Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements (original) (raw)
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NIGERIAN RESEARCH REACTOR-1 CORE NEUTRONICS CALCULATIONS - Azande T. S
Feasibility studies for the conversion of the Nigerian Research Reactor-1 (NIRR-1) have been performed using WIMS and CITATION codes at State. In this work, the core neutronics calculation of NIRR-1 concerning mass loading of U-235 in the core, shut down margin (SDM), safety reactivity factor (SRF), control rod worth, and control rod critical depth of insertion were investigated at low enrichment. Two fuel types (UAl 4 and UO 2) were considered and the uranium densities required for the conversion of NIRR-1 core to low enrichment were computed to be 1201g/cc with 20% enrichment, 1144g/cc with 19.75% enrichment, 1274g/cc with 15% enrichment, 1448g/cc with 10% enrichment for UAl 4 fuel type and 1141g/cc with 20% enrichment, 1144g/cc with 19.75% enrichment, 1216g/cc with 15% enrichment, and 1389g/cc with 10% enrichment for UO 2 fuel type. Significantly, higher uranium densities are required to convert NIRR-1 from HEU to LEU – indicating a drastic review of the NIRR-1 core. INTRODUCTION As part of the ongoing global effort to convert HEU reactor cores to Low Enriched Uranium (LEU) cores under the Reduced Enrichment for Research and Test Reactors (RERTR) program, there is need to study the possibilities of converting NIRR-1 core to less than 20% enrichment. Works have been done on Research and Test Reactor core conversion around the globe. In some proposed models, the total number of fuel pins (Khamis and Khattab, 1999) and the core radius/height ratio (Matos and Lell, 2005) has been drastically changed. This brings about noticeable changes in the relative flux values for both inner and outer irradiation sites. In this work, HEU and LEU cores are analyzed using the present UAl 4 fuel and a potential LEU fuel (UO 2 clad in zircalloy) are considered. The existing HEU core was also analyzed to validate the reactor model used. A significant feature of this work is the preservation of the technical and the geometric specification of the reactor so as to maintain the original designed thermal hydraulic of the reactor. A detailed description of the HEU core of NIRR-1 can be found in the Final Safety Analysis Report (FSAR); (Azande, et al, 2010; SAR, 2005).
Effects of various spacer grid modeling on the neutronic parameters of the VVER-1000 reactor
2011
The main objective of this paper is to study the effects of various spacer grid models on the neutronic parameters of a VVER-1000 reactor. Specifically, the data of the nuclear power plant at the Bushehr site, which is of a VVER-1000 type, will be studied. Three models, representing the spacer grids along the fuel assemblies are presented. These three models are the homogeneous and the heterogeneous local spacer grid models and the shroud spacer grid model. In the homogeneous and the heterogeneous models, the spacer grids are considered at their actual locations in the axial direction. The only difference between the two models is that in the homogeneous model, the spacer grids are homogenized with the coolant while in the heterogeneous model, the spacer grids are modeled around the fuel cells at their exact axial positions. In the shroud model, the spacer grids are modeled in the shroud region containing the coolant and are not necessarily placed at their appropriate axial positions. The aforementioned models and the core geometry are modeled in 3-D hexagonal geometry and solved by the CITATION code. The required cross sections are obtained from the WIMS code using the associated ENDF/B-7 based library. The visual FORTRAN programming 90 is used for analysis of the whole calculation process and provides a sophisticated link between the WIMS and the CITATION codes during core cycle burnup. The core excess reactivity and the critical boron concentration of the reactor are calculated at normal operating state. In addition, the changes in the critical boron concentration and power peaking factors during the first core cycle, for the homogeneous and shroud models are also predicted. The core calculation analysis is done for the core with fully withdrawn control rods. As will be shown, the heterogeneous model, which is a close representation of the actual case, provides a more accurate result than the homogeneous and the shroud models. However, the heterogeneous model is complicated as compared to the other two models. Therefore, the choice of which model to use depends on the accuracy required by the user.
