The PUMA project, a Specific Targeted Research Project (STREP) of the European Union EURATOM 6th ... more The PUMA project, a Specific Targeted Research Project (STREP) of the European Union EURATOM 6th Framework Program, is mainly aimed at providing additional key elements for the utilisation and transmutation of plutonium and minor actinides in contemporary and future (high temperature) gas-cooled reactor design, which are promising tools for improving the sustainability of the nuclear fuel cycle, hereby also contributing to the reduction of Pu and MA stockpiles, and to the development of safe and sustainable reactors for CO2-free energy generation. The project runs from
As a crucial core physics parameter, the control rod reactivity has to be predicted for the contr... more As a crucial core physics parameter, the control rod reactivity has to be predicted for the control and safety of the reactor. This paper studies the control rod reactivity calculation of the pebble-bed reactor with three scenarios of UO2, (Th,U)O2, and PuO2 fuel type without any modifications in the configuration of the reactor core. The reactor geometry of HTR-10 was selected for the reactor model. The entire calculation of control rod reactivity was done using the MCNP6 code with ENDF/B-VII library. The calculation results show that the total reactivity worth of control rods in UO2-, (U,Th)O2-, and PuO2-fueled cores is 15.87, 15.25, and 14.33%Δk/k, respectively. These results prove that the effectiveness of total control rod in thorium and uranium cores is almost similar to but higher than that in plutonium cores. The highest reactivity worth of individual control rod in uranium, thorium and plutonium cores is 1.64, 1.44, and 1.53%Δk/k corresponding to CR8, CR1, and CR5, respecti...
This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA re... more This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the " wallpaper " fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life.
The theoretical background of sensitivity and uncertainty calculations associated with burnup cal... more The theoretical background of sensitivity and uncertainty calculations associated with burnup calculations has been studied in depth. In general a burnup calculation scheme consists of consecutive sets ("steps") of spectrum-, cross section collaps-and one-group burnup step calculations. It was found that in order to be able to calculate uncertainties in final nuclide densities for such a scheme it is necessary to calculate, for each single step, the sensitivity of the nuclide densities at the end of this step with respect to uncertainties in the microscopic (one-group) cross sections and the nuclide densities at the beginning of the step. Hereby it should be noted that the flux/spectrum may be different for different steps. The sensitivity matrices for the consecutive steps are then combined into a single "one-group cross section-to-density" sensitivity matrix of the entire scheme, which can then be used to calculate the uncertainties and covariance matrix of the final nuclide density vector. The applicability of the method is demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. Although not all differences between measured and calculated densities can be explained by the propagated cross section uncertainties, the comparison still demonstrates the feasibility of the approach presented here.
Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANT... more Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, st...
In the context of the European Union Fifth Framework Program studies have been performed about th... more In the context of the European Union Fifth Framework Program studies have been performed about the incineration of plutonium from reprocessed uranic fuel and from MOX in a High Temperature Gas Cooled Reactor (HTR). In these studies only the steady-state behaviour and the destruction of the plutonium have been investigated.
The core physics investigations at the Nuclear Research Consultancy Group in the Netherlands, as ... more The core physics investigations at the Nuclear Research Consultancy Group in the Netherlands, as part of the activities within the HTR-N project of the European Fifth Framework Program, are focused on the incineration of pure (first- and second-generation) Pu fuels in the reference pebble bed high-temperature gas-cooled reactor (HTR) HTR-MODUL with a continuous reload [MEDUL, (MEhrfach DUrchLauf, multipass)] fueling strategy in which the spherical fuel elements, or pebbles, pass through the core a number of times before being permanently discharged. For pebbles fueled with different loadings of plutonium, the feasibility of a sustained fuel cycle under nominal reactor conditions was investigated by means of the reactivity and temperature coefficients of the reactor. The HTR-MODUL was found to be a very effective reactor to reduce the stockpile of first-generation plutonium. It reduces the amount of plutonium to about one-sixth of the original and reduces the risk of proliferation by...
Progress in Nuclear Energy - PROG NUCL ENERGY, 1991
ABSTRACT Two one-dimensional models are developed for the investigation of the gas dynamical beha... more ABSTRACT Two one-dimensional models are developed for the investigation of the gas dynamical behaviour of the fuel gas in a cylindrical gaseous core fission reactor. By numerical and analytical calculations, it is shown that, for the case that a direct energy extraction mechanisn (such as MHD) is not present, increasing density oscillations occur in the gas. Also an estimation is made of the attainable direct energy conversion efficiency, for the case that a direct energy extraction mechanism is present.
ABSTRACT A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at hig... more ABSTRACT A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters.
