• Journal of the Korean Society of Safety
• /
• v.24 no.5
• /
• pp.63-68
• /
• 2009

The urgent VAI method development is required since "The Act of Physical Protection and Radiological Emergency that is established in 2003" requires an evaluation of physical threats in nuclear facilities and an establishment of physical protection in Korea. The VAI methodology is developed to (1) make a sabotage model by reusing existing fire/flooding/pipe break PSA models, (2) calculate MCSs and TEPSs, (3) select the most cost-effective TEPS among many TEPSs, (4) determine the compartments in a selected TEPS as vital areas, and (5) provide protection measures to the vital areas. The developed VAI methodology contains four steps, (1) collecting the internal level 1 PSA model and information, (2) developing the fire/flood/pipe rupture model based on level 1 PSA model, (3) integrating the fire/flood/pipe rupture model into the sabotage model by JSTAR, and (4) calculating MCSs and TEPS. The VAT process is performed through the VIPEX that was developed in KAERI. This methodology serves as a guide to develop a sabotage model by using existing internal and external PSA models. When this methodology is used to identify the vital areas, it provides the most cost-effective method to save the VAI and physical protection costs.

• Journal of the Ergonomics Society of Korea
• /
• v.18 no.3
• /
• pp.41-53
• /
• 1999

More than twenty HRA (Human Reliability Analysis) methodologies have been developed and used for the safety analysis in nuclear field during the past two decades. However, no methodology appears to have universally been accepted, as various limitations have been raised for more widely used ones. One of the most important limitations of conventional HRA is insufficient analysis of the task structure and problem space. To resolve this problem, we suggest a framework of informational analysis for HRA. The proposed informational analysis consists of three parts. The first part is the scenario analysis that investigates the contextual information related to the given task on the basis of selected scenarios. The second is the goals-means analysis to define the relations between the cognitive goal and task steps. The third is the cognitive function analysis that identifies the cognitive patterns and information flows involved in the task. Through the three-part analysis. systematic investigation is made possible from the macroscopic information on the tasks to the microscopic information on the specific cognitive processes. It is expected that analysts can attain a structured set of information that helps to predict the types and possibility of human error in the given task.

• Nuclear Engineering and Technology
• /
• v.34 no.5
• /
• pp.421-432
• /
• 2002

As one of the advanced design features of the APR1400, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood phase of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, test results of direct ECC bypass performed in the steam-water test facility tailed MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) are presented. The test condition is determined, based on the preliminary analysis of TRAC code, by applying the ‘modified linear scaling method’with the l/4.93 length scale . From the tests, ECC direct bypass fraction, steam condensation rate and information on the flow distribution in the upper annulus downcomer region are obtained.

• Nuclear Engineering and Technology
• /
• v.44 no.8
• /
• pp.855-870
• /
• 2012

The paper deals with three main topics: a) the definition of System Thermal-Hydraulics (SYS TH), b) a historical outline for SYS TH and, c) the description of elements for reflection when planning research projects or improvement activities, this last topic being the main reason for the paper. Distinctions between basic thermal-hydraulics and computational Fluid-Dynamics (CFD) on the one side and SYS TH on the other side are considered under the first topic; stakeholders in the technology are identified. The proposal of Interim Acceptance Criteria for Emergency Core Cooling Systems in 1971 by US NRC (AEC at the time) is recognized as the starting date or the triggering event for SYS TH (second topic). The complex codes and the main experimental programs (list provided in the paper) constitute the pillars for SYS TH. Caution or warning statements are introduced in advance when discussing the third topic: a single person (or a researcher) has little to no possibility, or capability, of streamlining the forthcoming investments or to propose a roadmap for future activities. Nevertheless, the ambitious attempt to foresee developments in this area has been pursued without constraints connected with the availability of funds and with industrial benefits or interests. Demonstrating the acceptability of current SYS TH limitations and training in the application of those codes are mentioned as the main challenges for forthcoming research activities.

• Journal of Environmental Health Sciences
• /
• v.37 no.4
• /
• pp.237-249
• /
• 2011

The Chernobyl accident was a terrible catastrophe for humanity. Scientists are in concurrence about an increase of thyroid cancer incidence among children, but not among adults, because even areas less contaminated by radiation have also reported an increase in the incidence of thyroid cancer. In this case, the rise might be due to a screening effect. There is no convincing evidence that the incidence of leukemia and solid cancer has increased among the exposed populations, but it still remains a controversial issue. Additionally, apparent evidence of decreased fertility and increased hereditary effects have not been observed in the general population. WHO suggested 4,000 people could have died or may die in the future among emergency workers and residents of the most contaminated areas, while Greenpeace insists there will be 93,080 victims around the world. The radiation dose due to Chernobyl was mainly low, so if its health effects are to be found, more long-term and welldesigned research will be needed.

