The Experts below are selected from a list of 33405 Experts worldwide ranked by ideXlab platform

Tewfik Hamidouche - One of the best experts on this subject based on the ideXlab platform.

  • analysis of loss of coolant accident in mtr pool type Research reactor
    Progress in Nuclear Energy, 2011
    Co-Authors: Tewfik Hamidouche, Elkhider Siahmed
    Abstract:

    Abstract In MTR Research Reactors, heat removal is, safely performed by forced convection during normal operation and by natural convection after reactor shutdown for residual decay heat removal. However, according to the duration time of operation at full power, it may be required to maintain the forced convection, for a certain period of time after the reactor shutdown. This is among the general requirements for the overall safety engineering features of MTR Research Reactors to ensure a safe residual heat removal. For instance, in safety analysis of Research Reactors, initiating events that may challenge the safe removal of residual heat must be identified and analyzed. In the present work, it was assumed a total loss of coolant accident in a typical MTR nuclear Research reactor with the objective of examining the core behavior and the occurrence of any fuel damage. For this purpose, the IAEA 10 MW benchmark core, which is a representative of medium power pool type MTR Research Reactors, was chosen herein in order to investigate the evolution of cladding temperature through the use of a best estimate thermalhydraulic system code RELAP5/mod3.2.

  • application of coupled code technique to a safety analysis of a standard mtr Research reactor
    Nuclear Engineering and Design, 2009
    Co-Authors: Tewfik Hamidouche, Anis Bousbiasalah, Elkhider Siahmed, Mohamed Y Mokeddem, F Dauria
    Abstract:

    Abstract Accident analyses in nuclear Research Reactors have been performed, up to now, using simple computational tools based on conservative physical models. These codes, developed to focus on specific phenomena in the reactor, were widely used for licensing purposes. Nowadays, the advances in computer technology make it possible to switch to a new generation of computational tools that provides more realistic description of the phenomena occurring in a nuclear Research reactor. Recent International Atomic Energy Agency (IAEA) activities have emphasized the maturity in using Best Estimate (BE) Codes in the analysis of accidents in Research Reactors. Indeed, some assessments have already been performed using BE thermal–hydraulic system codes such as RELAP5/Mod3. The challenge today is oriented to the application of coupled code techniques for Research Reactors safety analyses. Within the framework of the current study, a Three-Dimensional Neutron Kinetics Thermal–Hydraulic Model (3D-NKTH) based on coupled PARCS and RELAP5/Mod3.3 codes has been developed for the IAEA High Enriched Uranium (HEU) benchmark core. The results of the steady state calculations are sketched by comparison to tabulated results issued from the IAEA TECDOC 643. These data were obtained using conventional diffusion codes as well as Monte Carlo codes. On the other hand, the transient analysis was assessed with conventional coupled point kinetics–thermal–hydraulic channel codes such as RELAP5 stand alone, RETRAC-PC, and PARET codes. Through this study, the applicability of the coupled code technique is emphasized with an outline of some remaining challenges.

  • overview of accident analysis in nuclear Research Reactors
    Progress in Nuclear Energy, 2008
    Co-Authors: Tewfik Hamidouche, Anis Bousbiasalah, Elkhider Siahmed, Francesco Saverio Dauria
    Abstract:

    Advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. Nowadays, it becomes possible to switch to new generation of computational tools in order to get better realistic simulations of complex phenomena and transients. The challenge today is to revisit safety features of the existing Research Reactors in order to verify that the safety requirements still met and when necessary to introduce some amendments, coming from not only the new requirements but also, in order to introduce new equipments from recent advancement of new technologies. The objective of this work is to give an overview of the state of the art in performing safety analysis of Research Reactors and to emphasize the need and the provision to achieve such goals.

