• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1338-1348
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• 2016

A novel fully passive small modular superheated water reactor (SWR) for underwater deployment is designed to produce 160 MWe with steam at $500^{\circ}C$ to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF). The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.138-151
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• 2022

Cladding ballooning and rupture are the important phenomena at the early stage of a severe accident. Most severe accident analysis codes determine the cladding rupture based on simple parameter models. In this paper, a FRTMB module was developed using the thermal-mechanical model to analyze the fuel mechanical behavior. The purpose is to judge the cladding rupture with the severe accident analysis code. The FRTMB module was integrated into the self-developed severe accident analysis code ISAA to simulate the PHEBUS FPT3 experiment. The predicted rupture time and temperature of the cladding were basically consistent with the measured values, which verified the correctness and effectiveness of the FRTMB module. The results showed that the rising of gas pressure in the fuel rod and high temperature led to cladding ballooning. Consequently, the cladding hoop strain exceeded the strain limit, and the cladding burst. The developed FRTMB module can be applied not only to rod-type fuel, but also to plate-type fuel and other types of reactor fuel rods. Moreover, the FRTMB module can improve the channel blockage model of ISAA code and make contributions to analyzing the effect of clad ballooning on transient and subsequent parts of core degradation.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.14 no.3
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• pp.403-414
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• 1994

Nonlinear analysis of RC containment structure under thermal load and pressure is presented to trace the behaviour after an assumed LOCA. The temperature distribution varying with time through the wall thickness is determined by transient finite element analysis with the two time level scheme in time domain. The layered shell finite elements are used to represent the containment structures in nuclear power plants. Both geometric and material nonlinearities are taken into account in the finite element formulation. The constitutive relation of concrete is modeled according to Drucker-Prager yield criteria in compression. Tension stiffening model is used to represent the tensile behaviour of concrete including bond effect. The reinforcing bars are modeled by smeared layer at the location of reinforcements accounting elasto-plastic axial behaviors. The steel liner model under Von Mises yield criteria is adopted to represent elastic-perfect plastic behaviour. Geometric nonlinearity is formulated to consider the large displacement effect. Thermal stress components are determined by the initial strain concept during each time step. The temperature differential between any two consecutive time steps is considered as a load incremental. The numerical results from this study reveal that nonlinear temperature gradient based on transient thermal analysis will produces excessive large displacement. Nonlinear behavior of containment structures up to ultimate stage can be traced reallistically. The present study allows more realistic analysis of concrete containment structures in nuclear power plants.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.937-946
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• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

• Tunnel and Underground Space
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• v.8 no.3
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• pp.184-193
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• 1998

In order to dispose radioactive wastes safely, it is needed to understand the mechanical, thermal, fluid behavior of rockmass and physico-chemical interactions between rockmass and water. Also, the knowledge about mechanical and hydraulic properties of rocks is required to predict and to model many conditions of geological structure, underground in-situ stress, folding, hot water interaction, intrusion of magma, plate tectonics etc. This study is based on researches about rock mechanics issues associated with a waste disposal in deep rockmass. This paper includes the mechanical and hydraulic behavior of rocks in varying temperature conditions, thermo-hydro-mechanical coupling analysis in rock mass and deformation behavior of discontinuous rocks. The mechanical properties were measured with Interaken rock mechanics testing systems and hydraulic properties were measured with transient pulse permeability measuring systems. In all results, rock properties were sensitive to temperature variation.

• Structural Engineering and Mechanics
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• v.50 no.6
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• pp.755-772
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• 2014

A simplistic approach towards evaluation of complete load deflection response of Reinforced Concrete (RC) flexural members under post fire (residual) scenario is presented in this paper. The cross-section of the RC flexural member is divided into a number of sectors. Thermal analysis is performed to determine the temperature distribution across the section, for given fire duration. Temperature-dependent stress-strain curves for concrete and steel are then utilized to perform a moment-curvature analysis. The moment-curvature relationships are obtained for beams exposed to different fire durations. These are then utilized to obtain the load-deflection plots following pushover analysis. Moreover one of the important issues of modeling the initial stiffness giving due consideration to stiffness degradation due to material degradation and thermal cracking has also been addressed in a rational manner. The approach is straightforward and can be easily programmed in spreadsheets. The presented approach has been validated against the experiments, available in literature, on RC beam subjected to different fire durations viz. 1hr, 1.5hrs and 2hrs. Complete load-deflection curves have been obtained and compared with experimentally reported counterparts. The results also show a good match with the results obtained using more complicated approaches such as those involving Finite element (FE) modeling and conducting a transient thermal stress analysis. Further evaluation of the beams during fire (at elevated temperatures) was performed and a comparison of the mechanical behavior of RC beams under post fire and during fire scenarios is made. Detailed formulations, assumptions and step by step approach are reported in the paper. Due to the simplicity and ease of implementation, this approach can be used for evaluation of global performance of fire affected structures.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.28 no.12
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• pp.1557-1566
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• 2004

