Hae-Dong Chung

Estimation of fracture toughness transition curves of RPV steels from Charpy impact test data

Abstract The major concern for reactor pressure vessels (RPVs) in terms of integrity is the reduction in fracture toughness of materials due to radiation embrittlement. In order to ensure the structural integrity of RPVs, a very conservative approach has been employed since the first commercial operation of a nuclear power plant (NPP). RT NDT has been used as a principal parameter to indicate the degree of irradiated degradation in RPV material, which is determined using Charpy impact and drop weight tests based on the ASME code requirements. Charpy test is very practical and easy, but it does not provide the fracture toughness itself. Therefore, the Master Curve method, as a direct method to determine the fracture toughness of RPV, was investigated by a number of researchers during the last decade. An alternative approach is proposed in this paper to estimate the reference transition temperature, T 0 , in the Master Curve method using Charpy impact test data, which are abundant for old NPPs. Two well-known correlations between Charpy absorbed energy and K Ic were used to estimate the fracture toughness transition curves.

Development of elastic-plastic integrity evaluation system for pressure vessel and pipings

Abstract A practically useful system for elastic-plastic fracture mechanics analysis has been developed. The developed system is comprised of the J- integral/Tearing modulus (J/T) approach and the deformation plasticity failure assessment diagram (DPFAD) approach. The program contains analysis routines for five types of fracture mechanics specimens and five types of flaws in cylindrical geometries. A double and triple interpolation schemes were adopted to interpolate J values from the EPRI developed EPFM handbooks and a material property database was also developed. Several case studies were performed to evaluate the accuracy and the usefulness of the code. It was found that the J/T approach and the DPFAD approach yielded similar results. However, the DPFAD approach is more convenient for quick assessment of cracked structures while the J/T approach is more useful in evaluating the full history of the fracture process.

Evaluation of Representative Piping Systems Designed by Implicit Fatigue Concept

NUREG-1801 provides generic aging lessons learned to manage aging effects that may occur during continued operation beyond the design life of nuclear power plant. According to this report, the metal fatigue, among several age-related degradation mechanisms, is identified as one of time-limited aging analysis item. The objective of this paper is to introduce fatigue life evaluation of representative surge line and residual heat removal system piping which was designed by implicit fatigue concept. For the back-fitting evaluation employing explicit fatigue concept, detailed parametric CFD as well as FE analyses results are used. The well-known ASME Section III NB-3600 procedure is adopted for the metal fatigue and NUREG/CR-5704 procedure is further investigated to deal with additional environmental water effects. With regard to the environmental effect evaluation, two types of fatigue life correction factors are considered, such as maximum Fen and individual Fen . As a result, it was proven that a thermal stratification phenomenon is the governing factor in metal fatigue life of the surge line and strain rate is the most important parameter affecting the environmental fatigue life of both piping. The evaluation results will be used as technical bases for continued operation of OPR 1000 plant.© 2008 ASME

Integrity Evaluation of Korean Nuclear Reactor Pressure Vessel Under Pressurized Thermal Shock Conditions According to JEAC

In nuclear power plant, reactor pressure vessel (RPV) is the primary equipment that contains reactor cores and coolant. The RPV integrity should be evaluated in consideration with transient operation conditions and material deterioration. Especially, the pressurized thermal shock (PTS) has been considered as one of the most important issues regarding the RPV integrity since Rancho Seco nuclear power plant accident in1978.In this paper, integrity evaluation of Korean RPV was performed by using finite element analysis. PTS conditions like small break loss of coolant accident (SBLOCA) and Turkey Point steam line break (TP-SLB) were applied as loading conditions. Neutron fluence data of actual RPV operated over 30 years was used to determine fracture toughness of RPV material.The 3-dimensional finite element model including circumferential surface crack was generated for fracture mechanics analysis. The RPV integrity was evaluated according to Japan Electric Association Code (JEAC).Copyright

Regulatory Experience of Leak-Before-Break (LBB) Technology to the High Energy Piping Systems in Korea

The application of the leak-before-break (LBB) technology to the newly constructed pressurized water reactors (PWRs) has been approved for the several high energy piping systems inside containment in Korea. The main purpose of the LBB application for these systems at the design stage is the removal of the dynamic effects associated with the postulated double-ended guillotine break (DEGB) from design basis loads, as well as to the elimination of the pipe whip restraints and jet impingement barriers so as to increase the access the inspections. LBB technology is based on the low probability of pipe ruptures in the candidate piping systems using fracture mechanics and the insights from the state-of-the-art technology including operating experience. The procedures for LBB application is fundamentally based on the Unite States Nuclear Regulatory Commission (US NRC) requirements as detailed in the standard review plan (SRP) 3.6.3. However, a number of the additional requirements and issues are not specified in the review procedure during regulatory review were imposed and addressed during the review process. The regulatory review is focused on the confirmation on the methods for the elements in the screening criteria and several technical concerns on the determination of material properties, the validation of crack evaluation methods and leak rate estimation in the LBB evaluation considering the adequate margin. Although the application of the LBB has been approved by the safety authority for some high energy systems, the validation of LBB is continuously maintained in consideration of operating experience. In this paper, the regulatory positions for LBB application are described for the areas of screening criteria, leak rate estimation including the capability of leak detection system, material properties, load combination, crack stability methods, and margins in the crack stability evaluations. The issues encountered during the regulatory review such as the dynamic fracture test to consider the dynamic strain aging (DSA) of carbon and low alloy steel, thermal stratification and striping in the pressurizer surge line, water/steam hammer in main steam lines, and estimation of the crack opening area at the pipe-to-nozzle interface considering the asymmetry are also introduced. In addition, several regulatory actions to improve the reliability in the capability of leak detection systems and to clarify the screening criteria such as the corrosion resistance is provided.Copyright

