[en] Fuels containment system is a facility to isolate fouled fuels so that their fission product could not contaminate the reactor pool. This system was designed to anticipate the occurrence of the fuels damage by leakage in consequence of burn up process in the reactor core. The proposed design contains two RSG-GAS fouled fuels bundles and could not distribute the fission product to reactor pool. From the analysis result showed that the proposed design is capable to the working stresses and without heat accumulation in the system during its operation

[en] At normal operation of 15 MW prior to shut down, the capability test of 3 unit emergency pool cooling system have been carried out at different time. The system is used to cool the decay heat after reactor shut down at normal operation as well as emergency condition. The systems are operated independently toward the reactor operation and other emergency system. Three cooler systems of reactor pool emergency have capability above requires capability. This paper describes comparison between capability of intake of heat which remain at the commissioning phase and at examination to know the degradation of performance. Capability of intake of heat at the examination, one line system is different from other the system and have degradation compared to capability at the commissioning. Degradation capability is caused by natural process and not caused by encumbering factor. The equipment seldom operated, so the level of degradation is 36.5 % still become safety limit. The degradation equipment does not affect alertness in operating of reactor. (author)

[en] Study for electric power improvement means that discussion for characteristics of electricity focused at the fluctuation of voltage waveforms caused by influence of lightning attack both directly or indirectly to RSG-GAS nuclear installation. The measuring BHA bus bar voltage waveform by used power line monitor will be done at normal conditions (there's no lightning attacks). The measuring result shows that the voltage waveform is sinusoidal. The measuring devices as describe above will be installs for continuous measuring to get voltage waveform in the case of lightning attack occurs. The influence of lightning attack into power line can be read from that device recording which show that the voltage waveform have changed. In order to recovery the voltage waveform can be done by used the lightning arrester, and over voltage arrester that installed at bus bar BHA-02. That device will be hoped could overcome the damage of waveform to normal condition. (author)

[en] The RSG-GAS reactor is a research reactor operated for radioisotope production, neutron activation analysis, research and material testing. The RSG-GAS reactor operates for 625 MWd each cycle and core fuel management uses a 5/1 pattern in the fuel loading. Every fuels replacement in the reactor core performed calculating and measured reactor parameters to determine the safety factor of reactor operation. Under normal operating conditions the irradiation facility is in a water or aluminium dummy. The addition of aluminum dummy to the irradiation facility needs to be known for its effect on the neutronic parameters of the RSG-GAS reactor. In this paper, we evaluated the calculation and measurement of reactor operating parameters and the dummy effect of aluminium on reactivity and neutron flux of 97 RSG-GAS reactors core. The calculation of operating parameters is done using a combination of WIMS/D5 for the cell generation cross section, Batan-2DIFF to calculate reactivity, radial power peacking factor and Batan-3DIFF to calculate the axial power peacking factor and neutron flux in the reactor core. Based on the results of calculations and measurements indicate that the 97 RSG-GAS reactors core meets the reactor operating safety limit. The reactivity effect due to the placement of the aluminium dummy on the CIP 0.17 % Δk/k. If the aluminium dummy is placed in the CIP and the IP will increase the flux in that position because aluminium has a very good moderating power. The value of power factor for peak radial and axial power is 1.2200 and 1.2902, respectively. The value is still within the safety limit of reactor operations. (author)

[en] Reactor power calibration is the process of adjusting or re-measuring the reactor power measuring tools or channels against standard values. At the time of calibration there is a correction factor which is important in the reactor power calibration that must be measured and analyzed periodically. The current problem is the absence of correction factor measurement of power calibration especially at the operation of reactor at 15 MW and 30 MW. Measurement of correction factor is intended to obtain the shift of correction factor that has been used in the calculation of RSG-GAS reactor power calibration, to know the effect of the change on reactor power calibration, and to know the difference of correction factor value when the secondary coolant is operated with 4 and 7 Unit blower. This paper involves the recording of reactor power calibration, and measuring calibration correction factor at 15 MW and 30 MW power. Correction factor measurements are made under the conditions of the entire system being operated as during high power operation, but the difference is that the reactor is not burdened by power (reactor is not operated). The recording of data is carried out until the temperature difference in and out of the stable terrace with prior initial data recording before the cooling system is operated. From the measurement results obtained a correction factor when the secondary cooling system operated with 4 units of blower (operation 15 MW) is 0.1 °C (previous correction factor is 0.41 °C) and correction factor when the secondary cooling system is operated with 7 units Blower (operation 30 MW) that is 0.12 °C. Changes and differences in correction factors affect the reactor power calibration results. It is expected that the measurement result can be a reference in conducting RSG-GAS reactor power calibration. (author)

