the transient. The comparison of the transient analyses results concluded that the fuel Doppler coefficient and the sodium density coefficient are the two most important coefficients in understanding the ULOF transient behaviour. In particular, the uncertainty in evaluating the sodium density coefficient distribution has the largest impact on the description of reactor dynamics. This is because the maximum sodium temperature rise takes place at the top of the core and in the sodium plenum.

  5. Assessment of neutral particle analysis abilities to measure the plasma hydrogen isotope composition in ITER burning scenarios

    International Nuclear Information System (INIS)

    The main object of the neutral particle analysis (NPA) on ITER is to measure the hydrogen isotope composition of the plasma using measurements of the neutralized fluxes of the corresponding hydrogen ions. In burning scenarios the reliable on-line measurements of the tritium/deuterium (T/D) density ratio stands out as the most important application of the diagnostics. This paper presents the results of the error analysis of the NPA signals for ‘official’ ITER burning scenarios—inductive and steady state. The goal of the study is to find the range of values of T/D density ratio, where NPA measurements meet the ITER requirements specified for this parameter, i.e. 10% of accuracy and time resolution equal to 0.1 s. This analysis takes into account both the statistics of the particle counts detected by NPA and of the noise counts induced in NPA detectors by neutrons and gammas. The calibration errors and errors of the external plasma parameters, used in the interpretation of NPA data, are also accounted. The results of the study for low (20–200 keV) and high (200 keV–2 MeV) energy ranges of the D and T neutral fluxes are discussed. It was shown that in the inductive scenario NPA can provide the required accuracy of the T/D density ratio measurement if the plasma composition varies within 0.2–10 (edge measurements) and 0.15–10 (core measurements). In the steady-state scenario these ranges are 0.01–10 (edge measurements) and 0.07–7 (core measurements). (paper)

  6. Phototherapy on the Treatment of Burning Mouth Syndrome: A Prospective Analysis of 20 Cases.

    Science.gov (United States)

    dos Santos, Lúcia de Fátima C; de Andrade, Samantha C; Nogueira, Gessé E C; Leão, Jair C; de Freitas, Patrícia M

    2015-01-01

    The aim of this study was to report the effect of laser phototherapy (LPT) on the treatment of burning mouth syndrome (BMS). This prospective clinical study reports on preliminary outcomes of twenty volunteers diagnosed with BMS who have undergone the conventional treatment prior to laser phototherapy. LPT consisted of weekly sessions of LPT (660 nm), for a period of 10 weeks. The laser protocol consisted of the following parameters: 40 mW, 10 J cm(2) and 0.4 J per point, irradiation time of 10 s. In all sessions, the burning intensity was evaluated with a 10 cm Visual Analogue Scale (VAS). The burning intensity evaluation by VAS was performed immediately before and after each LPT session. Nonparametric test of Wilcoxon was used for statistical analysis, considering a significance level of 5%. All volunteers reported reduced burning intensity in all sessions when compared to the previous one and reduction in VAS scores by up to 49% in the last clinical session when compared to the first session. When only the VAS baseline of the first session was compared with the consecutive sessions, there was a statistically significant reduction in VAS scores in almost all sessions. The LPT may be an alternative treatment for the relief of oral burning symptoms in patients with BMS. PMID:26138316

  7. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author).

  8. Liquid fuel accelerator-driven long-lived waste incinerator. Burning capability evaluation and safety aspects analysis

    International Nuclear Information System (INIS)

    Two Monte Carlo models are developed in order to perform criticality calculations and to determine the burning capability for both MAs and plutonium. 99Tc was selected as representative of the LLFPs. The MA burners can incinerate the production of ten or more PWR. The plutonium burners can destroy the production of one or two PWR. The analysis confirms that, for MA burners, the void fraction has a positive feedback on keff. The results of the calculations corresponding to the MA plus LLFP burner show how to solve this problem, the presence of 99Tc in the reflector markedly curtails the increase of keff. The problem of the positive reactivity, introduced in the Pu burner by temperature increase, can be solved by augmenting the absorption, adding 99Tc in the core. It is convenient to develop and ad hoc burner concept for each task, namely: minor actinides, long-lived fission products and plutonium (with perhaps its associated MAs). (author)

  9. COMPARATIVE ANALYSIS OF SINGLE-CORE AND MULTI-CORE SYSTEMS

    OpenAIRE

    Ogundairo Johnson; Omosehinmi Dinyo

    2015-01-01

    Overall performance of computer systems are better investigated and evaluated when its various components are considered, components such as the hardware, software and firmware. The comparative analysis of single-core and multi-core systems was carried out using Intel Pentium G640T 2.4GHz dualcore, Intel Pentium IV 2.4GHz single-core and Intel Pentium IV 2.8GHz single-core systems. The approach method was using hi-tech benchmarking and stress testing software(s) to examine systems...

  10. Burnup dependent core neutronic analysis for PBMR

    International Nuclear Information System (INIS)

    The strategy for core neutronics modeling is based on SCALE4.4 code KENOV.a module that uses Monte Carlo calculational methods. The calculations are based on detailed unit cell and detailed core modeling. The fuel pebble is thoroughly modeled by introducing unit cell modeling for the graphite matrix and the fuel kernels in the pebble. The core is then modeled by placing these pebbles randomly throughout the core, yet not loosing track of any one of them. For the burnup model, a cyclic manner is adopted by coupling the KENOV.a and ORIGEN-S modules. Shifting down one slice at each discrete time step, and inserting fresh fuel from the top, this cyclic calculation model continues until equilibrium burnup cycle is achieved. (author)

  11. Hybrid Analysis of Engine Core Noise

    Science.gov (United States)

    O'Brien, Jeffrey; Kim, Jeonglae; Ihme, Matthias

    2015-11-01

    Core noise, or the noise generated within an aircraft engine, is becoming an increasing concern for the aviation industry as other noise sources are progressively reduced. The prediction of core noise generation and propagation is especially challenging for computationalists since it involves extensive multiphysics including chemical reaction and moving blades in addition to the aerothermochemical effects of heated jets. In this work, a representative engine flow path is constructed using experimentally verified geometries to simulate the physics of core noise. A combustor, single-stage turbine, nozzle and jet are modeled in separate calculations using appropriate high fidelity techniques including LES, actuator disk theory and Ffowcs-Williams Hawkings surfaces. A one way coupling procedure is developed for passing fluctuations downstream through the flowpath. This method effectively isolates the core noise from other acoustic sources, enables straightforward study of the interaction between core noise and jet exhaust, and allows for simple distinction between direct and indirect noise. The impact of core noise on the farfield jet acoustics is studied extensively and the relative efficiency of different disturbance types and shapes is examined in detail.

  12. Analysis of the microcirculation after soft tissue reconstruction of the outer ear with burns in patients with severe burn injuries.

    Science.gov (United States)

    Medved, Fabian; Medesan, Raluca; Rothenberger, Jens Martin; Schaller, Hans-Eberhard; Schoeller, Thomas; Manoli, Theodora; Weitgasser, Lennart; Naumann, Aline; Weitgasser, Laurenz

    2016-07-01

    Reconstruction of soft tissue defects of the ear with burns remains one of the most difficult tasks for the reconstructive surgeon. Although numerous reconstructive options are available, the results are often unpredictable and worse than expected. Besides full and split skin grafting, local random pattern flaps and pedicled flaps are frequently utilized to cover soft tissue defects of the outer auricle. Because of the difficulty and unpredictable nature of outer ear reconstruction after burn injury, a case-control study was conducted to determine the best reconstructive approach. The microcirculatory properties of different types of soft tissue reconstruction of the outer ear with burns in six severely burned Caucasian patients (three men and three women; mean age, 46 years (range, 22-70)) were compared to those in the healthy tissue of the outer ear using the O2C device (Oxygen to See; LEA Medizintechnik, Gießen, Germany). The results of this study revealed that the investigated microcirculation parameters such as the median values of blood flow (control group: 126 AU), relative amount of hemoglobin (control group: 59.5 AU), and tissue oxygen saturation (control group: 73%) are most similar to those of normal ear tissue when pedicled flaps based on the superficial temporal artery were used. These findings suggest that this type of reconstruction is superior for soft tissue reconstruction of the outer ear with burns in contrast to random pattern flaps and full skin grafts regarding the microcirculatory aspects. These findings may improve the knowledge on soft tissue viability and facilitate the exceptional and delicate process of planning the reconstruction of the auricle with burns. PMID:26997326

  13. Analysis and research status of severe core damage accidents

    International Nuclear Information System (INIS)

    The Severe Core Damage Research and Analysis Task Force was established in Nuclear Safety Research Center, Tokai Research Establishment, JAERI, in May, 1982 to make a quantitative analysis on the issues related with the severe core damage accident and also to survey the present status of the research and provide the required research subjects on the severe core damage accident. This report summarizes the results of the works performed by the Task Force during last one and half years. The main subjects investigated are as follows; (1) Discussion on the purposes and necessities of severe core damage accident research, (2) proposal of phenomenological research subjects required in Japan, (3) analysis of severe core damage accidents and identification of risk dominant accident sequences, (4) investigation of significant physical phenomena in severe core damage accidents, and (5) survey of the research status. (author)

  14. Preliminaries on core image analysis using fault drilling samples; Core image kaiseki kotohajime (danso kussaku core kaisekirei)

    Energy Technology Data Exchange (ETDEWEB)

    Miyazaki, T.; Ito, H. [Geological Survey of Japan, Tsukuba (Japan)

    1996-05-01

    This paper introduces examples of image data analysis on fault drilling samples. The paper describes the following matters: core samples used in the analysis are those obtained from wells drilled piercing the Nojima fault which has moved in the Hygoken-Nanbu Earthquake; the CORESCAN system made by DMT Corporation, Germany, used in acquiring the image data consists of a CCD camera, a light source and core rotation mechanism, and a personal computer, its resolution being about 5 pixels/mm in both axial and circumferential directions, and 24-bit full color; with respect to the opening fractures in core samples collected by using a constant azimuth coring, it was possible to derive values of the opening width, inclination angle, and travel from the image data by using a commercially available software for the personal computer; and comparison of this core image with the BHTV record and the hydrophone VSP record (travel and inclination obtained from the BHTV record agree well with those obtained from the core image). 4 refs., 4 figs.

  15. Reactivity accidents analysis during natural core cooling operation of ETRR-2

    International Nuclear Information System (INIS)

    One of the main features of Egypt Test and Research Reactor Number 2 (ETRR-2), MTR type, is a continuous steady-state operation at low power level, <=400 kW, with core cooling by natural water circulation. Two flapper valves mounted on the return core cooling pipe lines and long chimney encloses the reactor core and assure natural convection phenomena through the reactor core and reactor pool. Many tests and experiments are carried out during this state of operation. A possible occurrence of reactivity insertion accidents (RIA) may be expected over this operation. The present work studies two types of possible RIA: 1-fast reactivity insertion accident (FRIA) with rate 1.04$/s and 2-slow reactivity insertion accident (SRIA) with rate 0.023$/s which may occur due to fast/slow withdrawal of a control rod or sudden cooling of the core inlet water temperature. Failure or success of the reactor scram system during the transient operation is considered. A computer code TRAP22 is developed for such analysis. It is verified against CONVEC code and commissioning tests for steady state operation. The results of verification show good agreement. The study demonstrates that the reactor can be scrammed safely due to either FRIA or SRIA, whenever the maximum expected hot channel HC clad temperature lies within the range 70.73-71.85 deg. C. While, in case of failure of scram system the maximum (HC) clad temperature reaches the burn out value at time 1.175s for FRIA and at 46.36s for SRIA. At the burn out point the clad surface heat flux exceeds its design critical value which results in partial fuel melt

  16. Tajoura reactor core conversion neutrons analysis

    International Nuclear Information System (INIS)

    This paper presents the preliminary neutronics studies and results of the Tajoura reactor core conversion calculations from currently used highly enriched (80% U235) fuel to low enriched fuel (36% U''2''3''5) by using the TAJN computer package. The compact core loading consists of 16 fuel assemblies type IRT-2M surrounded by removable and stationary beryllium reflector and ordinary water for moderation and cooling. The study was undertaken to compare results of TAJN computer package and the vendor documented results. The results of these calculations at the BOL and EOL conditions with equilibrium Xe at 10 MWt are shown. (author)

  17. Analysis of factorial time-course microarrays with application to a clinical study of burn injury

    Science.gov (United States)

    Zhou, Baiyu; Xu, Weihong; Herndon, David; Tompkins, Ronald; Davis, Ronald; Xiao, Wenzhong; Wong, Wing Hung; Toner, Mehmet; Warren, H. Shaw; Schoenfeld, David A.; Rahme, Laurence; McDonald-Smith, Grace P.; Hayden, Douglas; Mason, Philip; Fagan, Shawn; Yu, Yong-Ming; Cobb, J. Perren; Remick, Daniel G.; Mannick, John A.; Lederer, James A.; Gamelli, Richard L.; Silver, Geoffrey M.; West, Michael A.; Shapiro, Michael B.; Smith, Richard; Camp, David G.; Qian, Weijun; Storey, John; Mindrinos, Michael; Tibshirani, Rob; Lowry, Stephen; Calvano, Steven; Chaudry, Irshad; West, Michael A.; Cohen, Mitchell; Moore, Ernest E.; Johnson, Jeffrey; Moldawer, Lyle L.; Baker, Henry V.; Efron, Philip A.; Balis, Ulysses G.J.; Billiar, Timothy R.; Ochoa, Juan B.; Sperry, Jason L.; Miller-Graziano, Carol L.; De, Asit K.; Bankey, Paul E.; Finnerty, Celeste C.; Jeschke, Marc G.; Minei, Joseph P.; Arnoldo, Brett D.; Hunt, John L.; Horton, Jureta; Cobb, J. Perren; Brownstein, Bernard; Freeman, Bradley; Maier, Ronald V.; Nathens, Avery B.; Cuschieri, Joseph; Gibran, Nicole; Klein, Matthew; O’Keefe, Grant

    2010-01-01

    Time-course microarray experiments are capable of capturing dynamic gene expression profiles. It is important to study how these dynamic profiles depend on the multiple factors that characterize the experimental condition under which the time course is observed. Analytic methods are needed to simultaneously handle the time course and factorial structure in the data. We developed a method to evaluate factor effects by pooling information across the time course while accounting for multiple testing and nonnormality of the microarray data. The method effectively extracts gene-specific response features and models their dependency on the experimental factors. Both longitudinal and cross-sectional time-course data can be handled by our approach. The method was used to analyze the impact of age on the temporal gene response to burn injury in a large-scale clinical study. Our analysis reveals that 21% of the genes responsive to burn are age-specific, among which expressions of mitochondria and immunoglobulin genes are differentially perturbed in pediatric and adult patients by burn injury. These new findings in the body’s response to burn injury between children and adults support further investigations of therapeutic options targeting specific age groups. The methodology proposed here has been implemented in R package “TANOVA” and submitted to the Comprehensive R Archive Network at http://www.r-project.org/. It is also available for download at http://gluegrant1.stanford.edu/TANOVA/. PMID:20479259

  18. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  19. Impacts of Nuclear Data on the Core Characteristics of a Compact Breed-and-Burn Fast Reactor (B-BR)

    International Nuclear Information System (INIS)

    In our study, the spent nuclear fuels (SNF) from light water reactors after being metallized are used as the blanket fuels. Meanwhile, metallic low-enriched uranium (LEU) is located in the initial core. Besides providing the core first criticality, the initial core will supply neutrons to the blanket fuels so the fertile fuels can be converted to fissile fuels. In this paper, the impact of using different nuclear data libraries on the core characteristics of the B-BR was addressed. The performance of the B-BR depends on how many fissile fuels can be bred and utilized from the blanket fuels. On the other hand, the core inherent safety characteristic depends strongly on the TRU (transuranic) fuel composition generated in the fuel region. Three present major evaluated nuclear data libraries were considered in this study, i.e., ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0. The neutronics analyses were all performed by using the continuous-energy Monte Carlo code Serpent. A B-BR core has been characterized with three different evaluated nuclear data libraries, ENDF/BVII..0, ENDF/B-VII.1 and JENDL-4.0. The comparison results show that the core excess reactivity is rather sensitive to the type of the library which in turn affects the estimated core life time of the B-BR core. However, a good agreement within the standard deviations on the integral kinetic parameters, Doppler reactivity coefficient, coolant density reactivity coefficient, axial expansion reactivity coefficient and radial expansion reactivity coefficient can be observed among the libraries. The coolant void reactivity (CVR) values of the three libraries slightly deviate beyond the standard deviations. With the different neutron libraries, the buildup of the fission products and minor actinides (especially Am-242m) are slightly different

  20. Full-Core Pin by Pin Subchannel Analysis of SMART

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Kyong Won; Hwang, Dae Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    MATRA-S, a subchannel analysis code was modified to analyze full-core of SMART with pin by pin subchannel model without lumping channels. The SMART core has 57 fuel assemblies of 17x17 arrays with 264 fuel rods and 25 tubes and there are total of 15,048 fuel rods and 16,780 subchannels as shown in Fig.1. Subchannel analysis of a whole reactor core usually has modeled by lumping channels. It lumps several subchannels or assemblies into a big virtual channel and reduces a number of channels and a calculation size of the problem by more than few orders of magnitude. The subchannel analysis models for evaluation of thermal margin of SMART core are 44 and 39 channels for the 1/8-symmetry core as show in Fig.2. The models were developed to be simple that they can be evaluated within reasonable time but to be conservative to a reference model. The reference model is a pin by pin 1/8-core subchannel analysis model and it has 2,333 subchannels and total of 2,119 fuel rods and tubes. The lumped models adopted several engineering factors and assumptions for conservativeness. As the computing power increases drastically, a single stage, full core pin by pin subchannel analysis has become true. This full real subchannel analysis model can obtain more operating margins than lumped models if it is applicable within reasonable time and cost

  1. Core Competence Analysis--Toyota Production System

    Institute of Scientific and Technical Information of China (English)

    钱璐宜

    2013-01-01

      Core competencies are the wel spring of new business development. It is the sharpest sword to penetrate the mature market, hold and enlarge the existing share. Toyota makes wel use of its TPS and form its own style which other car manufacturers hard to imitate.In contrast,the Chinese company---FAW only imitating the superficial aspects from Toyota and ignoring its own problems in manufacture line.

  2. Radiochemical analysis of nuclear fuel burn-up and spent fuel key nuclides

    International Nuclear Information System (INIS)

    Destructive radiochemical analysis of spent nuclear fuels is an important tool to determine burn-up with high accuracy and to better understand the process of depletion and formation of actinides and fission products during irradiation as a result of fission and successive neutron capture. The resulting isotope inventories and nuclear databases that are created, are of high importance to evaluate the performance of nuclear fuels in a reactor, to evaluate computer codes applied for a safe transport, storage and disposal/reprocessing of spent fuels and to safeguard fissile material. The objective is to provide chemical and radiochemical analyses procedures for an accurate determination of isotopic compositions and concentrations of actinides and fission products in different types of industrial (UO2, MOX) and experimental nuclear fuels (UAlx, U3Si2, UMo, ...). For a burn-up determination program typically 21 actinides and fission products are analyzed. For an extended characterization program this can increase to up to approximately 50 isotopes

  3. Adapting the deep burn in-core fuel management strategy for the gas turbine - modular helium reactor to a uranium-thorium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)]. E-mail: alby@neutron.kth.se; Gudowski, Waclaw [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)

    2005-11-15

    evolution, reaction rates, neutron flux and spectrum at the equilibrium of the fuel composition, highlights the features of a deep burn in-core fuel management strategy for a uranium-thorium fuel.

  4. Adapting the deep burn in-core fuel management strategy for the gas turbine - modular helium reactor to a uranium-thorium fuel

    International Nuclear Information System (INIS)

    the fuel composition, highlights the features of a deep burn in-core fuel management strategy for a uranium-thorium fuel

  5. Examination of the Early Diagnostic Applicability of Active Dynamic Thermography for Burn Wound Depth Assessment and Concept Analysis.

    Science.gov (United States)

    Prindeze, Nicholas J; Fathi, Payam; Mino, Matthew J; Mauskar, Neil A; Travis, Taryn E; Paul, Dereck W; Moffatt, Lauren T; Shupp, Jeffrey W

    2015-01-01

    Despite advances in perfusion imaging, burn wound imaging technology continues to lag behind that of other fields. Quantification of blood flow is able to predict time for healing, but clear assessment of burn depth is still questionable. Active dynamic thermography (ADT) is a noncontact imaging modality capable of distinguishing tissue of different thermal conductivities. Utilizing the abnormal heat transfer properties of the burn zones, we examined whether ADT was useful in the determination of burn depth in a model of early burn wound evaluation. Duroc pigs (castrated male; n = 3) were anesthetized, and two burns were created with an aluminum billet at 3 and 12 seconds. These contact times resulted in superficial partial and deep partial thickness burn wounds, respectively. ADT and laser Doppler imaging (LDI) imaging were performed every 30 minutes postburn for a total of five imaging sessions ending 150 minutes postburn. For ADT, imaging excitation was performed for 42-120 seconds with dual quartz-infrared lamps, and subsequent infrared image capture was performed for 300 seconds. MATLAB-assisted image analysis was performed to determine burn zone region of interest thermal relaxation and characteristic patterns. LDI was performed with a moorLDI system, and biopsies were captured for histology following the 150-minute imaging session. Both ADT and LDI imaging modalities are able to detect different physical properties at 30, 60, 90 120, and 150 minutes postburn with statistical significance (P stimulation and a potentially more identifiable differential cooling characteristic. Histological analysis confirmed burn depth. This preliminary work confirms that ADT can measure burn depth and is deserving of further research either as a stand-alone imaging technology or in combination with a device to assess perfusion. PMID:25412050

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

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

    International Nuclear Information System (INIS)

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

  8. Ground-based aerosol characterization during the South American Biomass Burning Analysis (SAMBBA) field experiment

    Science.gov (United States)

    Brito, J.; Rizzo, L. V.; Morgan, W. T.; Coe, H.; Johnson, B.; Haywood, J.; Longo, K.; Freitas, S.; Andreae, M. O.; Artaxo, P.

    2014-11-01

    This paper investigates the physical and chemical characteristics of aerosols at ground level at a site heavily impacted by biomass burning. The site is located near Porto Velho, Rondônia, in the southwestern part of the Brazilian Amazon rainforest, and was selected for the deployment of a large suite of instruments, among them an Aerosol Chemical Speciation Monitor. Our measurements were made during the South American Biomass Burning Analysis (SAMBBA) field experiment, which consisted of a combination of aircraft and ground-based measurements over Brazil, aimed to investigate the impacts of biomass burning emissions on climate, air quality, and numerical weather prediction over South America. The campaign took place during the dry season and the transition to the wet season in September/October 2012. During most of the campaign, the site was impacted by regional biomass burning pollution (average CO mixing ratio of 0.6 ppm), occasionally superimposed by intense (up to 2 ppm of CO), freshly emitted biomass burning plumes. Aerosol number concentrations ranged from ~1000 cm-3 to peaks of up to 35 000 cm-3 (during biomass burning (BB) events, corresponding to an average submicron mass mean concentrations of 13.7 μg m-3 and peak concentrations close to 100 μg m-3. Organic aerosol strongly dominated the submicron non-refractory composition, with an average concentration of 11.4 μg m-3. The inorganic species, NH4, SO4, NO3, and Cl, were observed, on average, at concentrations of 0.44, 0.34, 0.19, and 0.01 μg m-3, respectively. Equivalent black carbon (BCe) ranged from 0.2 to 5.5 μg m-3, with an average concentration of 1.3 μg m-3. During BB peaks, organics accounted for over 90% of total mass (submicron non-refractory plus BCe), among the highest values described in the literature. We examined the ageing of biomass burning organic aerosol (BBOA) using the changes in the H : C and O : C ratios, and found that throughout most of the aerosol processing (O : C ≅ 0

  9. Ground based aerosol characterization during the South American Biomass Burning Analysis (SAMBBA field experiment

    Directory of Open Access Journals (Sweden)

    J. Brito

    2014-05-01

    Full Text Available This paper investigates the physical and chemical characteristics of aerosols at ground level at a site heavily impacted by biomass burning. The site is located near Porto Velho, Rondônia, in the Southwestern part of the Brazilian Amazon forest, and was selected for the deployment of a large suite of instruments, among them an Aerosol Chemical Speciation Monitor. Our measurements were made during the South American Biomass Burning Analysis (SAMBBA field experiment, which consisted of a combination of aircraft and ground based measurements over Brazil, aiming to investigate the impacts of biomass burning emissions on climate, air quality, and numerical weather prediction over South America. The campaign took place during the dry season and the transition to the wet season in September/October 2012. During most of the campaign, the site was impacted by regional biomass burning pollution (average CO mixing ratio of 0.6 ppm, occasionally superimposed by intense (up to 2 ppm of CO, freshly emitted biomass burning plumes. Aerosol number concentrations ranged from ∼1000 cm−3 to peaks of up to 35 000 cm−3 during biomass burning (BB events, corresponding to an average submicron mass mean concentrations of 13.7 μg m−3 and peak concentrations close to 100 μg m−3. Organic aerosol strongly dominated the submicron non-refractory composition, with an average concentration of 11.4 μg m−3. The inorganic species, NH4, SO4, NO3, and Cl, were observed on average at concentrations of 0.44, 0.34, 0.19, and 0.01 μg m−3, respectively. Equivalent Black Carbon (BCe ranged from 0.2 to 5.5 μg m−3, with an average concentration of 1.3 μg m−3. During BB peaks, organics accounted for over 90% of total mass (submicron non-refractory plus BCe, among the highest values described in the literature. We examined the ageing of Biomass Burning Organic Aerosol (BBOA using the changes in the H : C and O : C ratios, and found that throughout most of the aerosol

  10. Evaluation of Analysis Techniques for Fluted-Core Sandwich Cylinders

    Science.gov (United States)

    Lovejoy, Andrew E.; Schultz, Marc R.

    2012-01-01

    Buckling-critical launch-vehicle structures require structural concepts that have high bending stiffness and low mass. Fluted-core, also known as truss-core, sandwich construction is one such concept. In an effort to identify an analysis method appropriate for the preliminary design of fluted-core cylinders, the current paper presents and compares results from several analysis techniques applied to a specific composite fluted-core test article. The analysis techniques are evaluated in terms of their ease of use and for their appropriateness at certain stages throughout a design analysis cycle (DAC). Current analysis techniques that provide accurate determination of the global buckling load are not readily applicable early in the DAC, such as during preliminary design, because they are too costly to run. An analytical approach that neglects transverse-shear deformation is easily applied during preliminary design, but the lack of transverse-shear deformation results in global buckling load predictions that are significantly higher than those from more detailed analysis methods. The current state of the art is either too complex to be applied for preliminary design, or is incapable of the accuracy required to determine global buckling loads for fluted-core cylinders. Therefore, it is necessary to develop an analytical method for calculating global buckling loads of fluted-core cylinders that includes transverse-shear deformations, and that can be easily incorporated in preliminary design.

  11. Subchannel analysis of a small ultra-long cycle fast reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Han; Kim, Ji Hyun; Bang, In Cheol, E-mail: icbang@unist.ac.kr

    2014-04-01

    Highlights: • The UCFR-100 is small-sized one of 60 years long-life nuclear reactors without refueling. • The design safety limits of the UCFR-100 are evaluated using MATRA-LMR. • The subchannel results are below the safety limits of general SFR design criteria. - Abstract: Thermal-hydraulic evaluation of a small ultra-long cycle fast reactor (UCFR) core is performed based on existing safety regulations. The UCFR is an innovative reactor newly designed with long-life core based on the breed-and-burn strategy and has a target electric power of 100 MWe (UCFR-100). Low enriched uranium (LEU) located at the bottom region of the core play the role of igniter to operate the UCFR for 60 years without refueling. A metallic form is selected as a burning fuel region material after the LEU location. HT-9 and sodium are used as cladding and coolant materials, respectively. In the present study, MATRA-LMR, subchannel analysis code, is used for evaluating the safety design limit of the UCFR-100 in terms of fuel, cladding, and coolant temperature distributions in the core as design criteria of a general fast reactor. The start-up period (0 year of operation), the middle of operating period (30 years of operation), and the end of operating cycle (60 years of operation) are analyzed and evaluated. The maximum cladding surface temperature (MCST) at the BOC (beginning of core life) is 498 °C on average and 551 °C when considering peaking factor, while the MCST at the MOC (middle of core life) is 498 °C on average and 548 °C in the hot channel, respectively, and the MCST at the EOC (end of core life) is 499 °C on average and 538 °C in the hot channel, respectively. The maximum cladding surface temperature over the long cycle is found at the BOC due to its high peaking factor. It is found that all results including fuel rods, cladding, and coolant exit temperature are below the safety limit of general SFR design criteria.

  12. Analysis of derived optical parameters of atmospheric particles during a biomass burning event. Comparison with fossil fuel burning

    Science.gov (United States)

    Costa, A.; Mogo, S.; Cachorro, V.; de Frutos, A.; Medeiros, M.; Martins, R.; López, J. F.; Marcos, A.; Marcos, N.; Bizarro, S.; Mano, F.

    2015-12-01

    During the day November 26, 2014, a scheduled cleanup of the woods took place around the GOA-UVa aerosol measurement station located at the campus of the University of Beira Interior (40° 16’30”N, 7°30’35”W, 704m a.s.l.), Covilhã, Portugal. This cleanup included excessive vegetation removal during the morning, using fossil fuel-burning machinery, and burning of the vegetation during the afternoon. In situ measurements of aerosol optical properties were made and this study aims the characterization of the evolution of aerosol properties during the day. The optical parameters were monitored using a 3-wavelength nephelometer and a 3-wavelength particle soot absorption photometer. Selective sampling/exclusion of the coarse particles was done each 5 minutes. The scattering and absorption Ångström exponents as well as the single scattering albedo were derived and fully analyzed. The scattering and absorption coefficients increased dramatically during the event, reaching values as high as 720.3 Mm-1 and 181.9 Mm-1, respectively, for the green wavelength and PM10 size fraction. The spectral behavior of these parameters also changed wildly along the day and an inversion of the slope from positive to negative in the case of the single scattering albedo was observed.

  13. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    Energy Technology Data Exchange (ETDEWEB)

    Widiawati, Nina, E-mail: nina-widiawati28@yahoo.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10, Bandung (Indonesia)

    2015-09-30

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from −0.6695443 % at BOC to −0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  14. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    Science.gov (United States)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from -0.6695443 % at BOC to -0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  15. ORIGEN computer code use in non-destructive analysis of irradiated fuel elements for burn-up determination

    International Nuclear Information System (INIS)

    An iterative method for burn-up determination in the non-destructive analysis of irradiated fuel elements using the ORIGEN computer code is presented. On the bases of data obtained from ORIGEN code the calibration coefficient for the neutron flux is determined as a function of one fission product activity while the burn-up is determined as a function of the calibration coefficient for a given irradiation history. These functions are used for determining the burn-up of nuclear fuel elements measured by gamma-scanning. The method is tested for fuel elements irradiated in a TRIGA reactor facility. (Author)

  16. Analysis for core conversion HEU-LEU) of PARR-2

    International Nuclear Information System (INIS)

    Calculational methodology for conversion of Miniature Neutron Source Reactor (MNSR) from HEU to LEU was validated by doing analysis of HEU fuel (90.2% enriched). On the basis of HEU based reactor model, analysis of LEU (UO/sub 2/ fuel) core gives results, which qualify the UO/sub 2/ fuel for future LEU core of MNSR. However for LEU fuel, neutron flux at irradiation sites is slightly lower for the reactor operating at 30 kW power. Therefore reactor power will have to be increased to a level of 33 kW to get the same thermal flux values as obtained for HEU core. Use of the same control rod as being used in the current HEU core gives lower values of shut down margin and control rod worth. But the slightly increased diameter of control rod improves shut down margin to a value that is comparable to the corresponding value for HEU core. LEU (UO/sub 2/ fuelled) core with following characteristics provides replica of the currently operating HEU core: 'Enrichment: 12.46%' Guide tube and grid plate material: Zr-4 'Reactor power: 3.3kW' Cladding material of fuel pin: Zr-4/' Control rod absorber (cadmium) thickness: 4.5 mm All other materials and structures have been assumed to be same as are being used in the presently operating HEU core. There is no significant difference between the dose values for HEU and prospected LEU fuel. Therefore existing HEU core and prospected LEU core of MNSR are considered to be safe for the public even in case of an accident releasing radioactive gases from the fuel. (orig./A.B.)

  17. Assessment of fire emission inventories during the South American Biomass Burning Analysis (SAMBBA) experiment

    Science.gov (United States)

    Pereira, Gabriel; Siqueira, Ricardo; Rosário, Nilton E.; Longo, Karla L.; Freitas, Saulo R.; Cardozo, Francielle S.; Kaiser, Johannes W.; Wooster, Martin J.

    2016-06-01

    Fires associated with land use and land cover changes release large amounts of aerosols and trace gases into the atmosphere. Although several inventories of biomass burning emissions cover Brazil, there are still considerable uncertainties and differences among them. While most fire emission inventories utilize the parameters of burned area, vegetation fuel load, emission factors, and other parameters to estimate the biomass burned and its associated emissions, several more recent inventories apply an alternative method based on fire radiative power (FRP) observations to estimate the amount of biomass burned and the corresponding emissions of trace gases and aerosols. The Brazilian Biomass Burning Emission Model (3BEM) and the Fire Inventory from NCAR (FINN) are examples of the first, while the Brazilian Biomass Burning Emission Model with FRP assimilation (3BEM_FRP) and the Global Fire Assimilation System (GFAS) are examples of the latter. These four biomass burning emission inventories were used during the South American Biomass Burning Analysis (SAMBBA) field campaign. This paper analyzes and inter-compared them, focusing on eight regions in Brazil and the time period of 1 September-31 October 2012. Aerosol optical thickness (AOT550 nm) derived from measurements made by the Moderate Resolution Imaging Spectroradiometer (MODIS) operating on board the Terra and Aqua satellites is also applied to assess the inventories' consistency. The daily area-averaged pyrogenic carbon monoxide (CO) emission estimates exhibit significant linear correlations (r, p > 0.05 level, Student t test) between 3BEM and FINN and between 3BEM_ FRP and GFAS, with values of 0.86 and 0.85, respectively. These results indicate that emission estimates in this region derived via similar methods tend to agree with one other. However, they differ more from the estimates derived via the alternative approach. The evaluation of MODIS AOT550 nm indicates that model simulation driven by 3BEM and FINN

  18. Nuclear Core Analysis Technology Development for Future Fuel

    International Nuclear Information System (INIS)

    As for the supporting work for the irradiation test of MOX fuel rods, the fuel irradiation test data including the reactor operating conditions and the linear power generation rates for each MOX fuel rods for cycle 4 through 8 were collected. The estimated fuel burnup of test fuel rods and the radial power distribution within the fuel pellet were calculated and supplied to the fuel rod design group in order to assess the fuel rod behavior. The determination of the irradiation conditions of the test fuel rods for succeeding cycles were also carried out. As for the development of core analysis methodology, the nuclear analysis code verification data were acquired through the participation in the international programs organized by NEA. The acquired data includes the MOX core transient benchmark problem, the draft benchmark specification of VENUS-7, 9 and 17, and the B and W critical experiment data. The experiment data from the commercial power reactor operation with MOX fuel rods were also acquired. The capability of predicting the static and transient behavior of the MOX fueled core of the nuclear core analysis code, NUREC, was validated against the above the verification data. This core analysis technology was transferred to the industry related with nuclear core design

  19. Analysis and Application of the Series Core Snubber

    Science.gov (United States)

    Xie, Fei; Li, Ge; Cheng, Desheng; Chen, Qiangjian

    2013-05-01

    The transformer core snubber (CS), as one of the most important components in the EAST (experimental advanced superconducting tokamak) NBI (neutral beam injector) system, is designed to limit grid damage and protect the ion source during periods of electrical breakdowns. A transformer core snubber is analyzed in detail in this paper. Several kinds of soft magnetic cores are presented and compared. With analysis and experiment on the basic characteristics of the cores, the most suitable materials are suggested. The circuit simulation code is established which could simulate faulty conditions with concentrated and distributed CS concepts. Based on the above work, an ion source CS is developed with series type of distributed topology. The CS has been subjected to experimental validation at 80 kV with a peak short-current of approximately 400 A in a real NBI system, which proves the accuracy of the adopted assumptions and the analysis method.

  20. Analysis and Application of the Series Core Snubber

    International Nuclear Information System (INIS)

    The transformer core snubber (CS), as one of the most important components in the EAST (experimental advanced superconducting tokamak) NBI (neutral beam injector) system, is designed to limit grid damage and protect the ion source during periods of electrical breakdowns. A transformer core snubber is analyzed in detail in this paper. Several kinds of soft magnetic cores are presented and compared. With analysis and experiment on the basic characteristics of the cores, the most suitable materials are suggested. The circuit simulation code is established which could simulate faulty conditions with concentrated and distributed CS concepts. Based on the above work, an ion source CS is developed with series type of distributed topology. The CS has been subjected to experimental validation at 80 kV with a peak short-current of approximately 400 A in a real NBI system, which proves the accuracy of the adopted assumptions and the analysis method. (fusion engineering)

  1. Core conversion effects on the safety analysis of research reactors

    International Nuclear Information System (INIS)

    The safety related parameters of the 5 MW Democritus research reactor that will be affected by the scheduled core conversion to use LEU instead of HEU are considered. The analysis of the safety related items involved in such a core conversion, mainly the consequences due to MCA, DBA, etc., is of a general nature and can, therefore, be applied to other similar pool type reactors as well. (T.A.)

  2. TMI-2 accident: core heat-up analysis

    International Nuclear Information System (INIS)

    This report summarizes NSAC study of reactor core thermal conditions during the accident at Three Mile Island, Unit 2. The study focuses primarily on the time period from core uncovery (approximately 113 minutes after turbine trip) through the initiation of sustained high pressure injection (after 202 minutes). The transient analysis is based upon established sequences of events; plant data; post-accident measurements; interpretation or indirect use of instrument responses to accident conditions

  3. Characterization of Pseudomonas aeruginosa in Burn Patients Using PCR- Restriction Fragment Length Polymorphism and Random Amplified Polymorphic DNA Analysis

    OpenAIRE

    Abdolaziz Rastegar Lari; Bagher Yakhchali; Parviz Owlia; Hassan Salimi

    2010-01-01

    One of the major opportunistic pathogens in patients with burninjuries is Pseudomonas aeruginosa, which causes severe infectionsin burned patients. The objective of the study was to examinethe molecular epidemiology of P. aeruginosa colonization inthe burn unit of Shahid Motahari Hospital in Tehran, Iran. Restrictionfragment length polymorphism (RFLP) and random amplifiedpolymorphic DNA (RAPD) analysis were employed tostudy 127 clinical and two environmental P. aeruginosa isolatescollected fr...

  4. Development of VHTR Core Analysis and Verification Methodology

    International Nuclear Information System (INIS)

    The primary objective of this project is to develop a three dimensional cylindrical geometry code to analyze PBRs using the Analytic Function Expansion Nodal (AFEN) method, and the second objective is to produce the numerical data and to verify the deterministic code from commercial PBR core and prism reactor core using the Monte Carlo method. We developed the TOPS code and verified its validity with various benchmark problems for stead-state and transient conditions. Considering the pebble flow and temperature distribution within the core, the core analysis for commercial pebble-type reactor was carried out by using the Monte Carlo method and the spatial-dependent Dancoff factors, and also was evaluated with the Monte Carlo method. And the optimization of the decay chain model, and implementation of the multi-group cross section processing of DeCART using McCARD for double heterogeneity effect. The TOPS code can be used in VHTR's design and reactor core characteristics evaluation, and the Monte Carlo results of core analysis can be used to the verification of the deterministic code. Furthermore, they are expected that the analysis method can be installed in the deterministic code

  5. Development of VHTR Core Analysis and Verification Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Han Gyu; Kim, Chang Hyo; Park, Ho Jin [Seoul National University, Seoul (Korea, Republic of)] (and others)

    2009-03-15

    The primary objective of this project is to develop a three dimensional cylindrical geometry code to analyze PBRs using the Analytic Function Expansion Nodal (AFEN) method, and the second objective is to produce the numerical data and to verify the deterministic code from commercial PBR core and prism reactor core using the Monte Carlo method. We developed the TOPS code and verified its validity with various benchmark problems for stead-state and transient conditions. Considering the pebble flow and temperature distribution within the core, the core analysis for commercial pebble-type reactor was carried out by using the Monte Carlo method and the spatial-dependent Dancoff factors, and also was evaluated with the Monte Carlo method. And the optimization of the decay chain model, and implementation of the multi-group cross section processing of DeCART using McCARD for double heterogeneity effect. The TOPS code can be used in VHTR's design and reactor core characteristics evaluation, and the Monte Carlo results of core analysis can be used to the verification of the deterministic code. Furthermore, they are expected that the analysis method can be installed in the deterministic code.

  6. Transient analysis for PWR reactor core using neural networks predictors

    International Nuclear Information System (INIS)

    In this study, transient analysis for a Pressurized Water Reactor core has been performed. A lumped parameter approximation is preferred for that purpose, to describe the reactor core together with mechanism which play an important role in dynamic analysis. The dynamic behavior of the reactor core during transients is analyzed considering the transient initiating events, wich are an essential part of Safety Analysis Reports. several transients are simulated based on the employed core model. Simulation results are in accord the physical expectations. A neural network is developed to predict the future response of the reactor core, in advance. The neural network is trained using the simulation results of a number of representative transients. Structure of the neural network is optimized by proper selection of transfer functions for the neurons. Trained neural network is used to predict the future responses following an early observation of the changes in system variables. Estimated behaviour using the neural network is in good agreement with the simulation results for various for types of transients. Results of this study indicate that the designed neural network can be used as an estimator of the time dependent behavior of the reactor core under transient conditions

  7. Geoantineutrino Spectrum, 3He/4He - ratio radial distribution and Slow Nuclear Burning on the Boundary of the Liquid and Solid Phases of the Earth's core

    CERN Document Server

    Rusov, V D; Vaschenko, V N; Tarasov, V A; Zelentsova, T N; Bolshakov, V N; Litvinov, D A; Kosenko, S I; Byegunova, O A

    2006-01-01

    The problem of the geoantineutrino deficit and the experimental results of the interaction of uranium dioxide and carbide with iron-nickel and silica-alumina melts at high pressure (5-10 Gpa) and temperature (1600-22000 C) have motivated us to consider the possible consequences of the assumption made by V.Anisichkin and coauthors that there is an actinid shell on boundary of liquid and solid phases of the Earth's core. We have shown that the activation of a natural nuclear reactor operating as the solitary waves of nuclear burning in 238U- and/or 232Th-medium (in particular, the neutron-fission progressive wave of Feoktistov and/or Teller-Ishikawa-Wood) can be such a physical consequence. The simplified model of the kinetics of accumulation and burnup in U-Pu fuel cycle of Feoktistov is developed. The results of the numerical simulation of neutron-fission wave in two-phase UO2/Fe medium on a surface of the Earth's solid core are presented. The georeactor model of 3He origin and the 3He/4He-ratio distribution ...

  8. Analysis of predictor factors of limb amputation in patients with high-voltage electrical burns

    Directory of Open Access Journals (Sweden)

    Guillermo García Álvarez

    2015-09-01

    Full Text Available Background: Limb amputation is considered one of the most devastating consequences of electrical injury. Any factors that correlate with the degree of muscle damage can be used to predict the necessity of limb amputation. The aim of this study was to determine the factors that can be used to predict limb amputation in high-voltage electrically injured patients. Methods: Eighty-two high-voltage electrically injured patients were admitted to the Department of Plastic and Reconstructive Surgery and Burns of National Arzobispo Loayza Hospital on a 5 year period. A retrospective analysis of the possible related risk factors between amputation and non-amputation patients was performed. Results: A total of 68 patients were enrolled for analysis. Thirteen patients underwent limb amputations. Multivariate analysis of the risk factors between amputation and non-amputation groups showed statistical significance for first 24 hour creatine kinase-isoenzyme MB (CKMB level. A serum CK-MB level above 14,955 U/L predicted high risk of limb amputation with high specificity (84% and sensitivity (77%. Only one patient with a remarkable decrease of creatine kinase (CPKt and CK-MB levels after fasciotomy avoided a major limb amputation. Conclusion: Our results suggest that CPK-MB level is an independent factor for prediction of limb amputation in patients with high-voltage electrical burns. We suggest that the addition of CPK-MB evaluation to clinical symptom screening may be a valuable method for early detection of muscle damage.

  9. HTGR core seismic analysis using an array processor

    International Nuclear Information System (INIS)

    A Floating Point Systems array processor performs nonlinear dynamic analysis of the high-temperature gas-cooled reactor (HTGR) core with significant time and cost savings. The graphite HTGR core consists of approximately 8000 blocks of various shapes which are subject to motion and impact during a seismic event. Two-dimensional computer programs (CRUNCH2D, MCOCO) can perform explicit step-by-step dynamic analyses of up to 600 blocks for time-history motions. However, use of two-dimensional codes was limited by the large cost and run times required. Three-dimensional analysis of the entire core, or even a large part of it, had been considered totally impractical. Because of the needs of the HTGR core seismic program, a Floating Point Systems array processor was used to enhance computer performance of the two-dimensional core seismic computer programs, MCOCO and CRUNCH2D. This effort began by converting the computational algorithms used in the codes to a form which takes maximum advantage of the parallel and pipeline processors offered by the architecture of the Floating Point Systems array processor. The subsequent conversion of the vectorized FORTRAN coding to the array processor required a significant programming effort to make the system work on the General Atomic (GA) UNIVAC 1100/82 host. These efforts were quite rewarding, however, since the cost of running the codes has been reduced approximately 50-fold and the time threefold. The core seismic analysis with large two-dimensional models has now become routine and extension to three-dimensional analysis is feasible. These codes simulate the one-fifth-scale full-array HTGR core model. This paper compares the analysis with the test results for sine-sweep motion

  10. Century-long Record of Black Carbon in an Ice Core from the Eastern Pamirs: Estimated Contributions from Biomass Burning

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Mo; Xu, B.; Kaspari, Susan D.; Gleixner, Gerd; Schwab, Valerie; Zhao, Huabiao; Wang, Hailong; Yao, Ping

    2015-08-01

    We analyzed refractory black carbon (rBC) in an ice core spanning 1875-2000 AD from Mt. Muztagh Ata, the Eastern Pamirs, using a Single Particle Soot Photometer (SP2). Additionally a pre-existing levoglucosan record from the same ice core was used to differentiate rBC that originated from open fires, energy-related combustion of biomass, and fossil fuel combustion. Mean rBC concentrations increased four-fold since the mid-1970s and reached maximum values at the end of 1980s. The observed decrease of the rBC concentrations during the 1990s was likely driven by the economic recession of former USSR countries in Central Asia. Levoglucosan concentrations showed a similar temporal trend to rBC concentrations, exhibiting a large increase around 1980 AD followed by a decrease in the 1990s that was likely due to a decrease in energy-related biomass combustion. The time evolution of levoglucosan/rBC ratios indicated stronger emissions from open fires during the 1940s-1950s, while the increase in rBC during the 1980s-1990s was caused from an increase in energy-related combustion of biomass and fossil fuels.

