• Proceedings of the Computational Structural Engineering Institute Conference
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• 2004.04a
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• pp.301-308
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• 2004

The present study mainly focuses on the inelastic stress analysis of the 1/4 scale prestressed concrete containment vessel model(PCCV) under internal pressure and evaluates not only failure mode but also ultimate pressure capacity of the PCCV. Inelastic analysis is carried out 2D axisymmertic FE model and 3D FE model using four concrete material models which are Drucker-Prager Model, Chen-Chen Model, Damaged Plasticity Model and Menetrey-Willam Model. The uplift phenomenon of the basemat is considered in the 2D axisymmetric FE models. It is found from the 2D axisymmetric analysis results that both of Drucker-Prager model and Damaged Plasticity Model have a good performance and the uplift of the basemat is too small to influence on the global behavior of the PCCV. The FE analysis results on the ultimate pressure and failure mode have a good agreement with experimental results.

• 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.

• Computers and Concrete
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• v.14 no.2
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• pp.127-143
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• 2014

LOCA (Loss Of Coolant Accident) is one of the most important utmost accidents for Prestressed Concrete Containment Vessel (PCCV) due to its coupled effect of high temperature and inner pressure. In this paper, heat conduction analysis is used to obtain the LOCA temperature distribution of PCCV. Then the elastic internal force of PCCV under LOCA temperature is analyzed by using both simplified theoretical method and FEM (finite element methods) method. Considering the coupled effect of LOCA temperature, a nonlinear elasto-plasitic analysis is conducted for PCCV under utmost internal pressure considering three failure criteria. Results show that the LOCA temperature distribution is strongly nonlinear along the shell thickness at the early time; the moment result of simplified analysis is well coincident with the one of numerical analysis at weak constraint area; while in the strong constrained area, the value of moments and membrane forces fluctuate dramatically; the simplified and numerical analysis both show that the maximum moment occurs at 6hrs after LOCA.; the strain of PCCV under LOCA temperature is larger than the one of no temperature under elasto-plastic analysis; the LOCA temperature of 6hrs has the greatest influence on the ultimate bearing capacity with 8.43% decrease for failure criteria 1 and 2.65% decrease for failure criteria 3.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.575-602
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• 2009

In the wake of the Three Mile Island accident, vigorous research efforts were initiated to acquire a basic knowledge of the progression and consequences of accidents that involve a substantial degree of core degradation and melting. The primary emphasis of this research was placed on containment integrity, with: i) hydrogen combustion-detonation, ii) steam explosion, iii) direct containment heating (DCH), and iv) melt attack on the BWR Mark-I containment shell identified as energetic processes that could lead to early containment failure (i.e., within the first 24 hours of the accident). Should the core melt fail the reactor vessel, then non-condensable gas production from Molten Core-Concrete Interaction (MCCI) was identified as a mechanism that could fail the containment by pressurization over the long term. One signification question that arose as part of this investigation was the effectiveness of water in terminating an MCCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. Successful quenching of the core melt would prevent basemat melt through, as well as continued containment pressurization by non-condensable gas production, and so the accident progression would be successfully terminated without release of radioactivity to the environment. Based on these potential merits, ex-vessel corium coolability has been the focus of extensive research over the last 20 years as a potential accident management strategy for current plants. In addition, outcomes from this research have impacted the accident management strategies for the Gen III+LWR plant designs that are currently being deployed around the world. This paper provides: i) an historical overview of corium coolability research, ii) summarizes the current status of research in this area, and iii) highlights trends in severe accident management strategies that have evolved based on the findings from this work.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3493-3505
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• 2023

Reactor containment buildings serve as the last barrier to prevent radioactive leakage due to accidents and their safety is crucial in overpressurization conditions. Thus, the Regulatory Guide (RG) 1.216 has mentioned the global strain as one of failure criteria in the free-field for cylindrical prestressed concrete containment vessels (PCCV) subject to internal pressure. However, there is a limit that RG 1.216 shows the free-field without the specific locations of failure criteria and also the global strain corresponding to only azimuth 135° has been mentioned in NUREG/CR-6685, regardless of the elevations of the structure. Therefore, in order to reevaluate the failure criteria of the 1:4 scaled PCCV, the high fidelity simulation model based on the experimental test was significantly validated in this study, and it was interesting to find that the experimental and numerical result was very close to each other. In addition, for the consideration of the material uncertainties, the Latin hypercube method was used as a statistical approach. Consequently, it was revealed that the radial displacements of various azimuth area such as 120°, 135°, 150°, 180° and 210° at elevations 4680 mm and 6,200 mm can represent as the global deformation at the free-field, obtained from the statistical approach.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.336-347
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• 1997

