• 비파괴검사학회지
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• 제29권3호
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• pp.256-263
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• 2009

After several PWSCCs were found in Bugey(France), Ringhals(Sweden), Tihange(Belgium), Oconee, Arkansas, Crystal Fever, Davis-Basse, VC Summer(U.S.A.), Thuruga(Japan), USNRC and PNNL started the research on PWSCC, that is, the PINC project. USNRC required KINS to participate in the PINC project in May 2005. KINS organized the Korean consortium at March 2006 and Pre-RRT for RVHP were performed for the preparation of PINC RRT. Through these preliminary RRT, Korea NDE teams can learn and develop the detection and sizing technique for RVHP dissimilar metal weld. These techniques are now being prepared in Korea and need to be utilized for the In-service inspection of the RVHP and BMI of Korea Nuclear Power Plants. PINC RRT mock-ups will be helpful to training.

• 대한기계학회논문집A
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• 제22권2호
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• pp.311-320
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• 1998

The cryogenic fracture behaviors of austenitic stainless steel HN2 developed for nuclear fusion reactor were evaluated quantitatively by using the small punch(SP) test. The electrochemical polarization test was applied to study thermal aging degradation of HN2 steel. The X-ray diffraction(XRD) analysis was conducted to detect carbides and nitrides precipitated on the grain boundary of the heat treated HN2 steel. The mechanical properties of the HN2 steel significantly decreased with increasing time and temperature of heat treatment or with decreasing testing temperature. The integrated charge(Q) obtained from electrochemical polarization test showed a good correlation with the SP energy(ESP) obtained by means of SP tests. From the results observed in the x-ray diffraction and anodic polarization curve, it was known that the material the grain boundary. Combining SP test and electrochemical polarization test, it could be useful tools to non-destructively evaluate the cryogenic fracture behaviors and the aging degradation for cryogenic structural material.

• 대한기계학회논문집A
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• 제37권9호
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• pp.1127-1132
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• 2013

원자로 내부구조물은 파손시 원자로 안전 운전/정지에 주요한 영향을 미칠 수 있으며 중성자 조사 수준이 높아 중성자 조사와 관련된 다양한 열화가 발생하였거나 잠재적으로 발생할 수 있다. 원자로 내부구조물의 주요 재질인 오스테나이트 스테인리스 강은 중성자 조사에 따라 인장/크리프 물성, 파괴인성 등 기계적 재료 거동에 변화가 발생한다. 각종 열화기구에 대한 원자로 내부구조물의 구조 건전성이 설계수명 또는 계속운전 기간 동안 유지됨을 평가할 때 중성자 조사에 따른 기계적 재료거동의 변화를 고려하여야 한다. 본 연구에서는 중성자 조사에 따른 기계적 재료거동의 변화를 고려한 사용자 정의 보조 프로그램을 개발하였다. 개발된 사용자 정의 보조 프로그램을 다양한 조건에 대해 검증한 결과, 타당함을 확인하였다.

• Journal of Welding and Joining
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• 제16권3호
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• pp.121-128
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• 1998

The degradation of mechanical properties in the high nitrogen steel HN3 developed for nuclear fusion reactor has been evaluated quantitatively using the small punch(SP) test, X-ray diffraction (XRD) analysis has also been conducted to identify carbides or nitrides precipitated on grain boundaries of the heat treated samples. Mechanical properties of the steel HN3 significantly decreased with increasing heat treatment time and temperature or with decreasing testing temperature. Combination of XRD and metallurgical observation, revealed that the material degradation in the thermally aged steel was caused by precipitation of carbides on the grain boundaries. While the weld metal showed the lowest mechanical properties among various microstructures in GTA weldments. By combining SP test and XRD analysis, cryogenic fracture behaviors and aging degradation for high nitrogen steel could be successfully evaluated in nondestructive manner.

• Nuclear Engineering and Technology
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• 제42권2호
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• pp.203-210
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• 2010

The $\underline{H}igh$ flux $\underline{A}dvanced$ $\underline{N}eutron$ $\underline{A}pplication$ $\underline{R}eact\underline{O}r$ (HANARO), an open-tank-in-pool type reactor, is one of the multi-purpose research reactors in the world. Since the commencement of HANARO's operations in 1995, a significant number of experimental facilities have been developed and installed at HANARO, and continued efforts to develop more facilities are in progress. Owing to the stable operation of the reactor and its frequent utilization, more experimental facilities are being continuously added to satisfy various fields of study and diverse applications. The irradiation testing equipment for nuclear fuels and materials at HANARO can be classified into capsules and the Fuel Test Loop (FTL). Capsules for irradiation tests of nuclear fuels in HANARO have been developed for use under the dry conditions of the coolant and materials at HANARO and are now successfully utilized to perform irradiation tests. The FTL can be used to conduct irradiation testing of a nuclear fuel under the operating conditions of commercial nuclear power plants. During irradiation tests conducted using these capsules in HANARO, instruments such as the thermocouple, Linear Variable Differential Transformer (LVDT), small heater, Fluence Monitor (F/M) and Self-Powered Neutron Detector (SPND) are used to measure various characteristics of the nuclear fuel and irradiated material. This paper describes not only the status of HANARO and the status and perspective of irradiation devices and instrumentation for carrying out nuclear fuel and material tests in HANARO but also some results from instrumentation during irradiation tests.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2139-2146
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• 2023