  11. Century-long record of black carbon in an ice core from the Eastern Pamirs: Estimated contributions from biomass burning

    Science.gov (United States)

    Wang, Mo; Xu, Baiqing; Kaspari, Susan D.; Gleixner, Gerd; Schwab, Valérie F.; Zhao, Huabiao; Wang, Hailong; Yao, Ping

    2015-08-01

    We analyzed refractory black carbon (rBC) in an ice core spanning 1875-2000 AD from Mt. Muztagh Ata, the Eastern Pamirs, using a Single Particle Soot Photometer (SP2). Additionally a pre-existing levoglucosan record from the same ice core was used to differentiate rBC that originated from open fires, energy-related combustion of biomass, and fossil fuel combustion. Mean rBC concentrations increased four-fold since the mid-1970s and reached maximum values at end of the 1980s. The observed decrease of the rBC concentrations during the 1990s was likely driven by the economic recession of former USSR countries in Central Asia. Levoglucosan concentrations showed a similar temporal trend to rBC concentrations, exhibiting a large increase around 1980 AD followed by a decrease in the 1990s that was likely due to a decrease in energy-related biomass combustion. The time evolution of levoglucosan/rBC ratios indicated stronger emissions from open fires during the 1940s-1950s, while the increase in rBC during the 1980s-1990s was caused from an increase in energy-related combustion of biomass and fossil fuels.

  12. TRACE analysis of Phenix core response to an increase of the core inlet sodium temperature

    International Nuclear Information System (INIS)

    This work presents the analysis, using the TRACE code, of the Phenix core response to an inlet sodium temperature increase. The considered experiment was performed in the frame of the Phenix End-Of-Life (EOL) test program of the CEA, prior to the final shutdown of the reactor. It corresponds to a transient following a 40°C increase of the core inlet temperature, which leads to a power decrease of 60%. This work focuses on the first phase of the transient, prior to the reactor scram and pump trip. First, the thermal-hydraulic TRACE model of the core developed for the present analysis is described. The kinetic parameters and feedback coefficients for the point kinetic model were first derived from a 3D static neutronic ERANOS model developed in a former study. The calculated kinetic parameters were then optimized, before use, on the basis of the experimental reactivity in order to minimize the error on the power calculation. The different reactivity feedbacks taken into account include various expansion mechanisms that have been specifically implemented in TRACE for analysis of fast-neutron spectrum systems. The point kinetic model has been used to study the sensitivity of the core response to the different feedback effects. The comparison of the calculated results with the experimental data reveals the need to accurately calculate the reactivity feedback coefficients. This is because the reactor response is very sensitive to small reactivity changes. This study has enabled us to study the sensitivity of the power change to the different reactivity feedbacks and define the most important parameters. As such, it furthers the validation of the FAST code system, which is being used to gain a more in-depth understanding of SFR core behavior during accidental transients. (author)

  13. Assumed mode approach to fast reactor core seismic analysis

    International Nuclear Information System (INIS)

    The need for a time history approach, rather than a response spectrum approach, to the seismic analysis of fast breeder reactor core structures is described. The use of a Rayleigh-Ritz/Assumed Mode formalism for developing mathematical models of reactor cores is presented. Various factors including structural nonlinearity, fluid inertia, and impact which necessitate abandonment of response spectrum methods are discussed. The use of the assumed mode formalism is described in some detail as it applies to reactor core seismic analysis. To illustrate the use of this formal approach to mathematical modeling, a sample reactor problem with increasing complexities of modeling is presented. Finally, several problem areas--fluid inertia, fluid damping, coulomb friction, impact, and modal choice--are discussed with emphasis on research needs for use in fast reactor seismic analysis

  14. Core thermal hydraulic analysis for TNR power uprating

    International Nuclear Information System (INIS)

    This paper presents preliminary results of a study undertaken to investigate the possibility of raising the power of the Tajura Nuclear Research Reactor (TNRR) from 10 to 20 MWt keeping the same core configuration and with minimum changes in the primary cooling circuit. The study was carried out for a fresh core, with compact load (16 assemblies) under normal operation conditions. A computer program, TAJT, was used to simulate the core and perform the necessary thermal hydraulic analysis. The results obtained show that the reactor power could be raised to 15 MWt safely and with no changes in the primary cooling circuit. To raise the power to 20 MWt will require changes in the core configuration and primary circuit

  15. Ground based characterization of biomass burning aerosols during the South American Biomass Burning Analysis (SAMBBA) field experiment in Brazil during Sept - Oct 2012

    Science.gov (United States)

    Artaxo, Paulo; Ferreira de Brito, Joel; Varanda Rizzo, Luciana; Johnson, Ben; Haywood, Jim; Longo, Karla; Freitas, Saulo; Coe, Hugh

    2013-04-01

    Biomass burning is one of the major drivers for atmospheric composition in the Southern hemisphere. In Amazonia, deforestation rates have been steadily decreasing, from 27,000 Km² in 2004 to about 5,000 Km² in 2011. This large reduction (by factor 5) was not followed by similar reduction in aerosol loading in the atmosphere due to the increase in agricultural fires. AERONET measurements from 5 sites show a large year-to year variability due to climatic and socio-economic issues. Besides this strong reduction in deforestation rate, biomass burning emissions in Amazonia increases concentrations of aerosol particles, CO, ozone and other species, and also change the surface radiation balance in a significant way. To complement the long term biomass burning measurements in Amazonia, it was organized in 2012 the intensive campaign of the South American Biomass Burning Analysis (SAMBBA) experiment with an airborne and a ground based components. A sampling site was set up at Porto Velho, with measurements of aerosol size distribution, optical properties such as absorption and scattering at several wavelengths, organic aerosol characterization with an ACSM - Aerosol Chemical Speciation Monitor. CO, CO2 and O3 were also measured to characterize combustion efficiency and photochemical processes. Filters for trace elements measured by XRF and for OC/EC determined using a Sunset instrument were also collected. An AERONET CIMEL sunphotometer was operated in parallel with a multifilter radiometer (MFR). A large data set was collected from August to October 2012. PM2.5 aerosol concentrations up to 250 ug/m3 were measured, with up to 20 ug/m3 of black carbon. Ozone went up to 60 ppb at mid-day in August. At night time ozone was consumed completely most of the time. ACSM shows that more than 85% of the aerosol mass was organic with a clear diurnal pattern. The organic aerosol volatility was very variable depending on the air mass sampled over Porto Velho. Aerosol optical depth at

  16. Fuel burn analysis of a sodium fast reactor with KANEXT and Serpent

    International Nuclear Information System (INIS)

    The fast reactors cooled by sodium are one of the options considered in the Generation IV. Since most of the reactors of Fourth Generation are still in development stage, is necessary to have efficient and reliable computational tools, this in order to obtain accurate results in reasonable computational times. In this paper is introduced and describes the deterministic code KANEXT (KArlsruhe Neutronic EXtended Tool) and is compared against a Monte Carlo code of more diffusion: Serpent. KANEXT, being a modular code requires the interaction of different modules to perform a job, this interaction of modules is described in this article. The parameters to be compared are the results of the neutron multiplication effective factor and the evolution of isotopes during the burning. The mentioned comparison is carried out for a fast reactor cooled by sodium of relatively small size compared to commercial size reactors. In this paper the particularities of the reactor are described, important for the analysis such as geometry, enrichments, reflector, etc. The considerations in the implementation in both codes are also described, as are simplifications, length of the burning steps, possible solutions of the Bateman equations for the burning fuel in Serpent and the solution options for transport (P3) and diffusion (P1) in KANEXT. The results show good correspondence between Serpent and KANEXT, which give confidence to continue using KANEXT as the main tool. Respect to computation time, time saving is evident with the use of deterministic codes instead of Monte Carlo codes, in this particular case, the time savings using KANEXT is about 98.5% of the time used by Serpent. (Author)

  17. Analysis of Irradiation Holes of In-Core Region

    International Nuclear Information System (INIS)

    Test fuels and materials are irradiated in the in-core region in side of the chimney. The inner chimney is composed of In-Core and Out-Core regions. The In-Core region has 23 hexagonal vertical irradiation holes named from R01 to R20, CT, IR1 and IR2 and 8 cylindrical irradiation holes named from CAR1 to CAR4 and SOR1 to SOR4. The Out-Core region is composed of 8 cylindrical irradiation holes named from OR1 to OR8 which are installed near the inner shell of the reflector tank. HANARO is the multi-purpose research reactor which utilizes in-core irradiation holes, which is being used in various field. Over the past 7 years we have used CT 8 times, IR once, IR2 and OR3 twice, OR4 three times and OR5 ten times. These irradiation holes are used to perform an evaluation of the neutron irradiation properties and the tests were all completed and done successfully. HANARO has been used successfully, and it still will be used continuously in various fields such as nuclear in-pile tests, the production of radioisotopes, neutron transmutation doping, neutron activation analysis, neutron beam research, radiography, environmental science

  18. Analysis of Irradiation Holes of In-Core Region

    Energy Technology Data Exchange (ETDEWEB)

    In, Won-ho; Lee, Yong-sub; Kim, Tae-hwan; Lim, Kyoung-hwan; Ahn, Hyung-jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Test fuels and materials are irradiated in the in-core region in side of the chimney. The inner chimney is composed of In-Core and Out-Core regions. The In-Core region has 23 hexagonal vertical irradiation holes named from R01 to R20, CT, IR1 and IR2 and 8 cylindrical irradiation holes named from CAR1 to CAR4 and SOR1 to SOR4. The Out-Core region is composed of 8 cylindrical irradiation holes named from OR1 to OR8 which are installed near the inner shell of the reflector tank. HANARO is the multi-purpose research reactor which utilizes in-core irradiation holes, which is being used in various field. Over the past 7 years we have used CT 8 times, IR once, IR2 and OR3 twice, OR4 three times and OR5 ten times. These irradiation holes are used to perform an evaluation of the neutron irradiation properties and the tests were all completed and done successfully. HANARO has been used successfully, and it still will be used continuously in various fields such as nuclear in-pile tests, the production of radioisotopes, neutron transmutation doping, neutron activation analysis, neutron beam research, radiography, environmental science.

  19. Core Handling and Real-Time Non-Destructive Characterization at the Kochi Core Center: An Example of Core Analysis from the Chelungpu Fault

    Directory of Open Access Journals (Sweden)

    W. Lin

    2007-11-01

    Full Text Available As an example of core analysis carried out inactive fault drilling programs, we report the procedures of core handling on the drilling site and non-destructive characterization in the laboratory. This analysis was employed onthe core samples taken from HoleBof the Taiwan Chelungpu-fault Drilling Project (TCDP, which penetrated through the active fault that slipped during the 1999 Chi-Chi, Taiwan earthquake. We show results of the non-destructive physical property measurements carried out at the Kochi Core Center (KCC, Japan. Distinct anomalies of lower bulk density and higher magnetic susceptibilitywere recognized in all three fault zones encountered in HoleB. To keep the core samples in good condition before they are used for variousanalyses is crucial. In addition, careful planning for core handlingand core analyses is necessary for successfulinvestigations. doi:10.2204/iodp.sd.s01.35.2007

  20. Preliminary Core Analysis for Regulatory Evaluation of SFR Nuclear Designs

    International Nuclear Information System (INIS)

    In Korea, a conceptual design of SFR has been developed by Korea Atomic Energy Research Institute (KAERI). An application for the design approval of a prototype SFR is scheduled in 2017. In order to prepare the licensing of a prototype SFR, Korea Institute of Nuclear Safety (KINS) is developing the regulatory audit code system for SFR since 2012. The SFR nuclear evaluation system for regulatory verification has an object to verify core integral parameters, reactivity coefficients, and peaking power factor, etc. provided by the designers and generate reactivity coefficients and kinetic parameters for safety analyses. For these purpose, both stochastic (Monte Carlo) and deterministic tools can be used. In deterministic core analysis, the PARCS code is being considered for the whole-core steady-state and time-dependent, multi-group hexagonal diffusion calculations and additional code modules for cross-section generation and more accurate transport solution, etc. are planned to be developed from 2013 or be introduced. This paper presents the development strategy of SFR nuclear evaluation system for regulatory verification and the preliminary core analysis results of Korean SFR demonstration reactor with 600MWe (DEMO-600). The DEMO-600 core is preliminarily analyzed using PARCS, and global reactivity worths such as uniform expansion coefficients, rod worths are calculated by direct eigenvalue differences of the base and perturbed configurations. The results calculated by PARCS are compared with the DIF3D nodal solutions provided by KAERI. (author)

  1. Stress analysis of portable safety platform (Core Sampler Truck)

    International Nuclear Information System (INIS)

    This document provides the stress analysis and evaluation of the portable platform of the rotary mode core sampler truck No. 2 (RMCST number-sign 2). The platform comprises railing, posts, deck, legs, and a portable ladder; it is restrained from lateral motion by means of two brackets added to the drill-head service platform

  2. Methodology for reactor core physics analysis - part 2

    International Nuclear Information System (INIS)

    The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs

  3. Analysis of burn up effects on kinetic parameters in an Accelerator Driven Subcritical TRIGA reactor

    International Nuclear Information System (INIS)

    Highlights: • Among the kinetic parameters, the most important ones are βeff and Λ. • The kinetic parameters at the end of cycle for 20 days operation at 1 MW reactor power have been calculated using MCNPX code. • The system sub-criticality levels, the required accelerator current and power have been calculated for each burnup step. • Burnup effects on neutronic parameters and accelerator requirements are analysed for a life cycle of the reactor core. - Abstract: In this paper, the kinetic parameters at the end of cycle for 20 days operation at 1 MW reactor power have been calculated using MCNPX code for source multiplication factors (Ks) of critical and 0.97 levels. Accelerator Driven Subcritical TRIGA reactor has been considered as the case study of the problem. The system sub-criticality levels, the required accelerator current (Ip) and power (Pacc) have been calculated for each burn up step. According to the results, the required Ip and Pacc increased as system sub-criticality level decreased. Also, the results show that the effective delayed neutron fraction (βeff) does not depend on sub-criticality levels significantly but the neutron reproduction time (Λ) strongly depends on the number of fuel elements in the core and sub-criticality levels. In addition, it was observed that in comparison with the beginning-of-cycle values, at end-of-cycle, the Λ increases by 112.6% and 108.4% respectively but the βeff decreases by 6.1% and 1.8% respectively in critical and 0.97 levels

  4. Ground-based aerosol characterization during the South American Biomass Burning Analysis (SAMBBA field experiment

    Directory of Open Access Journals (Sweden)

    J. Brito

    2014-11-01

    Full Text Available This paper investigates the physical and chemical characteristics of aerosols at ground level at a site heavily impacted by biomass burning. The site is located near Porto Velho, Rondônia, in the southwestern part of the Brazilian Amazon rainforest, and was selected for the deployment of a large suite of instruments, among them an Aerosol Chemical Speciation Monitor. Our measurements were made during the South American Biomass Burning Analysis (SAMBBA field experiment, which consisted of a combination of aircraft and ground-based measurements over Brazil, aimed to investigate the impacts of biomass burning emissions on climate, air quality, and numerical weather prediction over South America. The campaign took place during the dry season and the transition to the wet season in September/October 2012. During most of the campaign, the site was impacted by regional biomass burning pollution (average CO mixing ratio of 0.6 ppm, occasionally superimposed by intense (up to 2 ppm of CO, freshly emitted biomass burning plumes. Aerosol number concentrations ranged from ~1000 cm−3 to peaks of up to 35 000 cm−3 (during biomass burning (BB events, corresponding to an average submicron mass mean concentrations of 13.7 μg m−3 and peak concentrations close to 100 μg m−3. Organic aerosol strongly dominated the submicron non-refractory composition, with an average concentration of 11.4 μg m−3. The inorganic species, NH4, SO4, NO3, and Cl, were observed, on average, at concentrations of 0.44, 0.34, 0.19, and 0.01 μg m−3, respectively. Equivalent black carbon (BCe ranged from 0.2 to 5.5 μg m−3, with an average concentration of 1.3 μg m−3. During BB peaks, organics accounted for over 90% of total mass (submicron non-refractory plus BCe, among the highest values described in the literature. We examined the ageing of biomass burning organic aerosol (BBOA using the changes in the H : C and O : C ratios, and found that throughout most of the

  5. Criticality qualification of a new Monte Carlo code for reactor core analysis

    International Nuclear Information System (INIS)

    In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.

  6. Criticality qualification of a new Monte Carlo code for reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Catsaros, N. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Gaveau, B. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Jaekel, M. [Laboratoire de Physique Theorique, Ecole Normale Superieure, 24 rue Lhomond, 75231 Paris (France); Maillard, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); CNRS-IDRIS, Bt 506, BP167, 91403 Orsay (France); CNRS-IN2P3, 3 rue Michel Ange, 75794 Paris (France); Maurel, G. [Faculte de Medecine, Universite Paris VI, 27 rue de Chaligny, 75012 Paris (France); MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Savva, P., E-mail: savvapan@ipta.demokritos.g [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Silva, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Varvayanni, M.; Zisis, Th. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece)

    2009-11-15

    In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.

  7. Magnetic resonance imaging in laboratory petrophysical core analysis

    Science.gov (United States)

    Mitchell, J.; Chandrasekera, T. C.; Holland, D. J.; Gladden, L. F.; Fordham, E. J.

    2013-05-01

    Magnetic resonance imaging (MRI) is a well-known technique in medical diagnosis and materials science. In the more specialized arena of laboratory-scale petrophysical rock core analysis, the role of MRI has undergone a substantial change in focus over the last three decades. Initially, alongside the continual drive to exploit higher magnetic field strengths in MRI applications for medicine and chemistry, the same trend was followed in core analysis. However, the spatial resolution achievable in heterogeneous porous media is inherently limited due to the magnetic susceptibility contrast between solid and fluid. As a result, imaging resolution at the length-scale of typical pore diameters is not practical and so MRI of core-plugs has often been viewed as an inappropriate use of expensive magnetic resonance facilities. Recently, there has been a paradigm shift in the use of MRI in laboratory-scale core analysis. The focus is now on acquiring data in the laboratory that are directly comparable to data obtained from magnetic resonance well-logging tools (i.e., a common physics of measurement). To maintain consistency with well-logging instrumentation, it is desirable to measure distributions of transverse (T2) relaxation time-the industry-standard metric in well-logging-at the laboratory-scale. These T2 distributions can be spatially resolved over the length of a core-plug. The use of low-field magnets in the laboratory environment is optimal for core analysis not only because the magnetic field strength is closer to that of well-logging tools, but also because the magnetic susceptibility contrast is minimized, allowing the acquisition of quantitative image voxel (or pixel) intensities that are directly scalable to liquid volume. Beyond simple determination of macroscopic rock heterogeneity, it is possible to utilize the spatial resolution for monitoring forced displacement of oil by water or chemical agents, determining capillary pressure curves, and estimating

  8. CORE

    DEFF Research Database (Denmark)

    Krigslund, Jeppe; Hansen, Jonas; Hundebøll, Martin;

    2013-01-01

    different flows. Instead of maintaining these approaches separate, we propose a protocol (CORE) that brings together these coding mechanisms. Our protocol uses random linear network coding (RLNC) for intra- session coding but allows nodes in the network to setup inter- session coding regions where flows...... increase the benefits of XORing by exploiting the underlying RLNC structure of individual flows. This goes beyond providing additional reliability to each individual session and beyond exploiting coding opportunistically. Our numerical results show that CORE outperforms both forwarding and COPE......-like schemes in general. More importantly, we show gains of up to 4 fold over COPE-like schemes in terms of transmissions per packet in one of the investigated topologies....

  9. Code Coupling for Multi-Dimensional Core Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-05-15

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  10. A Small LWR Core Design Using ThO{sub 2}-UO{sub 2} and Fully Ceramic Micro-encapsulated Fuels for TRU burning

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Gonghoon; Hong, Ser Gi [Kyung Hee Univ., Yongin (Korea, Republic of); Lee, Kyung Hoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    All calculations were performed by using DeCART (Deterministic Core Analysis based on Ray Tracing) and MASTER (Multipurpose Analyzer for static and Transient Effects of Reactors) code 4, 5. The results show that new core has the TRU destruction rates of ∼21% per 1400EFPD (Effective Full Power Day). We designed the new core having UO{sub 2}-ThO{sub 2} pins and TRU FCM pins to effectively destroy TRU nuclides from LWR spent fuel. The results of the analyses show that the new core has the TRU destruction rates of ∼21% per 1400EFPD (Effective Full Power Day) without degradation of MTC. In this paper, the new 308MWt LWR core using new fuel assembly designs is designed to efficiently destroy TRU (transuranics) nuclide without degradation of safety aspects by using ThO{sub 2}-UO{sub 2} fuel pins and FCM (Fully Ceramic Micro-encapsulated) fuel pins containing TRU fuel particles{sup 1,2,3.} Thorium oxide fuel has been known as fertile, meaning it produces a fissile isotope, which can contribute to the extension of the fuel cycle length. Additionally, the use of thorium fuel offers proliferation resistance because it produces much smaller amount of TRU nuclides than the uranium fuel in commercial reactors.

  11. Gas-core reactor power transient analysis. Final report

    International Nuclear Information System (INIS)

    The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of the study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process. (auth)

  12. Spectroscopic modeling and analysis of plasma conditions in implosion cores

    Science.gov (United States)

    Golovkin, Igor E.

    In this dissertation we discuss the effects of opacity and plasma gradients on the analysis and interpretation of Ar K-shell line emission from Ar-doped inertial confinement fusion (ICF) experiments, and introduce a spectroscopic technique for the determination of core plasma gradients. In particular, the Ar Heβ composite spectral feature is used for core plasma temperature and density diagnostics. We present a versatile, spectroscopic-quality Non-Local-Thermodynamic- Equilibrium radiation transport model that takes into account the effects of collisional-radiative atomic kinetics, plasma gradients, Stark-broadened line shapes and radiation transport. The code computes the radiative properties of the plasma, and it can be easily adapted to treat different problems of spectra formation. We discuss the importance of high-order satellite emission in the formation of Heβ spectral feature, and the interpretation of core averaged electron temperatures and densities extracted from space integrated spectra of non- uniform plasmas. We also present an application of Genetic Algorithms to the analysis of experimental X-ray spectra. This algorithm drives the search for plasma parameters that yield the best fits to experimental spectra. We discuss the applicability of Case Injected Genetic Algorithms to accelerate analysis of spectra. Furthermore, we introduce a novel method for the determination of plasma temperature and density gradients in imploded cores. The gradients are extracted from the self-consistent analysis of time-resolved X-ray spectra and spatial emissivity distributions obtained from X-ray monochromatic images. In this case, the search in the complex parameter space of gradient functions is driven by a multi-objective Niched Pareto Genetic Algorithm. We discuss the analysis of time resolved spectra recorded during Ar-doped ICF implosions at the NOVA laser facility. Time histories of core averaged electron densities and temperatures during the collapse of the

  13. Preliminary analysis of the proposed BN-600 benchmark core

    International Nuclear Information System (INIS)

    The Indira Gandhi Centre for Atomic Research is actively involved in the design of Fast Power Reactors in India. The core physics calculations are performed by the computer codes that are developed in-house or by the codes obtained from other laboratories and suitably modified to meet the computational requirements. The basic philosophy of the core physics calculations is to use the diffusion theory codes with the 25 group nuclear cross sections. The parameters that are very sensitive is the core leakage, like the power distribution at the core blanket interface etc. are calculated using transport theory codes under the DSN approximations. All these codes use the finite difference approximation as the method to treat the spatial variation of the neutron flux. Criticality problems having geometries that are irregular to be represented by the conventional codes are solved using Monte Carlo methods. These codes and methods have been validated by the analysis of various critical assemblies and calculational benchmarks. Reactor core design procedure at IGCAR consists of: two and three dimensional diffusion theory calculations (codes ALCIALMI and 3DB); auxiliary calculations, (neutron balance, power distributions, etc. are done by codes that are developed in-house); transport theory corrections from two dimensional transport calculations (DOT); irregular geometry treated by Monte Carlo method (KENO); cross section data library used CV2M (25 group)

  14. Uncertainty analysis of moderate- versus coarse-scale satellite fire products for quantifying agricultural burning: Implications for Air Quality in European Russia, Belarus, and Lithuania

    Science.gov (United States)

    McCarty, J. L.; Krylov, A.; Prishchepov, A. V.; Banach, D. M.; Potapov, P.; Tyukavina, A.; Rukhovitch, D.; Koroleva, P.; Turubanova, S.; Romanenkov, V.

    2015-12-01

    Cropland and pasture burning are common agricultural management practices that negatively impact air quality at a local and regional scale, including contributing to short-lived climate pollutants (SLCPs). This research focuses on both cropland and pasture burning in European Russia, Lithuania, and Belarus. Burned area and fire detections were derived from 500 m and 1 km Moderate Resolution Imaging Spectroradiometer (MODIS), 30 m Landsat 7 Enhanced Thematic Mapper Plus (ETM+), and Landsat 8 Operational Land Imager (OLI) data. Carbon, particulate matter, volatile organic carbon (VOCs), and harmful air pollutants (HAPs) emissions were then calculated using MODIS and Landsat-based estimates of fire and land-cover and land-use. Agricultural burning in Belarus, Lithuania, and European Russia showed a strong and consistent seasonal geographic pattern from 2002 to 2012, with the majority of fire detections occurring in March - June and smaller peak in July and August. Over this 11-year period, there was a decrease in both cropland and pasture burning throughout this region. For Smolensk Oblast, a Russian administrative region with comparable agro-environmental conditions to Belarus and Lithuania, a detailed analysis of Landsat-based burned area estimations for croplands and pastures and field data collected in summer 2014 showed that the agricultural burning area can be up to 10 times higher than the 1 km MODIS active fire estimates. In general, European Russia is the main source of agricultural burning emissions compared to Lithuania and Belarus. On average, all cropland burning in European Russia as detected by the MCD45A1 MODIS Burned Area Product emitted 17.66 Gg of PM10 while annual burning of pasture in Smolensk Oblast, Russia as detected by Landsat burn scars emitted 494.85 Gg of PM10, a 96% difference. This highlights that quantifying the contribution of pasture burning and burned area versus cropland burning in agricultural regions is important for accurately

  15. Fault tree analysis on BWR core spray system

    International Nuclear Information System (INIS)

    Fault Trees which describe the failure modes for the Core Spray System function in the Browns Ferry Nuclear Plant (BWR 1065MWe) were developed qualitatively and quantitatively. The unavailability for the Core Spray System was estimated to be 1.2 x 10-3/demand. It was found that the miscalibration of four reactor pressure sensors or the failure to open of the two inboard valves (FCV 75-25 and 75-53) could reduce system reliability significantly. It was recommended that the pressure sensors would be calibrated independently. The introduction of the redundant inboard valves could improve the system reliability. Thus this analysis method was verified useful for system analysis. The detailed test and maintenance manual and the informations on the control logic circuits of each active component are necessary for further analysis. (author)

  16. Optimization analysis of the nuclear fuel cycle transition to the last core

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L.; Blanco, J. [Union Fenosa Generacion, Madrid (Spain)

    2001-07-01

    The Zorita NPP was the first Spanish commercial nuclear reactor connected to the grid. It is a 160 MW one loop PWR, Westinghouse design, owned by UFG, in operation since 1968. The configuration of the reactor core is based on 69 fuel elements type 14 x 14, the standard reload of the present equilibrium cycle being based on 16 fuel elements with 3.6% enrichment in {sup 235}U. In order to properly plan the nuclear fuel management of the transition cycles to its end of life, presently foreseen by 2008, an based on the non-reprocessing option required by the policy of the Spanish Administration, a technical-economical optimization analysis has been performed. As a result, a fuel management strategy has been defined looking for getting simultaneously the minimum integral fuel cost of the transition from the present equilibrium cycle to the last core, as well as the minimum residual worth of the fuel remaining in the core after the final outage. Based on the ''lessons learned'' derived from the study, the time margin for the decision making has been determined, and a planning of the nuclear fuel supply for the transition reloads, specifying both the number of fuel elements and their enrichment in {sup 235}U, as been prepared. Finally, based on the calculated economical worth of the partially burned fuel of the last core, after the end of its operation cycle, a financial cover for yearly compensation from now on of the foreseen final lost has been elaborated. Most of the conceptual conclusions obtained are applicable to the other commercial nuclear reactors in operation owned by UFG, so that they are understood to be of general interest and broad application to commercial PWR. (author)

  17. Optimization analysis of the nuclear fuel cycle transition to the last core

    International Nuclear Information System (INIS)

    The Zorita NPP was the first Spanish commercial nuclear reactor connected to the grid. It is a 160 MW one loop PWR, Westinghouse design, owned by UFG, in operation since 1968. The configuration of the reactor core is based on 69 fuel elements type 14 x 14, the standard reload of the present equilibrium cycle being based on 16 fuel elements with 3.6% enrichment in 235U. In order to properly plan the nuclear fuel management of the transition cycles to its end of life, presently foreseen by 2008, an based on the non-reprocessing option required by the policy of the Spanish Administration, a technical-economical optimization analysis has been performed. As a result, a fuel management strategy has been defined looking for getting simultaneously the minimum integral fuel cost of the transition from the present equilibrium cycle to the last core, as well as the minimum residual worth of the fuel remaining in the core after the final outage. Based on the ''lessons learned'' derived from the study, the time margin for the decision making has been determined, and a planning of the nuclear fuel supply for the transition reloads, specifying both the number of fuel elements and their enrichment in 235U, as been prepared. Finally, based on the calculated economical worth of the partially burned fuel of the last core, after the end of its operation cycle, a financial cover for yearly compensation from now on of the foreseen final lost has been elaborated. Most of the conceptual conclusions obtained are applicable to the other commercial nuclear reactors in operation owned by UFG, so that they are understood to be of general interest and broad application to commercial PWR. (author)

  18. A Retrospective Analysis of the Burn Injury Patients Records in the Emergency Department, an Epidemiologic Study

    OpenAIRE

    Nilgün Aksoy; Senay Arli; Ozlem Yigit

    2014-01-01

    Introduction: Burns can be very destructive, and severely endanger the health and lives of humans. It maybe cause disability and even psychological trauma in individuals. . Such an event can also lead to economic burden on victim’s families and society. The aim of our study is to evaluate epidemiology and outcome of burn patients referring to emergency department. Methods: This is a cross-sectional study was conducted by evaluation of patients’ files and forensic reports of burned patients’ r...

  19. A comparative analysis of potential impact area of common sugar cane burning methods

    Science.gov (United States)

    Hiscox, A. L.; Flecher, S.; Wang, J. J.; Viator, H. P.

    2015-04-01

    The negative effects of agricultural burning are well-known, although the actual impact area of different activities has not previously been quantified. An elastic backscatter lidar system was used to examine the impact-area size and dispersion of smoke generated from different types of sugarcane burning activities; pre-harvest (standing) burning and post-harvest (ground) burning. Experiments were conducted in the sugarcane harvest season of 2010 and 2011 at two locations in Louisiana, USA. Current dispersion theory would suggest that the primary difference between burn types would be primarily in the initial plume rise, but that the overall plume shape would remain the same. However, remotely sensed lidar data with the capability to measure plume dispersion and the short time dynamics of plume location showed pre-harvest (standing) burning produced a larger plume with greater rise and more spread within the 300 m of the plume, but a decrease in dispersion, but not concentration further downwind. Post-harvest (ground) burning produced a more traditional plume shape, but still exceeded impact area predictions near the source. Moreover, large changes in plume size can occur with small increases in wind speed. These are the first instrumented measurements of the meteorological effects of the different types of sugarcane burning. These results indicate that ground burning is preferable, but should be avoided in lower wind speed conditions.

  20. Study of reference material for NMR core analysis

    International Nuclear Information System (INIS)

    Reference Materials of NMR core experimental analysis have been studied systematically in this paper. According to the national standard criterion, a set of NMR reference materials have been made successfully. These reference materials include fluid pattern-D2O and H2O, divergence of glass grain muster, and the solid ceramics. They have been applied to the core analysis in Daqing, Xinjiang, Dagang Oilfield, etc. The results show, fluid pattern, divergence of glass grain muster as well as ceramics reference materials have different calibration results of core NMR porosity as a result of their different relaxation mechanisms. Fluid pattern is suitable for NMR porosity calibration in Berea sandstone as well as terrestrial sediment sand with little shale content and well sorted. Ceramics reference material is applied to shaly sand with average sorted. Both fluid pattern and solid ceramics material are not suitable for calibration in sandstone sample with high clay content, complex lithology with rich paramagnetic materials. It is suggested that the representative natural core sample should be selected as regional reference material to calibrate NMR porosity of complex lithology, or internal magnetic field gradient in complex rock grain/fluid system should be studied in order to get corrected NMR porosity. (authors)

  1. Analysis of core damage frequency: Surry, Unit 1 internal events

    International Nuclear Information System (INIS)

    This document contains the accident sequence analysis of internally initiated events for the Surry Nuclear Station, Unit 1. This is one of the five plant analyses conducted as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 documents the risk of a selected group of nuclear power plants. The work performed and described here is an extensive of that published in November 1986 as NUREG/CR-4450, Volume 3. It addresses comments form numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved. The context and detail of this report are directed toward PRA practitioners who need to know how the work was performed and the details for use in further studies. The mean core damage frequency at Surry was calculated to be 4.05-E-5 per year, with a 95% upper bound of 1.34E-4 and 5% lower bound of 6.8E-6 per year. Station blackout type accidents (loss of all AC power) were the largest contributors to the core damage frequency, accounting for approximately 68% of the total. The next type of dominant contributors were Loss of Coolant Accidents (LOCAs). These sequences account for 15% of core damage frequency. No other type of sequence accounts for more than 10% of core damage frequency. 49 refs., 52 figs., 70 tabs

  2. Transient Safety Analysis of Fast Spectrum TRU Burning LWRs with Internal Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Zazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hill, Bob [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-31

    The objective of this proposal was to perform a detailed transient safety analysis of the Resource-Renewable BWR (RBWR) core designs using the U.S. NRC TRACE/PARCS code system. This project involved the same joint team that has performed the RBWR design evaluation for EPRI and therefore be able to leverage that previous work. And because of their extensive experience with fast spectrum reactors and parfait core designs, ANL was also part the project team. The principal outcome of this project was the development of a state-of-the-art transient analysis capability for GEN-IV reactors based on Monte Carlo generated cross sections and the US NRC coupled code system TRACE/PARCS, and a state-of-the-art coupled code assessment of the transient safety performance of the RBWR.

  3. High Resolution Continuous Flow Analysis System for Polar Ice Cores

    Science.gov (United States)

    Dallmayr, Remi; Azuma, Kumiko; Yamada, Hironobu; Kjær, Helle Astrid; Vallelonga, Paul; Azuma, Nobuhiko; Takata, Morimasa

    2014-05-01

    In the last decades, Continuous Flow Analysis (CFA) technology for ice core analyses has been developed to reconstruct the past changes of the climate system 1), 2). Compared with traditional analyses of discrete samples, a CFA system offers much faster and higher depth resolution analyses. It also generates a decontaminated sample stream without time-consuming sample processing procedure by using the inner area of an ice-core sample.. The CFA system that we have been developing is currently able to continuously measure stable water isotopes 3) and electrolytic conductivity, as well as to collect discrete samples for the both inner and outer areas with variable depth resolutions. Chemistry analyses4) and methane-gas analysis 5) are planned to be added using the continuous water stream system 5). In order to optimize the resolution of the current system with minimal sample volumes necessary for different analyses, our CFA system typically melts an ice core at 1.6 cm/min. Instead of using a wire position encoder with typical 1mm positioning resolution 6), we decided to use a high-accuracy CCD Laser displacement sensor (LKG-G505, Keyence). At the 1.6 cm/min melt rate, the positioning resolution was increased to 0.27mm. Also, the mixing volume that occurs in our open split debubbler is regulated using its weight. The overflow pumping rate is smoothly PID controlled to maintain the weight as low as possible, while keeping a safety buffer of water to avoid air bubbles downstream. To evaluate the system's depth-resolution, we will present the preliminary data of electrolytic conductivity obtained by melting 12 bags of the North Greenland Eemian Ice Drilling (NEEM) ice core. The samples correspond to different climate intervals (Greenland Stadial 21, 22, Greenland Stadial 5, Greenland Interstadial 5, Greenland Interstadial 7, Greenland Stadial 8). We will present results for the Greenland Stadial -8, whose depths and ages are between 1723.7 and 1724.8 meters, and 35.520 to

  4. Preliminary core mechanics analysis for KALIMER by CRAMP code

    International Nuclear Information System (INIS)

    CRAMP code is designed to solve the problem of mutually interacting and distorting sub-assemblies in a fast breeder reactor. It is the UK's main core mechanics design tool and is currently being used in the design of EFR. This report contains the results of preliminary core mechanics calculations for KALIMER core configuration by the updated version of CRAMP code. The base case calculation s on KALIMER core, and the sensitivity studies (to investigate effect of main design parameter) are carried out by the code which was updated with material subroutine in CRAMP to model the characteristics of HT9. Sensitivity studies include following cases; (1) with gaps at LRP and URP reduced to 0.4 mm at 386 dg C (2) with 0.2 mm radial clearance around both nosepiece at seals (3) with flexibility at LRP reduced by a factor of 2 (4) with stiffness of nosepiece increased by a factor of 2 (5) with reduced creep (6) combined with gap reduced 0.4 mm and 0.2 mm clearance at seals (7) with IVS position replaced to dummy ducts (8) with initial bow at every duct except S/A 1. From each calculation, the data obtained and compared are as follows; (a) contact forces between pads (b) gaps between pads (c) duct dynamic behavior of duct bowing and dilation i.e. the variation of bowing and dilation with time) (d) reactivity change (e) maximum LRP contact force, maximum URP contact force and maximum nosepiece force vs interval number for the base case. The design requirements and the specifications for KALIMER assembly ducts are reviewed, and preliminary core mechanics analysis for KALIMER core configuration are carried out. (Author). 7 refs., 2 tabs., 50 figs

  5. Preliminary Core Analysis of a Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Chang Keun; Chang, Jongwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Venneri, Francesco [Ultra Safe Nuclear Corporation, Los Alamos (United States); Hawari, Ayman [NC State Univ., Raleigh (United States)

    2014-05-15

    The Micro Modular Reactor (MMR) will be 'melt-down proof'(MDP) under all circumstances, including the complete loss of coolant, and will be easily transportable and retrievable, and suitable for use with very little site preparation and Balance of Plant (BOP) requirements for a variety of applications, from power generation and process heat applications in remote areas to grid-unattached locations, including ship propulsion. The Micro Modular Reactor design proposed in this paper has 3 meter diameter core (2 meter active core) which is suitable for 'factory manufactured' and has few tens year of service life for remote deployment. We confirmed the feasibility of long term service life by a preliminary neutronic analysis in terms of the excess reactivity, the temperature feedback coefficient, and the control margins. We are able to achieve a reasonably long core life time of 5 ∼ 10 years under typical thermal hydraulic condition of a helium cooled reactor. However, on a situation where longer service period and safety is important, we can reduce the power density to the level of typical pebble bed reactor. In this case we can design 10 MWt MMR with core diameter for 10 ∼ 40 years core life time without much loss in the economics. Several burnable poisons are studied and it is found that erbia mixed in the compact matrix seems reasonably good poison. The temperature feedback coefficients were remaining negative during lifetime. Drum type control rods at reflector region and few control rods inside core region are sufficient to control the reactivity during operation and to achieve safe cold shutdown state.

  6. Efficiency Analysis of Technological Methods for Reduction of NOx Emissions while Burning Hydrocarbon Fuels in Heat and Power Plants

    Directory of Open Access Journals (Sweden)

    S. Kabishov

    2013-01-01

    Full Text Available The paper contains a comparative efficiency analysis pertaining to application of existing technological methods for suppression of nitric oxide formation in heating boilers of heat generators. A special attention has been given to investigation of NOx  emission reduction while burning hydrocarbon fuel with the help of oxygen-enriched air. The calculations have demonstrated that while enriching oxidizer with the help of oxygen up to 50 % (by volume it is possible to reduce volume of NOx formation (while burning fuel unit by 21 %.

  7. Al4C3 Hydration Thermochemical Analysis for Burned Carbon-containing Refractories with Al

    Institute of Scientific and Technical Information of China (English)

    YANG Ding'ao; YU Zhiming; FAN Liuwu

    2003-01-01

    In this paper, X-ray diffractogram analysis and SEM observation of Al4 C3 formed at high temperature from carbon-containing refractories with Al have been carried out.Aluminum added to carbon-containing refractories reacts with C(s)to form Al4 C3(s) gradually during heating from 600 ℃ to 1200℃.It is considered that the interlocked structure of Al4 C3 plate crystals promotes the outstanding increase of hot modulus of rupture of carbon-containing refractories with Al. The HMOR of carbon-containing refractories added with Al additive from 0 to 5wt% increases by 2.8 times being from 6.5MPa to 18.2MPa.After a thermochemical calculation for hydration reaction processes of Al4 C3 and H2O(g), the equilibrium partial pressure chart of H2O(g)in H2O-Al4C3-Al(OH)3 system vs various temperatures has been attained . The H2O (g) partial pressure in the air needed for the Al4 C3 hydration reaction is no more than 10~18 atm at the temperature below 120℃.It is considered that the burned carbon-containing refractories with Al is extremely easy to hydrate and the cracking of burned carbon-containing refractories is generated because that the hydration expansion is 2.11 times during transforming from Al4 C3 to Al(OH)3.The fundamental measure against hydration of the refractories is to insulate the refractories from H2O(g)by various means such as pitch impregnation or other sealing materials.

  8. Analysis of burns caused by pre-filled gas canisters used for lamps or portable camping stoves.

    Science.gov (United States)

    Desouches, C; Salazard, B; Romain, F; Karra, C; Lavie, A; Volpe, C Della; Manelli, J C; Magalon, G

    2006-12-01

    The use of pre-filled valveless gas canisters for lamps or camping stoves has caused a number of serious burn incidents. We performed a retrospective analysis of all of the patients who were victims of such incidents admitted to the Marseille Burn Centre between January 1990 and March 2004. There were a total of 21 patients burned in such conditions. Adult males made up the majority of the victims of this sort. Lesions were often extensive (60% of the patients were burned over more than 10% of their body surface) and systematically deep. In order of frequency, burn locations were: the lower limbs, the upper limbs, the hands and the face. The incidents principally occurred during replacement of the canister near an open flame. The marketing of a canister with a valve in order to avoid gas leaks did not cause the old canisters to be taken off the market. On the contrary, European Safety Standard EN417, updated in October 2003, validated the use of these valveless canisters. The severity of the lesions caused and the existence of safe equivalent products requires the passage of a law that forbids valveless canisters. PMID:16982156

  9. Ice cores

    DEFF Research Database (Denmark)

    Svensson, Anders

    Ice cores from Antarctica, from Greenland, and from a number of smaller glaciers around the world yield a wealth of information on past climates and environments. Ice cores offer unique records on past temperatures, atmospheric composition (including greenhouse gases), volcanism, solar activity......, dustiness, and biomass burning, among others. In Antarctica, ice cores extend back more than 800,000 years before present (Jouzel et al. 2007), whereas. Greenland ice cores cover the last 130,000 years...

  10. Thermal Analysis of Air-Core Power Reactors

    OpenAIRE

    Zhao Yuan; Jun-jia He; Yuan Pan; Xiao-gen Yin; Can Ding; Shao-fei Ning; Hong-lei Li

    2013-01-01

    A fluid-thermal coupled analysis based on FEM is conducted. The inner structure of the coils is built with consideration of both the structural details and the simplicity; thus, the detailed heat conduction process is coupled with the computational fluid dynamics in the thermal computation of air-core reactors. According to the simulation results, 2D temperature distribution results are given and proved by the thermal test results of a prototype. Then the temperature results are used to calcu...

  11. Burn Rehabilitation

    OpenAIRE

    Koray Aydemir; Mehmet Ali Taşkaynatan

    2011-01-01

    Burn injuries are important in terms of causing serious disability and threatening life. With the establishment of modern burn treatment units and advances in acute care management contributed to a reduced mortality rate over the last decades. As a result of improved outcome, more attention has to be given to a comprehensive burn rehabilitation program. Burn rehabilitation is a process that starts from day of admission and continues for months or sometimes years after the initial event. The t...

  12. Analysis of Homogeneous BFS-73-1 MA Benchmark Core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Il; Yoo, Jae Woon; Song, Hoon; Jang, Jin Wook; Kim, Yeong Il

    2007-06-15

    Analysis of BFS-73-1 critical assembly for MA transmutation has been carried out by using K-CORE system mainly, DIF3D code. All of measured data are compared with the results of analysis and sensitiveness of calculation conditions, for example, number of neutron energy groups, mesh size used, and analysis method, are assessed. Effective multiplication factor was in good agreement within experimental uncertainty in both transport and diffusion calculations. Fission rate distribution of U-235 and U-238 is also fairly good agreed with experimental results within maximum 5% in core region. But large discrepancy was seen in blanket region and it tends to increase as the location closes to core boundary. Largest error of relative reaction rate ratio was seen in Am-243 fission and U-238 capture. For the case of Am-243, the error lay on appropriate range considering the measurement uncertainty of that as 4.6%. Sample reactivity worths for scattering dominant isotope was greatly differ from the experimental results, which can be explained in terms of sample heterogeneity effect, sample self shielding and finally resonance bilinear correction effect. These effects will be evaluated as future study. C/E of effective delayed neutron fraction is within 4%, which is within the measurement uncertainty.

  13. Monte Carlo analysis of Musashi TRIGA mark II reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Tetsuo [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan)

    1999-08-01

    The analysis of the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). Effective multiplication factors (k{sub eff}) for the several fuel-loading patterns including the initial core criticality experiment, the fuel element and control rod reactivity worth as well as the neutron flux measurements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated k{sub eff} overestimated the experimental data by about 1.0%{delta}k/k for both the initial core and the several fuel-loading arrangements. The calculated reactivity worths of control rod and fuel element agree well the measured ones within the uncertainties. The comparison of neutron flux distribution was consistent with the experimental ones which were measured by activation methods at the sample irradiation tubes. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicated that the Monte Carlo model is enough to simulate the Musashi TRIGA-II reactor core. (author)

  14. Core disruptive accident analysis using ASTERIA-FBR

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization (JNES) is developing a core disruptive accident analysis code, ASTERIA-FBR, which tightly couples the thermal-hydraulics and the neutronics to simulate the core behavior during core disruptive accidents (CDA) of fast breeder reactors (FBRs). ASTERIA-FBR consists of the three-dimensional thermal-hydraulics calculation module: CONCORD, the fuel pin behavior calculation module: FEMAXI-FBR, and the space-time neutronics module: Dynamic-GMVP or PARTISN/RKIN. This paper describes a comparison between characteristics of GMVP and PARTISN and summarizes the challenging issues on applying Dynamic-GMVP to the calculation against unprotected loss-of-flow (ULOF) event which is a typical initiator of core disruptive accident of FBR. It was found that Dynamic-GMVP is confirmed to be basically applicable to the CDA phenomena. It was found that, however, applying GMVP to the CDA calculation is less reasonable than PARTISN since the calculation load of GMVP is too large to meet the required calculation accuracy, although the Monte-Carlo method is based on the actual neutron behavior without any discretization of space and energy. The statistical error included in the calculation results may affect the super-prompt criticality during ULOF event and thus the amount of released energy

  15. An Efficient Analysis Methodology for Fluted-Core Composite Structures

    Science.gov (United States)

    Oremont, Leonard; Schultz, Marc R.

