Papers by Frederik Reitsma
Nuclear Engineering and Design, Jun 1, 2003
The problem of modelling a highly absorbing region in a diffusion calculation is well known and m... more The problem of modelling a highly absorbing region in a diffusion calculation is well known and many methods have been developed to accommodate the transport effects in diffusion theory. In this work the use of the equivalent cross-sections method for pebble bed type reactors is evaluated by applying it to calculations of control rod (CR) experiments performed at the ASTRA
Nuclear Engineering and Design, Jul 1, 2023
ASTM International eBooks, Mar 9, 2009
Neutron fluences were calculated for the first Reactor Vessel Surveillance (RVS) capsules extract... more Neutron fluences were calculated for the first Reactor Vessel Surveillance (RVS) capsules extracted from Koeberg Unit 1 and 2. For the calculation of the Koeberg RVS-capsule dosimetry cross sections a calculational system (called PCDC) was developed on a 32-bit i386/i486 personal computer (PC). This system consists of pre- and post-processors, a 171-group neutron reaction cross section library, a 171-group reactor zone flux spectrum library and an interface with the SAND-II, DOSCROS-84 and IRDF-90 dosimetry libraries. The sensitivity of these cross sections was determined for different reactor model parameters, core conditions and dosimetry libraries. Using the IRDF-90 dosimetry library, a maximum difference of 15% was obtained between the fluences predicted with the non-fissionable dosimeters.
Nuclear Engineering and Design, Oct 1, 2012
ABSTRACT In the multi-pass fuel management scheme employed for the pebble bed modular reactor the... more ABSTRACT In the multi-pass fuel management scheme employed for the pebble bed modular reactor the fuel pebbles are re-circulated until they reach the target burn-up. The rate at which fresh fuel is loaded and burned fuel is discharged is a result of the core neutronics cycle analysis but in practice (on the plant) this has to be controlled and managed by the fuel handling and storage system and use of the burnup measurement system. The excess reactivity is the additional reactivity available in the core during operating conditions that is the result of loading a fuel mixture in the core that is more reactive (less burned) than what is required to keep the reactor critical at full power operational conditions. The excess reactivity is balanced by the insertion of the control rods to keep the reactor critical. The excess reactivity allows flexibility in operations, for example to overcome the xenon build up when power is decreased as part of load follow. In order to limit reactivity excursions and to ensure safe shutdown the excess reactivity and thus the insertion depth of the control rods at normal operating conditions has to be managed. One way to do this is by operational procedures. The reactivity effect of long-term operation with the control rods inserted deeper than the design point is investigated and a control rod insertion limit is proposed that will not limit normal operations. The effects of other phenomena that can increase the power defect, such as higher-than-expected fuel temperatures, are also introduced. All of these cases are then evaluated by ensuring cold shutdown is still achievable and where appropriate by reactivity insertion accident analysis. These aspects are investigated on the PBMR 400 MW design.
Transactions of the American Nuclear Society - Volume 122, 2020
The Idaho National Laboratory's PEBBED code and simple probability considerations are used to est... more The Idaho National Laboratory's PEBBED code and simple probability considerations are used to estimate the likelihood and consequences of the accumulation of highly reactive pebbles in the region of peak power in a pebble-bed reactor. The PEBBED code is briefly described, and the logic of the probability calculations is presented in detail. The results of the calculations appear to show that hot-spot formation produces only moderate increases in peak accident temperatures, and no increases at all in normal operating temperatures.
