• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1264-1272
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• 2016

The accurate estimation of leak rate through cracks is crucial in applying the leak before break (LBB) concept to pipeline design in nuclear power plants. Because of its importance, several programs were developed based on the several proposed flow models, and used in nuclear power industries. As the flow models were developed for a homogeneous pipe material, however, some difficulties were encountered in estimating leak rates for bimaterial pipes. In this paper, a flow model is proposed to estimate leak rate in bimaterial pipes based on the modified Henry-Fauske flow model. In the new flow model, different crack morphology parameters can be considered in two parts of a flow path. In addition, based on the proposed flow model, a program was developed to estimate leak rate for a crack with linearly varying cross-sectional area. Using the program, leak rates were calculated for through-thickness cracks with constant or linearly varying cross-sectional areas in a bimaterial pipe. The leak rate results were then compared and discussed in comparison with the results for a homogeneous pipe. The effects of the crack morphology parameters and the variation in cross-sectional area on the leak rate were examined and discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.35-46
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• 2020

In this paper, a fracture mechanics evaluation system which can be used to assess the leak-before-break (LBB) of nuclear piping is developed. Existing solutions for calculating the fracture mechanics parameters (J-integral and crack opening displacement) required for LBB evaluation were firstly presented. Then a module for calculating J-integral and COD was developed, with an additional module for predicting the critical load based on the crack driving force diagram to finally develop a fracture mechanics evaluation system. To confirm the validity of the proposed evaluation system, finite element (FE) analysis was performed, and the FE J-integral and COD results were compared with prediction results using the J-integral and COD estimations program. Furthermore, the critical load assessment module was verified by comparing the actual pipe test results (Battelle test data) with prediction results using the proposed program.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.20 no.6
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• pp.1862-1871
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• 1996

The puopose of this paper is sto modify the current LBB criteria. The validity of current LBB criteria and current standard LBB analysis mehtod are evaluated using linear elastic fracture mechanics and elastic-plastic fracture mechanics. The results of evaluation demonstrate that the current LBB driteria are very conservative and some level of margins already exist in the standard LBB analysis method. Thus, the margin on load .root. and margin on crack size 2 can be eliminated to extend LBB application for the samller diameter pipe.

• Journal of the Korean Society for Precision Engineering
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• v.18 no.7
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• pp.85-90
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• 2001

In order to apply the Leak-Before-Break(LBB)concept to nuclear piping, accurate estimation of J-integral and crack opening displacement(COD) is essential for complex loading, such as combined tension and bending. This paper proposes a new engineering method to estimate J-integral and the COD for circumferential through-wall cracked pipes subject to combined tension and bending loading. The proposed method to estimate the COD is validated against three published pipe test data, generated from a monotonically increasing bending load with a constant internal pressure, which shows excellent agreements.

• Journal of the Korean Society for Precision Engineering
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• v.31 no.7
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• pp.623-626
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• 2014

The LBB (Leak Before Break) concept is based on evaluating the fracture toughness. NUREG 1061, Vol.3 announced that the specimen for evaluating fracture resistance needs to have same thickness or thicker than pipe. Therefore, it is difficult to collect specimen from pipe which has same thickness as a pipe. So, ASTM standard suggested the use of standard specimen with 1 inch thickness. However, it has been known that an application of LBB by test results of standard specimen is conservative compare with that by real pipe. In this study, to supplement such conservatism, the evaluation of fracture resistance characteristics was performed with curved CT specimen, which has same thickness and curvature as a pipe. In addition, fracture resistance characteristics of curved CT specimen were compared with those of CT specimen. For this, shape factor F, hpl and g were obtained from FEM analysis using the limit load method.

• The Journal of Korea Robotics Society
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• v.17 no.2
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• pp.178-182
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• 2022

Post-tension duct in nuclear reactor containment building is filled with grease to prevent steel strand from corroding. If grease leaks by break of duct, steel strand will corrode and cause problem in building safety. Therefore, grease leak should be checked preventatively. But currently used method is inefficient, since it has to remove grease and strand to check. And other methods for checking post-tension dust are not well-researched. In this paper, we develop screw-propelled robot that can move in grease and detect grease duct directly. Also, we make the test environment that is similar to real post-tension duct of containment building and test robot in that environment to verify that our robot is feasible in the post-tension duct. The robot can move forward and backward in grease duct by twin screw mechanism and has mount for sensors to detect grease leakage and strand corrosion.

