• Nuclear Engineering and Technology
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• v.40 no.3
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• pp.199-212
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• 2008

A small-break loss of coolant accident (SB-LOCA) test with a break size equivalent to a 3-inch cold leg break of the APR1400 was carried out as the first transient integral effect test using the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). This was the first integral effect test to investigate the integral performance of the test facility and to verify its simulation capability for one of the design-basis accidents. Reasonably good thermal hydraulic data was obtained so that an integral performance of the fluid sub-systems was identified and control performance of the ATLAS was confirmed under real thermal hydraulic conditions. Based on the measured data, a post-test calculation was carried out using the best-estimate thermal hydraulic safety analysis code, MARS 3.1, and the similarity between the expected and actual data was investigated. On the whole, the post-test calculation reasonably predicts the major thermal hydraulic parameters measured during the SB-LOCA test. The obtained data will be used to enhance the simulation capability of the ATLAS and to improve an input model of the ATLAS for simulation of other target scenarios.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.12
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• pp.3748-3754
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• 2009

An integral head loss test in a test apparatus was conducted to simulate chemical effects on a head loss across a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). The test was conducted during 30 days in the condition of a short spray, a long spray, and no materials with chemical effects. The result exhibited that the head loss was affected on amounts of the exposed materials according to spray conditions. XRD analysis of the collected precipitates showed that the precipitates were phosphate compounds. Comparison of the head loss with dissolved species concentration showed that high increase rate of the head loss resulted from the corrosion of aluminum and zinc but slow increase rate of the head loss resulted from the precipitates induced by Si, Mg, and Ca from leaching reaction at NUKON and concrete after passivation of metal specimens.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.11
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• pp.3260-3268
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• 2009

A test apparatus has been fabricated to simulate chemical effect on head loss through a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). Tests were conducted under condition of same ratio of strainer surface area to water volume between the test appratus and the containment sump. A series of tests have been performed to investigate the effects of spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the test screen is strongly affected by spray duration and is increased rapidly at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKONTM and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

• Nuclear Engineering and Technology
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• v.15 no.3
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• pp.179-187
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• 1983

The pH of containment spray and sump water following a LOCA for KNU 5'||'&'||'6 and KNU 1 was calculated to see if pH design criteria of containment spray system established by USNRC were met. The pH calculations have been made for the two cases; maximum pH and minimum pH. For KNU 5'||'&'||'6, results showed that long term sump pH values calculated for the maximum pH and minimum pH case well met the pH requirement of at least 8.5 and spray pH for the maximum case slightly exceeded the range of design criteria (8.5 to 11.0). For KNU 1, pH requirement of long term sump pH was also met, however, spray pH value for the maximum pH case was very largely greater than that of current pH requirement. (No pH requirement of containment spray water has been established at the time of designing KNU 1) In order to find the design parameters of containment spray system which are expected to meet the spray pH requirement, several calculations were wade, by changing the input parameters to "LCCAPH". Finally, it was shown that the boric acid concentration in RWST (refueling water storage tank), which was the primary sources of containment spray water during injection mode, be maintained the range of 2750 ppm to 2850 ppm, or tile flow rate of NaOH added to spray water he kept between 10 gpm to 24 gpm.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.229-239
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• 1997

To investigate the realistic behavior of mass and energy release and resultant containment response during large break Loss of Coolant accident (LOCA), analyses are performed for Yonggwang (YGN) 3&4 nuclear power plants by using a merged version of RELAP5/CONTEMPT4 computer code. Comparative analyses by using conservative design computer codes are also peformed. The break types analyzed are the double-ended guillotine breaks at the cold leg and hot leg. The design analysis resulted in containment peak pressure during post-blowdown phase for the cold leg break. However, the RELAP5/CONTEMPT4 analyses show that the containment pressure has a peak during blowdown phase, thereafter it decreases monotonously without the second port-blowdown peak. For the hot leg break, revised design analysis shows much lower pressure than that reported in YGN 3&4 final safety analysis report. The RELAP5/CONTEMPT4 analysis shoos similar trend and confirmed that the bypass flow through the broken loop steam generator during post-blowdown is negligibly small compared to that of cold leg break. The low pressure and temperature predicted tv realistic analysis presented in this paper suggest that the design analysis methodology contains substantial margin and it can be improved to provide benefit in investment protection, such as, relaxing plant technical specifications and reducing containment design pressure.

