• Nuclear Engineering and Technology
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• v.39 no.5
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• pp.647-654
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• 2007

Postulation of flaws, one of the most important areas in RPV integrity assessment, significantly affects the results. In the present work, several parameters, such as orientation, underclad vs. surface cracking, crack depth and shape, etc., are postulated and parametric studies are performed to investigate the influence of the flaw parameters on the structural integrity assessment of the reactor pressure vessel during pressurized thermal shock. The influence of individual parameters describing the crack is evaluated based on sensitivity study results.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.90-95
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• 2004

The probabilistic fracture analysis is used to determine the effects of uncertainties involved in material properties, location and size of flaws, etc, which can not be addressed using a deterministic approach. In this paper the probabilistic fracture analysis is applied for evaluating the RV(Reactor Vessel) under PTS(Pressurised Thermal Shock). A semi-elliptical axial crack is assumed in the inside surface of RV. The selected random parameters are initial crack depth, neutron fluence, chemical composition of material (copper, nickel and phosphorous) and $RT_{NDT}$. The deterministically calculated $K_I$ and crack tip temperature are used for the probabilistic calculation. Using Monte Carlo simulation, the crack initiation probability for fixed flaw and PNNL(Pacific Northwest National Laboratory) flaw distribution is calculated. As the results show initiation probability of fixed flaw is much higher than that of PNNL distribution, the postulated crack sizes of 1/10t in this paper and 1/4t of ASME are evaluated to be very conservative.

• Journal of Mechanical Science and Technology
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• v.17 no.9
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• pp.1380-1387
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• 2003

When a cold HPSI (High pressure Safety Injection) fluid associated with an overcooling transient, such as SGTR (Steam Generator Tube Rupture), MSLB (Main Steam Line Break) etc., enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters the downcomer of the reactor pressure vessel, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. As general thermal-hydraulic system analysis codes cannot properly predict the thermal stratification phenomena, RG 1.154 requires that a detailed thermal-mixing analysis of PTS (pressurized Thermal Shock) evaluation be performed. Also. previous PTS studies have assumed that the thermal stratification phenomena generated in the stagnated loop side of a partially stagnated primary coolant loop are neutralized in the vessel downcomer by the strong flow from the unstagnated loop. On the basis of these reasons, this paper focuses on the development of a 3-dimensional thermal-mixing analysis model using PHOENICS code which can be applied to both partial and total loop stagnated cases. In addition, this paper verifies the fact that, for partial loop stagnated cases, the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is almost neutralized by the strong flow of the unstagnated loop but is not fully eliminated.

• Proceedings of the KSME Conference
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• 2000.04a
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• pp.908-913
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• 2000

In the process of integrity evaluation for nuclear power plant components, a series of fracture mechanics evaluation on surface cracks in reactor pressure vessel(RPV) must be conducted. These fracture mechanics evaluations are based on stress intensity factor, K. However, under pressurized thermal shock(PTS) conditions, the combination of thermal and mechanical stress by steep temperature gradient and internal pressure causes considerably high tensile stress at the inside of RPV wall. Besides, the internal pressure during the normal operation produces high tensile stress at the RPV wall. As a result cracks on inner surface of RPVs may experience elastic-plastic behavior which can be explained with J-integral. In such a case, however, J-integral may possibly lose its validity due to constraint effect. In this paper, in order to verify the suitability of J-integral, two dimensional finite element analyses were applied for various surface crack. Total of 18 crack geometries were analyzed, and Q stresses were obtained by comparing resulting HRR stress distribution with corresponding actual stress distributions. In conclusion, HRR stress fields were found to overestimate the actual crack-tin stress field due to constraint effect.

