• Nuclear Engineering and Technology
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• 제56권3호
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• pp.1002-1012
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• 2024

To remove insoluble fission products, which could possibly cause reactor instability and significantly reduce heat transfer efficiency from primary system of molten salt reactor, a helium bubbling method is employed into a passive molten salt fast reactor. In this regard, two-phase flow behavior of molten salt and helium bubbles was investigated experimentally because the helium bubbles highly affect the circulation performance of working fluid owing to an additional drag force. As the helium flow rate is controlled, the change of key thermal-hydraulic parameters was analyzed through a two-phase experiment. Simultaneously, to assess the applicability of numerical model for the analysis of two-phase flow behavior, the numerical calculation was performed using the OpenFOAM 9.0 code. The accuracy of the numerical analysis code was evaluated by comparing it with the experimental data. Generally, numerical results showed a good agreement with the experiment. However, at the high helium injection rates, the prediction capability for void fraction of helium bubbles was relatively low. This study suggests that the multiphaseEulerFoam solver in OpenFOAM code is effective for predicting the helium bubbling but there exists a room for further improvement by incorporating the appropriate drag flux model and the population balance equation.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1513-1525
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• 2024

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.

• Journal of Radiation Protection and Research
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• 제35권1호
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• pp.12-20
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• 2010

원전 일차계통은 복잡한 수질환경으로 방사화생성물과 부식생성물 등 다양한 방사성핵종이 생성되고 있다. 이 방사성 핵종 중에서 원전종사자 피폭방사선량평가와 방사성유출물관리 측면에서 중요한 핵종으로는 $^3H,\;^{14}C,\;^{58}Co,\;^{60}Co,\;^{137}Cs,\;^{131}I$를 들 수 있다. 본 논문은 원전 방사선작업종사자와 원전 주변주민의 피폭방사선량에 기여가 큰 방사성핵종에 대해 살펴보고, 이들 핵종에 의한 선량평가 과정을 소개하였다. 특히 국내 원전에서 발생하였던 $^{131}I$ 내부피폭사건과 일차계통 냉각수의 탈염수 오염사건 등을 포함한 원전의 운영과정에서 일어났던 종사자와 원전주변주민에 대한 피폭 방사선량 평가 사례를 제시하였다. 또한 최근 이슈로 떠오른 삼중수소와 $^{14}C$의 선량평가에 대한 잠재적인 현안 등도 간단히 기술하였다.

• Nuclear Engineering and Technology
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• 제43권2호
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• pp.149-158
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• 2011

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17${\times}$17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and $^{90}Sr$, respectively. Finally, it was found that 88.7% of the metal waste from the 17${\times}$17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

• 대한기계학회논문집A
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• 제39권3호
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• pp.249-257
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• 2015

원자력 발전소의 반응로에는 핵분열 에너지를 생성하고 방사성 물질의 유출을 막는 핵연료 집합체가 있으며, 이러한 집합체는 핵연료와 피복관으로 구성되어 있는 핵 연료봉으로 구성되어 있다. 원자로에서 핵연료봉 거동의 안전성을 평가하기 위해 해석적인 방법을 적용하며 이러한 평가 코드를 핵 연료 성능 코드라 한다. 경수로 핵연료 해석에서는 간극의 두께에 따라 열전도도가 크게 영향을 받는 간극 열전도도가 주요 거동해석에 영향을 미친다. 본 연구에서는 간극 두께에 따라 열전도도가 변화하는 3 차원 간극 요소(Gap element)를 제안하였으며, 이를 적용하기 위해 3 차원 열탄성 모듈을 FORTRAN90을 이용하여 개발하였다. 제안된 3 차원 간극 요소를 이용하여 핵 연료봉에서 발생할 수 있는 비대칭적인 형상인 핵 연료 표면에 결함이 생긴 경우 MPS(Missing Pellet Surface)와 핵연료봉의 편심(Eccentricity of the nuclear fuel rod) 형상에 대하여 3 차원 해석을 진행하였다.

• 분석과학
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• 제16권2호
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• pp.117-124
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• 2003

고온 건식공정의 사용후핵연료 산화분말 ($U_3O_8$)과 경 중수로 연계 핵연료 제조공정의 $UO_2$ 소결체 물성 이해에 필요한 Oxygen/Metal 비를 습식 및 건식 분석방법으로 측정하였다. $UO_2$ 분말에 핵분열생성물 원소의 산화물을 일정량 첨가하고 $1,700^{\circ}C$의 수소분위기에서 소결시켜 20,000~60,000 MWd/MtU 연소도 범위의 사용후핵연료와 화학조성이 유사한 모의 사용후핵연료를 제조하였다. 습식법에 의한 O/M 비 측정을 위하여 혼합산 (10 M HCl : 8 M $HNO_3$, 2.5:1 V/V)에 의한 가압산분해법으로 모의 사용후핵연료를 용해하고 우라늄과 핵분열생성물 원소를 추출 크로마토그래피로 분리한 후 금속원소의 총량을 유도결합플라스마 원자방출분광분석법으로 결정하였다. 또한 $UO_2$가 산화될 때의 무게변화를 열중량 무게분석법 (thermogravimetric)으로 측정하여 O/M비를 계산하고 습식법으로 얻은 결과와 비교하였다. $Mo_{0.4}-Ru_{0.4}-Rh_{0.1}-Pd_{0.1}$ 합금이 O/M비 측정에 미치는 영향을 조사하였다.

