Kajian Performa Passive Residual Heat Removal System (PRHRS) pada System-Integrated Advanced Modular Reactor (SMART)

SMART (System-Integrated Advanced Modular Reactor) merupakan desain reaktor multifungsi Generasi III+ tipe SMR (Small Modular Reactor) yang dikembangkan oleh KAERI (Korean Atomic Energy Research Institute) dengan kapabilitas produksi listrik 107 MWe dan energi termal 365 MWt. Sistem SMART meliputi berbagai fitur keselamatan untuk mengatasi LOCA (Loss of Coolant Accident) dan skenario kecelakaan lainnya. Salah satu dari fitur tersebut adalah Passive Residual Heat Removal System (PRHRS) atau sistem pembuang sisa panas pasif yang bekerja tanpa membutuhkan sumber daya elektrik. Sistem ini bekerja sesuai dengan prinsip sirkulasi alam sehingga bergantung pada aspek termal, tekanan, dan pengaruhnya terhadap aliran massa. Ketiga aspek tersebut dapat mempengaruhi kapabilitas pembuangan panas pada sistem. Data performa PRHRS reaktor SMART pada beberapa kondisi kecelakaan yang diperoleh melalui studi eksperimental maupun simulasi termohidrolika dianalisis pada kajian ini. Hasil analisis menunjukkan unjuk kerja pembuangan sisa panas yang baik oleh PRHRS SMART dengan waktu aktuasi yang tepat dan pendinginan yang stabil. Dengan kapabilitas multifungsi dan kemampuan pendinginan yang baik pada berbagai skenario kecelakaan, SMART memiliki potensi tinggi untuk kelak diterapkan di Indonesia.

Effect of High-Current Pulsed Electron Beam Processing of Zr-1%Nb Alloy on Its Oxidation Kinetics at 1200C in Air and Steam

The paper reports the effect of high-current pulsed electron beam (HCPEB) processing of the Zr-1%Nb alloy, as one of the most widely used in water-cooled nuclear reactors, on the kinetics of its oxidation at 1200 °C in air and steam (these conditions are typical for potential loss-of-coolant accidents). It was shown that HCPEB processing caused a change in the surface morphology of the samples. In particular, craters with diameters of about 100 μm were found on the modified surfaces. They had initiated at an energy density of 5 J/cm2 and were characterized by relevant reliefs with microcracks. After HCPEB processing at 10 J/cm2, the craters were deeper with fractured surface layers. In addition, a pronounced surface relief corresponding to quenched martensitic microstructures was observed on the modified sample surfaces that had formed due to high heating and cooling rates. Due to sufficient degradation of the sample surfaces after HCPEB processing at 10 J/cm2, the kinetics of high-temperature oxidation was estimated only for the as-received samples and ones treated at 5 J/cm2. It was found that the as-received samples showed slightly greater weight gain levels in both air and steam environments, which fully correlated with the thickness ratio of the oxide, α-Zr(O) and prior-β layers. These phenomena and further research directions were discussed.

Simulation of VVER-1000 Guillotine Large Break Loss of Coolant Accident Using RELAP5/SCDAPSIM/MOD3.5

The safety performance of nuclear power plants (NPPs) is a very important factor in evaluating nuclear energy sustainability. Safety analysis of passive and active safety systems have a positive influence on reactor transient mitigation. One of the common transients is primary coolant leg rupture. This study focused on guillotine large break loss of coolant (LB-LOCA) in one of the reactor vessels, in which cold leg rupture occurred, after establishment of a steady-state condition for the VVER-1000. The reactor responses and performance of emergence core cooling systems (ECCSs) were investigated. The main safety margin considered during this simulation was to check the maximum value of the clad surface temperature, and it was then compared with the design licensing limit of 1474 K. The calculations of event progression used the engineering-level RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program, which also provide a more detailed treatment of coolant system thermal hydraulics and core behavior. The obtained results show that actuation of ECCSs at their actuation set points provided core cooling by injecting water into the reactor pressure vessel, as expected. The peak cladding temperature did not overpass the licensing limit during this LB-LOCA transient. The primary pressure above the core decreased rapidly from 15.7 MPa to 1 MPa in less than 10 s, then stabilizes up to the end of transient. The fuel temperature decreased from 847 K to 378 K during the first 30 s of the transient time. The coolant leakage reduced from 9945 kg/s to approximately 461 kg/s during the first 190 s in the transient. Overall, the study shows that, within the design of the VVER-1000, safety systems of the have inherent robustness of containing guillotine LB-LOCA.

Analysis of different factors on the features of thermal destruction of a fuel pin

The presented work studies the influence of various factors that affect the specific features of fuel pins melting. For this purpose, fuel pins with different geometries and energy release are considered. Numerical simulation of melting is carried out using a program module for calculating the destruction of fuel rods. Comparison with theoretical calculations is made. The analysis of the convergence of calculations with respect to the time step value and the number of calculated cells along the radius and height is carried out. As a result of work with the use of numerical methods, the characteristic times of destruction of fuel elements during an accident with a loss of coolant flow rate (an accident of the ULOF type) and the dependence of weight loss on time are obtained under various conditions.

Stress Analysis of HCPB BB under Ex-Vessel LOCA Condition

Helium Cooled Pebble Bed Breeding Blanket (HCPB BB) is a kind of concept for the European demonstration fusion reactor (DEMO). The blanket attachment system plays an important role in the mechanical connection of the BB and vacuum vessel. Typically, the mechanical and thermal loads should meet the requirement to avoid collapse of the system with off-normal conditions, e.g., under ex-vessel Loss of Coolant Accident (LOCA. This paper investigates the loading requirement corresponding to the maximum stress that can sustain to avoid the LOCA condition. Firstly, a model of the BB is constructed using SolidWorks. Then, stress analysis is carried out based on the cross section of the blanket. Through simulation, the critical condition for the LOCA case and the maximum stress value for the model are obtained. According to the relevant size dimension from the reference, the blanket’s cross section is drawn, and one can get the stress field under the ex-vessel LOCA through stress analysis. The stress distribution under the ex-vessel LOCA condition is simulated to find out the maximum stress field that the blanket can sustain through this paper. The significance is to predict the possible conditions leading to an accident and find possible methods to avoid them.