  • dynamic calculations of the iaea safety mtr Research reactor benchmark problem using relap5 3 2 code
    Annals of Nuclear Energy, 2004
    Co-Authors: Tewfik Hamidouche, Anis Bousbiasalah, M Adorni, F Dauria
    Abstract:

    Nowadays, increased attention to safety issues for nuclear Research Reactors has emerged as a consequence of their enlarged commercial exploitation. Almost all of the Research Reactors safety analyses were, so far, performed using conservative computational tools. Currently, the application of Best-Estimate methods constitutes a real necessity in order to get a more realistic vision of the system behavior, and overcome constraining limits related to conservative approaches. The aim of the current work is an attempt to apply this technique using the system thermal-hydraulic RELAP5/Mod3.2 code. For this purpose, the IAEA 10 MW MTR pool type Research Reactors Benchmark problem is considered. The exercise consists in performing some fast transients related to typical reactivity induced accidents, and relatively slow transients related to loss of flow accidents. The RELAP5 results were compared against previous data obtained by various conservative channel codes. Differences between the two modeling approaches are afterwards emphasized and discussed. � 2004 Elsevier Ltd. All rights reserved.

Francesco Saverio Dauria - One of the best experts on this subject based on the ideXlab platform.

  • thermal hydraulic analysis of reactivity accidents in mtr Research Reactors using relap5
    Kerntechnik, 2015
    Co-Authors: N Elsahlamy, A Khedr, Francesco Saverio Dauria
    Abstract:

    Abstract The present paper comes in the line with the international approach which use the best estimate codes, instead of conservative codes, to get more realistic prediction of system behavior under off-normal reactor conditions. The aim of the current work is to apply this approach using the thermal-hydraulic system code RELAP5/Mod3.3 in a reassessment of safety of the IAEA benchmark 10 MW Research Reactor. The assessment is performed for both slow and fast reactivity insertion transients at initial power of 1.0 W. The reactor power is calculated using the RELA5 point kinetic model. The reactivity feedback terms are considered in two steps. In the first step the feedback from changes in water density and fuel temperature (Doppler effects) are considered. In the second step the feedback from the water temperature changes is added. The results from the first step are compared with that published in IAEA-TECDOC-643 benchmarks. The comparison shows that RELAP5 over predicts the peak power and consequently ...

  • overview of accident analysis in nuclear Research Reactors
    Progress in Nuclear Energy, 2008
    Co-Authors: Tewfik Hamidouche, Anis Bousbiasalah, Elkhider Siahmed, Francesco Saverio Dauria
    Abstract:

    Advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. Nowadays, it becomes possible to switch to new generation of computational tools in order to get better realistic simulations of complex phenomena and transients. The challenge today is to revisit safety features of the existing Research Reactors in order to verify that the safety requirements still met and when necessary to introduce some amendments, coming from not only the new requirements but also, in order to introduce new equipments from recent advancement of new technologies. The objective of this work is to give an overview of the state of the art in performing safety analysis of Research Reactors and to emphasize the need and the provision to achieve such goals.

  • the effect of code user and boundary conditions on relap calculations of mtr Research reactor transient scenarios
    Nuclear Technology & Radiation Protection, 2005
    Co-Authors: Ahmed Khedr, M Adorni, Francesco Saverio Dauria
    Abstract:

    The safety evaluation of nuclear power and re search Reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR Research Reactors is presented in this pa per. To clarify this effect, two nodalizations for Research reactor different in the simulation of the open water surface boundary conditions of the reactor pool have been used. Very different results are obtained with few choices for code users. The core natural circulation flow with the be ginning of core boiling doesn't stop but in creases. The in creasing in the natural circulation flow shifts out the boiling from the core and the clad temperature decreases be low the local saturation temperature.

Elkhider Siahmed - One of the best experts on this subject based on the ideXlab platform.

