• Journal of Sensor Science and Technology
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• v.22 no.1
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• pp.71-75
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• 2013

Nuclear containment building is used as second blockage to protect us from a radiation leakage caused by the natural disaster or any accidents, so it's safety is important and must be kept with continuous surveillance. In this study, we measured the strain of a nuclear containment building's wall by using FBG sensor and investigated the structural safety of a nuclear containment building. 50 FBG strain sensors and 18 FBG strain sensors were attached on the side wall and upper dome of a nuclear containment building, respectively. We measured the strains of the outside concrete wall during the Structural Integrity Test (SIT) of a nuclear containment building. The strain of an upper dome was larger than that of a side wall, about $200{\mu}{\varepsilon}$. And the very small strain was measured at vertical direction of a side wall. These experimental results were used to evaluate the structural health of nuclear containment building.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2053-2068
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• 2023

Following a loss of coolant accident (LOCA), prestressing concrete containment vessels (PCCVs) may experience high thermal load as well as internal pressure. The high temperature stress would increase the risk of premature damage to the containment, which reduces the safety margin during the increasing internal pressure. However, current investigations cannot clearly address the issues of thermal-pressure coupling effect on damage propagation and thus safety of the containment. Thus, this paper offers three simple and powerful damage parameters to differentiate the severity of damage of the containment. Moreover, despite of the temperature action severely threatening the pressure performance of the containment, the research regarding the improvement of the resistant performance of the containment is quite scarce. Therefore, in this paper, a comprehensive comparison of damage propagation and mechanism between conventional and fiber-reinforced concrete (FRC) containments is performed. The effects of fiber characteristics parameters on damage propagation of structures following the LOCA are also specifically revealed. It is found that the proposed damage indices can properly indicate state of damage in the containment body and the addition of fiber can be used to obviously mitigate the damage propagation in PCCV considering the thermal-pressure coupling.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.434-441
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• 2017

Containment venting is one of several essential measures to protect the integrity of the final barrier of a nuclear reactor during severe accidents, by which the uncontrollable release of fission products can be avoided. The authors seek to develop an optimization approach to venting operations, from a simulation-based perspective, using an integrated severe accident code, THALES2/KICHE. The effectiveness of the containment-venting strategies needs to be verified via numerical simulations based on various settings of the venting conditions. The number of iterations, however, needs to be controlled to avoid cumbersome computational burden of integrated codes. Bayesian optimization is an efficient global optimization approach. By using a Gaussian process regression, a surrogate model of the "black-box" code is constructed. It can be updated simultaneously whenever new simulation results are acquired. With predictions via the surrogate model, upcoming locations of the most probable optimum can be revealed. The sampling procedure is adaptive. Compared with the case of pure random searches, the number of code queries is largely reduced for the optimum finding. One typical severe accident scenario of a boiling water reactor is chosen as an example. The research demonstrates the applicability of the Bayesian optimization approach to the design and establishment of containment-venting strategies during severe accidents.

• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.211-218
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• 2010

An analysis model on debris transport in the containment floor of pressurized water reactors is developed in which the flow field is calculated by Eulerian conservation equations of mass and momentum and the debris particles are traced by Lagrange equations of motion using the pre-determined flow field data. For the flow field calculation, two-dimensional Shallow Water Equations derived from Navier Stokes equations are solved using the Finite Volume Method, and the Harten-Lax-van Leer scheme is used for accuracy to capture the dry-to-wet interface. For the debris tracing, a simplified two-dimensional Lagrangian particle tracking model including drag force is developed. Advanced schemes to find the positions of particles over the containment floor and to determine the position of particles reflected from the solid wall are implemented. The present model is applied to calculate the transport fraction to the Hold-up Volume Tank in Advanced Power Reactors 1400. By the present model, the debris transport fraction is predicted, and the effect of particle density and particle size on transport is investigated.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1045-1052
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• 2009

A calculation model is developed to predict the transient free surface flow on the containment floor following a loss-of-coolant accident (LOCA) of pressurized water reactors (PWR) for the use of debris transport evaluation. The model solves the two-dimensional Shallow Water Equation (SWE) using a finite volume method (FVM) with unstructured triangular meshes. The numerical scheme is based on a fully explicit predictor-corrector method to achieve a fast-running capability and numerical accuracy. The Harten-Lax-van Leer (HLL) scheme is used to reserve a shock-capturing capability in determining the convective flux term at the cell interface where the dry-to-wet changing proceeds. An experiment simulating a sudden break of a water reservoir with L-shape open channel is calculated for validation of the present model. It is shown that the present model agrees well with the experiment data, thus it can be justified for the free surface flow with accuracy. From the calculation of flow field over the simplified containment floor of APR1400, the important phenomena of free surface flow including propagations and interactions of waves generated by local water level distribution and reflection with a solid wall are found and the transient flow rates entering the Holdup Volume Tank (HVT) are obtained within a practical computational resource.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3677-3691
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• 2023