• Journal of the Earthquake Engineering Society of Korea
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• v.4 no.1
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• pp.51-62
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• 2000

핵연료집합체 검증 프로그램의 일환으로 본 연구에서는 지진과 배과파단이 핵연료집합체의 건선성에 미치는 영향을 검토하였다 원자로 노심의 상세 동적해석을 이용하여 지진 및 배과파단시 핵연료 집합체에 발생하는 전단력 굽힘 모우멘트 및 변위를 계산하였고 또한 집합체를 지지하고 있는 지지격자체의 충격력을 검토하였다 이들 하중에 대한 핵연료집합체의 응력해석을 수행하여 사고조건하에서의 구조적 건전성에 대하여 언급하였고 추후 설계시 고려할 사항을 제시하였다.

• Journal of the Korean Society of Safety
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• v.24 no.5
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• pp.63-68
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• 2009

The urgent VAI method development is required since "The Act of Physical Protection and Radiological Emergency that is established in 2003" requires an evaluation of physical threats in nuclear facilities and an establishment of physical protection in Korea. The VAI methodology is developed to (1) make a sabotage model by reusing existing fire/flooding/pipe break PSA models, (2) calculate MCSs and TEPSs, (3) select the most cost-effective TEPS among many TEPSs, (4) determine the compartments in a selected TEPS as vital areas, and (5) provide protection measures to the vital areas. The developed VAI methodology contains four steps, (1) collecting the internal level 1 PSA model and information, (2) developing the fire/flood/pipe rupture model based on level 1 PSA model, (3) integrating the fire/flood/pipe rupture model into the sabotage model by JSTAR, and (4) calculating MCSs and TEPS. The VAT process is performed through the VIPEX that was developed in KAERI. This methodology serves as a guide to develop a sabotage model by using existing internal and external PSA models. When this methodology is used to identify the vital areas, it provides the most cost-effective method to save the VAI and physical protection costs.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.35 no.5
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• pp.1083-1096
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• 2015

This study utilised Computational Fluid Dynamics(CFD) for preliminary assessment of mixing efficiencies of 2-phase fluids in a pipe at which a slurry flow and an injected solidifier join, for the purposes of reducing trial-and-error-based instances of physical experiments and conducting the overall research in an economical way. Using OpenFOAM$^{(R)}$, we simulated behavior of 3-phase fluids under 18 different settings generated by changing diameters of a dredged soil transportation pipe, a quality improving material injection pipe and the confluence angle. While difference in mixing efficiencies amongst the instances was insignificant, discernible boundary layers amongst the materials were observed in all of the instances. In order to break the boundary layers, we designed a substructure inside a pipe and found out that it could remarkably improve mixing efficiencies particularly for short distance applications.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.126-133
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• 2016

Recently, the research and development of Sodium-cooled Fast Reactors (SFRs) have made progresses. However, liquid sodium, the coolant of an SFR, is chemically unstable and sodium fire can be occurred when liquid sodium leaks from sodium pipe. To reduce the damage by the sodium fire, many fire walls and fire extinguishers are needed for SFRs. LBB concept in SFR might reduce the scale of sodium fire and decrease or eliminate fire walls and fire extinguishers. Therefore, LBB concept can contribute to improve economic efficiency and to strengthen defense-in depth safety. The LBB assessment procedure has been well established, and has been used significantly in light water reactors (LWRs). However, an LBB assessment of an SFR is more complicated because SFRs are operated in elevated temperature regions. In such a region, because creep damage may occur in a material, thereby growing defects, an LBB assessment of an SFR should consider elevated temperature effects. The procedure and method for this purpose are provided in RCC-MRx A16, which is a French code. In this study, LBB assessment was performed for PGSFR IHTS hot-leg pipe according to RCC-MRx A16 and the applicability of the code was discussed.