Neutronic Design of the RA10 Research Reactor´s Core
The RA-10 Research reactor is a multipurpose open-pool type reactor, currently being designed jointly between INVAP and CNEA, and owned by this last institution. It has a nominal fission power of 30 MW, with the capability of producing several radioisotopes, while providing suitable conditions for many types of irradiation tests. The reactor core is located inside a chimney, surrounded by heavy water contained in the Reflector Vessel, filled with Heavy Water. The whole assembly is at the bottom of the Reactor Pool, which is full of de-mineralized light water acting as coolant, moderator and biological shielding. The Fuel compound is LEU-U 3 Si 2. Reactor shut down can be achieved by two different means, which are the insertion of six Control Rods into the core, or the partial drainage of the heavy water from the Reflector Vessel. Two types of neutron sources are foreseen: a cold neutron source with two tangential beams, and thermal neutron source with two beams; both with several ne...
Conceptual core model for the reactor core test
1970
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Neutronic analysis of denaturing plutonium in a thorium fusion breeder and power flattening
Energy Conversion and Management, 2005
The purpose of this study is to denature nuclear weapon grade quality plutonium in a thorium fusion breeder. Ten fuel rods containing the mixture of ThO 2 and PuO 2 are placed in a radial direction in the fissile zone where ThO 2 is mixed with variable amounts of PuO 2 to obtain a quasi-constant nuclear heat production density. The plutonium composition volume fractions in the fuel rods are gradually increased from 0.1% to 1% by 0.1% increments. The fissile fuel zone is cooled with four various coolants with a volume fraction ratio of 1 (V coolant /V fuel = 1). These coolants are helium gas, flibe ''Li 2 BeF 4 '', natural lithium and eutectic lithium ''Li 17 Pb 83 ''. Nuclear weapon grade quality 239 Pu in the fuel composition is denatured due to the accumulation of the 240 Pu isotope in the fissile zone after 18 months of plant operations. Under a first wall fusion neutron current load of 2.222 • 10 14 (14.1 MeV n/cm 2 s), which corresponds to 5 MW/m 2 , by a plant factor of 100%, at the end of the plant operation, the fissile fuel enrichment quality between 6.0% and 10% is obtained depending on the coolant types. During the plant operation, the tritium breeding ratio (TBR) should be at least 1.05. In the selected blanket, only the flibe coolant is already self sustaining at start up. The TBR increases steadily due to the higher neutron multiplication rate during the plant operation period. The highest TBR is obtained for the eutectic lithium coolant 1.4035, followed by the flibe coolant 1.3095, helium gas coolant 1.2172 and natural lithium coolant 1.0553 at the end of the operation period of 48 months. The energy multiplication factor M changed between 2.1731 and 6.6241 depending on coolant type during the operation period. The peak to average fission power density ratio C in the blanket decreases
Progress in Nuclear Energy, 2007
Pakistan Research Reactor-1 (PARR-1) is a typical MTR type swimming pool reactor utilizing low enriched uranium (LEU), i.e. 19.99% enriched in 235 U, silicide dispersion fuel of density 3.28 gU/cm 3 . This simulation study was conducted by employing the standard reactor physics simulation codes WIMS-D/4, CITATION, and a burnup analysis code FCAP along with a reactor thermal hydraulic simulation code PARET. The present study shows that by directly substituting LEU silicide dispersion fuel of density 4.8 gU/cm 3 in place of fuel currently in use in PARR-1 and by loading beryllium elements at the unused 9 Â 6 position of PARR-1 grid plate, a smaller equilibrium core can be designed that can provide 82% higher neutron fluxes at the central flux trap position at 14% lower cost than the existing core. Fuel cycle length of this core is also two days larger than the existing core and this core can be operated safely at the existing power of 10 MW with the existing coolant flow rate of 1000 m 3 /h. A possible use of LEU UeMo monolithic fuel of density 15.3 gU/cm 3 with some adjustment in fuel to moderator ratio and use of Be reflector would provide 89% higher neutron flux in PARR-1 at 29% lower cost. Fuel cycle length of this core will be five days shorter than the existing core and it will require 48% more coolant flow rate for its safe operation at 10 MW.