A method has been developed to include diffusion-theory-type (macroscopic) group cross sections i... more A method has been developed to include diffusion-theory-type (macroscopic) group cross sections into MCNP, while still running in continuous (point-wise) energy calculation mode. A good agreement is observed between results (k eff , reaction rates and flux profiles) from a heterogeneous -reference -MCNP calculation and those from MCNP calculations based on homogenised macroscopic group cross sections, obtained from the heterogeneous MCNP calculation and ELNINJO post-processing. Also a good agreement is observed between results from MCNP transport-theory calculations and PANTHER diffusion-theory calculations, based on exactly the same macroscopic group cross sections, although effects of the fundamental differences between transport-and diffusion theory are apparent as well. The presented results provide further evidence for the validity of the MCNP-ELNINJO method for the generation of homogenised (macroscopic) group cross sections for complicated geometries. Furthermore, the method ...
For the EFTTRA-T5 experiment, the irradiation of which is planned in the High Flux Reactor in Pet... more For the EFTTRA-T5 experiment, the irradiation of which is planned in the High Flux Reactor in Petten, Americium-containing zirconia-based inert matrix targets are envisaged, both with and without addition of Plutonium. The power of an Am-target (without Pu) is very low at BOI and slowly increases with irradiation time, because fissile nuclides are formed during irradiation. The addition of Plutonium increases the initial power of the target, such that a high fuel temperature can be obtained at Beginning Of Irradiation (BOI). In this paper the technical preparations of the EFTTRA-T5 experiment will be explained. The calculations of temperature and power profile, leading to the definition of the actual composition of the EFTTRA-T5 targets are presented. In addition, the results of the fabrication tests for the targets are discussed.
The High-Temperature gas-cooled Reactor (HTR) is a promising concept for the next generation of n... more The High-Temperature gas-cooled Reactor (HTR) is a promising concept for the next generation of nuclear power plants, and it is essential that validated analytical tools are available in the European nuclear community to perform conceptual design studies, industrial calculations (reload calculations and the associated core follow), safety analyses for licensing, etc., for new fuel cycles aiming at plutonium and minor actinide (MA) incineration/transmutation without multi-reprocessing of the discharged fuel. The “HTR-N” project of the European Union Fifth Framework Program includes activities concerning the validation of computational tools and the qualification of models. These activities are centred around the two HTR systems currently in operation, viz. the HTR-10 and the HTTR. The re-calculation of the HTTR first criticality with a Monte Carlo neutron transport code now yields acceptable correspondence with experimental data. Also calculations by 3-D diffusion theory codes yield ...
As early as the 1970s, attempts have been made to reduce the peak fuel temperature in pebble bed-... more As early as the 1970s, attempts have been made to reduce the peak fuel temperature in pebble bed-type high-temperature reactors (HTRs) by means of so-called "wallpaper fuel," in which the fuel is arranged in a spherical shell within a pebble. By raising the particle packing fraction, fuel kernels are condensed to the outer diameter of the fuel zone, leaving a central part of the pebble free of fuel. This modification prevents power generation in this central fuel-free zone and decreases the temperature gradient across the pebble. Besides the reduction of maximum and average particle temperature, the wallpaper concept also enhances neutronic performance through improved neutron economy, resulting in reduced fissile material and/or enrichment needs or providing the potential to achieve higher burnup. To assess such improvements, calculations were performed using the PANTHERMIX code. Among other tests, investigations of fuel cycle under steady-state conditions and loss-of-coo...
The code VAREX is described that calculates by first-order perturbation theory the contributions ... more The code VAREX is described that calculates by first-order perturbation theory the contributions of individual nuclides to reactivity changes induced by, for example, a change of temperature or nuclide densities. Furthermore, VAREX calculates the contributions that make up the well-known four-factor formula in reactor physics, which has proven to be a very useful tool to characterize the neutron spectrum. The (effective) delayed neutron fractions per nuclide and for the whole geometry are calculated by use of two nuclear data libraries that are based on JEF2.2. Finally, the mean neutron generation time can be calculated. Results of calculations are given to show the reader the type of information that VAREX can provide.
The high-temperature gas-cooled reactor (HTGR) appears as a good candidate for the next generatio... more The high-temperature gas-cooled reactor (HTGR) appears as a good candidate for the next generation of nuclear power plants. In the "HTR-N" project of the European Union Fifth Framework Program, analyses have been performed on a number of conceptual HTGR designs, derived from reference pebble-bed and hexagonal block-type HTGR types. It is shown that several HTGR concepts are quite promising as systems for the incineration of plutonium and possibly minor actinides.
