Hho Jung Kim

Round robin analysis for probabilistic structural integrity of reactor pressure vessel under pressurized thermal shock

Performed here is a comparative assessment study for the probabilistic fracture mechanics approach of the pressurized thermal shock of the reactor pressure vessel A round robin consisting of one prerequisite deterministic study and five cases for probabilistic approaches is proposed, and all organizations interested are invited The problems are solved by the paiticipants and their results are compared to issue some recommendation of best practices and to assure an understanding of the key parameters in this type of approach, like transient description and frequency, material properties, defect type and distribution, fracture mechanics methodology etc, which will be useful in the justification through a probabilistic approach for the case of a plant over-passing the screening criteria Six participants from 3 organizations responded to the problem and their results are compiled and analyzed in this study

Safety evaluation of socket weld integrity in nuclear piping

The purposes of this paper are to evaluate the integrity of socket weld in nuclear piping and prepare the technical basis for a new guideline on radiographic testing (RT) for the socket weld. Recently, the integrity of the socket weld is regarded as a safety concern in nuclear power plants because lots of failures and leaks have been reported in the socket weld. The root causes of the socket weld failure are known as unanticipated loadings such as vibration or thermal fatigue and improper weld joint during construction. The ASME Code Sec. III requires 1/16 inch gap between the pipe and fitting in the socket weld. Many failure cases, however, showed that the gap requirement was not satisfied. The Code also requires magnetic particle examination (MT) or liquid penetration examination (PT) on the socket weld, but not radiographic examination (RT). It means that it is not easy to examine the 1/16 inch gap in the socket weld by using the NDE methods currently required in the Code. In this paper, the effects of the requirements in the ASME Code Sec. III on the socket weld integrity were evaluated by using finite element method. The crack behavior in the socket weld was also investigated under vibration event in nuclear power plants. The results showed that the socket weld was very susceptible to the vibration if the requirements in ASME Code were not satisfied. The constraint between the pipe and fitting due to the contact significantly affects the integrity of the socket weld. This paper also suggests a new guideline on the RT for the socket weld during construction stage in nuclear power plants.

Plant Behavior Following a Loss-of-Residual-Heat-Removal Event Under a Shutdown Condition

The potential of the RELAP5/MOD3.2 code was assessed for a loss-of-residual-heat-removal (RHR) event during midloop operation, and the predictability of major thermal-hydraulic phenomena was evaluated for the long-term transient. The calculations were compared for two cases of experiments conducted at the Rig of Safety Assessment-IV (ROSA-IV)/Large-Scale Test Facility (LSTF) in Japan: the cold-leg-opening and the pressurizer-manway-opening cases. In addition, the real plant responses to the event were evaluated for Yong Gwang nuclear power plant Units 3 and 4 (YGN 3/4) in Korea, especially concerning the mitigation capability to remove the decay heat through the steam generators (SGs). From the LSTF simulation, it was found that the RELAP5 code was capable of simulating the plant behavior following the loss-of-RHR event under a shutdown condition. As a result, the thermal-hydraulic transport process including noncondensable gas behavior was reasonably predicted with an appropriate time step and CPU time, and the major thermal-hydraulic phenomena agreed well with the experiment. However, there were some code deficiencies such as an estimation of large system mass errors for the long transient and severe flow oscillations in the core region. These should be improved for more accurate and reliable calculation. In the YGN 3/4 simulation, the water-filled SG case delayed the coolant discharge to containment by ∼2 h and the core heatup by ∼1.3 h, as compared to the emptied-SG case, because of reduction of the pressurization rate that resulted from condensation on the SG U-tube wall. For the water-filled SGs, the amount ofheat transfer into the secondary side was estimated at more than 60% of the total core power throughout the transient.

