• Corrosion Science and Technology
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• 제5권3호
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• pp.94-98
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• 2006

The next generation nuclear power reactor will use Pb-Bi as the cooling material. The steel structure materials such as HT-9 used in the reactor suffer from corrosion when they are exposed to high temperature Pb-Bi. Therefore corrosion should be prevented to use Pb-Bi as the coolant material without any safety problem. One method is to control the oxygen content in Pb-Bi. An appropriate amount of oxygen in Pb-Bi can produce a thin oxide layer on steel, and this layer protects the steel from corrosion attack. Since the required oxygen content in Pb-Bi is in the range of $10^{-5}$ to $10^{-7}$ wt%, this small oxygen content can be controlled by flowing a mixture of hydrogen gas and water vapor. The stagnant corrosion test of HT-9 samples was performed by controlling the oxygen content up to 2,000 hours. The corrosion behavior of HT-9 was analyzed at the temperatures of $500^{\circ}C$ and $650^{\circ}C$ with a reduced condition and a oxygen content of $10^{-6}$ wt%.

• Nuclear Engineering and Technology
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• 제47권6호
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• pp.719-728
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• 2015

Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

• Journal of Energy Engineering
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• 제25권1호
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• pp.153-162
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• 2016

Experimental investigation was carried out on the CHF(Critical Heat Flux) under downward flow condition in narrow rectangular channels simulating subchannel of plate-type-fuel for JRTR(Jordan Research and Training Reactor). The experiments covers the license requirement of the research reactor. Two test sections used in this study simulate full scale subchannels for fission moly uranium target and plate-type-fuel, respectively. From the experimental results, the parameters affecting on the CHF are investigated. By using experimental data, the existing CHF prediction models were evaluated. Finally, the applicability of correlations were analysed to predict CHF in the narrow rectangular channel under the downward flow condition.

• Transactions of the Korean hydrogen and new energy society
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• 제24권5호
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• pp.428-436
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• 2013

Attrition characteristics of $CO_2$ absorbents for pre-combustion $CO_2$ capture were investigated to check attrition loss of those absorbents and to determine solid circulation direction and the better $CO_2$ absorbent. The cumulative attrition losses of two absorbents increased with increasing time. However, attrition loss under a humidified condition was lower than that under a non-humidified condition case. Between two absorbents, attrition loss of PKM1-SU absorbent was higher than that of P4-600 absorbent. The average particle sizes of the attrited particles were less than $2.5{\mu}m$ for two absorbents under a non-humidified condition case, and therefore, we could conclude that the main mechanism of attrition for two absorbents is not fragmentation but abrasion. Based on the results from the test for the effect of humidity on the attrition loss, we selected solid circulation direction from SEWGS reactor to regeneration reactor because the SEWGS reactor contains more water vapor than regeneration reactor. Attrition loss and make-up rate of two absorbents were compared based on the results from $CO_2$ sorption capacity tests and attrition tests. Required make-up rate of P4-600 absorbent was lower than that of PKM1-SU absorbent. However, more detail investigation on the optimum regeneration temperature, manufacturing cost, solid circulation rate, regeneration rate, and long-term sorption capacity should be considered to select the best $CO_2$ absorbent.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.218-227
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• 2016

Background: Steam explosions may occur in nuclear power plants by molten fuel-coolant interactions when the external reactor vessel cooling strategy fails. Since this phenomenon can threaten structural barriers as well as major components, extensive integrity assessment research is necessary to ensure their safety. Method: In this study, the influence of yield criteria was investigated to predict the failure of a reactor cavity under a typical postulated condition through detailed parametric finite element analyses. Further analyses using a geometrically simplified equivalent model with homogeneous concrete properties were also performed to examine its effectiveness as an alternative to the detailed reinforcement concrete model. Results: By comparing finite element analysis results such as cracking, crushing, stresses, and displacements, the Willam-Warnke model was derived for practical use, and failure criteria applicable to the reactor cavity under the severe accident condition were discussed. Conclusion: It was proved that the reactor cavity sustained its intended function as a barrier to avoid release of radioactive materials, irrespective of the different yield criteria that were adopted. In addition, from a conservative viewpoint, it seems possible to employ the simplified equivalent model to determine the damage extent and weakest points during the preliminary evaluation stage.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• 제25권12호
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• pp.807-814
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• 2015

A control rod drive mechanism(CRDM) is a reactor regulating system, which inserts, withdraws or maintains a control rod containing a neutron absorbing material within a reactor core to control the reactivity of the core. The top-mounted CRDM for Jordan Research and Training Reactor(JRTR) with 5 MW power has been designed and fabricated based on the HANARO's experience through KAERI and DAEWOO consortium project. This paper describes the performance qualification test results to demonstrate the operability of a prototype and four production CRDMs during the reactor lifetime. The driving performance, the drop performance and the endurance tests for CRDM are carried out at a test rig simulating the actual reactor conditions. A vibration of internal components due to the coolant flow is also measured using a laser vibrometer. As a result, the CRDMs are driven having a good driving performance without a malfunction between command and output signals for the stepping motor. Also, the pure drop time and the impact acceleration are within 0.72 s and 4.2 g to meet the design requirements, and the vibrational displacement of control rod is measured as maximum $5.2{\mu}m$.

• Proceedings of the Korean Nuclear Society Conference
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.109-110
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• 2005

• Proceedings of the Korean Radioactive Waste Society Conference
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.502-503
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 한국방사성폐기물학회 2017년도 추계학술논문요약집
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• pp.333-334
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• 2017

• Proceedings of the Korean Radioactive Waste Society Conference
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• 한국방사성폐기물학회 2017년도 춘계학술논문요약집
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• pp.223-224
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• 2017