Abstract Advanced Power Reactor (APR-1400) is a Generation III+ Pressurized Water Reactor (PWRs) ... more Abstract Advanced Power Reactor (APR-1400) is a Generation III+ Pressurized Water Reactor (PWRs) and has gained popularity among energy mix community. APR-1400 features enhanced safety limits to prevent a "Fukushima-type" accident scenario for a duration up to 8 hours. This work proposes a system that enhances APR 1400 passive cooling (Thermosyphon Cooling System, TCS) capabilities to guarantee at least double the coping time before core meltdown. TCS provides enough cooling to the reactor core in the case of loss of offsite power, known as Station Black-Out (SBO) accident. This work invites to use the In-containment Refueling Water Storage Tank (IRWST) to further extend the APR1400 capability to passively cool the reactor core after an SBO accident. This work is designed and tested by 3Keymaster simulator platform developed by Western Service Corporation. Simulation has modified the existing model of APR1400 in the simulator by adding pipes, valves and heat exchanger. Results show the system will cool the reactor core for almost 18 hours after the SBO accident. The system works initially in parallel with steam generators then it continues cooling the reactor core for extra 18 hours resulting in providing more time for restoring the on-site power sources and preventing core meltdown.
Abstract The United Arab Emirates is currently building and operating four units of the APR-1400 ... more Abstract The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.
This work presents a study on the fundamental aspects of a reliable gold determination in volumin... more This work presents a study on the fundamental aspects of a reliable gold determination in voluminous mineral samples by Neutron Activation Analysis, using a Linear Particle Accelerator as a source of irradiation and a single detector as a detection system. The report includes i) an evaluation of the detection efficiency of the system, particularly the evaluation of the geometric efficiency in samples with significant volumes (with approximate volumes from 20 to 30 cm3); ii) a few methods for the calculation of the gamma ray self attenuation in the sample; iii) some gold in-homogeneity tests in the sample; iv) several validation calculations obtained with the MCNP code - the complete system was modeled with this program. This group of results has produced a gamma-spectrometry protocol for this determination, in order to obtain accurately the gold concentration in mineral samples this study was required by a legal bureau to determine the gold abundance in a Patagonian small deposit. A...
The Indonesian Multipurpose Research Reactor namely Reaktor Serba Guna G.A. Siwabessy (RSG GAS) i... more The Indonesian Multipurpose Research Reactor namely Reaktor Serba Guna G.A. Siwabessy (RSG GAS) is a 30 MWth (max.) pool-type reactor loaded with plate-type low-enriched uranium fuel, using light water as coolant and moderator, and beryllium as reflector. The benchmark of the 1st criticality core of RSG GAS using different nuclear data libraries such as JENDL-4.0, JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1 have been performed in the previous work and compared with the experiment result. In this work, the newly released ENDF/B-VIII.0 neutron reaction and thermal neutron scattering libraries will be used and the important neu-tronics parameters such as multiplication factor, kinetics parameters, and fission reaction rate will be calculated using Monte Carlo code MCNP6.2 and compared against the previous work and the experiment result.
We present in this work the results of neutron measurements performed on electrolytic cells conta... more We present in this work the results of neutron measurements performed on electrolytic cells containing deuterated Pd cathodes, using a high efficiency thermal neutron detection system in combination with a procedure involving a non-stationary current through the cell's circuit. Experiments carried-out at our laboratory over a long period revealed a low level neutron production correlated with the current pulses, giving
In this work we present scattering kernels and cross sections for a few cryogenic materials of in... more In this work we present scattering kernels and cross sections for a few cryogenic materials of interest as cold neutron moderators. These calculations are based on a Synthetic Scattering Function (SSF), developed to describe the incoherent interaction of slow neutrons with hydrogeneous materials in a simple way. The main advantages of this model reside in the analytical expressions that it produces for several physics parameters of specific interest to the field of Neutron and Reactor Physics. These parameters include double-differential cross sections, energy-transfer kernels, and total cross sections, which in turn allow the evaluation of neutron scattering and transport properties. The possibility of calculating those quantities in a fast and accurate way, permits the production of group constants for any specific material, at any temperature, any degree in the Legendre expansion, and over any required energy mesh. We have included the SSF routines into the NJOY code, in such a w...