Numerical Analyses to Simulate Thermal Stratification Phenomenon in a Piping System

In some portions of nuclear piping systems, stratification phenomena may occur due to the density difference between hot and cold stream. When the temperature difference is large, the stratified flow under diverse operating conditions can produce high thermal stress, which leads to unanticipated piping integrity issues. The objectives of this research are to examine controvertible numerical factors such as model size, grid resolution, turbulent parameters, governing equation, inflow direction and pipe wall. Parametric threedimensional computational fluid dynamics analyses were carried out to quantify effects of these parameters on the accuracy of temperature profiles in a typical nuclear piping with complex geometries. Then, as a key finding, it was recommended to use optimized mesh of real piping with the conjugated heat transfer condition for accurate thermal stratification analyses.

A Residual Stress Analysis of a Corner Crack in Pressurizer Vent Nozzle Penetration Weld

There are some ferritic low alloy steel components including dissimilar metal welding parts in a nuclear power plant. Residual stress induced by welding process is one of the factors that may lead a sound component to have a defect. Therefore it is necessary that the distribution of residual stress is obtained to predict the behavior of a dissimilar metal welding part. In this paper, the distribution of residual stress obtained by finite element analysis is investigated to assess the stress intensity factor of a corner crack in pressurizer vent nozzle penetration weld. Then the stress intensity factor and plastic zone correction of a corner crack are calculated under internal pressure, thermal stress and residual stress in accordance with Electric Power Research Institute equation [1]. The resulting stress intensity factor and plastic-zone correction were compared with those obtained from Structural Integrity Associates.Copyright

Evaluation of Thermal Stratification and Primary Water Environment Effects on Fatigue Life of Austenitic Piping

Abstract During the last two decades, lots of efforts have been devoted to resolve thermal stratification phenomenon and primary water environment issues. While several effective methods were proposed especially in related to thermally stratified flow analyses and corrosive material resistance experiments, however, lack of details on specific stress and fatigue evaluation make it difficult to quantify structural behaviors. In the present work, effects of the thermal stratification and primary water are numerically examined from a structural integrity point of view. First, a representative austenitic nuclear piping is selected and its stress components at critical locations are calculated in use of four stratified temperature inputs and eight transient conditions. Subsequently, both metal and environmental fatigue usage factors of the piping are determined by manipulating the stress components in accordance with NUREG/CR-5704 as well as ASME B&PV Codes. Key findings from the fatigue evaluation with applicability of pipe and three-dimensional solid finite elements are fully discussed and a recommendation for realistic evaluation is suggested. 기호설명

Numerical Analyses of Surge Line Piping to Assess Thermal Stratification Phenomenon

During the last two decades, thermal stratification has been issued as a critical problem in the nuclear power industry. Since the problem caused by this phenomenon also became important in Korea, it is necessary to quantify the thermal stratification effect to ensure the safety of the piping system. In this paper, detailed stress analyses of the surge line, considering the thermal stratification, are conducted. Parametric sensitivity analyses to find out an optimum model were carried out using pipe element models and full 3-D element models. For instance, in case of the pipe element model, the effect of starting location of thermal stratification and boundary condition were investigated. And, in case of the 3-D solid element model, the effect of boundary condition and thermal loading condition were assessed. The stress analysis results showed that the thermal stratification phenomenon significantly affected the integrity of the surge line piping. Also, establishment of insurge and outsurge conditions was derived as one of the further investigations.Copyright

Parametric CFD Analyses to Simulate Stratified Flows

CFD analysis is widely adopted for determination of design characteristics of major equipment with flexible geometry and flow. However, the accuracy of CFD analysis strongly depends on the numerical model, governing equation and simulation parameters. A thermal stratification phenomenon can lead to unanticipated damages of nuclear piping because of the stresses caused by different fluid densities due to stratified flow. In this paper, systematic CFD analyses are performed for surge line which is one of primary piping system by using representative commercial code to investigate key parameters; (1) mesh size and time step effect (2) turbulence parameter effect (3) material property approximation effect (4) conjugate heat transfer effect (5) insurge and outsurge flow effect. From numerical analysis results related to the items (1) through (3), the optimum CFD model as well as reasonable input parameters was determined. With regard to the item (4), thermal difference was bigger as 82∼208% than without considering conjugate heat transfer. On the other hand, for the item (5), stratified flows were come out clearer in outsurge flow. Based on the parametric CFD analyses to simulate stratified flows, most of numerical issues were resolved while further investigation is required for the conjugate heat transfer effect.Copyright