• Energy Research
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• Annals of Nuclear Energy close

The accuracy of the moments method is examined by comparison of approximate and exact values of the first three moments of solutions of the aerosol dynamics equation when only condensation and removal are taken into account. The log-normal and γ initial distributions are considered. Errors are expressed in terms of nondimensional independent variables and are found to be small.

The accuracy of the moments method is examined by comparison of approximate and exact values of the first three moments of solutions of the aerosol dynamics equation when only condensation and removal are taken into account. The log-normal and γ initial distributions are considered. Errors are expressed in terms of nondimensional independent variables and are found to be small.

Abstract In case of a core melt-down accident in a light water nuclear reactor, hydrogen is produced during reactor core degradation and released into the reactor building. This subsequently creates a combustion hazard. A local ignition of the combustible mixture may generate standing flames or initially slow propagating flames. Depending on geometry, mixture composition and turbulence level, the flame can accelerate or be quenched after a certain distance. The loads generated by the combustion process (increase of the containment atmosphere pressure and temperature) may threaten the integrity of the containment building and of internal walls and equipment. Turbulent deflagration flames may generate high pressure pulses, temperature peaks, shock waves and large pressure gradients which could severely damage specific containment components, internal walls and/or safety equipment. The evaluation of such loads requires validated codes which can be used with a high level of confidence. Currently, turbulence and steam effect on flame acceleration, flame deceleration and flame quenching mechanisms are not well reproduced by combustion models usually implemented in safety tools and further model enhancement and validation are still needed. For this purpose, two hydrogen deflagration benchmark exercises have been organised in the framework of the SARNET network. The first benchmark was focused on turbulence effect on flame propagation. For this purpose, three tests performed in the ENACCEF facility were considered. They concern vertical flame propagation in an initially homogenous mixture with 13 vol.% hydrogen content and different geometrical configurations. Three blockage ratios of 0, 0.33 and 0.6 were considered to generate different levels of turbulence. The second benchmark objective was the investigation of the diluting effect on flame propagation. Thus, three tests performed in the ENACCEF facility using the same blockage ratio of 0.63 and three different initial gas compositions (with 10, 20 and 30 vol.% diluents) have been considered. Since ENACCEF runs at ambient temperature, a surrogate to steam was used consisting of a mixture of 0.6He + 0.4CO 2 on molar basis. This paper aims to present the benchmarks conclusions regarding the ability of LP and CFD combustion models to predict the effect of turbulence and diluent on flame propagation.

Abstract In case of a core melt-down accident in a light water nuclear reactor, hydrogen is produced during reactor core degradation and released into the reactor building. This subsequently creates a combustion hazard. A local ignition of the combustible mixture may generate standing flames or initially slow propagating flames. Depending on geometry, mixture composition and turbulence level, the flame can accelerate or be quenched after a certain distance. The loads generated by the combustion process (increase of the containment atmosphere pressure and temperature) may threaten the integrity of the containment building and of internal walls and equipment. Turbulent deflagration flames may generate high pressure pulses, temperature peaks, shock waves and large pressure gradients which could severely damage specific containment components, internal walls and/or safety equipment. The evaluation of such loads requires validated codes which can be used with a high level of confidence. Currently, turbulence and steam effect on flame acceleration, flame deceleration and flame quenching mechanisms are not well reproduced by combustion models usually implemented in safety tools and further model enhancement and validation are still needed. For this purpose, two hydrogen deflagration benchmark exercises have been organised in the framework of the SARNET network. The first benchmark was focused on turbulence effect on flame propagation. For this purpose, three tests performed in the ENACCEF facility were considered. They concern vertical flame propagation in an initially homogenous mixture with 13 vol.% hydrogen content and different geometrical configurations. Three blockage ratios of 0, 0.33 and 0.6 were considered to generate different levels of turbulence. The second benchmark objective was the investigation of the diluting effect on flame propagation. Thus, three tests performed in the ENACCEF facility using the same blockage ratio of 0.63 and three different initial gas compositions (with 10, 20 and 30 vol.% diluents) have been considered. Since ENACCEF runs at ambient temperature, a surrogate to steam was used consisting of a mixture of 0.6He + 0.4CO 2 on molar basis. This paper aims to present the benchmarks conclusions regarding the ability of LP and CFD combustion models to predict the effect of turbulence and diluent on flame propagation.