A thermal neutron imaging facility has been set up at the North Carolina State University PULSTAR... more A thermal neutron imaging facility has been set up at the North Carolina State University PULSTAR reactor. The PULSTAR is an open pool type light water moderated 1-MWth research reactor with six beam tubes. The present facility is set up on beam tube # 5 of the reactor. The facility is intended to have radiographic and tomographic capabilities. The design
ABSTRACT The interaction of slow neutrons with liquid H2 and D2 is described in a simple way with... more ABSTRACT The interaction of slow neutrons with liquid H2 and D2 is described in a simple way with the use of a synthetic scattering function. The analytical expressions derived from this model allow a very fast evaluation of the total cross section as well as the isotropic and the anisotropic energy-transfer kernels. From these, neutron scattering and transport properties are calculated and compared with available experimental results for both liquids.
ABSTRACT We discuss here some applications of synthetic thermal-neutron scattering function (SSF)... more ABSTRACT We discuss here some applications of synthetic thermal-neutron scattering function (SSF) with special emphasis on topics of interest in the field of nuclear engineering. This scattering function was basically devised to describe the interaction of slow neutrons with molecules, although the cases that we will discuss here involve hydrogeneous molecules, due to their importance as moderator materials. The analytic character of the expressions for scattering kernels and cross-sections derived from the SSF permits an easy modelling of any molecule and a very fast computation of the required quantities under different physical conditions.
ABSTRACT The traditional procedures employed to correct an observed scattering spectrum in order ... more ABSTRACT The traditional procedures employed to correct an observed scattering spectrum in order to isolate the structure factor, involve approximations that are not valid in many real cases. We present here Monte Carlo simulations based on a Synthetic Model to describe the (incoherent) neutron-molecule interaction, which allow a simultaneous description of multiple, inelastic and beam-attenuation processes into the sample, as well as the contributions due to the presence of a container. The procedure is applied to experiments performed on H2O and D2O at a reactor two-axes instrument. We show that our unified scheme produces very good agreement with the measurements and discuss some of its merits as compared with the traditional method.
Abstract Advanced Power Reactor (APR-1400) is a Generation III+ Pressurized Water Reactor (PWRs) ... more Abstract Advanced Power Reactor (APR-1400) is a Generation III+ Pressurized Water Reactor (PWRs) and has gained popularity among energy mix community. APR-1400 features enhanced safety limits to prevent a "Fukushima-type" accident scenario for a duration up to 8 hours. This work proposes a system that enhances APR 1400 passive cooling (Thermosyphon Cooling System, TCS) capabilities to guarantee at least double the coping time before core meltdown. TCS provides enough cooling to the reactor core in the case of loss of offsite power, known as Station Black-Out (SBO) accident. This work invites to use the In-containment Refueling Water Storage Tank (IRWST) to further extend the APR1400 capability to passively cool the reactor core after an SBO accident. This work is designed and tested by 3Keymaster simulator platform developed by Western Service Corporation. Simulation has modified the existing model of APR1400 in the simulator by adding pipes, valves and heat exchanger. Results show the system will cool the reactor core for almost 18 hours after the SBO accident. The system works initially in parallel with steam generators then it continues cooling the reactor core for extra 18 hours resulting in providing more time for restoring the on-site power sources and preventing core meltdown.
Abstract The United Arab Emirates is currently building and operating four units of the APR-1400 ... more Abstract The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.
This work presents a study on the fundamental aspects of a reliable gold determination in volumin... more This work presents a study on the fundamental aspects of a reliable gold determination in voluminous mineral samples by Neutron Activation Analysis, using a Linear Particle Accelerator as a source of irradiation and a single detector as a detection system. The report includes i) an evaluation of the detection efficiency of the system, particularly the evaluation of the geometric efficiency in samples with significant volumes (with approximate volumes from 20 to 30 cm3); ii) a few methods for the calculation of the gamma ray self attenuation in the sample; iii) some gold in-homogeneity tests in the sample; iv) several validation calculations obtained with the MCNP code - the complete system was modeled with this program. This group of results has produced a gamma-spectrometry protocol for this determination, in order to obtain accurately the gold concentration in mineral samples this study was required by a legal bureau to determine the gold abundance in a Patagonian small deposit. A...
