• Journal of the Korean Society for Nondestructive Testing
• /
• v.29 no.3
• /
• pp.248-255
• /
• 2009

After several damages by PWSCC were found in the world, USNRC and PNNL(Pacific Northwest National Laboratory) started the research on PWSCC under the project name of PINC. The aim of the project was 1) to fabricate representative NDE mock-ups with flaws to simulate PWSCCs, 2) to identify and quantitatively assess NDE methods for accurately detecting, sizing and characterizing PWSCCs, 3) to document the range of locations and morphologies of PWSCCs and 4) to incorporate results with other results of ongoing PWSCC research programs, as appropriate. Korea nuclear industries have also been participating in the project. Thermally and mechanically cracked-four mockups were prepared and phased array and manual ultrasonic testing(UT) techniques were applied. The results and lessons learned from the preliminary RRT are summarized as follows: 1) Korea RRT teams performed the RRT successfully. 2) Crack detection probability of the participating organizations was an average 87%, 80% and 80% respectively. 3) RMS error of the crack sizing showed comparatively good results. 4) The lessons learned may be helpful to perform the PINC RRT and PSI /ISI in Korea in the future.

• Nuclear Engineering and Technology
• /
• v.51 no.4
• /
• pp.1109-1116
• /
• 2019

It is essential to predict the radioactivity distribution around the reactor cooling system (RCS) during obligatory cyclic operation of AP-1000. A home-developed program CPA-AP1000 is upgraded to predict the response of activated corrosion products (ACPs) in the RCS. The program is written in MATLAB and it uses state of the art MCNP as a subroutine for flux calculations. A pair of cyclic power profiles were superimposed after initial full power operation. The effect of cyclic operation is noticed to be more prominent for in-core surfaces, followed by the primary coolant and out-of-core structures. The results have shown that specific activity trends of $^{56}Mn$ and $^{24}Na$ promptly follow the power variations, whereas, $^{59}Fe$, $^{58}Co$, $^{99}Mo$ and $^{60}Co$ exhibit a sluggish power-following response. The investigations pointed out that promptly power-following response of ACPs in the coolant is vital as an instant radioactivity source during leakage incidents. However, the ACPs with delayed power-following response in the out-of-core components are perceived to cause a long-term activity. The present results are found in good agreement with those for a reference PWR. The results are useful for source term monitoring and optimization of work procedures for an innovative reactor design.

• The Korean Journal of Nuclear Medicine Technology
• /
• v.13 no.3
• /
• pp.185-188
• /
• 2009

Purpose: Standard of retests were discrepant and inconsistent due to inaccuracy and lack of standardization within normal range limit of tumor marker test. To enhance the standardization of retests set standard value below normal range and the Order Communication System Quality Control (OCS QC) program was put in place. This program enables managing the results within lower limit of normal range which were used for tumor marker test in Health Center. Materials and Methods: At present the tumor marker study for AFP, CEA, CA19-9, CA125, and PSA included outpatients in Asan Medical Center from February to March, 2009. The standard value was obtained by using the percentage of CV of Inter Assay according to the normal range of each tumor test. The results were confirmed by using the OCS QC program via formatted assessment of screening test such as test items, standard value and medical department. The number of out-of-range results within plus and minus 30 percents regarding the five primary items of tumor marker test was assessed. The next step was to obtain the number of AFP, CEA, and CA125 according to the ratio of comparison between prior and post test result, 60%, 50%, and 40% within normal range, respectively. In addition, set standard value below normal range. Results: The first screening test with percentage of sample number was resulted between 30%-40% and the second one was AFP 26.1%, CEA 18.9%, CA19-9 17.3%, CA125 18.7%, and PSA 21.0% obtained screening percentage of average 20 percents. The limited value of retest was AFP less than 5.0 and more than 10.0, CEA less than 1.0 and more than 3.0, CA19-9 less than 10.0 and more than 30.0, and PSA less than 1.0 and more than 2.0 to set and the number of retest was obtained by applying to the limited value of retest to screening percentage of average 20 percents For two months, the number of retest was AFP 0, CEA 15, CA19-9 3, CA125 2, and PSA 5. Conclusions: Through using the OCS QC program in establishing the standard of retest systemically, there appeared to be reduced discrepancy among the examiners and to be expected improvement in relation to the error of results.

• Journal of Energy Engineering
• /
• v.8 no.2
• /
• pp.213-223
• /
• 1999

In recently, the safety system software of the nuclear power plant has been verified and validated according to ANSI/IEEE-ANS-7-4.3.2-1982 to improve the reliability. This standard requires that safety-related software should be tested in the static and dynamic environments. In case of Inadequate Core Cooling Monitoring System (ICCMS), the static test procedure and related techniques are developed but the dynamic test procedure and related techniques are not developed. Therefore, this paper discusses the undeveloped techniques, and suggests the dynamic test procedure and the program for generation of test input data. The performance of the program was identified using accident analysis report of Ulchin 3&4 Final Safety Analysis Report (FSAR).

• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.23 no.2
• /
• pp.235-243
• /
• 2010

Three-dimensional structural analysis system for nuclear containment building is presented in this paper. This system includes high-performance plate/shell elements as finite element library. It also adopts numerical modeling technique for unbonded tendon as well as bonded tendon in prestressed concrete structures. This system is constructed by connecting several in-house program to a commercial program DIANA, and then is capable of performing nonlinear analysis for ultimate pressure capacity of nuclear containment building. Finally, three-dimensional structural analysis of CANDU-type containment building is carried out in order to test the reliability of this system. These numerical results are compared with reference values, which obtained from axisymmetric structural analysis.

• The Transactions of The Korean Institute of Electrical Engineers
• /
• v.64 no.2
• /
• pp.254-259
• /
• 2015

On January 30, 2012 an auxiliary component of Byron Unit 2 was tripped on bus under voltage. The cause of the event was the failure of the C-phase insulator track for the Unit 2 station auxiliary transformer(SAT) revenue metering transformer. In addition to this event, other events have occurred at other plants resulting in an open-phase condition.[1] Therefore, Nuclear Regulatory Commission(NRC) has requested that not only nuclear power plant(NPP) operating company but also its Design Certification(DC) applicant have to prepare open-phase detecting system in their operating plants and design document. In this paper, various open-phase conditions are simulated in NPP using Electromagnetic Transient Program(EMTP) and Atpdraw, and open-phase condition detecting system is proposed for Main Transformer(MT), Unit Auxiliary Transformer(UAT) and SAT connected power line in NPP.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.05a
• /
• pp.1001-1007
• /
• 1995

A Comprehensive Vibration Assessment Program (CVAP) has been performed for Yonggwang Nuclear Power Plant Unit 4 (YGN 4) in order to verify the structural integrity of the reactor internals for flow induced vibrations prior to commercial operation. The theoretical evidence for the structural integrity of the reactor internals and the basis for measurement and inspection are provided by the analysis. Flow induced hydraulic loads and reactor internals vibration response data were measured during pre-core hot functional testing in YGN 4 site. Also, the critical areas in the reactor internals were inspected visually to check any existence of structural abnormality before and after the pre-core hot functional testing. Then, the measured data have been analyzed and compared with the predicted data by analysis. The measured stresses are less than the predicted values and the allowable limits. It is concluded that the vibration response of the reactor internals due to the flow induced vibration under normal operation is acceptable for long term operation.

• Nuclear Engineering and Technology
• /
• v.18 no.2
• /
• pp.71-77
• /
• 1986

A one dimensional neutron kinetics program, BIK which is applicable to the safety analyses of PWR's is developed to analyze the reactor core in axial dimension. The BIK employs the finite difference technique in space and $\theta$-time integration method in time. Detailed models for the Doppler and moderator feedbacks and control rod motion are included. The benchmark of the nuclear model is carried out through the ANL benchmark problem and the time dependent nuclear power change in the rod ejection accident of KNU1 is calculated by BIK code. The results indicate that the BIK can predict the neutron dynamics with fair accuracy within the limits of one dimensional analysis and it is useful for the safety analyses of PWR's.

• Nuclear Engineering and Technology
• /
• v.51 no.1
• /
• pp.325-336
• /
• 2019

The activity concentrations of $^{226}Ra$, $^{232}Th$, and $^{40}K$ from 102 building materials samples were determined using a high-purity germanium (HPGe) detector. The activity concentrations were evaluated for possible radiological hazards to the human health. The excess lifetime cancer risks (ELCR) were also estimated, and the average values were recorded as $0.42{\pm}0.24{\times}10^{-3}$, $3.22{\pm}1.83{\times}10^{-3}$, and $3.65{\pm}1.85{\times}10^{-3}$ for outdoor, indoor, and total ELCR respectively. The activity concentrations were further subjected to RESRAD-BUILD computer code to evaluate the long-term radiation exposure to a dweller. The indoor doses were assessed from zero up to 70 years. The simulation results were $92{\pm}59$, $689{\pm}566$, and $782{\pm}569{\mu}Sv\;y^{-1}$ for indoor external, internal, and total effective dose equivalent (TEDE) respectively. The results reported were all below the recommended maximum values. Therefore, the radiological hazards attributed to building materials under study are negligible.

• Journal of Welding and Joining
• /
• v.33 no.1
• /
• pp.14-23
• /
• 2015

Recent developing tendency of nuclear power plant are studied by searching of NDSL, KIPRIS, Science Direct and so on. Welding materials such as low alloyed steels, stainless steels, nickel-based alloys, zirconium alloy and welding methods such as narrow gap welding, laser beam welding, friction stir welding, overlay welding are investigated.