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.