• Nuclear Engineering and Technology
• /
• v.51 no.2
• /
• pp.384-393
• /
• 2019

An experimental validation of a nuclear forensics methodology for the source reactor-type discrimination of separated weapons-useable plutonium is presented. The methodology uses measured values of intra-element isotope ratios of plutonium and fission product contaminants. MCNP radiation transport codes were used for various reactor core modeling and fuel burnup simulations. A reactor-dependent library of intra-element isotope ratio values as a function of burnup and time since irradiation was created from the simulation results. The experimental validation of the methodology was achieved by performing two low-burnup experimental irradiations, resulting in distinct fuel samples containing sub-milligram quantities of weapons-useable plutonium. The irradiated samples were subjected to gamma and mass spectrometry to measure several intra-element isotope ratios. For each reactor in the library, a maximum likelihood calculation was utilized to compare the measured and simulated intra-element isotope ratio values, producing a likelihood value which is proportional to the probability of observing the measured ratio values, given a particular reactor in the library. The measured intra-element isotope ratio values of both irradiated samples and its comparison with the simulation predictions using maximum likelihood analyses are presented. The analyses validate the nuclear forensics methodology developed.

• Nuclear Engineering and Technology
• /
• v.56 no.3
• /
• pp.1116-1123
• /
• 2024

-In recent years, unmanned underwater vehicles (UUV) have been vigorously developed, and with the continuous deepening of marine exploration, traditional energy can no longer meet the energy supply. Nuclear energy can achieve a huge and sustainable energy supply. The heat pipe reactor has no flow system and related auxiliary systems, and the supporting mechanical moving parts are greatly reduced, the noise is relatively small, and the system is simpler and more reliable. It is more favorable for the control of unmanned systems. The use of heat pipe reactors in unmanned underwater vehicles can meet the needs for highly compact, long-life, unmanned, highly reliable, ultra-quiet power supplies. In this paper, a heat pipe reactor scheme named UPR-S that can be applied to unmanned underwater vehicles is designed. The reactor core can provide 1 MW of thermal power, and it can operate at full power for 5 years. UPR-S has negative reactive feedback, it has inherent safety. The temperature and stress of the reactor are within the limits of the material, and the core safety can still be guaranteed when the two heat pipes are failed.

• Nuclear Engineering and Technology
• /
• v.48 no.3
• /
• pp.702-710
• /
• 2016

Global concern and interest in the safety of nuclear power plants have increased considerably since the Fukushima accident. In the event of a severe accident, the reactor vessel water level cannot be measured. The reactor vessel water level has a direct impact on confirming the safety of reactor core cooling. However, in the event of a severe accident, it may be possible to estimate the reactor vessel water level by employing other information. The cascaded fuzzy neural network (CFNN) model can be used to estimate the reactor vessel water level through the process of repeatedly adding fuzzy neural networks. The developed CFNN model was found to be sufficiently accurate for estimating the reactor vessel water level when the sensor performance had deteriorated. Therefore, the developed CFNN model can help provide effective information to operators in the event of a severe accident.

• Nuclear Engineering and Technology
• /
• v.51 no.6
• /
• pp.1658-1668
• /
• 2019

Nuclear power generates a large portion of the energy used today and plays an important role in energy development. To ensure safe nuclear power generation, it is essential to conduct an accurate analysis of reactor structural integrity. Accordingly, in this study, a methodology for obtaining accurate structural responses to the combined seismic and reactor coolant loads existing prior to the shutdown of a nuclear reactor is proposed. By applying the proposed analysis method to the reactor vessel internals, it is possible to derive the seismic responses considering the influence of the hydraulic loads present during operation for the first time. The validity of the proposed methodology is confirmed in this research by using the finite element method to conduct seismic and hydraulic load analyses of the advanced APR1400 1400 MWe power reactor, one of the commercial reactors. The structural responses to the combined applied loads are obtained using displacement-based and stress-based superposition methods. The safety of the subject nuclear reactor is then confirmed by analyzing the design margin according to the American Society for Mechanical Engineers (ASME) evaluation criteria, demonstrating the promise of the proposed analysis method.

• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4373-4391
• /
• 2022

Marine reactor systems experience platform movement, and therefore, the system thermal-hydraulic analysis code needs to reflect the motion effect on the fluid to evaluate reactor safety. A moving reactor model for MARS-KS was developed to simulate the hydrodynamic phenomena in the reactor under motion conditions; however, its applicability does not cover the MULTID component used in multidimensional flow analyses. In this study, a moving reactor model is implemented for the MULTID component to address the importance of multidimensional flow effects under dynamic motion. The concept of the volume connection is generalized to facilitate the handling of the junction of MULTID. Further, the accuracy in calculating the pressure head between volumes is enhanced to precisely evaluate the additional body force. Finally, the Coriolis force is modeled in the momentum equations in an acceleration form. The improvements are verified with conceptual problems; the modified model shows good agreement with the analytical solutions and the computational fluid dynamic (CFD) simulation results. Moreover, a simplified gravity-driven injection is simulated, and the model is validated against a ship flooding experiment. Throughout the verifications and validations, the model showed that the modification was well implemented to determine the capability of multidimensional flow analysis under ocean conditions.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1901-1910
• /
• 2023

The safety-class (1E) digital control system (DCS) of nuclear power plant characterized structural multiple redundancies, therefore, it is important to quantitatively evaluate the reliability of DCS in different degree of backup loss. In this paper, a reliability evaluation model based on T-S fuzzy fault tree (FT) is proposed for 1E DCS of nuclear power plant, in which the connection relationship between components is described by T-S fuzzy gates. Specifically, an output rejection control system is chosen as an example, based on the T-S fuzzy FT model, the key indicators such as probabilistic importance are calculated, and for a further discussion, the T-S fuzzy FT model is transformed into Bayesian Network(BN) equivalently, and the fault diagnosis based on probabilistic analysis is accomplished. Combined with the analysis of actual objects, the effectiveness of proposed method is proved.

• 전기의세계
• /
• v.14 no.5
• /
• pp.1-7
• /
• 1965

Using "Yonsei$^{101}$ Analog Computer" the poisoning of the Xenon$^{135}$ in a thermal homogeneous nuclear reactor is analyzed. The simulator is constituted of high gain D.C. operational amplifiers and operational impedances. During the nuclear reactor operation, the Xenon poisoning increases against time until the equilibrium state reaches. After the reactor shut-down, it increases remarkably until the maximum value and then decreases. The simulated curves agree with theoretical values satisfactorily. The accuracy of the analog computer solution is 0.4387 per cent during the nuclear reactor operation and 6.7 per cent after the nuclear reactor shut-down respectively.pectively.

• Nuclear Engineering and Technology
• /
• v.30 no.2
• /
• pp.164-172
• /
• 1998

This paper presents the detailed modelling of reactor vessel ledge region for the dynamic analysis of the coupled internals and core model. The dynamic responses due to earthquake and pipe break are calculated using the input motions of reactor vessel taken from Ulchin nuclear power plant units 3 and 4. Two different representations for detailed and simplified models of the RV ledge region are made. The dynamic responses of the reactor internals components are compared between them. Response characteristics are reported and simplified model is suggested for earthquake and pipe break analysis for the future design of the reactor internals.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.10a
• /
• pp.153-153
• /
• 1998

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.05a
• /
• pp.149-149
• /
• 1999