• 한국압력기기공학회 논문집
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• 제16권2호
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• pp.18-24
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• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.

• Corrosion Science and Technology
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• 제23권3호
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• pp.203-214
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• 2024

A heat exchanger is a device designed to transfer heat between two or more fluids. In a vehicle's thermal management system, Al heat exchangers play a critical role in controlling and managing heat for efficient and safe operation of the engine and other components. The fluid used to prevent heat exchangers from overheating the engine is mostly tap water. Heat exchange performance can be maintained at sub-zero temperatures using a solution mixed with antifreeze. Although the fluid flowing through the heat exchanger can reduce the temperature inside the engine, it also has various problems such as cavitation corrosion. Cavitation corrosion characteristics in tap water and corrosion characteristics were evaluated in this study when antifreeze was added for test specimens where AA4045 was cladded on the inner surface of AA3003 tubes of a fin-type heat exchanger. The cavitation corrosion resistance of AA3003 was found to be superior to that of AA4045 regardless of the test solution due to higher corrosion resistance and hardness of AA3003 than those of AA4045. The cavitation corrosion rate of Al alloys increased with the addition of antifreeze.

• 자원리싸이클링
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• 제21권5호
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• pp.18-30
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• 2012

Zr은 고온에서의 높은 치수안정성, 내식성은 물론 낮은 중성자 흡수단면적을 지녀 원자력산업용 소재 중 1차 방사능 차폐재인 핵연료 피복관으로 사용되며 현재까지 다른 소재로 대체 불가능하다. 하지만 Hf을 정제한 Zr sponge 제조기술은 미국, 프랑스, 러시아만 가지고 있어 원자력의존도가 높은 한국에서는 국가전략물자로 분류 철저히 관리되고 있다. 국내 유통되는 Zr의 대부분은 원자력산업에 사용되어 지며 유통구조는 정제된 Zr합금을 국외로부터 수입하여 tube로 가공 후 핵연료집합체로 제조되고, 그 외 소량이 합금첨가원소 및 폭약재 등 고부가가치 일반산업에 사용된다. 본 논문에서는 Zr 제조기술에 대한 현재산업현황 및 정련기술을 살펴보고, 최근 연구되고 있는 Electrolytic reduction process와 Molten oxide electrolysis와 같은 신 제련기술에 대한 소개 및 Zr recycling의 전반적인 기술소개도 포함하였다.

• 한국화재소방학회논문지
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• 제30권4호
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• pp.121-127
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• 2016

본 논문에서는 대전류공급장치(PCITS)에 의해 소손된 강이음쇠형(iron fittings type) 금속플렉시블호스(CSST)의 구조 및 전기적 특성을 해석하였다. CSST는 보호 피막, 튜브, 너트, 클램프 링, 플레어 캡, 소켓, 볼 밸브 등으로 구성되어 있다. CSST의 내전압 평가는 전기 충전부와 비충전부 사이에 교류 전압 220 V를 1분간 인가하여 견뎌야 한다. 직류 500 V에 의한 절연 성능의 평가는 온도 상승 시험 전에 $1M{\Omega}$ 이상, 시험이 끝난 후에는 $0.3M{\Omega}$ 이상을 요구한다. 정상 제품의 평균 저항은 $11.5m{\Omega}$이었으나 PCITS로 130 A를 흘려 소손된 제품의 평균 저항 $11.50m{\Omega}$이었다. 또한 130 A가 약 10 s 흘렀을 때 튜브의 보호 피막이 일부 용융되었고, 검정색의 연기가 발생하였다. 60 s 경과되면 튜브의 대부분이 적색으로 발열되며, 전류가 120 s 흘렀을 때는 적열 범위가 넓어졌다. 95%의 신뢰 구간(CI)의 검증에서 P 값은 0.019로 정규 분포를 갖지 못하였으나 Anderson-Darling (AD) 통계량은 0.896, 표준 편차는 0.5573 등으로 양호한 특성을 나타냈다.

