• Nuclear Engineering and Technology
• /
• 제29권2호
• /
• pp.93-98
• /
• 1997

Air-oxidation experiments on unirradiated and irradiated UO$_2$ were performed at temperature from 150 to 375$^{\circ}C$ for investigating the long-term storage behavior of spent PWR fuel. The rate of oxidation was monitored by a thermogravimetric analyzer(TGA) and an X-ray diffraction(XRD). The correlation between the onset-time for U$_3$O$_{8}$ formation and temperature was given as follows, logt(hr) = -12.89+7650/T(K), 423$_2$ pellet, the oxidation rate of irradiated UO$_2$ increase more rapidly at the initial stage and shows a lower saturation point at the later Stage. The Oxidation rate of high bumup UO$_2$ and gadolinia-doped UO$_2$(Gd$_2$O$_3$-UO$_2$) were observed to be much slower than that of unirradiated UO$_2$ pellets.s.

• 한국압력기기공학회 논문집
• /
• 제12권1호
• /
• pp.134-140
• /
• 2016

The Control Rod Assembly (CRA) controls the reactor power by adjusting its position in the reactor core during normal operation and should be quickly inserted into the reactor core by free drop under scram condition to shut down chain reactions. Therefore, the drop time of the CRA is one of important factors for the safety of the nuclear reactor and must be experimentally verified. This study presents the drop performance test of the CRA which has been conceptually designed for the Proto-type Generation IV Sodium-cooled Fast Reactor. During the test, the CRA was free dropped from a height of 1 m under different flow rate conditions and its drop time was measured. The results showed that the drop time of the CRA increased as the flow rate increased; the average drop times of the CRA were approximately 1.527 seconds, 1.599 seconds and 1.676 seconds at 0%, 100% and 200% of design flow rates, respectively.

• 한국방사성폐기물학회:학술대회논문집
• /
• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
• /
• pp.280-289
• /
• 2005

A pyrochemical process has been introduced and utilized so that the transmutation of spent PWR fuel in PEACER can produce mainly low and intermediate level waste for near surface disposal. Major radioactive nuclides from PEACER pyroprocessing are composed of TRU and LLFP. In this study, the requirement for the final waste from PEACER is evaluated based on the methodology for establishment of waste acceptance criteria. Also, sensitivity analysis for several input parameters is conducted in order to determine acceptable decontamination factor (DF) and LLFP removal efficiency and to find out input parameter that extremely have an effect on DE As a result of the study, LLFP removal efficiency, especially Sr-90 and Tc-99, is proved to be a major nuclide which contributes to annual dose by human intrusion scenario rather than TRU DF. More than $98.5\%$ of LLFP have to be removed to meet below dose constraint within the DF more than 5.0E+03. Besides, because of the relative short half-life of Sr-90, the increasing of the institutional control period is recommended for most important input parameter to determine DF.

• Nuclear Engineering and Technology
• /
• 제23권4호
• /
• pp.410-419
• /
• 1991

국내 남동지역에서 수집된 4가지 벤토나이트 시료를 대상으로 X-선 회절과 화학조성을 분석한 결과 Ca-벤토나이트인 것으로 나타났다. 4가지 시료의 비표면적, 양이온교환능 및 팽윤도를 비교하여, 분배계수 측정을 위한 적절한 재료로서 동해 A 시료를 선택하였다. Cs, Co 및 Am의 흡착평형은 약 10일 정도에서 이루어졌으며, Sr의 경우는 이보다 훨씬 발리 이루어졌다. 분배계수 측정 결과로부터 국산 벤토나이트가 높은 흡착능을 가지고 있음을 알았으며, 농도변화에 대한 분배계수 감은 약 $10^{-7}$ mo1/$\ell$의 농도범위에서 최고를 나타내었다.

• Corrosion Science and Technology
• /
• 제17권4호
• /
• pp.147-153
• /
• 2018

Copper is used in various applications in environments favoring and enabling formation of biofilms by naturally occurring microbes. Copper is also the chosen corrosion barrier for nuclear waste in Finland. The copper canisters should have lifetimes of 100,000 years. Copper is commonly considered to be resistant to corrosion in oxygen-free water. This is an important argument for using copper as a corrosion protection in the planned canisters for spent nuclear-fuel encapsulation. However, microbial biofilm formation on metal surfaces can increase corrosion in various conditions and provide conditions where corrosion would not otherwise occur. Microbes can alter pH and redox potential, excrete corrosion-inducing metabolites, directly or indirectly reduce or oxidize the corrosion products, and form biofilms that create corrosive microenvironments. Microbial metabolites are known to initiate, facilitate, or accelerate general or localized corrosion, galvanic corrosion, and intergranular corrosion, as well as enable stress-corrosion cracking. Sulfate-reducing bacteria (SRB) are present in the repository environment. Sulfide is known to be a corrosive agent for copper. Here we show results from corrosion of copper in anoxic simulated ground water in the presence of SRB enriched from the planned disposal site.

