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

• International Journal of Air-Conditioning and Refrigeration
• /
• 제11권2호
• /
• pp.51-60
• /
• 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 occur 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, effects of turbulence penetration on the thermal stratification into T-branches with square cross-section in the modeled ECCS are analysed numerically. Standard k-$\varepsilon$ model is employed to calculate the Reynolds stresses in momentum equations. Results show that the length and strength of thermal stratification are primarily affected by the leak flow rate of coolant and the Reynolds number of duct. Turbulence penetration into the T-branch of ECCS shows two counteracting effects on the thermal stratification. Heat transport by turbulence penetration from main duct to leaking flow region may enhance thermal stratification while the turbulent diffusion may weaken it.

• 설비공학논문집
• /
• 제15권3호
• /
• pp.239-245
• /
• 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 occur 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, effects of turbulence penetration on the thermal stratification into T-branches with square cross-section in the modeled ECCS are analysed numerically. $textsc{k}$-$\varepsilon$ model is employed to calculate the Reynolds stresses in momentum equations. Results show that the length and strength of thermal stratification are primarily affected by the leak flow rate of coolant and the Reynolds number of the main flow in the duct. Turbulence penetration into the T-branch of ECCS shows two counteracting effects on the thermal stratification. Heat transport by turbulence penetration from the main duct to leaking flow region may enhance thermal stratification while the turbulent diffusion may weaken it.

• 시스템엔지니어링학술지
• /
• 제10권1호
• /
• pp.33-41
• /
• 2014

New design concept of Reactor Coolant System (RCS) including a reactor assembly for the SMR is introduced in this work. An exploration of new type of reactor that is advanced from proposed SMRs is performed by using systems engineering approach. In this point of view project structured on three main phases; needs analysis (NA), concept exploration (CE), and concept definition (CD). Main objectives as an output of the CE stage are a small size, low cost, shortening the schedule, and enhancing safety. The SMRs usually have a small size requirement. In order to meet the size requirement and to achieve a productivity, in other words, easiness to manufacture, this paper suggests an integrated PWR design concept through researching predecessors. Although the integrated PWR concept provides many advantages, it has disadvantages that composite of maintenance and a low availability problem. Therefore, this paper comes up with a run-to-fail design concept based on modular design to address the maintenance problem and to maximize the availability of SMRs as well as to be compatible with the overall-SMRs including Barge Mounted(BM)type.

• 방사성폐기물학회지
• /
• 제19권1호
• /
• pp.51-58
• /
• 2021

Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1995년도 추계학술발표회논문집(1)
• /
• pp.403-410
• /
• 1995

A loss of the residual heat removal (RHR) accident during mid-loop operation of Yong-gwang Nuclear Units 3/4 was analyzed using the RELAP5/MOD3.1.2 code. In this work the following assumptions are used; (i) initially the reactor coolant system (RCS) above the hot leg center line is filled with nitrogen gas, (ii) two 3/4-inch diameter vent valves on the reactor vessel head and the top of pressurizer in the reactor coolant system are always open, and a level indicator is connected to the RMR suction line, (iii) the two steam generators are in wet layup status and the steam generator atmospheric dump valve assemblies are removed so that the secondary side pressure remains at nearly atmospheric condition throughout the accident, and (iv) the loss of RHR is presumed to occur at 48 hours after reactor shutdown. Findings from the RELAP5 calculations are (i) the core boiling begins at ∼5 min, (ii) the peak RCS pressure is ∼3.0 bar, which implies a possibility of temporary seal break, (iii) ∼94 % of the decay heat is removed by reflux condensation in the steam generator U-tubes in spite of the presence of noncondensable gas, (iv) the core uncovery time is evaluated to be 7.2 hours. Significant mass errors were observed in the calculations.

