• Nuclear Engineering and Technology
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• 제47권6호
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• pp.700-711
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• 2015

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called "self-wastage phenomena." A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodiumwater reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm).

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2000년도 춘계학술대회논문집
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• pp.1829-1835
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• 2000

The experimental results of sodium-water reaction noise measurement in frequency range $1{/sim}200kHz$ are presented. The experiments of noise generation under the condition of sodium test facility, water leak rate $0.01{\sim}1.2g/s$ and temperature of sodium $250{\sim}500^{\circ}C$, were carried out. From theoretical study it is noted that the noise resonant attenuation on hydrogen bubbles in liquid sodium plays the significant role for leak noise spectra formation. Interaction of leak noise and hydrogen bubbles in sodium being accompanied by thermal, emission and viscosity energy dissipation was studied. Acoustic noise spectra were investigated from point of view of water leak detection in sodium/water steam generator. The results of sodium-water reaction noise absorption on hydrogen bubbles in liquid sodium by temperature $250{\sim}500^{\circ}C$ are presented. The theoretical model of noise absorption using the coefficients of attenuation was developed. From calculation the coefficients of attenuation were estimated.

• Korean Chemical Engineering Research
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• 제57권1호
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• pp.28-41
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• 2019

본 연구는 소듐냉각고속로 원형로 소듐-물 반응 압력완화계통의 성능 해석을 목적으로 한다. 증기발생기의 전열관 파단에 의한 대규모 물 누출 사고 발생 시, 증기발생기 전열관 내측의 물을 급수덤프탱크로 배출하고 전열관 외측의 소듐 및 반응생성물을 소듐덤프탱크로 배출 할 때 유체의 거동을 해석하여 계통 설계요건의 적절성을 평가하였다. 증기발생기 쉘 측의 액체와 중간열전달계통 내 소듐이 모두 배출되는데 소요되는 시간은 약 50초이고, 증기발생기 전열관 측의 급수가 모두 배출되는데 소요되는 시간은 약 2.5초로 계산되었다. 증기발생기와 중간열전달계통 내 유체가 덤프탱크로 배출되는 동안 전열관 측의 압력은 쉘 측의 압력보다 높게 유지되어 쉘 측의 소듐이 전열관 측으로 역류하는 현상은 없는 것으로 해석되었다. 본 연구의 결과는 SFR 원형로 소듐-물반응압력완화계통의 성능 평가에 대한 기초자료로 활용할 예정이다.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.893-899
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• 2024

In a secondary cooling system of a sodium-cooled fast reactor (SFR), rapid detection of hydrogen due to sodium-water reaction (SWR) caused by water leakage from a heat exchanger tube of a steam generator (SG) is important in terms of safety and property protection of the SFR. For hydrogen detection, the hydrogen detectors using atomic transmission phenomenon of hydrogen within Ni-membrane were used in Japanese proto-type SFR "Monju". However, during the plant operation, detection signals of water leakage were observed even in the situation without SWR concerning temperature up and down in the cooling system. For this reason, the study of a new hydrogen detector has been carried out to improve stability, accuracy and reliability. In this research, the authors focus on the difference in composition of hydrogen and the difference between the background hydrogen under normal plant operation and the one generated by SWR and theoretically estimate the hydrogen behavior in liquid sodium by using ultra-accelerated quantum chemical molecular dynamics (UA-QCMD). Based on the estimation, dissolved H or NaH, rather than molecular hydrogen (H₂), is the predominant form of the background hydrogen in liquid sodium in terms of energetical stability. On the other hand, it was found that hydrogen molecules produced by the sodium-water reaction can exist stably as a form of a fine bubble concerning some confinement mechanism such as a NaH layer on their surface. At the same time, we observed experimentally that the fine H₂ bubbles exist stably in the liquid sodium, longer than previously expected. This paper describes the comparison between the theoretical estimation and experimental results based on hydrogen form in sodium in the development of the new hydrogen detector in Japan.