Since containment building is the major shielding structure to ensure safety of nuclear power plant, the structural behavior and ultimate pressure capacity of containments must be studied in depth. This paper addresses ambiguous issue of determining free-field position for liner failure by suggesting an expanded free-field region and comparing internal pressure capacities obtained by test data, conservative assumption and suggested free-field region. For this purpose, a practical approach to determine the free-field position for the evaluation of liner tearing is carried out. The maximum principal strain histories versus internal pressure capacities among different free-field positions at various azimuths and elevations are compared with those at the equipment hatch as a conservative assumption. The comparison shows that there are considerable differences in the internal pressure capacity at liner failure within the expanded free-field region compared to the vicinity of the equipment hatch. Additionally, this study proposes an approximate correlation with conservative factors by considering the expanded free-field ranges and material characteristics to determine realistic failure criteria for liner. The applicability of the proposed correlation is demonstrated by comparing the internal pressure capacities of full-scale containment buildings following liner failure criteria according to RG 1.216 and an approximate correlation.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.884-894
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• 2015

Background: Fibers in concrete resist the growth of cracks and enhance the postcracking behavior of structures. The addition of fibers into a conventional reinforced concrete can improve the structural and functional performance of safety-related concrete structures in nuclear power plants. Methods: The influence of fibers on the ultimate internal pressure capacity of a prestressed concrete containment vessel (PCCV) was investigated through a comparison of the ultimate pressure capacities between conventional and fiber-reinforced PCCVs. Steel and polyamide fibers were used. The tension behaviors of conventional concrete and fiber-reinforced concrete specimens were investigated through uniaxial tension tests and their tension-stiffening models were obtained. Results: For a PCCV reinforced with 1% volume hooked-end steel fiber, the ultimate pressure capacity increased by approximately 12% in comparison with that for a conventional PCCV. For a PCCV reinforced with 1.5% volume polyamide fiber, an increase of approximately 3% was estimated for the ultimate pressure capacity. Conclusion: The ultimate pressure capacity can be greatly improved by introducing steel and polyamide fibers in a conventional reinforced concrete. Steel fibers are more effective at enhancing the containment performance of a PCCV than polyamide fibers. The fiber reinforcementwas shown to bemore effective at a high pressure loading and a lowprestress level.

• Journal of Advanced Marine Engineering and Technology
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• v.40 no.3
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• pp.165-173
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• 2016

On March 11, 2011, a massive earthquake measuring 9.0 on the Richter scale and subsequent 10-.14 m waves struck the Fukushima Daiichi (FD) Nuclear Power Plant. The main and backup electric power was damaged preventing the cooling system from functioning. Fuel rods overheated and led to hydrogen explosions. If heat in the fuel rods is not dissipated, the nuclear fuel coating material (e.g., Zircaloy) reacts with water vapor to generate hydrogen at high temperatures. This hydrogen is released into the containment area. If the released hydrogen burns, the stability of the containment area is significantly impacted. In this study, researchers performed an explosion analysis in a high-risk explosion area, analyzing the hydrogen distribution in a containment building [1] and the effects of a hydrogen explosion on containment safety. Results indicated that a hydrogen explosion was possible throughout the containment building except the middle area. If an explosion occurs at the top of the containment building with more than 40% of the hydrogen collected or in the bottom right or left side of the of containment building, safety of the containment building could be threatened.

• Journal of the Korean Institute of Gas
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• v.17 no.2
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• pp.85-89
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• 2013

The purpose of the development of KC-1 LNG cargo containment system is reduction in royalty and increase in competitiveness of shipbuilding industry. An assessment of structure safety for LNG cargo containment system under sloshing load due to ship motion has become an important design element. The ideal way is to implement fully interaction of the fluid domain and the cargo containment system. However the irregular sloshing pressure were idealized in the form of a triangular wave for safety assessment because the fluid- structure interaction analysis is taken the extensive computation time and difficult to ensure the accuracy of the results. In this study, the sloshing load was assumed to be a triangular wave with a maximum pressure of 10 bar during 15/1000 seconds. In the analytic results, the basic insulation panel of KC-1 LNG cargo containment system was assessed to be structurally safe for sloshing load.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.37-41
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• 2020

The Nuclear containment building is a main safety-related structure that performs shielding and conservation functions to prevent highly radioactive materials from leakage to the outside environment in the case of various environmental conditions and postulated accidents. The containment building contains a reactor, steam generator, pressurizer, tank, reactor coolant system, auxiliary system and engineering safety system, and is designed so that highly radioactive materials above the limits specified in 10 CFR 100 do not escape to the outside environment in the case of LOCA(Loss of Coolant Accident) for instance. The containment metal liner plate(CLP) is a carbon steel plate with a nominal plate thickness of 6 mm, which functions as a mold for the wall and dome of the containment building when concrete is filled, fulfills airtightness to prevent leakage of seriously radioactive materials. In recent years, backside corrosion was found on the liner plate in some domestic nuclear power plants. The main cause of backside corrosion was unfilled concrete. In this paper, an inspection technique of assessing filling suitability for CLP backside concrete is developed. Results show that the validity of inspection technique for CLP backside concrete using vibration sensor is successfully verified.