The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OEC... more The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor 400 MW design (PBMR-400) coupled neutronics/thermal hydraulics transient benchmark problem as part of their official activities. The scope of the benchmark is to establish a well-defined problem, based on a common given library of cross sections, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events through a set of multi-dimensional computational test problems. The benchmark includes three steady state exercises and six transient exercises. This paper describes the first two steady state exercises, their objectives and the international participation in terms of organization, country and computer code utilized. This description is followed by a comparison and analysis of the participants' results submitted for these two exercises. The comparison of results from different codes allows for an assessment of the sensitivity of a result to the method employed and can thus help to focus the development efforts on the most critical areas. The two first exercises also allow for removing of user-related modeling errors and prepare core neutronics and thermal-hydraulics models ofmore » the different codes for the rest of the exercises in the benchmark. (authors)« less
Social Science Research Network, 2022
Journal of Nuclear Science and Technology, May 23, 2019
In the 350 MW Modular High Temperature Gas-Cooled Reactor (MHTGR-350), not only is the fuel doubl... more In the 350 MW Modular High Temperature Gas-Cooled Reactor (MHTGR-350), not only is the fuel double heterogeneous but so are the lumped burnable poisons (LBPs). The LBPs are composed of Bi-Structural Isotropic (BISO) particles and the fuel is composed of Tri-Structural Isotropic (TRISO) particles. This work investigates different methods to model coated particles using KENO-VI and NEWT of SCALE 6. The most efficient way of modelling TRISO particles in terms of packing and randomization is established in continuous energy (CE) mode and its impact on k inf is investigated. In the multi-group (MG) treatment, coated particles are modelled with the DOUBLEHET function which is only designed for particles that contain fuel. The LBP BISO particles could therefore not be modelled. Hence a method called the LBP Trace method is developed to model the LBP BISO particles using the DOUBLEHET function. It was found that k inf changed by 1500 pcm compared to the conventional (homogenized) case, when using the LBP Trace method. However, no significant change was observed in the macroscopic absorption cross section that would be passed to a nodal core calculation. Furt hermore, the LBP Trace method showed small changes in the nuclear data uncertainty when compared to conventional case.
This paper discusses how high temperature gas-cooled reactors (HTGRs) could provide energy for ph... more This paper discusses how high temperature gas-cooled reactors (HTGRs) could provide energy for phosphate rock (PR) processing while extracting uranium (U) from the processed PR that can again be used as raw material for nuclear reactor fuel that may power the greenhouse gas lean energy source employed. First estimates using a HTGR presently constructed in China (HTR-PM) conclude that a concentration of approximately 80 mg/kg U in PR is sufficiently high for energy neutral wet acid PR processing with waste treatment and a concentration of approximately 110 mg/kg U is adequate to promote energy intensive high quality thermal phosphoric acid production. In addition, the recovery of U from PRs yields beneficial side-effects in a way that U loads on agricultural soils are reduced and consequently contamination of groundwater with U will be diminished.
Nuclear Engineering and Design, May 1, 2014
ABSTRACT A computer code was developed for the semi-automatic translation of input models for the... more ABSTRACT A computer code was developed for the semi-automatic translation of input models for the VSOP-A diffusion neutronics simulation code to the format of the newer VSOP 99/05 code. In this paper, this algorithm is presented as a generic method for producing codes for the automatic translation of input models from the format of one code version to another, or even to that of a completely different code. Normally, such translations are done manually. However, input model files, such as for the VSOP codes, often are very large and may consist of many thousands of numeric entries that make no particular sense to the human eye. Therefore the task, of for instance nuclear regulators, to verify the accuracy of such translated files can be very difficult and cumbersome. This may cause translation errors not to be picked up, which may have disastrous consequences later on when a reactor with such a faulty design is built. Therefore a generic algorithm for producing such automatic translation codes may ease the translation and verification process to a great extent. It will also remove human error from the process, which may significantly enhance the accuracy and reliability of the process. The developed algorithm also automatically creates a verification log file which permanently record the names and values of each variable used, as well as the list of meanings of all the possible values. This should greatly facilitate reactor licensing applications.
This research focuses on modelling reflectors in typical material testing reactors (MTRs). Equiva... more This research focuses on modelling reflectors in typical material testing reactors (MTRs). Equivalence theory is used to homogenise and collapse detailed transport solutions to generate equivalent nodal parameters and albedo boundary conditions for reflectors, for subsequent use in full core nodal diffusion codes. This approach to reflector modelling has been shown to be accurate for two-group large commercial light water reactor (LWR) analysis, but has not been investigated for MTRs. MTRs are smaller, with much larger leakage, environment sensitivity and multi-group spectrum dependencies than LWRs. This study aims to determine if this approach to reflector modelling is an accurate and plausible homogenisation technique for the modelling of small MTR cores. The successful implementation will result in simplified core models, better accuracy and improved efficiency of computer simulations. Codes used in this study include SCALE 6.1, OSCAR-4 and EQUIVA (the last two codes are developed and used at Necsa). The results show a five times reduction in calculational time for the proposed reduced reactor model compared to the traditional explicit model. The calculated equivalent parameters however show some sensitivity to the environment used to generate them. Differences in the results compared to the current explicit model, require more careful investigation includingmore » comparisons with a reference result, before its implementation can be recommended. (authors)« less
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Papers by Frederik Reitsma