During reactor operation, the divertor must withstand unprecedented simultaneous high heat fluxes and high-energy neutron irradiation. The extremely severe service environment of the divertor imposes a huge challenge to the bonding quality of divertor joints, i.e., the joints must withstand thermal, mechanical and neutron loads, as well as cyclic mode of operation. In this paper, potassium-doped tungsten (KW) is selected as the plasma facing material (PFM), oxygen-free copper (OFC) as the interlayer, oxide dispersion strengthened copper (ODS-Cu) alloy as the heat sink material, and reduced activation ferritic/martensitic (RAFM) steel as the structural material. In this study, a vacuum brazing technology is proposed and optimized to bond Cu and ODS-Cu alloy with the silver-free brazing material CuSnTi. The most appropriate brazing parameters are a brazing temperature of 940 ℃ and a holding time of 15 min. High-quality bonding interfaces have been successfully obtained by vacuum brazing technology, and the average shear strength of the as-obtained KW/Cu and ODS-Cu alloy joints is ~268 MPa. And a fabrication route for manufacturing the flat-type divertor target based on brazing technology is set. For evaluating the reliability of the fabrication technologies under the reactor relevant condition, the high heat flux test at 20 MW/m² for the as-manufactured flat-type KW/Cu/ODS-Cu/RAFM mockup is carried out by using the Electron-beam Material testing Scenario (EMS-60) with water cooling. This paper reports the improved vacuum brazing technology to connect Cu to ODS-Cu alloy and summarizes the production route, high heat flux (HHF) test, the pre and post non-destructive examination, and the surface results of the flat-type KW/Cu/ODS-Cu/RAFM mockup after the HHF test. The test results demonstrate that the mockup manufactured according to the fabrication route still have structural and interfacial integrity under cyclic high heat loads.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.791-798
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• 2012

Since standardized fracture test specimens cannot be easily extracted from in-service components, several alternative fracture toughness test methods have been proposed to characterize the deformation and fracture resistance of materials. One of the more promising alternatives is the local approach employing the SP(Small Punch) testing technique. However, this process has several limitations such as a lack of anisotropic yield potential and tediousness in the damage parameter calibration process. The present paper investigates estimation of ductile fracture resistance(J-R) curve by FE(Finite Element) analyses using an anisotropic damage model and enhanced calibration procedure. In this context, specific tensile tests to quantify plastic strain ratios were carried out and SP test data were obtained from the previous research. Also, damage parameters constituting the Gurson-Tvergaard-Needleman model in conjunction with Hill's 48 yield criterion were calibrated for a typical nuclear reactor material through a genetic algorithm. Finally, the J-R curve of a standard compact tension specimen was predicted by further detailed FE analyses employing the calibrated damage parameters. It showed a lower fracture resistance of the specimen material than that based on the isotropic yield criterion. Therefore, a more realistic J-R curve of a reactor material can be obtained effectively from the proposed methodology by taking into account a reduced load-carrying capacity due to anisotropy.

• 한국기계가공학회지
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• 제11권4호
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• pp.160-167
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• 2012

This paper is concerned with a particular fracture mechanics parameter ${\Delta}K_{th}$, known as the 'threshold stress intensity range', or 'fatigue threshold'. This threshold ${\Delta}K_{th}$ constitutes, as it were, a hinge between the notion of crack initiation and the notion of crack growth. It has often been thought that, like the endurance limit, it could be an intrinsic criterion of the material. The study was conducted on a SA-508 pressure vessel steel used in the nuclear power industry. This material exhibits a typical threshold effect in the range of the crack growth rates which were determined; that is, below approximately $da/dN=10^{-6}mm/cycle$, the slope of the da./dN versus ${\Delta}K$ curve is almost vertical. The value of ${\Delta}K_{th}$ was determined at a growth rate of $10^{-7}$ mm/cycle according to the ASTM Standard for threshold testing. The fatigue threshold values are in the range 21 $kg/mm^{3/2}$ to 12 $kg/mm^{3/2}$ depending on the stress ratio effect.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2011년도 정기 학술대회
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• pp.415-418
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• 2011

Thermal fatigue is one of the life-limiting damage mechanisms in the nuclear power plant conditions. The turbulent mixing of fluids of different temperatures induces rapid temperature changes to the pipe wall. The successive thermal transients cause varying cyclic thermal stresses. These cyclic thermal stresses cause fatigue crack nucleation and growth similar to the cyclic mechanical stresses. The aim of this study was to fulfil the need by developing an real crack manufacturing method, which would produce realistic cracks. The test material was austenitic STS 304, which is used as pipelines in the reactor coolant system of a nuclear power plants. In order to fabricate thermal fatigue crack similar to realistic crack, successive thermal transients were applied to the specimen. Thermal transient cycles were combined with heating (60sec) and cooling cycle (30sec). And, In order to identify ultrasonic characteristic, it was performed the ultrasonic reflection measuring method for the fabricated specimen. From the results of ultrasonic reflection measuring testing, it was conformed that A-scan results(average 83% of real crack depth) for the TFC reference specimen was more enhanced NDT reliability than results(average 38% of real crack depth) for the EDM notch reference specimen.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.499-507
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• 2020

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.