• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3892-3901
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• 2021

The pressure drop of a moisture separator in a steam generator is the important design parameter to ensure the successful performance of a nuclear power plant. The moisture separators have a wide range of operating conditions based on the arrangement of them. The prediction of the pressure drop in a moisture separator is challenging due to the complexity of the multi-dimensional two-phase vortex flow. In this study, the moisture separator test facility using the air/water two-phase flow was used to predict the pressure drop of a moisture separator in a Korean OPR-1000 reactor. The prototypical steam/water two-phase flow conditions in a steam generator were simulated as air/water two-phase flow conditions by preserving the centrifugal force and vapor quality. A series of experiments were carried out to investigate the effect of hydraulic characteristics such as the quality and liquid mass flux on the two-phase pressure drop. A new prediction model based on the scaling law was suggested and validated experimentally using the full and half scale of separators. The suggested prediction model showed good agreement with the steam/water experimental results, and it can be extended to predict the steam/water two-phase pressure drop for moisture separators.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.59-68
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• 2020

ECCMIX component was introduced in RELAP5/MOD3 for calculating the interfacial condensation. Compared to other existing components in RELAP5, user experience of ECCMIX component is restricted to developmental assessment applications. To evaluate the capability of the ECCMIX component, ECCS experiment was conducted which included single-phase and two-phase thermal mixing. The experiment was carried out with test sections containing a main pipe (70 mm inner diameter) and a branch pipe (21 mm inner diameter) under the atmospheric pressure. The steam mass flow in the main pipe ranged from 0 to 0.0347 kg/s, and the subcooled water mass flow in the branch pipe ranged from 0.0278 to 0.1389 kg/s. The comparison of the experimental data with the calculation results illuminated that although the ECCMIX component was more difficult to converge than Branch component, it was a more appropriate manner to simulate interfacial condensation under two-phase thermal mixing circumstance, while the two components had no differences under single-phase circumstance.

• 핵의학기술
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• 제12권3호
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• pp.181-183
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• 2008

We developed simple and highly efficient synthesis method for [$^{11}C$]methionine using solid-phase extraction method. For synthesis, we used C18 cartridge. [$^{11}C$]methionine was synthesized on C18 cartridge according to the solid-phase [$^{11}C$]methylation of precursor L-homocysteine thiolactone hydrochloride. The radiochemical yields of [$^{11}C$]methionine was $48.9{\pm}7.93%$ decay corrected (results of 30 syntheses, mean$\pm$SD), with average production higher than 180 mCi. This procedure showed high yield and simple synthesis of [$^{11}C$]methionine.

• 전기학회논문지
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• 제64권2호
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• pp.254-259
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• 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.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.280-291
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• 2018

In the framework of OECD/NEA Working Group on Fuel Safety, a RIA fuel-rod-code Benchmark Phase I was organized in 2010-2013. It consisted of four experiments on highly irradiated fuel rodlets tested under different experimental conditions. This benchmark revealed the need to better understand the basic models incorporated in each code for realistic simulation of the complicated integral RIA tests with high burnup fuel rods. A second phase of the benchmark (Phase II) was thus launched early in 2014, which has been organized in two complementary activities: (1) comparison of the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena; (2) assessment of the uncertainty of the results. The present paper provides a summary and conclusions of the second activity of the Benchmark Phase II, which is based on the input uncertainty propagation methodology. The main conclusion is that uncertainties cannot fully explain the difference between the code predictions. Finally, based on the RIA benchmark Phase-I and Phase-II conclusions, some recommendations are made.

• Nuclear Engineering and Technology
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• 제56권8호
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• pp.3435-3449
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• 2024

Considering that system analysis codes are used for the evaluation of the performance of Passive Safety Systems (PSSs), it is important to investigate the capability of the system analysis code to reliably predict the heat transfer and natural circulation flow, which are the main phenomena governing the performance of a PSS. Since MARS-KS has been widely validated for heat transfer models, this study focuses on evaluating its capability to predict the single and two-phase pressure drops and natural circulation flow. The straight pipe simulation results indicate that the pressure drop predictions are reliable within ±5 % error margin for the single-phase flow and the errors of pressure drop up to - 30 % for the two-phase flow. Through single-phase natural circulation flow analysis, it is concluded that the use of the appropriate K-factor modeling based on the flow regimes is important since the natural circulation flow rate in MARS-KS is mainly affected by the form loss factor modeling. With two-phase natural circulation flow analysis, this study emphasizes the behavior of the system could change significantly depending on the two-phase wall friction and pressure loss modeling. With the analysis results, modeling considerations for the PSS performance evaluation with the system analysis codes are proposed.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.886-892
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• 2024

