• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.673-682
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• 2018

One of the most widely used methods to estimate core damage during a nuclear power plant accident is containment radiation measurement. The evolution of severe accidents is extremely complex, leading to uncertainty in the containment dose rate (CDR). Therefore, it is difficult to accurately determine core damage. This study proposes to conduct uncertainty analysis of CDR for core damage assessment. First, based on source term estimation, the Monte Carlo (MC) and point-kernel integration methods were used to estimate the probability density function of the CDR under different extents of core damage in accident scenarios with late containment failure. Second, the results were verified by comparing the results of both methods. The point-kernel integration method results were more dispersed than the MC results, and the MC method was used for both quantitative and qualitative analyses. Quantitative analysis indicated a linear relationship, rather than the expected proportional relationship, between the CDR and core damage fraction. The CDR distribution obeyed a logarithmic normal distribution in accidents with a small break in containment, but not in accidents with a large break in containment. A possible application of our analysis is a real-time core damage estimation program based on the CDR.

• Steel and Composite Structures
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• v.17 no.2
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• pp.159-172
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• 2014

A finite element model with the consideration of damage initiation and evolution has been developed for the analysis of the dynamic response of a composite sandwich panel subject to low velocity impact. Typical damage modes including fiber breakage, matrix crushing and cracking, delamination and core crushing are considered in this model. Strain-based Hashin failure criteria with stiffness degradation mechanism are used in predicting the initiation and evolution of intra-laminar damage modes by self-developed VUMAT subroutine. Zero-thickness cohesive elements are adopted along the interface regions between the facesheets and the foam core to simulate the initiation and propagation of delamination. A crushable foam core model with volumetric hardening rule is used to simulate the mechanical behavior of foam core material at the plastic state. The time history curves of contact force and the core collapse area are obtained. They all show a good correlation with the experimental data.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.311-322
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• 2013

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.

• Journal of the Earthquake Engineering Society of Korea
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• v.19 no.6
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• pp.273-282
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• 2015

In this study, shaking table test has been carried out for the dual frame passive control system for seismic performance verification of the proposed system. The proposed system was separated into two independent frameworks that are strength resistant core and frame structure by connecting to the damper. Moreover, the seismic performance improvement of the proposed system has been verified by comparing and analyzing the experimental results of the proposed system with an existing core system. As a result of the shaking table test, acceleration and displacement responses of dual-frame vibration control system are decreased than those of the existing strength resistant type core system. In the case of the core system, while the damage was concentrated on the column of first floor, the damage of the dual system was dispersed in each layer. The damage also was concentrated on the damper, almost no damage occurs to the structural members. It has been emphasized that installed dampers in the proposed dual system reduce the input energy of whole structure by absorbing seismic input energy, which leads overall system damage to be reduced.

• Advanced Composite Materials
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• v.17 no.1
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• pp.41-56
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• 2008

In this research, optical fiber sensors and shape memory alloys (SMA) were incorporated into sandwich panels for development of a smart honeycomb sandwich structure with damage detection and shape recovery functions. First, small-diameter fiber Bragg grating (FBG) sensors were embedded in the adhesive layer between a CFRP face-sheet and an aluminum honeycomb core. From the change in the reflection spectrum of the FBG sensors, the debonding between the face-sheet and the core and the deformation of the face-sheet due to impact loading could be well detected. Then, the authors developed the SMA honeycomb core and bonded CFRP face-sheets to the core. When an impact load was applied to the panel, the cell walls of the core were buckled and the face-sheet was bent. However, after the panel was heated over the reverse transformation finish temperature of the SMA, the core buckling disappeared and the deflection of the face-sheet was relieved. Hence the bending stiffness of the panel could be recovered.

• Journal of the Korean Society of Safety
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• v.37 no.4
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• pp.129-138
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• 2022

A new method for injecting cooling water into the Korean research reactor (KRR) in the event of beam tube rupture is proposed in this paper. Moreover, the research evaluates the risk to the reactor core in terms of core damage frequency (CDF). The proposed method maintains the cooling water in the chimney at a certain level in the tank to prevent nuclear fuel damage solely by gravitational coolant feeding from the emergency water supply system (EWSS). This technique does not require sump recirculation operations described in the current procedure for resolving beam tube accidents. The reduction in the risk to the core in the event of beam tube rupture that can be achieved by the proposed change in the cooling water injection design is quantified as follows. 1) The total CDF of the KRR for the proposed design change is approximately 4.17E-06/yr, which is 8.4% lower than the CDF of the current design (4.55E-06/yr). 2) The CDF for beam tube rupture is 7.10E-08/yr, which represents an 84.1% decrease compared with that of the current design (4.49E-07/yr). In addition to this quantitative reduction in risk, the modified cooling water injection design maintains a supply of pure coolant to the EWSS tank. This means that the reactor does not require decontamination after an accident. Thermal hydraulic analysis proves that the water level in the reactor pool does not cause damage to the nuclear fuel cladding after beam tube rupture. This is because the amount of water in the chimney can be regulated by the EWSS function. The EWSS supplies emergency water to the reactor core to compensate for the evaporation of coolant in the core, thus allowing water to cover the fuel assemblies in the reactor core over a sufficient amount of time.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Journal of the Korean Society of Safety
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• v.35 no.3
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• pp.86-95
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• 2020

This paper describes the work and the results of the final Probabilistic Safety Assessment (PSA) for the Jordan Research and Training Reactor (JRTR). This final PSA was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA described here is a Level 1 PSA, which addresses the risks associated with core damage. After reviewing the documents and its conceptual design, nine typical initiating events were selected regarding internal events during the normal operation of the reactor. AIMS-PSA (Version 1.2c) was used for the accident quantification, and FTREX was used as the quantification engine. 1.0E-15/yr of the cutoff value was used to deliminate the non-effective Minimal Cut Sets (MCSs) when quantifying the JRTR PSA model. As a result, the final result indicates a point estimate of 2.02E-07/yr for the overall Core Damage Frequency (CDF) attributable to internal initiating events in the core damage state for the JRTR. A Loss of Primary Cooling System Flow (LOPCS) is the dominant contributor to the total CDF by a single initiating event (9.96E-08/yr), and provides 49.4% of the CDF. General Transients (GTRNs) are the second largest contributor, and provide 32.9% (6.65E-08/yr) of the CDF.

• Journal of the Earthquake Engineering Society of Korea
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• v.25 no.6
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• pp.251-259
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• 2021

Current seismic fragility functions for buildings were developed by defining damage state threshold based on story drift concerning foreign references and using the capacity spectrum method based on spectral displacement. In this study, insufficient details and dependence on the core location of piloti-type buildings were not reflected in the fragility function because it was developed before the Pohang earthquake. In order to develop an improved one for piloti-type buildings, several types of core were selected, damage state threshold was determined based on the capacity of structural members, and three-dimensional analyses were utilized. As a result, seismic fragility functions based on spectral acceleration were developed for various core locations and different shear strengths of the column stirrup. The fragility of piloti-type buildings significantly varied according to core location, an additional single wall, and whether the contribution of column stirrup was included or not. To estimate fragility more reasonably, it is necessary to prepare the parameters to reflect actual state well.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.949-958
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• 2024

In contrast to all-time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long-term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a MELCOR model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.