• Nuclear Engineering and Technology
• /
• 제27권3호
• /
• pp.447-452
• /
• 1995

경·중수로 연계 핵연료 주기 (Direct Use of Spent PWR Fuel in CANDU : DUPIC ) 기술개발의 핵물질 보장조치(Safeguards)는 경수로 사용후 핵연료를 중수로에 재 활용하기 위한 DUPIC 공정에 대한 최적 보장조치 시스템을 구축하여, 국제 원자력 기구(IAEA) 및 국제 원자력 사회에서 핵 투명성확보 및 신뢰도를 향상시키는 것을 기술개발의 목적으로 하고 있다. DUPIC 공정은 고립된 차폐시설내의 고준위 방사선장 하에서 가동되므로 타 시설에 비해 핵 물질 전용 가능성은 희박하지만, 전 공정이 원격제어 되야 하고, 조업조건이 정복해야 하므로 기존의 보장조치 기술보다 더욱 발전된 계량관리시스템, 측정시스템 및 감시시스템 등을 개발하여야 한다. 이를 위해 본 연구에서는 각 항목에 대한 요소 분석 및 각 항목별 향후 연구방향에 대해 분석하였다. DUPIC 공정 전반에 대한 핵물질 계량관리를 위해 물질수지구역 (Material Balance Area : MBA) 및 주요측정 지점 (Key Measurement Point : KMP )을 설정하여 각 측정지점별 측정방법 및 재고검증(Inventory Verification) 방법을 분석하였다. 최적 측정시스템을 개발하기 위해 적용 가능한 비파괴분석 방법들을 분석한 결과, 핵분열성 물질 함량을 정량적으로 측정할 수 있는 수동적 중성자 측정법이 가장 적합하다는 결론을 얻었다. 또한, 감시시스템을 개발하기 위해 전용전략의 주요 요소 및 전용경로 등을 분석하였으며, 핵물질 및 시설에 대한 물리적 방호체제를 DUPIC시설에 적용하기 위하여 물리적 방호에 필요한 방호체제 요소를 분석하여 DUPIC 시설을 위한 가상적인 방호체제를 구축하였다.

• Nuclear Engineering and Technology
• /
• 제26권2호
• /
• pp.277-284
• /
• 1994

폐기물 고화체로부터 주위의 Near-field로의 단일 핵종의 이동에는 많은 연구가 수행되고 있으나, 방사성 핵종 사슬의 이동에 관한 연구는 매우 제한되어 있다. 본 논문에서는 조화 유출되는 방사성 핵종 사슬의 이동을 분석하고, 유한 크기의 다공성 매질에서의 일반적인 비순환 해석해를 제시하였다. 또한 이해를 사용후 핵연료에서 가장 중요한 핵종사슬인 $^{234}$ U$\longrightarrow$$^{230}$ Th$\longrightarrow$$^{226}$ Ra 에 적용하여 보았다. 본 연구는 방사성 폐기물의 처분장 성능평가에 유용하고 중요하게 사용될 것이다.

• Nuclear Engineering and Technology
• /
• 제32권5호
• /
• pp.514-520
• /
• 2000

The mechanism of corrosion behavior has to be understood clearly to select an optimum material for handling molten salts to be used in the oxide reduction process of PWR spent fuel. In this study, the oxidation states of corrosion products on the surface of Inconel 600 and 800H as well as their chemical compositions and structural informations were determined by using XPS, ICP-AES, AAS, EPMA and XRD after the corrosion experiment with lithium molten salts at 75$0^{\circ}C$ for 25 hours. Nickel and oxygen were detected from the corrosion products on the surface of Inconel plates and chromium was found to be dissolved out into lithium molten salts leaving cracks on the surface. The corrosion products were identified as metal oxides such as Fe$_2$O$_3$, Cr$_2$O$_3$, NiO, NiFe$_2$O$_4$and MnO by using XPS and XRD.

• Nuclear Engineering and Technology
• /
• 제28권3호
• /
• pp.274-279
• /
• 1996

Diffusion coefficient is a critical parameter for predicting radiological source term(migration rate and flux of radionuclide) through given near field conditions in spent fuel or high level waste repository. The effect of exchangeable cation-$Na^+$ and $Ca^{2+} - on the diffusion of $I^- \;and^3H$ (as HTO) in compacted bentonite was examined using a through-diffusion method. Bentonite material used here was compacted to a density of 1.3 Mg/m$^3$, and Na-bentonite was saturated with a solution of 100 mol NaCl/m$^3$ and Ca-bentonite with 50 $mol\;CaCl_2$/m$^3$. The results show that effective diffusion coefficients are generally higher by a factor of two to five in Ca-than Na-clay. This is attributed to the larger particle size of Ca-compared to Na-bentonite; hence, Ca-bentonite has a greater proportion of relatively large pores, which make a greater contribution to mass transport than small pores. Although the nature of the exchangeable cation affects mass diffusion in compacted bentonite, the effect is small and not likely to influence performance assessment modeling of compacted bentonite-based barriers.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2005년도 추계학술대회논문집
• /
• pp.394-399
• /
• 2005

A case history is presented pertaining to high frequency piping vibration and noise caused by globe valve in the spent fuel pool cooling system of nuclear power plant. Frequency analyses were performed on the system to diagnose the problem and develop a solution to reduce the piping vibration and noise. The source of the high frequency and noise energy was traced to the globe valve located immediately downstream of the centrifugal pump by performing valve throttling test. Measurements of vibration and noise are presented to show that the high frequency vibration and noise amplitude was dependent upon the valve disc position and flow rate. Strouhal vortex shedding frequencies were generated at the exit of the globe valve which exited structural resonance of valve disc and amplified the high frequency vibration and noise. The problem was identified as an interaction between the flow inside globe valve and the valve disc structure. Attempts to reduce the vibration and noise amplitudes of the piping system were successfully achieved by the modification of guide-disc diameter and disc-edge figure The valve disc was replaced by an alternative to eliminate the source of the harmful high frequency vibration and noise.