    2012-01-01

    The primary loading condition in launch-vehicle barrel sections is axial compression, and it is therefore important to understand the compression behavior of any structures, structural concepts, and materials considered in launch-vehicle designs. This understanding will necessarily come from a combination of test and analysis. However, certain potentially beneficial structures and structural concepts do not lend themselves to commonly used simplified analysis methods, and therefore innovative analysis methodologies must be developed if these structures and structural concepts are to be considered. This paper discusses such an analysis technique for the fluted-core sandwich composite structural concept. The presented technique is based on commercially available finite-element codes, and uses shell elements to capture behavior that would normally require solid elements to capture the detailed mechanical response of the structure. The shell thicknesses and offsets using this analysis technique are parameterized, and the parameters are adjusted through a heuristic procedure until this model matches the mechanical behavior of a more detailed shell-and-solid model. Additionally, the detailed shell-and-solid model can be strategically placed in a larger, global shell-only model to capture important local behavior. Comparisons between shell-only models, experiments, and more detailed shell-and-solid models show excellent agreement. The discussed analysis methodology, though only discussed in the context of fluted-core composites, is widely applicable to other concepts.

  16. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author)

  17. Calculation and Analysis of B/T (Burning and/or Transmutation Rate of Minor Actinides and Plutonium Performed by Fast B/T Reactor

    Directory of Open Access Journals (Sweden)

    Marsodi

    2006-01-01

    Full Text Available Calculation and analysis of B/T (Burning and/or Transmutation rate of MA (minor actinides and Pu (Plutonium has been performed in fast B/T reactor. The study was based on the assumption that the spectrum shift of neutron flux to higher side of neutron energy had a potential significance for designing the fast B/T reactor and a remarkable effect for increasing the B/T rate of MA and/or Pu. The spectrum shifts of neutron have been performed by change MOX to metallic fuel. Blending fraction of MA and or Pu in B/T fuel and the volume ratio of fuel to coolant in the reactor core were also considered. Here, the performance of fast B/T reactor was evaluated theoretically based on the calculation results of the neutronics and burn-up analysis. In this study, the B/T rate of MA and/or Pu increased by increasing the blending fraction of MA and or Pu and by changing the F/C ratio. According to the results, the total B/T rate, i.e. [B/T rate]MA + [B/T rate]Pu, could be kept nearly constant under the critical condition, if the sum of the MA and Pu inventory in the core is nearly constant. The effect of loading structure was examined for inner or outer loading of concentric geometry and for homogeneous loading. Homogeneous loading of B/T fuel was the good structure for obtaining the higher B/T rate, rather than inner or outer loading

  18. LWR core safety analysis with Areva's 3-dimensional methods

    International Nuclear Information System (INIS)

    The quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools and an extensive validation base. Sophisticated 3-dimensional core models ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. The validation base includes measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models achieve reliable and comprehensive results for a wide range of applications. As an example an overview of the application experience as well as the validation base of AREVA's 3-dimensional codes is given. The importance and necessity of the comprehensive 3-dimensional methodology is illustrated with examples of a BWR and PWR safety analysis. For BWR transient application the analysis of regional power oscillations is considered and regarding the PWR safety analysis an example referring to fast enthalpy rise and the maximum fuel temperature caused by a rod ejection accident is shown. (orig.)

  19. Burn Rehabilitation

    Directory of Open Access Journals (Sweden)

    Koray Aydemir

    2011-07-01

    Full Text Available Burn injuries are important in terms of causing serious disability and threatening life. With the establishment of modern burn treatment units and advances in acute care management contributed to a reduced mortality rate over the last decades. As a result of improved outcome, more attention has to be given to a comprehensive burn rehabilitation program. Burn rehabilitation is a process that starts from day of admission and continues for months or sometimes years after the initial event. The term ‘burn rehabilitation’ incorporates the physical, physiological and social aspects of care. Burns can leave a patient with severely debilitating and deforming contractures, which can lead to significant disability when left untreated. Burn rehabilitation aims to prevent the possible complications, minimalize joint contractures and deformities, increase range of motion, control hypertrophic scarring, achieve the best possible functional capacity and to regain the patients vocational and recreational activities. (Journal of the Turkish Society Intensive Care 2011; 9 Suppl: 70-7

  20. Development of the transient analysis of the LMR core

    International Nuclear Information System (INIS)

    FX2-TH is a two-dimensional multigroup fine-mesh diffusion code for the transient analysis of the LMR core and has been designed to treat transients initiated by such mechanism as sodium voiding, motion of fuel and/or structural material, and control rod motion. It takes account of feedback effects from changes in both the fuel temperature and average coolant temperatures through Doppler broadening and change in coolant density, respectively. However, the thermal expansion of the fuel and structural material causes one of the most important reactivity feedback effects on the transient behavior of the LMR core loaded with metallic fuels. This paper describes the thermal reactivity feedback model implemented into the code and the three-dimensional, steady-state nodal diffusion code for the hexagonal-z geometry, which has been developed to serve as a standard neutronics solver of a future transient code for LMR's

  1. Seismic response analysis for prismatic fuel HTGR core

    International Nuclear Information System (INIS)

    For high-temperature gas cooled reactors (HTGR) with prismatic fuels, their resistance against an earthquake is not fully ascertained yet. Aseismic design studies and also experiments must therefore be made when such a reactor plant is to be installed in areas of high seismicity. This report describes analytical study on the seismic response of a prismatic fuel reactor core, including the following: aseismic core structure, the analysis model and calculation formulae, the effects of various design variables on response charactersitics, and the desired block shape. Three analysis models have been considered for the seismic vibration of the prismatic fuel HTGR core. The first is the impact model, the second ''the spring dashpot model'', and the third ''the dryfriction model''. The calculation has been performed with three models, and these results are nearly the same. The followings were revealed: (1) At low input-wave frequencies, the response value increases with the gap between the blocks. Beyond a certain point, however, the effect of gap is nearly negligible. (2) When the blocks are restrained horizontally by keys, the response value decreases with increase of the key stiffness. The key is thus effective in earthquake resistance. (3) The response value increases with block-stiffness, so that short massive blocks are better for earthquake resistance. (4) The response value decreases with increase of the block damping factor. But beyond a certain point, this effect is only small. (5) Stiffness and damping in the restraint structure for the reactor core do not have much effect in earthquake resistance. (author)

  2. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Su' ud, Zaki, E-mail: szaki@fi.itba.c.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) (Indonesia); Sekimoto, H., E-mail: hsekimot@gmail.com [Research Lab. For Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo (Japan)

    2014-09-30

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  3. EXPERIMENTAL ANALYSIS AND ISHIKAWA DIAGRAM FOR BURN ON EFFECT ON MANGANESE SILICON ALLOY MEDIUM CARBON STEEL SHAFT

    Directory of Open Access Journals (Sweden)

    AsmamawTegegne

    2013-12-01

    Full Text Available Burn on/metal penetration is one of the surface defects of metal castings in general and steel castings in particular. A research on the effect of burn on the six ton medium carbon steel shaft for making a roller of cold rolled steel sheet produced at one of the metals industry was carried out. The shaft was cast using sand casting by pouring through riser/feeding head step by step (with time interval of pouring. As it was required to use foam casting method for better surface finish and dimensional accuracy of the cast, the pattern was prepared from polystyrene and embedded by silica sand. Physical observations, photographic analysis, visual inspection, measurement of depth of penetration and fish bone diagram were used as method of results analysis. The shaft produced has strongly affected by sand sintering (burn on/metal penetration. Many reasons may be the case for these defects, however analysis results showed that the use of poorly designed gating system led to turbulence flow, uncontrollable high temperature fused the silica sand and liquid polystyrene penetrated the poorly reclaimed and rammed sand mold as a result of which eroded sand has penetrated the liquid metal deeply and reacted with it, consequently after solidification and finishing the required 240mm diameter of the shaft has reduced un evenly to 133mm minimum and 229mm maximum mm that end in the rejection of the shaft from the product since it is below the required standard for the designed application. In addition, it was not possible to remove the adhered sand by grinding. Thus burn on is included in mechanical type burn on.

  4. Burned Area Mapping in Greece Using SPOT-4 HRVIR Images and Object-Based Image Analysis

    Directory of Open Access Journals (Sweden)

    Anastasia Polychronaki

    2012-02-01

    Full Text Available The devastating series of fire events that occurred during the summers of 2007 and 2009 in Greece made evident the need for an operational mechanism to map burned areas in an accurate and timely fashion to be developed. In this work, Système pour l’Observation de la Terre (SPOT-4 HRVIR images are introduced in an object-based classification environment in order to develop a classification procedure for burned area mapping. The development of the procedure was based on two images and then tested for its transferability to other burned areas. Results from the SPOT-4 HRVIR burned area mapping showed very high classification accuracies ( 0.86 kappa coefficient, while the object-based classification procedure that was developed proved to be transferable when applied to other study areas.

  5. Analysis of Burning Processes in Turbulent Mixing Axial and Tangential Flows

    Directory of Open Access Journals (Sweden)

    R. I. Essmann

    2009-01-01

    Full Text Available The paper demonstrates that in the case of turbulent diffusion flame tongues the burning process of combined multiphase fuel is determined by flow structure and conditions for mixing various types of fuel and distributed oxidizer flows. It has been determined that the ratio of air  supplied for burning through axial and tangential channels governs a shape of the flame tongue, its size and process intensity that allows efficiently to optimize technological parameters.

  6. Analysis of Secondary Chemistry and Treatment of Burn Wounds with Nonthermal Plasma Induced Effluent

    Science.gov (United States)

    Golkowski, Mark; Plimpton, S. Reed; Golkowski, Czeslaw

    2013-10-01

    Exploitation of non-thermal plasmas in the biomedical setting is a rapidly growing field with a large number of diverse technologies under investigation. Potential applications of such devices range from instrument sterilization to clinical therapy. One of the key hurdles to the implementation of non-thermal plasma technologies in the relatively poor understanding of the chemical processes taking place. Our group has recently completed precise analysis of chemical species created by our indirect exposure non-thermal plasma device with hydrogen peroxide additives. Reactive nitrogen and oxygen species are observed using optical absorption spectroscopy. We report the unique detection of short lived hydroxyl radicals at a significant distance from the discharge using electron paramagnetic spin resonance trapping. The hydroxyl radicals are shown to be generated in secondary ozonide based chemical processes away from the discharge. The plasma device is applied to a porcine model of infected full thickness burn wounds. The bacteria load reduction after treatment with our device is shown to be 10-100 fold improvement over Silvadene which is the main treatment currently used in the clinic. Partially funded by NIH SBIR R43 AI096594.

  7. Analysis of core calculation schemes for advanced water reactors

    International Nuclear Information System (INIS)

    This research thesis addresses the analysis of the core control of sub-moderated water reactors with plutonium fuel and varying spectrum. Firstly, a calculation scheme is defined, based on transport theory for the three existing assembly configurations. It is based on the efficiency analysis of the control cluster and of the flow sheet shape in the assembly. Secondly, studies of the assembly with control cluster and within a theory of diffusion with homogenization or detailed assembly representation are performed by taking the environment into account in order to assess errors. Thirdly, due to the presence of a very efficient absorbent in control clusters, a deeper physical analysis requires the study of the flow gradient existing at the interface between assemblies. A parameter is defined to assess this gradient, and theoretically calculated by using finite elements. Developed software is validated

  8. Design and analysis of PCRV core cavity closure

    International Nuclear Information System (INIS)

    Design requirements and considerations for a core cavity closure which led to the choice of a concrete closure with a toggle hold-down as the design for the Gas-Cooled Fast Breeder Reactor (GCFR) plant are discussed. A procedure for preliminary stress analysis of the closure by means of a three-dimensional finite element method is described. A limited parametric study using this procedure indicates the adequacy of the present closure design and the significance of radial compression developed as a result of inclined support reaction

  9. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  10. Core analysis and CT imaging improve shale completions

    International Nuclear Information System (INIS)

    To improve hydraulic fracturing efficiency in Devonian shales, core analysis and computerized tomography (CT) can provide data for orienting perforations, determining fracture direction, and selecting deviated well trajectories. This article reports on technology tested in a West Virginia well for improving the economics of developing Devonian shale and other low permeability gas reservoirs. With slight production increase per well, Columbia Natural Resources Inc. (CNR) has determined that marginal gas well payout time can be shortened enough to encourage additional drilling. For eight wells completed by CNR in 1992, the absolute open flow (AOF) averaged 116 Mcfd before stimulation. After stimulation using long-standing fracture stimulation procedures, the AOF averaged 500 Mcfd

  11. Growth factor therapy in patients with partial-thickness burns: a systematic review and meta-analysis.

    Science.gov (United States)

    Zhang, Yi; Wang, Tao; He, Jinguang; Dong, Jiasheng

    2016-06-01

    Growth factor (GF) therapy has shown promise in treating a variety of refractory wounds. However, evidence supporting its routine use in burn injury remains uncertain. We performed this systematic review and meta-analysis assessing randomised controlled trials (RCTs) to investigate efficacy and safety of GFs in the management of partial-thickness burns. Electronic searches were conducted in PubMed and the Cochrane databases. Endpoint results analysed included wound healing and scar formation. Thirteen studies comprising a total of 1924 participants with 2130 wounds (1131 GF receiving patients versus 999 controls) were identified and included, evaluating the effect of fibroblast growth factor (FGF), epidermal growth factor (EGF) and granulocyte macrophage-colony stimulating factor (GM-CSF) on partial-thickness burns. Topical application of these agents significantly reduced healing time by 5·02 (95% confidence interval, 2·62 to 7·42), 3·12 (95% CI, 1·11 to 5·13) and 5·1 (95% CI, 4·02 to 6·18) days, respectively, compared with standard wound care alone. In addition, scar improvement following therapy with FGF and EGF was evident in terms of pigmentation, pliability, height and vascularity. No significant increase in adverse events was observed in patients receiving GFs. These results suggested that GF therapy could be an effective and safe add-on to standard wound care for partial-thickness burns. High-quality, adequately powered trials are needed to further confirm the conclusion. PMID:25040572

  12. Analysis of a basic core performance for FBR core nuclear design. 3

    International Nuclear Information System (INIS)

    The spatial distribution of reaction rates in the ZPPR-13A, having an axially heterogeneous core, has been analyzed. The ZPPR-13A core is treated as a 2-dimensional RZ configuration consisting of a homogeneous core. The analysis is performed by utilizing both probabilistic and deterministic treatments. The probabilistic treatment is performed with the Monte Carlo Code MVP running with continuous energy variable. By comparing the results obtained by both treatments and reviewing the calculation method of effective resonance cross sections, for deterministic treatment, utilized for the reaction rate distributions, it is revealed that the present treatment of effective resonance cross sections is not accurate, since there are observed effects due to dependence on energy group number or energy group width, and on anisotropic scattering. To utilize multi-band method for calculating effective resonance cross sections, widely used by the European researchers, the computer code GROUPIE is installed and the performance of the code is confirmed. Although, in order to improve effective resonance cross sections accuracy, the thermal neutron reactor standard code system SRAC-95 was introduced last year in which the ultra-fine group spectrum calculation module PEACO worked specially under the restriction that number of nuclei having resonance cross section, in any zone, should be less than three, because collision probabilities were obtained by an interpolation method. This year, the module is improved so that these collision probabilities are directly calculated, and by this improvement the highly accurate effective resonance cross sections below the energy of 40.868 keV can be calculated for whole geometrical configurations considered. To extend the application range of the module PEACO, the cross sections of sodium and structure material nuclei are prepared so that they are also represented as ultra-fine group cross sections. By such modifications of cross section library

  13. PWR core safety analysis with 3-dimensional methods

    International Nuclear Information System (INIS)

    Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

  14. Equilibrium Core Analysis of Two Types of Cores for the AHR

    International Nuclear Information System (INIS)

    A preliminary conceptual design of a research reactor core employing a rod type fuel assembly, has been performed. Two types of the reactor core configuration have been developed as a basic core of the Advanced HANARO Reactor (AHR). One is the aluminum block core whose coolant channels are made inside a hexagonal aluminum block for loading the fuel assemblies, the other is a flow tube core by using the same zircaloy flow tubes as the HANARO. These cores have four control rods of a shroud type. In the control rod sites, 18-element fuel assemblies of a circular type are loaded, whereas 36-element fuel assemblies of a hexagonal type are loaded in the hexagonal sites except the central site. The AHR is designed on the basis that the thermal power is 20MW and the maximum thermal neutron flux is about 4.0x1014 n/cm2/sec in the reflector region. For these two cores, MCNP calculations have been finished for the condition of loading fresh fuels and no irradiation holes and beam tubes in the reflector region. For the depleted core, the parameters such as the cycle length, fuel burnup and maximum linear power for the equilibrium core, are evaluated by using the HANARO fuel management code system

  15. Developing engineering design core competences through analysis of industrial products

    DEFF Research Database (Denmark)

    Hansen, Claus Thorp; Lenau, Torben Anker

    2011-01-01

    Most product development work carried out in industrial practice is characterised by being incremental, i.e. the industrial company has had a product in production and on the market for some time, and now time has come to design a new and upgraded variant. This type of redesign project requires...... challenge in staging a course module, in which students develop knowledge, understanding and skills, which will prepare them for being able to participate in and contribute to redesign projects in industrial practice. In the course module Product Analysis and Redesign that has run for 8 years we have...... that the engineering designers have core design competences to carry through an analysis of the existing product encompassing both a user-oriented side and a technical side, as well as to synthesise solution proposals for the new and upgraded product. The authors of this paper see an educational...

  16. A Retrospective Analysis of the Burn Injury Patients Records in the Emergency Department, an Epidemiologic Study

    Directory of Open Access Journals (Sweden)

    Nilgün Aksoy

    2014-08-01

    Full Text Available Introduction: Burns can be very destructive, and severely endanger the health and lives of humans. It maybe cause disability and even psychological trauma in individuals. . Such an event can also lead to economic burden on victim’s families and society. The aim of our study is to evaluate epidemiology and outcome of burn patients referring to emergency department. Methods: This is a cross-sectional study was conducted by evaluation of patients’ files and forensic reports of burned patients’ referred to the emergency department (ED of Akdeniz hospital, Turkey, 2008. Demographic data, the season, place, reason, anatomical sites, total body surface area, degrees, proceeding treatment, and admission time were recorded. Multinomial logistic regression was used to compare frequencies’ differences among single categorized variables. Stepwise logistic regression was applied to develop a predictive model for hospitalization. P<0.05 was defined as a significant level. Results: Two hundred thirty patients were enrolled (53.9% female. The mean of patients' ages was 25.3 ± 22.3 years. The most prevalence of burn were in the 0-6 age group and most of which was hot liquid scalding (71.3%. The most affected parts of the body were the left and right upper extremities. With increasing the severity of triage level (OR=2.2; 95% CI: 1.02-4.66; p=0.046, intentional burn (OR=4.7; 95% CI: 1.03-21.8; p=0.047, referring from other hospitals or clinics (OR=3.4; 95% CI: 1.7-6.6; p=0.001, and percentage of burn (OR=18.1; 95% CI: 5.42-62.6; p<0.001 were independent predictive factor for hospitalization. In addition, odds of hospitalization was lower in patients older than 15 years (OR=0.7; 95% CI: 0.5-0.91; p=0.035. Conclusion: This study revealed the most frequent burns are encountered in the age group of 0-6 years, percentage of <10%, second degree, upper extremities, indoor, and scalding from hot liquids. Increasing ESI severity, intentional burn, referring from

  17. Analysis and Assessment of the Spatial and Temporal Distribution of Burned Areas in the Amazon Forest

    Directory of Open Access Journals (Sweden)

    Francielle da Silva Cardozo

    2014-08-01

    Full Text Available The objective of this study was to analyze the spatial and temporal distribution of burned areas in Rondônia State, Brazil during the years 2000 to 2011 and evaluate the burned area maps. A Linear Spectral Mixture Model (LSMM was applied to MODIS surface reflectance images to originate the burned areas maps, which were validated with TM/Landsat 5 and ETM+/Landsat 7 images and field data acquired in August 2013. The validation presented a correlation ranging from 67% to 96% with an average value of 86%. The lower correlation values are related to the distinct spatial resolutions of the MODIS and TM/ETM+ sensors because small burn scars are not detected in MODIS images and higher spatial correlations are related to the presence of large fires, which are better identified in MODIS, increasing the accuracy of the mapping methodology. In addition, the 12-year burned area maps of Rondônia indicate that fires, as a general pattern, occur in areas that have already been converted to some land use, such as vegetal extraction, large animal livestock areas or diversified permanent crops. Furthermore, during the analyzed period, land use conversion associated with climatic events significantly influenced the occurrence of fire in Rondônia and amplified its impacts.

  18. A three decade analysis of factors affecting burn mortality in the elderly.

    Science.gov (United States)

    Lionelli, G T; Pickus, E J; Beckum, O K; Decoursey, R L; Korentager, R A

    2005-12-01

    This study's objective was to identify variables that affect the mortality of elderly burn patients and to assess their changes over time. A retrospective review was conducted on all patients 75 or older (n=201) admitted to a university-based burn center between 1972 and 2000. Variables examined were age, sex, TBSA, ABSI, inhalation injury, timing from burn to operative intervention, the number of surgical procedures, the number of pre-morbid conditions, and mortality. There were 95 fatalities. TBSA strongly correlated with mortality (p<0.0001). Adjusting for TBSA and inhalation injury, mortality significantly decreased (p=0.04, odds ratio=0.58). Mortality significantly increased with inhalation injury (p<0.01). Fatality risk increased by 400% with inhalation injury. Absence of inhalation injury was not significant with respect to mortality in the 1970s, however there was a significant decrease (p=0.02) in mortality without an inhalation injury in the 1980s and 1990s. ABSI was strongly predictive of mortality (p<0.0001). On average there was a 200% increase in mortality per unit increase of ABSI. The elderly are 58% less likely to die from burns now as compared to the 1970s. Although mortality rose with increasing TBSA equally in each decade, the absolute risk of mortality decreased over time. This data suggests major strides have been made in burn care, however similar success has not been achieved with inhalation injuries. PMID:16269217

  19. Global stability analysis of pressurized water reactor core nonlinear system

    International Nuclear Information System (INIS)

    Determining the global stability of a pressurized water reactor (PWR) core nonlinear system is the problem to be solved. In the paper, the core nonlinear system was modeled and the linearized model of the system was obtained via the small perturbation method. According to the distributing situation of the core nonlinearity measure in the power level range based on the equilibrium manifold, seven linear models corresponding to seven power levels respectively were chosen as local models of the core and the set of seven local models was used to approximately substitute the core system. The global stability of the PWR core nonlinear system was analyzed by utilizing Lyapunov stability theory. The calculated result shows that the core nonlinear system is globally and asymptotically stable. The modeling method of the core is effective in analyzing the global stability of a PWR core nonlinear system. (authors)

  20. Statistical analysis of dynamic parameters of the core

    International Nuclear Information System (INIS)

    The transients of various types were investigated for the cores of zero power critical facilities in RRC KI and NPP. Dynamic parameters of neutron transients were explored by tool statistical analysis. Its have sufficient duration, few channels for currents of chambers and reactivity and also some channels for technological parameters. On these values the inverse period. reactivity, lifetime of neutrons, reactivity coefficients and some effects of a reactivity are determinate, and on the values were restored values of measured dynamic parameters as result of the analysis. The mathematical means of statistical analysis were used: approximation(A), filtration (F), rejection (R), estimation of parameters of descriptive statistic (DSP), correlation performances (kk), regression analysis(KP), the prognosis (P), statistician criteria (SC). The calculation procedures were realized by computer language MATLAB. The reasons of methodical and statistical errors are submitted: inadequacy of model operation, precision neutron-physical parameters, features of registered processes, used mathematical model in reactivity meters, technique of processing for registered data etc. Examples of results of statistical analysis. Problems of validity of the methods used for definition and certification of values of statistical parameters and dynamic characteristics are considered (Authors)

  1. Neutronics analysis of the current core of the TRIGA Mark II reactor Vienna

    International Nuclear Information System (INIS)

    This paper presents the part of PhD work performed at the TRIGA Mark II Vienna. A detailed three dimensional MCNP model of the reactor was developed. The neutronics library JEFF3.1 was applied to this model. The model was completed by employing the fresh fuel composition experiments and was confirmed by the initial criticality, reactivity distribution and thermal flux distribution performed in 1962. To analyse the current burned core, burn up and its relevant material composition was calculated by ORIGEN2 and confirmed by gamma spectroscopy of six spent Fuel Elements FE(s). This new material composition of the current core was incorporated into the already developed MCNP model. This paper presents the current core calculations employing MCNP5 and its experimental validation through criticality and reactivity distribution experiments, performed at the TRIGA Mark II research reactor Vienna. The MCNP predicts the criticality of the current core on loading of 78th FE in the core which is also confirmed experimentally. Five FE(s) were calculated and measured for their reactivity worths. The deviations between theoretical results and experimental observations were in range from 3% to 17%. (author)

  2. A Burnup Analysis of PBMR-400MWth Reactor Core

    International Nuclear Information System (INIS)

    The purpose of this study is to analyze the burnup characteristics of 400MWth PBMR using Monte Carlo method. In the world, the deterministic method is widely used to model such that system but it still has a disadvantage which is not flexible in simulating the burnup cycle. Although this method applies some techniques to increase the accuracy of calculation results but it is necessary to model this system by a suitable computer code that can verify and validate the results of the deterministic method. A method which uses a Monte Carlo technique for simulating the burnup cycle was performed. A reactor physics computer code uses in this method is MONTEBURN 2.0 which accurately and efficiently computes the neutronic and material properties of the fuel cycle. MONTEBURN is a fully automated tool that links the MCNP Monte Carlo transport code with a radioactive decay and burnup code ORIGEN. In this model, the calculations are based on a detailed core modeling using MCNP. The fuel pebble is thoroughly modeled by introducing unit cell modeling for the graphite matrix and fuel kernels in the pebble. For the burnup model, a start-up core was studied with considering the movement of pebbles. By shifting down one layer at each discrete time step and inserting fresh fuel from the top, this cyclic calculation is continued until equilibrium burnup cycle is achieved. In this study, the time dependence of multiplication factor keff, the spatial dependence of flux profile, power distribution, burnup, and inventory of isotopes in the start up process are analyzed. The results will provide the basis data of the burnup process and be also utilized as the verified data to validate a compute code for PBMR core analysis which will be developed in near future

  3. Analysis of Ignition and Combustion in Otto Lean-Burn Engines with Prechambers

    OpenAIRE

    Norum, Viggo Lauritz

    2008-01-01

    Otto-engines in which the combustion chamber has richer fuel/air mix close to the ignition source and leaner charge further away from the ignition source are often called "stratified charge engines". Stratified charge can be used to increase the combustion speed in an internal combustion engine and thereby enable the engine to run on a fuel/air mix that would normally burn too slowly or not burn at all. The use of prechambers is one way to obtain stratified charge.This thesis presents and use...

  4. Application of Network Analysis Method to VHTR core

    International Nuclear Information System (INIS)

    A Very High Temperature Reactor (VHTR) is currently envisioned as a promising future reactor concept because of its high-efficiency and capability of generating hydrogen. Prismatic Modular Reactor (PMR) is one of the main VHTR concepts, which consists of hexagonal prismatic fuel blocks and reflector blocks made of nuclear grade graphite. However their shape could be changed by neutron damage during the reactor operation and the shape change can makes the gaps between the blocks inducing bypass flow. Most of reactor coolant flows through the coolant channel within the fuel block, but some portion of the reactor coolant bypasses to the interstitial gaps. The vertical gap and horizontal gap are called bypass gap and cross gap, respectively. CFD simulation for the full core of VHTR might be possible but it requires vast computational cost and time. Therefore, fast, flexible and reliable code is required to predict the flow distribution corresponding to the various bypass gap distribution. Consequently in this study, the flow network analysis method is applied to analyze the core flow of VHTR. The applied method was validated by comparing with SNU VHTR multiblock experiment. As a result, the calculated results show good agreements with experimental data although computational time and cost of the developed code was very small

  5. Stress analysis and assessment of RPV core support pads and surround bottom head

    International Nuclear Information System (INIS)

    Background: The core support pads were used to limit circumferential rotation of core barrel, and the structure integrity of the core support pads was an important factor to the safe operation of nuclear power plant. Purpose: To ensure the structure integrity of the core support pads. Methods: Three-dimensional FEA model for bottom head, core support pads and part cylinder of CAP1000 RPV was established. Thermal analysis, static analysis, fatigue analysis and fracture analysis were performed. The analysis results were evaluated according to ASME B and PVC-III-NB-3200 and ASME B and PVC-III-1-Appendix G. Results: The evaluation indicated that the core support pads and surround bottom head could satisfy related requirements of above code. Conclusions: The analysis methodology used in this paper could also be applied to the core support pads of RPV for above 1000 MW nuclear power plant. (authors)

  6. Time-resolved characterization and energy balance analysis of implosion core in shock-ignition experiments at OMEGA

    International Nuclear Information System (INIS)

    Time-resolved temperature and density conditions in the core of shock-ignition implosions have been determined for the first time. The diagnostic method relies on the observation, with a streaked crystal spectrometer, of the signature of an Ar tracer added to the deuterium gas fill. The data analysis confirms the importance of the shell attenuation effect previously noted on time-integrated spectroscopic measurements of thick-wall targets [R. Florido et al., Phys. Rev. E 83, 066408 (2011)]. This effect must be taken into account in order to obtain reliable results. The extracted temperature and density time-histories are representative of the state of the core during the implosion deceleration and burning phases. As a consequence of the ignitor shock launched by the sharp intensity spike at the end of the laser pulse, observed average core electron temperature and mass density reach T ∼ 1100 eV and ρ ∼ 2 g/cm3; then temperature drops to T ∼ 920 eV while density rises to ρ ∼ 3.4 g/cm3 about the time of peak compression. Compared to 1D hydrodynamic simulations, the experiment shows similar maximum temperatures and smaller densities. Simulations do not reproduce all observations. Differences are noted in the heating dynamics driven by the ignitor shock and the optical depth time-history of the compressed shell. Time-histories of core conditions extracted from spectroscopy show that the implosion can be interpreted as a two-stage polytropic process. Furthermore, an energy balance analysis of implosion core suggests an increase in total energy greater than what 1D hydrodynamic simulations predict. This new methodology can be implemented in other ICF experiments to look into implosion dynamics and help to understand the underlying physics

  7. Safety analysis of JMTR LEU fuel core, (3)

    International Nuclear Information System (INIS)

    Dose analysis in the safety evaluation and the site evaluation were performed for the JMTR core conversion from MEU fuel to LEU fuel. In the safety evaluation, the effective dose equivalents for the public surrounding the site were estimated in fuel handling accident and flow blockage to coolant channel which were selected as the design basis accidents with release of radioactive fission products to the environment. In the site evaluation, the flow blockage to coolant channel was selected as siting basis events, since this accident had the possibility of spreading radioactive release. Maximum exposure doses for the public were estimated assuming large amounts of fission products to release. It was confirmed that risk of radiation exposure of the public is negligible and the siting is appropriate. (author)

  8. Burning Issue: Handling Household Burns

    Science.gov (United States)

    ... take steps to avoid household burns. Never leave cooking food unattended on the stove. Set your water heater’s thermostat to 120 °F or lower to prevent scalding burns. And install smoke alarms on every floor of your home. Keep yourself and your family safe from unexpected ...

  9. Availability analysis of the AP600 passive core cooling system

    International Nuclear Information System (INIS)

    The reliability analysis of the AP600 Passive Core Cooling System (PXS) has been done. The fault tree analysis method was used for the quantitative analysis. The PXS can be grouped to several sub-systems i.e.: Reactor Coolant System (RCS) Injection Subsystem, Emergency Core Decay Heat Removal Subsystem, and Containment Sump pH Control Subsystem. The quantitative analysis results indicates that the system unavailability is highly dependent on the valves configuration of the Automatic Depressurization System (ADS). If the ADS valves is arranged in Option-1, the system unavailability is 2.347E-03, this means that the yearly contribution to plant down time can be estimated to be about 20.56 hours per year. Whereas, by using Option-2 of fourth stage ADS valves, the system unavailability is reduced to be 9.877E-04 or 8.65 hours per year and this value is consistent with the allocated goal value (8.0 hours per year). The ADS contributes 66.89% to the system unavailability if it is arranged in Option-1, and will reduced to be about 21.21% if its fourth stages are arranged in Option-2. If the ADS is not included as a subsystem of the PXS (relocate to RCS as a subsystem of RCS), then the PXS unavailability will be reduced to about 7.784E-04 or 6.82 hours per year and this is less then the allocated goal value. The major contributors to the system unavailability are mostly dominated by Stage-4 ADS valves (air piston operated valves and squib valves), inservice testing valves of ADS (solenoid operated valves), solenoid valves of Nitrogen Supply to Accumulator, and Passive Residual Heat Removal actuation valves (air operated valves). Therefore, it is recommended that those valves be analyzed more detail to gain the improvement in its reliability. It is also recommended that the fourth stage of ADS valves should be arranged according to Option-2, i.e. one 10-inch normally open motor operated gate valve in series with one 10-inch normally closed squib valve. (author). 13 refs, 3

  10. Characterization of Pseudomonas aeruginosa in Burn Patients Using PCR- Restriction Fragment Length Polymorphism and Random Amplified Polymorphic DNA Analysis

    Directory of Open Access Journals (Sweden)

    Abdolaziz Rastegar Lari

    2010-09-01

    Full Text Available One of the major opportunistic pathogens in patients with burninjuries is Pseudomonas aeruginosa, which causes severe infectionsin burned patients. The objective of the study was to examinethe molecular epidemiology of P. aeruginosa colonization inthe burn unit of Shahid Motahari Hospital in Tehran, Iran. Restrictionfragment length polymorphism (RFLP and random amplifiedpolymorphic DNA (RAPD analysis were employed tostudy 127 clinical and two environmental P. aeruginosa isolatescollected from January to June 2008. In RFLP, the PCR productsof 16S rRNA gene were digested with restriction enzyme Alu I,Hae III, and Rsa I, and the fragments generated were analyzed byagarose electrophoresis. Molecular typing by RFLP did show nodiscriminatory power for P. aeruginosa isolates, but RAPD-PCRrevealed eight different genotypes; RAPD1to RAPD8 in clinicaland environmental isolates. RAPD1 was the major genotype inclinical (n=64, 50.4% and environmental isolates (n=1, 50%.The findings suggest that RAPD might have a superior typeabilityand discriminatory power over RFLP to study P. aeruginusa.Moreover, they highlight the need for further attention to the controlof infection sources in Burn Units to prevent the transmissionof the bacterium.

  11. Numerical analysis of free-burning argon arcs based on the local thermodynamic equilibrium model at various electrical currents

    International Nuclear Information System (INIS)

    Free-burning arcs where the work piece acts as an anode were numerically analyzed using a computational domain including the arc itself and its anode region based on the local thermodynamic equilibrium model. Because the major arc parameters such as temperature, axial velocity, electric potential difference and pressure-rise from ambient atmospheric pressure are much dependent on the working current, our investigation was concerned with developing a capability to model free-burning argon arcs and considering the energy flux going into the anode at various values of the electrical current (I = 50, 100 and 200 A) by computational fluid dynamics analysis. Predicted temperatures along the z-axis between the electrodes were in fair agreement with existing experimental results. Particularly, reasonable relationships between the maximum velocity or temperature and the applied current were predicted, which matched well with other theoretical results. In addition, some discrepancies with other predictions were shown in the results about electric potential and pressure-rise. It should be related to the omission of the space-charge effect near the electrodes for a simplified unified model and the application of a turbulence model for the steep temperature gradient at the arc edges. - Highlights: • Free-burning argon arcs were investigated at various working currents numerically. • The relationships between the current, the velocity and the temperature were found. • Some discrepancies were shown in the results of pressure and electric potential. • Those should be supplemented by the non-equilibrium situation near electrodes

  12. Uncertainly propagation analysis for Yonggwang nuclear unit 4 by McCARD/MASTER core analysis system

    International Nuclear Information System (INIS)

    This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor (keff), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

  13. Burning Mouth Syndrome

    Science.gov (United States)

    ... OralHealth > Topics > Burning Mouth Syndrome > Burning Mouth Syndrome Burning Mouth Syndrome Main Content Key Points Symptoms Diagnosis Primary and Secondary BMS Treatment Helpful Tips Key Points Burning mouth syndrome is burning pain in the mouth that may ...

  14. Burn-out analysis. Part I: A burn-out correlation for water in round and rectangular channels uniformly heated at various pressures with forced convection

    Energy Technology Data Exchange (ETDEWEB)

    Macbeth, R.V. [Reactor Development Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1961-11-15

    The practical objective of research on 'burn-out' is a reliable method giving the maximum safe rating for any water cooled reactor. Experimental work, which began at numerous centres about 10 years ago, has been concerned principally with endeavouring to understand the phenomenon as it applies to simple geometries such as round and rectangular channels. Many millions of pounds have been spent on this work and several thousand separate experimental results obtained. This considerable effort has achieved little real success in providing an explanation of 'burn-out' however. Many conflicting views have arisen and correlations so far developed have been shown to give calculated 'burn-out' heat fluxes varying by a factor of the order of 5> when applied to a typical reactor situation. While some uncertainty may be due to experimental variations, inadequate analytical effort is considered to be the primary cause of the present confused situation. To overcome this various analytical studies are being initiated by the Reactor Development Division at Winfrith and a detailed plan is being evolved for bringing effort to bear on certain fundamental aspects of boiling which have been neglected and which in some oases will require the development of special experimental techniques. This report describes the result of some work already carried out. It concerns an initial examination made on a large group of 'burn-out' data and describes the development of a correlation which predicts 'burn-out' heat fluxes to within an R.M.S. error of less than 10% over a very wide range of operating conditions including pressure. (author)

  15. Analysis of xRAGE and flag high explosive burn models with PBX 9404 cylinder tests

    Energy Technology Data Exchange (ETDEWEB)

    Harrier, Danielle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Andersen, Kyle Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-08-05

    High explosives are energetic materials that release their chemical energy in a short interval of time. They are able to generate extreme heat and pressure by a shock driven chemical decomposition reaction, which makes them valuable tools that must be understood. This study investigated the accuracy and performance of two Los Alamos National Laboratory hydrodynamic codes, which are used to determine the behavior of explosives within a variety of systems: xRAGE which utilizes an Eulerian mesh, and FLAG with utilizes a Lagrangian mesh. Various programmed and reactive burn models within both codes were tested using a copper cylinder expansion test. The test was based on a recent experimental setup which contained the plastic bonded explosive PBX 9404. Detonation velocity versus time curves for this explosive were obtained using Photon Doppler Velocimetry (PDV). The modeled results from each of the burn models tested were then compared to one another and to the experimental results. This study validate

  16. Evaluation and Parameter Analysis of Burn up Calculations for the Assessment of Radioactive Waste - 13187

    International Nuclear Information System (INIS)

    Burn up calculations facilitate a determination of the composition and nuclear inventory of spent nuclear fuel, if operational history is known. In case this information is not available, the total nuclear inventory can be determined by means of destructive or, even on industrial scale, nondestructive measurement methods. For non-destructive measurements however only a few easy-to-measure, so-called key nuclides, are determined due to their characteristic gamma lines or neutron emission. From these measured activities the fuel burn up and cooling time are derived to facilitate the numerical inventory determination of spent fuel elements. Most regulatory bodies require an independent assessment of nuclear waste properties and their documentation. Prominent part of this assessment is a consistency check of inventory declaration. The waste packages often contain wastes from different types of spent fuels of different history and information about the secondary reactor parameters may not be available. In this case the so-called characteristic fuel burn up and cooling time are determined. These values are obtained from a correlations involving key-nuclides with a certain bandwidth, thus with upper and lower limits. The bandwidth is strongly dependent on secondary reactor parameter such as initial enrichment, temperature and density of the fuel and moderator, hence the reactor type, fuel element geometry and plant operation history. The purpose of our investigation is to look into the scaling and correlation limitations, to define and verify the range of validity and to scrutinize the dependencies and propagation of uncertainties that affect the waste inventory declarations and their independent verification. This is accomplished by numerical assessment and simulation of waste production using well accepted codes SCALE 6.0 and 6.1 to simulate the cooling time and burn up of a spent fuel element. The simulations are benchmarked against spent fuel from the real reactor

  17. BN-600 full MOX core benchmark analysis (PHYSOR 2004 paper)

    International Nuclear Information System (INIS)

    As a follow-up of the BN-600 hybrid core benchmark, a full MOX core benchmark was performed within the framework of the IAEA co-ordinated research project Discrepancies between the values of main reactivity coefficients obtained by the participants for the BN-600 full MOX core benchmark appear to be larger than those in the previous hybrid core benchmarks on traditional core configurations. This arises due to uncertainties in the proper modelling of the axial sodium plenum above the core. It was recognized that the sodium density coefficient strongly depends on the core model configuration of interest (hybrid core vs. fully MOX fuelled core with sodium plenum above the core) in conjunction with the calculation method (diffusion vs. transport theory). The effects of the discrepancies revealed between the participants' results on the ULOF and UTOP transient behaviours of the BN-600 full MOX core were investigated in simplified transient analyses. Generally the diffusion approximation predicts more benign consequences for the ULOF accident but more hazardous ones for the UTOP accident when compared with the transport theory results. The heterogeneity effect does not have any significant effect on the simulation of the transient The comparison of the transient analyses results concluded that the fuel Doppler coefficient and the sodium density coefficient are the two most important coefficients in understanding the ULOF transient behaviour. In particular, the uncertainty in evaluating the sodium density coefficient distribution has the largest impact on the description of reactor dynamics. This is because the maximum sodium temperature rise takes place at the top of the core and in the sodium plenum

  18. Overview of core simulation methodologies for light water reactor analysis

    International Nuclear Information System (INIS)

    The current in-core fuel management calculation methods provide a very efficient route to predict neutronics behavior of light water reactor (LWR) cores and their prediction accuracy for current generation LWRs is generally sufficient. However, since neutronics calculations for LWRs are based on various assumptions and simplifications, we should also recognize many implicit limitations that are 'embedded' in current neutronics calculation methodologies. Continuous effort for improvement of core simulation methodologies is also discussed. (author)

  19. Technical benefit and risk analysis on cement clinkering process with compact internal burning of carbon

    International Nuclear Information System (INIS)

    This article demonstrates the potential technical benefit and risk for cement clinkering process with compact internal burning of carbon, a laboratory-phase developing technique, from 9 aspects, including the heat consumption of clinkering and exhaust heat utilization, clinker quality, adaptability to alternative fuels, the disposal ability of industrial offal and civil garbage, adaptability to the raw materials and fuels with high content of chlorine, sulphur and alkali, the feasibility of process scale up, the briquetting process of the coal-containing cement raw meal pellet, NOx emission and the capital cost and benefit of conversion project. It is concluded that it will be able to replace the modern precalciner rotary kiln process and to become the main stream technique of cement clinkering process in low carbon economy times. - Highlights: • Compact internal burning of carbon enables cement shaft kiln to run stably. • Compact internal burning of carbon enables cement shaft kiln to scale up. • New process triples energy efficiency with excellent environmental performance. • It will be able to compete with and replace the existing precalciner kiln process. • It will become the mainstream clinkering process in low carbon economy

  20. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

  1. Evolution of Massive Stars Up to the End of Central Oxygen Burning

    OpenAIRE

    Eid, Mounib F. El; Meyer, Bradley S.; The, Lih-Sin

    2004-01-01

    We present a detailed study of the evolution of massive stars of masses 15, 20, 25 and 30 $\\msun$ assuming solar-like initial chemical composition. The stellar sequences were evolved through the advanced burning phases up to the end of core oxygen burning. We present a careful analysis of the physical characteristics of the stellar models. In particular, we investigate the effect of the still unsettled reaction $^{12}$C($\\alpha$,$\\gamma$)$^{16}$O on the advanced evolution by using recent comp...

  2. Non destructive determination of fuel burn up of the RA reactor at Vinca by gamma radiation spectra analysis

    International Nuclear Information System (INIS)

    The problem of non destructive determination of the burn-up of used up fuel of the powerful experimental reactor RA at Vinca, by the analysis of gamma radiation spectra using a gamma-spectrometer with a semiconductor Ge(Li) detector has been studied. The first part of this analytical problem is concerned with calculation of fuel burn up. In the preparation of its solution material and energetic balance of the fuel composition and conditions of fuel irradiation. The obtained solution uses numerical methods. In this solution fuel burn up is determined: 1) on the basis of the data on the composition of 106Ru, 134Cs and 137Cs, than 2) from a series of the data on the fuel and reactor and 3) on the basis of those, numerous, literature data which participate in defining of the balance of fuel burn up process. The second part of this problem refers to determination of the composition of the above gamma radioactive fission products from the obtained instrumental spectrum. Under strictly defined conditions of the measurement of gamma radiation from the fuel elements, when a determined type of the gamma ray collimator is used, the photo peak area of the corresponding line in the instrumental spectrum is defined as a function of the fission products activity, energy and yield of its gamma rays, the thickness of fuel and added absorbers as well as the dimensions of the collimator used. On this basis, the activity quotient of two fission products is a function of: 1) two areas of two photo peaks of their two lines and 2) some of the above cited values. Unknown magnitudes of the remaining values of the same relation are determined from the photo peak areas of the lines of the complex spectrum of one fission product by solving a system of equations. Accuracy of the solutions of these two separate parts of the observed analytical problems was confirmed experimentally. The results obtained are characterized by high repeatability. Total errors are generally greater, primarily due to

  3. Self-Healing Many-Core Architecture: Analysis and Evaluation

    Directory of Open Access Journals (Sweden)

    Arezoo Kamran

    2016-01-01

    Full Text Available More pronounced aging effects, more frequent early-life failures, and incomplete testing and verification processes due to time-to-market pressure in new fabrication technologies impose reliability challenges on forthcoming systems. A promising solution to these reliability challenges is self-test and self-reconfiguration with no or limited external control. In this work a scalable self-test mechanism for periodic online testing of many-core processor has been proposed. This test mechanism facilitates autonomous detection and omission of faulty cores and makes graceful degradation of the many-core architecture possible. Several test components are incorporated in the many-core architecture that distribute test stimuli, suspend normal operation of individual processing cores, apply test, and detect faulty cores. Test is performed concurrently with the system normal operation without any noticeable downtime at the application level. Experimental results show that the proposed test architecture is extensively scalable in terms of hardware overhead and performance overhead that makes it applicable to many-cores with more than a thousand processing cores.

