When a sudden rupture occurs in high energy lines such as MSL (Main Steam Line) and safety injection line of nuclear power plants, ejection of inner fluid with high temperature and pressure causes blast wave, and may lead to secondary damage of adjacent major components and/or structures. The objective of this study is to assess integrity of containment wall and steam generator due to the blast wave under a postulated high energy line break condition at the MSL piping. In this context, a preliminary analysis was conducted to examine the blast wave simulation using coupled Eulerian-Lagrangian technique. Subsequently, a finite element analysis was carried out to assess integrity of the structures. As typical results, strain and stress values were calculated at the containment wall and steam generator, which did not exceed their failure criteria.

Since a containment serves as the last barrier of NPP (Nuclear Power Plant), high level of safety should be ensured to isolate and protect internal devices against external events such as earthquake, hurricane and aircraft crash. The objective of this study is to assess structural integrity of the reinforced concrete with liner plate of a containment subjected to the aircraft crash, and to examine its effects on an interior structure and major components. In this context, systematic FE (Finite Element) analyses combined with SPH (Smoothed-Particle Hydrodynamics) model for fuel were carried out by taking into account different crash locations and colliding velocities. Damage of the reinforced concrete containment due to crushing and cracking as well as stress and strain of reinforced structure were evaluated. Subsequently, stresses of major components as well as damage of reactor cavity were calculated. As typical results, connector was derived as critical location of the containment with regard to interior components, high stress values occurred at penetration pipes and head fittings, which did not exceed their yield and ultimate strengths respectively.

Creep behavior of nuclear materials is susceptible factor for structural integrity assessment. Particularly, RPV (Reactor Pressure Vessel) under severe accident with a core melting conditions may go through its wall penetration caused by direct contact with corium as well as additional unexpected failure by creep damage. Even though various experiments and analyses were conducted to resolve this special integrity issue, there are still many uncertainties which come from high temperature range, availabilities in only isothermal analysis and so on. In the present study, the state-of-the-art thermal, mechanical and creep properties of a representative RPV material, SA508 Gr. 3 C1. 1 were examined, with which correlation of creep in various loading conditions were determined. Subsequently, a user subroutine which defines the condition-dependent creep constitutive equation was developed and applied to RPV simulation, and relevant Larson-Miller Parameter was also newly fitted. Systematic numerical analyses were conducted by taking into account a typical accident scenario of external reactor vessel cooling. Finally, consequences of the creep effects such as the deformation shape, total strain and rupture time of vessel wall were compared with those of the existing study to enhance the accuracy of numerical analysis.

Steam explosion may occur in nuclear power plants by fuel-coolant interactions when the external reactor vessel cooling strategy is failed. This phenomenon can cause shock wave that endangers surrounding reactor cavity wall due to resulting dynamic effects. Even though extensive researches have been performed to predict influences of the steam explosion, due to complexity of physical phenomena and thermal-hydraulic conditions, it is remained as one of possible hazards. The object of this study is to examine characteristics of reactor cavity and nuclear components under representative steam explosion conditions. In this context, an assembled finite element mesh was generated and evaluated by the trinitrotoluene model. As a result, stresses, strains and displacements of the reactor cavity and nuclear components were calculated. Subsequently, crack evaluation of reinforced concrete was performed and their results were discussed.

The head penetration nozzles of control rod driving mechanisms (CRDMs) are susceptible components on primary water stress corrosion cracking (PWSCC) due to the dissimilar metal welds. The accurate integrity assessment of the CRDM head penetration nozzles is important for the safe operation of nuclear power plants. To resolve the integrity issue, conventional finite element methods, a cohesive zone model, and a virtual crack closure technique have been employed; however, there are still many uncertainties in accuracy and efficiency. In the present study, a specific Strain Rate Damage Model (SRDM) with stress and thermal dependent parameters was adopted to calculate crack initiation time. Also, a level set method, which defines the crack location based on the crack surface and vertical surface of crack tip, was considered to simulate arbitrary crack growth. By taking into account these two features, the eXtended Finite Element Method (XFEM) was implemented to simulate the PWSCC initiation and growth with a user subroutine code. Finally, the validity of the proposed method was evaluated by comparing the reference cracks that occurred in the CRDM head penetration nozzles.

