• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3236-3246
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• 2021

Lessons learned from the Fukushima Daiichi nuclear power plant accident directed that multiple failures should be considered more seriously rather than single failure in the licensing bases and safety cases because attempts to take accident management measures could be unsuccessful under the high radiation environment aggravated by multiple failures, such as complete loss of electric power, uncontrollable loss of coolant inventory, failure of essential safety function recovery. In the case of the complete loss of electric power called station blackout (SBO), if there is no mitigation action for recovering safety functions, the reactor core would be overheated, and severe fuel damage could be anticipated due to the failure of the active heat sink. In such a transient condition at CANDU-6 plants, the seal failure of the primary heat transport (PHT) pumps can facilitate a consequent increase in the fuel sheath temperature and eventually lead to degradation of the fuel integrity. Therefore, it is necessary to specify the regulatory guidelines for multiple failures on a licensing basis so that licensees should prepare the accident management measures to prevent or mitigate accident conditions. In order to explore the efficiency of implementing accident management strategies for CANDU-6 plants, this study proposed a realistic accident analysis approach on the SBO transient with multiple-failure sequences such as seal failure of PHT pumps without operator's recovery actions. In this regard, a comparative study for two PHT pump seal failure modes with and without coolant seal leakage was conducted using a best-estimate code to precisely investigate the behaviors of thermal-hydraulic parameters during transient conditions. Moreover, a sensitivity analysis for different PHT pump seal leakage rates was also carried out to examine the effect of leakage rate on the system responses. This study is expected to provide the technical bases to the accident management strategy for unmitigated transient conditions with multiple failures.

• 한국안전학회지
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• 제35권4호
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• 한국안전학회지
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• 제37권5호
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• pp.80-88
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• 2022

The Korean nuclear industry had developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code, which adopts a two-fluid, three-field model that is comprised of gas, continuous liquid and droplet fields and has the capability to simulate three-dimensional models. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for the accident management plan of a nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification is required for the separate and integral effect experiments. Therefore, the goal of this work is to verify the calculation capability of the SPACE code for multiple failure accidents. For this purpose, an experiment was conducted to simulate a Control Element Drive Mechanism (CEDM) break with a safety injection failure using the ATLAS test facility, which is operated by Korea Atomic Energy Research Institute (KAERI). This experiment focused on the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The results of the overall system transient response using the SPACE code showed similar trends with the experimental results for parameters such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it can be concluded that the SPACE code has sufficient capability to simulate a CEDM break with a safety injection failure accident.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1210-1216
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• 2018

Site or multi-unit (MU) risk assessment has been a major issue in the field of nuclear safety study since the Fukushima accident in 2011. There have been few methods or experiences for MU risk assessment because the Fukushima accident was the first real MU accident and before the accident, there was little expectation of the possibility that an MU accident will occur. In addition to the lack of experience of MU risk assessment, since an MU nuclear power plant site is usually very complex to analyze as a whole, it was considered that a systematic method such as probabilistic safety assessment (PSA) is difficult to apply to MU risk assessment. This paper proposes a new MU risk assessment methodology by using the conventional PSA methodology which is widely used in nuclear power plant risk assessment. The logical failure structure of a site with multiple units is suggested from the definition of site risk, and a decomposition method is applied to identify specific MU failure scenarios.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.469-476
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• 2013

In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost? In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.474-483
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• 2021

To establish the exclusive role of hydrogen on burst behaviour of Zircaloy-4 during loss-of-coolant accident transients, an extensive single-rod burst tests were conducted on both unirradiated as-received and hydrogenated Zircaloy-4 cladding tubes at different heating rates and internal overpressures. The visual observations of cladding tubes during bursting as well as post-burst are presented in detail to understand the effect of hydrogen concentration, heating rate, and internal pressure. Impact of hydrogen on burst parameters-burst stress, burst strain, burst temperature-during loss-of-coolant accident transients are compared and discussed. Rupture at multiple locations for hydrogenated cladding at lower internal pressure and higher heating rate is reported for the very first time. A novel burst criterion accounting hydrogen concentration in nuclear fuel cladding is proposed.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• 대한조선학회논문집
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• 제59권3호
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• pp.164-172
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• 2022

