• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.118-123
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• 2004

Cast austenitic stainless steel have been widely used for primary coolant piping in light water reactors. This material is subject to thermal embrittlement at reactor operating temperature. CF-8M and CF-8A cast austenitic stainless steel is used for several components, such as primary coolant piping, elbow, pump casing, and valve bodies in light water reactors. Thermal embrittlement results in spinodal decomposition of delta-ferrite leading to decreased fracture toughness. In this study, the specimens were prepared using an accelerated aging method. The measurement of ferrite content, Charpy impact test and J-R test were performed to verify the predicting equation for aged material properties. In case of above 25% ferrite content, predicted result of J-R curve might be non-conservative.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.131-136
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• 2003

CF8M cast austenitic stainless steel is used for several components such as primary coolant piping, elbow, pump casing, and valve bodies in light water reactors. These components are subject to thermal aging at the reactor operating temperature. Thermal aging results in spinodal decomposition of the delta-ferrite leading to increased strength and decreased toughness. In this study, three kinds of the aged CF8M specimen were prepared using an artificially simulated aging method. The objective of this study is to summarize the method of estimating ferrite contents, Charpy impact energy and J-R curve, and to evaluate the thermal embrittlement of the CF8M cast austenitic stainless steel piping used in the domestic nuclear power plants.

• 대한기계학회논문집A
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• 제38권1호
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• pp.11-16
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• 2014

CF8M (11% ferrite) 주조 스테인리스강의 $310^{\circ}C$ 순수환경에서의 저주기피로 수명에 미치는 수소 및 미세구조의 영향을 분석하였다. CF8M 의 경우, 공기환경 대비 $310^{\circ}C$ 순수환경에서의 피로수명의 감소는 단조재인 316LN 에 비해 다소 작았다. 미세구조 및 파면 분석을 통해, CF8M 의 저주기피로 수명의 감소는 316LN 의 경우와 마찬가지로 수소유기균열에 의한 것으로 판단되었다. 그러나, CF8M 의 경우, 페라이트상 경계에 수소유기균열에 의한 2 차 균열이 빈번히 발생함에 따라 균열 선단에서의 응력집중이 저하되는 효과가 있었다. 이러한 응력집중의 완화로 인해 수소유기균열에 의한 피로균열진전이 둔화되어 결과적으로 저주기피로 수명의 저하가 완화되는 것으로 판단되었다.

• 대한기계학회논문집A
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• 제30권10호
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• pp.1173-1178
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• 2006

In a primary reactor cooling system(RCS), a dissimilar weld zone exists between cast stainless steel(CF8M) in a pipe and low-alloy steel(SA508 cl.3) in a nozzle. Thermal aging is observed in CF8M as the RCS is exposed for a long period of time to a reactor operating temperature between 290 and $330^{\circ}C$, while no effect is observed in SA508 cl.3. The specimens are prepared by an artificially accelerated aging technique maintained for 300, 1800 and 3600 hrs at $430^{\circ}C$, respectively. The specimens for elastic-plastic fracture toughness tests are according to the process in the thermal notch is created in the heat affected zone(HAZ) of CF8M and deposited zone. From the experiments, the $J_{IC}$ value notched in HAZ of CF8M presented a rapid decrease up to 300 hours at $430^{\circ}C$ and slowly decreased according to the process in the thermal aging time. Also, the $J_{IC}$ value presented a lower value than that of the CF8M base metal. And, the $J_{IC}$ of the deposited zone presented the lowest value of all other cases.

• Nuclear Engineering and Technology
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• 제43권1호
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• pp.83-88
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• 2011

To define the effect of strain rate variation from 0.04% to 0.004%/s on environmental fatigue of CF8M cast stainless steel, which is used as a primary piping material in nuclear power plants, low-cycle fatigue tests were conducted at operating pressure and temperature condition of a pressurized water reactor, 15 MPa and $315^{\circ}C$, respectively. A high-pressure and high-temperature autoclave and cylindrical solid fatigue specimens were used for the strain-controlled low-cycle environmental fatigue tests. It was observed that the fatigue life of CF8M stainless steel is shortened as the strain rate decreases. Due to the effect of test temperature, the fatigue data of NUREG-6909 appears a slightly shorter than that obtained by KEPRI at the same stress amplitude of $1{\times}10^3$ MPa. The environmental fatigue correction factor $F_{en}$'s calculated with inputs of the test data increases with high strain amplitude, while the $F_{en}$'s of NUREG-6909 remain constant regardless of strain amplitude.

• 대한기계학회논문집A
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• 제23권11호
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• pp.2014-2021
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• 1999

The present investigation is concerned with the degradation characteristics of cast stainless steel(CF8M), exposed to the $\sigma$-phase degradation at $700^{\circ}C$. In the present paper, the degradation of CF8 M at $700^{\circ}C$ is evaluated by a non-destructive test, DL-EPR(double loop electrochemical potentiokinetic reactivation). The surface of specimens is observed by using scanning electron microscopy after DL-EPR test. Also. chromium contents of matrix, grain boundary and ferrite phase are analyzed by electron probe X-ray micro analyzer. Through the experiments, the following results are obtained 1) The degree of sensitization(DOS) of CF8M aged up to 15hr at $700^{\circ}C$ is increased with acing time while that is decreased with aging time from 15hr to 150hr. 2) The impact energy decreases with increase of $\sigma$-phase while DOS increases with $\sigma$-phase until aging time reaches to 15hr. After the aging time. 15hr, the $\sigma$-phase and the rate of impact energy with respect to aging time decreases. Therefore the degradation behavior of the CF8M can be evaluated by comparing SEM micrographs and the value of DOS.