The steam explosion is a fuel-coolant interaction process where the heat transfer from the melt to water is quite intense and rapid. This phenomenon may threat integrity of nuclear components and structures. For instance, the dynamic loads on the reactor cavity and the reactor lower plenum could potentially lead to failure of the main steam lines connected to the steam generators. In addition, since the main steam line extends to the containment wall, failure of the containment building may occur. The object of present study is to examine characteristics of Main Steam Line (MSL) and containment building under the steam explosion conditions. In this context, the corresponding FE models were generated and previously determined displacements of penetration piping were used as loading conditions. Subsequent FE analyses were conducted for the main steam line and containment building to calculate stresses and crack evaluations. Finally, structural assessment of nuclear component and structure combined with concrete failure criteria was performed and their results were discussed.

The worst case scenario including melt-down of reactor core components has been recognized as one of the safety issues. Without any counter measure, this may lead to penetration of RPV (Reactor Pressure Vessel) due to significant thermal loads caused by the relocation of molten core debris into the lower plenum. The present paper shows applicability of ERVC (External Reactor Vessel Cooling), aimed at limiting radiological releases by adopting the in-vessel retention concept, under core melting conditions. Systematic numerical analyses were carried out by taking into account diverse melting progress mechanisms, properties and amounts of molten debris, and damage models. Consequences of the thermal loads under assumed core melting conditions, such as the critical locations and time to reach penetration of the RPV lower plenum, were compared as a part of sensitivity analyses. Also, influences of each parameter were examined and technical findings from the assessment were discussed to demonstrate effectiveness of the ERVC for severe accident mitigation features of new power plants.

Lots of defects due to primary water stress corrosion cracking have been observed at diverse penetration locations of major components in nuclear power plants. To resolve this issue, during the past couple of decades, numerous analytical and experimental researches have been carried out to determine residual stress distributions in dissimilar metal welds that are plausible to crack initiation caused by the primary water stress corrosion. The present research deals with prediction of J-integrals for axial inner surface cracks in a mock-up that is to resemble upper head of reactor pressure vessel with penetrations of control rod driving mechanism. At first, representative residual stress distributions at J-groove welded region of the mock-up were calculated through finite element analyses of un-cracked models according to practical welding procedure. Then, J-integrals were calculated by mapping the residual stress distributions to the cracked models with different shapes. Based on these parametric analyses, effects of varying crack geometries were investigated of which are fully discussed in the manuscript.

For integrity assessment of structure containing crack, evaluation method based on fracture mechanics such as linear-elastic and elastic-plastic fracture mechanics has been relatively common method and becoming more widespread. However, it can be used only if the crack opening or tearing is occurred. If the crack exists on the piping components subjected to internal pressure, net-section stress is occurred in the direction on crack opening no matter where crack locate. On the contrary to this, if the external pressure is applied to piping components, net-section stress is occurred opposite direction and it is expected crack opening not to be occur. The subject of this study is SMART steam generator tube which is designed as helical geometrical feature and to be pressurized outside. Three dimensional finite element analyses are carried out to investigate crack behavior under external pressure considered various crack geometries and locations. Furthermore, the possibility of failure of SMART steam generator tube under design pressure is investigated.

Since small-sized specimens are widely used for fracture toughness tests to assure safety of a reactor pressure vessel in service, as a part of surveillance program, various geometry parameters affecting on the stress level near the crack-tip should be investigated for realistic assessment of cleavage fracture behavior. The aim of the present paper is to improve the current master curve method for typical miniature specimens, especially pre-cracked Charpy V-notched (PCVN) specimens. In this context, effects of thickness and side-grooves were quantified from comparing finite element (FE) analyses results in use of various PCVN specimens with and without side-grooves. Then, a scale factor to deal with geometry effects was suggested by employing the fracture toughness diagram, which was derived from FE analyses data of compact tension specimens and PCVN specimens. The scale factor was applied to calculate equivalent stress intensity factors influencing on the reference temperature embodied in the master curve method. The approach proposed in this paper will be useful to estimate fracture toughness of PCVN specimen made of SA508 carbon steel.

Adjacent multiple cracks can be found in power plant components such as pressure vessel, piping and so forth. If multiple cracks are detected, it can be affected significant effect to structure integrity due to high interaction effect of the multiple cracks. Therefore general fitness-for-service (FFS) codes and standards propose assessment method for the multiple cracks. The Section XI of ASME Code suggests crack coalescence method for multiple cracks. It employs the approach combining neighbored cracks as single crack if the distance between them is close. In some cases, since this combined rule can be considered as a conservative approach, more accurate investigation of the interaction effect between multiple cracks is needed. In this study, the interaction effect between two aligned coplanar circumferential surface cracks in elbows subjected to internal pressure was investigated. Since most previous studies dealt with only plates or straight pipes, the present research was centered on cracked elbow. FE (Finite element) limit analyses were carried out by changing elbow geometries and crack shapes. Also, applicability of the current criterion for the multiple cracks was discussed.