• Proceedings of the Korean Institute of Industrial Safety Conference
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• 2000.11a
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• pp.263-268
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• 2000

원자력 압력용기의 건전성 평가 및 안전성 확보에 대한 관심은 1978년 미국 Rancho Seco 발전소에서 발생한 가압열충격 사고로 인해 크게 부각되기 시작하였다. 가압열충격(Pressurized Thermal Shock： PTS)이란 계통의 압력이 높은 상태이거나 증가중인 상태에서 급속한 냉각과 과도한 냉각이 발생하는 것을 의미한다. 이러한 냉각에 의해 원자로용기 외벽보다 내벽이 빨리 냉각되어 상당한 온도구배가 발생하고 이 온도구배에 따라 용기 내벽에 최대인장 열응력이 발생한다.(중략)

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1423-1432
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• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

• Nuclear Engineering and Technology
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• v.40 no.5
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• pp.349-364
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• 2008

Over the years, probabilistic risk assessment (PRA) research activities conducted at the U.S. Nuclear Regulatory Commission (NRC) have played an essential role in support of the agency's move towards risk-informed regulation. These research activities have provided the technical basis for NRC's regulatory activities in key areas; provided PRA methods, tools, and data enabling the agency to meet future challenges; supported the implementation of NRC's 1995 PRA Policy Statement by assessing key sources of risk; and supported the development of necessary technical and human resources supporting NRC's risk-informed activities. PRA research aimed at improving the NRC's understanding of risk can positively affect the agency's regulatory activities, as evidenced by three case studies involving research on fire PRA, human reliability analysis (HRA), and pressurized thermal shock (PTS) PRA. These case studies also show that such research can take a considerable amount of time, and that the incorporation of research results into regulatory practice can take even longer. The need for sustained effort and appropriate lead time is an important consideration in the development of a PRA research program aimed at helping the agency address key sources of risk for current and potential future facilities.

• Nuclear Engineering and Technology
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• v.24 no.1
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• pp.86-97
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• 1992

The purpose of this paper is to evaluate the direct vessel injection design from a pressurized thermal shock(PTS) viewpoint for the Combustion Engineering System 80+ A break of the main steam line from zero power and a 0.05 ft$^2$small break loss-of-coolant accident (LOCA) from full power were selected as the potential PTS events. In order to investigate the stratification effects in the reactor downcomer region, the fluid mixing analysis was performed using the COMMIX-IB code for steam line break and using the REMIX code for 0.05 ft$^2$small break LOCA. The stress distributions within the reactor vessel walls experiencing the pressure and the temperature transients were calculated using the OCA-P code for both events. The results of the analysis showed that a small break LOCA without decay heat presented the greatest challenge to the vessel, however, there is no crack initiation through end-of-life of the vessel with consideration of decay heat.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.29 no.12
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• pp.654-662
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• 2017

When emergency core cooling system (ECCS) is operated during loss of coolant accident (LOCA) in a pressurized water reactor (PWR), pressurized thermal shock (PTS) phenomenon can occur as cooling water is injected into a cold leg, mixed with hot primary coolant, and then entrained into a reactor vessel. Insufficient flow mixing may cause temperature stratification and steam condensation. In addition, flow vibration may cause thermal stresses in surrounding structures. This will reduce the life of the reactor vessel. Due to the importance of PTS phenomenon, in this study, calculation was performed for Test 1 among six types of OECD/NEA ROSA tests with ANSYS CFX R.17. Predicted results were then compared to measured data. Additionally, because temperature difference between the hot coolant at the inlet of the cold leg and the cold cooling water at the inlet of the ECCS injection line is 200 K or more, buoyancy force due to density difference might have significant effect on thermal-hydraulic characteristics of flow. Therefore, in this study, the necessity to include buoyancy force term in governing equations for accurate prediction of single phase thermal stratification in both cold legs and downcomer by ECCS injection was numerically studied.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2803-2811
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• 2020

With the aging of nuclear reactors, embrittlement of the reactor pressure vessel (RPV) steel, as a consequence of routine operations, is highly probable. To ensure operational integrity and safety, prediction and mitigation of compromising damage, brought on by pressurized thermal shock (PTS) following an emergency procedure, is of utmost importance. Computational fluid dynamics (CFD) codes can be employed to predict these events and have therefore been an acceptable method for such assessments. In this paper, CFD simulations of a density driven ECC state in the ROCOM facility are analyzed. Obtained numerical results are validated with the experimental measurements. Considerable attention is attributed to the boundary conditions and their influence, specifically outlet definitions, in order to determine and adequately replicate the non-active pumps in the facility. Consequent analyses focused on initial conditions as well as on the temporal discretization and inner iterations. Disparities due to different turbulent modelling approaches are investigated for standard RANS models. Based on observed trends for different cases, a definitive simulation setup has been established, results of which have been ultimately compared to the measurements.