• Nuclear Engineering and Technology
• /
• v.32 no.6
• /
• pp.618-635
• /
• 2000

A total of 17 experimental data for the onset of slugging, which is assumed to be the precursor of the condensation-induced waterhammer (CIWH), have been obtained for various How rates of water Incorporating the most recent correlations of interfacial heat transfer and friction factor developed for a circular geometry and using an improved criterion of transition from stratified to a slug flow, two existing analytical models to predict lower and upper bounds for CIWH have been upgraded. Applicability of the present as well as existing CIWH models has been tested by comparison with two sets of CIWH data. The result of this comparison shows that the applicability of the present as well as existing models is reasonably good. Based on the present models for CIWH, a computer code entitled as“KAIST-CIWH”has been developed and sample guide charts to find CIWH free regions for a given combination of major flow parameters in a long horizontal pipe have been presented along with the results of parametric studies of major parameters (D, P, $T_{f,in}$, and L/D) on the critical inlet water flow rate（$W_{f,in}_crit$ for both lower and upper bounds. In addition, two simple formulas for lower and upper bounds that can be used in an emergency for quick results have been presented.

• Nuclear Engineering and Technology
• /
• v.28 no.3
• /
• pp.257-264
• /
• 1996

The conventional Equipment Environment Qualification (EEQ) envelope is developed based on the containment responses during the design basis events. The Safety Depressurization System (SDS) design without In-containment Refueling Water Storage Tank (IRWST) adopted in the Ulchin 3&4 challenges the conventional EEQ envelope during long term Feed and Bleed (F&B) operation due to the direct discharge of high mass and energy into the containment. Therefore, it is necessary to confirm that the containment pressure and temperature history during the long term F&B operation does not violate the conventional EEQ envelope. However, this subject has never been quantitatively assessed before. To investigate the success path of long term F&B operation this paper analyzes the thermal hydraulic response of the containment and Reactor Coolant System (RCS) until the completion of depressurization and cooldown of RCS into Shutdown Cooling System (SCS) entry condition. It is found that the SCS entry condition can be reached within 6 hours without violating the EEQ curve by proper operation of SDS valves, High Pressure Safety Injection (HPSI) pumps and active Containment Heat Removal System (CHRS). The suggested strategy not only demonstrates the feasibility of long term F&B operation but also can be utilized in the preparation of Emergency Procedure Guidelines (EPGs)

• Nuclear Engineering and Technology
• /
• v.30 no.2
• /
• pp.112-127
• /
• 1998

The present study is to assess the applicability of the best-estimate thermal-hydraulic codes, RELAP5/MOD3.2 and CATHARE2V1.3U, to the analysis of thermal-hydraulic behavior in PWRs during midloop operation following the loss of RHRS. The codes simulate an integral test, BETHSY 6.94, which was conducted in the large scale test facility of BETHSY in France. The test represents the accident where the loss of RHRS occurs during midloop operation with the pressurizer and upper head vents open and the sight level indicator broken. Besides, the hot legs are half filled with water and the upper parts of the primary cooling system are filled with nitrogen, with a letdown line open and only one SG available. The purposes of this study are to understand the physical phenomena associated with reflux cooling in the 5G U-tubes when noncondensable gas is present under low pressure and to assess the applicability of the codes to simulate the loss of RHRS event by comparing the predictions with the test results. The results of the study may contribute to actual applications for plant safety evaluation and description of the emergency operating procedure.

• Nuclear Engineering and Technology
• /
• v.39 no.4
• /
• pp.327-336
• /
• 2007

The influence of density differences on the mixing of the primary loop inventory and the Emergency Core Cooling (ECC) water in the downcomer of a Pressurised Water Reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. ROCOM is a 1:5 scaled model of a German PWR, and has been designed for coolant mixing studies. It is equipped with advanced instrumentation, which delivers high-resolution information for temperature or boron concentration fields. This paper presents a ROCOM experiment in which water with higher density was injected into a cold leg of the reactor model. Wire-mesh sensors measuring the tracer concentration were installed in the cold leg and upper and lower part of the downcomer. The experiment was run with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water especially for the validation of the Computational Fluid Dynamics (CFD) software ANSYS CFX. A mesh with two million control volumes was used for the calculations. The effects of turbulence on the mean flow were modelled with a Reynolds stress turbulence model. The results of the experiment and of the numerical calculations show that mixing is dominated by buoyancy effects: At higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this circumferential propagation. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.

• Nuclear Engineering and Technology
• /
• v.51 no.1
• /
• pp.303-309
• /
• 2019

Tritium is an important nuclide that must be monitored for radiation safety management. In this study, HTO was orally administered to rats at the level of 37 kBq ($1{\mu}Ci$) or 370 kBq ($10{\mu}Ci$) to examine tissue distribution and excretion levels. After sacrifice, wet and dry tissue samples were weighed and analyzed for tissue free-water tritium (TFWT) and organically bound tritium (OBT). The mean tissue concentrations of TFWT (OBT) were 30.9 (17.8) and 4.4 (8.1) Bq/g on days 7 and 13 at the 37 kBq level and 30.8 (64.6) Bq/g on day 17 at the 370 kBq level. To assess the cytogenetic damage due to tritium exposure, a cytokinesis-blocked micronucleus (MN) assay was performed in blood samples from rats exposed to HTO for 14 and 21 days after oral administration. There was no significant difference in the MN frequencies between the control and exposed rats.