• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4499-4513
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• 2022

Spent fuel behavior of dry storage was simulated in a continuous state from steady-state operation by modifying FRAPCON-4.0 to incorporate spent fuel-specific fuel behavior models. Spent fuel behavior of a typical PWR was compared with that of NuScale Power Module (NPM^TM). Current PWR discharge burnup (60 MWd/kgU) gives a sufficient margin to the hoop stress limit of 90 MPa. Most hydrogen precipitation occurs in the first 50 years of dry storage, thereby no extra phenomenological safety factor is identified for extended dry storage up to 100 years. Regulation for spent fuel management can be significantly alleviated for LWR-based SMRs. Hydride embrittlement safety criterion is irrelevant to NuScale spent fuels; they have sufficiently lower plenum pressure and hydrogen contents compared to those of PWRs. Cladding creep out during dry storage reduces the subchannel area with burnup. The most deformed cladding outer diameter after 100 years of dry storage is found to be 9.64 mm for discharge burnup of 70 MWd/kgU. It may deteriorate heat transfer of dry storage by increasing flow resistance and decreasing the view factor of radiative heat transfer. Self-regulated by decreasing rod internal pressure with opening gap, cladding creep out closely reaches the saturated point after ~50 years of dry storage.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2002년도 춘계학술대회 논문집
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• pp.684-687
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• 2002

The authors have developed two devices which cuts the spend fuel rod to an optimal size and extracts fuel pellet from the pieces of cut fuel rods. These devices are so important to reduce radioactive wastes that some advanced countries developed their own methods and devices. The authors have benchmarked from these methods and devices. For spent fuel rod cutting, the tube cutting method has been chosen. some mechanical properties of the fuel tube and pellet has been carefully considered for an optimal cutting size. For fuel pellet extraction, a mechanically extracting method has been adopted. The existing chemical method have turned out to be inappropriate because it produced large amount of radioactive wastes, in spite of its high fuel recovery characteristics. The developed method has an advantage that it can be applied to other fuel rods that have different shapes and sizes. The two devices are set up and operated in the hot cell where people can not go in, so that the devices have been designed to be controlled remotely and modulated for easy maintenance. And the performance of the devices has been tested by using simulated fuel rod. From the experimental results, the devices are supposed to be useful for reducing radioactive wastes.

• 분석과학
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• 제17권5호
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• pp.381-387
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• 2004

가압 경수로 사용후핵연료의 화학특성을 규명하기 위하여 극미량 함유되어 있는 삼중수소 ($^3H$)의 정량기술을 확립하였다. 분석과정에서 발생하는 방사성 폐액의 양을 줄이고 분석자의 방사선 피폭을 줄이기 위하여 하나의 시료로부터 $^{14}C$과 $^3H$를 순차적으로 회수할 수 있도록 분리조건을 최적화하였다. 사용후핵연료를 질산으로 용해하는 과정에서 $^{14}CO_2$와 함께 휘발하는 $^{129}I_2$는 $AgNO_3$가 침윤되어 있는 흡착제로 제거하였다. $^{14}CO_2$는 1.5 M NaOH에 포집시키고 $^3H_2O$는 증류시켜 회수하였다. $^3H$의 평균 회수율은 97.9%, 상대표준편차는 0.9% (n = 3) 이었으며, 37,000 MWd/MtU 연소도의 사용후핵연료를 대상으로 $^3H$를 분석하고 표준물첨가법으로 분석신뢰도를 평가하였다.

• Corrosion Science and Technology
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• 제19권1호
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• pp.1-7
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• 2020

Chalk River Unidentified Deposits (CRUDs) were collected from the Korean pressurized water reactor (PWR) plant (A, B, and C) where the axial offset anomaly (AOA) occurred. AOA, also known as a CRUD-induced power shift, is one of the key issues in maintaining stable PWR plant operations. CRUDs were sampled from spent nuclear fuel rods and analyzed using an electron probe micro-analyzer (EPMA). This paper describes the characteristics of boron-deposits from the CRUDs sampled from twice-burnt assemblies from the Korean PWR. The primary coolant of a PWR contains boron and lithium. It is known that boron deposition occurs in a thick CRUD layer under substantial sub-cooled nucleate boiling (SNB). The results of this study are summarized as follows. Boron was not found at the locations where the existence was confirmed in simulated CRUDs, in other words, the cladding and CRUD boundaries. Nevertheless, we clearly observed the presence of boron and confirmed that boron existed as a lump in crystalline form. In addition, the study confirmed that CRUD existed in a crystal form with a unique size of about 10 ㎛.

