Perhitungan sel bahan bakar teras reaktor PWR telah dilakukan dengan menggunakan program komputer WIMSD-5B. Perhitungan dilakukan untuk mengetahui karakteristik neutronik bahan bakar teras reaktor PWR dengan variasi daya. Karakteristik neutronik diketahui dengan memperoleh konstanta makroskopik seperti k-inf, koefisien difusi, tampang lintang serapan dan fisi. Generasi sel bahan bakar dilakukan dengan 69 grup energi neutron pada program transport satu dimensi (WIMSD-5B) menggunakan ENDF-BVII.1 data file. Sel satuan diperhitungkan pada perangkat elemen bakar dengan model cluster dengan susunan square pitch, kemudian dihitung dimensi satuan selnya. Satu satuan sel terdiri dari satu satuan bahan bakar dan moderator. Dari satu satuan sel ekivalen tersebut diperoleh data dimensi sel sebagai data masukan program WIMSD-5B yang dikenal dengan annulus. Bahan bakar yang digunakan adalah UO2 dan bentuk geometrinya pin cell bahan bakar. Hasil perhitungan faktor multiplikasi tak terhingga sel teras PWR yang dihitung dengan menggunakan paket program WIMSD-5B adalah 1,302338 dan fraksi bakar 37,12 GWD/TU. Dari hasil perhitungan dapat dinyatakan bahwa nilai faktor multiplikasi tak terhingga, konstanta difusi, tampang lintang serapan dan nu-fisi sangat dipengaruhi oleh bentuk model yang digunakan.Kata kunci: bahan bakar, teras PWR, WIMSD-5B, energi neutron, konstanta makroskopik
The Fuel Temperature Reactivity Coefficient (FTRC) is an important parameter in design, control, and safety, particularly in PWR reactor. It is then very important to validate any new library for an accurate prediction of this parameter. The objective of this work is to determine the value of the FTRC parameter using the new WIMDS library based on ENDF/BVIII.0 nuclear data files. For this purpose, it is used a set of light water moderated lattice experiments as the PWR-1175 MWe experiment critical reactors, the reactor using UO2 fuel pellet. The analysis is used with WIMSD-5B lattice code with original cross-section libraries and WIMSD-5B with ENDF/B-VIII.0 new cross-section libraries. The results showed that the fuel temperatures reactivity coefficients for the PWR reactor using original libraries is – 3.10 pcm/K with enrichment of 3.1% but for ENDF/B-VlII.0 libraries is – 3.00 pcm/K. Compared to the experimental data of the reactor core, the difference is in the range of 6.9 % for ENDF/B-VIII.0 libraries. It can be concluded that for the reactor, it is better to use ENDF/B-VIII.0 libraries because the original library is not accurate anymore.
A thorium-fueled benchmark comparison was made in this study between state-of-the-art codes, WIMSD-5B code to MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations as part of efforts to examine the possible benefits of using thorium in PWR fuel. WIMSD-5B calculations employ the same model as a reference, MOCUP, and CASMO, however, there are some variances in methodology and cross-section libraries. On a PWR pin cell model, eigenvalue and isotope concentrations were examined up to high burnup. The eigenvalue comparison as a function of burnup is good, with a maximum difference of less than 5% and an average absolute difference of less than 1%. The isotope concentration comparisons outperform a set of ThO2-UO2 fuel benchmarks and are comparable to a set of uranium fuel benchmarks previously published in the literature. As a function, the eigenvalue comparison The actinide and fission product data sources for a typical thorium fuel are reported in the WIMSD-5B burnup calculations. The reasons for discrepancies in coding are examined and explored.Keywords: Thorium, PWR Fuel, Burn up, Pin Cell, WIMSD-5B
NEUTRONIC CALCULATION FOR PWR MOX FUEL PIN CELLS WITH WIMSD-5B CODE. The WIMSD-5B thermal reactor lattice cell code is used in many laboratories for research reactor calculations and power reactors. The program uses the Wigner-Seitz approximation for cell pin calculations. The approximation has been widely applied to the pin of UO2 cells and has shown good results in previous studies but can produce incorrect results when used for pin cells in MOX fuels. This paper investigates the use of the WIMS-5B code to calculate the neutron multiplication factor and depletion for MOX fuel pin cells. Calculations were performed using the WIMSD-5B code updated with the ENDF-BVIII.0 library. The outer scattering boundary condition was used to overcome the effect of the Wigner-Seitz approach on the lack of MOX fuel. Results of this study indicates that most of the results obtained using ENDF-BVIII.0 are better than ENDF-BVII.1, and this is in line with expectations. The difference in the maximum k-inf value obtained from this library occurs in the fuel that has the greatest enrichment. On the other hand, the addition of the outer scattering limit improves the results obtained using ENDF-BVIII.0, causing a slight improvement for other libraries. This shows that by using appropriate libraries and the addition of the scattering outer limit, the Wigner Seitz approximation for the MOX pin cell pins in WIMS-D5 can yield quite accurate results.Keywords: Wigner-Seitz approximation, WIMS-D5 code, MOX fuel, Doppler reactivity.
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