  4. Real time analysis of light water core neutronics

    International Nuclear Information System (INIS)

    A method is described for determining the neutronics parameters of a reactor core comprising the steps of: representing the reactor core as a plurality of nodes having a coarse nodal representation; monitoring selected neutronic parameters of the reactor core; providing time-dependent two group neutron diffusion equations coupled to delayed neutron precursor concentrations that have been subjected to space-time factorization by shape and amplitude functions in response to the plurality of nodes; sensing the monitored parameters; and determining the core neutronics parameters in response to the sensed parameters and the provided two group neutron diffusion equations in constant time steps for sensing the monitored parameters and determining the core neutronics parameters in a real-time environment, the time steps being not less than one quarter second

  5. Unsteady thermal analysis of gas-cooled fast reactor core

    International Nuclear Information System (INIS)

    This thesis presents numerical analysis of transient heat transfer in an equivalent coolant-fuel rod cell of a typical gas cooled, fast nuclear reactor core. The transient performance is assumed to follow a complete sudden loss of coolant starting from steady state operation. Steady state conditions are obtained from solving a conduction problem in the fuel rod and a parabolic turbutent convection problem in the coolant section. The coupling between the two problems is accomplished by ensuring continuity of the thermal conditions at the interface between the fuel rod and the coolant. to model turbulence, the mixing tenght theory is used. Various fuel rod configurations have been tested for optimal transient performance. Actually, the loss of coolant accident occurs gradually at an exponential rate. Moreover, a time delay before shutting down the reactor by insertion of control rods usually exists. It is required to minimize maximum steady state cladding temperature so that the time required to reach its limiting value during transient state is maximum. This will prevent the escape of radioactive gases that endanger the environment and the public. However, the case considered here is a limiting case representing what could actually happen in the worst probable accident. So, the resutls in this thesis are very indicative regarding selection of the fuel rode configuration for better transient performance in case of accidents in which complete loss of collant occurs instantaneously

  6. Homogeneous protein analysis by magnetic core-shell nanorod probes

    KAUST Repository

    Schrittwieser, Stefan

    2016-03-29

    Studying protein interactions is of vital importance both to fundamental biology research and to medical applications. Here, we report on the experimental proof of a universally applicable label-free homogeneous platform for rapid protein analysis. It is based on optically detecting changes in the rotational dynamics of magnetically agitated core-shell nanorods upon their specific interaction with proteins. By adjusting the excitation frequency, we are able to optimize the measurement signal for each analyte protein size. In addition, due to the locking of the optical signal to the magnetic excitation frequency, background signals are suppressed, thus allowing exclusive studies of processes at the nanoprobe surface only. We study target proteins (soluble domain of the human epidermal growth factor receptor 2 - sHER2) specifically binding to antibodies (trastuzumab) immobilized on the surface of our nanoprobes and demonstrate direct deduction of their respective sizes. Additionally, we examine the dependence of our measurement signal on the concentration of the analyte protein, and deduce a minimally detectable sHER2 concentration of 440 pM. For our homogeneous measurement platform, good dispersion stability of the applied nanoprobes under physiological conditions is of vital importance. To that end, we support our measurement data by theoretical modeling of the total particle-particle interaction energies. The successful implementation of our platform offers scope for applications in biomarker-based diagnostics as well as for answering basic biology questions.

  7. 100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT

    Energy Technology Data Exchange (ETDEWEB)

    HARRINGTON RA

    2010-01-15

    On December 15-16, 2009, a 100-KE Reactor Core Removal Project Alternative Analysis Workshop was conducted at the Washington State University Consolidated Information Center, Room 214. Colburn Kennedy, Project Director, CH2M HILL Plateau Remediation Company (CHPRC) requested the workshop and Richard Harrington provided facilitation. The purpose of the session was to select the preferred Bio Shield Alternative, for integration with the Thermal Shield and Core Removal and develop the path forward to proceed with project delivery. Prior to this workshop, the S.A. Robotics (SAR) Obstruction Removal Alternatives Analysis (565-DLV-062) report was issued, for use prior to and throughout the session, to all the team members. The multidisciplinary team consisted ofrepresentatives from 100-KE Project Management, Engineering, Radcon, Nuclear Safety, Fire Protection, Crane/Rigging, SAR Project Engineering, the Department of Energy Richland Field Office, Environmental Protection Agency, Washington State Department of Ecology, Defense Nuclear Facility Safety Board, and Deactivation and Decommission subject matter experts from corporate CH2M HILL and Lucas. Appendix D contains the workshop agenda, guidelines and expectations, opening remarks, and attendance roster going into followed throughout the workshop. The team was successful in selecting the preferred alternative and developing an eight-point path forward action plan to proceed with conceptual design. Conventional Demolition was selected as the preferred alternative over two other alternatives: Diamond Wire with Options, and Harmonic Delamination with Conventional Demolition. The teams preferred alternative aligned with the SAR Obstruction Removal Alternative Analysis report conclusion. However, the team identified several Path Forward actions, in Appendix A, which upon completion will solidify and potentially enhance the Conventional Demolition alternative with multiple options and approaches to achieve project delivery

  8. Aspects of cell calculations in deterministic reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Varvayanni, M. [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Savva, P., E-mail: savvapan@ipta.demokritos.gr [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N. [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    {Tau}he capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available

  9. Simulation of Core Support Barrel Vibration Monitoring Using Ex-Core Neutron Noise Analysis and Fuzzy Logic Algorithm

    International Nuclear Information System (INIS)

    The application of ex-core Neutron Noise Analysis (NNA) to monitor the vibration characteristics of a reactor Core Support Barrel (CSB) was studied. Ex-core flux data was obtained using a non-analog Monte Carlo neutron transport method in a simulated CSB model. The implicit capture and Russian Roulette technique was optimized through a sensitivity study to simulate the neutron transport. A combination of two-dimensional and three-dimensional beam and shell mode vibration of CSB was modelled. Parallel processing was employed to reduce the simulation time. An NNA module was developed to analyze the ex-core flux data based on its time variation, Normalized Power Spectral Density (NPSD), Normalized Cross-Power Spectral Density (NCPSD), Coherence and phase differences. The data was then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core flux signal fluctuation was directly proportional to the CSB's vibration observed at 8 and 15 Hz in the beam mode vibration, and at 8 Hz in the shell mode vibration. The Coherence result between flux pairs was unity at the vibration peak frequencies. A set of out-of-phase and in-phase unique pattern of phase differences was observed for each of the vibration models. The fuzzy logic module successfully recognized the correct vibration frequencies, modes, orders, directions, and phase differences within 4.1 milliseconds for the three-dimensional beam and shell mode vibrations

  10. Simulation of Core Support Barrel Vibration Monitoring Using Ex-Core Neutron Noise Analysis and Fuzzy Logic Algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Christian, Robby; Song, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Seon, Ho [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-08-15

    The application of ex-core Neutron Noise Analysis (NNA) to monitor the vibration characteristics of a reactor Core Support Barrel (CSB) was studied. Ex-core flux data was obtained using a non-analog Monte Carlo neutron transport method in a simulated CSB model. The implicit capture and Russian Roulette technique was optimized through a sensitivity study to simulate the neutron transport. A combination of two-dimensional and three-dimensional beam and shell mode vibration of CSB was modelled. Parallel processing was employed to reduce the simulation time. An NNA module was developed to analyze the ex-core flux data based on its time variation, Normalized Power Spectral Density (NPSD), Normalized Cross-Power Spectral Density (NCPSD), Coherence and phase differences. The data was then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core flux signal fluctuation was directly proportional to the CSB's vibration observed at 8 and 15 Hz in the beam mode vibration, and at 8 Hz in the shell mode vibration. The Coherence result between flux pairs was unity at the vibration peak frequencies. A set of out-of-phase and in-phase unique pattern of phase differences was observed for each of the vibration models. The fuzzy logic module successfully recognized the correct vibration frequencies, modes, orders, directions, and phase differences within 4.1 milliseconds for the three-dimensional beam and shell mode vibrations.

  11. Analysis of the seismic response of a fast reactor core

    International Nuclear Information System (INIS)

    This report deals with the methods to apply for a correct evaluation of the reactor core seismic response. Reference is made to up-to-date design data concerning the PEC core, taking into account the presence of the core-restraint plate located close to the PEC core elements top and applying the optimized iterative procedure between the vessel linear calculation and the non-linear ones limited to the core, which had been described in a previous report. It is demonstrated that the convergence of this procedure is very fast, similar to what obtained in the calculations of the cited report, carried out with preliminary data, and it is shown that the cited methods allow a reliable evaluation of the excitation time histories for the experimental tests in support of the seismic verification of the shutdown system and the core of a fast reactor, as well as relevant data for the experimental, structural and functional, verification of the core elements in the case of seismic loads

  12. HEU-LEU mixed core analysis for TR-2

    International Nuclear Information System (INIS)

    Core conversion calculation have been carried out for different core loadings of the TR-2 reactor in order to find out the optimum design for radioisotope production. Using HEU and LEU fuel elements in the mixed core also introduced additional peaking problems to be eliminated. Five group structure is used for the burnup dependent cross-section libraries that are generated by EPRI-CELL code. 20 diffusion-depletion code GEREBUS is used for the reactivity and burnup calculations. New graphite reflectors have been added to the periphery of the core to enhance the reactivity and the discharge burnup levels. Two water boxes have been placed inside reactor core in order to increase the radioisotope production. The activity levels of the irradiation samples, core excess reactivities, power peaking factors, and the anti-reactivities of the control blades have been calculated for various loadings. After the optimization studies, it is found that these modifications have been yielded higher production rates and an uniform distribution in the activity levels of the irradiation samples. One irradiation and two standard LEU fuel elements have already been loaded to the TR-2 core without any operational or safety related problems.The agreement between the calculation and the experiments are quite good for the operated 13 cycles

  13. Sugar Cane Burning and Human Health: An Analysis Using Spatial Propensity Score Matching

    OpenAIRE

    Chagas, Andre; Almeida, Alex; Azzoni, Carlos

    2013-01-01

    The production of ethanol and sugar from sugar cane has sharply increased for the last 20 years. If there are overall incentives to substitute the consumption of fossil fuels by biofuels, the increase of production and the expansion of new cultivated areas of sugar cane have eventually an impact on human health and employment mainly at regional levels. To harvest the crop--mostly manually done by low-skill workers--the practice of burning to clean dry grasses and poisonous insects has been ex...

  14. Study on burn-up credit and minor actinide in post-irradiation analysis

    International Nuclear Information System (INIS)

    Accuracy of burnup calculation for actinide is very important as to the study of burn-up credit. For minor-actinides such as Am243 and Cm244, however, typical burnup calculation codes are not accurate enough. The accuracy for both nuclides was studied by using the SWAT code. The study showed that the C/E values of both nuclides could be improved at the same time by changing the cross section of Pu242. A study of burnup calculation related to the cross section of Pu242 should be performed to improve the accuracy for both nuclides. (author)

  15. European ERANOS formulaire for fast reactor core analysis

    International Nuclear Information System (INIS)

    ERANOS code scheme was developed within the European collaboration on fast reactors. It contains all the functions required to calculate a complete set of core, shielding and fuel cycle parameters for LMFR cores. Nuclear data are taken from recent evaluations (JEF2.2) and adjusted on integral experiments (ERALIB1). Calculational scheme uses the ECCO cell code to generate cross section data. Whole core calculations are carried out using the spatial modules BISTRO (Sn) and TGVNARIANT (nodal method). Validation is based on integral and power reactor experiments. Integral experiments are also used for adjustment of nuclear data

  16. Safety analysis of the Topaz behavior during irradiation, Its effect on the core performance and the in-core fuel management strategy

    International Nuclear Information System (INIS)

    . The results of safety analysis showed that no safety criteria was violated. The maximum worth of any box was found to be 80 pcm, the total worth of all boxes were found to be 350 pcm. Moreover the paper describes the administrative procedure to limit the reactivity insertion rate of any box to less than 10 pcm/sec. To verify the calculated values, the indirect measurements of Topaz reactivity was performed. Where during the irradiation, a maneuver is performed to extract a Topaz box which have completed the specified irradiation time and another box on schedule is inserted. In this paper the position of the highest influence on Topaz reactivity is presented. On 23-11-2004 the first maneuver performed, the reactor was operating at the nominal power, and while extracting one box and inserting the other box the automatic reactor power control system (ARPCS) was connected, where the ARPCS controls and compensates the reactivity change during the reactor operation due to fuel burn up, temperature changes and Topaz maneuvering. CR-1 was selected by the ARPCS selection logic to compensate the reactivity during the maneuvering, and among the charts of the six control rods withdraw positions during the whole operation period the rod being selected by the ARPCS selection logic shows damps at the maneuver time. The control rod with drawl position chart was analysed, the rod percent extracted to compensate the reactivity during the Topaz box extraction and the rod percent inserted to compensate the reactivity during Topaz box insertion were obtained. Then from the characterization curves of control rods (s-curves) and taking into account the change of the control rod worth during the cycle, the value of the whole Topaz box reactivity worth in a specific irradiation position was obtained and were compared with the calculated values. The effect of Topaz on fuel burn up was determined. While performing the core load follow, the core is being depleted using different methodologies

  17. Ultra-sensitive detection of defective fuel rods (sipping test) with low core burn-up and long-decay period

    International Nuclear Information System (INIS)

    The wet sipping method is employed by KWU to detect defective fuel rods: water containing fission products are expelled from the damaged fuel rods and analysed by γ-spectrometry. The preferred defect indicators are 131I, 134Cs and 137Cs. During a fuel defect inspection carried out at the Biblis Plant. Unit B, in September 1976, the fission caesium of the sipping samples was concentrated and separated selectively from the radionuclide mixture. The AMP (ammonium-molybdate-phosphate) precipitation process employed for the first time in this context attained the same degree of reliability as with sipping tests performed on highly burned-up fuel rods. (orig.)

  18. Methods adopted to account for vessel-core seismic interactions in the PEC analysis

    International Nuclear Information System (INIS)

    In the PEC reactor a restraint was provided around the core in order to control the top-core displacements under seismic conditions. Clearances between adjacent elements and between the external elements and the core-restraint ring introduce some amount of non-linearity in the seismic motion of the core. On the other hand a linear analysis shows that the core strongly affects the dynamical behaviour of the vessel. Therefore a method was developed by ENEA and Ansaldo to take into account the effects of non-linearities in the decoupled detailed analysis of the core and reactor block. Studies were also performed in order to account for the possibility of influences of fluid coupling on the core vessel interaction. This paper briefly describes the adopted methods and provides an overview of the main results. (author). 2 refs, 6 figs, 1 tab

  19. Analysis of dismantling possibility and unloading efforts of fuel assemblies from core of WWER

    International Nuclear Information System (INIS)

    The computation methods of optimal dismantling sequence of fuel assemblies (FA) from core of WWER after different operating periods and accident conditions are considered. The algorithms of fuel dismantling sequence are constructed both on the basis of analysis of mutual spacer grid overlaps of adjacent fuel assemblies and numerical structure analysis of efforts required for FA removal as FA heaving from the core. Computation results for core dismantling sequence after 3-year operating period and LB LOCA are presented in the paper

  20. Transport criticality analysis for FBR MONJU initial critical core in whole core simulation by NSHEX and GMVP

    International Nuclear Information System (INIS)

    FBR MONJU Initial Critical Core (ICC) criticality problem has been solved by deterministic and Monte Carlo transport methods by the codes NSHEX and GMVP. The analysis has been carried out in different energy-groups approximations. As a result the effect of cross-section (XS) condensation from 70 into few energy-group structures by different collapsing methods has been evaluated. The 3D discrete-ordinate code NSHEX has been applied for wide range of core simulations-from whole core, considering the fissile, fertile and shielding regions to simplified models that simulate an increased neutron leakage. It has been found that there is room for improvement in the assessment of the neutron leakage in the few energy-group approximations. The good agreement between NSHEX and GMVP results, especially without XS collapsing, is pointed out as a conformation for the applicability of the code NSHEX in FBR 3D whole core calculations. Some practical conclusions have been extracted that are important for the implementation of the code NSHEX in the standard criticality analysis. (author)

  1. Plastic-dynamic analysis on shock absorber of reactor core

    International Nuclear Information System (INIS)

    The plastic-dynamic process under the condition of impact is studied for the shock absorbing device. The safety of the reactor core and vessel can be ensured by reasonably selecting the dimensions to lessen the dynamic loading factor

  2. Acoustic emission strand burning technique for motor burning rate prediction

    Science.gov (United States)

    Christensen, W. N.

    1978-01-01

    An acoustic emission (AE) method is being used to measure the burning rate of solid propellant strands. This method has a precision of 0.5% and excellent burning rate correlation with both subscale and large rocket motors. The AE procedure burns the sample under water and measures the burning rate from the acoustic output. The acoustic signal provides a continuous readout during testing, which allows complete data analysis rather than the start-stop clockwires used by the conventional method. The AE method helps eliminate such problems as inhibiting the sample, pressure increase and temperature rise, during testing.

  3. 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)

  4. TREAT Transient Analysis Benchmarking for the HEU Core

    International Nuclear Information System (INIS)

    This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used to determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term reported values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core's performance.

  5. Knowledge Economy Core Journals: Identification through LISTA Database Analysis

    OpenAIRE

    Nouri, Rasool; Karimi, Saeed; Ashrafi-rizi, Hassan; Nouri, Azadeh

    2013-01-01

    Background Knowledge economy has become increasingly broad over the years and identification of core journals in this field can be useful for librarians in journal selection process and also for researchers to select their studies and finding Appropriate Journal for publishing their articles. Present research attempts to determine core journals of Knowledge Economy indexed in LISTA (Library and Information Science and Technology). Methods The research method was bibliometric and research popu...

  6. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries

  7. Performance analysis of the 840 MWt PRISM reference burner core

    International Nuclear Information System (INIS)

    The General Electric PRISM (Power Reactor, Innovative Small Module) is a modular, pool-type sodium-cooled fast reactor employing innovative, passive features to provide an extremely high level of public safety. A PRISM power block consists of two 840 MWt reactor modules, each with a vessel diameter of 9.15 m (30 ft), tied to a turbine generator and producing 622 MWe. A full-size plant consists of three power blocks producing 1866 MWe of electrical power. Two core configurations have been analyzed. The reference is a 'burner' core (conversion ratio of 0.8) and the alternative is a breakeven' core (plutonium consumption balanced by plutonium generation). The core nuclear designs are largely governed by passive safety and reactivity control issues. The key features employed to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters (rod stops) and gas expansion modules (GEMs). A passive reactor vessel auxiliary cooling system (RVACS) assures safety-grade decay heat removal. This paper summarizes the operational and safety performance of the 840 MWt PRISM modular reactor, with emphasis on the reference burner core. (author)

  8. Adult patients are more catabolic than children during acute phase after burn injury: a retrospective analysis on muscle protein kinetics

    Science.gov (United States)

    Tuvdendorj, Demidmaa; Chinkes, David L.; Zhang, Xiao-Jun; Ferrando, Arny A.; Elijah, Itoro E.; Mlcak, Ronald P.; Finnerty, Celeste C.; Wolfe, Robert R.; Herndon, David N.

    2011-01-01

    Purpose This study was performed to determine if there is an age-related specificity in the response of muscle protein metabolism to severe burn injury during acute hospitalization. This is a retrospective analysis of previously published data. Methods: Nineteen adult and 58 pediatric burn-injured patients (age 43.3 ± 14.3 vs. 7.2 ± 5.3 years, adult vs. children) participated in stable isotope [ring-2H5]phenylalanine (Phe) infusion studies. Femoral arterial and venous blood samples and muscle biopsy samples were collected throughout the study. Data are presented as means ± standard deviation (SD). A p value less than 0.05 was considered statistically significant. Results Muscle net protein balance (NB) was higher in children (adult vs. children, -43 ± 61 vs. 8 ± 68 nmol Phe/min/100 ml leg volume, p < 0.05). Muscle protein fractional synthesis rate (FSR) was higher in children (adult vs. children, 0.11 ± 0.05 vs. 0.16 ± 0.10 %/h, p < 0.05). Leg muscle protein breakdown was not different between the groups (adult vs. children, 179 ± 115 vs. 184 ± 124 nmol Phe/ min/100 ml leg volume, p < 0.05; synthesis rate was 134 ± 96 and 192 ± 128 nmol Phe/min/100 ml leg volume in adults and children, respectively (p = 0.07). Age significantly correlated with muscle protein NB (p = 0.01) and FSR (p = 0.02); but not with breakdown (p = 0.67) and synthesis (p = 0.07) rates measured by using a three-pool model. Conclusion In burn injury, the muscle protein breakdown may be affected to the same extent in adults and children, whereas synthesis may have age-related specificities, resulting in a better but still low NB in children. PMID:21647721

  9. Single assembly preliminary analysis for horizontal seismic analysis on fast breeder reactor core

    International Nuclear Information System (INIS)

    Seismic analysis is one of important parts of fast breeder reactor (FBR) core design. It is necessary for structural integrity assessment and analysis of variation of reactivity under the earthquake. Moreover some important data for qualification of the scram capability of the control rods during the earthquake. Moreover some important data for qualification of the scram capability of the control rods during the earthquake could be provided. In the paper, FINAS, one finite element code developed by Japanese engineers, was used. The calculation model and method were studied on single assembly in China Experimental Fast Reactor (CEFR), as an example. Some preliminary analyses were carried out, which prepare for the seismic analysis on multiple assemblies in FBR core. (authors)

  10. TREAT Transient Analysis Benchmarking for the HEU Core

    Energy Technology Data Exchange (ETDEWEB)

    Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-05-01

    This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used to determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term “reported” values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core’s performance.

  11. Quantitative spectral analysis of the sdB star HD 188112: a helium-core white dwarf progenitor

    CERN Document Server

    Latour, M; Irrgang, A; Schaffenroth, V; Geier, S; Hillebrandt, W; Roepke, F K; Taubenberger, S; Kromer, M; Fink, M

    2016-01-01

    HD 188112 is a bright (V = 10.2 mag) hot subdwarf B (sdB) star with a mass too low to sustain core helium burning and is therefore considered as a pre-extremely low mass (ELM) white dwarf (WD). ELM WDs (M $\\le$ 0.3 Msun) are He-core objects produced by the evolution of compact binary systems. We present in this paper a detailed abundance analysis of HD 188112 based on high-resolution Hubble Space Telescope (HST) near and far-ultraviolet spectroscopy. We also constrain the mass of the star's companion. We use hybrid non-LTE model atmospheres to fit the observed spectral lines and derive the abundances of more than a dozen elements as well as the rotational broadening of metallic lines. We confirm the previous binary system parameters by combining radial velocities measured in our UV spectra with the already published ones. The system has a period of 0.6065858 days and a WD companion with M $\\geq$ 0.70 Msun. By assuming a tidally locked rotation, combined with the projected rotational velocity (v sin i = 7.9 $\\...

  12. The reactor core analysis code CITATION-1000VP for High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Reactor core analysis with full core model has been necessary for the High Temperature Engineering Test Reactor (HTTR) design. The CITATION-1000VP code has been developed to enable reactor core analysis of HTTR with full core model through extending the number of zones and meshes, and enhancing the calculation speed of CITATION code. This report describes the program changes for extending the number of zones and meshes, and for vectorization. The maximum number of zones and meshes becomes 999 and 500, respectively. The calculation speed is enhanced up to 21 times. (author)

  13. The integrated code system CASCADE-3D for advanced core design and safety analysis

    International Nuclear Information System (INIS)

    The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

  14. Influence of aerosols from biomass burning on the spectral analysis of Cherenkov telescopes

    CERN Document Server

    Reyes, R de los; Bernloehr, K; Krueger, P; Deil, C; Gast, H; Kosack, K; Marandon, V

    2013-01-01

    During the last decade, imaging atmospheric Cherenkov telescopes (IACTs) have proven themselves as astronomical detectors in the very-high-energy (VHE; E>0.1 TeV) regime. The IACT technique observes the VHE photons indirectly, using the Earth's atmosphere as a calorimeter. Much of the calibration of Cherenkov telescope experiments is done using Monte Carlo simulations of the air shower development, Cherenkov radiation and detector, assuming certain models for the atmospheric conditions. Any deviation of the real conditions during observations from the assumed atmospheric model will result in a wrong reconstruction of the primary gamma-ray energy and the resulting source spectra. During eight years of observations, the High Energy Stereoscopic System (H.E.S.S.) has experienced periodic natural as well as anthropogenic variations of the atmospheric transparency due to aerosols created by biomass burning. In order to identify data that have been taken under such long-term reductions in atmospheric transparency, ...

  15. Breeding ratio analysis for the improved Flower-SCWFR core

    International Nuclear Information System (INIS)

    Supercritical Water-cooled Fast Reactor (SCWFR) presents features of combining fast and light water reactor characteristics in one design. The coolant mass flow rate is just as 1/8 as in the BWR, and the neutron energy is harder than in PWR, so it would has the breeding ability. In this paper, using different models of improved Flower-SCWFR core, the void reactivity effect, power distribution,and breeding ratio are analyzed by core zoning scheme, axial coolant densities zoning,seed and blanket assembly with suitable P/D value, MOX fuel with different design and enrichment zoning, and solid uranium matrix cooled by internal clad channels in blanket assembly. Finally, an optimized modal of the improved SCWFR cores-'Flower type' is obtained. (authors)

  16. Analysis of RA-8 critical facility core in some configurations

    International Nuclear Information System (INIS)

    The RA-8 critical facility was designated and built to be used in the experimental plan of the 'CAREM' Project but is, in itself, very versatile and adequate to perform many types of other experiments. The present paper includes calculated estimates of some critical configurations and comparisons with experimental results obtained during its start up. Results for Core 1 with homogeneous arrangement of rods containing 1.8 % enriched uranium, showed very good agreement. In fact, an experimentally critical configuration was reached with 1.300 rods and calculated values were: 1.310 using the WIMS code and 1.148 from the CONDOR code. Moreover, it was verified that the estimated number of 3.4% enriched uranium rods to be fabricated is enough to build a heterogeneous core or even a homogeneous core with this enrichment. The replacement of 3.4 % enriched uranium by 3.6 % will not present problems related with the original plan. (author)

  17. analysis of reactivity accidents in MTR for various protection system parameters and core condition

    International Nuclear Information System (INIS)

    Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99Mo production. The old core comprising 29 fuel elements and one Co Irradiation Device (CID) and the new core comprising 27 fuel elements, CID, and two 99Mo production boxes. The in core irradiation has the advantage of no special cooling or irradiation loop is required. The purpose of the present work is the analysis of reactivity accidents (RIA) for ETRR-2 cores. The analysis was done to evaluate the accidents from different point of view:1- Analysis of the new core for various Reactor Protection System (RPS) parameters 2- Comparison between the two cores. 3- Analysis of the 99Mo production boxes.PARET computer code was employed to compute various parameters. Initiating events in RIA involve various modes of reactivity insertion, namely, prompt critical condition (p=1$), accidental ejection of partial and complete CID uncontrolled withdrawal of a control rod accident, and sudden cooling of the reactor core. The time histories of reactor power, energy released, and the maximum fuel, clad and coolant temperatures of fuel elements and LEU plates were calculated for each of these accidents. The results show that the maximum clad temperatures remain well below the clad melting of both fuel and uranium plates during these accidents. It is concluded that for the new core, the RIA with scram will not result in fuel or uranium plate failure.

  18. Extraction of trapped gases in ice cores for isotope analysis

    International Nuclear Information System (INIS)

    The use of ice cores for paleoclimatic investigations is discussed in terms of their application for dating, temperature indication, spatial time marker synchronization, trace gas fluxes, solar variability indication and changes in the Dole effect. The different existing techniques for the extraction of gases from ice cores are discussed. These techniques, all to be carried out under vacuum, are melt-extraction, dry-extraction methods and the sublimation technique. Advantages and disadvantages of the individual methods are listed. An extensive list of references is provided for further detailed information. (author)

  19. Recent Development of the Inter-Assembly Flow Analysis Tools for SFR Core Thermal Hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. G.; Kim, E. K.; Lee, Y. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    A typical SFR core is generally comprised of hundreds of hexagonal type ducted subassemblies. And these subassemblies have hundreds of fuel rods with a triangular channel arrangement forming a closed circuit by themselves without any flow path between them. Subchannel analysis is considered to be the most suitable method for the LMR subassembly analysis when considering the geometrical complexities and computational resources needs. MATRA-LMR was developed as an analysis code to predict flow and temperature fields in SFR subassemblies. In the SFR core, flow redistribution can be occurred in the inter-assembly region of the core. The hotter counter flow from the upper center region of the LMR core may have a significant effect on the thermo-mechanical integrity of the duct wall. This paper describes the recent development of the inter-assembly flow analysis tools for SFR core thermal hydraulics and shows a few calculation results.

  20. Magnetic, Structural, and Particle Size Analysis of Single- and Multi-Core Magnetic Nanoparticles

    DEFF Research Database (Denmark)

    Ludwig, Frank; Kazakova, Olga; Barquin, Luis Fernandez;

    2014-01-01

    We have measured and analyzed three different commercial magnetic nanoparticle systems, both multi-core and single-core in nature, with the particle (core) size ranging from 20 to 100 nm. Complementary analysis methods and same characterization techniques were carried out in different labs and the...... results are compared with each other. The presented results primarily focus on determining the particle size—both the hydrodynamic size and the individual magnetic core size—as well as magnetic and structural properties. The used analysis methods include transmission electron microscopy, static and...... dynamic magnetization measurements, and Mössbauer spectroscopy. We show that particle (hydrodynamic and core) size parameters can be determined from different analysis techniques and the individual analysis results agree reasonably well. However, in order to compare size parameters precisely determined...

  1. McCARD for neutronics design and analysis of research reactor cores

    International Nuclear Information System (INIS)

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO research reactor, and YALINA subcritical facility. (authors)

  2. Beacon: A three-dimensional structural analysis code for bowing history of fast breeder reactor cores

    International Nuclear Information System (INIS)

    The core elements of an LMFBR are bowed due to radial gradients of both temperature and neutron flux in the core. Since all hexagonal elements are multiply supported by adjacent elements or the restraint system, restraint forces and bending stresses are induced. In turn, these forces and stresses are relaxed by irradiation enhanced creep of the material. The analysis of the core bowing behavior requires a three-dimensional consideration of the mechanical interactions among the core elements, because the core consists of different kinds of elements and of fuel assemblies with various burnup histories. A new computational code BEACON has been developed for analyzing the bowing behavior of an LMFBR's core in three dimensions. To evaluate mechanical interactions among core elements, the code uses the analytical method of the earlier SHADOW code. BEACON analyzes the mechanical interactions in three directions, which form angles of 600 with one another. BEACON is applied to the 600 sector of a typical LMFBR's core for analyzing the bowing history during one equilibrium cycle. 120 core elements are treated, assuming the boundary condition of rotational symmetry. The application confirms that the code can be an effective tool for parametric studies as well as for detailed structural analysis of LMFBR's core. (orig.)

  3. Analysis of a multigroup stylized CANDU half-core benchmark

    International Nuclear Information System (INIS)

    Highlights: → This paper provides a benchmark that is a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. → An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core CANDU benchmark problem. → Reference eigenvalues and selected pin and bundle fission rates are included. → 2-, 4- and 47-group Monte Carlo solutions are compared to analyze homogenization-free transport approximations that result from energy condensation. - Abstract: An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem. Reference eigenvalues and selected pin and bundle fission rates are also included. This benchmark is intended to provide computational reactor physicists and methods developers with a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. In addition to transport theory code verification, the 8-group energy structure provides reactor physicist with an ideal problem for examining cross section homogenization and collapsing effects in a full-core environment. To this end, additional 2-, 4- and 47-group full-core Monte Carlo benchmark solutions are compared to analyze homogenization-free transport approximations incurred as a result of energy group condensation.

  4. MOX燃料堆芯热工特性及设计限值研究%Analysis of MOX core thermal characteristics and design criteria

    Institute of Scientific and Technical Information of China (English)

    刘一哲; 喻宏; 田和春

    2009-01-01

    For the MOX core of a Sodium-cooled Fast Reactor(SFR) nuclear power plant, the advantages are higher linear power, burn-up and outlet temperature. Core thermal hydraulic design meets some new issues. Based on the analysis of MOX fuel characteristics, the thermal design criteria of core were raised in this paper. Core thermal characteristics and margin were analyzed for the 870 MWe nuclear power plant. The results showed that higher thermal parameters were reasonable and feasible for the MOX core, and the thermal margin satisfied the requirements.%使用MOX燃料的快堆核电站以其线功率高、燃耗高、堆芯出口温度高等特点,对堆芯热工设计提出了新的问题.本文在对MOX燃料热工性能分析的基础上,给出了主要的热工设计限值,并以电功率870 MW电站为参考,初步分析了其堆芯热工特性和设计裕量.结果表明对于MOX燃料,较高的堆芯热工参数合理可行,且具有足够的裕量.

  5. Modelling and analysis of the behavior of LWRs at severe core accidents

    International Nuclear Information System (INIS)

    With respect to the assessment of the consequences of severe accidents in light water reactors from the initiation of the accident up to the thermal failure of the reactor pressure vessel (RPV), a modular program system has been developed. Experimental results will be considered with respect to the modeling of the fuel rod behavior, e.g. deformation of the fuel rod, metal water reaction and the melting of the fuel rods. The fuel and core models allow to estimate the coolability of fuel rods and core as well as the consequences of core meltdown accidents at various pressure levels. After partial failure of the lower core retention structure, the core material will drop into the lower plenum and heat up the RPV. This strong interaction between the thermal behavior of the remaining core and the partially dropped core material has been modeled because of an accident sequence analysis. The analyses described here show, that not the entire core will fail, but a partial drop of core material into the lower plenum is likely to occur. With respect to the validation of the program system, comparison calculations with the fuel rod behavior and melt models SSYST and EXMEL will be performed. Moreover, the program system will be applied to the bundle behavior in meltdown experiments, the TMI-2 core behavior and the course of a core meltdown accident in risk studies. (orig.)

  6. 2011 annual report of validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    This report summarizes results obtained by the validation study of core analysis methods for Full MOX BWR. Validation Study of Core Analysis Methods for Full MOX BWR is aimed at compiling technical data base, which is used to assess calculated characteristics of MOX cores from initial (MOX-one-third-loaded core) to full MOX cores in a full MOX BWR nuclear plant and review the validation of nuclear design methods in safety assessment of high burnup MOX fuel cores in future. For this purpose, experiments and analysis of obtained data have been implemented for core physics experiments on irradiated MOX fuel and core physics experiments on full MOX cores. Analysis of isotopic composition measurements of MOX fuel irradiated in a MOX-one-third-loaded core, and isotopic composition measurements and analysis of the obtained data on MOX fuel irradiated in the full MOX BWR are planned. Results of core physics experiments on irradiated BWR MOX fuel (initial fissile Pu content: 5.5 wt%, fuel burnup: 61 GWd/t) in the REBUS program have been analyzed, and it was observed that biases of effective neutron multiplication factors keff's, (keff - 1.0), were -1.0%Δk (Nuclear data library: JENDL-3.2), -0.3%Δk (JENDL-3.3) and 0.2%Δk(JENDL-3.2) for the diffusion, transport and Monte Carlo calculations, respectively. The analysis results of each analysis method reproduced well the measured fission rate and activation rate distributions for fuel rods. The core analysis results obtained by using typical core analysis codes (CASMO-4/ SIMULATE-3 (JNES's version)) with coarse geometrical meshes and a few energy groups showed keff's of 0.986 and 0.993 for the six critical cores of the REBUS program, which showed underestimation. Core physics experiments of the FUBILA program included a full MOX core controlled by a B4C control rod, a core containing Gd2O3-UO2, fuel and UO2 fuel rods, a 10x10 MOX assembly core and a 9x9 reference core consisting of about 17 month elapsed MOX fuel. Analysis

  7. Methodology for seismic analysis of FBR core assembly using variable added damping

    International Nuclear Information System (INIS)

    It is necessary to consider the fluid-solid coupling effect due to the interaction between coolant and core assembly when analyzing the seismic performance of the FBR core assembly. The added damping was treated mostly as a constant in previous researches. In fact, the effect on assemblies from the coolant depends strongly on the gap between the core assemblies, and the damping should be considered as a variable. In order to simulate the vibration of the core assembly more accurately, the methodology for the seismic analysis of FBR core assemblies using variable added damping was studied. In this paper, the seismic analysis model of one single row of core assemblies (5 assemblies) of FBR was established. By comparing the two kinds of added damping models, the constant and variable ones respectively, the results show that the seismic analysis of the core assemblies with variable added damping is feasible and effective. Meanwhile, the simulation method used in this paper can obtain more precise approximation of the vibration of the core assembly and lays the foundation for more realistically simulating the seismic response of reactor core assemblies. It also helps to reduce the conservative margin of the structural design and is meaningful in engineering application. (authors)

  8. Heterogeneous Multi core processors for improving the efficiency of Market basket analysis algorithm in data mining

    OpenAIRE

    L, Aashiha Priyadarshni.

    2014-01-01

    Heterogeneous multi core processors can offer diverse computing capabilities. The efficiency of Market Basket Analysis Algorithm can be improved with heterogeneous multi core processors. Market basket analysis algorithm utilises apriori algorithm and is one of the popular data mining algorithms which can utilise Map/Reduce framework to perform analysis. The algorithm generates association rules based on transactional data and Map/Reduce motivates to redesign and convert the existing sequentia...

  9. Nuclear data needs for the analysis of generation and burn-up of actinide isotopes in nuclear reactors

    International Nuclear Information System (INIS)

    A reliable prediction of the in-pile and out-of-pile physics characteristics of nuclear fuel is one of the objectives of present-day reactor physics. The paper describes the main production paths of important actinides for light water and fast breeder reactors. The accuracy of recent nuclear data is examined by comparisons of theoretical predictions with the results from post-irradiation analysis of nuclear fuel from power reactors, and partly with results obtained in zero-power facilities. A world-wide comparison of nuclear data to be used in large fast power reactor burn-up and long term considerations is presented. The needs for further improvement of nuclear data are discussed. (orig.)

  10. Chorioretinal burn: body temperature dependence

    International Nuclear Information System (INIS)

    Irradiance thresholds for chorioretinal damage in rhesus monkeys vary linearly with core temperatures between 34 and 390C. Damage results from the combined thermal effects of retinal irradiation and the body temperature. Visible damage is calculated to occur at a tissue temperature of 42.50C. Fever increases the retina's susceptibility to burns from the sun, lasers, and other radiant energy sources

  11. Analysis of rugae in burn victims and cadavers to simulate rugae identification in cases of incineration and decomposition.

    Science.gov (United States)

    Muthusubramanian, M; Limson, K S; Julian, R

    2005-06-01

    The most challenging situations in Forensic Odonto-Stomatology are mass disasters, where the forensic dentist is usually confronted with charred human remains or heavily decomposed or fragmented bodies. This article determines the extent of preservation of palatal rugae for use as an alternative identification tool in such situations, using a study group comprising burn victims and cadavers simulating forensic cases of incineration and decomposition. The thermal effects and the decomposition changes on the palatal rugae of burn victims with panfacial third degree burns and human cadavers in storage were respectively assessed and graded on a new scale. Ninety three percent of burn victims and 77% of human cadavers had Grade 0 changes (normal). When changes were noted, they were less pronounced than the generalized body involvement of burns in burn victims and the generalized body decomposition of human cadavers. PMID:16223023

  12. Broadband radio-frequency spectrum analysis in spectral-hole-burning media.

    Science.gov (United States)

    Colice, Max; Schlottau, Friso; Wagner, Kelvin H

    2006-09-01

    We demonstrate a 20 GHz spectrum analyzer with 1 MHz resolution and >40 dB dynamic range using spectral-hole-burning (SHB) crystals, which are cryogenically cooled crystal hosts lightly doped with rare-earth ions. We modulate a rf signal onto an optical carrier using an electro-optic intensity modulator to produce a signal beam modulated with upper and lower rf sidebands. Illuminating SHB crystals with modulated beams excites only those ions resonant with corresponding modulation frequencies, leaving holes in the crystal's absorption profile that mimic the modulation power spectrum and persist for up to 10 ms. We determine the spectral hole locations by probing the crystal with a chirped laser and detecting the transmitted intensity. The transmitted intensity is a blurred-out copy of the power spectrum of the original illumination as mapped into a time-varying signal. Scaling the time series associated with the transmitted intensity by the instantaneous chirp rate yields the modulated beam's rf power spectrum. The homogeneous linewidth of the rare-earth ions, which can be 20 GHz, determines the spectral bandwidth. PMID:16912776

  13. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated

  14. Magnetic loss analysis in Mn-Zn ferrite cores

    International Nuclear Information System (INIS)

    Magnetic losses have been measured and analyzed upon a wide range of frequencies in Mn-Zn ferrite ring cores. Exploiting the concept of loss separation and modeling the conductivity process in the heterogeneous material as a function of frequency, the role of the different energy dissipation mechanisms has been elucidated. It is shown, in particular, that eddy current effects can be appreciated, in standard materials and cores, only on approaching and overcoming the MHz range. The basic mechanism for hysteresis and low-frequency losses is therefore identified with the domain wall relaxation engendered by spin damping processes. Resonant absorption of energy associated with magnetization rotation is in turn deemed to chiefly contribute to the loss upon the practical range of frequencies going from a few 104Hz to a few MHz

  15. A TEM analysis of nanoparticulates in a Polar ice core

    International Nuclear Information System (INIS)

    This paper explores the prospect for analyzing nanoparticulates in age-dated ice cores representing times in antiquity to establish a historical reference for atmospheric particulate regimes. Analytical transmission electron microscope (TEM) techniques were utilized to observe representative ice-melt water drops dried down on carbon/formvar or similar coated grids. A 10,000-year-old Greenland ice core was melted, and representative water drops were transferred to coated grids in a clean room environment. Essentially, all particulates observed were aggregates and either crystalline or complex mixtures of nanocrystals. Especially notable was the observation of carbon nanotubes and related fullerene-like nanocrystal forms. These observations are similar with some aspects of contemporary airborne particulates including carbon nanotubes and complex nanocrystal aggregates

  16. Joyo MK-II core bowing analysis based on thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    A study on the inherent safety test at the Experimental Fast Reactor Joyo has been performed to demonstrate the inherent safety of fast breeder reactors. In this study, emphasis was placed on the improvement on the feedback reactivity calculation accuracy. The investigation work for core bowing calculation has been continued because it is expected to cause negative feedback reactivity that would improve the passive safety of a fast breeder reactor. The core bowing behavior in JOYO has been analyzed by the system which consists of the MK-II core management code system MAGI, the interface code TETRAS and the core bowing calculation code BEACON. As it was supposed that the coolant flow inside of the reactor vessel might effects on wrapper tube temperature, detailed coolant flow was calculated by single phase multidimensional thermal-hydraulic analysis code AQUA. (1) As a result of the AQUA calculation, it was made clear that the coolant flow effect on the coolant temperature was negligible in fuel region. (2) The coolant temperature at the outlet of reflectors adjacent to a fuel subassembly are affected by the coolant flow that comes from the outlet of reflectors in the 6th and the 7th row. It decreases the outlet temperature of the reflectors in the 5th row in AQUA calculation. (3) High temperature coolant flow exists in neighbor of the outlet of reflectors in the 8-10th row. As a result, coolant temperature calculated by AQUA are higher in 30-40degC than that calculated by TETRAS. It was made clear that the coolant flow inside of the reactor vessel had no effect on driver fuel bowing, which was the dominant factor of the core bowing reactivity. On the other hand, in reflectors region, it affects the wrapper tube temperature, which determine the irreversible swelling and creep. Essentially, in order to verify the feedback reactivity effect caused by the core bowing, it is desired to measure the mechanical behavior of the subassemblies under power operation, but it is

  17. Experimental Analysis and Modeling of the Crushing of Honeycomb Cores

    Science.gov (United States)

    Aminanda, Y.; Castanié, B.; Barrau, J.-J.; Thevenet, P.

    2005-05-01

    In the aeronautical field, sandwich structures are widely used for secondary structures like flaps or landing gear doors. The modeling of low velocity/low energy impact, which can lead to a decrease of the structure strength by 50%, remains a designer’s main problem. Since this type of impact has the same effect as quasi-static indentation, the study focuses on the behavior of honeycomb cores under compression. The crushing phenomenon has been well identified for years but its mechanism is not described explicitly and the model proposed may not satisfy industrial purposes. To understand the crushing mechanism, honeycomb test specimens made of Nomex™, aluminum alloy and paper were tested. During the crushing, a CCD camera showed that the cell walls buckled very quickly. The peak load recorded during tests corresponded to the buckling of the common edge of three honeycomb cells. Further tests on corner structures to simulate only one vertical edge of a honeycomb cell show a similar behavior. The different specimens exhibited similar load/displacement curves and the differences observed were only due to the behavior of the different materials. As a conclusion of this phenomenological study, the hypothesis that loads are mainly taken by the vertical edge can be made. So, a honeycomb core subjected to compression can be modeled by a grid of nonlinear springs. A simple analytical model was then developed and validated by tests on Nomex™ honeycomb core indented by different sized spherical indenters. A good correlation between theory and experiment was found. This result can be used to satisfactorily model using finite elements the indentation on a sandwich structure with a metallic or composite skin and honeycomb core.

  18. Supermode analysis of the 18-core photonic crystal fiber laser

    Institute of Scientific and Technical Information of China (English)

    王远; 姚建铨; 郑一博; 温午麒; 陆颖; 王鹏

    2012-01-01

    The modal of 18-core photonic crystal fiber laser is discussed and calculated.And corresponding far-field distribution of the supermodes is given by Fresnel diffraction integral.For improving beam quality,the mode selection method based on the Talbot effect is introduced.The reflection coefficients are calculated,and the result shows that an in-phase supermode can be locked better at a large propagation distance.

  19. Core support block thermal mixing test analysis report

    International Nuclear Information System (INIS)

    The extent of gas mixing and pressure drop within the core support block was experimentally investigated for various geometric and height configurations. These tests were conducted by the Experimental Engineering Branch of General Atomic Company. As a result of this investigation, the core support block thermal mixing and pressure drop has been quantified. Thermal mixing and the temperature sensor accuracy can be substantially improved at the cost of higher pressure drop. A 70-degree miter angle configuration is recommended for the reference design of the HTGR core support block (CSB). The recommended CSB height will depend on further evaluation of the possible range of variations in fuel region reactor conditions to be determmined by the Systems Engineering Department. The average temperature in a rodded region (a region with control rods in the lowered position) can be measured by the temperature sensor to within a 450F accuracy, a big improvement from an early CSB design tested by the Commissariat a La Energie Atomique at Saclay, France in 1974 and 1975

  20. Application of nodal equivalence parameters to prismatic VHTR core analysis

    International Nuclear Information System (INIS)

    The generation of nodal cross sections and equivalence parameters for prismatic VHTR core components is discussed. For fuel-block cross section generation, a conventional single-block model with a reflective boundary condition is used. A one-dimensional fuel-reflector model is proposed for reflector cross section generation in order to accurately represent the significant neutron spectrum variation at the core-reflector interface. Two-dimensional multi-block models are used for obtaining control rod cross sections for rodded fuel and reflector blocks to best approximate actual spectra in the blocks. The verification of the models was performed by generating cross sections with the DRAGON and HELIOS codes, using the cross section data in 2-D and 3-D DIF3D nodal calculations, and comparing the results to MCNP4C ones. The results show that the use of discontinuity factors reduces errors in nodal solutions for the multiplication factor and power distribution. Surface-dependent discontinuity factors are found essential for improving the accuracy of the power distribution of cores with asymmetrically rodded blocks when nodal calculations are performed with one node per hexagonal block. (authors)

  1. Performance modeling and analysis of parallel Gaussian elimination on multi-core computers

    Directory of Open Access Journals (Sweden)

    Fadi N. Sibai

    2014-01-01

    Full Text Available Gaussian elimination is used in many applications and in particular in the solution of systems of linear equations. This paper presents mathematical performance models and analysis of four parallel Gaussian Elimination methods (precisely the Original method and the new Meet in the Middle –MiM– algorithms and their variants with SIMD vectorization on multi-core systems. Analytical performance models of the four methods are formulated and presented followed by evaluations of these models with modern multi-core systems’ operation latencies. Our results reveal that the four methods generally exhibit good performance scaling with increasing matrix size and number of cores. SIMD vectorization only makes a large difference in performance for low number of cores. For a large matrix size (n ⩾ 16 K, the performance difference between the MiM and Original methods falls from 16× with four cores to 4× with 16 K cores. The efficiencies of all four methods are low with 1 K cores or more stressing a major problem of multi-core systems where the network-on-chip and memory latencies are too high in relation to basic arithmetic operations. Thus Gaussian Elimination can greatly benefit from the resources of multi-core systems, but higher performance gains can be achieved if multi-core systems can be designed with lower memory operation, synchronization, and interconnect communication latencies, requirements of utmost importance and challenge in the exascale computing age.