• Nuclear Engineering and Technology
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• 제28권4호
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• pp.366-372
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• 1996

The simulated spent PWR fuel pellet which is corresponding to the turnup of 33,000 MWD/MTU is prepared by adding 11 fission-product elements to UO$_2$. The simulated spent fuel pellet is treated at 40$0^{\circ}C$ in air (oxidation), at 110$0^{\circ}C$ in air (high-temperature treatment), and at $600^{\circ}C$ in hydrogen (reduction). The product is treated through additional addition and reduction up to 3 cycles. Pellets are completely pulverized by the first oxidation, and the high-temperature treatment causes particle and crystallite to grow and surface to be smooth, and thus particle size significantly increases and surface area decreases. The reduction following the high-temperature treatment decreases much the particle size by means of the formation of intercrystalline cracks. The particle size decreases a little during the second oxidation and reduction cycle and then remains nearly constant during the third and fourth cycles. Surface area of pounder increases progressively with the repetition of oxidation and reduction cycles, mainly due to the formation of Surface cracks. The degradation of surface area resulting from high-temperature treatment is restored by too subsequent resulting oxidation and reduction cycles.

• 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.

• Nuclear Engineering and Technology
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• 제30권2호
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• pp.99-111
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• 1998

Chemical composition of the insoluble residue in a simulated spent PWR fuel(SIMRJEL) were studied. SIMFUELS were prepared by adding calculated amount of FP(fission product) elements with a burnup of 3.6% FIMA(fission per initial metal atom) to uranium in nitrate solution, evaporating the mixed solution to dryness, calcining at 90$0^{\circ}C$ in a stream of 4% H$_2$ + 96% He, and heating the pellet at 140$0^{\circ}C$ under high and low oxygen potentials. Insoluble residue was obtained from the dissolution of the SIMFUEL with HNO$_3$(1 : 1). The chemical composition of the SIMFUELs and the insoluble residues was determined by EPMA(electron probe microanalysis), XPS(X-ray photoelectron spectroscopy) and by XRD (X-ray diffraction) measurements. All of the insoluble residues suspended and precipitated were composed mainly of Mo, Ru with a small amount of Zr, Rh, Pd and Cd. The amount of insoluble residue(<1 wt.%) and a Mo/Ru ratio decreased with increasing oxygen potential. Formation of the zirconium molybdate precipitate, ZrMo$_2$O$_{7}$(OH)$_2$($H_2O$)$_2$, was observed in the residues. The possible role of Mo on the phase formation was discussed in regard to oxygen potential.l.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

• 분석과학
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• 제13권6호
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• pp.751-758
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• 2000

사용후핵연료에 미량 함유되어 있는 요오드를 정량하기 위하여 분리 및 회수를 위한 최적조건을 실험하였다. 사용후핵연료의 용해를 위하여 모의 사용후핵연료를 이용하였으며 용해과정에서 요오드가 휘발되지 않는 혼합산(질산:염산)의 최적 혼합비는 80:20 mol%이었으며 비휘발성 요오드의 생성을 촉진시키기 위하여 오존을 사용하였을 때 요오드의 회수율이 6회 평균 측정값이 9.6% 증가를 나타내었다. 사염화탄소에 의한 요오드 추출과정에서 ${IO_3}^-$를 $I_2$으로 환원시키기 위하여 $NH_2OH{\cdot}HCl$을 사용하였으며 요오드 환원에 적합한 용해용액의 산도는 2.5 M 그리고 $NH_2OH{\cdot}HCl$의 양은 $1.5{\times}10^{-3}mole$ 이상이었다. 분리된 요오드는 이온 크로마토그래피로 정량하였으며 이때 회수율은 $82.8{\pm}4.1%$로서 $^{131}I$를 추적자로 사용하여 보정하였다.

• 방사성폐기물학회지
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• 제2권2호
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• pp.125-133
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• 2004

최대 선출력 61 ㎾/m 및 평균 연소도 1,770 ㎿d/tU의 조건으로 하나로에서 조사한 DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) 핵 연료를 EPMA (Electron Probe Micro Analyzer)를 이용하여 핵분열 생성물을 분석하였다. EPMA의 정확한 분석 방법을 확립하고자, 핵분열생성물 대신 시약을 첨가하여 제조한 모의 DUPIC 핵연료로 EPMA 분석을 수행하였고, 그 결과를 습식 화학 분석의 결과와도 비교하여 평가하였다. 모의 DUPIC 핵연료 중심부의 금속 석출물은 약 1 $\mu\textrm{m}$ 정도의 크기로 관찰되었으며, 이들의 조성은 Mo-53.89 at.%, Ru-37.40 at.% 및 Pd+Rh-8.71 at%이었다. 모의 DUPIC 핵연료 시험에서 정립한 시험방법으로 조사한 DUPIC 핵연료 시편의 금속 석출물 특성을 분석하였다. 핵연료 중앙부에서 관찰된 금속 석출물들의 크기는 2∼2.5 $\mu\textrm{m}$ 정도이었으며, Mo-47.34 at.%, Ru-46 at.%, Pd+Rh-6.65 at.%의 조성임을 확인하였다. 이 실험을 위하여, 특별히 시료의 전도성을 향상시키기 위한 처리를 하였으며, 작은 금속 석출물에 EPMA의 전자빔을 정확히 조사할 수 있는 실험 조건을 제시하였다.