The PUMA project, a Specific Targeted Research Project (STREP) of the European Union EURATOM 6th ... more The PUMA project, a Specific Targeted Research Project (STREP) of the European Union EURATOM 6th Framework Program, is mainly aimed at providing additional key elements for the utilisation and transmutation of plutonium and minor actinides in contemporary and future (high temperature) gas-cooled reactor design, which are promising tools for improving the sustainability of the nuclear fuel cycle, hereby also contributing to the reduction of Pu and MA stockpiles, and to the development of safe and sustainable reactors for CO2-free energy generation. The project runs from
As a crucial core physics parameter, the control rod reactivity has to be predicted for the contr... more As a crucial core physics parameter, the control rod reactivity has to be predicted for the control and safety of the reactor. This paper studies the control rod reactivity calculation of the pebble-bed reactor with three scenarios of UO2, (Th,U)O2, and PuO2 fuel type without any modifications in the configuration of the reactor core. The reactor geometry of HTR-10 was selected for the reactor model. The entire calculation of control rod reactivity was done using the MCNP6 code with ENDF/B-VII library. The calculation results show that the total reactivity worth of control rods in UO2-, (U,Th)O2-, and PuO2-fueled cores is 15.87, 15.25, and 14.33%Δk/k, respectively. These results prove that the effectiveness of total control rod in thorium and uranium cores is almost similar to but higher than that in plutonium cores. The highest reactivity worth of individual control rod in uranium, thorium and plutonium cores is 1.64, 1.44, and 1.53%Δk/k corresponding to CR8, CR1, and CR5, respecti...
This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA re... more This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the " wallpaper " fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life.
The theoretical background of sensitivity and uncertainty calculations associated with burnup cal... more The theoretical background of sensitivity and uncertainty calculations associated with burnup calculations has been studied in depth. In general a burnup calculation scheme consists of consecutive sets ("steps") of spectrum-, cross section collaps-and one-group burnup step calculations. It was found that in order to be able to calculate uncertainties in final nuclide densities for such a scheme it is necessary to calculate, for each single step, the sensitivity of the nuclide densities at the end of this step with respect to uncertainties in the microscopic (one-group) cross sections and the nuclide densities at the beginning of the step. Hereby it should be noted that the flux/spectrum may be different for different steps. The sensitivity matrices for the consecutive steps are then combined into a single "one-group cross section-to-density" sensitivity matrix of the entire scheme, which can then be used to calculate the uncertainties and covariance matrix of the final nuclide density vector. The applicability of the method is demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. Although not all differences between measured and calculated densities can be explained by the propagated cross section uncertainties, the comparison still demonstrates the feasibility of the approach presented here.
Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANT... more Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, st...
In the context of the European Union Fifth Framework Program studies have been performed about th... more In the context of the European Union Fifth Framework Program studies have been performed about the incineration of plutonium from reprocessed uranic fuel and from MOX in a High Temperature Gas Cooled Reactor (HTR). In these studies only the steady-state behaviour and the destruction of the plutonium have been investigated.
The core physics investigations at the Nuclear Research Consultancy Group in the Netherlands, as ... more The core physics investigations at the Nuclear Research Consultancy Group in the Netherlands, as part of the activities within the HTR-N project of the European Fifth Framework Program, are focused on the incineration of pure (first- and second-generation) Pu fuels in the reference pebble bed high-temperature gas-cooled reactor (HTR) HTR-MODUL with a continuous reload [MEDUL, (MEhrfach DUrchLauf, multipass)] fueling strategy in which the spherical fuel elements, or pebbles, pass through the core a number of times before being permanently discharged. For pebbles fueled with different loadings of plutonium, the feasibility of a sustained fuel cycle under nominal reactor conditions was investigated by means of the reactivity and temperature coefficients of the reactor. The HTR-MODUL was found to be a very effective reactor to reduce the stockpile of first-generation plutonium. It reduces the amount of plutonium to about one-sixth of the original and reduces the risk of proliferation by...
Progress in Nuclear Energy - PROG NUCL ENERGY, 1991
ABSTRACT Two one-dimensional models are developed for the investigation of the gas dynamical beha... more ABSTRACT Two one-dimensional models are developed for the investigation of the gas dynamical behaviour of the fuel gas in a cylindrical gaseous core fission reactor. By numerical and analytical calculations, it is shown that, for the case that a direct energy extraction mechanisn (such as MHD) is not present, increasing density oscillations occur in the gas. Also an estimation is made of the attainable direct energy conversion efficiency, for the case that a direct energy extraction mechanism is present.
ABSTRACT A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at hig... more ABSTRACT A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters.
