• Proceedings of the KIEE Conference
• /
• 1999.07c
• /
• pp.1311-1313
• /
• 1999

This paper presents the selection of suitable locations for PSSs in large power system using Speed Participation Factor(SPF). And practical considerations for optimal locations of PSSs are discussed. In this paper, the simulation results for SPF in KEPCO's power system using SSSP by EPRI are also included.

• The KSFM Journal of Fluid Machinery
• /
• v.13 no.3
• /
• pp.12-17
• /
• 2010

The orifice establishment which is improper does to change the entity differential pressure and occurs an error in flow measurement data. Because of this, the thermal power of nuclear power plant could be evaluated excessively and the safety margin could be decreased. In this paper, characters of orifice which is established abnormally was investigated. Specially, the orifice plate which is established in opposition case was modeled and analyzed. Finally, 14.4% was lowly measured differential pressure, when being established in the resultant opposition. And this result with EPRI and NRC experiences was similar.

• Proceedings of the KSME Conference
• /
• 2003.04a
• /
• pp.610-616
• /
• 2003

Recently, the common concern of nuclear power industry in the development of technology mitigating and preventing the aging of steam generator tubes prevails, because the trends of steam generator flaws at Uljin unit #1,2 and KSNP(Korea Standard Nuclear Power Plant) impose a burden on the operation of nuclear power plant. While the regulatory agency is demanding the establishment of the advanced general performance maintenance system, the steam generator management program adapting advanced technology is being developed which may comply with EPRI PWR SG Guidelines based on NEI 97-06 ‘ General Guidelines including all the maintenance aspects consist of the tube integrity assessment criteria, repair limit, allowable leakage level, water chemistry will be composed in order to obtain the approval of regulatory agency and be applied to Nuclear power plant early 2005. This presentation is to introduce maintenance state including SG tube degradation and main contents of advanced SG management program being developed, and futhermore update present and future plan, and estimate the alternation after the completion.

• Proceedings of the Computational Structural Engineering Institute Conference
• /
• 2006.04a
• /
• pp.575-580
• /
• 2006

Air-operated valves are extensively used for process control and system isolation functions in nuclear power plant, where the safety is primary issue. The purpose of this study is to develop structural analysis code of various air-operated valves such as globe valve, gate valve, and butterfly valve. The thrust formula is derived for valve with the expected weak areas. The expected weak areas are referred from EPRI data. The structural stress analysis is carried out by analytical and commercial FEM code, ANSYS 8.0. The numerical results are compared together and verified on program procedures.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.11a
• /
• pp.103-109
• /
• 1996

고리 원자력 1호기 14주기(‘95년도) 운전기간 중 증기발생기 세관 열전달 용량 저하로 전출력 운전 기간동안 정격출력보다 15% 감발 운전한 경험이 있었는데, 이 기간중 냉각재내 방사성 부식생성물(CRUD) 농도가 약 80% 감소됨을 발견하였다. 이때 출력감소 비율보다 많은 CRUD 감소현상 규명을 위해 냉각재 수질관리인자와 EPRI 피복재 부식모델인 PFCC코드를 사용한 피 복재 산화물 두께변화 등을 비교한 결과, 운전중 용출되는 방사성 부식생성물은 핵연료 표면의 피복재 산화물에 흡착된 Co핵종이 피복재 산화물 이탈시 함께 거동하는 것으로 확인되었으며, 피복재 산화물 이탈은 산화막 두께 및 열유속에 주로 의존함이 밝혀졌다. 따라서 냉각재내에서 방사성 부식 생성물의 생성률 저감을 위해서는 정상운전시 핵연료 표면의 산화막 증가를 억제할 수 있는 수질 조건을 도출하고 그에따른 운전을 통해 원전 작업자의 방사선 피폭량 저감 및 방사성폐기물의 발생을 줄일 수 있을 것으로 여겨진다.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05b
• /
• pp.417-422
• /
• 1996

The KNGR has an advanced ECCS design feature which employs four mechanically-separated SI trains where each train consisting of one HPSI pump and one SIT injects ECC water directly into the reactor vessel downcomer annulus. To demonstrate that the KNGR ECCS design features meet the EPRI ALWR requirements of no core uncovery for a break of up to 6 inch diameter, small break LOCA cases with various break sizes were analyzed using a best-estimate analytical procedure. Two kinds of break locations are considered: cold leg and DVI line breaks. It was observed that the KNGR ECC design can tolerate a cold leg break of up to 10 inches with no core uncovery. However. since DVI line break with 6 inch diameter undergoes slight core uncovery. further investigation is required for KNGR SIS optimization.

• Proceedings of the KIEE Conference
• /
• 2015.07a
• /
• pp.416-417
• /
• 2015

In this study, the parameters of the lightning flashover rate calculation algorithm in South Korea, USA, Japan are investigated for the suitable algorithm for South Korea. Each parameters used in each calculation algorithms have a obvious difference by the environment of the area and transmission lines, because they have different environmental parameters respectively. In this paper, the environmental parameters are weighted according to the change of outputs by different inputs, environmental parameters. In addition, the methods of the selection CRIEPI algorithm and the application EPRI, CRIEPI parameters for improvement the KEPRI algorithm are proposed.

• Proceedings of the KSME Conference
• /
• 2001.06a
• /
• pp.616-621
• /
• 2001

Flow-accelerated corrosion(FAC) is a phenomenon that results in metal loss from piping, vessels, and equipment made of carbon steel. FAC occurs only under certain conditions of flow, chemistry, geometry, and material. Unfortunately, those conditions are in much of the high-energy piping in nuclear and fossil-fueled power plants. Also, for domestic NPP secondary pipings whose operating time become longer, more evidences of FAC have been reported. The authors are studying on FAC management using CHECWORKS, computer code developed by EPRI. This paper is on the prediction results of metal loss by FAC in the one of CANDU type NPP secondary piping systems.

• Proceedings of the KSME Conference
• /
• 2000.11a
• /
• pp.328-332
• /
• 2000

Following the method described in ASTM E1737, J resistance curves are measured for Cr-Mo steel SA387, and Cr steel A240 which are used as piping materials in nuclear industry. Crack driving force diagrams are generated in order to find out instability points in crack growth. The $J_{appl}$ curves, which are used in the crack driving force diagram, are obtained from EPRI J estimation method and the finite element analysis. Crack growth instability points are plotted in load-crack length plane and the results are discussed.

• Proceedings of the KIEE Conference
• /
• 2002.07c
• /
• pp.1950-1952
• /
• 2002

원자력 발전소 방사성 잡음 규제기준을 설정하기 위하여 발전소 내의 전자파 잡음 환경을 측정하였다. 대상 발전소는 1000 MW급 울진 원자력발전소 3호기로 발전소 출력 0 %에서부터 100 %까지 10일간 주 제어실 및 전기설비실 주변을 대상으로 US NRC Reg. Guide 1.180 및 Mil-Std-462D를 측정기준으로 하여 전기장 10 kHz ${\sim}$ 7 GHz. 자기장 30 Hz ${\sim}$ 100 kHz의 방사성 잡음을 측정하였다. 측정결과를 최소자승법을 이용한 회귀분석으로 잡음 한계값을 유도하여 미국 EPRI 및 NRC 측정결과와 비교한 결과 자기장의 경우 낮은 한계값을 보였고 전기장의 경우는 20 dB 이상 높게 결정되었다.