• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1343-1354
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• 2022

Experiments were performed to measure the critical heat flux (CHF) on a vertical surface abutting a coarse packed bed of spherical particles. This geometry is representative of a CANDU reactor calandria tubesheet facing the end shield cavity during the in-vessel retention (IVR) phase of a severe accident. Deionized light water was used as the working fluid. Low carbon steel shielding balls with diameters ranging from 6.4 to 12.7 mm were used, allowing for the development of an empirical correlation of CHF as a function of shielding ball diameter. Previously published data is used to develop a more comprehensive empirical correlation accounting for the impacts of both shielding ball diameter and heating surface height. Tests using borosilicate shielding balls demonstrated that the dependence of CHF on shielding ball thermal conductivity is insignificant. The deposition of iron oxide particles transported from shielding balls to the heating surface is verified to increase CHF non-trivially. The results presented in this paper improve the state of the knowledge base permitting quantitative prediction of CHF in the CANDU end shield, refining our ability to assess the feasibility of IVR. The findings clarify the mechanisms governing CHF in this scenario, permitting identification of potential future research directions.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.536-545
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• 2024

Radioactive iodine is a representative fission product to be quantified for the safety assessment of nuclear facilities. In integral severe accident analysis codes, the iodine behavior is usually described by a multi-physical model of iodine chemistry in aqueous phase under radiation field and mass transfer through gas-liquid interface. The focus of studies on iodine source term evaluations using the combination approach is usually put on the chemical aspect, but each contribution to the iodine amount released to the environment has not been decomposed so far. In this study, we attempted the decomposition by revising the two-film theory of molecular-iodine mass transfer. The model involves an effective overall mass transfer coefficient to consider the iodine chemistry. The decomposition was performed by regarding the coefficient as a product of two functions of pH and the overall mass transfer coefficient for molecular iodine. The procedure was applied to the EPICUR experiment and suppression chamber in BWR.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.824-829
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• 1998

미국의 개량형 원자력 발전소 개념설계단계에서 중대사고시 사고완화를 위한 전략으로 원자로 압력용기 외부냉각 개념이 제안되었다. 중대사고 진행과정에서 노심용융물이 원자로 압력용기 하부헤드로 재배치 되었을 때 압력용기 외벽을 냉각함으로서 노심용융물을 압력용기 내부에 가두어 두어 격납건물 내로의 유출을 방지하는 방식이다. 이 연구에서는 원자로 압력용기 하부헤드 내의 노심용융물 거동중 자연 순환에 의한 거동을 수치적으로 모의하여 보았다. 연구결과, 정상상태의 온도 및 속도분포는 현상학적으로 적절하게 모의되나 고화와 액화의 경우에는 고유모델의 필요성이 요구되었다.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.180-189
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• 2023

We measured the heat load to a reactor vessel with and without the in-vessel debris bed under an IVR-ERVC condition. Mass transfer methodology was adopted based on heat and mass transfer analogy to achieve high Ra'_H of order ~10¹⁵ with compact test rigs. We postulated the in-vessel debris bed has a flat top and particulate debris was simulated as an identical diameter spheres. We conducted experiments varying the height of the debris bed and the results showed that Nusselt numbers decreased in both uppermost and curved surfaces with the increasing bed height. Once the debris bed is formed, it acts as an obstacle to the natural convective flow, which reduces the buoyancy. The reduction of driving force results in the impaired heat transfer in both upward and downward heat transfers.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.165-178
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• 2013

A comprehensive review on the technical standards about human factors (HF) design and software reliability maintenance for digital instrumentation and control (I&C) and human-machine interface technology (HMIT) in Japanese light water reactor nuclear power plants (NPPs) was given in this paper mainly by introducing the relevant activities at the Japan Electric Association to set up many industrial standards within the traditional framework of nuclear safety regulation in Japan. In Japan, the Fukushima Daiichi accident that occurred on March 11, 2011 has great impact on nuclear regulation and nuclear industries where concerns by the general public about safety have heightened significantly. However for the part of HF design and software reliability maintenance of digital I&C and HMIT for NPP, the author believes that the past practice of Japanese activities with the related technical standards can be successfully inherited in the future, by reinforcing the technical preparedness for the prevention and mitigation against any types of severe accident occurrence.

• Nuclear Engineering and Technology
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• v.16 no.3
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• pp.141-154
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• 1984

An analysis is presented of key phenomena and scenario which imply some general trends for beyond design-basis-accident in Kori-1 PWR dry containment. The study covers a wide range of severe accident sequences initiated by small break LOCA. The MARCH computer code, with KAERI modifications was used in this analysis. The major emphasis of the paper are two folds, 1) the phenomenologic understanding of severe accident and 2) a study of H2 combustion and debris/ water interactions in a specific small break LOCA for Kori-1 plant. The sensitivity studies for the specific plant data and thermal interaction modelings used in the SASA were performed. The results show that if hydrogen burning does occur at low concentration, the resulting peak pressure does not exceed the design value, while the lower concentration assumption results in repeated burning due to the continuing H$_2$ generation. For debris/water interaction, the particle size has no effect on the magnitude of peak pressure for the amount of water assumed to be in the reactor cavity. But, the occurrence of peak pressure is considerably delayed in case of using the dryout correlation. The peak containment pressure predicted from the hydrogen combustion and steam pressure spite during full core meltdown scenario does not present a severe threat to the containment integrity.

• Nuclear Engineering and Technology
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• v.53 no.10
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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.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.989-998
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• 2023

Under nuclear reactor severe accidents, the resuspension of radioactive aerosol may occur in the containment due to the disturbing airflow generated by hydrogen combustion, hydrogen explosion and containment depressurization resulting in the increase of radioactive source term in the containment. In this paper, for containment conditions, by considering the contact between particle and rough deposition surface, the distribution of the distance between two contact points of particle and deposition surface, rolling and lifting separation mechanism, resuspension model based on random contact with rough surface (RRCR) is established. Subsequently, the detailed torque and force analysis is carried out, which indicates that particles are more easily resuspended by rolling under low disturbing airflow velocity. The simulation result is compared with the experimental result and the prediction of different simulation methods, the RRCR model shows equivalent and better predictive ability, which can be applicable for simulation of aerosol resuspension in containment during severe accident.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.448-456
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• 2016

Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies themass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The $H_2O$ content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of $CO_2$ necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

• Nuclear Engineering and Technology
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• v.41 no.9
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• pp.1157-1170
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• 2009

A variety of Generation III/III+ reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world to solve the future energy supply shortfall. Nanofluid coolants showing an improved thermal performance are being considered as a new key technology to secure nuclear safety and economics. However, it should be noted that there is a lack of comprehensible design works to apply nanofluids to Generation III+ reactor designs. In this work, the review of accident scenarios that consider expected nanofluid mechanisms is carried out to seek detailed application spots. The Axiomatic Design (AD) theory is then applied to systemize the design of nanofluid-engineered nuclear safety systems such as Emergency Core Cooling System (ECCS) and External Reactor Vessel Cooling System (ERVCS). The various couplings between Gen-III/III+ nuclear safety features and nanofluids are investigated and they try to be reduced from the perspective of the AD in terms of prevention/mitigation of severe accidents. This study contributes to the establishment of a standard communication protocol in the design of nanofluid-engineered nuclear safety systems.