A HAZID is a brainstorming workshop to identify hazards in an early phase of a project. It should be flexible to capture all probable accidents allowing experienced participants to exploit their expertise and experiences. A bowtie analysis is a graphical representation of major accident hazards elaborating safety measures i.e. barriers. The result of these workshops should be documented in an organized manner to share as good as possible details of the discussion through the lifetime of the project. Currently results are documented using a three-step representation of an accident; causes, top event and consequences, which cannot capture correctly sequence of events leading to various accidents and roles of barrier between two events. Another problem is that barriers would be shown repeatedly leading to a misunderstanding that there are an enough number of safety measures. A new bowtie analysis method is proposed to describe an accident in multiple steps showing relations among causes or consequences. With causes and consequences shown in a format of a tree, the frequencies of having the top event (Fault tree analysis) and various consequences (Event tree analysis) are evaluated automatically based on the frequency of initiating causes and the probabilities of failure of barriers. It will provide a good description of the accident scenario and help the risk to be assessed transparently.

• Nuclear Engineering and Technology
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• 제38권3호
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• pp.231-238
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• 2006

Whole-body exposure to high-dose radiation causes injury involving multiple organs that depends on their sensitivity to radiation. This acute radiation syndrome (ARS) is caused by a brief exposure of a major part of the body to radiation at a relatively high dose rate. ARS is characterized by an initial prodromal stage, a latent symptom-free period, a critical or manifestation phase that usually takes one of four forms (three forms): hematologic, gastrointestinal, or cardiovascular and neurological (neurovascular), depending upon the exposure dose, and a recovery phase or death. One of the most important factors in treating victims exposed to radiation is the estimation of the exposure dose. When high-dose exposure is considered, initial dose estimation must be performed in order to make strategy decisions for treatment as soon as possible. Dose estimation can be based on onset and severity of prodromal symptoms, decline in absolute lymphocyte count post exposure, and chromosomal analysis of peripheral blood lymphocytes. Moreover, dose assessment on the basis of calculation from reconstruction of the radiation event may be required. Experience of a criticality accident occurring in 1999 at Tokai-mura, Japan, showed that ARS led to multiple organ failure (MOF). This article will review ARS and discuss the possible mechanisms of MOF developing from ARS.

• Journal of Chest Surgery
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• 제28권9호
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• pp.847-850
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• 1995

For the purpose of evaluation of clinical characteristics in multiple rib fracture due to accident , 24 cases treated by surgical rib fixation using Judet`s strut for multiple rib fracture and flail chestduring the period from June 1993 to October 1994 were reviewed. There were 17 males and 7 females.They ranged in age from 19 years old to 56 years old. The causes of rib fracture were traffic accident in 18 cases, fall down in 3 cases, compression in 2 cases stab wound in 1 case. The number of rib fracture were five in 7 cases, six in 5 cases, four in 5 cases, three in 3 cases. Associated intrathoracic injuries were hemopneumothorax in 12 cases, hemothorax in 10 cases, lung laceration or hemorrhagic contusion in 7 cases. Associated extrathoracic injuries were abdominal injuries in 21 cases, orthopedic problem in 7 cases, head trauma in 4 cases. The most common fractured site was posterolateral portion of the ribs. The causes for operation were flail chest, severe rib displacement and pain, hemothorax or hemopneumothorax with continuous air leakage and stab wound. There were 6 postoperative complications ; one with hydrothorax, two with fibrothorax, two with wound infection and one case of death due to multiful organ failure. Postoperatively, all patients became comfortable and complained less painful. Twenty patients restored spontaneous breathing without ventilator support, three patients were ventilated during a day and one patient expired after 2 days. There were no morbidity and mortality related to operation.