Since several defects caused by PWSCC (Primary Water Stress Corrosion Cracking) have been observed at the piping system of nuclear power plants, during the last two decades, lots of analytical and experimental researches have been performed to find out the relationship between the residual stress and PWSCC initiations. The present research deals with prediction of the residual stress at welding parts of CRDM (Control Rod Drive Mechanism). Especially, numerical investigations are carried out to support relevant experimental set-up because it is not easy to prepare large-sized mock-ups comparable to real geometry. First, preliminary examination was performed to establish an optimized welding process simulation procedure to accurately predict weld residual stresses. Then, detailed parametric FE analyses were carried out to examine effects of varying geometries and experimental conditions. Key findings were obtained from the FE analyses, which were used for finalizing the configuration of mock-up, are fully discussed in this manuscript.

Lots of investigations on failures of wall thinned piping have been carried out since the accident of Surry unit 2 in USA. From these preceding efforts, flow accelerated corrosion (FAC) which is a kind of wall thinning phenomenon is revealed main factor of failure of pipes in nuclear power plants. However, there are a few researches which directly take into account of flow characteristics and geometric changes for stress assessment of FAC-caused wall thinned piping. In this paper, structural integrity assessment employing a fluid-structure interaction (FSI) analysis scheme is performed on pipes representing secondary piping system of PWR which consists of straight pipes and elbows of various bend angles. Prior to the assessment, CFD analyses are conducted to predict plausible wall thinning location by considering flow and geometric parameters such as bend angle and radius of elbow. Then, for typical pipe geometry, detailed limit load analyses are performed to calculate maximum stress caused by turbulence and velocity of flow near the wall thinned part. Through these kinds of detailed parametric analyses, effects of FSI were observed, which should be considered for assessment of FAC-caused wall thinned piping.

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.

A system-integrated modular advanced reactor is being developed for multi purposes such as electricity production, sea water desalination and so on in Korea. While ASME Codes provide simplified design and operation procedures to determine allowable loadings for pressure retaining materials in components, the procedures are applicable when a temperature change rate associated with startup and shutdown is less than about 56°C/hr. If the procedures are applied to a rapid temperature change, results would be overly conservative. The objective of this research is to assess an applicability of the simplified design procedures to reactor coolant system of the integrated modular reactor with the change rates of 56°C/hr and 100°C/hr. To investigate effects of cooldown rate, heatup rate and surface crack location, systematic three-dimensional finite element analyses are carried out. The resulting pressure-temperature limit curves are compared with those obtained from the ASME Sec. XI operating procedure as well as Sec. III design procedure. Thereby, it was proven that the specific design features significantly affect the safe design region in the pressure-temperature limit curve to prevent a nonductile failure.

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.

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.

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 F en and individual F en . 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.

As the increase of operating times of petro-chemical facilities, possibility of flaw occurrences has been rapidly increased also. From a recent in-service inspection, three different sized surface cracks (crack depths of 4.5, 9.9 and 14.2 mm) were detected in welding parts of a spherical shell in Korea. While API579 code provides corresponding assessment procedures to determine crack driving forces, in the present work, numerical analyses are carried out for the cracked spherical shell to investigate effects of complex geometry, analysis model and residual stress. From detailed finite element analysis, stress distributions are determined for the full three-dimensional model. Then, stress intensity factors are calculated for simplified model by employing the well-known influence stress function technique. Finally, parametric structural integrity evaluation of the cracked spherical shell is conducted by using failure assessment diagram method. And the numerical results reflecting different geometry, analysis model and residual stress are fully discussed.

During the last decade, several defects due to primary water stress corrosion cracking (PWSCC) have been reported at bottom-mounted instrumentation (BMI) and control rod drive mechanism nozzles. The exact locations were dissimilar metal weld parts which are greatly important because the cracking could lead to leakage of primary coolant. The PWSCC of BMI mock-up’s penetration with dissimilar metal welds was examined by using doped steam test method by Korea Atomic Energy Research Institute. In this work, numerical analyses are performed for the same environment condition with the doped steam test. With respect to the numerical analyses, heat transfer analyses are carried out based on thermal conduction. The welding paths are simulated by using lumped path method for conservative evaluation and model change (remove/rebirth) method. Then residual stress analyses are conducted using the heat transfer analysis results, in which annealing effect of welding process simulation is considered for resetting the plastic deformation. However, the plastic behaviour of steels during phase transformations is not considered with experimental data. In addition, the consequence of weld residual stress that is known as the cause of PWSCC is being investigated.