The cladding temperatures and axial mass distribution computed by MAAP5 were compared with their measured values in the test bundle of the Phebus Test FPT0. The computed cladding temperatures were in good agreed with the measured values in the pre-transient phase. In the transient heat-up phase, the computed temperatures were overestimated by the Baker-Just correlation in MAAP5, but the computed temperatures could simulate the subsequently measured values. The computed mass distribution in the axial direction was in qualitative agreement with the measured one for post-test fuel damage observations. The calculated flow areas of inner and outer regions in the test bundle were compared with the photographic observations. MAAP5 computed them at the height of 0.2 m where the molten pool formed was in qualitative agreement with the photographic observations. It was found that the remaining steam flow paths might be caused by the gas-liquid two-phase flow counter-current flow limitation.

• Nuclear Engineering and Technology
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• 제44권5호
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• pp.551-560
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• 2012

The Wolsong-1 (W-1) Phase-B pre-simulations were carried out in preparation for tests to be conducted for the restart of the reactor after a major refurbishment project that included replacement of the pressure tube. These pre-simulations for Wolsong-1 Phase-B differ from those in the past that were performed for the Wolsong-1,2,3,4 tests in that these tests use the WIMS/DRAGON/RFSP-IST code suite for verification of the tests and gadolinium instead of the traditional PPV/MULTICELL/RFSP code system and boron as poison in the moderator system. The use of gadolinium is deemed not to have domestically accumulated experience gained from the previous Phase-B tests. Thus, it is appropriate to conduct a study in order to gain a correct understanding and interpretation of potential differences in test results stemming from using gadolinium rather than boron. Although the calibration of the reactivity device will not be noticeably different using boron and gadolinium at a constant moderator temperature, the temperature dependency of the neutronic behavior due to the presence of gadolinium in the moderator system might be pronounced. The results of the pre-simulations using gadolinium revealed that the moderator temperature reactivity coefficients indeed showed significant differences in comparison with those with boron. In order to secure the validity of the analysis results, the newly acquired WIMS/DRAGON/RFSP-IST code suite was verified against the W-2,3,4 Phase-B test results. The results of the new code suite verifications revealed some overall improvements in accuracy; justification of the use of the code can be claimed for the validation of the W-1 Phase-B test results.

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.525-536
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• 2005

The interfacial area transport equation dynamically models the changes in interfacial structures along the flow field by mechanistically modeling the creation and destruction of dispersed phase. Hence, when employed in the numerical thermal-hydraulic system analysis codes, it eliminates artificial bifurcations stemming from the use of the static flow regime transition criteria. Accounting for the substantial differences in the transport mechanism for various sizes of bubbles, the transport equation is formulated for two characteristic groups of bubbles. The group 1 equation describes the transport of small-dispersed bubbles, whereas the group 2 equation describes the transport of large cap, slug or chum-turbulent bubbles. To evaluate the feasibility and reliability of interfacial area transport equation available at present, it is benchmarked by an extensive database established in various two-phase flow configurations spanning from bubbly to chum-turbulent flow regimes. The geometrical effect in interfacial area transport is examined by the data acquired in vertical fir-water two-phase flow through round pipes of various sizes and a confined flow duct, and by those acquired In vertical co-current downward air-water two-phase flow through round pipes of two different sizes.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.69-78
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• 2021

Venturi meter is frequently used in feed water flow control system in a nuclear power plant. Its accurate measurement plays a vital role in the safe operation of the plant. This paper firstly investigates the influence of the length of each section of pipeline, the throat inner diameter of Venturi and the flow characteristics in a single-phase flow on the accuracy of Venturi measurement by numerical calculation. Then the flow and the accuracy are discussed in a multi-phase flow. Numerical results show that the geometrical parameters and the characteristics of complex turbulent flow in the single-phase flow have little impact on the accuracy of Venturi flow rate measurement. In the multi-phase flow, the calculated flow rate of Venturi deviated from the actual flow rate and this deviation value is closely related to the amount of steam in the pipeline and increases sharply with the increase of the amount of steam. The over-reading of Venturi flow rate is present.