  2. Prediction of Storage Life of Propellants having Different Burning Rates using Dynamic Mechanical Analysis

    Directory of Open Access Journals (Sweden)

    V.S. Wani

    2012-09-01

    Full Text Available Propellants, visco-elastic in nature, show time and temperature dependent behaviour on deformation. Hence, the time–temperature superposition principle may be applied to the visco-elastic properties of propellants. In the present study, dynamic mechanical analyser (DMA was used to evaluate the dynamic mechanical properties and quantify the storage life of four different propellants based on hydroxyl terminated polybutadiene, aluminium and ammonium perchlorate having different burning rates ranging from 5 mm/s to 25 mm/s. Each sample was given a multi-frequency strain of 0.01 per cent at three discrete frequencies (3.5 Hz, 11 Hz, 35 Hz in the temperature range - 80 °C to + 80 °C. The storage modulus, loss modulus, tan delta and glass transition temperature (Tg for each propellant samples have been evaluated and it is observed that all the propellants have shown time (frequency and temperature dependent behaviour on deformation. A comparison of the log aT versus temperature curves (where aT is horizontal (or time shift factor for all four propellants indicate conformance to the Williams–Landel–Ferry (WLF equation. The master curves of storage modulus (log É versus log ω plots were generated for each propellant. A plot of É versus time for all propellants was generated up to 3 years, 6 years, and 10 years of time, respectively. The drop in the storage modulus below the acceptable limit with time may be used to predict the shelf life of the propellant.Defence Science Journal, 2012, 62(5, pp.290-294, DOI:http://dx.doi.org/10.14429/dsj.62.2480

  3. Analysis of the critical and first full power operating cores for PARR using leu oxide fuel

    International Nuclear Information System (INIS)

    This paper explains the analysis for determining the first full power operating core for PARR using LEU oxide fuel. The core configuration selected for this first full power operation contains about 6.13 kg of U-235 distributed in 19 standard and five control fuel elements. The neutron flux level is doubled when core is shifted from 5MW to 10 MW. Total nuclear power peaking factor of the core is 2.03. The analysis shows that the core can be operated safely at 5 MW with a flow rate of 520 meter cube per hour and at 10 MW with a flow rate of 900 meter cube per hour. (A.B.). 10 figs

  4. Tank 241-B-203 push mode core sampling and analysis plan. Revision 1

    International Nuclear Information System (INIS)

    This Sampling and Analysis Plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for two push-mode core samples from tank 241-B-203 (B-203)

  5. Validation of helios for ATR core follow analysis

    International Nuclear Information System (INIS)

    This work summarizes the validation analyses for the HELIOS code to support core design and safety assurance calculations of the Advanced Test Reactor (ATR). Past and current core safety assurance is performed by the PDQ-7 diffusion code; a state of the art reactor physics simulation tool from the nuclear industry's earlier days. Over the past twenty years, improvements in computational speed have enabled the use of modern neutron transport methodologies to replace the role of diffusion theory for simulation of complex systems, such as the ATR. More exact methodologies have enabled a paradigm-shift away from highly tuned codes that force compliance with a bounding safety envelope, and towards codes regularly validated against routine measurements. To validate HELIOS, the 16 ATR operational cycles from late-2009 to present were modeled. The computed power distribution was compared against data collected by the ATR's on-line power surveillance system. It was found that the ATR's lobe-powers could be determined with ±10% accuracy. Also, the ATR's cold startup shim configuration for each of these 16 cycles was estimated and compared against the reported critical position from the reactor log-book. HELIOS successfully predicted criticality within the tolerance set by the ATR startup procedure for 13 out of the 16 cycles. This is compared to 12 times for PDQ (without empirical adjustment). These findings, as well as other insights discussed in this report, suggest that HELIOS is highly suited for replacing PDQ for core safety assurance of the ATR. Furthermore, a modern verification and validation framework has been established that allows reactor and fuel performance data to be computed with a known degree of accuracy and stated uncertainty. (author)

  6. Determination and analysis of trace metals and surfactant in air particulate matter during biomass burning haze episode in Malaysia

    Science.gov (United States)

    Ahmed, Manan; Guo, Xinxin; Zhao, Xing-Min

    2016-09-01

    Trace metal species and surface active agent (surfactant) emitted into the atmosphere from natural and anthropogenic source can cause various health related and environmental problems. Limited data exists for determinations of atmospheric particulate matter particularly trace metals and surfactant concentration in Malaysia during biomass burning haze episode. We used simple and validated effective methodology for the determination of trace metals and surfactant in atmospheric particulate matter (TSP & PM2.5) collected during the biomass burning haze episode in Kampar, Malaysia from end of August to October 2015. Colorimetric method of analysis was undertaken to determine the concentration of anionic surfactant as methylene blue active substance (MBAS) and cationic surfactant as disulphine blue active substance (DBAS) using a UV-Visible spectrophotometer. Particulate samples were also analyzed for trace metals with inductive coupled plasma mass spectrometer (ICP-MS) followed by extraction from glass microfiber filters with close vessel microwave acid digestion. The result showed that the concentrations of surfactant in both samples (TSP & PM2.5) were dominated by MBAS (0.147-4.626 mmol/m3) rather than DBAS (0.111-0.671 mmol/m3) and higher than the other researcher found. Iron (147.31-1381.19 μg/m3) was recorded leading trace metal in PM followed by Al, Zn, Pb, Cd, Cr and others. During the haze period the highest mass concentration of TSP 313.34 μg/m3 and 191.07 μg/m3 for PM2.5 were recorded. Furthermore, the backward air trajectories from Kampar in north of peninsular Malaysia confirmed that nearly all the winds paths originate from Sumatera and Kalimantan, Indonesia.

  7. The results of BN600 hybrid benchmark core analysis

    International Nuclear Information System (INIS)

    The present paper includes the results of phase 1 (RZ, two dimensional model) calculations of the BN-600 hybrid core benchmark problem. The methods applied consisted of: diffusion approximation with calculation of direct and adjoint problems; calculation of reactivity coefficients; ABBN-93 nuclear data library processing, (18 group calculations). Phase 2 (Hex-Z, three dimensional model) consists of diffusion approximation with calculation of direct and adjoint problems; calculation of reactivity coefficients using first order perturbation theory; nuclear data processing code for the ABBN-78 data library. Results presented include: multiplication factors, Doppler coefficients, fuel and structure density coefficients, expansion coefficients, power distribution, beta-effective values, reaction rate distributions

  8. Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

    International Nuclear Information System (INIS)

    Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

  9. TRAC-BF1 thermal-hydraulic, ANSYS stress analysis for core shroud cracking phenomena

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission sent Generic Letter 94-03 informing all licensees about the intergranular stress corrosion cracking (IGSCC) of core shrouds found in both Dresden unit I and Quad Cities unit 1. The letter directed all licensees to perform safety analysis of their boiling water reactor (BWR) units. Two transients of special concern for the core shroud safety analysis include the main steam line break (MSLB) and recirculation line break transient

  10. Preliminary Verification Calculation of DeCART/CAPP System by HTTR Core Analysis

    International Nuclear Information System (INIS)

    In this study, the DeCART/CAPP system verification calculations have been performed against the Japan's HTTR (High Temperature Engineering Test Reactor) configurations. The calculations are carried out for single cell and single block models. The reference calculations are performed by the McCARD code. The two step core analysis system HELIOS/CAPP or DeCART/CAPP has been developed for VHTR core analysis by KAERI. In the system, first the HELIOS or DeCART code is used for homogenized cross-section generation, and second the CAPP is used to calculate the core physics parameters

  11. Preliminary Verification Calculation of DeCART/CAPP System by HTTR Core Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Han, Tae Young; Lee, Hyun Chul; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In this study, the DeCART/CAPP system verification calculations have been performed against the Japan's HTTR (High Temperature Engineering Test Reactor) configurations. The calculations are carried out for single cell and single block models. The reference calculations are performed by the McCARD code. The two step core analysis system HELIOS/CAPP or DeCART/CAPP has been developed for VHTR core analysis by KAERI. In the system, first the HELIOS or DeCART code is used for homogenized cross-section generation, and second the CAPP is used to calculate the core physics parameters.

  12. Fast reactor core monitoring by analysis of temperature noise

    International Nuclear Information System (INIS)

    The study shows, with the results obtained, how it is possible to approach the problem of diagnosis with a technique based on the use of algorithms for statistical pattern recognition was justifiable. The results presented here, with a view to their use for fast breeder reactor core surveillance, are very encouraging, the most important point being the data representation. For this study, it was difficult to find the most suitable parameters for characterizing the various simulated core states, however, despite this handicap, the classification algorithm provided quite acceptable results. The second point concerns the characterization of a system's evolution. The criterion defined was chosen for adaptation to our algorithm. One acertained that it was possible to characterize evolution on the basis of this criterion as long as the rejected points were not too far from the known learning sets. Under these circumstances, the advantage in characterizing evolution in that the changes in evolution occur when the rejected points have a tendency to agglomerate in a small area of space could be seen. This phenomenon thus makes it possible to forsee whether the creation of a new class is possible. Where the rejected points are far away from the known learning sets, the criterion used proved to be too sensitive and the characterization of evolution was less satisfactory

  13. Burning plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Furth, H.P.; Goldston, R.J.; Zweben, S.J. (Princeton Univ., NJ (USA). Plasma Physics Lab.); Sigmar, D.J. (Massachusetts Inst. of Tech., Cambridge, MA (USA))

    1990-10-01

    The fraction of fusion-reaction energy that is released in energetic charged ions, such as the alpha particles of the D-T reaction, can be thermalized within the reacting plasma and used to maintain its temperature. This mechanism facilitates the achievement of very high energy-multiplication factors Q, but also raises a number of new issues of confinement physics. To ensure satisfactory reaction operation, three areas of energetic-ion interaction need to be addressed: single-ion transport in imperfectly symmetric magnetic fields or turbulent background plasmas; energetic-ion-driven (or stabilized) collective phenomena; and fusion-heat-driven collective phenomena. The first of these topics is already being explored in a number of tokamak experiments, and the second will begin to be addressed in the D-T-burning phase of TFTR and JET. Exploration of the third topic calls for high-Q operation, which is a goal of proposed next-generation plasma-burning projects. Planning for future experiments must take into consideration the full range of plasma-physics and engineering R D areas that need to be addressed on the way to a fusion power demonstration.

  14. Epidemiology of U.K. military burns.

    Science.gov (United States)

    Foster, Mark Anthony; Moledina, Jamil; Jeffery, Steve L A

    2011-01-01

    The authors review the etiology of U.K. military burns in light of increasing hybrid warfare. Analysis of the nature of these injured personnel will provide commanders with the evidence to plan for on-going and future operations. Case notes of all U.K. Armed Forces burn injured patients who were evacuated to the Royal Centre for Defence Medicine were reviewed. Demographics, burn severity, pattern, and mortality details were included. There were 134 U.K. military personnel with burns requiring return to the United Kingdom during 2001-2007. The median age was 27 (20-62) years. Overall, 60% of burns seen were "accidental." Burning waste, misuse or disrespect of fuel, and scalds were the most prevalent noncombat burns. Areas commonly burned were the face, legs, and hands. During 2006-2007 in the two major conflicts, more than 59% (n = 36) of the burned patients evacuated to the United Kingdom were injured during combat. Burns sustained in combat represent 5.8% of all combat casualties and were commonly associated with other injuries. Improvised explosive device, minestrike, and rocket-propelled grenade were common causes. The mean TBSA affected for both groups was 5% (1-70). The majority of combat burn injuries have been small in size. Greater provision of flame retardant equipment and clothing may reduce the extent and number of combat burns in the future. The numbers of noncombat burns are being reduced by good military discipline. PMID:21422938

  15. Reconstruction and analysis of temperature and density spatial profiles inertial confinement fusion implosion cores

    International Nuclear Information System (INIS)

    We discuss several methods for the extraction of temperature and density spatial profiles in inertial confinement fusion implosion cores based on the analysis of the x-ray emission from spectroscopic tracers added to the deuterium fuel. The ideas rely on (1) detailed spectral models that take into account collisional-radiative atomic kinetics, Stark broadened line shapes, and radiation transport calculations, (2) the availability of narrow-band, gated pinhole and slit x-ray images, and space-resolved line spectra of the core, and (3) several data analysis and reconstruction methods that include a multi-objective search and optimization technique based on a novel application of Pareto genetic algorithms to plasma spectroscopy. The spectroscopic analysis yields the spatial profiles of temperature and density in the core at the collapse of the implosion, and also the extent of shell material mixing into the core. Results are illustrated with data recorded in implosion experiments driven by the OMEGA and Z facilities

  16. Geochemical analysis of core from a geothermal anomaly

    International Nuclear Information System (INIS)

    A mild geothermal area in western Montana, USA, has been studied, as a natural analog, to learn about the effects that long-term heat generated by a repository containing spent nuclear fuel might have on the surrounding rock mass. The results of previous geological, geophysical and hydrogeological studies are briefly summarized. Extensive petrological studies have been undertaken on core samples obtained from a 2 km deep borehole drilled into the Empire Creek Stock. These include a detailed petrographic study, x-ray diffraction analyses, scanning electron microscope and electron microprobe analyses, porosity and permeability measurements, oxygen isotope analyses, uranium disequilibrium analyses and K-Ar age determinations. The implications to deep burial of nuclear wastes are discussed. 40 refs

  17. The effects of core zoning on optimization of design analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Highlights: • 1/8 of core is simulated by MCNP and thermal-hydraulic code simultaneously. • Effects of core zoning are studied by dividing the core into two regions. • Both the neutronics and thermal-hydraulic behavior are investigated. • The flat flux distribution is achieved in the optimization analysis. • The flat flux can lead to worse thermal-hydraulic behavior occasionally. - Abstract: The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly

  18. Behavior of an heterogeneous annular FBR core during an unprotected loss of flow accident: Analysis of the primary phase with SAS-SFR

    International Nuclear Information System (INIS)

    In the framework of a substantial improvement on FBR core safety connected to the development of a new Gen IV reactor type, heterogeneous core with innovative features are being carefully analyzed in France since 2009. At EDF R and D, the main goal is to understand whether a strong reduction of the Na-void worth - possibly attempting a negative value - allows a significant improvement of the core behavior during an unprotected loss of flow accident. Also, the physical behavior of such a core is of interest, before and beyond the (possible) onset of Na boiling. Hence, a cutting-edge heterogeneous design, featuring an annular shape, a Na-plena with a B4C plate and a stepwise modulation of fissile core heights, was developed at EDF by means of the SDDS methodology, with a total Na-void worth of -1 $. The behavior of such a core during the primary phase of a severe accident, initiated by an unprotected loss of flow, is analyzed by means of the SAS-SFR code. This study is carried-out at KIT and EDF, in the framework of a scientific collaboration on innovative FBR severe accident analyses. The results show that the reduction of the Na-void worth is very effective, but is not sufficient alone to avoid Na-boiling and, hence, to prevent the core from entering into the primary phase of a severe accident. Nevertheless, the grace time up to boiling onset is greatly enhanced in comparison to a more traditional homogeneous core design, and only an extremely low fraction of the fuel (<0.1%) enters into melting at the end of this phase. A sensitivity analysis shows that, due to the inherent neutronic characteristics of such a core, the gagging scheme plays a major role on the core behavior: indeed, an improved 4-zones gagging scheme, associated with an enhanced control rod drive line expansion feed-back effect, finally prevents the core from entering into sodium boiling. This major conclusion highlights both the progress already accomplished and the need for more detailed future

  19. Experimental burn plot trial in the Kruger National Park: history, experimental design and suggestions for data analysis

    Directory of Open Access Journals (Sweden)

    R. Biggs

    2003-12-01

    Full Text Available The experimental burn plot (EBP trial initiated in 1954 is one of few ongoing long-termfire ecology research projects in Africa. The trial aims to assess the impacts of differentfire regimes in the Kruger National Park. Recent studies on the EBPs have raised questions as to the experimental design of the trial, and the appropriate model specificationwhen analysing data. Archival documentation reveals that the original design was modified on several occasions, related to changes in the park's fire policy. These modifications include the addition of extra plots, subdivision of plots and changes in treatmentsover time, and have resulted in a design which is only partially randomised. The representativity of the trial plots has been questioned on account of their relatively small size,the concentration of herbivores on especially the frequently burnt plots, and soil variation between plots. It is suggested that these factors be included as covariates inexplanatory models or that certain plots be excluded from data analysis based on resultsof independent studies of these factors. Suggestions are provided for the specificationof the experimental design when analysing data using Analysis of Variance. It is concluded that there is no practical alternative to treating the trial as a fully randomisedcomplete block design.

  20. Inspection of spar-core bond in helicopter rotor blades using finite element analysis

    Science.gov (United States)

    Chakrapani, Sunil Kishore; Barnard, Daniel J.; Dayal, Vinay

    2015-03-01

    This work focuses on inspection of spar-core bond of a helicopter rotor blade using finite element analysis. Structures which have high density, high stiffness cores can be very difficult to inspect due to various mode conversions. FEM was used to capture these mode conversions effectively. The structure consists of a thin spar section followed by a spar-core half space and another thin spar section. A Lamb wave excited in the spar section can mode convert into a Rayleigh wave in the spar-core section due to the coupling of the core material. This in turn mode converts back into a Lamb wave upon interacting with the next spar section. This work focuses solely on capturing the mode conversions between Rayleigh and Lamb modes at different discontinuities in the geometry.

  1. Analysis of impurity effect on Silicide fuels of the RSG-GAS core

    International Nuclear Information System (INIS)

    Simulation of impurity effect on silicide fuel of the RSG-GAS core has been done. The aim of this research is to know impurity effect of the U-234 and U-236 isotopes in the silicide fuels on the core criticality. The silicide fuels of 250 g U loading and 19.75 of enrichment is used in this simulation. Cross section constant of fuels and non-structure material of core are generated by WIMSD/4 computer code, meanwhile impurity concentration was arranged from 0.01% to 2%. From the result of analysis can be concluded that the isotopes impurity in the fuels could make trouble in the core and the core can not be operated at critical after a half of its cycle length (350 MW D)

  2. New time-line technique for station blackout core-melt analysis

    International Nuclear Information System (INIS)

    Florida Power Corporation (FPC) has developed a new method for analyzing station blackout (SBO) core-melt accidents. This method, created during the recent probabilistic risk assessment (PRA) of Crystal River Unit 3 (CR-3), originated from the need to analyze the interactions among the two-train emergency feedwater (EFW) system, station batteries, and diesel generators (DGs) following a loss of off-site power (LOSP) event. SBO core-melt sequences for CR-3 are unique since the time core-melt commences depends on which DG fails last. The purpose of this paper is to outline the new method of analysis of SBO core-melt accidents at CR-3. The significance of SBO core-melt accidents to total plant risk, along with the efficacy of various methods to reduce SBO risk, are also discussed

  3. Reactor analysis methods. 7. MCNP4B Analysis of the HTR-10 Startup Core

    International Nuclear Information System (INIS)

    A benchmark criticality analysis was performed of the HTR-10 pebble bed reactor recently constructed at the Institute of Nuclear Energy Technology, Beijing, using the MCNP4B continuous-energy Monte Carlo code. This analysis was part of the U.S. contribution to the International Atomic Energy Agency Coordinated Research Program (CRP-5) on the evaluation of high-temperature gas-cooled reactor performance. The HTR-10 is a 10-MW(thermal) pebble bed reactor that uses graphite spheres that are 6 cm in diameter and contain embedded coated fuel particles (CFPs) with 17% enriched UO2. The uranium loading per fuel sphere is 5.0 g. The full core consists of ∼27 000 spheres randomly packed in a cylindrical cavity with a mean height of 1.97 m, a diameter of 1.8 m, and a volume of 5.0m3. The core is surrounded by a structure consisting of a graphite reflector and a borated carbon shield. The radial reflector, which is 1 m thick, contains channels for the control and shutdown systems, irradiation sites, and helium coolant. The initial approach to critical was achieved by filling the discharge tube and cone at the bottom of the core with moderator spheres, then adding a random mixture of fuel and moderator spheres until the critical mass was achieved. The total number of spheres needed to reach criticality was 16 890, with a fuel-to-moderator sphere ratio (F/M) of 57 to 43%. Although the physics benchmark problem consists of three parts, only the first part is considered here. Problem B1 calls for the prediction of the initial, cold, critical core loading with the control and shutdown absorbers completely withdrawn at 20 deg. C and a helium pressure of 3.0 MPa. The detailed MCNP4B model of the reactor included the double-heterogeneity of the CFPs and the graphite spheres, and an explicit representation of the graphite reflector. A body-centered cubic lattice was used to approximate the packing of spheres in the core, with the size of the moderator sphere reduced in a manner that

  4. Numerical analysis of the burn-through at in-service welding of 316 stainless steel pipeline

    International Nuclear Information System (INIS)

    The purpose of this study is to develop an appropriate numerical model to predict the onset of the failure of a pipeline-wall during an in-service welding process. Therefore, the thermo-mechanical stresses as well as the temperature across the pipe wall have been obtained and the former have been compared against the temperature-dependent yield stress of the material. The results show that this is a more accurate criterion in order to check the burn-through risk and when the effective stress at two thirds of the pipe wall thickness is larger than the yield stress at the associated temperature, there is a risk of burn-through. The results show that burn-through may occur under the welding pool, and it is more likely to happen in the primary passes of the welding. -- Highlights: • To evaluate the risk of burn-through, the critical level of 980 °C is not only sufficient. • We find that the mechanical stresses also play a major role in the burn-through. • The effective stress should be compared against the temperature-dependent yield stress. • The geometry of welding passes has a major role in the occurrence of burn-through. • The risk of burn-through is high during the first pass than the second pass

  5. The evolution of the epidemic of charcoal-burning suicide in Taiwan: a spatial and temporal analysis.

    Directory of Open Access Journals (Sweden)

    Shu-Sen Chang

    2010-01-01

    Full Text Available BACKGROUND: An epidemic of carbon monoxide poisoning suicide by burning barbecue charcoal has occurred in East Asia in the last decade. We investigated the spatial and temporal evolution of the epidemic to assess its impact on the epidemiology of suicide in Taiwan. METHODS AND FINDINGS: Age-standardised rates of suicide and undetermined death by charcoal burning were mapped across townships (median population aged 15 y or over = 27,000 in Taiwan for the periods 1999-2001, 2002-2004, and 2005-2007. Smoothed standardised mortality ratios of charcoal-burning and non-charcoal-burning suicide and undetermined death across townships were estimated using Bayesian hierarchical models. Trends in overall and method-specific rates were compared between urban and rural areas for the period 1991-2007. The epidemic of charcoal-burning suicide in Taiwan emerged more prominently in urban than rural areas, without a single point of origin, and rates of charcoal-burning suicide remained highest in the metropolitan regions throughout the epidemic. The rural excess in overall suicide rates prior to 1998 diminished as rates of charcoal-burning suicide increased to a greater extent in urban than rural areas. CONCLUSIONS: The charcoal-burning epidemic has altered the geography of suicide in Taiwan. The observed pattern and its changes in the past decade suggest that widespread media coverage of this suicide method and easy access to barbecue charcoal may have contributed to the epidemic. Prevention strategies targeted at these factors, such as introducing and enforcing guidelines on media reporting and restricting access to charcoal, may help tackle the increase of charcoal-burning suicides. Please see later in the article for the Editors' Summary.

  6. Monju core physics test analysis with various nuclear data libraries

    International Nuclear Information System (INIS)

    JAEA has been re-analyzing Monju core physics tests to validate the JAEA's neutronics calculation system to be used in the next Monju core physics tests. Precedent results presented in PHYSOR2008 have demonstrated the validity of the system based on the basic physical parameters, such as criticality, control rod worth, isothermal temperature coefficient, and power coefficient. This paper is a continuation of the validation study focusing on the other parameters, such as fixed absorber reactivity worth, fuel sub-assembly reactivity worth, coolant reactivity worth, burnup coefficient, and reaction rate. The fixed absorber reactivity worth is a reactivity induced by the replacement of a blanket sub-assembly to a fixed absorber sub-assembly. The fuel sub-assembly reactivity worth is a reactivity induced by the replacement of a fuel sub-assembly to a non-fissile dummy sub-assembly. The coolant reactivity worth is a reactivity induced by the replacement of a non- fissile dummy sub-assembly containing sodium to that containing helium. The reaction rate data include the reaction rate ratio of 238U capture to 239Pu fission. Each of the data is useful to check the calculation system in a particular aspect. For example, the first two data are suitable to check the calculation accuracy of a blanket region and a fuel sub-assembly, respectively. The parameters are simulated using the JAEA's neutronics calculation system with various nuclear date libraries, JENDL-3.2, JENDL-3.3, JENDL/AC-2008, JEFF-3.1, and ENDF/B-VII. A continuous energy Monte Carlo calculation code, MVP, is employed to check calculation methods. Figure 1 shows an example of the C/E (Calculation over Experiment) values. The C/E values are within experimental errors for the fixed absorber reactivity worth and the fuel sub- assembly reactivity worth. Those for the burnup reactivity coefficient are around the experimental error and show a tendency of overestimation. About the comparison with the Monte Carlo

  7. Development of pin-by-pin core analysis method using three-dimensional direct response matrix

    International Nuclear Information System (INIS)

    A three-dimensional direct response matrix method using a Monte Carlo calculation has been developed. The direct response matrix is formalized by four subresponse matrices in order to respond to a core eigenvalue k and thus can be recomposed at each outer iteration in core analysis. The subresponse matrices can be evaluated by ordinary single fuel assembly calculations with the Monte Carlo method in three dimensions. Since these subresponse matrices are calculated for the actual geometry of the fuel assembly, the effects of intra- and inter-assembly heterogeneities can be reflected on global partial neutron current balance calculations in core analysis. To verify this method, calculations for heterogeneous systems were performed. The results obtained using this method agreed well with those obtained using direct calculations with a Monte Carlo method. This means that this method accurately reflects the effects of intra- and inter-assembly heterogeneities and can be used for core analysis. A core analysis method, in which neutronic calculations using this direct response matrix method are coupled with thermal-hydraulic calculations, has also been developed. As it requires neither diffusion approximation nor a homogenization process of lattice constants, a precise representation of the effects of neutronic heterogeneities is possible. Moreover, the fuel rod power distribution can be directly evaluated, which enables accurate evaluations of core thermal margins. A method of reconstructing the response matrices according to the condition of each node in the core has been developed. The test revealed that the neutron multiplication factors and the fuel rod neutron production rates could be reproduced by interpolating the elements of the response matrix. A coupled analysis of neutronic calculations using the direct response matrix method and thermal-hydraulic calculations for an ABWR quarter core was performed, and it was found that the thermal power and coolant

  8. Criticality Benchmark Analysis of the HTTR Annular Startup Core Configurations

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2009-11-01

    One of the high priority benchmarking activities for corroborating the Next Generation Nuclear Plant (NGNP) Project and Very High Temperature Reactor (VHTR) Program is evaluation of Japan's existing High Temperature Engineering Test Reactor (HTTR). The HTTR is a 30 MWt engineering test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. A large amount of critical reactor physics data is available for validation efforts of High Temperature Gas-cooled Reactors (HTGRs). Previous international reactor physics benchmarking activities provided a collation of mixed results that inaccurately predicted actual experimental performance.1 Reevaluations were performed by the Japanese to reduce the discrepancy between actual and computationally-determined critical configurations.2-3 Current efforts at the Idaho National Laboratory (INL) involve development of reactor physics benchmark models in conjunction with the International Reactor Physics Experiment Evaluation Project (IRPhEP) for use with verification and validation methods in the VHTR Program. Annular cores demonstrate inherent safety characteristics that are of interest in developing future HTGRs.

  9. Criticality Benchmark Analysis of the HTTR Annular Startup Core Configurations

    International Nuclear Information System (INIS)

    One of the high priority benchmarking activities for corroborating the Next Generation Nuclear Plant (NGNP) Project and Very High Temperature Reactor (VHTR) Program is evaluation of Japan's existing High Temperature Engineering Test Reactor (HTTR). The HTTR is a 30 MWt engineering test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. A large amount of critical reactor physics data is available for validation efforts of High Temperature Gas-cooled Reactors (HTGRs). Previous international reactor physics benchmarking activities provided a collation of mixed results that inaccurately predicted actual experimental performance.1 Reevaluations were performed by the Japanese to reduce the discrepancy between actual and computationally-determined critical configurations.2-3 Current efforts at the Idaho National Laboratory (INL) involve development of reactor physics benchmark models in conjunction with the International Reactor Physics Experiment Evaluation Project (IRPhEP) for use with verification and validation methods in the VHTR Program. Annular cores demonstrate inherent safety characteristics that are of interest in developing future HTGRs.

  10. Thermal buckling analysis of truss-core sandwich plates

    Institute of Scientific and Technical Information of China (English)

    陈继伟; 刘咏泉; 刘伟; 苏先樾

    2013-01-01

    Truss-core sandwich plates have received much attention in virtue of the high values of strength-to-weight and stiffness-to-weight as well as the great ability of impulse-resistance recently. It is necessary to study the stability of sandwich panels under the influence of the thermal load. However, the sandwich plates are such complex three-dimensional (3D) systems that direct analytical solutions do not exist, and the finite element method (FEM) cannot represent the relationship between structural parameters and mechanical properties well. In this paper, an equivalent homogeneous continuous plate is idealized by obtaining the effective bending and transverse shear stiffness based on the characteristics of periodically distributed unit cells. The first order shear deformation theory for plates is used to derive the stability equation. The buckling temperature of a simply supported sandwich plate is given and verified by the FEM. The effect of related parameters on mechanical properties is investigated. The geometric parameters of the unit cell are optimized to attain the maximum buckling temperature. It is shown that the optimized sandwich plate can improve the resistance to thermal buckling significantly.

  11. Development of core thermal hydraulic analysis methodology using multichannel code system

    International Nuclear Information System (INIS)

    A multi-channel core analysis model using a subchannel code TORC is developed to improve the thermal margin, and is assessed and compared with the existing single-channel analysis model. To apply the TORC code to the w-type reactor core, a hot subchannel DNBR analysis model is developed using the lumping technology. In addition, the sensitivity of TORC to various models and input parameters are carried out to appreciate the code characteristics. The developed core analysis model is applied to the evaluation of the thermal margin for 17 x 17 KOFA loaded core. For this calculation, the KRB1 CHF correlation is developed on the basis of w and Siemens bundle CHF data, and the DNB design limit is established using the STDP method. From the result of the steady-state and transient analysis of the 17 x 17 KOFA loaded core, it is found that the extra 10% DNBR margin can be obtained compared with the existing single-channel analysis methodology. (Author) 65 figs., 12 tabs

  12. Extracting the core indicators of pulverized coal for blast furnace injection based on principal component analysis

    Science.gov (United States)

    Guo, Hong-wei; Su, Bu-xin; Zhang, Jian-liang; Zhu, Meng-yi; Chang, Jian

    2013-03-01

    An updated approach to refining the core indicators of pulverized coal used for blast furnace injection based on principal component analysis is proposed in view of the disadvantages of the existing performance indicator system of pulverized coal used in blast furnaces. This presented method takes into account all the performance indicators of pulverized coal injection, including calorific value, igniting point, combustibility, reactivity, flowability, grindability, etc. Four core indicators of pulverized coal injection are selected and studied by using principal component analysis, namely, comprehensive combustibility, comprehensive reactivity, comprehensive flowability, and comprehensive grindability. The newly established core index system is not only beneficial to narrowing down current evaluation indices but also effective to avoid previous overlapping problems among indicators by mutually independent index design. Furthermore, a comprehensive property indicator is introduced on the basis of the four core indicators, and the injection properties of pulverized coal can be overall evaluated.

  13. A SAS2H/KENO-V Methodology for 3D Full Core depletion analysis

    International Nuclear Information System (INIS)

    This paper describes the use of a SAS2H/KENO-V methodology for 3D full core depletion analysis and illustrates its capabilities by applying it to burnup analysis of the IRIS core benchmarks. This new SAS2H/KENO-V sequence combines a 3D Monte Carlo full core calculation of node power distribution and a 1D Wigner-Seitz equivalent cell transport method for independent depletion calculation of each of the nodes. This approach reduces by more than an order of magnitude the time required for getting comparable results using the MOCUP code system. The SAS2H/KENO-V results for the asymmetric IRIS core benchmark are in good agreement with the results of the ALPHA/PHOENIX/ANC code system. (author)

  14. Development of Computer Program for Whole Core Thermal-Hydraulic Analysis of Fast Reactors

    International Nuclear Information System (INIS)

    A whole core thermal-hydraulic analysis program ACT was developed for the purpose of evaluating detailed in-core thermal-hydraulic phenomena of sodium cooled fast reactors under various reactor operation conditions. ACT consists of four kinds of calculation modules, i.e., fuel-assembly, inter-wrapper gap (core barrel), upper plenum and heat transport system modules. The latter two modules give proper boundary conditions for the reactor core thermal-hydraulic analysis. These four modules are coupled with each other by using MPI and calculate simultaneously on a cluster workstation. ACT was applied to analyzing a sodium experiment performed at JNC, which simulated the natural circulation decay heat removal under PRACS and DRACS operation condition. In the experiment, not only inter-wrapper flows but also reverses flows in the fuel assemblies were observed. ACT succeeded in simulating such complicated phenomena. (authors)

  15. Steady state analysis of SFR cores using DYN3D-Serpent codes sequence

    International Nuclear Information System (INIS)

    A few-group cross section generation methodology for the deterministic analysis of SFR cores with DYN3D code has been proposed. The full core DYN3D results obtained using the few-group constants produced by Serpent agreed very well with that of the reference full core MC simulations. Such an agreement demonstrates the feasibility of the proposed few-group cross section generation procedure. In summary, this study showed that the Serpent-DYN3D code sequence can be successfully used for modeling fast spectrum reactor systems. (orig.)

  16. European contribution to Phase 3 of the benchmark core analysis for the BN-600 hybrid reactor

    International Nuclear Information System (INIS)

    This European participation in Phase 3 of the benchmark (BN-600) analysis consist of a joint contribution from France and the UK. Calculations were performed by ERANOS code and data system which has been developed in the framework of European cooperation on fast reactors. Results are presented for all the core neutronic parameters, both for homogeneous and heterogeneous core models and both for beginning and end of fuel cycle

  17. Design of the core of a breed/burn fast reactor with the deterministic code KANEXT; Diseno del nucleo de un reactor rapido de cria/quemado con el codigo deterministico KANEXT

    Energy Technology Data Exchange (ETDEWEB)

    Lopez S, R. C.; Francois L, J. L., E-mail: rcarlos.lope@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    The breeding fast reactors are interesting because they generate more plutonium than they consume, however, the fuel has to be reprocessed for the generated plutonium is used in another reactor. In a breed/burn reactor (BBR) the plutonium is generated and used -in situ- inside the same reactor, reducing this way costs and the proliferation possibility. In this work, the core of a BBR was designed; cooled by sodium that consists of 210 active assemblies and 7 spaces for control rods, each assembly consists of 169 pines. The design differs from other BBR it includes a blanket in the reactor center. The above-mentioned was to take advantage of the fact by geometry that the population of fast and epithermal neutrons will be high in the area, due to the fissions in adjacent fissile areas. Favorable results were obtained, although not definitive with exchange scheme of spent fuel. Efforts should be made in the future to homogenize the power generation within the reactor and replace the spent assemblies more efficiently. (Author)

  18. Burning Mouth Syndrome and "Burning Mouth Syndrome".

    Science.gov (United States)

    Rifkind, Jacob Bernard

    2016-03-01

    Burning mouth syndrome is distressing to both the patient and practitioner unable to determine the cause of the patient's symptoms. Burning mouth syndrome is a diagnosis of exclusion, which is used only after nutritional deficiencies, mucosal disease, fungal infections, hormonal disturbances and contact stomatitis have been ruled out. This article will explore the many causes and treatment of patients who present with a chief complaint of "my mouth burns," including symptomatic treatment for those with burning mouth syndrome. PMID:27209717

  19. Clinical characteristics analysis of 656 burn children%儿童烧伤656例临床分析

    Institute of Scientific and Technical Information of China (English)

    刘继松; 张开俊; 章祥洲; 徐东卫; 杜娟

    2013-01-01

    目的:分析儿童烧伤的特点,探讨预防或减少儿童烧伤的措施.方法:对656例烧伤患儿的性别、年龄、致伤因素、烧伤部位、烧伤程度、季节分布、地点、家庭及监护人文化程度等进行分析.结果:656例儿童烧伤中,男女比为1.66:1,1~3岁患儿为最多(383例,58.38%),以热液烫伤为最多(481例,73.33%),以头面颈部及躯干为主(50.30%、64.63%),中度为主(47.37%),多发生在夏季(42.63%),烧伤发生的地点大部分在家中(81.86%),农村患儿多于城市,家中有父母或监护人在场的占75.46%.结论:儿童烧烫伤主要是监护人的疏忽大意、照顾不周所致,提高监护人的安全意识和加强烧伤知识宣传、普及,是预防或减少儿童烧烫伤发生的有效措施.%Objective: To analyse the clinical features of child burn and explore the measures of preventing or decreasing the child burn. Methods: The data of the sex, age, factors, site and extent of injury, seasonal distribution, place, family situation and educational level of the supervisors of 656 burn children were analysed. Results:For 656 burn children,the rate of male versus female was 1. 66: 1, 1 to 3 years old children were often happened(383 cases,58. 38% ) , the most risk factors were hydrotherm(481 cases,73. 33% ). Head,face, neck and trunk were the most common burn sites (50. 30% and 64. 63% ) , moderate level of burn was most usual (47. 37% ) ,the peak season was summer(42. 63% ) ,most burns occurred at home(81. 86% ). Burn children in rural areas were more than those in urban areas and parents or guardian were on the scene (75. 46% ) . Conclusions: Child burn is principally caused by guardians carelessness and careless for child. Improving safety consciousness, strengthening the burn knowledge publicity and popularization are effective measures to prevent and decrease the child burn.

  20. Steady state thermal hydraulic analysis of LMR core using COBRA-K code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Kwang; Kim, Young Gyun; Kim Young In; Kim Young Cheol

    1997-02-01

    A thermal hydraulics analysis code COBRA-K is being developed by the KAERI LMR core design technology development team. COBRA-K is a part of the integrated computation system for LMR core design and analysis, the K-CORE system. COBRA-K is supposed to predict the flow and temperature distributions in LMR core. COBRA-K is an extension of the previously published COBRA-IV-I code with several functional improvements. Specially COBRA-K has been improved to analyze single and multi-assembly, and whole-core in the transient condition. This report describes the overall features of COBRA-K and gives general input descriptions. The 19 pin assembly experimental data of ORNL were used to verify the accuracy of this code for the steady state analysis. The comparative results show good agreements between the calculated and the measured data. And COBRA-K can be used to predict flow and temperature distributions for the LMR core design. (author). 7 refs., 6 tabs., 13 figs.

  1. Development of flow network analysis code for block type VHTR core by linear theory method

    International Nuclear Information System (INIS)

    VHTR (Very High Temperature Reactor) is high-efficiency nuclear reactor which is capable of generating hydrogen with high temperature of coolant. PMR (Prismatic Modular Reactor) type reactor consists of hexagonal prismatic fuel blocks and reflector blocks. The flow paths in the prismatic VHTR core consist of coolant holes, bypass gaps and cross gaps. Complicated flow paths are formed in the core since the coolant holes and bypass gap are connected by the cross gap. Distributed coolant was mixed in the core through the cross gap so that the flow characteristics could not be modeled as a simple parallel pipe system. It requires lot of effort and takes very long time to analyze the core flow with CFD analysis. Hence, it is important to develop the code for VHTR core flow which can predict the core flow distribution fast and accurate. In this study, steady state flow network analysis code is developed using flow network algorithm. Developed flow network analysis code was named as FLASH code and it was validated with the experimental data and CFD simulation results. (authors)

  2. Decay heat analysis of a VHTR core using the HELIOS and origen-2 codes

    International Nuclear Information System (INIS)

    This paper describes the procedure and results of a decay heat analysis in a relatively short time after a shutdown for the safety analysis of a VHTR core. In this analysis, HELIOS provides the one-group actinide cross sections to ORIGEN-2 through the 190 group lattice calculation for a single fuel block. Then, ORIGEN-2 performs the depletion and decay heat calculations using these actinide cross sections. After benchmarking this procedure against a PWR core, it was applied to a 200 MWth prismatic VHTR core. The results showed that the decay heat per unit operating power is very comparable to that for a typical large power PWR core, although the decay heat per unit heavy metal mass is three times higher than that in the PWR core. It was also found from the results that the decay heat fraction to operating power decreases very slowly with the core burnup after it reaches a maximum value of 6.1 percents at 5 GWd/tHM. (authors)

  3. Cellular burning in lean premixed turbulent hydrogen-air flames: Coupling experimental and computational analysis at the laboratory scale

    Science.gov (United States)

    Day, M. S.; Bell, J. B.; Cheng, R. K.; Tachibana, S.; Beckner, V. E.; Lijewski, M. J.

    2009-07-01

    One strategy for reducing US dependence on petroleum is to develop new combustion technologies for burning the fuel-lean mixtures of hydrogen or hydrogen-rich syngas fuels obtained from the gasification of coal and biomass. Fuel-flexible combustion systems based on lean premixed combustion have the potential for dramatically reducing pollutant emissions in transportation systems, heat and stationary power generation. However, lean premixed flames are highly susceptible to fluid-dynamical combustion instabilities making robust and reliable systems difficult to design. Low swirl burners are emerging as an important technology for meeting design requirements in terms of both reliability and emissions for next generation combustion devices. In this paper, we present simulations of a lean, premixed hydrogen flame stabilized on a laboratory-scale low swirl burner. The simulations use detailed chemistry and transport without incorporating explicit models for turbulence or turbulence/chemistry interaction. Here we discuss the overall structure of the flame and compare with experimental data. We also use the simulation data to elucidate the characteristics of the turbulent flame interaction and how this impacts the analysis of experimental measurements.

  4. Cellular burning in lean premixed turbulent hydrogen-air flames: Coupling experimental and computational analysis at the laboratory scale

    International Nuclear Information System (INIS)

    One strategy for reducing US dependence on petroleum is to develop new combustion technologies for burning the fuel-lean mixtures of hydrogen or hydrogen-rich syngas fuels obtained from the gasification of coal and biomass. Fuel-flexible combustion systems based on lean premixed combustion have the potential for dramatically reducing pollutant emissions in transportation systems, heat and stationary power generation. However, lean premixed flames are highly susceptible to fluid-dynamical combustion instabilities making robust and reliable systems difficult to design. Low swirl burners are emerging as an important technology for meeting design requirements in terms of both reliability and emissions for next generation combustion devices. In this paper, we present simulations of a lean, premixed hydrogen flame stabilized on a laboratory-scale low swirl burner. The simulations use detailed chemistry and transport without incorporating explicit models for turbulence or turbulence/chemistry interaction. Here we discuss the overall structure of the flame and compare with experimental data. We also use the simulation data to elucidate the characteristics of the turbulent flame interaction and how this impacts the analysis of experimental measurements.

  5. Transient and stability analysis of a BWR core with thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779 Col. Narvarte, 03020 Mexico, DF (Mexico); Espinosa-Paredes, Gilberto [Division de Ciencias Basicas e Ingenieria, Universidad Autonoma Metropolitana, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, DF (Mexico)], E-mail: gepe@xanum.uam.mx; Francois, Juan-Luis [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec Mor. (Mexico)

    2008-08-15

    The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket-seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal-hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO{sub 2} core, even during transient conditions. The stability and transient analysis show that the thorium-uranium fuel can be operated safely in current BWRs.

  6. HORECA. Hoger onderwijs reactor elementary core analysis system. User's manual

    International Nuclear Information System (INIS)

    HORECA is developed at IRI Delft for quick analysis of power distribution, burnup and safety for the HOR. It can be used for the manual search of a better loading of the reactor. HORECA is based on the Penn State Fuel Management Package and uses the MCRAC code included in this package as a calculation engine. (orig./HP)

  7. SMAD-2A mass spectrometer core isotopic analysis of water

    International Nuclear Information System (INIS)

    SMAT-2A is a small mass spectrometer for water isotopes analysis, double collector method (HD/H2 ratio). Improvements in maintenance automation data acquisition and processing are presented. Analyses of liquid and gaseous samples were made and internal reproducibility and precision are calculated. (Author)

  8. Analysis of void coefficient in fast spectrum BWR core with Monte Carlo code 'MVP'

    International Nuclear Information System (INIS)

    An innovative large BWR core concept has been proposed for aiming at fuel breeding as well as negative void reactivity coefficient. The core consists of two types of MOX fuel assemblies. One is a triangular tight lattice bundle 1.6 m in active core height and the other is the same bundle 0.8 m. The ratio of flow area to fuel area of the bundle is set at about 0.5 in order to increase breeding ratio. A neutron-streaming channel that consists of a cavity-can containing helium gas and a flow gap between the cavity-can and the channel box is located above each short bundle. It will decrease void reactivity coefficient by enhancing neutron leakage from the core when the void fraction is increased in the flow gap. A core composed of tight lattice bundles provides a much harder neutron spectrum than that of conventional BWRs but a slightly softer one than that of typical FBRs. The cavity-can and the flow gap will cause a steep gradient of neutron flux. The neutronics for such a complicated core structure could not be properly analyzed by conventional analysis methods. In particular, the analysis of void reactivity coefficient requires a sophisticated method because it deals with a small change in core composition. In the analysis of the void reactivity coefficient, we adopted a three-dimensional Monte Carlo code 'MVP', which has been developed by JAERI and has many advantages such as an easy input form for lattice structures, a short run time and a continuous neutron energy method. The continuous neutron energy method is important for the analysis of this core because fission reactions occur mainly in the resonance energy region, where the evaluation of accurate cross sections is difficult with conventional methods. The library used is JENDL-3.2. The multi-layer structure of lattices is also essential for the analysis because its hard spectrum and relatively long neutron mean free path require a modeling for the full core with a lot of bundles. The analysis indicates that

  9. Sensitivity analysis of core parameters to fuel manufacturing uncertainties

    International Nuclear Information System (INIS)

    Heterogeneities cannot be avoided at the manufacturing stage of nuclear fuel. In the project, control parameters are always best estimate computed, so that heterogeneities in the fuel affect them as an uncertainty. These heterogeneities cannot be only represented by a mean value and a standard deviation: their joint distribution must be evaluated according to post-manufacture statistical analysis. In order to improve the nuclear fuel performances and respect the safety rules, FRAMATOME has developed an original analysis method based on stochastic approach and the generalized perturbation theory. The joint low of the pointwise power variation is given by the sensitivity matrix, determined by the generalized perturbation theory, using the first order Taylor stochastic development. (authors). 8 refs., 5 figs

  10. Analysis of core samples from the BPXA-DOE-USGS Mount Elbert gas hydrate stratigraphic test well: Insights into core disturbance and handling

    Energy Technology Data Exchange (ETDEWEB)

    Kneafsey, Timothy J.; Lu, Hailong; Winters, William; Boswell, Ray; Hunter, Robert; Collett, Timothy S.