• 대한기계학회논문집A
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• 제36권7호
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• pp.769-776
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• 2012

본 논문의 유한요소 손상해석기법은 재료 인장물성과 파괴기준을 필요로 한다. 기존에 연구된 재료들은 노치인장 시험결과로부터 하나의 인장물성과 파괴기준을 구할 수 있었다. 그러나 본 논문에서 사용된 cast stainless steel(CF8M)의 경우, 동일한 조건의 노치인장 시험결과들에 분산이 존재하여 해석자에 따라 다른 인장물성과 파괴기준이 구해질 수 있다. 따라서 해석자에 관계없이 일관된 인장물성 및 파괴기준을 구할 수 있는 적절한 절차가 필요하다. 본 논문에서는 노치반경 16mm 의 인장시편 시험결과로부터 평균 인장물성을 구하였고, 이를 유한요소 해석에 적용하여 3 개의 파괴기준을 구하였다. 구해진 인장물성과 파괴기준을 적용하여 J-R 파괴인성 시험에 대한 손상해석을 수행하였고, 시험결과와 비교함을 통해 제시된 절차의 타당성을 검증하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.17-22
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• 2007

Low-cycle environmental fatigue tests of cast austenitic stainless steel CF8M at the condition of fatigue strain rate 0.04%/sec were conducted at the pressure and temperature, 15MPa, $315^{\circ}C$ of a operating pressurized water reactor. The used test rig was limited to install an extensometer at the gauge length of the cylindrical fatigue specimen inside the small autoclave. So the magnet type LVDT's were used to measure the fatigue displacement at the specimen shoulders inside the high temperature and high pressure water autoclave. However, the displacement and strain measured at the specimen shoulders is different from the one at the gauge length for the geometry and the cyclic strain hardening effect. FEM calculated the displacement and the strain of the gauge length from the data measured at the shoulders. Tensile test properties in elastic and plastic behavior of CF8M material were used in the FEM analysis. A series of low cycle fatigue tests simulating the cyclic strain hardening effect verified that the FEM calculation was well agreed with the simulated tests. The process and method developed in this study would be so useful to produce reliable environmental fatigue curves of CF8M stainless steel in pressurized water reactors.

• 대한기계학회논문집A
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• 제28권12호
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• pp.1968-1973
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• 2004

Structural degradations are often experienced on the components of nuclear power plants in reactor pressure vessels (RPV) and steam generators (SG) when these components are exposed to high temperature and high pressure for a long period of time. Such conditions result in the change of microstructures and of mechanical properties of materials, which requires an evaluation of the safeguards related to structural integrity. In a primary reactor cooling system (RCS), a dissimilar weld zone exists between cast stainless steel (CF8M) in a pipe and low-alloy steel (SA508 cl.3) in a nozzle. Thermal aging is observed in CF8M as the RCS is exposed for a long period of time under the operating temperature between 290 and 33$0^{\circ}C$. Under the same conditions, it is well known that degradation is not observed in low alloy steel. An investigation of the effect of thermal aging on the various mechanical properties of the dissimilar weld zone is required. The purpose of the present investigation is to find the effect of thermal aging on the dissimilar weld zone. The specimens are prepared by an artificially accelerated aging technique maintained for various times at 43$0^{\circ}C$, respectively. Then, The various mechanical test for the dissimilar welds are performed.

• 대한기계학회논문집A
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• 제32권2호
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• pp.177-185
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• 2008

Low-cycle environmental fatigue tests of cast austenitic stainless steel CF8M at the condition of fatigue strain rate 0.04%/sec were conducted at the pressure and temperature, 15MPa, $315^{\circ}C$ of a operating pressurized water reactor (PWR). The used test rig was limited to install an extensometer at the gauge length of the cylindrical fatigue specimen inside a small autoclave. So the magnet type LVDT#s were used to measure the fatigue displacement at the specimen shoulders inside the high temperature and high pressure water autoclave. However, the displacement and strain measured at the specimen shoulders is different from the one at the gauge length for the geometry and the cyclic strain hardening effect. Displacement of the fatigue specimen gauge length calculated by FEM (finite element method) used to modify the measured displacement and fatigue life at the shoulders. A series of low cycle fatigue life tests in air and PWR conditions simulating the cyclic strain hardening effect verified that the FEM modified fatigue life was well agreed with the simulating test results. The process and method developed in this study for the environmental fatigue test inside the small sized autoclave would be so useful to produce reliable environmental fatigue curves of CF8M stainless steel in pressurized water reactors.