Molybdenum-99 (99Mo) is a parent radioisotope of Technetium-99m (99mTc) widely used in nuclear diagnostics. The production of this radioisotope by PT. INUKI generated radioactive fission waste (RFW) that theoretically contains239Pu and235U, posing a nuclear proliferation risk. This paper discusses the determination of radionuclides inventory in the RFW and the proposed strategy for its management. The radionuclides inventory in the RFW was calculated using ORIGEN 2.1 code. The input parameters were obtained from one batch of 99Mo production using high enriched uranium in PT. INUKI. The result showed that the RFW contained activation products, actinides, and fission products, including239Pu and235U. This result was then used for consideration of the management of the RFW. The concentration of 235U was reduced by a down-blending method. The proposed strategy to further manage the down-blended RFW was converting it to U3O8 solid form, placed in a canister, and eventually stored in the interim storage for high-level waste located in The Radioactive Waste Technology Center.

In the safety assessment of radioactive waste disposal, it is critical to understand the porewater chemistry in compacted bentonite in order to predict long-term migration behavior of radionuclides in the engineered barrier. This study estimates the activity coefficients of dissolved ions in the porewater of compacted bentonite from the concentrations of ions at which CaCO3 precipitation occurred. Solutions containing CaCl2 and NaHCO3 were introduced under electrical potential gradient from the opposite sides of the compacted Na-bentonite packed at the dry density of 1.0 kg/dm 3. After the electromigration, the spatial distribution of ions along the compacted bentonite sample was determined. Sequential extraction method was developed to distinctly determine the concentrations of free ions in the porewater and in solid phase in bentonite. The results show that the exchangeable Na + ions were progressively replaced by the incoming Ca 2+ ions, and the compacted bentonite sample can be divided into three zones: Ca-, Ca-/Na-, and Na-bentonite zones. Precipitates of CaCO3 were observed both in Ca-and Ca/Na-bentonite zones. The experimentally determined activity coefficients were at least smaller by a factor of three compared to the theoretical approximation calculated using PHREEQC code assuming dilute-solution conditions with no electrostatic interactions between ions and bentonite surface.

EVALUATION OF NEUTRON SHIELDING PERFORMANCE OF CD-SS 316L AS A CANDIDATE ALLOY FOR DRY CASK OF RESEARCH REACTOR SPENT FUEL Development of dry casks is necessary to support the national strategy for management of spent fuels. One of the requirements for the dry cask is shielding performance for neutron emitted by the spent fuels to be stored in the dry cask. The objectives of this study are to determine the emitted neutrons by the spent fuel generated from GAS research reactor and to evaluate the neutron shielding performance of Cd-SS316L alloy as a candidate material to be used in dry cask for the spent fuels.  The former was carried out using Origen 2.1 software, while the latter using MCNP5. The result shows that the emitted neutrons by a spent fuel after 5 years discharged from GAS research reactor were 2.81×103 and 3.32×106 n/s for reactor core power of 15 and 30 MW, respectively. Addition of Cd improves the neutron shielding performance of SS 316L. The evaluation of neutron shielding performance of SS 316L with addition of Cd which is the candidate material for dry cask of the spent fuels from the GAS research reactor can be evaluated using Origen 2.1 software for neutron emission, while the neutron shielding performance was evaluated by the simulation using MNCP 5 software. This study shows the Cd-SS 316L alloy can be used for further study to develop the dry cask design for the GAS research reactor.Key words: Neutron shielding, cadmium, stainless steel, spent fuel.

A key issue contributing to the success of NPP technology is the safe handling of radioactive waste, particularly spent nuclear fuel. According to the IAEA safety standard, the spent fuel must be stored in interim wet storage for several years so the radiation and the decay heat of the spent fuel will decrease to the safe limit values, after which the spent fuel can be moved to dry storage. In this study, we performed a theoretical analysis of heat removal by natural convection airflow in spent nuclear fuel dry storage. The temperature difference between the air inside and outside dry storage produces an air density difference. The air density difference causes a pressure difference, which then generates natural airflow. The result of the theoretical analysis was validated with simulation software and experimental investigation using a reduced-scale dry storage prototype. The dry storage prototype consisted of a dry cask body and two canisters stacked to store materials testing reactor (MTR) spent fuel, which generates decay heat. The cask body had four air inlet vents on the bottom and four air outlet vents at the top. To simulate the decay heat from the spent fuel in the two canisters, the canisters were wrapped with an electric wire heater that was connected to a voltage regulator to adjust the heat power. The theoretical analysis results of this study are relatively consistent with the experimental results, with the mean relative deviation (MRD) values for the prediction of air velocity, the heat rate using natural airflow, and the heat rate using the thermal resistance network equation are +0.76, −23.69, and −29.54%, respectively.

