Preliminary Study of Probabilistic Safety Assessment in Evaluating Spent Fuel Interim Management of South Korea
Kyung Hee University
In South Korea, the master plan for high-level radioactive waste management, announced in 2016, suggested the construction and operation of intermediate storage facilities on a permanent disposal site and specified the adoption of dry storage in consideration of the ease of operation and expansion. As of 2019, the government is again reviewing its overarching policy on the back-end fuel cycles, including intermediate storage and permanent disposal. In the case of dry storage facilities, safety evaluation is being conducted using a combination of deterministic and probabilistic approaches, similar to general nuclear power plants. The two methods are complementary, of which Probabilistic Safety Assessment (PSA) has the advantage of being able to identify key scenarios affecting safety, but its use has not been highlighted so far. However, depending on the spent fuel management phases such as loading, transportation, and storage, it may be not enough to capture effective and efficient safety evaluation only deterministically, and probabilistic methods may contribute to the evaluation of long-term operation or external events such as earthquake. There have already been cases where PSA has been performed on a part of the nuclear fuel cycle through previous studies. In this study, the preliminary study on such application of PSA was carried out based on open sources. The model was considered into loading, transportation, and storage, respectively.
Measurement methods for radioactive concentrations of waste generated by decommissioning of uranium processing facilities
Japan Atomic Energy Agency
As a result of the decommissioning of the uranium conversion and uranium enrichment facilities at the Ningyo-toge Center, waste mainly containing uranium has been generated. Both solid and liquid wastes are stored in drums and containers, and it is necessary to treat them for disposal or reuse. To reasonably proceed with decommissioning and reuse, it is necessary to accurately estimate the radioactivity of these wastes.
Uranium amount in simulated waste drums was evaluated using the measuring device with a steel shielding box and NaI (Tl) detectors. The drums were measured for 60 minutes while rotating at 10 rpm. As a method of evaluating the amount of uranium in a drum, a method of representing the state of average shielding of gamma rays from the radiation sources in a drum by using the ratio of gamma-rays of two types of energy was examined. The average state was expressed as one coefficient and it has an approximately linear correlation against count rates of the measurement target. When this relation was used to quantify uranium, the relative errors from the true uranium concentrations became less than 40%. The proposed method was applied to the evaluation of radium concentration in a container. The system for measuring the radioactivity of a container is equipped with four NaI detectors. By rotating the container 90 degrees, four sides can be measured to assess the radioactivity of the entire container. As a result of the examination, it was found that even if there are uneven source arrangements, radium can be quantified within 30 % relative error.
New Neutron Absorber in Spent Fuel Casks Aiming at Improved Nuclear Safety and Better Economics
1University of West Bohemia; 2Czech Technical University
The recent increasing demand for better nuclear fuel utilization requires higher enriched uranium fuels, which is a challenge for spent fuel handling facilities in all countries with nuclear power plants. The operation with higher enriched fuels leads to reduced reserves to legislative and safety margins of spent fuel transport and storage facilities.
This study shows the solution with significantly increased nuclear safety and improved economics where a new concept of inseparable fixed neutron absorber is introduced to achieve spent fuel reactivity decrease.
It is possible to reduce the boron content in the cask basket, reduce fuel assembly pitch and reduce the neutron dose in the vicinity of the cask. The efficiency of this concept is demonstrated on the criticality safety analysis of the GBC-32 spent fuel cask.
Pitting and Intergranular Corrosion Behavior of Austenite Stainless Steels for Understanding of CISCC in Dry Storage Canister Materials
Korea Institute of Materials Science
Worldwide demand on interim storage of spent nuclear fuels arises due to the limited capacity of in-reactor-site storage. Among many kinds of storage types, dry storage has attracted assorted interest from engineering and industrial fields due to its technical benefits. Dry storage canister is generally made of austenite stainless steels because of its excellence in mechanical properties, corrosion resistance, weldability and economic benefits. However, as the dry storage canisters would have been considered to be installed in seashore, chloride-induced stress corrosion cracking (CISCC) is considered as potential threat to hinder materials reliability. CISCC generally occurs by combining residual stress originated from welding, chloride induced pitting and intergranular corrosion. Since pitting and/or attacked grain boundary would play a role as crack initiation site, it is essential to understand the corrosion behavior of the materials with considering of residual stress. In this study, we adopt multiple electrochemical techniques, such as electrochemical impedance spectroscopy (EIS), potentiodynamic electro-polarization, and double loop electrochemical potentiodynamic repassivation (DL-EPR) to evaluate corrosion behavior of the materials (i.e., 304L, 316L and 316LN) in chloride-containing solution. The results are quantified and statistically analyzed in terms of corrosion rate, pitting potential and degree of sensitization. Also, the results are compared with microstructure and chemical analysis to crystallize correlation between chemical composition of the alloys, environmental factors and corrosion behavior. In addition, further works to evaluate CISCC initiation and monitoring of crack growth will be introduced.
Spent Fuel Cooling and Analysis of Natural Convection Heat Transfer From a Vertical Cylinder Submerged in a Water Pool
Military Institute of Science and Technology
Natural convection flows around submerged heated cylinders have gained attention in recent years due to their use in many engineering applications such as flow around tubes and rods (as in nuclear reactors and spent fuel cooling ponds). The purpose of this paper is to establish the modelling strategy for simulating natural convection heat transfer and flow around vertically positioned cylinders submerged in a water tank heated with a constant heat flux.Spent Nuclear Fuel (SNF), nuclear waste, needs to be submerged under water in cooling ponds for several years until became less radioactive. SNF cooling ponds, in principle, based on a phenomenon to get rid of the decay heat released from the Fuel racks. And the Phenomenon is natural convection heat transfer between the fuel racks and the surrounding water. In this thesis study, for understanding the importance of SNF cooling, there had been considered one fuel rod first and then compared with three fuel rods to know the heat transfer rate with varying temperature. The height and diameter considered in simulation as 3.73m and 1cm respectively. The height, length and width of the pool was 4.5m*1m*1m. For example, Using the Comsol Multiphysics Simulator, at temperature, T=343 Kelvin, total net heat rate for 1 fuel rod was q=100.93Watt and for 3 fuel rod was q=166.32watt. These q values were further increased with increasing the temperature for both 1 versus 3 fuel rods respectively. Numerical simulation was conducted and the obtained data were validated against the experimental data .An experimental facility has been designed for studying the heat transfer through natural convection and comparing the experimental results with Comsol Multiphysics Simulator. In experimental facility, the dimension of the water box considered as a pool was 40cm*25cm*25cm and the dimension of heater considered as fuel rod was 30cm*1cm.