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황성식 ( Seong Sik Hwang ),권준현 ( Junhyun Kwon ),김동진 ( Dong Jin Kim ),김성우 ( Sung Woo Kim ) 한국부식방식학회 2021 Corrosion Science and Technology Vol.20 No.5
Neutron dose level at bottom head of a reactor pressure vessel (RPV) was calculated using reactor vessel neutron transport for a Korean nuclear power plant A. At 34 EFPY with a 40-year (2042) design life after plating repair, irradiation fast neutron effect was 6.6x10<sup>15</sup> n/Nuclear power plant, Corrosion of RPV steel, Ni plating, Neutron effect, He generation. As helium(He) gas can be generated by Ni only at 1/10<sup>6</sup> level of 5 × 10<sup>21</sup> n/㎠, He generation possibility in the Ni plating layer is very little during 40 years of operation (2042, 34 EFPY). Thermal neutrons can significantly affect the generation of He from Ni metal. At 10 years after a repair, He can be generated at a level of about 0.06 appm, a level that can add general welding repair without any consideration. After 40 years of repair, 9.8 appm of He may be generated. Although this is a rather high value, it is within the range of 0.1 to 10 appm when welding repair can be applied. Clad repair by Ni electroplating technology is expected to greatly improve the operation efficiency by improving the safety and shortening the maintenance period of the nuclear power plant.
선진 핵연료주기 시설(AFC)의 부식건전성 조사, 분석
황성식 ( Seong Sik Hwang ) 한국부식방식학회 2012 Corrosion Science and Technology Vol.11 No.6
The amount of spent fuel from nuclear power plants has been increasing. An effective management plan of the spent fuel becomes a critical issue because the storage capacity of each plant will reach its storage limit in a few years. The volume of high toxic spent fuel can be reduced through a fuel processing. Advanced Fuel Cycle system is considered to be one of the options to reduce the toxicity and volume of the spent fuel. It is necessary to set up a test facility to demonstrate the feasibility of the process at the engineering scale. The objective of the work is a development of the safety evaluation technology for the AFC system. The evaluation technology of the AFC structural integrity and processes were surveyed and reviewed. Key evaluation parameters for the main processes such as electrolytic reduction electrorefining and electrowinning were obtained the survey results may be used for the establishment of the AFC regulatory licensing procedure. The establishment of the licensing criteria minimizes the trials and errors of the AFC facility design. Issues taken from the survey on the regulatory procedure and design safety features for the AFC facility provide a chance to resolve potential issues in advance.
원자로 내부구조물 균열개시 민감도에 미치는 영향인자 고찰
황성식 ( Seong Sik Hwang ),최민재 ( Min Jae Choi ),김성우 ( Sung Woo Kim ),김동진 ( Dong Jin Kim ) 한국부식방식학회 2021 Corrosion Science and Technology Vol.20 No.4
To safely operate domestic nuclear power plants approaching the end of their design life, the material degradation management strategy of the components is important. Among studies conducted to improve the soundness of nuclear reactor components, research methods for understanding the degradation of reactor internals and preparing management strategies were surveyed. Since the IGSCC (Intergranular Stress Corrosion Cracking) initiation and propagation process is associated with metal dissolution at the crack tip, crack initiation sensitivity was decreased in the hydrogenated water with decreased crack sensitivity but occurrence of small surface cracks increased. A stress of 50 to 55% of the yield strength of the irradiated materials was required to cause IASCC (Irradiation Assisted Stress Corrosion Cracking) failure at the end of the reactor operating life. In the threshold-stress analysis, IASCC cracks were not expected to occur until the end of life at a stress of less than 62% of the investigated yield strength, and the IASCC critical dose was determined to be 4 dpa (Displacement Per Atom). The stainless steel surface oxide was composed of an internal Cr-rich spinel oxide and an external Fe and Ni-rich oxide, regardless of the dose and applied strain level.
기술노문특집 : 원자력 발전설비의 부식현황 및 대처방안 ; 기술논문 : 원자력 발전소 Alloy 600 부품의 PWSCC-Part 1
황성식 ( Seong Sik Hwang ) 한국부식방식학회(구 한국부식학회) 2013 부식과 방식 Vol.12 No.-
Alloy 600 재료의 PWSCC의 개념을 소개하고 그 발생과 전파에 마치는 미세조직, 온도, 응력, 수화학 환경등의 주요인자를 정리하였다. ○PWSCC란 니켈 기지 합금인 Alloy 600와 그 용정재인 Alloy 82/182 재료가 원자로 l차수 환경 에서 보이는 응력부식균열을 의미한다. ○Alloy 600의 PWSCC에 미치는 주요 인자에는 재료의 미세조직, 응력, 온도, 환경등이 있으며 그 중에서 재료의 미세조직이 가장 지배적인 인자이다. ○ 재료내의 탄화물은 탄소 함량과 열처리 조건에 따라 달리 형성되며 입계를 따라 준연속적으로 잘 발달된 입계탄화율을 가지는 재료가 PWSCC 에 저항성을 가진다. ○ 손상속도는 부가 응력의 네 제곱에 비례하여 증 가하는 것으로 알려져 있다. ○PWSCC는 Arrhenius 관계의 열활성화 과정 (thermally detavitca process)이다. ○ 용존수소량에 따라 재료의 부식전위가 정해지는 데 전극전위가 Ni/NiO 평형전위 부근에서 가장 큰 균열 성장 민감도를 보인다는 데는 연구자들 사이에 이견이 없다 그러나 균열의 개시에 대한 용존수소량의 영향에 대해서는 이견이 있다.