• Korean Chemical Engineering Research
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• 제56권3호
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• pp.388-403
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• 2018

본 연구는 소듐냉각고속로 원형로 소듐-물 반응 압력완화계통의 소듐, 물, 가스 배출배관 설계에 필요한 요건 도출을 목적으로 한다. 증기발생기의 전열관 파단에 의한 대규모 물 누출 사고 발생 시, 증기발생기 전열관 측의 물과 전열관 외측의 소듐 및 반응생성물을 물배출조와 소듐배출조로 신속하게 배출하기 위해 증기발생기의 소듐 배출배관 파열판 면적, 소듐배출조의 가스 방출배관 직경, 물배출조의 기체 방출배관 직경, 증기발생기의 물 배출배관 직경 등을 설정하기 위한 계산을 수행하였다. 이를 바탕으로 대규모 물 누출 사고 발생 시 증기발생기 내 유체 배출 소요시간 및 압력거동 해석을 수행하였고, 증기발생기 물 배출배관 격리밸브의 차단 설정압력 등의 설계인자를 도출하였다. 본 연구에서 도출된 설계인자들은 소듐냉각고속로 원형로 소듐-물 반응 압력완화계통 설계에 기초자료로 활용할 예정이다.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.952-964
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• 2016

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium-water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium-water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

• 한국세라믹학회지
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• 제45권12호
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• pp.845-850
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• 2008

Hydrated sodium silicate was synthesized by hydrothermal reaction using anhydrous sodium silicate. The optimum additions of water was 25wt% to make hydrated sodium silicate with homogeneous and purposed water contents. Porous ceramics with homogeneous microstructure and spherical closed pore can be fabricated by elimination of the large pores(a few mm in size) which was formed during first heat treatment through the decomposition of water. Spherical closed pore was formed above $600^{\circ}C$ and the pore size was increased with increasing second heat treatment temperature due to growth of pores. The size of spherical closed pore was varied from 35 to $233\;{\mu}m$ and specific gravity was varied from 0.2 to 1.02 depending on the combinations of the first and second heat treatment temperature.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.368-375
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• 2016

A sodiume-water reaction takes place when high-pressured water vapor leaks into sodium through a tiny defect on the surface of the heat transfer tube in a steam generator of the sodium-cooled fast reactor. The sodiume-water reaction brings deterioration of the mechanical strength of the heat transfer tube at the initial leakage site. As a result, it damages the crack itself, which may eventually enlarge into a larger opening. This self-enlargement is called "self-wastage phenomenon." In this study, a simulant experiment was proposed to reproduce the self-enlargement of a crack and to evaluate the mechanism of the self-wastage. The damage on the surface of the crack was simulated by making the neutralization reaction with hydrochloric acid solution and sodium hydroxide solution. A numerical investigation was carried out to validate the feasibility of the approach and to determine experimental conditions. From the computation results, it is observed that when 5M HCl is injected into 5M of NaOH with 0.05 m/s inlet velocity, the temperature at the surface near the crack increased over 319.26 K. The computational results show that the self-wastage phenomenon is capable of being reproduced by the simulant experiment.

• 공업화학
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• 제9권2호
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• pp.268-272
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• 1998

액체금속로 냉각재인 액체 소듐에서 시편의 누출특성을 소듐-물 반응 실험에 의해 조사하였다. 소듐-물 반응 현상의 확인은 물 누출 실험 전후에 Fe, Cr 및 Ni 등과 같은 시편의 조성 변화로 확인하였다. $100kg/cm^2$의 누출 압력으로 4시간 동안 시편의 누출 경로를 통해 물을 누출시킨 결과, 누출경로에서 소듐-물 반응생성물들이 침적되어 있는 것을 확인하였으나, 부식에 의해 누출경로가 완전 파열되어 다량의 수증기가 액체 소듐속으로 빠져나가는 re-openning 현상은 관찰되지 않았다. 시편의 누출경로가 막히는 self-plugging 현상은 소듐-물 반응에 의한 반응생성물과 시편의 부식에 의한 부식 생성물이 주 원인으로 추정되고, re-openning 현상은 시편의 누출경로에서 열적인 transient로 추정되었다.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2262-2275
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• 2022

The Korea Atomic Energy Research Institute (KAERI) studied the sodium-water reaction (SWR) minimized steam generator for the safety of the sodium-cooled fast reactor (SFR), and selected the copper bonded steam generator (CBSG) as the optimal concept. This paper introduces the conceptual design of the CBSG and the development of the CBSG sizing analyzer (CBSGSA). The CBSG consists of multiple heat transfer modules with a crossflow heat transfer configuration where sodium flows horizontally and water flows vertically. The heat transfer modules are stacked along a vertical direction to achieve the targeted large heat transfer capacity. The CBSGSA code was developed for the thermal-hydraulic analysis of the CBSG in a multi-pass crossflow heat transfer configuration. Finally, we conducted a preliminary sizing and rating analysis of the CBSG for the trans-uranium (TRU) core system using the CBSGSA code proposed by KAERI.