During postulated severe accidents in Light water Reactors, molten corium which was ejected from the reactor vessel bottom, may erode the concrete basemat of the containment and there by threaten the containment integrity. This study experimentally examines the molten core-concrete interaction (MCC) using 20kg of thermite melt (Fe + $Al_2$O$_3$) and the concrete, used in Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3 & 4) in Korea. The measured data are the downward heat fluxes, concrete erosion rate, gases and particle generation rates during MCCI. Transient results ore compared with those of TURCIT experiment conducted by SNL in USA. The peak downward heat flux to the concrete was measured to be about 2.1㎿/$m^2$. The initial concrete erosion rate was 175cm per hour, decreasing to 30cm per hour. It was shown from the post-test that the erosion was progressed downward up to 18mm in the concrete slug.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.177-185
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• 2022

Hydrogen risk in the containment filtered venting system (CFVS) vessel was analyzed, considering operation pressure and modes with the effect of PAR and accident scenarios. The CFVS is to depressurize the containment by venting the containment atmosphere through the filtering system. The CFVS could be subject to hydrogen risk due to the change of atmospheric conditions while the containment atmosphere passes through the CFVS. It was found that hydrogen risk increased as the CFVS opening pressure was set higher because more combustible gases generated by Molten Core Concrete Interaction flowed into the CFVS. Hydrogen risk was independent of operation modes and found only at the early phase of venting both for continuous and cyclic operation modes. With PAR, hydrogen risk appeared only at the 0.9 MPa opening pressure for Station Black-Out accidents. Without PAR, however, hydrogen risk appeared even with the CFVS opening set-point of 0.5 MPa. In a slow accident like SBO, hydrogen risk was more threatening than a fast accident like Large Break Loss-of-Coolant Accident. Through this study, it is recommended to set the CFVS opening pressure lower than 0.9 MPa and to operate it in the cyclic mode to keep the CFVS available as long as possible.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1174-1183
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• 2016

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

• Structural Engineering and Mechanics
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• v.60 no.2
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• pp.337-350
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• 2016

Even though extensive researches have been performed for steam explosion due to their complex mechanisms and inherent uncertainties, establishment of severe accident management guidelines and strategies is one of state-of-the arts in nuclear industry. The goal of this research is primarily to examine effects of vessel failure modes and locations on nuclear facilities under typical steam explosion conditions. Both discrete and integrated models were employed from the viewpoint of structural integrity assessment of steel components and evaluation of the cracking and crushing in reinforced concrete structures. Thereafter, comparison of systematic analysis results was performed; despite the vessel failure modes were dominant, resulting maximum stresses at the all steel components were sufficiently lower than the corresponding yield strengths. Two failure criteria for the reinforced concrete structures such as the limiting failure ratio of concrete and the limiting strains for rebar and liner plate were satisfied under steam explosion conditions. Moreover, stresses of steel components and reinforced concrete structures were reduced with maximum difference of 12% when the integrated model was adopted comparing to those of discrete models.

• Proceedings of the Korea Concrete Institute Conference
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• 1994.10a
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• pp.285-290
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• 1994

The influence of elevated temperatures on the mechanical properties of concrete is important for fire-resistance studies and also for understanding the behavior of containment vessel, such as nuclear reactor pressure vessels, during service and ultimate condition. The present study is to clarify the damage/deterioration of concrete structures that are subjected to high temperature exposure. To this end, comprehensive experiments are conducted. The major test variables are the peak temperatures, rate of temperature increase, and sustained duration at peak temperature. The results include weight loss residual compressive strength and stress-strain curve. From those results, residua compressive strength formula and stress-strain relationship are proposed.