    2009-09-01

    Collecting and preserving undamaged core samples containing gas hydrates from depth is difficult because of the pressure and temperature changes encountered upon retrieval. Hydrate-bearing core samples were collected at the BPXA-DOE-USGS Mount Elbert Gas Hydrate Stratigraphic Test Well in February 2007. Coring was performed while using a custom oil-based drilling mud, and the cores were retrieved by a wireline. The samples were characterized and subsampled at the surface under ambient winter arctic conditions. Samples thought to be hydrate bearing were preserved either by immersion in liquid nitrogen (LN), or by storage under methane pressure at ambient arctic conditions, and later depressurized and immersed in LN. Eleven core samples from hydrate-bearing zones were scanned using x-ray computed tomography to examine core structure and homogeneity. Features observed include radial fractures, spalling-type fractures, and reduced density near the periphery. These features were induced during sample collection, handling, and preservation. Isotopic analysis of the methane from hydrate in an initially LN-preserved core and a pressure-preserved core indicate that secondary hydrate formation occurred throughout the pressurized core, whereas none occurred in the LN-preserved core, however no hydrate was found near the periphery of the LN-preserved core. To replicate some aspects of the preservation methods, natural and laboratory-made saturated porous media samples were frozen in a variety of ways, with radial fractures observed in some LN-frozen sands, and needle-like ice crystals forming in slowly frozen clay-rich sediments. Suggestions for hydrate-bearing core preservation are presented.

  11. RAVE code system for 3-D core non-LOCA accident analysis

    International Nuclear Information System (INIS)

    Full text of publication follows: This paper provides an overview of the application of the Westinghouse updated RAVE three dimensional (3-D) core transient analysis code system for PWR non-LOCA accident analysis. The RAVE code system consists of a linkage of the following USNRC-approved codes: the EPRI RETRAN-02 (RETRAN) system transient analysis code, the Westinghouse SPNOVA (also referred to as ANC-K) reactor core neutron kinetic nodal code, and the EPRI VIPRE-01 (VIPRE) reactor core thermal-hydraulic (T/H) code. The RETRAN code is used for calculating transient conditions in the reactor coolant system (RCS), including reactor vessel, RCS loops, pressurizer and steam generators. RETRAN also models reactor trips, engineering safety feature (ESF) functions, and the control systems. The SPNOVA code is used to perform 3-D core neutronic calculations for core average power and power distributions in the core. Its reactivity feedback calculation is based on transient fluid conditions and fuel temperatures obtained from the VIPRE code. Based on core inlet temperature, inlet flow and core exit pressure from RETRAN, and the nodal nuclear power from SPNOVA, VIPRE provides back to RETRAN transient nodal heat flux in the reactor core region. An effective 3-D analysis requires RETRAN, SPNOVA and VIPRE calculations to be closely linked for the entire reactor core. The linking architecture uses a standard external communication interface protocol for communication among the running programs on the same or different computers. The RAVE code system currently uses the Parallel Virtual Machine (PVM) software for the data transfer. Besides the necessary changes for data transfer, no other changes were made to RETRAN, SPNOVA or VIPRE fundamental code algorithms or solution methods. The RETRAN model in the RAVE system uses the same detailed reactor vessel, RCS loops, pressurizer, and steam generator, and control and protection models as has been licensed for current plant Safety

  12. Verification of JUPITER standard analysis method for upgrading Joyo MK-III core design and management

    International Nuclear Information System (INIS)

    In the experimental fast reactor Joyo, loading of irradiation test rigs causes a decrease in excess reactivity because the rigs contain less fissile materials than the driver fuel. In order to carry out duty operation cycles using as many irradiation rigs as possible, it is necessary to upgrade the core performance to increase its excess reactivity and irradiation capacity. Core modification plans have been considered, such as the installation of advanced radial reflectors and reduction of the number of control rods. To implement such core modifications, it is first necessary to improve the prediction accuracy in core design and to optimize safety margins. In the present study, verification of the JUPITER fast reactor standard analysis method was conducted through a comparison between the calculated and the measured Joyo MK-III core characteristics, and it was concluded that the accuracy for a small sodium-cooled fast reactor with a hard neutron spectrum was within 5% of unity. It was shown that, the performance of the irradiation bed core could be upgraded by the improvement of the prediction accuracy of the core characteristics and optimization of safety margins. (author)

  13. Preliminary analysis of a large 1600 MWe PWR core loaded with 30% MOX fuel

    International Nuclear Information System (INIS)

    The paper presents a full-core 3-D analysis of the performances of a large 1600 MWe PWR core, loaded with 30% MOX fuel, in accordance with the European Utility Requirements (EUR). These requirements state that the European next generation power plants have to be designed capable to use MOX (UO2 - PuO2) fuel assemblies up to 50% of the core, together with UO2 fuel assemblies. The use of MOX assemblies has a significant impact on key physic parameters and on safety. A lot of studies have been carried out in the past to explore the feasibility of plutonium recycling strategies by loading LWR reactors with MOX fuel. Many of these works were based on lattice codes, in order to perform detailed analyses of the neutronic characteristics of MOX assemblies. With the aim to take into account their interaction with surrounding UO2 fuel elements, and the global effects on the core at operational conditions, an integrated approach making use of a 3-D core simulation is required. In this light, the present study adopts the state-of-art numerical models CASMO-5 and SIMULATE-3 to analyze the behavior of the core fueled with 30% MOX and to compare it with that of a large PWR reference core, fueled with UO2. (author)

  14. Calculation analysis of TRIGA MARK II reactor core composed of two types of fuel elements

    International Nuclear Information System (INIS)

    The most important properties of mixed cores are treated for TRIGA MARK II reactor, composed of standard (20% enriched, 8.5w% U content) and FLIP (70% enriched, 8.5w% U content) fuel elements. Large difference in enrichment and presence of burnable poison in FLIP fuel have strong influence on the main core characteristics, such as: fuel temperature coefficient, power defect, Xe and Sm worth, power and flux distributions, etc. They are significantly different for both types of fuel. Optimal loading of mixed cores therefore strongly depends on the loading pattern of both types of fuel elements. Results of systematic calculational analysis of mixed cores are presented. Calculations on the level of fuel element are performed with WIMSD-4 computer code with extended cross-section library. Core calculations are performed with TRIGAP two-group 1-D diffusion code. Results are compared to measurements and physical explanation is provided. Special concern is devoted to realistic mixed cores, for which optimal in-core fuel management is derived. Refs, figs and tabs

  15. Analysis of ringing due to magnetic core materials used in pulsed nuclear magnetic resonance applications

    Science.gov (United States)

    Prabhu Gaunkar, Neelam; Nlebedim, Cajetan; Hadimani, Ravi; Bulu, Irfan; Song, Yi-Qiao; Mina, Mani; Jiles, David

    Oil-field well logging instruments employ pulsed nuclear magnetic resonance (NMR) techniques and use inductive sensors to detect and evaluate the presence of particular fluids in geological formations. Acting as both signal transmitters and receivers most inductive sensors employ magnetic cores to enhance the quality and amplitude of signals recorded during field measurements. It is observed that the magnetic core also responds to the applied input signal thereby generating a signal (`ringing') that interferes with the measurement of the signals from the target formations. This causes significant noise and receiver dead time and it is beneficial to eliminate/suppress the signals received from the magnetic core. In this work a detailed analysis of the magnetic core response and in particular loading of the sensor due to the presence of the magnetic core is presented. Pulsed NMR measurements over a frequency band of 100 kHz to 1MHz are used to determine the amplitude and linewidth of the signals acquired from different magnetic core materials. A lower signal amplitude and a higher linewidth are vital since these would correspond to minimal contributions from the magnetic core to the inductive sensor response and thus leading to minimized receiver dead time.

  16. The Burning Saints

    DEFF Research Database (Denmark)

    Xygalatas, Dimitris

    The Anastenaria are Orthodox Christians in Northern Greece who observe a unique annual ritual cycle focused on two festivals, dedicated to Saint Constantine and Saint Helen. The festivals involve processions, music, dancing, animal sacrifices, and culminate in an electrifying fire-walking ritual....... Carrying the sacred icons of the saints, participants dance over hot coals as the saint moves them. The Burning Saints presents an analysis of these rituals and the psychology behind them. Based on long-term fieldwork, The Burning Saints traces the historical development and sociocultural context of the...... Greek fire-walking rituals. As a cognitive ethnography, the book aims to identify the social, psychological and neurobiological factors which may be involved and to explore the role of emotional and physiological arousal in the performance of such ritual. A study of participation, experience and meaning...

  17. LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Shengcheng; Wu, Hongchun; Cao, Liangzhi; Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn; Huang, Kai; He, Mingtao; Li, Xunzhao

    2014-10-15

    Highlights: • A new code system for design studies of accelerator driven subcritical reactors (ADSRs) is developed. • S{sub N} transport solver in triangular-z meshes, fine deletion analysis and multi-channel thermal-hydraulics analysis are coupled in the code. • Numerical results indicate that the code is reliable and efficient for design studies of ADSRs. - Abstract: Accelerator driven subcritical reactors (ADSRs) have been proposed and widely investigated for the transmutation of transuranics (TRUs). ADSRs have several special characteristics, such as the subcritical core driven by spallation neutrons, anisotropic neutron flux distribution and complex geometry etc. These bring up requirements for development or extension of analysis codes to perform design studies. A code system named LAVENDER has been developed in this paper. It couples the modules for spallation target simulation and subcritical core analysis. The neutron transport-depletion calculation scheme is used based on the homogenized cross section from assembly calculations. A three-dimensional S{sub N} nodal transport code based on triangular-z meshes is employed and a multi-channel thermal-hydraulics analysis model is integrated. In the depletion calculation, the evolution of isotopic composition in the core is evaluated using the transmutation trajectory analysis algorithm (TTA) and fine depletion chains. The new code is verified by several benchmarks and code-to-code comparisons. Numerical results indicate that LAVENDER is reliable and efficient to be applied for the steady-state analysis and reactor core design of ADSRs.

  18. A probabilistic SSYST-3 analysis for a PWR-core during a large break LOCA

    International Nuclear Information System (INIS)

    This report demonstrates the SSYST-3 analysis and application for a German PWR of 1300 MW. The report is concerned with the probabilistic analysis of a PWR core during a loss-of-coolant accident due to a large break. With the probabilistic analysis, the distribution functions of the maximum temperatures and cladding elongations occuring in the core can be calculated. Parameters like rod power, the thermohydraulic boundary conditions, stored energy in the fuel rods and the heat transfer coefficient were found to be the most important. The expected value of core damage was determined to be 2.9% on the base of response surfaces for cladding temperature and strain deduced from SSYST-3 single rod results. (orig./HP)

  19. Safety analysis for operating the Annular Core Research Reactor with the central cavity liner removed

    International Nuclear Information System (INIS)

    Isotope production in the Annular Core Research Reactor requires highly enriched uranium targets to be irradiated in the high flux central region of the core. In order to accomplish this goal, the central cavity liner has been removed to allow for the eventual placement of targets in that region. This safety evaluation presents the analysis associated with operating the reactor in the steady state mode with the central cavity liner removed and the central region of the core filled with water and aluminum void targets. The reactor operation with enriched, uranium loaded targets will be analyzed in a future analysis document. This analysis describes only the operation of the reactor in the steady state mode; consideration of pulse mode operations with the liner removed is not presented

  20. Neutronic Analysis of HTTR Core Using DeCART Code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Lee, Hyun Chul; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The neutronic analysis for the High Temperature Engineering Test Reactor (HTTR) has been performed. The HTTR is a graphite-moderated and helium gas cooled reactor with an outlet temperature of 950 .deg. C and thermal output of 30 MW. It has been studied as one of the Generation-IV (Gen-IV) reactor. In this study, the DECART code is used with a 190-group KARMA library. The calculation results are compared with those of the McCARD with ENDFB/VII.0 library.

  1. Neutronic Analysis of HTTR Core Using DeCART Code

    International Nuclear Information System (INIS)

    The neutronic analysis for the High Temperature Engineering Test Reactor (HTTR) has been performed. The HTTR is a graphite-moderated and helium gas cooled reactor with an outlet temperature of 950 .deg. C and thermal output of 30 MW. It has been studied as one of the Generation-IV (Gen-IV) reactor. In this study, the DECART code is used with a 190-group KARMA library. The calculation results are compared with those of the McCARD with ENDFB/VII.0 library

  2. Prompt Gamma Activation Analysis of the Nyírlugos obsidian core depot find

    Directory of Open Access Journals (Sweden)

    Zsolt Kasztovszky

    2014-03-01

    Full Text Available The Nyírlugos obsidian core depot find is one of the most important lithic assemblages in the collection of the Hungarian National Museum (HNM. The original set comprised 12 giant obsidian cores, of which 11 are currently on the permanent archaeological exhibition of the HNM. One of the cores is known to be inDebrecen. The first publication attributed the hoard, on the strength of giant (flint blades known from the Early and Middle Copper Age Tiszapolgár and Bodrogkeresztúr cultures, to the Copper Age. In the light of recent finds it is more likely to belong to the Middle Neolithic period. The source area was defined as Tokaj Mts., about100 kmto the NW from Nyírlugos. The size and beauty of the exceptional pieces exclude any invasive analysis. Using Prompt Gamma Activation Analysis (PGAA, we can measure major chemical components and some key trace elements of stone artefacts with adequate accuracy to successfully determine provenance of obsidian. Recent methodological development also facilitated the study of relatively large objects like the Nyírlugos cores. The cores were individually measured by PGAA. The results show that the cores originate from the Carpathian 1 sources, most probably the Viničky variety (C1b. The study of the hoard as a batch is an important contribution to the assessment of prehistoric trade and allows us to reconsider the so-called Carpathian, especially Carpathian 1 (Slovakian sources.

  3. Core and Conal Component Analysis of Pulsar B1933+16 --- Investigation of the Segregated Modes

    CERN Document Server

    Mitra, Dipanjan; Arjunwadkar, Mihir

    2016-01-01

    Radio pulsar B1933+16 is brightest core-radiation dominated pulsar in the Arecibo sky, and here we carry out a comprehensive high resolution polarimetric study of its radiation at both 1.5 and 4.6 GHz. At 1.5 GHz, the polarization is largely compatible with a rotating-vector model with $\\alpha$ and $\\beta$ values of 125 and --1.2$^{\\circ}$, such that the core and conal regions can be identified with the primary and secondary polarization modes and plausibly with the extraordinary and ordinary propagation modes. Polarization modal segregation of profiles shows that the core is comprised of two parts which we associate with later X-mode and earlier O-mode emission. Analysis of the broad microstructures under the core shows that they have similar timescales to those of the largely conal radiation of other pulsars studied earlier. Aberration/retardation analysis was here possible for both the conal and core radiation and showed average physical emission heights of about 200 km for each. Comparison with other core...

  4. A model of two-stage core calculation method coupled with subchannel analysis for boiling water reactors

    International Nuclear Information System (INIS)

    The two-stage core analysis method is widely used for BWR core analysis. The purpose of this study is to develop the core analysis model coupled with subchannel analysis within the two-stage calculation scheme using an assembly-based thermal-hydraulics calculation in the core analysis. This model does not change the thermal-hydraulics scheme of the core analysis. Rather, it appends the subchannel void distribution to the previous uniform analysis in lattice physics, and couples that with the subchannel analysis which axially calculates full assembly and uses the flow condition that produces the maximum void fraction in the operation core. The subchannel void distribution of one node from the subchannel analysis is only normalized and used for the lattice physics. The developed model was evaluated for the heterogeneous problem with multiple enrichments. The developed model could decrease the eigenvalue differences by more than half of that of the uniform case and made the differences of assembly power the same as the uniform case. Furthermore, it could reduce the root mean square differences to more than half of those of the uniform case in the low and high enrichment fuels. The computation times of the lattice physics become 2.3 times longer. The extended computing time does not prevent core analysis because the nuclear data are prepared in advance of the core analysis. As the result of the evaluation, the model can incorporate the subchannel effect to the core analysis. (author)

  5. Burn Injuries: Burn Depth, Physiopathology and Type of Burns

    OpenAIRE

    Kemalettin Koltka

    2011-01-01

    A significant burn injury is a serious and mortal event. The most important threat to life is hypovolemic shock with complex pathophysiologic mechanisms. Burn depth is classified as first, second, or third degree. Local inflammatory response results a vasodilatation and an increase in vascular permeability. A burn injury is a three dimensional ischemic wound. Zone of coagulation is the zone with maximum damage. Zone of stasis consists of damaged but viable tissues, the tissue is salvageable. ...

  6. Dynamical analysis of machining tool body with reinforced inner core of circular shape

    Directory of Open Access Journals (Sweden)

    Naď M.

    2009-06-01

    Full Text Available The vibration analysis of a clamped beam structure representing vibrating machining body tool is solved in this paper. The required modal properties of beam are obtained by application the reinforcing core with circular cross section. The perfect adhesion between core material and basic beam material is considered. The different material properties of beam and core are considered. The fundamental mathematical formulations describing the bending vibrations of this composite beam structure are presented. The effect of material properties and geometrical parameters of reinforcing core on natural frequencies of cantilever composite beam structure with circular and rectangular cross section is presented. This form of composite beam structure provides effective tool to modification of its dynamical properties.

  7. Zero-dimensional analysis of burn control with compression-decompression

    International Nuclear Information System (INIS)

    A simple but realistic method of active feedback control of a self-ignited tokamak reactor by major radius compression-decompression is presented, in which the thermal fusion output power and the ion temperature are envisaged as the variables that could be adopted as object of control. Error inherent in the measuring system for the controlled variable is taken into account in the analysis. Numerical calculations based on a point reactor model are performed, which indicate that the proposed method of control is capable of completely suppressing thermal runaway without involving any significant change in the major radius or in the fusion output power. Matching the reactor output to changes in load is shown also to be possible from a numerical example. The thermal stability of an igniting plasma governed by this feedback control is analyzed without prescribing any particular transport scaling. It is revealed that the control is achievable for a suitably chosen gain even with a response time lag longer than the thermal runaway time. Strong dependence of the stability on the scaling law is indicated. (author)

  8. Noise analysis of Forsmark 1 data to investigate BWR core local instability

    International Nuclear Information System (INIS)

    BWR core local instability was experienced at Forsmark 1 (F1) during reactor operation in cycle 16. The event has been studied by applying noise analysis and stability calculations to get insight into the event as well as to identify the cause of local instability. The present report is concerned with noise analysis of data collected during start-up in cycle 17. The results of the current study indicates: The F1 core is quite stable in cycle 17. The max. decay ratio (DR) value of 0.37 was obtained from the stability evaluation of an APRM (average power range monitor) and LPRM (local power range monitor) signals measured at 66% (APRM) of reactor power and 4252 Kg/s (SA-HC) of core flow. Compared with the power profile in cycle 17 (as well as in reactor F2), the core in cycle 16 had an extreme power profile with high power and bottom-shifted axial peak in the core periphery esp. at the four quadrant corners. Such a profile decreases the stability margin in the region. It is a common observation that the DR obtained from APRM tends to be higher than that from LPRM if the global instability mechanism is dominant in the core, and vice versa. The comparison of global and local DR values should be an effective method for detecting local instability during the reactor operation. In order to detect the local instability it is important to evaluate the core stability with sufficient number of LPRMs so as to cover the whole core cross section together with APRMs

  9. Tropospheric O3 over Indonesia during biomass burning events measured with GOME (Global Ozone Monitoring Experiment) and compared with trajectory analysis

    OpenAIRE

    Wittrock, F.; Richter, A; J. Meyer-Arnek; Ladstätter-Weißenmayer, A.; Burrows, J.P.

    2005-01-01

    Tropospheric ozone columns of up to 50 DU were observed by GOME (Global Ozone Monitoring Experiment) above Indonesia in September 1997, while only background amounts were measured in September 1998. The Traj.x trajectory model along with BRemen's Atmospheric PHOtochemical model (BRAPHO) were used to investigate the higher than average ozone columns above Indonesia. The transport analysis reveals that biomass burning over central Africa and northern Australia does not significantly influence o...

  10. Computed tomography as a core analysis tool: Applications, instrument evaluation, and image improvement techniques

    International Nuclear Information System (INIS)

    In recent years, the use of computerized tomography (CT) to characterize two-phase fluid flow through porous media has become increasingly popular. This paper describes a different application of CT: it use as a core analysis tool. The advantages and disadvantages of the different technological generations of commercial medical CT scanners available as core analysis instruments are also discussed. Additionally, methods are presented for improving images and reducing CT-number errors inherent in the scanning of high-density rock samples on instruments whose software was designed for the scanning of the low-density human patients

  11. Analysis on Roles for Components of Passive Emergency Core Cooling System

    International Nuclear Information System (INIS)

    International nuclear industry has been adopting a passive safety system to enhance safety and reliability of nuclear power plant with an advanced technology. Also, domestic nuclear industry issued the necessity for the development of key technologies for passive safety system design. It is necessary to develop the original technology for the improved technology, economics, and safety features. For this purpose, a Passive Emergency Core Cooling System (PECCS) is to be adopted as an improved safety design feature of APR+. When unfavorable accidents such as Station Black Out(SBO) happen, the PECCS should be able to make up the core and then cool down the core. This study discusses the applicability of PECCS and the proper design combinations especially during SBO. In this study, the applicability of PECCS and analysis on roles of components during SBO were assessed. RELAP5 calculations show that PECCS can make up the core and then prevent the core from being damaged during SBO with PAFS unavailable. Resultant analysis shows the role of the ADV for RCS depressurization, and SITs for RCS making up. When PAFS is available, ADVs is not required. Further study is required to sensitivity analysis such as actuation signal and setpoint

  12. Heterogeneous Multi Core Processors for Improving the Efficiency of Market Basket Analysis Algorithm in Data Mining

    Directory of Open Access Journals (Sweden)

    Aashiha Priyadarshni .L

    2014-09-01

    Full Text Available Heterogeneous multi core processors can offer diverse computing capabilities. The efficiency of Market Basket Analysis Algorithm can be improved with heterogeneous multi core processors. Market basket analysis algorithm utilises apriori algorithm and is one of the popular data mining algorithms which can utilise Map/Reduce framework to perform analysis. The algorithm generates association rules based on transactional data and Map/Reduce motivates to redesign and convert the existing sequential algorithms for efficiency. Hadoop is the parallel programming platform built on Hadoop Distributed File Systems(HDFS for Map/Reduce computation that process data as (key, value pairs. In Hadoop map/reduce, the sequential jobs are parallelised and the Job Tracker assigns parallel tasks to the Task Tracker. Based on single threaded or multithreaded parallel tasks in the task tracker, execution is carried out in the appropriate cores. For this, a new scheduler called MB Scheduler can be developed. Switching between the cores can be made static or dynamic. The use of heterogeneous multi core processors optimizes processing capabilities and power requirements for a processor and improves the performance of the system.

  13. Analysis on Roles for Components of Passive Emergency Core Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Soon Il; Hong, Soon Joon [FNC Tech, Yongin (Korea, Republic of); Kang, Sang Hee; Kim, Han Gon [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    International nuclear industry has been adopting a passive safety system to enhance safety and reliability of nuclear power plant with an advanced technology. Also, domestic nuclear industry issued the necessity for the development of key technologies for passive safety system design. It is necessary to develop the original technology for the improved technology, economics, and safety features. For this purpose, a Passive Emergency Core Cooling System (PECCS) is to be adopted as an improved safety design feature of APR+. When unfavorable accidents such as Station Black Out(SBO) happen, the PECCS should be able to make up the core and then cool down the core. This study discusses the applicability of PECCS and the proper design combinations especially during SBO. In this study, the applicability of PECCS and analysis on roles of components during SBO were assessed. RELAP5 calculations show that PECCS can make up the core and then prevent the core from being damaged during SBO with PAFS unavailable. Resultant analysis shows the role of the ADV for RCS depressurization, and SITs for RCS making up. When PAFS is available, ADVs is not required. Further study is required to sensitivity analysis such as actuation signal and setpoint.

  14. Analysis of core physics test data and sodium void reactivity worth calculation for MONJU core with ARCADIAN-FBR computer code system

    International Nuclear Information System (INIS)

    In order to evaluate core characteristics of fast reactors, a computer code system ARCADIAN-FBR has been developed by utilizing the existing analysis codes and the latest nuclear data library JENDL-3.3. The validity of ARCADIAN-FBR was verified by using the experimental data obtained in the MONJU core physics tests. The results of analyses are in good agreement with the experimental data and the applicability of ARCADIAN-FBR for fast reactor core analysis is confirmed. Using ARCADIAN-FBR, the sodium void reactivity worth, which is an important parameter in the safety analysis of fast reactors, was analyzed for MONJU core. 241Pu in the core fuel is transmuted to 241Am due to disintegrations. Therefore, the effect of 241Am accumulation on the sodium void reactivity worth was evaluated for MONJU core. As a result of calculation, it was confirmed that the accumulation of 241Am significantly influences on the sodium void reactivity worth and hence on the safety analysis of sodium-cooled fast reactors. (author)

  15. Analysis of the containment spray effect for severe accident management during Molten Core-Concrete Interaction

    International Nuclear Information System (INIS)

    Massive combustible gases generated by MCCI during a severe accident in NPP causes a problem of when we should spray the containment. The increase of hydrogen concentration due to the steam condensation caused by spraying might lead to a hydrogen burning and thus intimidate the containment integrity. In case the containment is designed to be robust enough to sustain the AICC (Adiabatic Isochoric Complete Combustion) load and to prevent DDT (Deflagration to Detonation Transition), it might be effective to spray and thus burn the hydrogen at early phase of MCCI to keep the containment integrity. Spraying the containment at late phase of MCCI might cause the containment to fail because of the increased combustible gases generation. MELCOR analysis for APR1400 shows that spraying the containment at early phase can delay the time to reach containment failure pressure by steam inerting and oxygen depletion. This kind of analysis helps us to better establish a spray actuation time for an accident management procedure against a postulated severe accident

  16. 儿童烧伤1036例特点分析%Feature Analysis of 1036 Cases of Children Burns

    Institute of Scientific and Technical Information of China (English)

    韩利坤

    2013-01-01

    Objective To explore the characteristics of etiology and clinical manifestation ,and to prevent children burns happen . Methods Their age ,sex ratio ,burn area ,burn place ,seasonal variation ,body location of the injury ,guardian is present or not ,par-ents’occupation ,education level were analyzed .Results The proportion of 1036 burn children is 40 .84% in burn patients in hospital at the same time ,gender ratio is 1 .7 :1 .Children who are one to three years old got the highest incidence-up to 77 .32% .The most frequent reason is hydrothermal burn (91 .31% ) .The most common is limbs burn at the same time (60 .33% ) .Light ,moderate burns which are less than 30% are the common injury (95 .66% ) .The most common is summer burns (31 .27% ) .Burn site occurs mostly in the home (97 .10% ) .There is guardian present(81 .27% ) .Most of children’s Parents are migrant workers(72 .98% ) . Education level of primary and secondary education is more common (71 .81% ) .Conclusion We should strengthen monitoring and watch of children ,and carry out emergency rescue burn prevention health education ,reduce the harm of children burn .%  目的探寻儿童烧伤的发病规律、特点,预防儿童烧伤发生。方法回顾性分析12岁及以下儿童烧伤病例,对性别、年龄、致伤原因、烧伤部位、烧伤发生的季节、烧伤地点、烧伤时监护人是否在场、烧伤患儿的父母职业和文化程度等进行分析。结果1036例烧伤儿童占同期烧伤住院患者(2537例)的比例为40.84%,男女性别比例为1.7∶1;年龄段以1岁~3岁阶段儿童烧伤最多,达到77.32%;致伤原因主要为热液烫伤(91.31%);以多部位同时受伤,四肢烧伤常见(60.33%);烧伤面积以30%以下的轻、中度烧伤为主(95.66%);烧伤季节以6月8月夏季烧伤最为常见(31.27%);烧伤地点大多发生在家中(97.10%),有监护人在场的达81.27

  17. Ion beam analysis techniques for the elemental fingerprinting of fine particle smoke from vegetation burning in NSW

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, D. [Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW (Australia)

    1996-12-31

    Accelerator based ion beam analysis (IBA) techniques, including PIXE, PIGME, RBS and PESA, have been used to analyse elemental compositions of airborne particles covering a 60,000 square kilometres area of Wollongong, Sydney and Newcastle. These IBA techniques provide elemental concentrations for over 20 different elements from hydrogen to lead, they include H, C, N, O, F, Na, Al, Si, P, S, Cl, K, Ca, Ti, V, Cr, Mn, Fe, Cu, Ni, Zn, Br and Pb. The four ion beam techniques are performed simultaneously on the 3MV Van de Graaff accelerator at ANSTO and have been described in detail elsewhere. They are sufficiently sensitive to analyse for many of these elements to levels around 10 ng/m{sup 3} or less in about five minutes of accelerator running time per filter. This is more than adequate for aerosol analyses as most filters contain around 150 {mu}g/cm{sup 2} of material which corresponds to about 10{mu}g/m{sup 3} of fine particles in the atmosphere. For this work fine particles are those with diameters less than 2.5{mu}m. Fine particle data has been collected twice a week and analysed for each of the above elements by ANSTO since 1991 at more than 25 different sites throughout NSW. This large dataset set allows us to not only determine the composition of fine particles and to look for signature elements for particular sources but also to use multivariate statistics to define elemental source fingerprints and then to determine the percentage contributions of these fingerprints to the total fine particle mass in the atmosphere. This paper describes the application of these techniques to the study of domestic wood fires and vegetation burning in NSW over a two year period from 1992-93. It also presents, for the first time, fine particle data related to the January 1994 bushfires in NSW. 6 refs., 1 tab., 5 figs.

  18. Ion beam analysis techniques for the elemental fingerprinting of fine particle smoke from vegetation burning in NSW

    International Nuclear Information System (INIS)

    Accelerator based ion beam analysis (IBA) techniques, including PIXE, PIGME, RBS and PESA, have been used to analyse elemental compositions of airborne particles covering a 60,000 square kilometres area of Wollongong, Sydney and Newcastle. These IBA techniques provide elemental concentrations for over 20 different elements from hydrogen to lead, they include H, C, N, O, F, Na, Al, Si, P, S, Cl, K, Ca, Ti, V, Cr, Mn, Fe, Cu, Ni, Zn, Br and Pb. The four ion beam techniques are performed simultaneously on the 3MV Van de Graaff accelerator at ANSTO and have been described in detail elsewhere. They are sufficiently sensitive to analyse for many of these elements to levels around 10 ng/m3 or less in about five minutes of accelerator running time per filter. This is more than adequate for aerosol analyses as most filters contain around 150 μg/cm2 of material which corresponds to about 10μg/m3 of fine particles in the atmosphere. For this work fine particles are those with diameters less than 2.5μm. Fine particle data has been collected twice a week and analysed for each of the above elements by ANSTO since 1991 at more than 25 different sites throughout NSW. This large dataset set allows us to not only determine the composition of fine particles and to look for signature elements for particular sources but also to use multivariate statistics to define elemental source fingerprints and then to determine the percentage contributions of these fingerprints to the total fine particle mass in the atmosphere. This paper describes the application of these techniques to the study of domestic wood fires and vegetation burning in NSW over a two year period from 1992-93. It also presents, for the first time, fine particle data related to the January 1994 bushfires in NSW. 6 refs., 1 tab., 5 figs

  19. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    International Nuclear Information System (INIS)

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public

  20. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    Energy Technology Data Exchange (ETDEWEB)

    PARMA JR.,EDWARD J.

    2000-01-01

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

  1. DNBR calculation in digital core protection system by a subchannel analysis code

    International Nuclear Information System (INIS)

    The DNBR calculation uncertainty and DNBR margin were evaluated in digital core protection system by a thermal-hydrualic subchannel analysis code MATRA. A simplified thermal-hydraulic code CETOP is used to calculate on-line DNBR in core protection system at a digital PWR. The DNBR tuning process against a best-estimate subchannel analysis code is required for CETOP to ensure accurate and conservative DNBR calculation but not necessary for MATRA. The DNBR calculations by MATRA and CETOP were performed for a large number of operating condition in Yonggwang nulcear units 3-4 where the digitial core protection system is initially implemented in Korea. MATRA resulted in a less negative mean value (i.e., reduce the overconservatism) and a somewhat larger standard deviation of the DNBR error. The uncertainty corrected minimum DNBR by MATRA was shown to be higher by 1.8% -9.9% that the CETOP DNBR

  2. Analysis of Stainless Steel Sandwich Panels with a Metal Foam Core for Lightweight Fan Blade Design

    Science.gov (United States)

    Min, James B.; Ghosn, Louis J.; Lerch, Bradley A.; Raj, Sai V.; Holland, Frederic A., Jr.; Hebsur, Mohan G.

    2004-01-01

    The quest for cheap, low density and high performance materials in the design of aircraft and rotorcraft engine fan and propeller blades poses immense challenges to the materials and structural design engineers. The present study investigates the use of a sandwich foam fan blade mae up of solid face sheets and a metal foam core. The face sheets and the metal foam core material were an aerospace grade precipitation hardened 17-4 PH stainless steel with high strength and high toughness. The resulting structures possesses a high stiffness while being lighter than a similar solid construction. The material properties of 17-4 PH metal foam are reviewed briefly to describe the characteristics of sandwich structure for a fan blade application. A vibration analysis for natural frequencies and a detailed stress analysis on the 17-4 PH sandwich foam blade design for different combinations of kin thickness and core volume are presented with a comparison to a solid titanium blade.

  3. Nonlinear dynamic analysis of prismatic elements for high-temperature gas-cooled reactor cores

    International Nuclear Information System (INIS)

    The high-temperature gas-cooled reactor (HTGR) core consists of several thousand prismatic graphite fuel elements arranged in columns within a prestressed concrete vessel. A major research and development effort was initiated in 1970 at General Atomic Company to study the dynamic response of the HTGR core arrangement to seismic excitation. A discussion is pesented of the history and some of the results of this effort with respect to the advances made in the development of analytical methods. The computer programs developed to perform the analysis are described, along with certain techniques and the modeling required to utilize them. The nonlinear dynamic analysis techniques employed to analyze the HTGR core are described

  4. CoreFlow: A computational platform for integration, analysis and modeling of complex biological data

    DEFF Research Database (Denmark)

    Pasculescu, Adrian; Schoof, Erwin; Creixell, Pau;

    2014-01-01

    the relationships between the data, the assumptions that have been made, and the manipulations that have been performed. Since the scripts use commonly available programming languages, they can easily be transferred to and from other computational environments for debugging or faster processing. This focus on ‘on...... provides programmers with a framework to manage data in real-time. It allows users to upload data into a relational database (MySQL), and to create custom scripts in high-level languages such as R, Python, or Perl for processing, correcting and modeling this data. CoreFlow organizes these scripts...... the fly’ analysis sets CoreFlow apart from other workflow applications that require wrapping of scripts into particular formats and development of specific user interfaces. Importantly, current and future releases of data analysis scripts in CoreFlow format will be of widespread benefit to the proteomics...

  5. Nuclear analysis and performance of the Light Water Breeder Reactor (LWBR) core power operation at Shippingport

    International Nuclear Information System (INIS)

    This report presents the nuclear analysis and discusses the performance of the LWBR core at Shippingport during power operation from initial startup through end-of-life at 28,730 EFPH. Core follow depletion calculations confirmed that the reactivity bias and power distributions were well within the uncertainty allowances used in the design and safety analysis of LWBR. The magnitude of the core follow reactivity bias has shown that the calculational models used can predict the behavior of U233-Th systems with closely spaced fuel rod lattices and movable fuel. In addition, the calculated final fissile loading is sufficiently greater than the initial fissile inventory that the measurements to be performed for proof-of-breeding evaluations are expected to confirm that breeding has occurred

  6. Thermal-hydraulic analysis of the MIT research reactor low enrichment uranium (LEU) Core

    International Nuclear Information System (INIS)

    The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the MIT reactor is evolving. The in-house multi-channel thermal-hydraulics code, MULCH, was developed specifically for the MITR. This code has been benchmarked against PLTEMP for steady-state analysis, and RELAP5 and temperature measurements for the loss of primary flow transient. In this paper, thermal hydraulic analyses using MULCH and RELAP5 in support of the MITR conversion tasks are described. Various fuel configurations are evaluated in order to support the LEU core design optimization study. The results show that a preferable LEU core design employs a fuel meat thickness of 20 mils with 18 plates per element with a hot channel factor less than 1.76. Simulation results also show that the LEU-fueled MITR can potentially operate at a higher power level, about 30 % higher than the current core. (authors)

  7. Analysis of heterogeneous boron dilution transients during outages with APROS 3D nodal core model

    International Nuclear Information System (INIS)

    A diluted water plug can form inside the primary coolant circuit if the coolant flow has stopped at least temporarily. The source of the clean water can be external or the fresh water can build up internally during boiling/condensing heat transfer mode, which can occur if the primary coolant inventory has decreased enough during an accident. If the flow restarts in the stagnant primary loop, the diluted water plug can enter the reactor core. During outages after the fresh fuel has been loaded and the temperature of the coolant is low, the dilution potential is the highest because the critical boron concentration is at the maximum. This paper examines the behaviour of the core as clean or diluted water plugs of different sizes enter the core during outages. The analysis were performed with the APROS 3D nodal core model of Loviisa VVER-440, which contains an own flow channel and 10 axial nodes for each fuel assembly. The widerange cross section data was calculated with CASMO-4E. According to the results, the core can withstand even large pure water plugs without fuel failures on natural circulation. The analyses emphasize the importance of the simulation of the backflows inside the core when the reactor is on natural circulation.

  8. Analysis of pan-genome to identify the core genes and essential genes of Brucella spp.

    Science.gov (United States)

    Yang, Xiaowen; Li, Yajie; Zang, Juan; Li, Yexia; Bie, Pengfei; Lu, Yanli; Wu, Qingmin

    2016-04-01

    Brucella spp. are facultative intracellular pathogens, that cause a contagious zoonotic disease, that can result in such outcomes as abortion or sterility in susceptible animal hosts and grave, debilitating illness in humans. For deciphering the survival mechanism of Brucella spp. in vivo, 42 Brucella complete genomes from NCBI were analyzed for the pan-genome and core genome by identification of their composition and function of Brucella genomes. The results showed that the total 132,143 protein-coding genes in these genomes were divided into 5369 clusters. Among these, 1710 clusters were associated with the core genome, 1182 clusters with strain-specific genes and 2477 clusters with dispensable genomes. COG analysis indicated that 44 % of the core genes were devoted to metabolism, which were mainly responsible for energy production and conversion (COG category C), and amino acid transport and metabolism (COG category E). Meanwhile, approximately 35 % of the core genes were in positive selection. In addition, 1252 potential essential genes were predicted in the core genome by comparison with a prokaryote database of essential genes. The results suggested that the core genes in Brucella genomes are relatively conservation, and the energy and amino acid metabolism play a more important role in the process of growth and reproduction in Brucella spp. This study might help us to better understand the mechanisms of Brucella persistent infection and provide some clues for further exploring the gene modules of the intracellular survival in Brucella spp. PMID:26724943

  9. Assessment of the vegetation cover in a burned area 22-years ago using remote sensing techniques and GIS analysis (Sierra de las Nieves, South of Spain).

    Science.gov (United States)

    Martínez-Murillo, Juan F.; Remond, Ricardo; Ruiz-Sinoga, José D.

    2015-04-01

    The study aim was to characterize the vegetation cover in a burned area 22-years ago considering the previous situation to wildfire in 1991 and the current one in 2013. The objectives were to: (i) compare the current and previous vegetation cover to widlfire; (ii) evaluate whether the current vegetation has recovered the previous cover to wildfire; and (iii) determine the spatial variability of vegetation recovery after 22-years since the wildfire. The study area is located in Sierra de las Nieves, South of Spain. It corresponds to an area affected by a wildfire in August 8th, 1991. The burned area was equal to 8156 ha. The burn severity was spatially very high. The main geographic features of the burned area are: mountainous topography (altitudes ranging from 250 m to 1500 m; slope gradient >25%; exposure mainly southfacing); igneous (peridotites), metamorphic (gneiss) and calcareous rocks (limestones); and predominant forest land use (Pinus pinaster sp. woodlands, 10%; pinus opened forest + shrubland, 40%; shrubland, 35%; and bare soil + grassland, 15%). Remote sensing techniques and GIS analysis has been applied to achieve the objectives. Landsat 5 and Landsat 8 images were used: July 13th, 1991 and July 1st, 2013, for the previous wildfire situation and 22-years after, respectively. The 1990 CORINE land cover was also considered to map 1991 land uses prior the wildfire. Likewise, the Andalucía Regional Government wildfire historic records were used to select the burned area and its geographical limit. 1991 and 2013 land cover map were obtained by means of object-oriented classifications. Also, NDVI and PVI1 vegetation indexes were calculated and mapped for both years. Finally, some images transformations and kernel density images were applied to determine the most recovered areas and to map the spatial concentration of bare soil and pine cover areas in 1991 and 2013, respectively. According to the results, the combination of remote sensing and GIS analysis let

  10. Burn Injuries: Burn Depth, Physiopathology and Type of Burns

    Directory of Open Access Journals (Sweden)

    Kemalettin Koltka

    2011-07-01

    Full Text Available A significant burn injury is a serious and mortal event. The most important threat to life is hypovolemic shock with complex pathophysiologic mechanisms. Burn depth is classified as first, second, or third degree. Local inflammatory response results a vasodilatation and an increase in vascular permeability. A burn injury is a three dimensional ischemic wound. Zone of coagulation is the zone with maximum damage. Zone of stasis consists of damaged but viable tissues, the tissue is salvageable. In zone of hyperemia tissue perfusion is increased. At the beginning, cardiac output falls and systemic vascular resistance increases; cardiac performance improves as hypovolemia is corrected with fluid resuscitation. While cardiac output increases systemic vascular resistance falls below normal values and a hypermetabolic state develops. Pulmonary vascular resistance increases immediately after thermal injury and this is more prolonged. To avoid secondary pulmonary complications, the smallest resuscitation volume of fluids that maintains adequate tissue perfusion should be given. Changes parallel to the cardiovascular response develop in other organ systems. The reasons of burn injury can be thermal, electrical, chemical or radiation. It is important to know the exact mechanism of burn injury because of different therapies for a specific cause. In this review information about burn depth, local and systemic responses to burn injury and major causes of burn injury are presented. (Journal of the Turkish Society Intensive Care 2011; 9 Suppl:1-6

  11. Computational Design and Analysis of Core Material of Single-Phase Capacitor Run Induction Motor

    Directory of Open Access Journals (Sweden)

    Gurmeet Singh

    2014-07-01

    Full Text Available A Single-phase induction motor (SPIM has very crucial role in industrial, domestic and commercial sectors. So, the efficient SPIM is a foremost requirement of today's market. For efficient motors, many research methodologies and propositions have been given by researchers in past. Various parameters like as stator/rotor slot variation, size and shape of stator/rotor slots, stator/rotor winding configuration, choice of core material etc. have momentous impact on machine design. Core material influences the motor performance to a degree. Magnetic flux linkage and leakage preliminary depends upon the magnetic properties of core material and air gap. The analysis of effects of core material on the magnetic flux distribution and the performance of induction motor is of immense importance to meet out the desirable performance. An increase in the air gap length will result in the air gap performance characteristics deterioration and decrease in air gap length will lead to serious mechanical balancing concern. So possibility of much variation in air gap beyond the limits on both sides is not feasible. For the optimized performance of the induction motor the core material plays a significant role. Using higher magnetic flux density, reduction on a magnetizing reactance and leakage of flux can be achieved. In this thesis work the analysis of single phase induction motor has been carried out with different core materials. The four models have been simulated using Ansys Maxwell 15.0. Higher flux density selection for same machine dimensions result into huge amount of reduction in iron core losses and thereby improve the efficiency. In this paper 2% higher efficiency has been achieved with Steel_1010 as compared to the machine using conventional D23 material. Out of four models result reflected by the machine using steel_1010 and steel_1008 are found to be better.

  12. Tropospheric O3 over Indonesia during biomass burning events measured with GOME (Global Ozone Monitoring Experiment and compared with trajectory analysis

    Directory of Open Access Journals (Sweden)

    F. Wittrock

    2005-05-01

    Full Text Available Tropospheric ozone columns of up to 50 DU were observed by GOME (Global Ozone Monitoring Experiment above Indonesia in September 1997, while only background amounts were measured in September 1998. The Traj.x trajectory model along with BRemen's Atmospheric PHOtochemical model (BRAPHO were used to investigate the higher than average ozone columns above Indonesia. The transport analysis reveals that biomass burning over central Africa and northern Australia does not significantly influence ozone columns over Indonesia in September 1997. El Niño conditions, leading to extreme dryness and uncontrolled fires in Indonesia, produced ozone precursors, which are initially only slowly advected westwards to the central Indian Ocean. Joint transport and chemistry modelling was able to reproduce the spatial distribution and amounts of ozone, NO2 and formaldehyde columns over Indonesia. The chemistry modelling shows a net production of 3.1 Tg of ozone produced by biomass burning in Indonesia in September 1997. Transport analysis further reveals that ozone columns over the Indian Ocean, between 10 and 20° S can be accounted for by the mixing of air masses containing NOx from lightning over the Congo Basin with air masses containing volatile organic compounds from biomass burning.

  13. Innovative research reactor core designed. Estimation and analysis of gamma heating distribution

    International Nuclear Information System (INIS)

    The Gamma heating value is an important factor needed for safety analysis of each experiments that will be realized on research reactor core. Gamma heat is internal heat source occurs in each irradiation facilities or any material irradiated in reactor core. This value should be determined correctly because of the safety related problems. The gamma heating value is in general depend on. reactor core characteristics, different one and other, and then each new reactor design should be completed by gamma heating data. The Innovative Research Reactor is one of the new reactor design that should be completed with any safety data, including the gamma heating value. For this reasons, calculation and analysis of gamma heating in the hole of reactor core and irradiation facilities in reflector had been done by using of modified and validated Gamset computer code. The result shown that gamma heating value of 11.75 W/g is the highest value at the center of reactor core, higher than gamma heating value of RSG-GAS. However, placement of all irradiation facilities in reflector show that safety characteristics for irradiation facilities of innovative research reactor more better than RSG-GAS reactor. Regarding the results obtained, and based on placement of irradiation facilities in reflector, can be concluded that innovative research reactor more safe for any irradiation used. (author)

  14. LMR core thermal-hydraulic analysis accounting for interassembly heat transfer =

    International Nuclear Information System (INIS)

    In is essential to have an accurate prediction of core coolant and fuel temperature distribution in the liquid metal reactor (LMR) core themal hydraulic design, because of the design limits are imposed on the maximum temperatures of claddings and fuel pins in the sodium cooled LMRs. Due to the high thermal conductivity of the sodium, the transverse interassembly heat transfer may have a significant effect on the temperature profile within the subassembly, especially when it is adjacent to considerably hotter or colder subassemblies. Therefore, the interassembly heat transfer calculation should be considered in the LMR core thermal hydraulic design and analysis. For multi-assembly analysis, the interassembly heat transfe model was added in the MATRA-LMR code and the code was extended a single assembly analysis to multi-assembly analysis, i. e., a whole core code. For the assessment of the development status with interassembly heat transfer, the benchmark calculations were performed with SLTHE and THI3D codes using the 7-assembly problems. It is founded that the subassembly mixed mean coolant temperature has been changed as an effect of the interassembly heat transfer. And the maximum temperature change was found in the non-fueled subassembly which is considerabl colder than the fueled subassemblies

  15. Microscopic analysis on showers recorded as single core on X-ray films

    International Nuclear Information System (INIS)

    Cosmic-ray particles recorded as single dark spots on X-ray films with use of the emulsion chamber data of Brazil-Japan Collaboration are studied. Some results of microscopic analysis of such single-core-like showers on nuclear emulsion plates are reported. (Author)

  16. A study on the core analysis methodology for SMART CEA ejection accident-I

    International Nuclear Information System (INIS)

    A methodology to analyze the fuel enthalpy is developed based on MASTER that is a time dependent 3 dimensional core analysis code. Using the proposed methodology, SMART CEA ejection accident is analyzed. Moreover, radiation doses are estimated at the exclusion area boundary and low population zone to confirm the criteria for the accident. (Author). 31 refs., 13 tabs., 18 figs

  17. Modeling Overlapping Laminations in Magnetic Core Materials Using 2-D Finite-Element Analysis

    DEFF Research Database (Denmark)

    Jensen, Bogi Bech; Guest, Emerson David; Mecrow, Barrie C.