  • analysis of loss of coolant accident in mtr pool type Research reactor
    Progress in Nuclear Energy, 2011
    Co-Authors: Tewfik Hamidouche, Elkhider Siahmed
    Abstract:

    Abstract In MTR Research Reactors, heat removal is, safely performed by forced convection during normal operation and by natural convection after reactor shutdown for residual decay heat removal. However, according to the duration time of operation at full power, it may be required to maintain the forced convection, for a certain period of time after the reactor shutdown. This is among the general requirements for the overall safety engineering features of MTR Research Reactors to ensure a safe residual heat removal. For instance, in safety analysis of Research Reactors, initiating events that may challenge the safe removal of residual heat must be identified and analyzed. In the present work, it was assumed a total loss of coolant accident in a typical MTR nuclear Research reactor with the objective of examining the core behavior and the occurrence of any fuel damage. For this purpose, the IAEA 10 MW benchmark core, which is a representative of medium power pool type MTR Research Reactors, was chosen herein in order to investigate the evolution of cladding temperature through the use of a best estimate thermalhydraulic system code RELAP5/mod3.2.

  • application of coupled code technique to a safety analysis of a standard mtr Research reactor
    Nuclear Engineering and Design, 2009
    Co-Authors: Tewfik Hamidouche, Anis Bousbiasalah, Elkhider Siahmed, Mohamed Y Mokeddem, F Dauria
    Abstract:

    Abstract Accident analyses in nuclear Research Reactors have been performed, up to now, using simple computational tools based on conservative physical models. These codes, developed to focus on specific phenomena in the reactor, were widely used for licensing purposes. Nowadays, the advances in computer technology make it possible to switch to a new generation of computational tools that provides more realistic description of the phenomena occurring in a nuclear Research reactor. Recent International Atomic Energy Agency (IAEA) activities have emphasized the maturity in using Best Estimate (BE) Codes in the analysis of accidents in Research Reactors. Indeed, some assessments have already been performed using BE thermal–hydraulic system codes such as RELAP5/Mod3. The challenge today is oriented to the application of coupled code techniques for Research Reactors safety analyses. Within the framework of the current study, a Three-Dimensional Neutron Kinetics Thermal–Hydraulic Model (3D-NKTH) based on coupled PARCS and RELAP5/Mod3.3 codes has been developed for the IAEA High Enriched Uranium (HEU) benchmark core. The results of the steady state calculations are sketched by comparison to tabulated results issued from the IAEA TECDOC 643. These data were obtained using conventional diffusion codes as well as Monte Carlo codes. On the other hand, the transient analysis was assessed with conventional coupled point kinetics–thermal–hydraulic channel codes such as RELAP5 stand alone, RETRAC-PC, and PARET codes. Through this study, the applicability of the coupled code technique is emphasized with an outline of some remaining challenges.

  • overview of accident analysis in nuclear Research Reactors
    Progress in Nuclear Energy, 2008
    Co-Authors: Tewfik Hamidouche, Anis Bousbiasalah, Elkhider Siahmed, Francesco Saverio Dauria
    Abstract:

    Advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. Nowadays, it becomes possible to switch to new generation of computational tools in order to get better realistic simulations of complex phenomena and transients. The challenge today is to revisit safety features of the existing Research Reactors in order to verify that the safety requirements still met and when necessary to introduce some amendments, coming from not only the new requirements but also, in order to introduce new equipments from recent advancement of new technologies. The objective of this work is to give an overview of the state of the art in performing safety analysis of Research Reactors and to emphasize the need and the provision to achieve such goals.

Nasir Ahmad - One of the best experts on this subject based on the ideXlab platform.