Reaktor Daya Non-Komersial (RDNK) with a 10 MW thermal power has been proposed as one of the technology options for the first nuclear power plant program in Indonesia. The reactor is a High Temperature Gas-Cooled Reactor-type with spherical fuel elements called pebbles. To support this program, it is necessary to prepare dry cask to safely store the spent pebble fuels that will be generated by the RDNK. The dry cask design has been proposed based on the Castor THTR/AVR but modified with air gaps to facilitate decay heat removal. The objective of this study is to evaluate criticality safety through keff  value of the proposed dry cask design for the RDNK spent fuel. The keff  values were calculated using MCNP5 program for the dry cask with 25, 50, 75, and 100% of canister capacity. The values were calculated for dry casks with and without air gaps in normal, submerged, tumbled, and both tumbled and submerged conditions. The results of calculated keff  values for the dry cask with air gaps at 100% of canister capacity from the former to the latter conditions were 0.127, 0.539, 0.123, and 0.539, respectively. These keff values were smaller than the criticality threshold value of 0.95. Therefore, it can be concluded that the dry cask with air gaps design comply the criticality safety criteria in the aforementioned conditions.

Near-surface disposal (NSD) has been applied in several countries to dispose of low-level radioactive waste. The demo plant of this disposal type is planned to be constructed in Serpong Nuclear Area, Banten. An assessment of radiation exposure is necessary to ensure the safety requirement of the facility in order to support this program. This study aims to estimate radionuclide migration from the proposed NSD demo facility to the environment and the corresponding total human dose using AMBER mathematical modeling. The representative radionuclide,137Cs, was selected because of its high mobility in the environment and the relatively long half-life in the low-level waste inventory. The scenario considered in the modeling was the normal release to the environment through groundwater. Parameters such as initial radionuclide concentration, soil physical parameters of the study site, and disposal design were entered into AMBER software to be calculated using mathematical formulas. The results show that the radionuclide concentration value in the environment is below the safe limit recommended by the Environmental Supervisory Agency. Likewise, the maximum dose received by the community around the facility is 7.40×10-11 mSv/y, 550 years after the post-closure of the facility, which is also below the regulatory limit of 1 mSv/y for the public.

In the past, 99mTc produced by PT. Industri Nuklir Indonesia (PT. INUKI) was carried out through the processing of a 99Mo parent radioisotope that was produced using high enriched uranium target. Due to restrictions on the use of high-enriched uranium (HEU) by the International Atomic Energy Agency, PT. INUKI has converted the process using a low enriched uranium (LEU) target. This study aims to determine the change of radioactive fission waste characteristics in 99Mo production due to the conversion process. The characteristics of radioactive fission waste generated from low-enriched uranium targets were calculated using ORIGEN 2.1 program. The results show that the radioactive fission waste contains radionuclides of activation products, actinides, and fission products. The conversion produced smaller activity of 99Mo compared to that of using the HEU target, which was 397 compared to 1010 Ci. The conversion generated radioactive fission waste with a smaller content of remaining 235U but greater content of actinides, particularly 239Pu. The activity of 239Pu in the radioactive fission waste from LEU was 29.1 μCi, approximately 19 times higher compared to that of using HEU, which was 1.52 μCi. After 50 years of decay, this radioactive fission waste was calculated to have a specific activity of 6.54x108 Bq/g, lower than that of the highly enriched uranium target, which was 3.01x109 Bq/g. This radioactive fission waste requires management with a high safety level.

Neutron shielding material is necessary for the container of radioactive waste emitting neutrons, such as spent fuel from research reactor. This study aimed to determine the shielding performance of Al-Cd material for neutron radiation emitted by radioactive radiation. The calculations of neutron shielding were carried out using MNCP5 software. The Al-Cd shielding was simulated to withstand a neutron radiation source with 2.538×10−8 MeV energy and 4.5 MeV energy moderated with paraffin. The Al-Cd was simulated as a homogeneously-mixed alloy and Al coated with Cd. The results show that addition of Cd to Al increased the neutron shielding performance for thermal and fast neutron. With the addition of 10% Cd to Al, the neutron shielding performance increased by more than 90% for thermal neutron, and 30% for fast energy neutron moderated with 5 cm thick paraffin. It shows that Al-Cd alloy is potential for radioactive waste container which requires high neutron shielding performance.