Claim Missing Document
Check
Articles

Found 7 Documents
Search

PERHITUNGAN INVENTORI NUKLIDA PADA PIN SEL BAHAN BAKAR REAKTOR PWR Santo Paulus; Syaiful Bakhri; Tukiran Surbakti
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 25, No 2 (2021): November 2021
Publisher : Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/sigma.2021.25.2.6451

Abstract

Perhitungan fisika reaktor untuk deplesi bahan bakar telah dilakukan, yang mengarah pada inventori isotop Pu di dalam bahan bakar sisa. Perhitungan inventori sotop bahan bakar dilakukan dengan program computer WIMSD-5B menggunakan data nuklir ENDFB-VII.1. Tujuan penelitian ini adalah untuk memprediksi jumlah atom Pu didalam bahan bakar selama reactor dioperasikan 3 tahun. Nilai parameter fluks dihitung program WIMSD dengan model  pin bahan bakar yang terletak di zona bahan bakar aktif. Bahan bakar yang dimodelnya terdiri dari tipe A dan B.  Hasil perhitungan faktor perkalian tak hingga pin sel PWR yang dihitung menggunakan paket program WIMSD berturut-turut adalah 1,13614 dan 1,19171 untuk bahan bakar tipe A dan B. Dari hasil perhitungan dapat dinyatakan bahwa jumlah Pu yang tersisa tergantung pada model bahan bakar yang digunakan. Nilai faktor perkalian tak hingga juga  sangat dipengaruhi oleh bentuk model bahan bakar yang digunakan
ANALYSIS OF FUEL TEMPERATURE REACTIVITY COEFFICIENT OF THE PWR USING WIMS CODE Santo Paulus Rajagukguk; Syaiful Bahkri; Tukiran Surbakti
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 1 (2022): February (2022)
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.1.6329

Abstract

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.
ANALYSES OF NEUTRON ABSORBER MATERIALS ON THE SAFETY PARAMETERS IN THE RSG-GAS REACTOR Lily Suparlina; Tukiran Surbakti; Surian Pinem; Purwadi Purwadi
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 2 (2021): June 2021
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.2.6278

Abstract

Shutdown system in RSG-GAS reactor is using neutron absorber. There are 3 kinds of absorber material in research reactors including Ag-In-Cd alloy, B4C, and Hf. In this works, analyses of different neutron absorbers on the main safety core parameters in the RSG-GAS research reactor are selected for analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, PPF and neutron flux . The RSG-GAS core silicide fuel is selected as the case study to verify calculations. A three-dimensional, four-group diffusion model is selected for core calculations. The well-known WIMSD-5B and Batan-3DIFF reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the B4C; also the lowest PPF is gained using the Hf material. The maximum point power densities belong to the inside fuel regions surrounding the CIP (centre irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the berrylium reflector. The greatest and least fluctuation of the point power densities are gained by using B4C and Ag-In-Cd alloy, respectively.
ANALYSIS OF COGENERATION ENERGY CONVERSION SYSTEM DESIGN IN IPWR REACTOR Ign. Djoko Irianto; Sriyono Sriyono; Sukmanto Dibyo; Djati Hoesen Salimy; Tukiran Surbakti; Rahayu Kusumastuti
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 1 (2022): February (2022)
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.1.6414

Abstract

The acceleration of national development, especially in the industrial sector, requires an adequate national energy supply. There are various types of energy sources which include conventional energy sources as well as new and renewable energy sources including nuclear energy. The problem is how to utilize these energy sources into energy that is ready to be utilized. BATAN as a research and development institution in the nuclear field has taken the initiative to contribute to the development of technology for providing electricity and other thermal energy, particularly reactor technology as a power plant and a provider of thermal energy. This research aims to analyze the design of the IPWR type SMR reactor cogeneration energy conversion system. The IPWR reactor cogeneration energy conversion system which also functions as a reactor coolant is arranged in an indirect cycle configuration or Rankine cycle. Between the primary cooling system and the secondary cooling system is mediated by a heat exchanger which also functions as a steam generator. The analysis was carried out using ChemCAD computer software to study the temperature characteristics and performance parameters of the IPWR reactor cogeneration energy conversion system. The simulation results show that the temperature of saturated steam coming out of the steam generating unit is around 505.17 K. Saturated steam is obtained in the reactor power range between 40 MWth to 100 MWth. The results of the calculation of the energy utilization factor (EUF) show that the IPWR cogeneration configuration can increase the value of the energy utilization factor up to 91.20%.
NEUTRONIC AND THERMAL HYDRAULICS ANALYSIS OF CONTROL ROD EFFECT ON THE OPERATION SAFETY OF TRIGA 2000 REACTOR Surian Pinem; Tukiran Surbakti; Iman Kuntoro
Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol 25, No 3 (2019): Oktober, 2019
Publisher : website

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (500.016 KB) | DOI: 10.17146/urania.2019.25.3.5576

Abstract

NEUTRONIC AND THERMAL HYDRAULICS ANALYSIS OF CONTROL ROD EFFECT ON THE OPERATION SAFETY OF TRIGA 2000 REACTOR. Analysis of neutronic and thermal-hydraulics parameters of whole operation cycle is very important for the safety of reactor operation. During the reactor operation cycle, the position of the control rods will change due to reactivity changes. The purpose of this study is to determine the effect of control rods position on neutronic and thermal-hydraulics parameters in relation to the safety of reactor operation of the TRIGA 2000 reactor using silicide fuel of MTR plate type. Those parameters are power peaking factor, reactivity coefficients, and steady-state thermohydraulic parameters. Neutronic calculations are performed using a combination of WIMSD/5 and Batan-3DIFF codes and for thermal-hydraulics the calculations are done using WIMSD/5 and MTRDYN codes. The calculation results show that the reactivity coefficient values are negative for all control rod positions both at CZP and HFP conditions. The MTC value decreases when the control rod is inserted into the active core while the FTC value increases. The total ppf results and temperature in steady-state rise when the control rods are inserted of into the active core whereby the maximum value occurs at the position of the control rods of 20 cm from the bottom of the active core. The calculation results of ppf, reactivity coefficient, and thermal-hydraulics parameters lay below safety limits, indicating that the TRIGA 2000 reactor can safely use U3Si2-Al silicide fuel as a substitute fuel for cylindrical type fuel.Keywords: neutronic, thermal-hydraulic parameter, control rod effect, TRIGA 2000, silicide fuel.
ANALISIS PENGARUH DENSITAS BAHAN BAKAR SILISIDA TERHADAP PARAMETER KINETIK TERAS REAKTOR RSG-GAS Tukiran Surbakti; Surian P; Tagor S
Jurnal Penelitian Fisika dan Aplikasinya (JPFA) Vol. 3 No. 1 (2013)
Publisher : Universitas Negeri Surabaya

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.26740/jpfa.v3n1.p19-30

Abstract

Saat ini RSG-GAS menggunakan elemen bakar silisida 2,96 g U/cc. Untuk meningkatkan waktu operasi reaktor maka akan direncanakan untuk mengganti elemen bakar silisida dengan kerapatan yang lebih tinggi. Keuntungan reaktor dengan bahan bakar kerapatan tinggi adalah dapat lebih efektif dan efisien. Maka perlu dilakukan perhitungan parameter kinetik teras silisida kerapatan tinggi mengingat pengaruhnya sangat penting untuk keselamatan operasi reaktor. Parameter kinetik yang dihitung yaitu fraksi neutron kasip efektif, konstanta peluruhan neutron kasip, umur neutron serempak yang merupakan faktor utama dalam kontrol dan keselamatan. Bahan bakar silisida tipe pelat dengan densitas 2,96 - 4,8 gU/cm3 digunakan pada teras RSG-GAS untuk menganalisis perhitungan parameter kinetik. Perhitungan sel dilakukan dengan paket program WIMSD-5B dan paket program Batan-2DIFF digunakan untuk perhitungan teras. Hasil perhitungan menunjukkan bahwa harga fraksi neutron kasip turun dengan naiknya densitas bahan bakar. Turunnya nilai parameter kinetik ini tidak mengganggu pergantian bahan bakar ke densitas yang lebih tinggi. Turunnya nilai parameter kinetik rata-rata dari densitas 2,96 gU/cm3 ke 3,55 gU/cm3 adalah 1,3 % sedangkan dari densitas 2,96 gU/cm3 ke 4,8 gU/cm3 adalah 2,2 % . Sehingga jika dilakukan pergantian bahan bakar maka ditinjau dari segi neutronik dan parameter kinetiknya tidak akan mengalami perubahan dalam pola operasi reaktor atau manajemen bahan bakar dan tidak akan berpengaruh terhadap keselamatan operasi reaktor.
NEUTRONIC CALCULATION FOR PWR MOX FUEL PIN CELLS WITH WIMSD-5B CODE Santo Paulus Rajagukguk; Syaiful Bakhri; Tukiran Surbakti
Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol 28, No 1 (2022): Februari, 2022
Publisher : website

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2022.28.1.6610

Abstract

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.