• 방사성폐기물학회지
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• 제19권3호
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• pp.307-322
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• 2021

This study developed an analytical methodology for the mechanical integrity of spent nuclear fuel (SNF) cladding tubes under external pinch loads during transportation, with reference to the failure mode specified in the relevant guidelines. Special consideration was given to the degraded characteristics of SNF during dry storage, including oxide and hydride contents and orientations. The developed framework reflected a composite cladding model of elastic and plastic analysis approaches and correlation equations related to the mechanical parameters. The established models were employed for modeling the finite elements by coding their physical behaviors. A mechanical integrity evaluation of 14 × 14 PWR SNF was performed using this system. To ensure that the damage criteria met the applicable legal requirements, stress-strain analysis results were separated into elastic and plastic regions with the concept of strain energy, considering both normal and hypothetical accident conditions. Probabilistic procedures using Monte Carlo simulations and reliability evaluations were included. The evaluation results showed no probability of damage under the normal conditions, whereas there were small but considerably low probabilities under accident conditions. These results indicate that the proposed approach is a reliable predictor of SNF mechanical integrity.

• Nuclear Engineering and Technology
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• 제35권5호
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• pp.378-386
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• 2003

Recently failures of nuclear fuel rods in Korean nuclear power plants were reported and their failure causes have been investigated by using PIE techniques. Destructive and physico-chemical examinations reveal that the clad hydriding phenomena had caused the rod failures primarily and secondarily in each case. In this study, the basic mechanisms of the primary and the secondary hydriding failures are reviewed, PIE data such as cladding inner and outer surface oxide thickness and the restructuring of the fuel pellets are analyzed, and they are compared with the predicted behaviors by a fuel performance code. In addition, post-defected fuel behaviors are reviewed and qualitatively analyzed. The results strongly support that the hydriding processes, primary and secondary, played critical roles in the respective fuel rods failures and the secondary hydriding failure can take place even in the fuel rod with low linear heat generation rate.

• Nuclear Engineering and Technology
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• 제34권6호
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• pp.574-585
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• 2002

The microstructural characteristics and its related mechanical properties of RPV cladding have been investigated using small punch (SP) tests. SA508 Cl.3 RPV steel plates were overlay cladded with the type ER309L welding consumables by submerged arc welding process. Although the RPV clad material had a small portion of 5 ferrite phase, it still showed the ductile to brittle transition behavior The transition temperature was determined by the SP test and it depended on the content of $\sigma$ phase, specimen size, and determination methods. The fracture appearance of SP specimen was changed from circumferential to radial cracking as test temperature became low, and below the transition temperature region, ER309L cladding usually fractured along the 6 ferrite by the low temperature failure of ferrite phase.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• Nuclear Engineering and Technology
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• 제20권4호
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• pp.241-245
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• 1988

핵연료봉의 PCI 파손은 원자로의 운전제한과 밀접한 관계가 있기 때문에, 출력급증 조건에서 핵연료봉의 PCI 파손을 지배하는 파손인자들의 거동을 검토하는 것은 매우 중요하다. 본 연구에서는 피복관에서의 원주방향 응력, 원주방향 변형도, 원주방향 주름 높이, 크립 변형율 및 변형도 에너지등의 파손인자들에 대한 거동특성을 핵연료봉 성능해석용 전산코드인 FEMAXI-IV를 이용하여 출력급증량 및 출력증가율의 운전인자들의 함수로 검토하였다.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2012년도 추계학술대회 논문집
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• pp.204-205
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• 2012

• Nuclear Engineering and Technology
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• 제28권1호
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• pp.17-26
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• 1996

