Analysis of the control rod insertion is important as it is closely related to reactor safety. Previously, the analysis has been carried out in RSG-GAS during static condition, not as a function of the fuel fraction. The RSG-GAS reactor in one cycle is a function of the fuel burn-up. It is necessary to analyze RSG-GAS core reactivity insertion as a function of the fuel burn-up to determine the behavior of the reactor, especially in uncontrolled operations such as continuous pulling of control rods. This analysis is carried out by the computer simulation method using WIMSD-5B and MTR-DYN codes, by observing power behavior as a function of time due to neutron chain reactions in the reactor core. Calculations are performed using point kinetics equation, and the feedback effect will be evaluated using static power coefficient and fuel burn-up function. Analyzes were performed for the core configuration of the core no. 99, by lifting the control rod or inserting positive reactivity to the core. The calculation results show that with the reactivity insertion of 0.5% Δk/k at start-up power of 1 W and 1 MW, safety limit is not exceeded either at the beginning, middle, or end of the cycle. The maximum temperature of the fuel is 135°C while the safety limit is 180°C. The margin from the safety limit is large, and therefore fuel damage is not possible when power excursion were to occur.
BATAN has three aging research reactors, so it is necessary to design a new, more modern MTR type reactor using high-density, low enrichment uranium molybdenum fuel. The thermal neutron flux at the irradiation position is an important concern in the design of research reactors. This analysis is performed using standard computer codes WIMSD-5B and Batan-FUEL. The purpose of this study is to analyze the effect of the core configuration with safety control rods and neutronic parameters using the diffusion method calculation. The reactor core consists of 16 fuel elements and four control rods placed in the 5 x 5 position of the grid plate and is loaded the reflector elements outside the core. The cycle length is also a concern, not less than 20 days, and the reactor can be operated safely with a power of 50 MW. The calculation results show that for the highest fuel loading, which is 450 grams of U7Mo/Al fuel with D2O as a reflector, it will provide the lowest thermal neutron flux at the center of the core irradiation position, namely 1.0 x1015 n/cm2s. The core fuel cycle length will be up to 39 days, meeting the expected acceptance and safety criteria.
Manfaat yang luas penggunaan reaktor riset membuat banyak negara membangun reaktor riset baru. Kecenderungan saat ini adalah tipe reaktor serbaguna dengan teras kompak untuk mendapatkan fluks neutron yang tinggi dengan daya yang relatif rendah. Reaktor riset di Indonesia realtif sudah tua. Oleh karena itu diperlukan desain reaktor riset baru sebagai alternative atau modefikasi desain, kelak pengganti reaktor riset yang sudah ada. Tujuan dari riset ini untuk melengkapi data desain teras TRIGA Bandung sebagai salah satu parameter yang penting dan dibutuhkan untuk menyusun LAK serta persyaratan untuk perizinan desain. Perhitungan dilakukan untuk memahamipola operasi dan majmeen mahan bakar reaktor TRIGA Bandung dengan konfigurasi teras setimbang yang optimal terdiri dari 16 bahan bakar dan 4 batang kendali dengan grid teras 5x5 dan daya 2 MW. Perhitungan manajemen bahan bakar desain teras TRIGA Bandung dilakukan untuk bahan bakar U3Si2-Al dengan kerapatan 2,96 gU/cc. Perhitungan dilakukan dengan paket program WIMSD-5B dan BATAN-FUEL. Hasil pehitungan menunjukankan bahwa dengan polaoperasi satu dan dua parameter operasi tidakada yang dilampaui. Namun hal ini tidak dapat digunakan untuk menambah peningkatan fraksibakar. Dalam hal ini tidak ditemukan peningkatan fraksi bakar yang signifikan dengan hanya merubah konfigurasi teras, hanya bisa memperpanjang siklus operasi.Kata kunci: desain konseptual, bahan bakar uranium-silisida, manajemen bahan bakar, WIMSD-5B, BATAN-FUEL
The control rod worth is one of the important parameters for the operation of a nuclear reactor. Proper measurement and calculation of the control rod worth are essential for the safe reactor operation under normal and transient conditions that are initiated by a postulated event such as a stuck rod, control rods ejection, etc. This paper presents calculation results of integral reactivity of the RSG-GAS research reactor first core and its comparison with the experimental data. Calculations were performed using the continuous energy transport code Serpent 2 with ENDF/B-VIII.0 nuclear data. Integral reactivity measurement was done by compensating method with control rod bank, regulating rod, and reactivity meter. Calculations are carried out for each method used in control rod measurement data with an aim to validate calculated results to experimental data. Compared with the measured experiment data, there are no significant differences in calculation results of integral reactivity. The maximum difference of the control rod's total reactivity is 1.26% compared to the measurement carried out by compensating method with regulating rod.
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