    2015-01-01

    This paper describes a technique for modeling overlapping laminations in magnetic core materials using two-dimensional finite-element (2-D FE) analysis. The magnetizing characteristic of the overlapping region is captured using a simple 2-D FE model of the periodic overlapping geometry and a comp...

  18. On the Paleostress Analysis Using Kinematic Indicators Found on an Oriented Core

    Czech Academy of Sciences Publication Activity Database

    Nováková, Lucie; Brož, Milan

    2014-01-01

    Roč. 2, č. 2 (2014), s. 76-83. ISSN 2331-9593 R&D Projects: GA MPO(CZ) FR-TI1/367 Institutional support: RVO:67985891 Keywords : paleostress analysis * borehole core * kinematic indicators * bias sampling * recent stress Subject RIV: DC - Siesmology, Volcanology, Earth Structure http://www.hrpub.org/download/20140105/UJG6-13901884.pdf

  19. Analysis of phase W neutral connection burn-out on No. 1 main transformer in Daya Bay Nuclear Power Station

    International Nuclear Information System (INIS)

    No.1 main transformer in Daya Bay Nuclear Power Station experienced a burn-out failure of Phase W neutral connection. By analyzing the failure process, the condition of the protective system and the damages to the equipment, the root causes are summarized, the corrective measures put forward and successful results achieved

  20. Impact of biomass burning on the atmosphere

    International Nuclear Information System (INIS)

    Fire has played an important part in biogeochemical cycling throughout most of the history of our planet. Ice core studies have been very beneficial in paleoclimate studies and constraining the budgets of biogeochemical cycles through the past 160,000 years of the Vostok ice core. Although to date there has been no way of determining cause and effect, concentration of greenhouse gases directly correlates with temperature in ice core analyses. Recent ice core studies on Greenland have shown that significant climate change can be very rapid on the order of a decade. This chapter addresses the coupled evolution of our planet's atmospheric composition and biomass burning. Special attention is paid to the chemical and climatic impacts of biomass burning on the atmosphere throughout the last century, specifically looking at the cycles of carbon, nitrogen, and sulfur. Information from ice core measurements may be useful in understanding the history of fire and its historic affect on the composition of the atmosphere and climate

  1. Hydraulic analysis of the emergency core cooling system of the RP-10 reactor

    International Nuclear Information System (INIS)

    This work shows calculation for the hydraulic analysis of the Emergency Core Cooling System (ECCS) of the RP-10 Reactor. This analysis is necessary for the design of such system. According to calculation results shown in the graphics, a pipe line of two inches of nominal diameter should be selected for such system and a maximum flow of 5 m3/h should be reached

  2. Core seismic analysis SNR-2 preliminary design and R and D program

    International Nuclear Information System (INIS)

    This paper reviews the work done at Belgonucleaire in the framework of the seismic analysis of FBR cores and subassemblies. The main features of the computer programme CLASH are first reviewed. Next, the main conclusions of the preliminary analysis of SNR-2 are examined. Finally, the objectives of future research are outlined and the status of the R and D program at Belgonucleaire is given. (author). 8 refs

  3. Optimization of High-Resolution Continuous Flow Analysis for Transient Climate Signals in Ice Cores

    DEFF Research Database (Denmark)

    Bigler, Matthias; Svensson, Anders; Kettner, Ernesto;

    2011-01-01

    Over the past two decades, continuous flow analysis (CFA) systems have been refined and widely used to measure aerosol constituents in polar and alpine ice cores in very high-depth resolution. Here we present a newly designed system consisting of sodium, ammonium, dust particles, and electrolytic...... meltwater conductivity detection modules. The system is optimized for high- resolution determination of transient signals in thin layers of deep polar ice cores. Based on standard measurements and by comparing sections of early Holocene and glacial ice from Greenland, we find that the new system features a...

  4. Consistency evaluation of JUPITER experiment and analysis for large FBR cores

    International Nuclear Information System (INIS)

    A series of critical experiments for study of large FBR cores, JUPITER, was analyzed with the latest analytical methods. These results were evaluated from various physical viewpoints by means of comparison with other cores or other nuclear characteristics by full use of sensitivity analysis, effect of different nuclear data libraries and application of most-detailed analytical tools. It is concluded that the JUPITER experiments and analytical results possess sufficient consistency on the whole, though there is some room for further improvements. The proper use of JUPITER data will enhance the accuracy and reliability of design work for the large FBRs. (author)

  5. Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (<20%) fuel with light water flowing from top to bottom under gravity. Standard correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core, OFI (onset of flow instability) and DNB (departure from nucleate boiling)

  6. Research reactor core conversion guidebook. V.2: Analysis (Appendices A-F)

    International Nuclear Information System (INIS)

    Volume 2 consists of detailed Appendices, covering safety analyses for generic 10 MW reactor, safety analysis - probabilistic methods, methods for preventing LOCA, radiological consequence analyses, examples of safety report amendments and safety specifications. Included in Volume 2 are example analyses for cores with with highly enriched uranium and low enriched uranium fuels showing differences that can be expected in the safety parameters and radiological consequences of postulated accidents. There are seven examples of licensing documents related to core conversion and two examples of methods for determining power limits for safety specifications in the document. Refs, figs, bibliographies and tabs

  7. Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, I.H. [SDTP User Group, Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore 45650, Islamabad (Pakistan)], E-mail: ihbokhari@yahoo.co.uk; Mahmood, T.; Chaudri, K.S. [SDTP User Group, Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore 45650, Islamabad (Pakistan)

    2007-10-15

    Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (<20%) fuel with light water flowing from top to bottom under gravity. Standard correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core, OFI (onset of flow instability) and DNB (departure from nucleate boiling)

  8. Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2015-01-01

    Full Text Available A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.

  9. 重度烧伤22例临床分析%Clinical Analysis of 22 Patients with Severe Burn

    Institute of Scientific and Technical Information of China (English)

    杨宏伟

    2013-01-01

    目的:总结重度烧伤及特重度烧伤的救治经验,寻找存在的问题,为进一步实施救治提供依据.方法:对通辽市科尔沁区第一人民医院自2010年10月-2012年10月收治的22例重度及特重度烧伤患者的临床资料进行回顾性分析.结果:22例中除2例死于多器官衰竭外,其余均治愈,治愈率90.90%.结论:如何减少或避免烧伤早期损害及并发症的防治仍是烧伤救治的关键.%Objective:To provide a basis for clinical treatment through summarizing the experience in treatment of se-vere burn and extraordinarily severe burn patients, finding out the existing problems. Methods: The general clinical data of 22 patients with severe burn and extraordinarily severe burn who were treated in the First People’s Hospital of Korqin District from October 2010 to October 2012 were retrospectively analyzed. Results: In 22 cases, 2 died of multi-organ failure, the others were cured, the cure rate was 90.90%. Conclusion:How to reduce or avoid early dam-age, prevent complications is the key to treat burn.

  10. Tokamak burn control

    International Nuclear Information System (INIS)

    Research of the fusion plasma thermal instability and its control is reviewed. General models of the thermonuclear plasma are developed. Techniques of stability analysis commonly employed in burn control research are discussed. Methods for controlling the plasma against the thermal instability are reviewed. Emphasis is placed on applications to tokamak confinement concepts. Additional research which extends the results of previous research is suggested. Issues specific to the development of control strategies for mid-term engineering test reactors are identified and addressed. 100 refs., 24 figs., 10 tabs

  11. Development and assessment of a subchannel analysis code system for SMART core design

    International Nuclear Information System (INIS)

    A subchannel code system is developed for the thermal-hydraulic analysis of SMART core, and the applicability and accuracy of the code is assessed for various experimental data with rod bundles. MATRA is a subchannel analysis code calculating the enthalpy and flow distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. MATRA has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-IV-I. MATRA has been provided with an improved structure and code functions to give more convenient user environment. Improvement of various models enhances the convergence and accuracy of the code: those include the numerical solution scheme for the crossflow, the void fraction model, and the lateral transport model, and so on. A turbulent mixing model considering void drift phenomenon is devised by employing the two-phase mixing test data under PWR and BWR conditions. MATRA/SR-1 CHF correlation system is developed from local conditions of rod bundle CHF data calculated by MATRA. The optimized 1/8 core lumping models are developed for the analysis of the thermal margins of SMART core at steady-state and transient conditions

  12. Analysis on criticality properties and control strategies after reflooding of a damaged reactor core

    International Nuclear Information System (INIS)

    Highlights: • Criticality analyses based on a severe accident condition including a core meltdown. • The critical mass assessment of corium within a core. • The analysis of criticality with the change of molten level of corium. • The change of coolant void reactivity (CVR) following a core reflood. • The establishment of practical strategies for a criticality control using H3BO3. - Abstract: In order to evaluate the crucial physical phenomena in a re-criticality of the reactor corium following a core reflood, a variety of criticality analyses have been performed with the design of a hypothetical core based on a severe accident condition including a core melt-down. This study aims to assess the critical mass of the corium within the core, to analyze the criticality change with the molten level and the coolant void reactivity (CVR), and to establish of practical strategies for a criticality control using H3BO3. The MCNPX 2.5.0 code was used to design the hypothetical accident core and to calculate the effective multiplication factor (keff), and the US Peach Bottom and Japan Fukushima nuclear power plant (NPP) was chosen for a reference model. The critical level was set to 0.95 for a conservative evaluation, and the consequential critical mass was indicated to be about 60 kg only. The 548 fuel assemblies loaded in the core were assumed to be molten up to approximately 77%, the keff value therefore has a range of 1.03241 ± 0.00194 to 1.40801 ± 0.00157. The negative reactivity was increasingly inserted as the coolant density was decreased from 95% to 0%, and the keff value approaches the sub-critical state when the void fraction is above 30%. Finally, according to the boron injection into water, it was found that at least 600 ppm 10B is required to have the sufficient criticality safety and to assure the sub-criticality condition following a severe core damage accident

  13. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    International Nuclear Information System (INIS)

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  14. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  15. Uncertainty analysis for the BEACON-COLSS core monitoring system application

    International Nuclear Information System (INIS)

    This paper will cover the measurement uncertainty analysis of BEACON-COLSS core monitoring system. The uncertainty evaluation is made by using a BEACON-COLSS simulation program. By simulating the BEACON on-line operation for analytically generated reactor conditions, accuracy of the 'Measured' results can be evaluated by comparing to analytically generated 'Truth'. The DNB power margin is evaluated based on the Combustion Engineering's Modified Statistical Combination of Uncertainties (MSCU) using the CETOPD code for the DNBR calculation. A BEACON-COLSS simulation program for the uncertainty evaluation function has been established for plant applications. Qualification work has been completed for two Combustion Engineering plants. Results of the BEACON-COLSS measured peaking factors and DNBR power margin are plant type dependent and are applicable to reload cores as long as the core geometry and detector layout are unchanged. (authors)

  16. High Level Analysis, Design and Validation of Distributed Mobile Systems with CoreASM

    Science.gov (United States)

    Farahbod, R.; Glässer, U.; Jackson, P. J.; Vajihollahi, M.

    System design is a creative activity calling for abstract models that facilitate reasoning about the key system attributes (desired requirements and resulting properties) so as to ensure these attributes are properly established prior to actually building a system. We explore here the practical side of using the abstract state machine (ASM) formalism in combination with the CoreASM open source tool environment for high-level design and experimental validation of complex distributed systems. Emphasizing the early phases of the design process, a guiding principle is to support freedom of experimentation by minimizing the need for encoding. CoreASM has been developed and tested building on a broad scope of applications, spanning computational criminology, maritime surveillance and situation analysis. We critically reexamine here the CoreASM project in light of three different application scenarios.

  17. CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

    International Nuclear Information System (INIS)

    The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

  18. Precision measurement system and analysis of low core signal loss in DCF couplers

    Science.gov (United States)

    Yan, P.; Wang, X. J.; Fu, Ch; Li, D.; Sun, J. Y.; Gong, M. L.; Xiao, Q. R.

    2016-07-01

    In order to achieve higher output power of double cladding fiber lasers, low signal loss has become a focus in researches on optical technology, especially double-clad fiber (DCF) couplers. According to the analysis, DCF couplers with low core signal loss (less than 1%) are produced. To obtain higher precision, we use the first-proposed method for core signal transfer efficiency measurement based on the fiber propagation field image processing. To the best of our knowledge, we report, for the first time, the results of the core signal loss less than 1% in DCF coupler measured by our measurement with high stability and relative precision. The measurement values can assess the quality of DCF couplers and be used as a signal to suggest the improvement on the processing technology of our self-made DCF couplers.

  19. Low-background neutron activation analysis. A powerful tool for atmospheric mineral dust analysis in ice cores

    International Nuclear Information System (INIS)

    The application of instrumental neutron activation analysis (INAA) for multi-elemental analysis of samples of extremely reduced mass such as dust samples extracted from ice cores requires specific efforts towards the development of a 'low level counting' analytical technique. An analytical protocol specifically designed for this kind of samples, based on low-background INAA (LBNAA) is here presented. A first application of the method was successfully performed on samples from the new alpine firn core NextData-LYS12. Sub-ng detection limits were reached for many elements. According to this point the technique is also potentially suitable to be applied to polar ice core samples. (author)

  20. O2C Laser Doppler and Digital Photo Analysis for Treatment Evaluation of Beta-Glucan versus Provitamin Pantothenic Acid of Facial Burns.

    Science.gov (United States)

    Thieme, Daniel; Spilker, Gerald; Lefering, Rolf; Weinand, Christian

    2016-04-01

    Various creams are available for superficial second-degree burns (SSDB) of the face. We evaluated provitamin pantothenic acid versus β-glucan for SSDB of the face using the O2C laser Doppler system and digital photo analysis. Out of 20 patients (January to December 2012) with facial burns, 7 with SSDB of both cheeks were included to our study. Burned cheek wounds were treated using pantothenic acid or β-glucan. Digital photos of marked regions were taken daily from predefined distances. Microcirculation was measured at marked regions for 7 days at scheduled time points using the O2C laser Doppler. Data were evaluated using the SPSS program (SPSS Inc., Chicago, IL). Wounds treated with β-glucan showed faster reepithelialization. O2C laser Doppler measurements showed faster increase in SO2, microvascular perfusion, hemoglobin content, and blood flow. This correlated good with clinical Vancouver Scar Scale results. Although not statistically significant, β-glucan cream therapy of SSDB results in aesthetically superior outcome and faster reepithelialization. PMID:27097145

  1. Clinical analysis of 136 cases of eye burns%眼烧伤136例临床分析

    Institute of Scientific and Technical Information of China (English)

    刘建伟; 邓爱军; 刘娜

    2012-01-01

    目的 探讨眼烧伤的临床特点及手术效果.方法 回顾性分析眼烧伤136例(206眼)的致伤原因、致伤程度、视力、并发症及手术治疗效果.结果 碱烧伤最多,热烧伤、酸烧伤次之.206眼中Ⅲ度烧伤107眼,占51.9%;Ⅳ度烧伤16眼,占7.8%.早期行羊膜移植者术后视力0.3及以上者95眼,占69.3%;无严重并发症者38眼,占27.8%.早期未行羊膜移植者视力0.3及以上者29眼,占51.8%;无严重并发症者7眼,占12.5%.化学烧伤者与热烧伤者相比,视力恢复差异无统计学意义(x2=2.14,P>0.05),但发生的严重并发症较多(x2=4.10,P<0.05).并发症期接受自体角膜缘移植联合板层角膜移植术者中,视力有显著提高者占79.3%.结论 早期诊疗、早期的羊膜移植和晚期的自体角膜缘移植联合板层角膜移植术对于减少并发症发生及术后视力的恢复有重要意义.%Objective To investigate the clinical characteristics and the result of surgery treatments of eye burns.Methods 136 patients(206 eyes)were evaluated retrospectively.The causes,degree of damage,visual acuity,complications and outcome of operation were analyzed.Results Alkali bum was the principal cause(85 eyes,41.3%),thermal burn(77eyes,37.4%)and acid burn(36eyes,17.5%)were next.to it.Most patients were severely burned,among them,107(51.9%)were Ⅲ degree burn and 16 were Ⅳ degree bum.Vision of 95 among 137(69.3%)eyes with early amniotic membrane transplantation(AMT)recover a vision better than 0.3,the number was 29 among 56(51.8%)in group of eyes with non-early AMT.Among eyes with early AMT,38 eyes(27.8%)didn't have severe complications,only 7 eyes(12.5%)didn' have severe complications in non-early AMT group.There was no significant difference of vision between chemical burn patients and thermal bum patients(P > 0.05),while the complications in chemical burn patients were more than that in thermal burn patients(P < 0.05).Most patients

  2. Improvement of core effective thermal conductivity model of GAMMA+ code based on CFD analysis

    International Nuclear Information System (INIS)

    Highlights: • We assessed the core effective thermal conductivity (ETC) model of GAMMA+ code. • The analytical model of GAMMA+ code was compared with the result of CFD analysis. • Effects of material property of composite and geometric configuration were studied. • The GAMMA+ model agreed with the CFD result when the fuel gap is ignored. • The GAMMA+ model was improved by the ETC model of fuel compact including fuel gap. - Abstract: The GAMMA+ code has been developed for the thermo-fluid and safety analyses of a high temperature gas-cooled reactor (HTGR). In order to calculate the core effective thermal conductivity, this code adopts a heterogeneous model derived from the Maxwell’s theory that accounts for three distinct materials in a fuel block of the reactor core. In this model, the fuel gap is neglected since the gap thickness is quite small. In addition, the configuration of the fuel block is assumed to be homogeneous, and the volume fraction and material properties of each component are taken into account. In the accident condition, the conduction and radiation are major heat transfer mechanism. Therefore, the core effective thermal conductivity model should be validated in order to estimate the heat transfer in the core appropriately. In this regard, the objective of this study is to validate the core effective thermal conductivity model of the GAMMA+ code by a computational fluid dynamics (CFD) analysis using a commercial CFD code, CFX-13. The effects of the temperature condition, material property and geometric modeling on the core effective thermal conductivity were investigated. When the fuel gap is not modeled in the CFD analysis, the result of the GAMMA+ code shows a good agreement with the CFD result. However, when the fuel gap is modeled, the GAMMA+ model overestimates the core effective thermal conductivity considerably for all cases. This is because of the increased thermal resistance by the fuel gap which is not taken into account in

  3. Toroidal superconducting transformer with cold magnetic core – results of analysis and measurements

    International Nuclear Information System (INIS)

    The paper is focused on a toroidal superconducting transformer with cold magnetic core. The transformer was developed aiming at the solution where magnetic core could operate immersed in LN2, not having much higher power losses than a core operating in room temperature. The second aim is concerned to the perpendicular component of the magnetic flux. It was assumed that it should be minimized. The third aim is required maximal coupling coefficient between windings. Magnetic material METGLAS 2605 SA1 (Metglas Ltd) was selected as the best taking electrical and mechanical properties into account. Numerical analysis of the transformer was carried using ANSYS software. It was assumed that magnetic core could be put inside windings and/or outside of it. The analysis and measurements yield that the best solution is the transformer with magnetic core put inside the windings. Such a construction reduces significantly perpendicular component of the magnetic field in HTS wire. Because outer winding is put on the inner one it gives maximal possible coupling coefficient. The primary and the secondary have the same number of turns. It means bifilar way of winding. The described transformer has the following constructional data: i) diameter of the main circle: 144 mm, ii) maximal diameter: 208 mm, iii) diameter of the small cross-section of the transformer (without external magnetic core): 50 mm and it results with the following parameters: output power 1.59 kVA, power density 700 VA/kg, efficiency: 99.5, coupling coefficient 0.99 at 100 ARMS of input current and maximum flux density 1.2 T.

  4. Analysis of reaction rate and sample worth measured in simulated LMFBR meltdown cores

    International Nuclear Information System (INIS)

    An analysis of fission rate, fission rate ratio and sample worth has been made on FCA VIII-2 Assembly built to simulate LMFBR meltdown cores. To check the adoptability of computational methods used in analysing core disruptive accidents, the results obtained by the three methods, transport (Sn), conventional and modified diffusion methods, were compared with the measured ones. Group constants used in the analysis were prepared from JAERI Fast Set Version II. Conventional diffusion theory does not represent the measured fission rates both in the compacted and voided region. The results obtained by transport theory with S4P0 approximation agree fairly well with the measured ones. The use of modified diffusion theory, which changes the diffusion coefficient in the voided region does not significantly improve the agreement. For Pu sample worth, problems still remain partially which need a more detailed analysis, although the use of transport theory improves the agreement to a large extent. (author)

  5. Development of a FBR fuel bundle-duct interaction analysis code-BAMBOO. Analysis model and verification by Phenix high burn-up fuel subassemblies

    International Nuclear Information System (INIS)

    The bundle-duct interaction analysis code ''BAMBOO'' has been developed for the purpose of predicting deformation of a wire-wrapped fuel pin bundle of a fast breeder reactor (FBR). The BAMBOO code calculates helical bowing and oval-distortion of all the fuel pins in a fuel subassembly. We developed deformation models in order to precisely analyze the irradiation induced deformation by the code: a model to analyze fuel pin self-bowing induced by circumferential gradient of void swelling as well as thermal expansion, and a model to analyze dispersion of the orderly arrangement of a fuel pin bundle. We made deformation analyses of high burn-up fuel subassemblies in Phenix reactor and compared the calculated results with the post irradiation examination data of these subassemblies for the verification of these models. From the comparison we confirmed that the calculated values of the oval-distortion and bowing reasonably agreed with the PIE results if these models were used in the analysis of the code. (author)

  6. Secondary Ion Mass Spectrometry Analysis of Materials to Develop In-core Safeguards Reactor Monitoring Devices

    International Nuclear Information System (INIS)

    During reactor operations and fuel burn up, some isotopic abundances change due to nuclear reactions and provide sensitive indicators of neutron fluence and fuel burnup. Secondary ion mass spectrometry (SIMS) analysis has been used to directly measure isotope ratios of selected impurity elements in irradiated nuclear reactor materials. Direct in situ SIMS measurements were made in graphite and metal samples, following shaping and surface cleaning. Other elements such as Be must be chemically separated and purified prior to SIMS analyses. Elements such as pre-existing impurity U and Pu produced from the U, are in low abundance and must also be chemically separated and are measured by thermal ionization mass spectrometry (TIMS). Studies combining SIMS and TIMS analyses demonstrate the value of this approach in determining reactor fluence profiles, power production, and other parameters. Future work proceeding from this analytical work will include developing monitoring devices designed for relatively easy placement and retrieval in a reactor, and direct SIMS analyses after exposure.

  7. Analysis of severe hypothetical accidents for the SNR-300 MARK 1A core

    International Nuclear Information System (INIS)

    Two types of hypothetical accidents have been analysed for the fresh and irradiated SNR-300 MARK 1A cores: 1) Loss of flow accidents caused by a coast down of all primary pumps and simultaneous failure of both independent shutdown systems. 2) Transient overpower accidents caused by a reactivity input ramp and simultaneous failure of both independent shutdown systems. The analysis was done by using the CAPRI-2/KADIS computer-program-system which was developed at the Nuclear Research Center Karlsruhe. Detailed parametric variations were performed for both accident types in case of the fresh core to determine the most important and influential parameters. These parameter studies were also used to allow a conservative parameter choice for the reference cases within a reasonable parameter band. In addition the parametric variations gave some insight under which circumstances the accident would lead into energetic disassembly, early shutdown with in-place cooling possibility, or a transition phase with extended fuel motion. The thermal energy in the molten fuel at the end of the nuclear excursion is one important quantity for the severity of the accident. In case of the loss of flow accidents these energies were 3,239 MWs and 3,605 MWs for the reference cases of the fresh and irradiated cores, respectively. The corresponding energies for the reference transient overpower accidents caused by a 15 cent/sec reactivity input ramp were 1,182 MWs and 2,940 MWs for the fresh and irradiated cores respectively. Besides the thermal energy release the analysis provides much more information, for example, about the core conditions at the end of the nuclear excursion. More important the analysis automatically gives the input data for the computer programs which analyse the mechanical response of the reactor tank and the tankinternal mechanical structures. (orig./HP)

  8. Dating a tropical ice core by time-frequency analysis of ion concentration depth profiles

    Science.gov (United States)

    Gay, M.; De Angelis, M.; Lacoume, J.-L.

    2014-09-01

    Ice core dating is a key parameter for the interpretation of the ice archives. However, the relationship between ice depth and ice age generally cannot be easily established and requires the combination of numerous investigations and/or modelling efforts. This paper presents a new approach to ice core dating based on time-frequency analysis of chemical profiles at a site where seasonal patterns may be significantly distorted by sporadic events of regional importance, specifically at the summit area of Nevado Illimani (6350 m a.s.l.), located in the eastern Bolivian Andes (16°37' S, 67°46' W). We used ion concentration depth profiles collected along a 100 m deep ice core. The results of Fourier time-frequency and wavelet transforms were first compared. Both methods were applied to a nitrate concentration depth profile. The resulting chronologies were checked by comparison with the multi-proxy year-by-year dating published by de Angelis et al. (2003) and with volcanic tie points. With this first experiment, we demonstrated the efficiency of Fourier time-frequency analysis when tracking the nitrate natural variability. In addition, we were able to show spectrum aliasing due to under-sampling below 70 m. In this article, we propose a method of de-aliasing which significantly improves the core dating in comparison with annual layer manual counting. Fourier time-frequency analysis was applied to concentration depth profiles of seven other ions, providing information on the suitability of each of them for the dating of tropical Andean ice cores.

  9. Experimental investigation and CFD analysis on cross flow in the core of PMR200

    International Nuclear Information System (INIS)

    Highlights: • An experimental facility for the cross flow in the core of PMR200 was constructed. • The cross flow experimental data were produced with wedge-shaped gap and parallel gap. • The experimental results were compared with CFD results. • The results of the CFD analysis and experimental data are in good agreement. • The pressure loss coefficient for the cross gap of PMR200 was derived. - Abstract: The Prismatic Modular Reactor (PMR) is one of the major Very High Temperature Reactor (VHTR) concepts, which consists of hexagonal prismatic fuel blocks and reflector blocks made of nuclear grade graphite. However, the shape of the graphite blocks could be easily changed by neutron damage during the reactor operation and the shape change can create gaps between the blocks inducing the bypass flow. In the VHTR core, two types of gaps, a vertical gap and a horizontal gap which are called bypass gap and cross gap, respectively, can be formed. The cross gap complicates the flow field in the reactor core by connecting the coolant channel to the bypass gap and it could lead to a loss of effective coolant flow in the fuel blocks. Thus, a cross flow experimental facility was constructed to investigate the cross flow phenomena in the core of the VHTR and a series of experiments were carried out under varying flow rates and gap sizes. The results of the experiments were compared with CFD (Computational Fluid Dynamics) analysis results in order to verify its prediction capability for the cross flow phenomena. Fairly good agreement was seen between experimental results and CFD predictions and the local characteristics of the cross flow was discussed in detail. Based on the calculation results, pressure loss coefficient across the cross gap was evaluated, which is necessary for the thermo-fluid analysis of the VHTR core using a lumped parameter code

  10. Treating and Preventing Burns

    Science.gov (United States)

    ... Issues Listen Español Text Size Email Print Share Treating and Preventing Burns Page Content Article Body Burns ... home, out of children’s reach, and away from heat or ignition sources. Lower the temperature of your ...

  11. Burns and Fire Safety

    Science.gov (United States)

    ... Tap water burns most often occur in the bathroom and tend to be more severe and cover a larger portion of the body than other scald burns. 9 10 11 A survey found that only 8 percent of adults felt ...

  12. Burn-up credit criticality benchmark. Phase 4-B: results and analysis of MOX fuel depletion calculations

    International Nuclear Information System (INIS)

    The DECD/NEA Expert Group on Burn-up Credit was established in 1991 to address scientific and technical issues connected with the use of burn-up credit in nuclear fuel cycle operations. Following the completion of six benchmark exercises with uranium oxide (UOX) fuels irradiated in pressurised water reactors (PWRs) and boiling water reactors (BWRs), the present report concerns mixed uranium and plutonium oxide (MOX) fuels irradiated in PWRs. The exercises consisted of inventory calculations of MOX fuels for two initial plutonium compositions. The depletion calculations were carried out using three representations of the MOX assemblies and their interface with UOX assemblies. This enabled the investigation of the spatial and spectral effects during the irradiation of the MOX fuels. (author)

  13. WIMCIT: An in-house development integrated code for neutron analysis in MTR reactors with mixed core

    International Nuclear Information System (INIS)

    So far, Peruvian nuclear research reactors RP-10 and RP-0 have been operating only with uranium oxide fuel elements (U3O8+Al). In the future RP-10 reactor will operate with mixed cores (U3Si2+Al, U3O8+Al). For this reason the calculus capacity has to be increased in order to be able to handle the new fuel management. Since preparation of nuclear libraries for diffusion calculation requires a significant amount of computational and human resources a new tool that processes libraries generation and diffusion calculations is necessary. For this reason the WIMCIT code was developed. The main feature of this code is to integrate WIMS and CITATION codes. The user can generate nuclear libraries in different groups (from 1 to 18) in the range of 0 - 10MeV automatically. They are generated from the WIMS output and then interpolated to create an input to CITATION according to the user specifications. Both the flux and power profiles can be calculated in different transverse sections. The output form the diffusion calculations is processed and different physical parameters of the reactor are displayed to the user. Others WIMCIT capabilities are, flux calculation, power profiles, burn-up average by element, control rod interaction in the core and fuel management in mixed core. The WIMCIT code has been validated on the experimental data from the RP-10 fresh core to the current core (Number 24). (author)

  14. A meta-analysis of core stability exercise versus general exercise for chronic low back pain.

    Directory of Open Access Journals (Sweden)

    Xue-Qiang Wang

    Full Text Available OBJECTIVE: To review the effects of core stability exercise or general exercise for patients with chronic low back pain (LBP. SUMMARY OF BACKGROUND DATA: Exercise therapy appears to be effective at decreasing pain and improving function for patients with chronic LBP in practice guidelines. Core stability exercise is becoming increasingly popular for LBP. However, it is currently unknown whether core stability exercise produces more beneficial effects than general exercise in patients with chronic LBP. METHODS: Published articles from 1970 to October 2011 were identified using electronic searches. For this meta-analysis, two reviewers independently selected relevant randomized controlled trials (RCTs investigating core stability exercise versus general exercise for the treatment of patients with chronic LBP. Data were extracted independently by the same two individuals who selected the studies. RESULTS: From the 28 potentially relevant trials, a total of 5 trials involving 414 participants were included in the current analysis. The pooling revealed that core stability exercise was better than general exercise for reducing pain [mean difference (-1.29; 95% confidence interval (-2.47, -0.11; P = 0.003] and disability [mean difference (-7.14; 95% confidence interval (-11.64, -2.65; P = 0.002] at the time of the short-term follow-up. However, no significant differences were observed between core stability exercise and general exercise in reducing pain at 6 months [mean difference (-0.50; 95% confidence interval (-1.36, 0.36; P = 0.26] and 12 months [mean difference (-0.32; 95% confidence interval (-0.87, 0.23; P = 0.25]. CONCLUSIONS: Compared to general exercise, core stability exercise is more effective in decreasing pain and may improve physical function in patients with chronic LBP in the short term. However, no significant long-term differences in pain severity were observed between patients who engaged in core stability exercise

  15. Analysis of core-melt states for the development of detection methods for filling level change and deformation of the core in PWR-type reactors

    International Nuclear Information System (INIS)

    The project ''noninvasive status monitoring of nuclear reactors for detection of filling level changes and core deformation'' (NIZUK) is aimed to develop a measuring system for the core status diagnosis during severe accidents in PWR-type reactors. For the development of an appropriate measuring technology the knowledge on the processes during the in-vessel phase of the accident sequence is of main importance. Using the analysis of the accident sequence nine in-vessel phases were defined that are the basis for the development of the measuring system. The differences between the individual core-melt states include the different core geometries and a varying gamma radiation distribution at the reactor pressure vessel outer surface. Especially the appearance of local flow-off paths during a late in-vessel phase requires that several measuring probes with gamma radiation sensors have to be installed around the reactor pressure vessel in order to detect the gamma radiation distribution at the outside. The definition of further core-melt states would be possible in case of a re-flooding of the reactor pressure vessel. However, the increasing filling level would not significantly change the core deformation and the gamma distribution at the outside.

  16. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

  17. Measurement of laminar burning velocities and analysis of flame stabilities for hydrogen-air-diluent premixed mixtures

    Institute of Scientific and Technical Information of China (English)

    HU ErJiang; HUANG ZuoHua; HE JiaJia; JIN Chun; MIAO HaiYan; WANG XiBin

    2009-01-01

    The laminar burning velocities and Markstein lengths of the hydrogen-air-diluent mixtures were meas-ured at different equivalence ratios (0.4-1.5), different diluents (N2, CO2 and 15%CO2+85%N2) and di-lution ratios (0, 0.05, 0.10 and 0.15) by using the outwardly expanding flame. The influences of flame stretch rate on the flame propagation characteristics were analyzed. The results show that both the laminar burning velocities and the Markstein lengths of the hydrogen-air-diluent mixtures decrease with the increase of dilution ratio. The decrease in Markstein lengths means that adding diluents into the hydrogen-air mixtures will decrease the diffusional-thermal instability of the flame front. For a specified dilution ratio, the laminar burning velocities give their maximum values at an equivalence ratio of 1.8. The Markstein lengths increase with the increase of the equivalence ratio monotonously regardless of the diluents. The study shows that CO2 as the diluent has a greater impact on the laminar flame speed and the flame front stability than N2 as the diluent.

  18. Optimization of burn referrals

    DEFF Research Database (Denmark)

    Reiband, Hanna K; Lundin, Kira; Alsbjørn, Bjarne;

    2014-01-01

    INTRODUCTION: Correct estimation of the severity of burns is important to obtain the right treatment of the patient and to avoid over- and undertriage. In this study we aimed to assess how often the guidelines for referral of burn injured patients are met at the national burn centre (NBC), Denmar...

  19. First Aid: Burns

    Science.gov (United States)

    ... Story" 5 Things to Know About Zika & Pregnancy First Aid: Burns KidsHealth > For Parents > First Aid: Burns Print A A A Text Size Scald ... THIS TOPIC Kitchen: Household Safety Checklist Fireworks Safety First Aid: Sunburn Firesetting Fire Safety Burns Household Safety: Preventing ...

  20. Spatial and model-order based reactor signal analysis methodology for BWR core stability evaluation

    International Nuclear Information System (INIS)

    A new methodology for the boiling water reactor core stability evaluation from measured noise signals has been recently developed and adopted at the Paul Scherrer Institut (PSI). This methodology consists in a general reactor noise analysis where as much as possible information recorded during the tests is investigated prior to determining core representative stability parameters, i.e. the decay ratio (DR) and the resonance frequency, along with an associated estimate of the uncertainty range. A central part in this approach is that the evaluation of the core stability parameters is performed not only for a few but for ALL recorded neutron flux signals, allowing thereby the assessment of signal-related uncertainties. In addition, for each signal, three different model-order optimization methods are systematically employed to take into account the sensitivity upon the model-order. The current methodology is then applied to the evaluation of the core stability measurements performed at the Leibstadt NPP, Switzerland, during cycles 10, 13 and 19. The results show that as the core becomes very stable, the method-related uncertainty becomes the major contributor to the overall uncertainty range while for intermediate DR values, the signal-related uncertainty becomes dominant. However, as the core stability deteriorates, the method-related and signal-related spreads have similar contributions to the overall uncertainty, and both are found to be small. The PSI methodology identifies the origin of the different contributions to the uncertainty. Furthermore, in order to assess the results obtained with the current methodology, a comparative study is for completeness carried out with respect to results from previously developed and applied procedures. The results show a good agreement between the current method and the other methods

  1. CDFR design and performance implications of extended burn-up

    International Nuclear Information System (INIS)

    The restrained core was adopted for the UK Commercial Demonstration Fast Reactor (CDFR) in the mid 1970's. The reasons for its adoption are still valid today although increased importance is now given to its seismic performance. During an earthquake of ''safe shutdown'' intensity the rigid restraint barrel, small clearances of the restraint pads and the cross sectional strength of the sub-assemblies, particularly at the upper restraint level adequately maintain the relative positions of sub-assemblies. A general analysis of sub-assembly motions with simplifying but conservative assumptions indicates that the transient reactivity insertions due to horizontal core movements in the safe shutdown earthquakes will be less than 1$. A programme of experiments is being initiated to support the analysis and demonstrate that the present predictions are pessimistic. The effect of higher burn-up on CDFR performance will be discussed in this paper

  2. Radiometric dating of sediment core from waterwork reservoir Rozgrund and analysis of mercury concentration depth profile

    International Nuclear Information System (INIS)

    Radioisotope dating of lake sediments combined with analysis of chemical properties of the sediment layers allow us to study the history of the human impact on nature. Undisturbed sediment layers in the core samples serve as chronicle database with information about lake ecosystem and surrounding environment in the time of deposition. A sediment core sample from the bottom of the water-work reservoir Rozgrund was collected and separated into 2 cm thick layers. Samples were analysed by HPGe spectrometry for anthropogenous Cs-137 activity. From identified peaks corresponding to nuclear tests and Chernobyl accident the sedimentation rate was calculated and the chronology of layers established. Sub-samples from each layer were prepared separately for the analysis of the Hg concentration by atomic absorption spectrometry. The results show very small variations in Hg concentrations and there is no significant trend present in the profile. (author)

  3. Core-Shell Columns in High-Performance Liquid Chromatography: Food Analysis Applications

    Science.gov (United States)

    Preti, Raffaella

    2016-01-01

    The increased separation efficiency provided by the new technology of column packed with core-shell particles in high-performance liquid chromatography (HPLC) has resulted in their widespread diffusion in several analytical fields: from pharmaceutical, biological, environmental, and toxicological. The present paper presents their most recent applications in food analysis. Their use has proved to be particularly advantageous for the determination of compounds at trace levels or when a large amount of samples must be analyzed fast using reliable and solvent-saving apparatus. The literature hereby described shows how the outstanding performances provided by core-shell particles column on a traditional HPLC instruments are comparable to those obtained with a costly UHPLC instrumentation, making this novel column a promising key tool in food analysis. PMID:27143972

  4. Core-scale solute transport model selection using Monte Carlo analysis

    CERN Document Server

    Malama, Bwalya; James, Scott C

    2013-01-01

    Model applicability to core-scale solute transport is evaluated using breakthrough data from column experiments conducted with conservative tracers tritium (H-3) and sodium-22, and the retarding solute uranium-232. The three models considered are single-porosity, double-porosity with single-rate mobile-immobile mass-exchange, and the multirate model, which is a deterministic model that admits the statistics of a random mobile-immobile mass-exchange rate coefficient. The experiments were conducted on intact Culebra Dolomite core samples. Previously, data were analyzed using single- and double-porosity models although the Culebra Dolomite is known to possess multiple types and scales of porosity, and to exhibit multirate mobile-immobile-domain mass transfer characteristics at field scale. The data are reanalyzed here and null-space Monte Carlo analysis is used to facilitate objective model selection. Prediction (or residual) bias is adopted as a measure of the model structural error. The analysis clearly shows ...

  5. Application of noise analysis to investigate core degradation process during PHEBUS-FPT1 test

    International Nuclear Information System (INIS)

    Noise analysis has been performed for measurement data obtained during PHEBUS-FPT1 test. The purpose of the study is to evaluate the applicability of the noise analysis to the following problems: To get more knowledge about the physical processes going on during severe core conditions; To better understand the core melting process; To establish appropriate on-line shut-down data. Results of the study indicate that the noise analysis is quite promising as a tool for investigating physical processes during the experiment. Compared with conventional approach of evaluating the signal's mean value behaviour, the noise analysis can provide additional, more detailed information: It was found that the neutron flux signal is subjected to additional reactivity perturbations in conjunction with fuel melting and relocation. This can easily be detected by applying noise analysis for the neutron flux signal. It has been demonstrated that the method developed in the present study can provide more accurate estimates of the onset of fuel relocation than using temperature signals from thermocouples in the thermal shroud. Moreover, the result suggests a potential of the present method for tracking the whole process of relocation. The result of the data analysis suggests a possibility of sensor diagnostics which may be important for confirming the quality and reliability of the recorded data. Based on the results achieved it is believed that the combined use of noise analysis and thermocouple signals will provide reliable shut-down criteria for the experiment. 8 refs

  6. Application of noise analysis to investigate core degradation process during PHEBUS-FPT1 test

    Energy Technology Data Exchange (ETDEWEB)

    Oguma, Ritsuo [GSE Power Systems AB, Nykoeping (Sweden)

    1997-01-01

    Noise analysis has been performed for measurement data obtained during PHEBUS-FPT1 test. The purpose of the study is to evaluate the applicability of the noise analysis to the following problems: To get more knowledge about the physical processes going on during severe core conditions; To better understand the core melting process; To establish appropriate on-line shut-down data. Results of the study indicate that the noise analysis is quite promising as a tool for investigating physical processes during the experiment. Compared with conventional approach of evaluating the signal`s mean value behaviour, the noise analysis can provide additional, more detailed information: It was found that the neutron flux signal is subjected to additional reactivity perturbations in conjunction with fuel melting and relocation. This can easily be detected by applying noise analysis for the neutron flux signal. It has been demonstrated that the method developed in the present study can provide more accurate estimates of the onset of fuel relocation than using temperature signals from thermocouples in the thermal shroud. Moreover, the result suggests a potential of the present method for tracking the whole process of relocation. The result of the data analysis suggests a possibility of sensor diagnostics which may be important for confirming the quality and reliability of the recorded data. Based on the results achieved it is believed that the combined use of noise analysis and thermocouple signals will provide reliable shut-down criteria for the experiment. 8 refs.

  7. Imaging and analysis of microcalcifications and lipid/necrotic core calcification in fibrous cap atheroma.

    Science.gov (United States)

    Maldonado, Natalia; Kelly-Arnold, Adreanne; Laudier, Damien; Weinbaum, Sheldon; Cardoso, Luis

    2015-06-01

    The presence of microcalcifications (µCalcs) >5 µm within the cap of human fibroatheroma has been shown to produce a 200-700% increase in peak circumferential stress, which can transform a stable plaque into a vulnerable one, whereas µCalcs 5 µm based on the gross morphological features of fibroatheromas, and the correlation between the size and distribution of µCalcs in the cap and the calcification in the lipid/necrotic core beneath it. Atherosclerotic lesions (N = 72) were imaged using HR-μCT at 2.1-μm resolution for detailed analysis of atheroma morphology and composition, and validated using non-decalcified histology. At 2.1-μm resolution one observes four different patterns of calcification within the lipid/necrotic core, and is able to elucidate the 3D spatial progression of the calcification process using these four patterns. Of the gross morphological features identified, only minimum cap thickness positively correlated with the existence of µCalcs > 5 µm in the cap. We also show that µCalcs in the cap accumulate in the vicinity of the lipid/necrotic core boundary with few on the lumen side of the cap. HR-μCT enables three-dimensional assessment of soft tissue composition, lipid content, calcification patterns within lipid/necrotic cores and analysis of the axial progression of calcification within individual atheroma. The distribution of µCalcs within the cap is highly non-uniform and decreases sharply as one proceeds from the lipid pool/necrotic core boundary to the lumen. PMID:25837377

  8. The adult intestinal core microbiota is determined by analysis depth and health status

    OpenAIRE

    Salonen, A.; Salojärvi, J.; Lahti, L.M.; De Vos

    2012-01-01

    High-throughput molecular methods are currently exploited to characterize the complex and highly individual intestinal microbiota in health and disease. Definition of the human intestinal core microbiota, i.e. the number and the identity of bacteria that are shared among different individuals, is currently one of the main research questions. Here we apply a high-throughput phylogenetic microarray, for a comprehensive and high-resolution microbiota analysis, and a novel computational approach ...

  9. Two-dimension calculation of proposed benchmark core analysis for the BN-600 hybrid reactor

    International Nuclear Information System (INIS)

    This paper presents primary calculation results of the proposed benchmark for a hybrid UOX/MOX fuelled core of the BN-600 reactor. The analysis in this paper uses a R-Z homogeneous model of the BN-600 reactor. Calculation results include effective multiplication factors obtained by both diffusion and Monte Carlo methods; fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution

  10. Three dimension calculation of proposed benchmark core analysis for the BN-600 hybrid reactor

    International Nuclear Information System (INIS)

    This paper presents primary calculation results of the proposed benchmark for a hybrid UOX/MOX fuelled core of the BN-600 reactor. The analysis in this paper uses a HEX-Z homogeneous model of the BN-600 reactor. Calculation results include effective multiplication factors obtained by both diffusion and Monte Carlo methods; fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution

  11. Tank 241-B-203 Push Mode Core Sampling and Analysis Plan

    Energy Technology Data Exchange (ETDEWEB)

    Jo, J. [Westinghouse Hanford Co., Richland, WA (United States)

    1995-05-16

    This Sampling and Analysis Plan (SAP) will identify characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements in accordance with the Tank Safety Screening Data Quality Objective (Babad and Redus 1994). This Data Quality Objective (DQO) is described in the Tank Characterization Plan (Jo, 1995) for tank 241-B-203 (B-203). This SAP will also identify procedures and requirements for collecting and characterizing samples from tank B-203 by the core sampling method.

  12. Analysis of core soil and water samples from the Cactus Crater Disposal Site at Enewetak atoll

    International Nuclear Information System (INIS)

    Core soil samples and water samples were collected from the Cactus Crater Disposal Site at Enewetak for analysis of 137Cs, 90Sr, 239+240Pu and 241Am by both gamma spectroscopy and, through a contractor laboratory, by wet chemistry procedures. The samples processing methods, the analytical methods and the analytical quality control are all procedures developed for the continuing Marshall Island radioecology and dose assessment work

  13. Analysis of the chondroitin sulfate proteoglycan core protein (CSPGCP) gene in achondroplasia and pseudoachondroplasia.

    OpenAIRE

    Finkelstein, J E; Doege, K; Yamada, Y; Pyeritz, R E; Graham, J M; Moeschler, J.B.; Pauli, R. M.; Hecht, J T; Francomano, C A

    1991-01-01

    Achondroplasia and pseudoachondroplasia are autosomal dominant skeletal dysplasias resulting in short-limbed dwarfism. Histologic and ultrastructural studies of the cartilage in pseudoachondroplasia and in homozygous achondroplasia have suggested a structural abnormality in chondroitin sulfate proteoglycan (CSPG), a major structural protein in the extra-cellular matrix. The gene encoding CSPG core protein (CSPGCP) is thus a logical "candidate gene" for analysis in these conditions. cDNA probe...

