• Journal of Welding and Joining
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• v.16 no.4
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• pp.55-62
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• 1998

While excellent joint quality has been obtained using dry chamber underwater welding methods, the size limitations imposed by this process restrict its use for underwater construction work. The wet underwater shielded metal-arc welding eliminates this restriction but suffers from poor weld properties by the 1-pass bead-on-plate welding due to the excessive diffusible hydrogen. On the other hand, in the wet underwater welding, it is well known that the quantity of diffusible hydrogen in multi-pass welded parts reduce to less than that in 1-pass welded parts. Therefore, in this paper, welding experiments are made the 3-pass bead-on-plate welds by using TMCP and normalized steel plates and E4301 and cellulose coated electrode. After that, The amounts of the hydrogen absorbed into the 3-pass welded area were measured according to the JIS Z 3118 specification. The microstructural changes as well as the microhardness distribution after the underwater 3-pass welding were also investigated using Vickers microhardness tester and S.E.M and O.M. The results indicated that the quantity of diffusible hydrogen in 3-pass welded areas was reduced little less than a half of one of that in 1-pass welded areas at the specific welding condition. As a result, the cold cracking of 3-pass welded areas decreased by reduced effect of diffusible hydrogen. In the underwater 3-pass welding, the micrography of cold cracking fracture surface showed mainly the cleavage of hydrogen embrittlement.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.168-171
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• 2003

The IMEF(Irradiated Materials Examination Facility), located in KAERI site, is a hot cell facility to test and evaluate the irradiation defects or embrittlement through post-irradiation examination(PIEs) of irradiated nuclear fuels and structural materials which are come from HANARO research reactor and commercial nuclear power plants. Therefore, to carry out its own function, the high level solid radioactive wastes, produced through PIEs, are periodically carried out and managed from hot cell to monolith. So far, approximately 30 drums which contains 50 liters are transported to monolith, and it is shown that the quantity is slowly increasing, In this paper, the procedures and work contents of the high level solid radwaste carrying out and management for IMEF are described in detail.

• Korean Journal of Materials Research
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• v.28 no.5
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• pp.286-294
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• 2018

Hot-press forming(HPF) steel can be applied successfully to auto parts because of its superior mechanical properties. However, its resistances to aqueous corrosion and the subsequent hydrogen embrittlement(HE) decrease significantly when the steel is exposed to corrosive environments. Considering that the resistances are greatly dependent on the properties of coating materials formed on the steel surface, the characteristics of the corrosion and hydrogen diffusion behaviors regarding the types of coating material should be clearly understood. Electrochemical polarization and impedance measurements reveal a higher corrosion potential and polarization resistance and a lower corrosion current of the Al-coating compared with Zn-coating. Furthermore, it was expected that the diffusion kinetics of the hydrogen atoms would be much slower in the Al-coating, and this would be due mainly to the much lower diffusion coefficient of hydrogen in the Al-coating with a face-centered cubic structure. The superior surface inhibiting effect of the Al-coating, however, is degraded by the formation of local cracks in the coated layer under severe stress conditions, and therefore further study will be necessary to gain a clearer understanding of the effect of cracks formed on the coated layer on the subsequent corrosion and hydrogen diffusion behaviors.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.74-83
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• 1995

The accurate determination of the fast neutron flux/fluence onto the pressure vessel is an essential part of the reactor pressure vessel surveillance program. It has been reported recently that the iron cross section data in ENDF/B versions III through V might underestimate the flux/fluence of fast neutrons in steel structures such as reactor pressure vessel. In this study, for the comparison of iron data of ENDF/B-IV and VI we produced two 47-group cross section sets, CXFe-IV and CXFe-Ⅵ, which are based on Yonggwang nuclear unit-3/4 model and the iron data of ENDF/B-IV and VI, respectively. A comparison was made of the results obtained from DOT4.3 calculation using CXFe-IV and CXFe-VI. From the results, it was found that the fast flux(E 〉 1.0 MeV), which is important for the pressure vessel embrittlement analysis, increases by about 7.6％ at the inner wall and 20％ at the outer wall of the vessel, if the iron data are used from ENDF/B-VI instead of ENDF/B-IV.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.29 no.4
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• pp.347-354
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• 2016

Because typical gas turbine systems have frequent startup and shutdown operations, it is likely to cause cracks at the gas turbine casing and gas leakages at casing flanges due to thermal fatigue and embrittlement. Therefore, the evaluation of structural integrity and gas leakage at the gas turbine casings must be performed. In this paper, we have evaluated the structural integrity of the turbine casing and bolts under a normal operation in accordance with ASME B&PVC and evaluated the leakage at casing flanges by examination of contact pressure calculated using the finite element analysis. Finally, we propose a design flow including finite element modeling, the interpretation and evaluation methods for gas turbine casings. This may be utilized in the design and development of gas turbine casings.