Abstract The evolutionary multi-physics tool developed at the Karlsruhe Institute of Technology is the homogeneous pin-by-pin reactor simulator DYNSUB, an internal coupling of the 3D neutron kinetics code DYN3D developed by Helmholtz Zentrum Dresden Rossendorf and the in-house sub-channel code SUBCHANFLOW. The ultimate goal of the on-going efforts concerning DYNSUB is to provide a cost-effective improved description of light water reactor core behavior with pin-by-pin resolution for both static and transient safety relevant scenarios. A cost-effective computer code is defined to be executable on commodity computing clusters which users/customers commonly have access to. Efforts undertaken to improve DYNSUB’s numerical performance and parallelize the code system are presented in this work. Moreover, the coupled code system has been extended in terms of fuel pin level homogenization corrections and flexible mapping schemes. After optimization and extension DYNSUB is successfully applied to study the OECD/NEA and U.S. NRC PWR MOX/UO 2 core transient benchmark with both fuel assembly/channel and pin level/sub-channel model resolution. Even though further improvements in terms of numerical performance and accuracy of physical models are required, the applicability of DYNSUB pin-by-pin simulations for light water reactor safety analysis is proven in principle in this work.

Abstract The evolutionary multi-physics tool developed at the Karlsruhe Institute of Technology is the homogeneous pin-by-pin reactor simulator DYNSUB, an internal coupling of the 3D neutron kinetics code DYN3D developed by Helmholtz Zentrum Dresden Rossendorf and the in-house sub-channel code SUBCHANFLOW. The ultimate goal of the on-going efforts concerning DYNSUB is to provide a cost-effective improved description of light water reactor core behavior with pin-by-pin resolution for both static and transient safety relevant scenarios. A cost-effective computer code is defined to be executable on commodity computing clusters which users/customers commonly have access to. Efforts undertaken to improve DYNSUB’s numerical performance and parallelize the code system are presented in this work. Moreover, the coupled code system has been extended in terms of fuel pin level homogenization corrections and flexible mapping schemes. After optimization and extension DYNSUB is successfully applied to study the OECD/NEA and U.S. NRC PWR MOX/UO 2 core transient benchmark with both fuel assembly/channel and pin level/sub-channel model resolution. Even though further improvements in terms of numerical performance and accuracy of physical models are required, the applicability of DYNSUB pin-by-pin simulations for light water reactor safety analysis is proven in principle in this work.

Abstract Multi-scale, multi-physics problems reveal significant challenges while dealing with coupled neutronic/thermal-hydraulic solutions. Current generations of reactor dynamic codes applied to Light Water Reactors (LWRs), and in the context of this paper, pressurized water reactors (PWRs), are based on 3D neutronic nodal methods coupled with single or two phase flow thermal hydraulic system or sub-channel codes. This paper describes the extension of the coupling scheme between the 3D neutron diffusion codes COBAYA3 and DYN3D, and the sub-channel thermal-hydraulic code SUBCHANFLOW for the simulation of boron dilution transients. This includes the validation of the boron transport model of SUBCHANFLOW. The coupling of COBAYA3 and DYN3D with SUBCHANFLOW is performed inside the NURESIM platform making use of the novel automatic mesh superposition for code coupling. Boron transport models are needed for the simulation of boron dilution transients following a SBLOCA (after loop-seal clearing). In this case, the mixing is a key mechanism determining the positive reactivity insertion in the core. The results obtained with the coupled codes COBAYA3/SUBCHANFLOW, DYN3D/SUBCHANFLOW will be presented and discussed for two transient scenarios; a homogeneous and a heterogeneous boron dilution problem for a mini-core and for a PWR core defined in the NURISP boron dilution Benchmark.

Abstract Multi-scale, multi-physics problems reveal significant challenges while dealing with coupled neutronic/thermal-hydraulic solutions. Current generations of reactor dynamic codes applied to Light Water Reactors (LWRs), and in the context of this paper, pressurized water reactors (PWRs), are based on 3D neutronic nodal methods coupled with single or two phase flow thermal hydraulic system or sub-channel codes. This paper describes the extension of the coupling scheme between the 3D neutron diffusion codes COBAYA3 and DYN3D, and the sub-channel thermal-hydraulic code SUBCHANFLOW for the simulation of boron dilution transients. This includes the validation of the boron transport model of SUBCHANFLOW. The coupling of COBAYA3 and DYN3D with SUBCHANFLOW is performed inside the NURESIM platform making use of the novel automatic mesh superposition for code coupling. Boron transport models are needed for the simulation of boron dilution transients following a SBLOCA (after loop-seal clearing). In this case, the mixing is a key mechanism determining the positive reactivity insertion in the core. The results obtained with the coupled codes COBAYA3/SUBCHANFLOW, DYN3D/SUBCHANFLOW will be presented and discussed for two transient scenarios; a homogeneous and a heterogeneous boron dilution problem for a mini-core and for a PWR core defined in the NURISP boron dilution Benchmark.