The Indonesian Multipurpose Research Reactor namely Reaktor Serba Guna G.A. Siwabessy (RSG GAS) i... more The Indonesian Multipurpose Research Reactor namely Reaktor Serba Guna G.A. Siwabessy (RSG GAS) is a 30 MWth (max.) pool-type reactor loaded with plate-type low-enriched uranium fuel, using light water as coolant and moderator, and beryllium as reflector. The benchmark of the 1st criticality core of RSG GAS using different nuclear data libraries such as JENDL-4.0, JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1 have been performed in the previous work and compared with the experiment result. In this work, the newly released ENDF/B-VIII.0 neutron reaction and thermal neutron scattering libraries will be used and the important neu-tronics parameters such as multiplication factor, kinetics parameters, and fission reaction rate will be calculated using Monte Carlo code MCNP6.2 and compared against the previous work and the experiment result.
We present in this work the results of neutron measurements performed on electrolytic cells conta... more We present in this work the results of neutron measurements performed on electrolytic cells containing deuterated Pd cathodes, using a high efficiency thermal neutron detection system in combination with a procedure involving a non-stationary current through the cell's circuit. Experiments carried-out at our laboratory over a long period revealed a low level neutron production correlated with the current pulses, giving
In this work we present scattering kernels and cross sections for a few cryogenic materials of in... more In this work we present scattering kernels and cross sections for a few cryogenic materials of interest as cold neutron moderators. These calculations are based on a Synthetic Scattering Function (SSF), developed to describe the incoherent interaction of slow neutrons with hydrogeneous materials in a simple way. The main advantages of this model reside in the analytical expressions that it produces for several physics parameters of specific interest to the field of Neutron and Reactor Physics. These parameters include double-differential cross sections, energy-transfer kernels, and total cross sections, which in turn allow the evaluation of neutron scattering and transport properties. The possibility of calculating those quantities in a fast and accurate way, permits the production of group constants for any specific material, at any temperature, any degree in the Legendre expansion, and over any required energy mesh. We have included the SSF routines into the NJOY code, in such a w...
A thermal neutron imaging facility has been set up at the North Carolina State University PULSTAR... more A thermal neutron imaging facility has been set up at the North Carolina State University PULSTAR reactor. The PULSTAR is an open pool type light water moderated 1-MWth research reactor with six beam tubes. The present facility is set up on beam tube # 5 of the reactor. The facility is intended to have radiographic and tomographic capabilities. The design
ABSTRACT The interaction of slow neutrons with liquid H2 and D2 is described in a simple way with... more ABSTRACT The interaction of slow neutrons with liquid H2 and D2 is described in a simple way with the use of a synthetic scattering function. The analytical expressions derived from this model allow a very fast evaluation of the total cross section as well as the isotropic and the anisotropic energy-transfer kernels. From these, neutron scattering and transport properties are calculated and compared with available experimental results for both liquids.
ABSTRACT We discuss here some applications of synthetic thermal-neutron scattering function (SSF)... more ABSTRACT We discuss here some applications of synthetic thermal-neutron scattering function (SSF) with special emphasis on topics of interest in the field of nuclear engineering. This scattering function was basically devised to describe the interaction of slow neutrons with molecules, although the cases that we will discuss here involve hydrogeneous molecules, due to their importance as moderator materials. The analytic character of the expressions for scattering kernels and cross-sections derived from the SSF permits an easy modelling of any molecule and a very fast computation of the required quantities under different physical conditions.
ABSTRACT The traditional procedures employed to correct an observed scattering spectrum in order ... more ABSTRACT The traditional procedures employed to correct an observed scattering spectrum in order to isolate the structure factor, involve approximations that are not valid in many real cases. We present here Monte Carlo simulations based on a Synthetic Model to describe the (incoherent) neutron-molecule interaction, which allow a simultaneous description of multiple, inelastic and beam-attenuation processes into the sample, as well as the contributions due to the presence of a container. The procedure is applied to experiments performed on H2O and D2O at a reactor two-axes instrument. We show that our unified scheme produces very good agreement with the measurements and discuss some of its merits as compared with the traditional method.
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Papers by Victor Gillette