• Journal of Radiation Industry
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• v.17 no.3
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• pp.265-273
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• 2023

To effectively and safely manage the radiation exposure to nuclear power plant (NPP) workers in accidents, major overseas NPP operators such as the United States, Germany, and France have developed and applied realistic 3D model radiation dose assessment software for workers. Continuous research and development have recently been conducted, such as performing NPP accident management using 3D-VR based on As Low As Reasonably Achievable (ALARA) planning tool. In line with this global trend, it is also required to secure technology to manage radiation exposure of workers in Korea efficiently. Therefore, in this paper, it is described the application method and assessment results of radiation exposure scenarios for workers in response to accidents assessment technology, which is one of the fundamental technologies for constructing a realistic platform to be utilized for radiation exposure prediction, diagnosis, management, and training simulations following accidents. First, the post-accident sampling after the Loss of Coolant Accident(LOCA) was selected as the accident and response scenario, and the assessment area related to this work was established. Subsequently, the structures within the assessment area were modeled using MCNP, and the radiation source of the equipment was inputted. Based on this, the radiation dose distribution in the assessment area was assessed. Afterward, considering the three principles of external radiation protection (time, distance, and shielding) detailed work scenarios were developed by varying the number of workers, the presence or absence of a shield, and the location of the shield. The radiation exposure doses received by workers were compared and analyzed for each scenario, and based on the results, the optimal accident response scenario was derived. The results of this study plan to be utilized as a fundamental technology to ensure the safety of workers through simulations targeting various reactor types and accident response scenarios in the future. Furthermore, it is expected to secure the possibility of developing a data-based ALARA decision support system for predicting radiation exposure dose at NPP sites.

• Nuclear Engineering and Technology
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• v.53 no.2
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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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• v.52 no.9
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• pp.1926-1938
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• 2020

We validated the performance of RELAP MOD3.3 code regarding the hybrid SIT with available experimental data. The concept of the hybrid SIT is to connect the pressurizer to SIT to utilize the water inside SIT in the case of SBO or SB-LOCA combined with TLOFW. We investigated how well RELAP5 code predicts the physical phenomena in terms of the equilibrium time, stratification, condensation against Separate Effect Test (SET) data. We also conducted the validation of RELAP5 code against Integrated Effect Test (IET) experimental data produced by the ATLAS facility. We followed conventional approach for code validation of IET data, which are pre-test and post-test calculation. RELAP5 code shows substantial difference with changing number of nodes. The increase of the number of nodes tends to reduce the condensation rate at the interface between liquid and vapor inside the hybrid SIT. The environmental heat loss also contributes to the large discrepancy between the simulation results of RELAP5 and the experimental data.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.811-820
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• 2013

SiC has been under investigation as a potential cladding for LWR fuel, due to its high melting point and drastically reduced chemical reactivity with liquid water, and steam at high temperatures. As SiC is a brittle material its behavior during the reflood phase of a Loss of Coolant Accident (LOCA) is another important aspect of SiC that must be examined as part of the feasibility assessment for its application to LWR fuel rods. In this study, an experimental assessment of thermal shock performance of a monolithic alpha phase SiC tube was conducted by quenching the material from high temperature (up to $1200^{\circ}C$) into room temperature water. Post-quenching assessment was carried out by a Scanning Electron Microscopy (SEM) image analysis to characterize fractures in the material. This paper assesses the effects of pre-existing pores on SiC cladding brittle fracture and crack development/propagation during the reflood phase. Proper extension of these guidelines to an SiC/SiC ceramic matrix composite (CMC) cladding design is discussed.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.225-236
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• 2012

The Fukushima-Daiichi accident shook the world, as a well-known plant design, the General Electric BWR Mark I, was heavily damaged in the tsunami, which followed the Great Japanese Earthquake of 11 March 2011. Plant safety functions were lost and, as both AC and DC failed, manoeuvrability of the plants at the site virtually came to a full stop. The traditional system of Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMG) failed to protect core and containment, and severe core damage resulted, followed by devastating hydrogen explosions and, finally, considerable radioactive releases. The root cause may not only have been that the design against tsunamis was incorrect, but that the defence against accidents in most power plants is based on traditional assumptions, such as Large Break LOCA as the limiting event, whereas there is no engineered design against severe accidents in most plants. Accidents beyond the licensed design basis have hardly been considered in the various designs, and if they were included, they often were not classified for their safety role, as most system safety classifications considered only design basis accidents. It is, hence, time to again consider the Design Basis Accident, and ask ourselves whether the time has not come to consider engineered safety functions to mitigate core damage accidents. Associated is a proper classification of those systems that do the job. Also associated are safety criteria, which so far are only related to 'public health and safety'; in reality, nuclear accidents cause few casualties, but create immense economical and societal effects-for which there are no criteria to be met. Severe accidents create an environment far surpassing the imagination of those who developed EOPs and SAMG, most of which was developed after Three Mile Island - an accident where all was still in place, except the insight in the event was lost. It requires fundamental changes in our present safety approach and safety thinking and, hence, also in our EOPs and SAMG, in order to prevent future 'Fukushimas'.