  14. Tank 241-TX-113 rotary mode core sampling and analysis plan

    International Nuclear Information System (INIS)

    This sampling and analysis plan (SAP) identities characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for push mode core samples from tank 241-TX-113 (TX-113). The Tank Characterization Technical Sampling Basis document identities Retrieval, Pretreatment and Immobilization as an issue that applies to tank TX-113. As a result, a 150 gram composite of solids shall be made and archived for that program. This tank is not on a Watch List

  15. Lumped parameter analysis of Pb-Bi cooled fast reactor PEACER core using MATRA

    International Nuclear Information System (INIS)

    A lead-bismuth (Pb-Bi) eutectic cooled fast reactor PEACER (Proliferation-resistant, Environment-friendly, Accident-tolerant, Continuable-energy and Economical Reactor) is under development at the Seoul National University. This study is intended to examine the liquid metal coolant behavior along the subchannels and to find out whether the given flux profiles and geometrical arrangement of fuel rods yield reasonable flow distribution during nominal operation using the subchannel analysis code MATRA (Multi-channel Analyzer for Transient and steady-state in Rod Arrays). MATRA was developed at the Korea Atomic Energy Research Institute based on the subchannel approach to calculate the enthalpy and flow distribution in fuel rod bundle elements for both steady-state and transient conditions. The best-estimate analysis was carried out utilizing MATRA for the PEACER-300 quadrant core under the nominal operation condition. Subchannel analysis was performed for the hottest assembly of the PEACER-300 core. The calculation result showed that during normal operation the core material temperature distribution stays well below the thermal design limits. Comparison of the code results with those by hand calculation resulted in good agreement. Hand calculations are in further progress to include the finite difference scheme in the radial direction

  16. Simple Sequence Repeat Analysis of Genetic Diversity in Primary Core Collection of Peach (Prunus persica)

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    In this study, the genetic diversity of 51 cultivars in the primary core collection of peach (Prunus persica (L.) Batsch) was evaluated by using simple sequence repeats (SSRs). The phylogenetic relationships and the evolutionary history among different cultivars were determined on the basis of SSR data. Twenty-two polymorphic SSR primer pairs were selected, and a total of 111 alleles were identified in the 51 cultivars, with an average of 5 alleles per locus. According to traditional Chinese classification of peach cultivars, the 51 cultivars in the peach primary core collection belong to six variety groups. The SSR analysis revealed that the levels of the genetic diversity within each variety group were ranked as Sweet peach > Crisppeach > Flat peach > Nectarine > Honey Peach > Yellow fleshed peach. The genetic diversity among the Chinese cultivars was higher than that among the introduced cultivars. Cluster analysis by the unweighted pair group method with arithmetic averaging (UPGMA)placed the 51 cultivars into five linkage clusters. Cultivar members from the same variety group were distributed in different UPGMA clusters and some members from different variety groups were placed under the same cluster. Different variety groups could not be differentiated in accordance with SSR markers. The SSR analysis revealed rich genetic diversity in the peach primary core collection, representative of genetic resources of peach.

  17. Application of noise analysis for the study of core local instability at Forsmark 1

    International Nuclear Information System (INIS)

    Core local instability was recently experienced at Forsmark 1 BWR. The event has been studied by applying noise analysis to data collected in January 1997 for the stability test. The result indicated that there was a region in the left corner of the core which was subject to instability due to neutronic and thermal-hydraulic coupling. The result of the noise analysis suggested two types of disturbance source, one in the vicinity of the detector string LPRM10 having resonant oscillation at 0.5 Hz and another relatively wide band noise in the neighbourhood of LPRM18. Three hypotheses have been examined as the possible cause, operational factor, abnormal fuel assembly, and wide band low frequency disturbance. Although the real cause has not been made clear from the noise analysis, it is likely that the operational factor played an important role as the cause. Further investigations are expected to be performed in the future. In order to detect the local instability it is important to have a stability monitor with a capability of monitoring a sufficient number of LPRMs so as to cover the whole core. This is important since local instability is a type of anomaly which should not occur during reactor operation

  18. Analysis and synthesis of the theoretical studies performed on the control and safety of LWR's burning plutonium fuel

    International Nuclear Information System (INIS)

    This report presents the comparative investigations of parameters for plutonium fuelled power stations (PWR and BWR) under steady state and dynamic conditions for typical accidents. The recycling of about 30% of mixed oxide fuel in the large LWR cores should not induce special problems, if some cautions are taken in core design to minimize the differences with UO2 cores taking into account a limited margin fo uncertainty. The influence on the core behaviour, during the investigated accidents, is not very important and does not induce restrictions for at least a 30% Pu fraction in the core. The operation with high plutonium amounts may be considered. From the steady state and safety point-of-views, the maximum allowable quantity into the cores should be sought for each reactor. In principle, a 100% UO2-PuO2 core could be operated under certain conditions of loading pattern and shutdown margins. For what concerns the storage and handling, the studies show the following results: storage pool design with respect to criticality will not be affected by the use of UO2-PuO2 fuel asemblies

  19. Analysis of a German BWR core with TRACE/PARCS using different cross section sets

    International Nuclear Information System (INIS)

    'Full text:' Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools as well as appropriate cross sections (XS) libraries depending on the individual thermal hydraulic state parameters. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running and user-friendly lattice codes such as Casco and Helios. At research institutions and universities, alternative tools to the commercial ones with full access to the source code as well as moderate cost are urgently needed. The Scale system is being developed and improved for lattice physics calculations of real core loading of Light Water Reactors (LWR). It represents a promising alternative to the commercial lattice codes. At Karlsruhe Institute of Technology (Kit) a computational route based on Scale/Triton/Newt for BWR core loading is under development. The generated XS-data sets have to be transformed in PMAXS-format for use in the reactor dynamic code PARCS. This task is performed by the module GenPMAXS being developed and tested at the Michigan University. To verify the computational route, a BWR fuel assembly depletion problem was calculated by PARCS and compared to the CASMO results. Since the SCALE/TRITON XS-file does actually not contain all required neutronic data, FORTRAN routines have been developed to incorporate the missing data e.g. the yields of Iodine, Xenon and Promethium into the XS-data sets in the PMAXS-format. The comparison of the results obtained with PARCS (using the corrected PMAXS file) and CASMO for the depletion problem exhibited a good agreement. Consequently, this approach was followed for the generation of a complete XS-set for a real BWR core to be used in subsequent transient analysis. Then 3D neutronic and thermal hydraulic core model were elaborated for a TRACE/PARCS analysis. The thermal hydraulic model is based on the 3D VESSEL

  20. Lumped parameter analysis of Pb-Bi cooled fast reactor PEACER core using MATRA

    International Nuclear Information System (INIS)

    PEACER core is designed to produce 1560MW of thermal output with electric output of 550MW. PEACER uses B4C control rods and lead-bismuth (Pb-Bi) coolant in the primary system. This work examines the Pb-Bi coolant behavior along the PEACER fuel channels and to check on whether the given heat flux profiles and geometrical arrangement of the fuel rods yield reasonable fluid dynamic distribution under nominal operation resorting to a subchannel approach using MATRA. MATRA is a thermal hydraulic analysis code based on the subchannel approach for calculating the enthalpy and flow distribution in the fuel rod bundle during steady-state and transient conditions. The calculational result revealed that the input data based on the current design of PEACER core yielded reasonable results mostly satisfying the thermal design limits. The calculation results, however, indicated a potential for fuel damage in the hottest assembly of the core. This was found to be mainly due to excessively conservative assumptions made in generating the input conditions. Work is underway to apply physically-based conditions of the PEACER core and more reliable rod-to-coolant heat transfer correlations. (author)

  1. Hydrodynamic and elastoplastic structural analysis of fast breeder reactor core accident

    International Nuclear Information System (INIS)

    This paper describes the principles and examples of applications of an explicit Lagrangian coupled finite difference-finite element code HEMP-ESI developed in order to calculate the structural consequences of hypothetical core disruptive accidents (HCDA) in nuclear reactors. The explicit solution algorithm of the finite difference scheme used to discretize the hydrodynamic fluid domains is shown to be very similar to that used for the solution of the finite element discretized shell structures, hence permitting an easy and efficient coupling. Two examples of simulation show the applicability of the method to nuclear reactor core safety analysis (test problem). Core explosion in a loop-type reactor including a shell containment: the calculation shows the energy absorbing function of the shell and enables the evaluation of the forces acting on the reactor containment. Hypothetical Core Disruptive Accident in a fast breeder reactor: the calculation shows the main features of this accident: lifting of the liquid sodium above the explosion and impact on the cover head inducing upward deformations; radial outflow of the sodium which induces large deformations of the inner and outer shell; zones of compressive circumferential stresses in the main shell at the junction of the spherical head and the cylindrical part

  2. Development of the enigma fuel performance code for whole core analysis and dry storage assessments

    Energy Technology Data Exchange (ETDEWEB)

    Rositer, Glyn [Fuel Cycle Solutions, UK National Nuclear Laboratory, Lancashire (United Kingdom)

    2011-11-15

    UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage-this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

  3. Reflooding of a severely damaged reactor core. Experimental analysis and modelling

    International Nuclear Information System (INIS)

    The understanding of the reflood process of a severely damaged reactor core represents a challenge in the prediction of safety margin of existing and future pressurized water reactors. After the TMI-2 accident, the understanding of coolability of severely damaged reactor core became an objective of many theoretical and experimental studies. Currently, the French Institute of Radioprotection and Nuclear Safety (IRSN) has started two experimental programs, PRELUDE and PEARL, to investigate the physical phenomena during a reflood process at high temperature and to provide relevant data in order to improve predictive models. The purpose of this paper is to propose a consistent thermo-hydraulic model of reflood of severely damaged reactor core. The presented model is based on the theory of heat transfer and two-phase flow in porous media and in small hydraulic diameter channels. The proposed model is implemented into the European computer code for severe accident analysis ICARE-CATHARE. The comparison of the calculations with PRELUDE experimental results is presented. Finally, the issue of transposition to the reactor scale is discussed and some answers are proposed using calculation results for a debris bed in a configuration similar to what could be expected in a severely damaged reactor core. (author)

  4. Safety analysis for core conversion (from HEU to LEU) of Pakistan research reactor-2 (PARR-2)

    International Nuclear Information System (INIS)

    PARR-2 (Pakistan Research Reactor-2), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (High Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. The present report presents thermal hydraulic and safety analyses of PARR-2 using existing HEU fuel as well as proposed LEU fuel. Presently, the core is comprised of 90.2% enriched UAl4-Al fuel. There are 344 fuel pins of 5.5 mm diameter. The core has a total of 994.8 g of U235. Standard computer code PARET/ANL (version 1992) was employed to perform steady-state and transient analyses. Various parameters were computed, which included: coolant outlet, maximum clad surface and maximum fuel centerline temperatures; and peak power and corresponding peak core temperatures resulting from a transient initiated by 4 mK positive reactivity insertion. Results were compared with the reported data in Final Safety Analysis Report (FSAR). It was found that the PARET results were in reasonable agreement with the manufacturer's results. Calculations were also carried out for the proposed LEU core with two suggested fuel pin sizes (5.5 mm and 5.1 mm diameter with 12.6% and 12.3% enrichment, respectively). Comparison of the LEU results with the existing HEU fuel has been made and discussed.

  5. Development and performance analysis of EPICS channel access server on FPGA based soft-core processor

    International Nuclear Information System (INIS)

    A soft core processor is a flexible hardware description language (HDL) model of a specific processor (CPU) that can be customized for a given application and synthesized for an FPGA as opposed to a hard core processor which is fixed in silicon. Combined with an on-board ethernet port, the technology incorporates integrating the IOC and digital control hardware within a single FPGA thus reducing the overall hardware complexities of field devices. In this paper, the technical details of porting EPICS Channel Access Server on MicroBlaze soft-core processor are explained. The EPICS performance on the MicroBlaze processor is analyzed. For this, the CPU load and server processing time for different numbers of Process Variables (PVs) have been studied for this platform. On the basis of the analysis, critical parameters of EPICS on this embedded platform have been derived and a few modifications in the channel access protocol are proposed for MicroBlaze soft-core processor. (author)

  6. Development of the enigma fuel performance code for whole core analysis and dry storage assessments

    International Nuclear Information System (INIS)

    UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage-this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

  7. Analysis of core damage frequency, Surry, Unit 1 internal events appendices

    International Nuclear Information System (INIS)

    This document contains the appendices for the accident sequence analyses of internally initiated events for the Surry Nuclear Station, Unit 1. This is one of the five plant analyses conducted as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 documents the risk of a selected group of nuclear power plants. The work performed is an extensive reanalysis of that published in November 1986 as NUREG/CR-4450, Volume 3. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved. The context and detail of this report are directed toward PRA practitioners who need to know how the work was performed and the details for use in further studies. The mean core damage frequency at Surry was calculated to be 4.0E-5 per year, with a 95% upper bound of 1.3E-4 and 5% lower bound of 6.8E-6 per year. Station blackout type accidents (loss of all AC power) were the largest contributors to the core damage frequency, accounting for approximately 68% of the total. The next type of dominant contributors were Loss of Coolant Accidents (LOCAs). These sequences account for 15% of core damage frequency. No other type of sequence accounts for more than 10% of core damage frequency

  8. Performance analysis of saturated iron core superconducting fault current limiter using Jiles–Atherton hysteresis model

    International Nuclear Information System (INIS)

    In this paper study of the Saturated Iron Core Superconducting Fault Current Limiter (SISFCL) has been carried out. Since in an SISFCL, the iron core plays a key role in distributing the magnetic flux, the hysteresis property of the core material has been introduced in a mathematical model to get a more accurate result. In this paper the Jiles–Atherton hysteresis model has been used for modeling the core. The equations are solved through numerical method and performances of SISFCL are analyzed for both normal and fault conditions. On further analysis it is observed that for suppression of higher value of fault current a high voltage develops across the DC source. Hence there is a chance of the DC source being damaged by the rise in voltage under fault condition. In order to protect the DC source, a shorted ring is introduced in the SISFCL circuit and its effects have been analyzed. It is noticed that the shorted ring has successfully reduced the voltage across the DC coil during fault condition while the performance of the limiter remains the same. - Highlights: • Mathematical modeling of SISFCL has been established. • Hysteresis effect has been included using J–A model. • Mathematical model of SISFCL using the Shorted Ring have been established. • Comparisons are drawn between the responses with and without shorted ring

  9. The accuracy of frozen section analysis in ultrasound- guided core needle biopsy of breast lesions

    Directory of Open Access Journals (Sweden)

    Riss Paul

    2009-09-01

    Full Text Available Abstract Background Limited data are available to evaluate the accuracy of frozen section analysis and ultrasound- guided core needle biopsy of the breast. Methods In a retrospective analysis data of 120 consecutive handheldultrasound- guided 14- gauge automated core needle biopsies (CNB in 109 consecutive patients with breast lesions between 2006 and 2007 were evaluated. Results In our outpatient clinic120 CNB were performed. In 59/120 (49.2% cases we compared histological diagnosis on frozen sections with those on paraffin sections of CNB and finally with the result of open biopsy. Of the cases 42/59 (71.2% were proved to be malignant and 17/59 (28.8% to be benign in the definitive histology. 2/59 (3.3% biopsies had a false negative frozen section result. No false positive results of the intraoperative frozen section analysis were obtained, resulting in a sensitivity, specificity and positive predicting value (PPV and negative predicting value (NPV of 95%, 100%, 100% and 90%, respectively. Histological and morphobiological parameters did not show up relevance for correct frozen section analysis. In cases of malignancy time between diagnosis and definitive treatment could not be reduced due to frozen section analysis. Conclusion The frozen section analysis of suspect breast lesions performed by CNB displays good sensitivity/specificity characteristics. Immediate investigations of CNB is an accurate diagnostic tool and an important step in reducing psychological strain by minimizing the period of uncertainty in patients with breast tumor.

  10. The accuracy of frozen section analysis in ultrasound- guided core needle biopsy of breast lesions

    International Nuclear Information System (INIS)

    Limited data are available to evaluate the accuracy of frozen section analysis and ultrasound- guided core needle biopsy of the breast. In a retrospective analysis data of 120 consecutive handheldultrasound- guided 14- gauge automated core needle biopsies (CNB) in 109 consecutive patients with breast lesions between 2006 and 2007 were evaluated. In our outpatient clinic120 CNB were performed. In 59/120 (49.2%) cases we compared histological diagnosis on frozen sections with those on paraffin sections of CNB and finally with the result of open biopsy. Of the cases 42/59 (71.2%) were proved to be malignant and 17/59 (28.8%) to be benign in the definitive histology. 2/59 (3.3%) biopsies had a false negative frozen section result. No false positive results of the intraoperative frozen section analysis were obtained, resulting in a sensitivity, specificity and positive predicting value (PPV) and negative predicting value (NPV) of 95%, 100%, 100% and 90%, respectively. Histological and morphobiological parameters did not show up relevance for correct frozen section analysis. In cases of malignancy time between diagnosis and definitive treatment could not be reduced due to frozen section analysis. The frozen section analysis of suspect breast lesions performed by CNB displays good sensitivity/specificity characteristics. Immediate investigations of CNB is an accurate diagnostic tool and an important step in reducing psychological strain by minimizing the period of uncertainty in patients with breast tumor

  11. Development of fuel performance code FEMAXI-6 and analysis of mechanical loading on cladding during power ramp for high burn-up fuel rod

    International Nuclear Information System (INIS)

    A fuel performance code FEMAXI-6 has been developed for the analysis of LWR fuel rod behaviors in normal operation and transient (not accident) conditions. The code uses FEM for mechanical analysis, and has incorporated thermal and mechanical models of phenomena anticipated in high burn-up fuel rods, such as fuel thermal conductivity degradation and pellet-clad bonding. In the present study, PCMI induced by swelling in a high burn-up BWR type fuel rod has been analyzed by the FEMAXI-6 code. During a power ramp for the high burn-up fuel, instantaneous pellet swelling can significantly exceed the level that is predicted by a 'steady-rate' swelling model, causing a large circumferential strain in cladding. This phenomenon has been simulated by a new swelling model to take into account the fission gas bubble growth, and we found that the new model can give satisfactory predictions on cladding diametral expansion in comparison with measurements in test rod. The bubble growth model assumes an equilibrium between bubble size and external pressure on the bubble, and simultaneous solution is obtained with both bubble size determination and diffusion equation of fission gas atoms. In addition, a pellet-clad bonding model which assumes firm mechanical coupling between pellet outer surface and cladding inner surface predicts an elevated tensile stress in the axial direction of cladding during ramp, indicating the generation of bi-axial stress state in the cladding. These analyses by the FEMAXI-6 code enable us to predict the magnitude of mechanical loading on cladding during transient and also serve for failure investigation. Clearly, prediction by code calculation depends on the creep and stress-strain properties of highly irradiated cladding. (author)

  12. Validation of the burn-up code EVOLCODE 2.0 with PWR experimental data and with a Sensitivity/Uncertainty analysis

    International Nuclear Information System (INIS)

    Highlights: • A successful validation of the burn-up simulation system EVOLCODE is presented here. • A Sensitivity/Uncertainty model was applied for uncertainty propagation/assessment. • Cross sections are for most cases the main contributors to inventory uncertainties. • The improved model helps to explain some simulation-experiment discrepancies. • Some hints for the improvement of basic data libraries are provided. - Abstract: A validation of the burn-up simulation system EVOLCODE 2.0 is presented here, involving the experimental measurement of U and Pu isotopes and some fission fragments production ratios after a burn-up of around 30 GWd/tU in a Pressurized Light Water Reactor (PWR). This work provides an in-depth analysis of the validation results, including the possible sources of the uncertainties. An uncertainty analysis based on the sensitivity methodology has been also performed, providing the uncertainties in the isotopic content propagated from the cross sections uncertainties. An improvement of the classical Sensitivity/Uncertainty (S/U) model has been developed to take into account the implicit dependence of the neutron flux normalization, that is, the effect of the constant power of the reactor. The improved S/U methodology, neglected in this kind of studies, has proven to be an important contribution to the explanation of some simulation-experiment discrepancies for which, in general, the cross section uncertainties are, for the most relevant actinides, an important contributor to the simulation uncertainties, of the same order of magnitude and sometimes even larger than the experimental uncertainties and the experiment-simulation differences. Additionally, some hints for the improvement of the JEFF3.1.1 fission yield library and for the correction of some errata in the experimental data are presented

  13. Steady-state and transient core feasibility analysis for a thorium-fuelled reduced-moderation PWR performing full transuranic recycle

    International Nuclear Information System (INIS)

    Highlights: • We present a core analysis for a thorium-transuranic fuelled reduced-moderation PWR. • There is the possibility of positive reactivity in severe large break LOCAs. • Mechanical shim is used to control reactivity within power peaking constraints. • Adequate shutdown margin can be achieved with B4C control rods are required. • The response to a rod ejection accident is within likely licensing limits. - Abstract: It is difficult to perform multiple recycle of transuranic (TRU) isotopes in PWRs as the moderator temperature coefficient (MTC) tends to become positive after a few recycles and the core may have positive reactivity when fully voided. Due to the favourable impact on the MTC fostered by use of thorium (Th), the possibility of performing Th–TRU multiple-recycle in reduced-moderation PWRs (RMPWRs) is under consideration. Heterogeneous fuel design with spatial separation of Th–U and Th–TRU is necessary to improve neutronic performance. This can take the form of a heterogeneous fuel assembly (TPUC), or whole assembly heterogeneity (WATU). Satisfactory discharge burn-up can be maintained while ensuring negative MTC, with the pin diameter of a standard PWR increased from 9.5 to 11 mm. However, the reactivity becomes positive when the coolant density in the core becomes extremely low. This could lead to positive reactivity in some loss of coolant accident (LOCA) scenarios, for example a surge line break, if the reactor does not trip. To protect against this beyond design basis accident, a second redundant set of shutdown rods is added to the reactor, so that either the usual or secondary rods can trip the reactor when there is zero coolant in the core. Even so, this condition is likely to be concerning from a regulatory standpoint. Reactivity control is a key challenge due to the reduced worth of neutron absorbers and their detrimental effect on the void coefficients, especially when diluted, as is the case for soluble boron. Mechanical

  14. An approach to model reactor core nodalization for deterministic safety analysis

    Science.gov (United States)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  15. An approach to model reactor core nodalization for deterministic safety analysis

    International Nuclear Information System (INIS)

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M

  16. An approach to model reactor core nodalization for deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my [Nuclear Energy Department, Regulatory Economics & Planning Division, Tenaga Nasional Berhad (Malaysia); Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my [Prototypes & Plant Development Center, Malaysian Nuclear Agency (Malaysia); Roslan, Ridha, E-mail: ridha@aelb.gov.my; Sadri, Abd Aziz [Nuclear Installation Divisions, Atomic Energy Licensing Board (Malaysia); Farid, Mohd Fairus Abd [Reactor Technology Center, Malaysian Nuclear Agency (Malaysia)

    2016-01-22

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  17. Shuffling strategy study of breeding-burning integrated fast reactor

    International Nuclear Information System (INIS)

    The breeding-burning integrated fast reactor uses burning assemblies to generate thermal power, meanwhile, converts 238U into 239Pu in the fertile assemblies. With periodical shuffling of assemblies, the reactor can maintain criticality for decades of years. To maintain long-term stability of the core reactivity, the core layout and shuffling strategy should balance the burning and the breeding of the assemblies. The scattered core layout and shuffling strategy ensures fast breeding of the fertile assemblies, and keeps stable core power distribution in whole life of the reactor. Moreover, at the end of the reactor life, the discharge burnups of different fuel assemblies are close to each other, which are about 250300 GW · d/t. This is important for breeding-burning integrated fast reactor to achieve very efficient utilization of uranium resource without reprocessing. (authors)

  18. [Evaluation of NMR relaxation method as a diagnostic tool for donor blood analysis and patients with hematologic diseases and burns].

    Science.gov (United States)

    Gangardt, M G; Popova, O V; Shmarov, D A; Kariakina, N F; Papish, E A; Kozinets, G I

    2002-08-01

    Diagnostic value of the NMR-relaxation method in the blood plasma was estimated in the patients with different pathologies. The time of hydrogen nuclei longitudinal relaxation (T1) in the health donors of the blood, in the patients with oncopathology (hemoblastoses) and in the cases with anemia and burning disease were investigated. The time of the longitudinal relaxation (T1) was measured by automated NMR-relaxometer "Palma" (Russia). The working frequency was equal to 35 MHz, the temperature was 45 +/- 0.1 degrees C. For the single measurement 0.2 ml of blood obtained from heparinized venous blood 1.5 hours after its taking was used. The time of the longitudinal relaxation (T1) was shown to be 1.78 +/- 0.02 in the health donors, 1.70 +/- 0.06 s in cases with anemia, 1.97 +/- 0.48 c in patients with leucosis, 2.40 +/- 0.12 s in patients with burns. The sensitivity and the specificity of diagnostics of leucosis based upon the results of the only single T1 measurement in blood plasma were concluded to be 75%. It proves the significant T1 change both in patients with anemia and burning disease of the II-III degree. However it is evidently insufficient for selective use of NMR-relaxation blood plasma (serum) in the diagnostics of anemia and leucosis. The data obtained prove also the possibility of use of NMR-relaxation blood plasma (serum) for control of the hemostasis state during treatment or remission. PMID:12362635

  19. Pan-genome sequence analysis using Panseq: an online tool for the rapid analysis of core and accessory genomic regions

    Directory of Open Access Journals (Sweden)

    Villegas Andre

    2010-09-01

    Full Text Available Abstract Background The pan-genome of a bacterial species consists of a core and an accessory gene pool. The accessory genome is thought to be an important source of genetic variability in bacterial populations and is gained through lateral gene transfer, allowing subpopulations of bacteria to better adapt to specific niches. Low-cost and high-throughput sequencing platforms have created an exponential increase in genome sequence data and an opportunity to study the pan-genomes of many bacterial species. In this study, we describe a new online pan-genome sequence analysis program, Panseq. Results Panseq was used to identify Escherichia coli O157:H7 and E. coli K-12 genomic islands. Within a population of 60 E. coli O157:H7 strains, the existence of 65 accessory genomic regions identified by Panseq analysis was confirmed by PCR. The accessory genome and binary presence/absence data, and core genome and single nucleotide polymorphisms (SNPs of six L. monocytogenes strains were extracted with Panseq and hierarchically clustered and visualized. The nucleotide core and binary accessory data were also used to construct maximum parsimony (MP trees, which were compared to the MP tree generated by multi-locus sequence typing (MLST. The topology of the accessory and core trees was identical but differed from the tree produced using seven MLST loci. The Loci Selector module found the most variable and discriminatory combinations of four loci within a 100 loci set among 10 strains in 1 s, compared to the 449 s required to exhaustively search for all possible combinations; it also found the most discriminatory 20 loci from a 96 loci E. coli O157:H7 SNP dataset. Conclusion Panseq determines the core and accessory regions among a collection of genomic sequences based on user-defined parameters. It readily extracts regions unique to a genome or group of genomes, identifies SNPs within shared core genomic regions, constructs files for use in phylogeny programs

  20. The Correlation Between the Burning Features, the Burning Agent and Motivation in Burn Victims Attending Shahid Motahari Hospital in Tehran During 2009: letter to Editor

    Directory of Open Access Journals (Sweden)

    Kamran Aghakhani M.D.

    2011-06-01

    respiratory tract involvement. The results of logistic regression analysis revealed that age more than 50 years, female sex, burn area greater than 10% (previous studies 40%, 2 third degree burns, respiratory tract involvement and suicide increased the risk of death in burn patients.

  1. Burn wound coverage and burn wound closure

    Czech Academy of Sciences Publication Activity Database

    Konigová, R.; Matoušková, Eva; Brož, L.

    2000. s. 9. [International Symposium and Course on Burns and Fire Desaster Management. Jerusalem Meeting /3./. 13.02.2000-16.02.2000, Jerusalem] R&D Projects: GA MZd IZ4368 Subject RIV: EB - Genetics ; Molecular Biology

  2. MORBIDITY AND SURVIVAL PROBABILITY IN BURN PATIENTS IN MODERN BURN CARE

    Science.gov (United States)

    Jeschke, Marc G.; Pinto, Ruxandra; Kraft, Robert; Nathens, Avery B.; Finnerty, Celeste C.; Gamelli, Richard L.; Gibran, Nicole S.; Klein, Matthew B.; Arnoldo, Brett D.; Tompkins, Ronald G.; Herndon, David N.

    2014-01-01

    Objective Characterizing burn sizes that are associated with an increased risk of mortality and morbidity is critical because it would allow identifying patients who might derive the greatest benefit from individualized, experimental, or innovative therapies. Although scores have been established to predict mortality, few data addressing other outcomes exist. The objective of this study was to determine burn sizes that are associated with increased mortality and morbidity after burn. Design and Patients Burn patients were prospectively enrolled as part of the multicenter prospective cohort study, Inflammation and the Host Response to Injury Glue Grant, with the following inclusion criteria: 0–99 years of age, admission within 96 hours after injury, and >20% total body surface area burns requiring at least one surgical intervention. Setting Six major burn centers in North America. Measurements and Main Results Burn size cutoff values were determined for mortality, burn wound infection (at least two infections), sepsis (as defined by ABA sepsis criteria), pneumonia, acute respiratory distress syndrome, and multiple organ failure (DENVER2 score >3) for both children (<16 years) and adults (16–65 years). Five-hundred seventy-three patients were enrolled, of which 226 patients were children. Twenty-three patients were older than 65 years and were excluded from the cutoff analysis. In children, the cutoff burn size for mortality, sepsis, infection, and multiple organ failure was approximately 60% total body surface area burned. In adults, the cutoff for these outcomes was lower, at approximately 40% total body surface area burned. Conclusions In the modern burn care setting, adults with over 40% total body surface area burned and children with over 60% total body surface area burned are at high risk for morbidity and mortality, even in highly specialized centers. PMID:25559438

  3. [Quantification of crop residue burned areas based on burning indices using Landsat 8 image].

    Science.gov (United States)

    Ma, Jian-hang; Song, Kai-shar; Wen, Zhi-dan; Shao, Tian-tian; Li, Bo-nan; Qi, Cai

    2015-11-01

    Crop residue burning leads to atmospheric pollution and is an enormous waste of crop residue resource. Crop residue burning can be monitored timely in large regions as the fire points can be recognized through remotely sensed image via thermal infrared bands. However, the area, the detailed distribution pattern and especially the severity of the burning areas cannot be derived only by the thermal remote sensing approach. The burning index, which was calculated with two or more spectral bands at where the burned and unburned areas have distinct spectral characteristics, is widely used in the forest fire investigation. However its potential application for crop residue burning evaluation has not been explored. With two Landsat 8 images that cover a part of the Songnen Plain, three burning indices, i.e., the normalized burned ratio (NBR), the normalized burned ratio incorporating the thermal band (NBRT), and the burned area index (BAI), were used to classify the crop residue burned and unburned areas. The overall classification accuracies were 91.9%, 92.3%, and 87.8%, respectively. The correlation analysis between the indices and the crop residue coverage indicated that the NBR and NBRT were positively correlated with the crop residue coverage (R2 = 0.73 and 0.64, respectively) with linear regression models, while the BAI was exponentially correlated with the crop residue coverage (R2 = 0.68). The results indicated that the use of burning indices in crop residue burning monitoring could quantify crop residue burning severity and provide valuable data for evaluating atmospheric pollution. PMID:26915202

  4. Analysis of measurements for a uranium-free core experiment at the BFS-2 critical assembly

    International Nuclear Information System (INIS)

    This document describes a series of calculations that were carried out to model various measurements from the BFS-58-1-I1 experiment. BFS-58-1-I1 was a mock-up of a uranium-free, Pu burning core at BFS-2, a Russian critical assembly operated by IPPE. The experiment measured values of Keff, Na void reactivity worth, material sample reactivity worths and reaction rate ratios. The experiments were modelled using a number of different methods. Basic nuclear data was taken from JENDL-3.2, in either 70 or 18 groups. Cross-section data for the various material regions of the assembly were calculated by either SLAROM or CASUP; the heterogeneous structure of the core regions was modelled in these calculations, with 3 different options considered for representing the (essentially 2D) geometry of the assembly components in a 1D cell model. Whole reactor calculations of flux and Keff were done using both a diffusion model (CITATION) and a transport model (TWOTRAN2), both using an RZ geometry. Reactivity worths were calculated both directly from differences in Keff values and by using the exact perturbation calculations of PERKY and SN-PERT (for CITATION and TWOTRAN2, respectively). Initial calculations included a number of inaccuracies in the assembly representation, a result of communication difficulties between JNC and IPPE; these errors were removed for the final calculations that are presented. Calculations for the experiments have also been carried out in Russia (IPPE) and France (CEA) as part of an international comparison exercise, some of those results are also presented here. The calculated value of Keff was 1.1%Δk/k higher than the measured value, Na void worth C/E values were ∼1.06; these results were considered to be reasonable. (Discrepancies in certain Na void values were probably due to experimental causes , though the effect should be investigated in any future experiments.) Several sample worth values were small compared with calculational uncertainties

  5. False-negative results of breast core needle biopsies – retrospective analysis of 988 biopsies

    International Nuclear Information System (INIS)

    Breast cancer is the most common malignant neoplasm and the most common cause of death among women. The core needle biopsy is becoming a universal practice in diagnosing breast lesions suspected of malignancy. Unfortunately, breast core needle biopsies also bear the risk of having false-negative results. 988 core needle breast biopsies were performed at the Maria Skłodowska-Curie Memorial Cancer Center and Institute of Oncology, Gliwice Branch, between 01 March 2006 and 29 February 2008. Malignant lesions were diagnosed in 426/988 (43.12%) cases, atypical hyperplasia in 69/988 (6.98%), and benign lesions in 493/988 (49.90%) cases. Twenty-two out of 988 biopsies (2.23%) were found to be false negative. Histopathological assessment of tissue specimens was repeated in these cases. In 14/22 (64%) cases, the previous diagnosis of a benign lesion was changed. In 8/22 (36%) cases, the diagnosis of a benign lesion was confirmed. False-negative rate was calculated at 2.2%. The rate of false-negative diagnoses resulting from a radiological mistake was estimated at 36%. The rate of false-negative diagnoses, resulting from histopathological assessment, was 64%. False-negative results caused by a radiological error comprised 1.5% of all histopathologically diagnosed cancers and atypias (sensitivity of 98.5%). There were no false-positive results in our material - the specificity of the method was 100%. Histopathological interpretation is a substantial cause of false-negative results of breast core needle biopsy. Thus, in case of a radiological-histopathological divergence, histopathological analysis of biopsy specimens should be repeated. The main radiological causes of false-negative results of breast core needle biopsy are as follows: sampling from an inappropriate site and histopathological non-homogeneity of cancer infiltration

  6. Analysis of containment venting following a core damage at a BWR Mark I using THALES-2

    International Nuclear Information System (INIS)

    Analysis of containment venting following a core damage at a boiling water reactor (BWR) Mark I using THALES-2 was performed. In this analysis, the effect of various parameters, namely, the areas of the vent path, containment venting pressure, and accident sequences on the containment thermodynamic response, and radionuclide transport and release in the containment venting at a BWR was examined. The code THALES-2B developed by the Japan Atomic Energy Research Institute (JAERI) was used in this analysis. The model plant in this analysis was the Browns Ferry plant. From this analysis was found that the 4-inch pipe of containment venting flow path is sufficient to maintain the containment pressure in the specified range if the containment was pressurized by the decay heat power. The entrainment by the pool swelling as well as by the flashing was not occurred during the containment venting. The source terms are not sensitive to the variation of containment venting flow path area. The containment venting pressure operation setting point has important rule in the containment venting. In the containment venting, the source terms are not sensitive to the accident sequence, except for Sr source term. In order to get better understanding on the containment venting strategy, the following analyses are necessary. Analyses of accident sequence which has a high power such as anticipated transient without scram are necessary, as well as analyses of accident sequence which pressurize the containment before the core damage. (author)

  7. [Effective core formulae for lung cancer based on complex network and survival analysis].

    Science.gov (United States)

    Yang, Ming; Li, Jia-qi; Jiao, Li-jing; Chen, Pei-qi; Xu, Ling

    2015-11-01

    The study on the effective core formulae (CEF) not only summarized traditional chinese medicine (TCM) treatment experience, but also helped reveal the underlying knowledge in the formulation of TCM prescriptions. The aim of the present paper was to investigate the method of data mining for the discovery of core effective formulae for lung cancer. In the present study, a prescription fingerprint approach was used to characterize the staged prescription information of patients. The D index was used to screen potential beneficial herbs. Then, based on a herbal compatibility network, the maximal clique searching algorithm (BK algorithm) and survival analysis were applied to discover CEF for lung cancer, and a mining analysis was made for the 322 cases from Longhua hospital. The correlation between prescriptions and survival time was analyzed by prescription fingerprints. Forty-three potentially beneficial herbs were obtained, and two CEFs were significant for the survival time by a parametric survival model based on lognormal distribution, the results were verified by a multivariate survival model. The rules of combination of the two CEFs basically conform to TCM onco-therapeutic theory of strengthening the body resistance and the actual conditions in clinic. All results showed that the established approach was feasible for discovering the core effective formulae for lung cancer and mining survival data for complex TCM onco-therapy. PMID:27097428

  8. Development of computer code packages for molten salt reactor core analysis

    International Nuclear Information System (INIS)

    This paper presents the implementations of the Oak Ridge National Laboratory (ORNL) approach for Molten Salt Reactor (MSR) core analysis with two nuclear reactor core analysis computer code systems. The first code system has been set up with the MCNP6 Monte Carlo code, its depletion module CINDER90 and the PYTHON script language. The second code system has been set up with the NEWT transport calculation module and ORIGEN depletion module connected by TRITON sequence in SCALE code, and the PYTHON script language. The PYTHON script language is used for implementing the online reprocessing of molten-salt fuel, and feeding new fertile material in the computer code simulations. In this paper, simplified nuclear reactor core models of a Molten Salt Breeder Reactor (MSBR), designed by ORNL in the 1960's, and FUJI-U3 designed by Toyohashi University of Technology (TUT) in the 2000's, were analyzed by the two code systems. Using these, various reactor design parameters of the MSRs were compared, such as the multiplication factor, breeding ratio, amount of material, total feeding, neutron flux distribution, and temperature coefficient. (author)

  9. Neutronic Analysis of the Pavia University TRIGA Mark II Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Tigliole, A. Borio Di [University of Pavia, Department of Nuclear and Theoretical Physics, Via Bassi 4, 27100 Pavia (Italy); University of Pavia, Laboratory of Applied Nuclear Energy, Via Aselli 41, 27100 Pavia (Italy); Cammi, A. [Polytechnic of Milano, Department of Energy, Via La Masa 34, 20156 Milano (Italy); Chiesa, D.; Clemenza, M. [University of Milano-Bicocca, Physics Department, Piazza della Scienza 3, 20126, Milano (Italy); Pattavina, L.; Previtali, E. [University of Milano-Bicocca, Physics Department, Piazza della Scienza 3, 20126, Milano (Italy); INFN section of Milano-Bicocca, Piazza della Scienza 3, 20126, Milano (Italy); Scian, G. [University of Pavia, Laboratory of Applied Nuclear Energy, Via Aselli 41, 27100 Pavia (Italy)

    2011-07-01

    An advanced computational tool based on Monte Carlo code was developed for the neutronic analysis of the TRIGA Mark II research reactor of the University of Pavia and the results from simulations were benchmarked with the experimental data. The neutron analysis was performed by means of the 3-D continuous-energy Monte Carlo codes MCNP4C and MCNP5. All core components were reproduced in good detail and the model allows the description of different core and fuel configurations with good versatility. In a first step, the results from the simulations with fresh fuel at room temperature were compared with the experimental data about critical core parameters at zero power level, quoted in the reactor First Criticality Final Report; a very good agreement was found and the goodness of the model was proved. Afterwards, neutron flux and reactivity dependence on the fuel and moderator temperature was studied in detail; all the available methods for the simulation of thermal effects were analyzed in order to model the reactor in the stationary operating condition at 250 kW. The model was updated to the present configuration and the use of specific S({alpha},{beta}) cross sections was found to be of primary importance for a correct simulation of low-energy neutrons scattering within moderator. (author)

  10. Neutronic Analysis of the Pavia University TRIGA Mark II Reactor Core

    International Nuclear Information System (INIS)

    An advanced computational tool based on Monte Carlo code was developed for the neutronic analysis of the TRIGA Mark II research reactor of the University of Pavia and the results from simulations were benchmarked with the experimental data. The neutron analysis was performed by means of the 3-D continuous-energy Monte Carlo codes MCNP4C and MCNP5. All core components were reproduced in good detail and the model allows the description of different core and fuel configurations with good versatility. In a first step, the results from the simulations with fresh fuel at room temperature were compared with the experimental data about critical core parameters at zero power level, quoted in the reactor First Criticality Final Report; a very good agreement was found and the goodness of the model was proved. Afterwards, neutron flux and reactivity dependence on the fuel and moderator temperature was studied in detail; all the available methods for the simulation of thermal effects were analyzed in order to model the reactor in the stationary operating condition at 250 kW. The model was updated to the present configuration and the use of specific S(α,β) cross sections was found to be of primary importance for a correct simulation of low-energy neutrons scattering within moderator. (author)

  11. Contributed Review: Nuclear magnetic resonance core analysis at 0.3 T

    International Nuclear Information System (INIS)

    Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general

  12. A 3D transport-based core analysis code for research reactors with unstructured geometry

    International Nuclear Information System (INIS)

    Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal SN method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of keff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

  13. Analysis of core damage frequency from internal events: Peach Bottom, Unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Kolaczkowski, A.M.; Lambright, J.A.; Ferrell, W.L.; Cathey, N.G.; Najafi, B.; Harper, F.T.

    1986-10-01

    This document contains the internal event initiated accident sequence analyses for Peach Bottom, Unit 2; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Peach Bottom, Unit 2, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provided additional insights regarding the dominant contributors to the Peach Bottom core damage frequency estimate. The mean core damage frequency at Peach Bottom was calculated to be 8.2E-6. Station blackout type accidents (loss of all ac power) were found to dominate the overall results. Anticipated Transient Without Scram accidents were also found to be non-negligible contributors. The numerical results are largely driven by common mode failure probability estimates and to some extent, human error. Because of significant data and analysis uncertainties in these two areas (important, for instance, to the most dominant scenario in this study), it is recommended that the results of the uncertainty and sensitivity analyses be considered before any actions are taken based on this analysis.

  14. Analysis of core damage frequency from internal events: Peach Bottom, Unit 2

    International Nuclear Information System (INIS)

    This document contains the internal event initiated accident sequence analyses for Peach Bottom, Unit 2; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Peach Bottom, Unit 2, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provided additional insights regarding the dominant contributors to the Peach Bottom core damage frequency estimate. The mean core damage frequency at Peach Bottom was calculated to be 8.2E-6. Station blackout type accidents (loss of all ac power) were found to dominate the overall results. Anticipated Transient Without Scram accidents were also found to be non-negligible contributors. The numerical results are largely driven by common mode failure probability estimates and to some extent, human error. Because of significant data and analysis uncertainties in these two areas (important, for instance, to the most dominant scenario in this study), it is recommended that the results of the uncertainty and sensitivity analyses be considered before any actions are taken based on this analysis

  15. Contributed Review: Nuclear magnetic resonance core analysis at 0.3 T

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell, Jonathan, E-mail: JMitchell16@slb.com; Fordham, Edmund J. [Schlumberger Gould Research, High Cross, Madingley Road, Cambridge CB3 0EL (United Kingdom)

    2014-11-15

    Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general.

  16. BGCore - A Comprehensive Package for Reactor Core and Fuel Storage Analysis

    International Nuclear Information System (INIS)

    Recent interest in Fast Gas Cooled Reactors requires major adaptations or evolutions of calculation tools to accommodate the innovative features of core design (new fuel and subassembly forms), fuel composition (homogeneous recycling of minor actinides). Fast neutron spectrum renders inadequate the familiar group reduction schemes and homogenization methods used in LWR analysis. In addition, the specificities of Fast Gas Cooled Reactors (materials, subassembly design, preferential direction for neutron leakage (streaming), high temperatures, particular reactivity effects, etc.) require at least an increase in the number of nuclides to be taken into account in the neutronic libraries with an extended tabulation in temperature. Enhancement of neutronic calculational tools is needed for S/A heterogeneity and anisotropy and to accurately model control elements and other non fueled regions. Several computational systems recently developed are widely used and several others are currently under development. All of the systems are based on Monte-Carlo codes for a 3 Dimensional representation of core and ORIGEN(8) code for fuel composition calculations. This abstract presents the outline and current progress of a development of a comprehensive calculational system for Fast Gas Cooled Reactors carried out in Ben-Gurion University. The flow chart of the system (temporarily) designated as BG-CORE is shown in Fig. 1. Our approach follows, in general, that adapted in other systems, i.e. interfacing the core model (MCNP) with a SARAF - an independently developed code for calculating fuel composition in-core and spent fuel emissions following discharge. Two novel features are added: temperature distributions and feedback, and fuel management. The following modules comprise the BG-CORE system: MCNP - Monte-Carlo code for 3 dimensional core representation using cross-section data sets based on JEF-2/JEF-3 and ENDFB-VI. This module provides Kerf, flux and power density distributions

  17. Low time resolution analysis of polar ice cores cannot detect impulsive nitrate events

    CERN Document Server

    Smart, D F; Melott, A L; Laird, C M

    2015-01-01

    Ice cores are archives of climate change and possibly large solar proton events (SPEs). Wolff et al. (2012) used a single event, a nitrate peak in the GISP2-H core, which McCracken et al. (2001a) time associated with the poorly quantified 1859 Carrington event, to discredit SPE-produced, impulsive nitrate deposition in polar ice. This is not the ideal test case. We critique the Wolff et al. analysis and demonstrate that the data they used cannot detect impulsive nitrate events because of resolution limitations. We suggest re-examination of the top of the Greenland ice sheet at key intervals over the last two millennia with attention to fine resolution and replicate sampling of multiple species. This will allow further insight into polar depositional processes on a sub-seasonal scale, including atmospheric sources, transport mechanisms to the ice sheet, post-depositional interactions, and a potential SPE association.

  18. Review of core disruptive accident analysis for liquid-metal cooled fast reactors

    International Nuclear Information System (INIS)

    Analysis methodologies of core disruptive accidents (CDAs) are reviewed. The role of CDAS in the overall safety evaluation of fast reactors has not always been well defined nor universally agreed upon. However, they have become a traditional issue in LMR safety, design, and licensing. The study is for the understanding of fast reactor behavior under CDA conditions to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features for the KALIMER developments. The methods used to analyze CDAs from initiating event to complete core disruption are described. Two examples of CDA analyses for CRBRP and ALMR are given and R and D needed for better understanding of CDA phenomena are proposed. (author). 10 refs., 2 tabs., 3 figs

  19. Joint European contribution to phase 4 of the BN-600 hybrid reactor benchmark core analysis

    International Nuclear Information System (INIS)

    Hybrid UOX/MOX fuelled core of BN-600 reactor was endorsed as an international benchmark. Phase 4 of the RCM benchmark studies consider full MOX core differentiated by design measures to reduce the sodium void worth. Parameters to be calculated were: fuel and steel Doppler coefficients; fuel density coefficient; sodium density coefficient; power distribution for fuel and non-fuelled regions; β effective and prompt neutron life time. Heterogeneity effects are evaluated. Analysis was carried out using ERANOS code and data system for fast reactors. Nuclear data library is based on JEF2.2. Accurate calculations of control rod heterogeneity effects with homogeneous equivalent cross sections for control rod absorbers were prepared using reactivity equivalence technique

  20. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    Energy Technology Data Exchange (ETDEWEB)

    Prabhu Gaunkar, N., E-mail: neelampg@iastate.edu; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C. [Department of Electrical and Computer Engineering, Iowa State University, Ames, Iowa 50011 (United States); Bulu, I.; Ganesan, K.; Song, Y. Q. [Schlumberger-Doll Research, Cambridge, Massachusetts 02139 (United States)

    2015-05-07

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors.