RSG-GAS Reactor Research Articles

Optimizing asset management for Structure System Components of RSG-GAS: A reliability-centric approach

This study focuses on the Structure, System, and Components (SSC) of G.A Siwabessy Multipurpose Reactor (RSG-GAS), emphasizing the integration of reliability improvement within asset management optimization. The research aims to derive Maintenance Priority Index (MPI) values, contributing to system reliability ratings, crucial for assessing component functionality and estimating Remaining Useful Life (RUL) The methodology involves measuring critical aspects of system quality, safety, and cost. The MPI value, representing 10 % of SSC critical components, guides subsequent reliability calculations. Utilizing failure data from RSG-GAS components (2010–2018) and reactor core configurations (70th-96th core), SSC reliability is simulated using Monte Carlo simulation, with RUL based on real data. High MPI values are identified for critical components such as KBE01/AP01-02_B in the primary purification system, JE-01/AP01-02_A in the primary cooling system, and PA01-02/CR001_A in the secondary cooling system. Maintenance intervals of 100 days are recommended for KBE01/AP 01–02_B and JE-01/AP01-02_A, exceeding 50 % reliability, while PA01-02/CR001_A maintenance can extend to 75 days. Monte Carlo simulation results, with 1000 samples, closely align with real reliability values. The RUL result projects operational lifetimes of 225.6, 221.4, and 171.5 days for KBE01/AP 01–02_B, JE-01/AP01-02_A, and PA01-02/CR001_A components. This study improves asset management, offering practical insights for critical safety system component maintenance planning.

Neutronic Analysis of the RSG-GAS Fuel Using Burnable Poison

Control and safety of nuclear reactors are significantly influenced by the determination of safety parameters. The three most crucial safety factors for assessing reactor status are the infinite multiplication factor, reactivity coefficients, and power peaking factor. The objective of the present study is to examine how the RSG-GAS fuel safety parameters behave in a typical reactor operation state. A lattice cell fuel model of the fuel lattice of the RSG-GAS reactor core was modeled using WIMSD-5Bwith cross-section library data based on ENDF/B-VIII.0. The value of the infinite multiplication factor with various burnable poison concentrations, as well as the moderator and fuel temperatures, were the variables that were examined. The reactivity coefficient parameters were similarly analyzed. By comparing the WIMSD-5B code results with information from the SAR document, the WIMS model for RSG-GAS fuel was verified, and it was inferred that the parameters are in good agreement. Safe behavior uses the predicted reactivity coefficient values as an example.

Evaluation of neutronics parameters during RSG-GAS commissioning by using Monte Carlo code

Several reactor physics commissioning experiments were conducted to obtain the neutronic parameters at the beginning of the G.A. Siwabessy Multi-purpose Reactor (RSG-GAS) operation. These parameters are essential for the reactor to safety operate. Leveraging the experimental data, this study evaluated the calculated core reactivity, control rod reactivity worth, integral control rod reactivity curve, and fuel reactivity. Calculations were carried out with Serpent 2 code using the latest neutron cross-section data ENDF/B-VIII.0. The criticality calculations were carried out for the RSG-GAS first core up to the third core configuration, which has been done experimentally during these commissioning periods. The excess reactivity for the second and third cores showed a difference of 510.97 pcm and 253.23 pcm to the experiment data. The calculated integral reactivity of the control rod has an error of less than 1.0% compared to the experimental data. The calculated fuel reactivity value is consistent with the measured data, with a maximum error of 2.12%. Therefore, it can be concluded that the RSG-GAS reactor core model is in good agreement to reproduce excess reactivity, control rod worth, and fuel element reactivity.

Weibull model for RUL estimation at RSG-GAS reactor implemented on PA01-AP01 secondary pump

Abstract Remaining Useful Life (RUL) estimation has been extensively explored in recent years. RUL could be used in deciding the maintenance timeline or inspection interval for the Reaktor Serba Guna – G. A. Siwabessy (RSG-GAS reactor). RSG-GAS reactor is a pool-type research reactor (built by the Interatom Internationale of Germany) and has been operating for more than 30 years to date. This study aimed to propose a Weibull model to find the RUL estimation value of the distribution parameters of the mean time to failure (MTTF). Therefore, the RSG-GAS reactor would be higher safety, longer lifetime and higher reliability with a smaller failure rate including for the PA01-AP01 secondary pump. The research methodology is processing data collection and estimating the parameters of the Weibull model to determine maintenance timeline or inspection intervals based on the MTTF value in case the reliability has reached the targeted percentage. Results show that the RUL estimation has been obtained for the RSG-GAS reactor. In the implemented study, a maintenance timeline has been stipulated for the PA01-AP01 secondary pump (with the model of KSB and type of CPK-S350-400) for the reliability of 90% and RUL estimation of circa 29 days.

OXIDE LAYER CHARACTERIZATION OF AlMg2 CLADDING OF IRRADIATED U3Si2/Al FUEL WITH 4,8 gU/cm3 DENSITY

OXIDE LAYER CHARACTERIZATION OF AlMg2 CLADDING OF IRRADIATED U3Si2/Al FUEL WITH 4,8 gU/cm3 DENSITY. To investigate the performance of AlMg2 cladding in the U3Si2/Al dispersion fuel, oxide layer characterization of AlMg2 cladding of the irradiated U3Si2/Al fuel with 4.8 gU/cm3 density was conducted. The oxide layer on the surface of AlMg2 cladding is one of the changes that occur on the cladding after the U3Si2/Al fuel plate has been irradiated in the RSG-GAS reactor to a burn-up of ∼40%. The characterization and observation of the oxide layer was conducted using SEM (Scanning Electron Microscope) and Energy-dispersive X-ray spectroscopy (EDS). Samples with a size of 3x3 mm were taken from the middle of the fuel plate (middle position). After cutting, metallographic preparation includes mounting, grinding, polishing, and ultrasonic cleaning. SEM preparation was carried out by sputter coating using Au layer. The oxide layer on the AlMg2 cladding has a thickness of 10.3 µm with a uniformly distributed cracks along the oxide layer.Keyword: LEU, uranium-silicide, post-irradiation examination, AlMg2 cladding, oxide layer.

Neutron Beam Characterization on the Beam Tubes of 30 MW G.A. Siwabessy Reactor Using Monte Carlo

The neutron beam characterization had been conducted on six tubes of RSG-GAS reactor for BNCT application in a simulation using Monte Carlo method with MCNP computer code. The simulation was conducted by modeling the geometry and material of RSG-GAS reactor with a model of radiation source from 235U fission reaction in 40 fuel bundles type U3Si2Al with 235U levels of 19.75%. The distribution of neutron and gamma fluxes was simulated from the reactor core to the edge of beam tube of S1, S2, S3, S4, S5, and S6 with a tube length of 400 cm and a diameter of 30 cm. The highest neutron flux produced by the beam tube of S5 was 4.3×1010 cm−2s−1 with the highest gamma dose was 1362 Sv/j. The lowest neutron flux produced by the beam tube of S6 was 5.9×109 cm−2s−1 with a gamma dose was 51 Sv/j. Based on the characterization result, it was shown that the output neutron beam of each RSG-GAS beam tube had the potential to be used for many application.

An Improved Steady-State and Transient Analysis of the RSG-GAS Reactor Core under RIA Conditions Using MTR-DYN and EUREKA-2/RR Codes

Steady-state and transient analysis of reactor core under Reactivity-Initiated Accident (RIA) conditions are important for reactor operation safety. The reactor dynamics are influenced by neutronic and thermal-hydraulic aspects of the core. In this study, steady-state and transient analysis under RIA conditions of the RSG-GAS multipurpose reactor was carried out using MTR-DYN and EUREKA-2/RR programs. Neutronic calculations were performed using a few group cross-sections generated by Serpent 2 with the latest cross-section data ENDF/B-VIII.0. Steady-state conditions were carried out with a nominal power of 30 MW, while transient under RIA conditions occurred because the control rod was pulled too quickly while the reactor operated. These transient RIA conditions were performed for two cases, during start-up with an initial power of 1 W, and within power range with an initial power of 1 MW. Thermal-hydraulic parameters considered in this study are reactor power, the temperature of the fuel, cladding, and coolant. The calculated maximum fuel temperature at a steady state is 126.02°C. Meanwhile, the calculated maximum fuel temperature during RIA conditions at the initial power of 1 W and 1 MW are 64.38°C and 137.14°C, respectively. There are no significant differences in thermal-hydraulic parameters between each used program. The thermal-hydraulic parameters such as the maximum temperature of the coolant, cladding, and fuel under this postulated RIA condition are within the acceptable reactor operation safety limits.

ANALYSIS OF THE PPF VALUE DEPENDENCE ON THE FUEL BURNUP

The 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.

Heat balance analysis of RSG-GAS coolant system due to cooling tower refurbishment

The RSG-GAS reactor is designed to operate at a power level of 30MW, as stated in the Safety Analysis Report, and currently RSG GAS has been operating for 34 years. Aging of components has been anticipated to ensure the reactor operates safely. Some parts of the structure, system and components (SSC) will experience degradation/aging and must be replaced. Cooling tower is one of them, where the sub components are maintained and tested regularly. Cooling tower aging indication are known from the decreased performance of the reactor cooling system. For this reason, it is necessary to evaluate, both in terms of measurement and calculation analysis, before and after the refurbishment. The efficiency of the cooling tower has been calculated based on the measurement results of the cooling temperature, at the RSG-GAS operation at a power level of 30 MW. From the measurement results, it can be seen that there is a difference in air velocity inlet and outlet of module 1 and module 2. As well as the working temperature of the blower is uneven, the blower temperature ranges from 60.5°C-91.9°C. So that the operation of the RSG GAS could not last long. The efficiency of the cooling tower during the day ranges from 54.72% to 73.23%. The effect of decreasing cooling tower capability can be seen from the calculation results on the performance of each module individually. while its effect on the overall performance is known from the increase in the temperature of the coolant entering the reactor from the HE. The results of the cooling temperature measurement, after refurbishment, show that the cooling tower performance has been able to dissipate heat according to the initial design.

Analysis of control rod effect on the safety operation parameter of the RSG-GAS Core

The operation cycle of the RSG-GAS reactor is regulated by control rod positions from the lowest to uppest positions. It is of necessary to know the effect of this control rods positions to the safety operation parameter a long the cycle. The objective of this research is to confince that the changes of safety operation parameter still meet the its safety limits. The neutronics and thermalhydraulics parameter to be observed are power peaking factor, kinetic and steady state parameters. Calculations are conducted by a combination code of WIMSD/5 and MTRDYN. The main results of maximum axial power peaking factor is 1.84 at control rod position of 20 cm. The maximum temperature of cooling, cladding and fuel are consecutively 67.18 °C, 122.26 °C, and 123.24 °C. These results lay far under safety limits of maximum temperature of coolant and fuel of 100 °C dan 200 °C at steady state operation of the RSG-GAS reactor which means that the reactor operation is safe as long as the cicle operation.

The effect of adding impurities to the high density fuel matrix on the reactivity of the RSG-GAS core

The safety calculation of the RSG-GAS reactor core needs to be done if new high-density fuel is used in the RSG-GAS core. The use of new high-density fuels is very much needed in the RSG-GAS core so that it is necessary to calculate its safety. When using high density fuel there will be a “bottle neck” effect where there is a problem with the fuel and cladding interaction. According to previous researchers this can be overcome by adding metal elements Si, Ni or Ti to fuel. However, with the addition of metal elements Si, Ni or Ti to the fuel, it needs to be analyzed in the RSG-GAS core. The analysis was carried out calculations using a computer program by simulating the Si metal used 1%, 2%, 3% 4% and 5% in the fuel. The calculation of the additional metal element in the high-density fuel effect on the reactivity of the RSG-GAS core was carried out with the WIMSD-5B and Batan-FUEL programs. The WIMSD-5B program is a computer program consisting of several modules including the transport module which is used for the calculation of the generation of macroscopic constants for the 19.75% enriched uranium silicide fuel and the density of 4.8 gU/cc as a function of degree of burn. Core calculations were performed using a neutron diffusion program, namely Batan-FUEL. By calculating the core criticality at operating temperature, the core neutronic parameter values are obtained. The results of the analysis show that the amount of Si 5% in the high density fuel matrix can solve the bottle neck problem and the neutronic parameters are still within the safety limit. This value will determine how much the value of the fuel burn up and control rods are in accordance with the acceptance criteria so that the safety analysis can be carried out on the RSG-GAS core using high density. From the analysis, it was found that the addition of the element Si in the high-density fuel matrix did not find any significant changes in core reactivity and neutronic parameter.

THE VERIFICATION OF THE RSG-GAS REACTOR COOLING TOWER HEAT TRANSFER CAPACITY

The RSG-GAS reactor has been replaced and the technical specifications for the new cooling tower specify that the heat transfer capacity from the secondary cooling water to the environment is 5500 kW per module. Therefore, this study aims to verify the theoretical calculations of the heat transfer capacity using performance test data collected on the 30 MW power operation on December 20, 2018, such as the temperature of the primary and secondary coolant entering and exiting the cooling tower, wet bulb, and environmental dry bulb temperature, as well as the inlet and outlet air temperature. Furthermore, the data were used to calculate the heat transfer capacity from the secondary cooling water to the environment. The results showed that each cell of the RSG-GAS cooling tower reactor transfers the heat of approximately 5528.52 kW. This value is consistent with the technical specifications written in the revised RSG-GAS Safety Analysis Report 11.

MAP OF RADIOISOTOPE PRODUCTION AND BATAN RESEARCH REACTOR UTILIZATION

Currently, Indonesia through BATAN is operating three research reactors, namely the RSG-GAS reactor with the power of 30 MWt at Puspiptek south Tangerang (the first criticality in 1987), the TRIGA 2000 reactor with the power of 2 MW in Bandung which the first criticality in 1965 with the power of 250 kW, was increased to 1 MW in 1971, and further upgraded to 2 MW in 2000. Beside that, there is Kartini reactor with a power of 100 kW located in Yogyakarta (first criticality in 1979). These reactors are quite old, and in accordance with Bapeten regulations, have carried out the first periodic safety review, to obtain a reactor license for the next 10 years of operation. In line with this, one of BATAN's current national research programs is to increase the production of radioisotopes and radiopharmaceuticals, where reactors play a very important role in the production of certain isotopes. In tracing the data obtained from operational reports related to irradiation requests from reactor users, namely PTRR, PSTNT, and PT INUKI for radioisotope production, which has been carried out in the last 5 years, May 2015 until 25 August 2020, show that the irradiation request at RSG-GAS is still not optimal. In term of the utilization of RSG-GAS, it can still be optimized, which in this case needs to be balanced with post-irradiation processing capabilities. Meanwhile, from the results of tracing and data collection, it can be shown that at this time the reactors are still operating. The utilization activities of the reactors complement each other according to their age and facilities.

ANALYSES OF NEUTRON ABSORBER MATERIALS ON THE SAFETY PARAMETERS IN THE RSG-GAS REACTOR

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.

COMPARISON OF MEASURED AND CALCULATED CONTROL ROD REACTIVITY OF THE RSG-GAS CORE

COMPARISON OF MEASURED AND CALCULATED CONTROL ROD REACTIVITY OF THE RSG-GAS CORE . The reactivity worth and a calibration curve of control rods are important physical parameters on account of the nuclear safety of the RSG-GAS reactor core. Reactivity characteristics of control rods must be determined not only, but also in the first criticality and also after any substantial changes in the reactor core. The required time to measure the reactivity worth by various methods depends mostly on the quality of the reactivity measurement devices. The rod-insertion method has some specific advantages which make such compensation unnecessary while yielding both integral and differential reactivity worth curves, as well as total reactivity worth. In this study, the total control rod worth of RSG-GAS is calibrated experimentally using a rod-insertion method to verify calculation results for the new core configuration. Calculations were done by diffusion and Monte Carlo methods using Batan-3DIFF and MCNP5 codes. Total control rod reactivity worth of the control rod is obtained 17.54 $, 17.03 $, and 17.87 $ by experiment, Batan-3DIFF, and MCNP5 calculations, respectively. The relative difference between the experimental and calculated values of control rod reactivity worth is about 3.0 %, which indicates a good agreement between the applied experimental method and calculation.

EVALUATION OF EQUILIBRIUM CORE OPERATION OF THE RSG-GAS REACTOR

The Indonesian Multipurpose Reactor, RSG-GAS reactor will accomplish its first lifetime in December 2020. The reactor has been operated in safe and reliable manner for about 33 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. The paper is intended to evaluate its in-core fuel management that is the conformance between the theory and implementation of the equilibrium core. Evaluation of the reactor operation parameter was carried out for core numbers 91 – 100. The data show that excess reactivity, shutdown reactivity and control rod reactivity have no significant difference at each core. The result shows that the BATAN-FUEL accurately determine the equilibrium core and its fuel loading pattern.This in-core fuel management of the RSG-GAS reactor supports the safety of reactor operation.

ANALYSIS OF REACTIVITY INSERTION AS A FUNCTION OF THE RSG-GAS FUEL BURN-UP

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.

Validation and improvement of gamma heating calculation methods for the G.A. Siwabessy multipurpose reactor

Gamma heating, which is deposited in irradiated targets or samples, is an important issue in research reactors because it affects the safety of samples and the reactor operation. Gamma heating in the Reaktor Serba Guna G.A. Siwabessy (RSG GAS) can be measured or calculated using computer code. In this paper, we present the results obtained by using gamma calorimeters to measure the gamma heating in the central irradiation position in RSG GAS. The measurement results were verified and then compared with the calculation results obtained using the Gamset code. However, the accuracy of the calculation results obtained using Gamset was inadequate, with more than 20% error. Moreover, Gamset was initially created to calculate gamma heating in 35 MWth reactors, and the RSG GAS reactor has an operating power level of 30 MWth. To address these issues, we developed a new program called NewGamset, built with an informative graphical user interface-based display. NewGamset can employ new analytical approaches as it utilizes a calculation base that comprises 18 energy groups and can provide updated physical parameters of RSG GAS. The results of gamma heating measurements obtained using gamma calorimeters made of graphite, aluminum, iron, and zirconium were 2.20 ± 0.03 W/g, 2.25 ± 0.02 W/g, 2.58 ± 0.03 W/g, and 2.91 ± 0.10 W/g, respectively. In general, the calculation results were larger than the measurement results, and the average ratio of calculation to measurement result was 1.37 for Gamset and 1.02 for NewGamset. Hence, it can be concluded that NewGamset provides more accurate results than Gamset. We also report the gamma heating estimation for common elements irradiated in the RSG GAS silicide core, with an operating power level of 15 MWth and 30 MWth. We highly recommend that NewGamset be used for validating the gamma heating data of various target materials in RSG GAS.

FMEA for maintenance criterion at RSG-GAS reactor implemented on JE01-AP01 primary pump

Abstract Reaktor Serba Guna-G.A. Siwabessy (RSG-GAS reactor) is a pool type research reactor, located in Serpong village, Indonesia. For safety of RSG-GAS reactor, the maintenance level criterion of System, Structure and Component (SSC) can be carried out by assessing on occurrence likelihood of a potential risks due to ageing degradation and a Risk Priority Number (RPN) value for assigning maintenance criterion, respectively. The objective of present study is to propose using the Failure Mode and Effects Analysis (FMEA) for assigning the maintenance level criterion of SSC and implementing on JE01-AP01 primary pump within a RSG-GAS reactor. In this research paper, SSC screening and failure based maintenance methodology is identified and classified into suitable classes at radar diagram. Results show that SSC can be prioritized based on failure analysis so that maintenance level criterion for SSC can be determined. On the implemented study, preventive maintenance can be performed to JE01-AP01 primary pump because of its RPN value 270.

Reactivity insertion accident analysis during uranium foil target irradiation in the RSG-GAS reactor core

Analysis of the steady-state and reactivity insertion accident is very important for the safety of reactor operations. In this study, steady-state and reactivity insertion accident analysis when the low enriched uranium foil target is irradiated in the reactor core has been carried out. The analysis is carried out by the best estimate method by using a coupled neutronic, kinetic, and thermal-hydraulic code, MTR-DYN. The MTR-DYN code is based on the 3-D multigroup neutron diffusion method. The cell calculations for the target are carried out by the WIMSD/5 and MTR-DYN code. After reactivity insertion, the coolant, fuel, and clad temperature are observed. The calculation results for the initial power of 1 W showed that the maximum temperature of the coolant, clad, and fuel are 49.76?C, 65.01?C, and 65.26?C, respectively. Meanwhile, when the reactivity insertion at the initial power of 1 MW, the maximum temperature of the coolant, clad, and fuel are 72.23?C, 140.79?C, and 141.97?C, respectively. Based on those calculation results during irradiation low enriched uranium foil target, the temperature in the steady-state and reactivity insertion accident does not exceed the allowable safety limit.