• Proceedings of KOSOMES biannual meeting
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• 2005.05a
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• pp.161-165
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• 2005

Recently, it is on the increase interest for Al alloy with new material for ship application to substitute for FRP ship. The reason is thatAl alloy ship has beneficial characteristics such as high sea speed, increase of loadage and easy to recycle compared with FRP ship. In this paper, mechanical and electrochemical properties are investigated by slow strain rate test experiment in various applied potential condition. These results will provide as reference data to design ship by deciding optimum protection potential regard to hydrogen embrittlement and stress corrosion cracking. In general, Al and Al alloys are not corroded with forming film which has the corrosion resistance property in neutral solution. However, it was observed that formation and destruction of passive film by $Cl^-$ ion in sea water environment. At comparison of current density after 1200 sec in potentiostatic experiment, the current density in the potential range of -0.68 $\~$-1.5 V is shown low value. The low current density means protection potential range. Elongation in applied potential of 0 V was high. However, the corrosion protection application in this condition is impossible potential because the toughness is low value by decreasing strength by active dissolution reaction at parallel part of specimen. The film composed with $CaCO_3$ and $Mg(OH)_2$ has a corrosion resistance property. However, the uniform electrodeposition coating at below -1.6 V potential is not formed since the time to form the uniform electrodeposition coating is short. Therefore, it is concluded that mechanical property is poor because effect by hydrogen gas generation is larger than that of electrodeposition coating. It is concluded that the optimum protection potential range from comparison of_maxim urn tensile strength, elongation and time to fracture is -1.3$\~$0.7 V (SSCE).

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.16-25
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• 2016

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconiumalloy fuel claddingmaterials are rapidlyheateddue to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF) design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI) is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in $1,200-1,500^{\circ}C$ steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstratedcorrosionresistance.Asthese composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Moalloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR) challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.

• Journal of the Korean Magnetics Society
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• v.8 no.5
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• pp.249-254
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• 1998

In relation to the application of magnetic method to the evaluation of irradiation damage (embrittlement) changes in the magnetic parameters(hysteresis loop and Barkhausen noise) and Vickers microhardness due to neutron irradiation and heat treatment were measured and compared. In the case of irradiation $(2.3{\times}10^{19}\;n/cm^2,\; E{\ge}1\;Mev,\; 288{\circ}C)$ hysteresis loop measurements show that susceptibility decreases as coercivity increase. Saturation magnetization do not show any change. Barkhausen noise amplitude and Barkhausen noise energy have decreased while Vickers microhardness has increased. For isothermally heat treated condition of irradiated specimen at 470 $^{\circ}C$ and 540 $^{\circ}C$, Barkhausen noise energy has increased while Vickers microhardness has decreased. Results of BNE and Vickers microhardness are reversed to the results on irradiated condition. All these consistent changes in magnetic parameter and Vickers microhardness measurement, which are thought to be resulted from the interaction between irradiation-induced defects and dislocation, and magnetic domain, respectively, show a possibility that magnetic measurement may be used to the evaluation of material degradation and recovery due to neutron irradiation and heat treatment, respectively, if a relevant large database in prepared.

• Proceedings of the Korean Institute of Surface Engineering Conference
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• 1998.05a
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• pp.22-22
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• 1998

304 는 BWR(boiling water reactor)의 reactor 구조용 재료로 사용되고 있고, 합금 600 은 PWR(pressurized water reator) 의 증기 발생기 세관으로 쓰이고 있으며 모두 약 $280{\;}^{\circ}C$ 이상 의 원자로 냉각수에 노출되어 있다. 원자로 냉각수 분위기에서 두 합금의 공통적인 특정은 입계응력부식균열(IGSCC)에 민감한것과 IGSCC가 예민화(sensitization)와 관련이 있는 것이 다. 두 합금에서 일어나는 IGSCC는 원자력발전소의 부식피해중 가장 빈도가 높고 발생시 방사능 누출로 인하여 원전의 신뢰성을 저하시키고, 가동중단으로 인한 경제적 손실을 초 래하여 지난 20 년 동안 가장 심도있게 연구된 주제다. 304 은 크롬 탄화물의 업계 석출로 언하여 예민화된경우 IGSCC 에 민감한 반면 600 은 예민화된 경우 뿐만 아니라 용체화처리된 상태에서도 IGSCC에 민감하다. 오히려 600은 용 체화처리 후 700 C에서 15~20시간 시효처리를 하여 크롬탄화물을 업계에 석출 시커었을 때 IGSCC 저항성이 향상된다. 두 합금의 IGSCC 특정 중 큰 차이는 304는 임계균열전위 ( (critical cracking potential) 이 존재하여 부식전위(corrosion potential) 가 엄계균열전위보다 낮 은 경우 IGSCC 가 일어나지 않지만 그 반대인 경우 IGSCC 에 민감하게된다. 반면에 600 은 뚜렷한 임계균열전위가 존재하지 않고 양극 분극(anodic polarization) 뿐만 아니라 음극분극 시에도 IGSCC 가 일어난다. 이련 이유로 600의 IGSCC 가구로 피막파괴-양극용해(film rupture-anodic dissolution)외에 수소취성(hydrogen embrittlement)기구도 제안되고 었다. 원전의 냉각수는 고 순도의 물이지만 수 처리 과정과 웅축기 배관의 누수로 인한 산소, $Cu^{2+},{\;}S_xO_6{\;}^{2-}(x=3~6)$ 등이 유입되어 오염되는데 이려한 오염물질들이 수 ppm정도 소량 포함된 경우 응 력부식민감도는 상당히 증가된다. 산성분위기 흑은 산소, $Cu^{2+}$, 등이 소량 포합된 산화성 분위기 그리고 sufur oxyanion 에 오염된 고온의 물에서 600 의 IGSCC 민감도는 예민화도가 증가할 수록 민감하여 304 의 IGSCC 와 매우 유사한 거동을 보인다. 본 강연에서는 304 와 600 의 고온 물에서 일어나는 IGSCC 민감도에 미치는 환경, 예민화처리, 합금원소의 영향을 고찰하고 이에 대한 최근의 연구 동향과 방식 방법을 다룬다.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1423-1432
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• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.