프레팅마모 기인 연료봉 손상을 방지할 수 있는 노내 연료봉 지지조건은 잔여 지지격자스프링 변위량 또는 연료봉 /지지격자 갭에 의해 평가될 수 있다. 핵연료 설계 인자들이 프레팅마모 손상에 미치는 영향을 평가하기 위해 연소도의 함수로서 노내 연료봉 지지조건을 모사할 수 있는 방법론을 사용하여 GRID-FORCE프로그램을 개발하였다. 이 프로그램에서는 노내 연료봉 지지조건에 영향을 주는 주요 인자로서 피복관 크립, 초기 스프링 변위, 초기 스프링힘 그리고 스프링힘 조사이완이 고려된다. 이 주요 인자들에 대한 민감도 분석 결과, 초기 스프링 변위, 스프링힘 조사이완, 피복관 크립 순으로 노내 연료봉 지지조건에 영향을 주는 것으로 나타났다. 이 프로그램을 실제 노내에서 발생한 프레팅마모 기인 연료봉 손상에 적용한 결과를 토대로 판단해 볼 때 이 프로그램을 새로 개발된 피복관 재질 및 /또는 새로 개발된 지지격자 설계가 프레팅마모 기인 연료봉 손상을 방지할 수 있는 설계여유도를 효과적으로 평가할 수 있음을 알 수 있다.

• Nuclear Engineering and Technology
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• 제16권3호
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• pp.169-179
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• 1984

FIPREL 전산코드를 사용하여 원자로 냉각수 내의 핵분열 생성물에 의한 방사능을 분석함으로써 PWR의 운전시에 발생하는 핵연료 피복관 파손을 평가할 수 있는 효과적인 절차를 모색하였다. 이 코드를 이용하여 핵연료의 농축도, 연소도, 가동온도 및 갭유출계수의 크기로 정량화되는 실제적 파손 크기등의 물리적 파라미터에 대해서 핵분열 생성물의 방사능이 나타내는 민감도에 대한 방대한 계산을 실시하였으며 그 결과는 PROFIP방법에 의한 것과 대체적으로 일치한다. 노출 우라늄이 존재하는 경우에는 옥소보다도 화학적으로 더 안정된 핵종간의 방사능비에 근거하여 반복계산을 실시함으로써 파손된 핵연료 봉에서 유출된 방사능만을 분리해 낸다. 개발된 전산코드로 파손 핵연료봉의 선형출력 밀도, 갯수, 실제적 파손 크기 및 노출우라늄의 질량등을 계산할 수 있다. 고리 1호기의 4주기에 걸친 운전 경험을 이 모텔에 의해 분석한 결과에 의하면 본 모델은 원자력발전소 정상운전시 핵연료봉의 상태를 감시·평가하는데 아주 적합한 것으로 판명되었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.97-102
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• 1996

A methodology for estimating the number of failed fuel rods bused on the primary coolant activity in operating PWRs has been developed. This method deals with both the diffusion and the kinetic models. In case of small or medium cladding failures, the diffusion model which can consider different sizes of failure is used, whereas for large cladding failures the kinetic model is used. From the kinetic model, the release-to-birth rate ratio (R/B) is represented as a linear function of the number of failed fuel rods. This has been done by expressing the escape rate coefficient in terms of the slope of log(R/B) versus $log\;{\lambda}$. The present method has been applied to the cases of 26 cycles of several nuclear power plants for which ultrasonic testings were performed. The results show that the present method gives better predictions than the existing computer codes such as IODYNE and CADE.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2638-2651
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• 2020

Nozzle corner cracks present at the intersection of reactor pressure vessels (RPVs) and inlet or outlet nozzles have been a persistent problem for a number of years. The fracture analysis of such nozzle corner cracks is very important and critical for the efficient design and assessment of the structural integrity of RPVs. This paper aims to perform an engineering critical assessment of RPVs with nozzle corner cracks subjected to several transients accompanied by pressurized thermal shocks. The critical crack size of the RPV model with nozzle corner cracks under transient loading is evaluated on failure assessment curve. In particular, the influence of cladding on the crack initiation of nozzle corner crack under thermal transients is studied. The influence of primary internal pressure and secondary thermal stress on the stress field at nozzle corner and SIF at crack front is analyzed. Finally, the influence of different crack size and crack shape on the final critical crack size is analyzed.