  • reactivity feedback and control margins in natural uranium fueled and heavy water moderated nuclear Research Reactors
    Progress in Nuclear Energy, 2006
    Co-Authors: Mohammad Javed Khan, Nasir Ahmad
    Abstract:

    Abstract Three-dimensional neutronics analyses were performed to find the reactivity feedback coefficients and control margins under normal and presumed accidental conditions for the novel proliferation resistant, safer and economical natural uranium fueled and heavy water moderated nuclear Research reactor cores [Annals of Nuclear Energy 31 (2004) 1331–1356 & 32 (2005) 612–620, Progress in Nuclear Energy (2006a,b) in press]. The results were compared with the reference core similar in design to National Experimental Reactor (NRX) and Canadian Indian Reactor (CIR). Standard reactor physics simulation codes WIMS-D/4 and CITATION were employed for this study. It was found that as opposed to the reference design, the new proposed cores are inherently stable as far as temperature reactivity feedback is concerned. Voids formation (loss of coolant) adds positive reactivity in all the cores. However, in the case of proposed cores, the reactivity added would be about 10% of that added in the reference core for loss of a given fraction of coolant. Moreover, as opposed to the NRX/CIR core, the shut-off rods, under the one stuck rod criterion, can shutdown and maintain the proposed cores subcritical in the accidental conditions.

  • effect of high density fuel loading on criticality of low enriched uranium fueled material test Research Reactors
    Annals of Nuclear Energy, 2005
    Co-Authors: Rizwan Ahmed, Nasir Ahmad
    Abstract:

    Abstract The effect of high-density fuel loading on the criticality of low enriched uranium fueled material test Reactors was studied using the standard reactor physics simulation codes WIMS-D/4 and CITATION. Three strategies were considered to increase the fuel loading per plate: (1) by substituting the high-density fuel in place of low-density fuel keeping meat thickness and water channel width constant, (2) by substituting the high-density fuel in place of low-density fuel keeping fuel meat thickness fixed and optimizing the water channel width between the fuel plates and (3) by increasing the fuel meat thickness of fixed density fuel and optimizing the water channel width between the fuel plates. The fuel requirements for critical and first high power cores were determined in each case for higher fuel loadings per plate. It has been found that in the first case, core volume reduces with increasing fuel loadings per plate but requirement of fuel also increases. In the second and third case, core volume as well as fuel requirement decreases with increasing fuel loadings per plate. However in the second case, core volume reduces more rapidly than in case 3 with increasing fuel loadings per plate. Employing standard computer code PARET, steady state thermal hydraulic analysis of all these cores was performed. The thermal hydraulic analysis reveals that cores with higher densities and fixed water channel width are better from thermal hydraulic point of view and have fuel and clad temperatures within the acceptable limits. But the core with higher densities and optimum water channel width is a better choice in terms of core compaction, less 235 U loading and higher neutron fluxes. Finally, the core was compacted in three steps to exploit the benefits of both types of cores. The strategy resulted in 36% reduction in the core volume, 50% increase in thermal neutron flux for irradiation and isotope production and a slight reduction in 235 U loading. All this was achieved with acceptable peak clad and peak fuel centerline temperatures.

  • effect of coolant channel width on group constants and multiplication factor of Research Reactors using mtr type low enriched uranium fuel
    Annals of Nuclear Energy, 2002
    Co-Authors: Nasir Ahmad
    Abstract:

    One dimensional transport theory lattice code wims-d/4 and three dimensional diffusion theory code citation have been used to study the effect of fuel loading on critical cores of low enriched uranium (LEU) fuelled material testing Reactors (MTRs). The fuel loading in a fuel element was varied by changing the fuel density in the fuel meat. In order to keep the reactor critically moderated, the optimal coolant channel width for a given fuel loading was calculated. For the purpose of optimization, the group constants D, Σa and νΣf, and infinite multiplication factor (k∞) were calculated as a function of coolant channel width using wims-d/4. An increase in 235U loading per fuel plate results in an increase in the optimal coolant channel width and k∞. The calculated values were found to be in good agreement with the typical design of MTR. citation was then used to determine the critical cores for different fuel loading with optimized fuel dimensions. Both critical mass and volume were found to decrease with an increase in the fuel loading. The criticality studies of Pakistan Research reactor-1 (PARR-1) are in good agreement with the predictions.

P M Egorenkov - One of the best experts on this subject based on the ideXlab platform.