• 분석과학
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• 제13권3호
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• pp.309-314
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• 2000

리튬 용융염($LiC+Li_2O$) 내에 미량으로 존재하는 핵분열생성물을 유도결합 플라스마 원자방출분광기(ICP-AES)를 이용하여 분석하였다. 분석대상 원소 중 감도가 가장 좋은 파장을 선택해 이들 파장에서 리튬 500, 1,000, 2,000 mg/L에 따른 분광학적 간섭 여부를 확인한 결과, 분석원소중 0.1 mg/L 이하의 Y, Nd, Sr, La, Eu의 방출세기는 리튬의 농도가 2,000 mg/L까지 증가시켜도 분광학적 간섭을 받지 않은 반면, Mo, Ba, Ru, Pd, Rh, Zr, Ce는 10% 에서 50% 이상 분광학적 간섭과 매트릭스에 의한 스펙트럼 방해가 나타났다. 리튬 매질로부터 미량 금속원소들을 군분리하기 위하여 모의 용융염 용액을 조제해 암모니아수를 가한 후 분리하고 다시 산처리하여 얻은 용액을 ICP-AES로 회수율을 측정하였다. 분석원소 중 Ru, Y, Rh, Zr, Nd, Ce, La, Eu의 회수율은 90% 이상인 반면 Mo, Ba, Pd, Sr는 낮은 회수율을 보여주었으며, 가해준 암모니아수 양이 증가할수록 회수율이 증가하는 경향을 보였다.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3367-3378
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• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

• 방사성폐기물학회지
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• 제17권1호
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• pp.37-46
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• 2019

사고저항성 핵연료의 일환으로 $UO_2$ 입자가 세라믹 셀 벽으로 둘러싸인 미세구조를 갖는 세라믹 미소셀 $UO_2$ 소결체를 개발 중이다. 이는 핵분열생성물들을 $UO_2$ 펠렛 내에 포집하여 펠렛 외부로의 방출을 저감함으로써 봉내압 상승을 완화하고 응력부식균열 발생률을 낮춘다. 생성량이나 방사능 측면에서 위험한 핵분열생성물 중 하나로 여겨지는 세슘은 세라믹 미소셀소결체 내에서 셀 물질과 화학반응 하여 포집될 수 있다. 따라서, 세슘 포집능은 해당 화학반응의 열역학적, 속도론적 특성에 의해 결정된다. 역으로, 미소셀 소결체의 조성설계 시 해당 반응에 대한 열역학적 예측이 필수적이다. 본 연구는 세라믹 현재 개발 중인 여러 미소셀 조성(Si-Ti-O, Si-Cr-O, Si-Al-O)에 대해 세슘의 포집능을 평가하는 열역학적 계산을 다룬다. 계산에 앞서 먼저 HSC Chemistry를 이용해 세슘과 셀 물질의 물리/화학적 상태를 정의한 후, LWR 정상운전 모사환경에서 계산된 세슘포텐셜(${\Delta}G_{Cs}$)과 산소포텐셜(${\Delta}G_{O_2}$)에 근거하여 세슘포집 반응성을 평가하였다. 계산 결과에 근거하면, 세슘 포집반응은 상기 모든 조성에서 자발적일 것으로 예상되며 이로써 조성설계의 근거를 제시함과 동시에 세슘의 포집능을 평가하는 효과적인 방법을 제공한다.

• 한국세라믹학회지
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• 제51권5호
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• pp.418-423
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• 2014

Fully ceramic micro encapsulated (FCM) nuclear fuel is a concept recently proposed for enhancing the stability of nuclear fuel. FCM nuclear fuel consists of tristructural-isotropic (TRISO) fuel particles within a SiC matrix. Each TRISO fuel particle is composed of a $UO_2$ kernel and a PyC/SiC/PyC tri-layer which protects the kernel. The SiC ceramic matrix is created by sintering. In this FCM fuel concept, fission products are protected twice, by the TRISO coating layer and by the SiC ceramic. The SiC ceramic has proven attractive for fuel applications owing to its low neutron-absorption cross-section, excellent irradiation resistivity, and high thermal conductivity. In this study, a SiC-matrix composite containing TRISO particles was sintered by hot pressing with $Al_2O_3-Y_2O_3$ additive system. Various sintering conditions were investigated to obtain a relative density greater than 95%. The internal distribution of TRISO particles within the SiC-matrix composite was observed using an x-ray radiograph. The fracture of the TRISO particles was investigated by means of analysis of the cross-section of the SiC-matrix composite.