A flame extinction phenomenon is a typical unsteady process in combustion. Flame extinction is characterized by various physical phenomena, such as convection, diffusion, and the production of heat and mass. Flame extinction can be achieved by either increasing the strain rate or curvature, by diluting an inert gas or inhibitor, or by increasing the thermal or radiant energy loss. Though the extinction is an inherently transient process, steady and quasi-steady approaches have been used as useful tools for understanding the flame extinction phenomenon. Recently, unsteady characteristics of flames have been studied by many researchers, and various attempts have been made to understand unsteady flame behavior, by using various extinction processes. Representative parameters for describing flame, such as flame temperature, important species related to reactions, and chemi-luminescence of the flame have been used as criterions of flame extinction. In these works, verification of each parameter and establishing the proper criterions of the extinction has been very important. In this study, a time-dependent flame temperature and an OH radical concentration were measured using optical methods, and the instantaneous change of the flame luminosity was also measured using a high-speed ICCD (HICCD) camera. We compare the unsteady extinction points obtained by three different methods, and we discuss transient characteristics of maximum flame temperature and OH radical distribution near the extinction limit.

• Journal of the Korean Society of Safety
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• v.36 no.3
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• pp.74-80
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• 2021

Korea Atomic Energy Research Institute (KAERI) has been operating an integral effects test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), according to APR1400 for transient experimental and design basis accident simulation. Moreover, based on the experimental data, the domestic standard problem (DSP) program has been conducted in Korea to validate system codes. Recently, through DSP-05, the performance of the passive auxiliary feedwater system (PAFS) in the event of multiple steam generator tube rupture (MSGTR) has been analyzed. However, some errors exist in the reference input model distributed for DSP-05. Furthermore, the calculation results of the heat loss correlation for the secondary system presented in the technical report of the reference indicate that a large difference is present in heat loss from the target value. Thus, in this study, the reference model is corrected using the geometric information from the design report and drawings of ATLAS. Additionally, a new heat loss correlation is suggested by fitting the results of the heat loss tests. Herein, MSGTR-PAFS accident analysis is performed using MARS-KS 1.5 with the improved model. The steady-state calculation results do not significantly differ from the experimental values, and the overall physical behavior of the transient state is properly predicted. Particularly, the predicted operating time of PAFS is similar to the experimental results obtained by the modified model. Furthermore, the operating time of PAFS varies according to the heat loss of the secondary system, and the sensitivity analysis results for the heat loss of the secondary system are presented.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1066-1080
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• 2024

SMART100 has a containment pressure and radioactivity suppression system (CPRSS) for passive containment cooling system (PCCS). This prevents overheating and over-pressurization of a containment through direct contact condensation in an in-containment refueling water storage tank (IRWST) and wall condensation in a CPRSS heat exchanger (CHX) in an emergency cool-down tank (ECT). The Korea Atomic Energy Research Institute (KAERI) constructed scaled-down test facilities, SISTA1 and SISTA2, for the thermal-hydraulic validation of the SMART100 CPRSS. Three separate effect tests were performed using SISTA1 to confirm the heat removal characteristics of SMART100 CPRSS. When the low mass flux steam with or without non-condensable gas is released into an IRWST, the conditions for mitigation of the chugging phenomenon were identified, and the physical variables were quantified by the 3D reconstruction method. The local behavior of the non-condensable gas was measured after condensation inside heat exchanger using a traverse system. Stratification of non-condensable gas occurred in large tank of the natural circulation loop. SISTA2 was used to simulate a small break loss-of-coolant accident (SBLCOA) transient. Since the test apparatus was a metal tank, compensations of initial heat transfer to the material and effect of heat loss during long-term operation were important for simulating cooling performance of SMART100 CPRSS. The pressure of SMART100 CPRSS was maintained below the design limit for 3 days even under sufficiently conservative conditions of an SBLOCA transient.

• Nuclear Engineering and Technology
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• v.19 no.1
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• pp.22-33
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• 1987

The confirmed integrity of nuclear fuel cladding materials is an important object during steady state and transient operations at nuclear power plant. In this context, the clad material yielding behavior is especially important because of pellet-clad gap expansion. During the steep power excursion, the in-pile irradiation behavior differences between uranium-dioxide fuel pellet and zircaloy clad induce the contact pressure between them. If this pressure reaches the zircaloy clad yield pressure, the zircaloy clad will be plastically deformed. After the reactor power resumed to normal state, this plastic permanent expansion of clad tube give rise to the pellet-clad gap expansion. In this paper, the simple mandrel expansion test method which utilizes thermal expansion difference between copper mandrel and zircaloy tube was adopted to simulate this phenomenon. That is, copper mandrel which has approximately three times of thermal expansion coefficient of zircaloy-4 (PWR fuel cladding material) were used in this experiment at the temperature range from 400C to 700C. The measured plastic expansion of zircaloy outer radius and derived mathematical relations give the yield pressure, yield stress of zircaloy-4 clad at the various clad wall temperatures, the activation energy of zircaloy tube yielding, and pellet-clad gap expansion. The obtained results are in good agreement with previous experimental results. The mathematical analysis and simple test method prove to be a reliable and simple technique to assess the yielding behavior and gap expansion measurement between zircaloy-4 tube and uranium-dioxide fuel pellet under biaxial stress conditions.