• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.110-114
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• 2001

A numerical analysis has been perfonned to estimate the effect of turbulent penetration and thermal stratified flow in the branch lines piping. This phenomenon of thermal stratification are usually observed in the piping lines of the safety related systems and may be identified as the source of fatigue in the piping system due to the thermal stress loading which are associated with plant operating modes. The turbulent penetration length reaches to $1^{st}$ valve in safety injection piping from reactor coolant system (RCS) at normal operation for nuclear power plant when a coolant does not leak out through valve. At the time, therefore, the thermal stratification does not appear in the piping between RCS piping and $1^{st}$ valve of safety injection piping. When a coolant leak out through the $1^{st}$ valve by any damage, however, the thermal stratification can occur in the safety injection piping. At that time, the maximum temperature difference of fluid between top and bottom in the piping is estimated about $50^{\circ}C$.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.1633-1638
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• 2003

In case that in-leakage through the valve disk occurs, a numerical study is performed to estimate on thermal stratification phenomenon in the Safety Injection piping connected with the Reactor Coolant System piping of Nuclear Power Plant. As the leakage flow rate increases, the temperature difference between top and bottom of horizontal piping has the inflection point. In the connection point of valve and piping, the maximum temperature difference between top and bottom was 185K and occurred in the condition of 10 times of standard leakage flow rate. In the connection point of elbow and horizontal piping, the maximum temperature difference was 145K and occurred in the condition of 15 times of standard leakage flow rate. In the vertical piping of Safety Injection piping, the near of connection point between elbow and vertical piping showed the outstanding thermal stratification phenomenon in comparison with another region because of turbulent penetration from Reactor Coolant System piping. In order to prevent damage of piping due to the thermal stratification when in-leakage through the valve disk occurs, the connection points between valve and piping, and the connection points between elbow and piping need to be inspected continually.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.237-248
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• 2013

Thermal stratification has continuously caused several piping failures in nuclear power plants since the early 1980s. However, this critical thermal effect was not considered when the old nuclear power plants were designed. Therefore, it is urgent to evaluate this unexpected thermal effect on the structural integrity of piping systems. In this paper, the thermal effects of stratified flow in two different safety injection piping systems were investigated by using a coupled CFD-FE method. Since stratified flow is generally generated by turbulent penetration and/or valve leakage, thermal stress analyses as well as CFD analyses were carried out considering these two primary causes. Numerical results show that the most critical factor governing thermal stratification is valve leakage and that temperature distribution significantly changes according to the leakage path. In particular, in-leakage has a high possibility of causing considerable structural problems in RCS piping.

• 한국안전학회지
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• 제14권2호
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• pp.32-41
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• 1999

If the Leak-Before-Break (LBB) concept is applicable to the nuclear piping design, it is not necessary to consider the dynamic effect due to pipe rupture. Therefore, the construction cost can be significantly reduced by eliminating unnecessary pipe whip restraints and jet impingement devices. The objective of this paper is to develop the Piping Evaluation Diagram (PED) for efficient application of LBB concept to piping system at an initial piping design stage. For this purpose, the 3-D finite element analyses were performed to evaluate the crack stability. And the stress-strain curve based on the pipe material tests were used to calculate the detectable leakage crack length. Finally, the present PED which was composed as a function of NOP load and allowable SSE load, was developed for an application of LBB concept to the safety injection and shutdown cooling line in Korean Next Generation Reactor (KNGR).

• 한국압력기기공학회 논문집
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• 제7권1호
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• pp.1-8
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• 2011

Advanced SITs of the evolutionary PWRs have the advantage that they can passively control the ECC water discharge flow rate. Thus, the LPSI pumps can be eliminated from the safety injection system owing to the benefit of the advanced SITs. In the present study, a passive sealing plate was designed in order to overcome the shortcoming of the advanced SITs, i.e., the early nitrogen discharge through the stand pipe. The operating principle of the sealing plate depends only on the natural phenomena of buoyancy and gravity. The performance of the sealing plate was evaluated using the VAPER test facility, equipped with a full-scale SIT. It was verified that the passive sealing plate effectively prevented the air discharge during the entire duration of the ECC water discharge. Also, the major performance parameters of the advanced SIT were not changed with the installation of the sealing plate.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.432-437
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• 2000

The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The the radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.1833-1838
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• 2003

In the nuclear power plant, emergency core coolant system(ECCS) is furnished at reactor coolant system(RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can be occurred due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack(TFC) accident. In the present study, when the turbulence penetration occurs in the branch piping, the maximum temperature differences of fluid at the pipe cross-sections of the T-branch with thermal stratification are examine

• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.54-60
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• 2010

Piping failures due to thermal fatigue have been widely reported in normally stagnant non-isolable reactor coolant branch lines. Since the thermal fatigue due to thermal stratification was not considered in the piping fatigue design in old NPPs, it is important to evaluate the effect of thermal stratification on the integrity of branch lines. In this study, geometrical screening criteria for Up-horizontal branch lines in MRP-132 were applied to SI(Safety Injection) lines of KSNP 2-loop and WH 3-loop. Some computational fluid dynamic(CFD) analyses on the Reactor Coolant System(RCS) branch lines were also performed to develop the regulatory guidelines for screening criteria. As a result of applying MRP-132 screening criteria, KSNP 2-loop and WH 3-loop SI lines are determined to need further detailed evaluation. Results of CFD analyses show that both valve isolation and amount of leakage through valve can be used as technical bases for the screening criteria on the thermal fatigue analysis.

• 대한기계학회논문집
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• 제19권6호
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• pp.1451-1459
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• 1995

The objective of this paper is to evaluate the material properties of SA312 TP316 and SA312 TP304 stainless steels and their associated welds manufactured for safety injection system of Yonggwang 3,4 nuclear generating stations. A total of 62 tensile tests and 46 fracture toughness tests were conducted and the effects of various parameters such as pipe size, crack plane orientation, tests were conducted and the effects of various parameters such as pipe size, crack plane orientation, test temperature, welding on material properties were discussed. Test results show that the effect of test temperature on fracture toughness was significant while the effects of pipe size and crack plane orientation on fracture toughness were negligible. Fracture toughness of the weld metal was in general higher than that of the base metal.

• Journal of Mechanical Science and Technology
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• 제19권5호
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• pp.1206-1215
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• 2005

The thermal stratification phenomena, frequently occurring in the component of nuclear power plant system such as pressurizer surge line, steam generator inlet nozzle, safety injection system (SIS), and chemical and volume control system (CVCS), can cause through-wall cracks, thermal fatigue, unexpected piping displacement and dislocation, and pipe support damage. The phenomenon is one of the unaccounted load in the design stage. However, the load have been found to be serious as nuclear power plant operation experience accumulates. In particular, the thermal stratification by the turbulent penetration or valve leak in the SIS and SCS pipe line can lead these safety systems to failure by the thermal fatigue. Therefore in this study an 1/10 scaledowned experimental rig had been designed and installed. And a series of experimental works had been executed to measure the temperature distribution (thermal stratification) in these systems by the turbulent penetration, valve leak, and heat transfer through valve. The results provide very valuable informations such as turbulent penetration depth, the possibility of thermal stratification by the heat transfer through valve, etc. Also the results are expected to be useful to understand the thermal stratification in these systems, establish the thermal strati­fication criteria and validate the calculation results by CFD Codes such as Fluent, Phenix, CFX.