[en] The power reactivity coefficient (KRD) is a neutronic parameter which is very important for the safety of reactor operation. The power reactivity coefficient is a combination of the doppler, moderator and void reactivity coefficients. The power reactivity coefficient of RSG-GAS was measured at the time of the first core, the RSG-GAS reactor core had been converted from an oxide core (U₃O₈-AL) to a silicide core (U₃Si₂Al) so it was necessary to re-measure it. The power reactivity coefficient is designed to be negative. The purpose of this study is to calculate the power reactivity coefficient of the RSG-GAS reactor through an experimental power increase. The power reactivity coefficient is determined by increasing the power from 1 - 15 MW gradually with the position of the bank control rod fixed and the control rod changing. The change in reactivity is determined according to the position of the control rod. From the calculation, it is known that the average power reactivity coefficient is -1.028 cents/MW and will be more negative as the power increases. This happens because the increase in reactor power will increase the temperature of the fuel which results in the Doppler effect. In addition, the moderating power of the moderator will decrease due to the increase in reactor coolant temperature. The reactivity coefficient value of the silicide core was lower than that of the oxide core (KRD of the oxide core = 1.97 cents/MW). This is because the heat generation in the oxide core for the uranium charge level is 250 gr higher than the silicide (1.87 W/gr oxide, 1.85 W/gr silicide) so that the doppler effect that occurs is higher. However, when compared with several other research reactors, the KRD RSG-GAS value is relatively higher. With a negative and more negative KRD value following the increase in reactor power, the reactor can be operated stably and safely. (author)

[en] Shredder machine is a shredder or crusher which in its application at the Radioactive Waste Technology Center (PTLR) is used to treat solid radioactive waste of contaminated material which has large dimensions such as jerry cans, High-density polyethylene (HDPE) drums, corrosion drums and others. The final objective of this activity is to provide information related to the shredder machine and to find out the relationship between the working mechanism of the shredder machine operating process, as well as failure analysis in the process of treating solid radioactive waste of contaminated material. The method used includes operating the tool, observing and analyzing the failure of the operation of the tool. The results showed that the production capacity of the shredder machine operation was 1,008 kg / hour for 160 liter HDPE drum solid waste. To keep the machine operating properly, periodic maintenance is required. The failure analysis in the shredder tool operation process is influenced by: the operator, the machine operating system, the method of operation, and the material being processed. (author)

[en] ¹⁷⁷Lu (Lutetium-¹⁷⁷) is a radioisotope in the Lanthanide group which is now widely used as a cancer radiotherapy agent. To produce this radioisotope, it can be done by irradiating the Lu₂O₃ target at the CIP (Central Irradiation Position) of the RSG-GAS Reactor. One important step for the safety of radioisotope production operations is the calculation of the resulting target activity. Therefore, this study aims to determine the optimum time in the calculation of radioactivity for the production of ¹⁷⁷Lu. This calculation uses the Origen 2.1 program package by entering input data such as neutron flux (1 x 10⁴ n/m²s), the mass of Lutetium oxide (3 milligrams), and duration of irradiation (4 days, 8 days, and 12 days). In determining the mass of Lutetium, it is divided into two components, namely the mass of ¹⁷⁶Lu and ¹⁷⁵Lu, which are 2 milligrams and 0.6 milligrams, respectively. This calculation resulted in ¹⁷⁷Lu of radioactivity for 12 days of irradiation of 30.939 GBq, of the radioactivity of ¹⁷⁷Lu of 8 days of irradiation of 25.511 GBq, and of the radioactivity of ¹⁷⁷Lu of 4 days of irradiation of 15.939 GBq. Based on the minimum dose of radioisotope used as a therapeutic agent, which is 20 GBq, the results of this radioisotope production starting with a variation of the irradiation time for 8 days can be used as therapy. (author)

[en] It has been implemented the measurement of compensation factors on the KNK 50 power range neutron detectors installed around the RSG-GAS core. The measurements are needed to check performance of the compensating part of the detectors after they are used for more than a year. While the reactor are in operation, the core produce neutrons and also gamma rays other which will contribute to current signal of the detectors. The compensation factor indicate capability of the detectors to neutralize of effect of gamma radiations affect so as its design, that it only measures neutrons. The measurements were performed on two installed KNK 50 power range detector one day after reactor shutdown. The constant positive voltage is applied to the positive electrodes of the neutron and gamma sensitive chambers. While varied negative voltages were applied to the negative electrodes of compensation chamber. For each variation of negative voltage applied, the resulting current at the signal electrode was measured and noted. Eventually, the minimum resulting currents are measured and noted. The compensation factor of the detector is the percentage of the minimum resulting current at the signal electrode to the resulting current when the negative electrode was grounded. The results of measurements of the detector JKT 03 CX 831 has a compensation factor of 0.45 % and a detector JKT 03 CX 841 has a compensation factor of 0.41 %. Both of these detectors is within the specification limits 2 %^[3*] so that the ability of the gamma compensation detector is functioning properly. (author)

[en] Nuclear safety and security cultures have been established but for the safeguard culture has not been established internationally. International safeguard system consists of the agreement, inspections and evaluations and it never considers safeguard culture of the country or facility. Historically, cultural indicators do not play a role in the IAEA verification activities against safeguards even though since the case of Iraq appears in the early 1990, IAEA considering the safeguard culture through the theory of organizational culture and the development of a safeguard it self. The aim of this paper is to discuss definition and indication of safeguards. The assessment is done by studying various literature which pertinent to culture, safeguards culture and its implementation in Indonesia and Hungary. It is expected that safeguard culture may strengthen and increase safeguards performance effectively that is verification activities based on correctness and completeness information submitted by the member State to the IAEA. Those it can be ensured that all nuclear material under the IAEA is only for peaceful purposes. (author)