• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.125-130
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• 2010

Sipping measurements were implemented at the Open Pool Australian Light water reactor (OPAL) to test for failure in reactor fuel elements. Fission product released by the fuel element into the pool water was measured using both High Purity Germanium (HPGe) detection via samples and a NaI(Tl) detection in-situ with the sipping device. Results from two fuel elements are presented.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.179-186
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• 2009

This paper presents a thermal creep test under internal pressure and post-test characterization performed on high burnup (68 MWd/kgU) ZIRLO. This research has been done by the CSN, ENRESA, and ENUSA in order to investigate the behavior of advanced cladding materials in contemporary PWRs at higher burnup under dry cask storage conditions. Also, to investigate the hydride reorientation, the cool-down of the samples after the test has been done in a coordinated manner with the internal pressure. The creep results obtained are consistent with the expected behavior from reference CWSR material, Zr-4. During the test, the material retained significant ductility: one specimen leaked during the test at an engineering strain of the tube section of 17%; remarkably, the crack closed due to de-pressurization. Although significant hydride reorientation occurred during the cool-down under pressure, no specimen failed during the cool-down.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 1997.04a
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• pp.217-224
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• 1997

The safety related equipments for use in nuclear power plants should be subjected to the seismic qualification in order to insure the safety of the nuclear power plant. This paper summarizes the seismic qualification test on the Low Voltage Motor Control Centers(MCC's) for use in Wolsong Nuclear Power Plants, Units 2, 3 and 4. The seismic qualification test was performed on the two prototype MCC's(a two-bay wide unit for Phase #1 Test and a five-bay wide unit for Phase #2 Test). The specimens were electrically powered and monitored during the test process. It was demonstrated that the specimens possessed sufficient structural and electrical integrity to withstand the required seismic conditions.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.104-104
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.693-694
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• 2005

• Journal of Korean Society for Quality Management
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• v.45 no.4
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• pp.969-980
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• 2017

Purpose: The purpose of this study was to estimate the lifetime, and verify the target lifetime at steady state temperature, of communication cable jackets used in nuclear power plants. Method: This study was completed according to test and analysis methods required by international standards. After measuring the residual elongation(%) of specimens at specific points in time with the accelerated degradation test, average failure time of each temperature was computed. Thus, the activation energy could be derived by applying the temperature-Arrhenius law to estimate cable jacket lifetime at steady state temperature. Results: The cable jacket lifetime was estimated as 363.8 years assuming a normal nuclear power plant operating temperature of $90^{\circ}C$. Conclusion: To ascertain stable operating conditions for a nuclear power plant, accelerated degradation tests were performed according to the Arrhenius law for components of the nuclear power plants. The lifetime was estimated from the degradation data collected during the accelerated degradation test.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.382-383
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• 2005

• Nuclear Engineering and Technology
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• v.38 no.7
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• pp.639-656
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• 2006

In order to apply the leak before break (LBB) concept to nuclear piping systems, the dynamic strain aging effect of low carbon steel materials has to be taken into account, in compliance with the requirements of the Korean Standard Review Guide (KSRG) 3.6.3-1. For this goal, J-R tests are needed for a range of various temperatures and loading rates, including dynamic loading conditions. In the dynamic loading J-R test, the unloading compliance method can not be applied to measure the crack growth and direct current potential drop (DCPD) method; this method also has a problem defining the crack initiation point. The normalization method is known as a very useful method to determine the J-R curve under dynamic loading because it does not need additional equipment or complicated loading sequences such as electric current or unloading. This method was accepted by the American Society for Testing and Materials (ASTM) as a standard test method E1820 A15 in 2001. However, it has not yet been clearly verified yet if the normalization method is sufficiently reliable to be applied to LBB. In this study, the basic background of the J-integral, LBB and dynamic loading J-R test are explained, and the current status for dynamic loading J-R test methods are reviewed from the view point of LBB for nuclear piping. In particular, the theoretical and historical background of the normalization method which has received attention recently, is summarized. Recent studies for this method are introduced and future works are suggested that may improve the reliability of LBB for nuclear piping.

• 제어로봇시스템학회:학술대회논문집
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• 2005.06a
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• pp.1384-1386
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• 2005

This paper focuses on development of load test simulator of a steam turbine-generator in a nuclear power plant. When load is taken off from electrical power network, it is very difficult to effectively control the steam flow to turbine of the nuclear turbine-generator, because of disturbances, such as electrical load and network unbalance on electrical network. Up to the present time, the conventional control system has been used for the load control on nuclear steam generator, owing to the easy control algorithms and the advantage which have been proven on the nuclear power plant. However, since there are problems with stability control during low power and start-up, only a highly experienced operator can operate during those procedures. Also, a great deal of time and an expensive simulator is needed for the training of an operator. The KEPRI is developed simulator for 600MW nuclear power plant to take a test of generator load rejection, throttle valve, and turbine load control. Total load test is implemented before start up.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.928-940
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• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.