Mitigation Measures Following a Loss-of-Residual-Heat-Removal Event During Shutdown

The transient following a loss-of-residual-heat-removal event during shutdown was analyzed to determine the containment closure time (CCT) to prevent uncontrolled release of fission products and the gravity-injection path and rate (GIPR) for effective core cooling using the RELAP5/MOD3.2 code. The plant conditions of Yonggwang Units 3 and 4, a pressurized water reactor (PWR) of 2815-MW(thermal) power in Korea, were reviewed, and possible event sequences were identified. From the CCT analysis for the five cases of typical plant configurations, it was estimated for the earliest CCT to be 40 min after the event in a case with a large cold-leg opening and emptied steam generators (SGs). However, the case with water-filled SGs significantly delayed the CCT through the heat removal to the secondary side. From the GIPR analysis for the six possible gravity-injection paths from the refueling water storage tank (RWST), the case with the injection point and opening on the other leg side was estimated to be the most suitable path to avoid core boiling. In addition, from the sensitivity study, it was evaluated for the plant to be capable of providing the core cooling for the long-term transient if nominal RWST water is available. As a result, these analysis methods and results will provide useful information in understanding the plant behavior and preparing the mitigation measures after the event, especially for Combustion Engineering-type PWR plants. However, to directly apply the analysis results to the emergency procedure for such an event, additional case studies are needed for a wide range of operating conditions such as reactor coolant inventory, RWST water temperature, and core decay heat rate.

RELAP5 Prediction of Transient Tests in the RD-14 Test Facility

Abstract Although the RELAP5 computer code has been developed for best-estimate transient simulation of a pressurized water reactor and its associated systems, it could not assess the thermal-hydraulic behavior of a Canada deuterium uranium (CANDU) reactor adequately. However, some studies have been initiated to explore the applicability for simulating a large-break loss-of-coolant accident in CANDU reactors. In the present study, the small-reactor inlet header break test and the steam generator secondary-side depressurization test conducted in the RD-14 test facility were simulated with the RELAP5/MOD3.2.2 code to examine its extended capability for all the postulated transients and accidents in CANDU reactors. The results were compared with experimental data and those of the CATHENA code performed by Atomic Energy of Canada Limited. In the RELAP5 analyses, the heated sections in the facility were simulated as a multichannel with five pipe models, which have identical flow areas and hydraulic elevations, as well as a single-pipe model. The results of the small-reactor inlet header break and the steam generator secondary-side depressurization simulations predicted experimental data reasonably well. However, some discrepancies in the depressurization of the primary heat transport system after the header break and consequent time delay of the major phenomena were observed in the simulation of the small-reactor inlet header break test.

Root Cause Analysis of SI Nozzle Thermal Sleeve Breakaway Failures Occurring at PWR Plants

At conventional pressurized water reactors (PWRs), cold water stored in the refueling water tank of emergency core cooling system is injected into the primary coolant system through a safety injection (SI) line, which is connected to each cold leg pipe between the main coolant pump and the reactor vessel during the SI operation, which begins on the receipt of a loss of coolant accident signal. In normal reactor power operation mode, the wall of SI line nozzle maintains at high temperature because it is the junction part connected to the cold leg pipe through which the hot main coolant flows. To prevent and relieve excessive transient thermal stress in the nozzle wall, which may be caused by the direct contact of cold water in the SI operation mode, a thermal sleeve in the shape of thin wall cylinder is set in the nozzle part of each SI line. Recently, mechanical failures that the sleeves are separated from the SI branch pipe and fall into the connected cold leg main pipe occurred in sequence at some typical PWR plants in Korea. To find out the root cause of thermal sleeve breakaway failures, the flow situation in the junction of primary coolant main pipe-SI branch pipe and the vibration modal characteristics of the thermal sleeve are investigated in detail by using both computational fluid dynamics code and structure analysis finite element code. As a result, the transient response in fluid pressure exerting on the local part of thermal sleeve wall surface to the primary coolant flow through the pipe junction area during the normal reactor operation mode shows oscillatory characteristics with the frequencies ranging from 15 Hz to 18 Hz. These frequencies coincide with the lower mode natural frequencies of thermal sleeve, which has a pinned support condition on the outer surface with the circumferential prominence set into the circumferential groove on the inner surface of SI nozzle at the midheight of thermal sleeve. In addition, the variation of pressure on the thermal sleeve surface yields alternating forces and torques in the directions of two rectangular axes perpendicular to the longitudinal axis of cylindrical thermal sleeve, which causes both rolling and pitching motions of the thermal sleeve. Consequently, it is seen that this flow situation surrounding the thermal sleeve during the normal reactor operation can induce resonant vibrations accompanying the shaking motion of the thermal sleeve at the pinned support condition, which finally leads to the failures of thermal sleeve breakaway from the SI nozzle.

