JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA
Not a member yet
289 research outputs found
Sort by
RISK ASSESSMENT ON THE DECOMMISSIONING STAGE OF INDONESIAN TRIGA 2000 RESEARCH REACTOR
No full textDecommissioning is the final stage of a nuclear reactor. In preparing the decommissioning plan, one of the important elements that need to be considered is safety assessment. During decommissioning, there are many complex tasks to be done where the radiological and non-radiological hazards arise and can significantly affect not only the workers but also the general public and the environment. Indonesia has no experience with nuclear reactor decommissioning, so it is necessary to study various experiences of decommissioning activities in the world. This study proposes a framework to implement the safety assessment on the decommissioning of the TRIGA 2000 research reactor. The framework was developed on desk-based research and analysis. The proposed framework involves the facility and decommissioning activities, hazard identification, hazard analysis, hazard evaluation, hazard or risk control, and independent review
MEASURED AND CALCULATED INTEGRAL REACTIVITY OF CONTROL RODS IN RSG-GAS FIRST CORE
Full text linkThe 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
PREDICTION OF AP1000’S NUCLEAR REACTOR PRESSURE VESSEL TEMPERATURE DURING NORMAL OPERATION
Full text linkModeling of thermal-hydraulic calculations for the AP1000 core to predict the reactor pressure vessel (RPV) temperature has been carried out. The reactor’s primary coolant system transfers the heat produced in the reactor fuel during reactor operation to the steam generator. Part of the heat will also be transferred from the coolant to the reactor vessel and the pipe. This paper presents the calculation result of the RPV temperature prediction during AP1000 normal operation. Calculations were performed using COBRA-EN code for analyzing the core thermal hydraulics and using analytics for predicting the RPV temperature. These methods were carried out with the aim to predict the RPV temperature as well as at steady state nominal power conditions, at the function of flow, and at power fluctuation conditions. The calculation results at nominal power 3400 MWt (100% heat generated in fuel was assumed) and thermal design flow with 10% tube plugging (TDF2) of 48,443.7 ton/hr, for the minimum system pressure of 15.1 MPa, nominal system pressure of 15.513 MPa, and design system pressure of 17.133 MPa, show that the core outlet coolant temperature is 326.96°C, 327.01°C, and 327.22°C, and the RPV temperature is 303.65°C, 303.87°C, and 306.67°C, and the minimum departure from nucleate boiling ratio (MDNBR) is 3.21, 3.29, and 3.01, respectively. During reactor operation at a fixed nominal power of 3400 MWt, nominal system pressure, and under the condition of flow fluctuation, the maximum RPV temperature is shown to be 303.87°C
ACKNOWLEDGEMENT
No full textThis Section consists of Abstract Collection Page, Keywords Index Page and Acknowledgment Pag
DOSE DISTRIBUTION ANALYSIS OF PROTON THERAPY FOR MEDULLOBLASTOMA CANCER WITH PHITS 3.24
Full text linkOne of the developments in particle therapy is proton radiation therapy. Meanwhile, a limited number of available proton therapy facilities makes research related to proton therapy difficult. Therefore, there is a need for alternative proton therapy simulations using programs other than those in proton therapy facilities. This research was aimed to simulate medulloblastoma brain cancer which children often experience.The program used in this research was PHITS version 3.24. The human body was modeled with the revised ORNL-MIRD phantom for a 10-year-old child. The therapy scheme was a whole posterior fossa boost of 19.8 Gy. The proton passive scattering was simulated by passing a uniform proton beam through the aperture and compensator with energy variations. The proton pencil beam scanning was simulated with small cylindrical beams with a radius of 0.5 cm, which were adjusted to the planning target volume with layers variations.The total duration to give the prescription dose was 550 seconds with passive scattering and 605 seconds with pencil beam scanning. In passive scattering, the OAR(s) with the most significant percentage of absorbed dose were the skin, cranium, and muscle, i.e., 8.22 ± 0.15 %, 5.51 ± 0.05 % and 1.39 ± 0,04 % respectively to their maximum tolerated dose, while in the pencil beam scanning, the OAR(s) with the most significant percentage of absorbed dose were the skin, cranium, and muscle, i.e., 5.42 ± 0.08 %, 4.43 ± 0.05 % and 0.51 ± 0.05 % respectively to their maximum tolerated dose. Dose distribution in passive scattering was relatively better than in pencil beam scanning in terms of dose homogeneity using dose sampling analysis at some points within the planning target volume
NEUTRONIC ANALYSIS OF THE VVER-1200 LATTICE CELL FUEL USING WIMSD-5B CODE
Full text linkThe calculation of safety parameters in nuclear reactors has an important influence on nuclear reactor control and safety. The infinite multiplication factor, reactivity coefficients, and power peaking factor parameters are the most important safety parameters for determining reactor status. The aim of the present study is to analyze the behavior of the nuclear safety parameters for the VVER-1200 core in a normal state of reactor operation. A lattice cell fuel model of the VVER-1200 reactor core was performed using WIMSD-5B. The cross-section library data based on the ENDF/B-VIII.0 was used. The investigated parameters were the value of infinite multiplication factor with different pitch, temperature, enrichment, and boron concentration. The calculation also investigated the reactivity coefficient parameters. The verification of WIMS model VVER-1200 was performed by comparing the results of the WIMSD-5B code with VVER-1200 data in the SAR document, and it was implied that they are in good agreement. The calculated values of reactivity coefficients illustrated a safe behavior
ANALYSIS OF FUEL TEMPERATURE REACTIVITY COEFFICIENT OF THE PWR USING WIMS CODE
Full text linkThe 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
Bagian Depan Vol.24 No. 1 (2022): February 2022
No full textThis section consists of Cover Page, Editorial Board Page, Peer Reviewer Page, Table of Content Page and Preface Pag
ANALYSIS OF THE PPF VALUE DEPENDENCE ON THE FUEL BURNUP
No full textThe RSG-GAS reactor has been operated in a safe and reliable manner for about 35 years since it commenced in operation in 1987 to serve radioisotopes production, NAA, neutron beam experiments, material irradiation, and reactor physics experimental activities as well as training. PPF value is necessary to determine by calculation because it is impossible to determine by experiment and also has a strong relation to the operation safety. The paper is intended to analyze the PPF values of the RSG-GAS reactor core as a function of burn up. The analysis is using WIMSD-5B/BATAN-3DIFF computer codes calculation. The result shows that the PPF values are significantly different for each burn-up or energy in MWD. The result also shows that the BATAN-3DIFF code accurately determines the PPF values of the RSG-GAS reactor core and supports the safety of reactor operation
Editorial Board
No full textThis section consists of Cover Page, Editorial Board Page, Peer Reviewer Page, Table of Content Page and Preface Pag