STUDI PARAMETER DESAIN TERAS PWR INTEGRAL DENGAN BAHAN BAKAR MOXMENGGUNAKAN PROGRAM SRAC. PWR Integral menggunakan bahan bakar MOX dapat menjadi solusi untuk mengatasi permasalahan elektrifikasi di negara kepulauan dan mengakomodasi pergantian bahanbakar dari UO2 menuju MOX. PWR Integral merupakan reaktor nuklir modular dengan daya 160 MWt untuksatu modulnya. Saat ini, PWR Integral menggunakan bahan bakar UO2 sebagai bahan bakar utama. Penelitianini bertujuan untuk mengetahui perubahan performa teras reaktor dari yang sebelumnya menggunakan bahanbakar UO2 menjadi bahan bakar MOX. Penelitian ini dilakukan dengan cara memvariasikan rasio PuO2 dalambahan bakar MOX, jumlah bahan bakar, dan jenis kelongsong dengan menggunakan sistem kode SRAC2006.Studi parameter dilakukan dengan memperhatikan nilai keff, rasio konversi, dan jumlah aktinida, serta akandilakukan pula perbandingan dengan teras reaktor yang menggunakan bahan bakar UO2. Hasil penelitianmenunjukkan performa teras reaktor dengan bahan bakar MOX lebih baik daripada UO2. Desain teras PWRIntegral yang paling optimum adalah teras dengan rasio PuO2 dalam MOX 12%, diameter teras 165 cm, danjenis kelongsong Zircalloy-4. Hal ini dapat dilihat dari periode kekritisan teras reaktor yang mencapai 1521hari dengan nilai CR paling rendah adalah 0,622004. Excess reactivity yang dimiliki juga lebih rendah yaitu1,0745932 dimana desain UO2 bernilai 1,1035821. Aktinida yang dihasilkan mengalami tren penurunan seiringreaktor beroperasi.
Diffusion approximation is an important approximation used to model a nuclear reactor core with a quite good accuracy and less computational cost. This approximation has been used widely around the globe for various kinds of nuclear reactors. This diffusion approximation is applied in a two-step method, a method combining a high fidelity transport code and low fidelity diffusion code. Meanwhile, innovations in the nuclear core model continue to make the nuclear reactor core safer, more robust, and smaller. The trend of creating smaller and more modular reactor core is emerging in the last ten years. These innovations will affect the core modeling system. Consequently, for smaller reactors, it is important to evaluate the capability of diffusion approximation if one wants to use a computationally cheaper method to model the reactor core. In this paper, neutron diffusion calculation for 160 MWth integral pressurized water reactor (IPWR) core was conducted using the PARCS nodal diffusion code employing the few-group spatially homogenized cross-sections generated by the Serpent Monte Carlo code. Due to its capability to model any reactor geometry in the high-resolution calculation, the results from Serpent were also used as a reference. Two important parameters are compared between PARCS and Serpent: effective neutron multiplication factor and core power distribution. For the full IPWR core model, a discrepancy of 564 pcm between PARCS and Serpent k eff was observed, while the radial power distribution had a maximum error of 4.71 %. It can be said, to some extent, that the diffusion approximation can be applied to IPWR core analysis. However, further improvement is indeed required if one wants more accurate results with low computational costs.
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