• Nuclear Engineering and Technology
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• 제21권2호
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• pp.111-122
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• 1989

고온 수증기(1323 K)분위기에서 산화시킨 지르칼로이-4 피복관으로, 구리 맨드렐 팽창실험(Copper Mandrel Expension Test)을 변형률(Strain Rate)이 $3.0\times10^{-5}$/sec일때 673-1173 K 온도 범위에서 수행하였다. 본 연구에서, 산화매개변수(Ki)는 시간(t)의 제곱근에 비례하고 $(Ki=\delta_{kit}\frac{1}{2}$), 비례상수($\delta_{ki}$)는 무게증가(Weight Gain), Zr02의 두께, $\alpha$(0) 층에 대하여 각각 0.281, 2.82, 2.313을 사용하였다. 지르칼로이-4의 고온(873-1073 K) 소성변형에 의한 활성화 에너지는 Zr02가 높은 강도를 갖기 때문에 산화 시간이 5분에서 60분으로 증가함에 따라 251 KJ/mol에서 323KJ/mo1로 증가하였다. 산화막 두께, K와 항복 응력의 관계는 ($\sigma/C)^n=K^m$exp (Q/RT)인 관계식을 얻었다. 여기서 n은 6.9, m은 5.7, 그리고 Q가 251, 258, 316, 323 KJ/mo1에 대해 C는 0.155, 0.138, 0.051, 0.046MPa이다.

• 한국산학기술학회논문지
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• 제7권2호
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• pp.113-119
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• 2006

중수로형 핵연료 제조공정 중 연료봉 피복관에 간격체와 지지체 등의 부착물이 브레이징으로 접합된다. 본 연구에서는 베릴륨을 물리 증착법(PVD)으로 접합될 부착물의 표면에 증착한 것과 비정질 용가재[$Zr_{1-x}Be_{x}(0.3{\le}x{\le}0.5)$]를 사용하여 브레이징된 접합부의 미세조직과 경도 등의 특성을 비교하고 브레이징 온도가 접합부에 미치는 영향 조사하였다. 비정질 용가재에 의한 접합층의 두께는 PVD-Be의 경우와 비교하여 더 얇았고, Be 함량이 감소할수록 접합층의 두께는 감소하였으며 모재의 침식은 거의 없었다. PVD-Be의 경우 공정 반응, 액상 출현, 모세관 현상과 확산으로 브레이징 되나 비정질 합금은 용가재 만이 용융되어 액상 접합되는 것으로 사료된다. PVD-Be 접합부의 미세조직은 계면에서 수지상이 형성되어 내부로 성장하나, 비정질 합금에 의한 접합부는 석출된 제2상들이 구상으로 구성되며 브레이징 온도가 증가할수록 구상은 더욱 커졌다. 비정질 합금 접합부의 경도는 Be 함량이 감소할수록 경도는 증가하였다. 본 연구에 사용된 비정질 합금 중 $Zr_{0.7}Be_{0.3}$ 합금은 접합부에서 Be의 모재로의 확산이 적어 부드러운 계면과 모재의 침식이 없었고 높은 경도 때문에 핵연료 피복재 접합에 가장 적합한 용가재로 사료된다.

• 한국재료학회지
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• 제26권12호
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• pp.681-687
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• 2016

Zirconium(Zr) nuclear fuel cladding tubes are made using a three-time pilgering and annealing process. In order to remove the oxidized layer and impurities on the surface of the tube, a pickling process is required. Zr is dissolved in HF and $HNO_3$ mixed acid during the process and pickling waste acid, including dissolved Zr, is totally discarded after being neutralized. In this study, the waste acid was recycled by adding $BaF_2$, which reacted with the Zr ion involved in the waste acid; $Ba_2ZrF_8$ was subsequently precipitated due to its low solubility in water. It is very difficult to extract zirconium from the as-recovered $Ba_2ZrF_8$ because its melting temperature is $1031^{\circ}C$. Hence, we tried to recover Zr using an electrowinning process with a low temperature molten salt compound that was fabricated by adding $ZrF_4$ to $Ba_2ZrF_8$ to decrease the melting point. Change of the Zr redox potential was observed using cyclic voltammetry; the voltage change of the cell was observed by polarization and chronopotentiometry. The structure of the electrodeposited Zr was analyzed and the electrodeposition characteristics were also evaluated.

