A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the nodal few-group neutron diffusion theory in 3-dimensional Cartesian geometry for a typical pressurized water reactor (PWR) static and transient analyses, especially for reactivity initiated accidents (RIA).The spatial variables are treated by using a polynomial nodal method (PNM) while for the neutron dynamic solver the adiabatic and improved quasi-static methods are adopted. A simple single channel thermal-hydraulics module and its steam table is implemented into the code. Verification works on static and transient benchmarks are being conducted to assess the accuracy of the code. For the static benchmark verification, the IAEA-2D, IAEA-3D, BIBLIS and KOEBERG light water reactor (LWR) benchmark problems were selected, while for the transient benchmark verification, the OECD NEACRP 3-D LWR Core Transient Benchmark and NEA-NSC 3-D/1-D PWR Core Transient Benchmark (Uncontrolled Withdrawal of Control Rods at Zero Power). Excellent agreement of the NODAL3 results with the reference solutions and other validated nodal codes was confirmed
This paper presents the generation and the verification of the few-group homogenized cross-sections for Batan-Fuel. This code is routinely used for fuel management in the RSG-GAS. The Monte Carlo code Serpent 2 code, in conjunction with the latest nuclear data library ENDF/B-VIII.0, was used. Calculations using the existing newly generated few-group cross-section data were carried out for the 88th core. The calculated core parameters such as excess reactivity and control rod worth are compared to the available experimental data. On the other hand, the fuel burnup fraction and radial power peaking factor (PPF) are compared to the results of Serpent 2. It was shown that the new cross-section data have more consistency and a better agreement excess reactivity and control rod worth compared to the experimental data. Similarly, the U-235 burnup fraction and radial power peaking factor by the new cross-section data are also shown to concur well with Serpent 2. The newly generated few-group cross-sections are recommended to replace the existing ones for the fuel management of RSG-GAS.
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