• 한국압력기기공학회 논문집
• /
• 제19권1호
• /
• pp.44-51
• /
• 2023

The RCP is one of the main components in nuclear power plants and plays an important role in circulating coolant to the RCS system. Currently, nuclear plants are monitored using various monitoring systems. However, since they operate independently according to their functional purpose, it is not able to analyze vibration and operation/performance information comprehensively, and thus failure diagnosis accuracy is limited. In addition, these systems do not provide some important information (such as fault type, parts and cause) necessary for emergency actions, but provide only alarm information. To improve these technical problems, this study proposes a diagnosis technique (M/L, Rule-based model, Data-driven model, Narrow band model) and methodology for comprehensive analysis.

• 한국압력기기공학회 논문집
• /
• 제4권2호
• /
• pp.40-45
• /
• 2008

There exist many types of pipe and component fatigue through vibrations, thermal fatigues or shifting. In some cases of thermal stratification/thermal fatigue, pipes & components are receiving thermal stress by means of material expansion and shrinkage by continuous thermal repetitive variation. Small cracks initially occur on the inside surface by thermal stress. These cracks grow in depth the pipe wall and finally come to a rupture. Pipe parts of susceptibility to thermal stratification and thermal fatigue are now being examined by conventional UT(ultrasonic test) as volumetric examination. It is difficult to fully satisfy the code & standards requirements because 3" weldolet weldments of RCS 16" pipe to 3" branch connection lines have complex structural shape. To solve the problems of conventional UT examination, we made a realistic mock-up and UT calibration block. We performed a simulation of phased array UT utilizing CIVA as NDE(Non-Destructive Examination) simulation software. Also we designed phased array UT transducer and wedge, optimal frequency by using simulation data. We performed phased array UT experiment through mock-up including artificial flaws(notch). The phased array UT technique is finally developed to improve the reliability of ultrasonic test at RCS 16" pipe to 3" branch connection weld.

• Nuclear Engineering and Technology
• /
• 제51권4호
• /
• pp.1109-1116
• /
• 2019

It is essential to predict the radioactivity distribution around the reactor cooling system (RCS) during obligatory cyclic operation of AP-1000. A home-developed program CPA-AP1000 is upgraded to predict the response of activated corrosion products (ACPs) in the RCS. The program is written in MATLAB and it uses state of the art MCNP as a subroutine for flux calculations. A pair of cyclic power profiles were superimposed after initial full power operation. The effect of cyclic operation is noticed to be more prominent for in-core surfaces, followed by the primary coolant and out-of-core structures. The results have shown that specific activity trends of $^{56}Mn$ and $^{24}Na$ promptly follow the power variations, whereas, $^{59}Fe$, $^{58}Co$, $^{99}Mo$ and $^{60}Co$ exhibit a sluggish power-following response. The investigations pointed out that promptly power-following response of ACPs in the coolant is vital as an instant radioactivity source during leakage incidents. However, the ACPs with delayed power-following response in the out-of-core components are perceived to cause a long-term activity. The present results are found in good agreement with those for a reference PWR. The results are useful for source term monitoring and optimization of work procedures for an innovative reactor design.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2014년도 추계학술대회 논문집
• /
• pp.110-115
• /
• 2014

The nuclear power plant was launched by Kori unit 1 in 1978 years. Currently, 23 nuclear power plants have been operating in Korea since 1978 years. The localization was completed for most of the reactor facility from Hanbit(Youngkwang) unit 3&4. However, RCP(Reactor Coolant Pump) and MMIS(Man Machine Interface System) is an important technology that has been excluded from the scope of the technical transfer has been dependent on a specific overseas vendor. Recent success in RCP development through co-operation with government and industries. Developed RCP will be applied to Shin-Hanul unit 1&2 nuclear power plants. The RCP operates in high speed and high pressure condition and only rotating component in the NSSS(Nuclear Steam Supply System). Therefore, the problem of vibration has arisen caused by the hydraulic forces of the working fluid. These forces can influence on the stability characteristics for entire RCS(Reactor Coolant System) loop, and can act as significant destabilizing forces. In this study, vibration evaluation of the pump shaft of development RCP estimated under normal operation and over speed conditions. In order to predict the vibration characteristics and dynamic behavior, modal analysis, critical speed analysis and unbalance response spectrum analysis were performed.