A method has been developed to include diffusion-theory-type (macroscopic) group cross sections i... more A method has been developed to include diffusion-theory-type (macroscopic) group cross sections into MCNP, while still running in continuous (point-wise) energy calculation mode. A good agreement is observed between results (k eff , reaction rates and flux profiles) from a heterogeneous -reference -MCNP calculation and those from MCNP calculations based on homogenised macroscopic group cross sections, obtained from the heterogeneous MCNP calculation and ELNINJO post-processing. Also a good agreement is observed between results from MCNP transport-theory calculations and PANTHER diffusion-theory calculations, based on exactly the same macroscopic group cross sections, although effects of the fundamental differences between transport-and diffusion theory are apparent as well. The presented results provide further evidence for the validity of the MCNP-ELNINJO method for the generation of homogenised (macroscopic) group cross sections for complicated geometries. Furthermore, the method ...
For the EFTTRA-T5 experiment, the irradiation of which is planned in the High Flux Reactor in Pet... more For the EFTTRA-T5 experiment, the irradiation of which is planned in the High Flux Reactor in Petten, Americium-containing zirconia-based inert matrix targets are envisaged, both with and without addition of Plutonium. The power of an Am-target (without Pu) is very low at BOI and slowly increases with irradiation time, because fissile nuclides are formed during irradiation. The addition of Plutonium increases the initial power of the target, such that a high fuel temperature can be obtained at Beginning Of Irradiation (BOI). In this paper the technical preparations of the EFTTRA-T5 experiment will be explained. The calculations of temperature and power profile, leading to the definition of the actual composition of the EFTTRA-T5 targets are presented. In addition, the results of the fabrication tests for the targets are discussed.
The High-Temperature gas-cooled Reactor (HTR) is a promising concept for the next generation of n... more The High-Temperature gas-cooled Reactor (HTR) is a promising concept for the next generation of nuclear power plants, and it is essential that validated analytical tools are available in the European nuclear community to perform conceptual design studies, industrial calculations (reload calculations and the associated core follow), safety analyses for licensing, etc., for new fuel cycles aiming at plutonium and minor actinide (MA) incineration/transmutation without multi-reprocessing of the discharged fuel. The “HTR-N” project of the European Union Fifth Framework Program includes activities concerning the validation of computational tools and the qualification of models. These activities are centred around the two HTR systems currently in operation, viz. the HTR-10 and the HTTR. The re-calculation of the HTTR first criticality with a Monte Carlo neutron transport code now yields acceptable correspondence with experimental data. Also calculations by 3-D diffusion theory codes yield ...
As early as the 1970s, attempts have been made to reduce the peak fuel temperature in pebble bed-... more As early as the 1970s, attempts have been made to reduce the peak fuel temperature in pebble bed-type high-temperature reactors (HTRs) by means of so-called "wallpaper fuel," in which the fuel is arranged in a spherical shell within a pebble. By raising the particle packing fraction, fuel kernels are condensed to the outer diameter of the fuel zone, leaving a central part of the pebble free of fuel. This modification prevents power generation in this central fuel-free zone and decreases the temperature gradient across the pebble. Besides the reduction of maximum and average particle temperature, the wallpaper concept also enhances neutronic performance through improved neutron economy, resulting in reduced fissile material and/or enrichment needs or providing the potential to achieve higher burnup. To assess such improvements, calculations were performed using the PANTHERMIX code. Among other tests, investigations of fuel cycle under steady-state conditions and loss-of-coo...
The code VAREX is described that calculates by first-order perturbation theory the contributions ... more The code VAREX is described that calculates by first-order perturbation theory the contributions of individual nuclides to reactivity changes induced by, for example, a change of temperature or nuclide densities. Furthermore, VAREX calculates the contributions that make up the well-known four-factor formula in reactor physics, which has proven to be a very useful tool to characterize the neutron spectrum. The (effective) delayed neutron fractions per nuclide and for the whole geometry are calculated by use of two nuclear data libraries that are based on JEF2.2. Finally, the mean neutron generation time can be calculated. Results of calculations are given to show the reader the type of information that VAREX can provide.
The high-temperature gas-cooled reactor (HTGR) appears as a good candidate for the next generatio... more The high-temperature gas-cooled reactor (HTGR) appears as a good candidate for the next generation of nuclear power plants. In the "HTR-N" project of the European Union Fifth Framework Program, analyses have been performed on a number of conceptual HTGR designs, derived from reference pebble-bed and hexagonal block-type HTGR types. It is shown that several HTGR concepts are quite promising as systems for the incineration of plutonium and possibly minor actinides.
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Papers by Jim Kuijper