• Journal of Mechanical Science and Technology
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• 제15권3호
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• pp.291-299
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• 2001

Recently, severe wear on the shutdown rod cladding of Ulchin Nuclear Power Plant #1, #2 were observed by the Eddy Current Test(E.C.T.). In particular, the wear at the sixth card location was up to 75%. The test results indicated that the Flow Induced Vibration(F.I.V.) might be the cause of the fretting wear resulting from the contact between Rod Cluster Control Assemblies(RCCAs) and their spacing cards(guide plates) arranged in the guide tube. From reviewing RCCAs fretting wear repots and analyzing the general characteristics of F.I.V. mechanism in the reactor, geometric layout and flow conditions around the control rod, it is concluded that the turbulence excitation is the most probable vibration mechanism of RCCA. To identify the governing mechanism of RCCA vibration, an experiment was performed for a representative rod position in which the most serious fretting wear experienced among the six rod positions. The experimental rig was designed and set up to satisfy the governing nondimensional numbers which are Reynolds number and mass damping parameter. The vibration amplitude measurement by the non-contact laser displacement sensor showed good agreements in the frequency and the maximum wearing(vibration) location with Ulchin E.C.T. results and Framatome report, respectively. The sudden increase in the vibration amplitude was sensed around the 6th guide plate with mass flow rate variation. Comparing the similitude rod behaviour with the idealized response of a cylinder in flow induced vibration, it was found that he dominant mechanism of vibration was transferred from turbulence excitation to periodic shedding at the mass flow ate 90ι/min. Also the critical velocity of the vibration in RCCAs was determined and the vibration can be prevented by reducing the bypass flow rate below the critical velocity.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.416-424
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• 2018

Microstructure of oxide formed on Zr-Nb-Sn tube sample was intensively examined by scanning transmission electron microscopy after exposure to simulated primary water chemistry conditions of various concentrations of Zn (0 or 30 ppb) and dissolved hydrogen ($H_2$) (30 or 50 cc/kg) for various durations without applying desirable heat flux. Microstructural analysis indicated that there was no noticeable change in the microstructure of the oxide corresponding to water chemistry changes within the test duration of 100 days (pretransition stage) and no significant difference in the overall thickness of the oxide layer. Equiaxed grains with nano-size pores along the grain boundaries and microcracks were dominant near the water/oxide interface, regardless of water chemistry conditions. As the metal/oxide interface was approached, the number of pores tended to decrease. However, there was no significant effect of $H_2$ concentration between 30 cc/kg and 50 cc/kg on the corrosion of the oxide after free immersion in water at $360^{\circ}C$. The adsorption of Zn on the cladding surface was observed by X-ray photoelectron spectroscopy and detected as ZnO on the outer oxide surface. From the perspective of $OH^-$ ion diffusion and porosity formation, the absence of noticeable effects was discussed further.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3095-3105
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• 2022

Solid Microencapsulated Fuel (SMF) is a type of solid fuel rod design that disperses TRISO coated fuel particles directly into a kind of matrix. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study focused on the neutronics and some of the fuel cycle characteristics of SMF by using OpenMC. Two kinds of SMFs have been designed and evaluated - fuel particles dispersed into a silicon carbide matrix and fuel particles dispersed into a zirconium matrix. A 7×7 fuel assembly with increased rod diameter transformed from the standard NHR200-II 9×9 array was also introduced to increase the heavy metal inventory. A preliminary study of two kinds of burnable poisons (Erbia & Gadolinia) in two forms (BISO and QUADRISO particles) was also included. This study found that SMF requires about 12% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in NHR200-II, which is about 7% for SMF with increased rod diameter. Feedback coefficients are less negative through the life of SMF than the reference. And it is estimated that the average center temperature of fuel kernel at fuel rod centerline is about 60 K below that of reference in this paper.