Aging Management of Nuclear Power Plants in Korea

Kori Unit 1, which is the oldest nuclear power plant (NPP) in Korea has been operated since 1978. In addition, 10 other NPPs have been operating more than 10 years. As the number of aging plants rise, public concern over the safety of operating NPPs has increased. Periodic safety review (PSR) in addition to the existing safety assessments are proposed by IAEA as an effective way to verify that operating NPPs maintain the high level of safety. In this regard, the Ministry of Science and Technology (MOST), Korea’s nuclear regulatory body, recently established an institutional process through revision to the atomic energy act to introduce PSR. This PSR considers, among other factors, improvements in safety standards and practices, the cumulative effects of plant aging, operating experience, and the evolution of science and technology. In particular, the assessment and management of plant aging is one of the major areas. It includes identification of the system, structure and components (SSCs) for aging management, assessment of aging effects and planning of aging management implementation program. PSR results could be one of the procedural requirements that are utilized to renew an operating license of a NPP. This paper describes safety assessment requirements including PSR and aging management activities in Korea. This paper also includes the strategy and method for the application of PSR results to the aging management and continued operation of NPPs.

Root Cause Analysis of SI Nozzle Thermal Sleeve Breakaway Failures Occurred at PWR Plants

Thermal sleeves in the shape of thin wall cylinder seated inside the nozzle part of each safety injection (SI) line at pressurized water reactors (PWRs) have such functions as prevention and relief of potential excessive transient thermal stress in the wall of SI line nozzle part which is initially heated up with hot water flowing in the primary coolant piping system when cold water is injected into the system through the SI nozzles during the SI operation. Recently, mechanical failures that the sleeves are separated from the SI branch pipe and fall into the connected cold leg main pipe occurred in sequence at some typical PWR plants in Korea. To find out the root cause of thermal sleeve breakaway failures, the flow situation in the in the junction of primary coolant main pipe and SI branch pipe and the vibration modal characteristics of the thermal sleeve are investigated in details by using both computational fluid dynamic (CFD) code and structure analysis finite element code. As the results, the transient response in fluid pressure exerting on the local part of thermal sleeve wall surface to the primary coolant flow through the pipe junction area during the normal reactor operation mode shows oscillatory characteristics with the frequencies ranging from 15 to 18, which coincide with the lower mode natural frequencies of thermal sleeve having a pinned support condition on the circumferential prominence on the outer surface of thermal sleeve which is put into the circumferential groove on the inner surface of SI nozzle at the mid-height of thermal sleeve. In addition, the variation of pressure on the thermal sleeve surface yield alternating forces and torques in the directions of two rectangular axes perpendicular to the longitudinal axis of cylindrical thermal sleeve, which causes both rolling and pitching motions of the thermal sleeve. Consequently, it is seen that this flow situation surrounding the thermal sleeve during the normal reactor operation can induce resonant vibrations accompanying the shaking motion of the thermal sleeve at the pinned support condition, which finally leads to the failures of thermal sleeve breakaway from the SI nozzle.© 2006 ASME

Flow and Modal Analysis for the Investigation of PWR Safety Injection Line-Installed Thermal Sleeve Separation Failure Mechanism

In conventional pressurized water reactors, a thermal sleeve (named simply ‘sleeve’ hereafter) is seated inside the nozzle part of each safety injection (SI) branch pipe to prevent and relieve potential excessive transient thermal stress in the nozzle wall when cold water is injected during the safety injection mode. Recently, mechanical failures that the sleeves are separated from the SI branch pipe and fall into the connected cold leg main pipe occurred in sequence at some typical PWR plants in Korea. This paper investigates the flow field in the pipe junction through a numerical simulation and vibration characteristics of thermal sleeves through a modal analysis to analyze the root cause of sleeve separation mechanism. By performing both flow simulation in the SI pipe junction and modal analysis of thermal sleeve, the fluid force and modal characteristics have been identified to be able to lead or contribute to separate thermal sleeves inside safety injection branch pipes in PWR plants from their original seating locations.Copyright

Fretting-wear Characteristics of Steam Generator Helical Tubes

This study investigates the safety assessment of the potential for fretting-wear damages caused by foreign object in operating nuclear power plants. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for the helical type tubes with various conditions. The wear rate of helical type tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the vibration of the tube on the remaining life of the tube. In addition, addressed is the effect of the external pressure on the vibration and fretting-wear characteristics of the tube.