• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1665-1670
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• 2023

It is shown that in the IBR-2M reactor by the end of the reactor cycle, changes in dynamics are observed associated with a strong weakening of the fast power feedback (PF), as a result of which the reactor becomes oscillatorily unstable. After each week of zero-power operation the negative changes in reactor dynamics disappear and the stability of the reactor is restored. Thus, the reactor undergoes cyclic changes in the oscillatory instability. The correlation between of a fast PF and a slow PF is experimentally observed, which makes it possible to almost completely eliminate the cyclic component of instability by changing the control mode of rods of the control system.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.820-830
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• 2016

Pipe wall thinning by flow-accelerated corrosion and various types of erosion is a significant and costly damage phenomenon in secondary piping systems of nuclear power plants (NPPs). Most NPPs have management programs to ensure pipe integrity due to wall thinning that includes periodic measurements for pipe wall thicknesses using nondestructive evaluation techniques. Numerous measurements using ultrasonic tests (UTs; one of the nondestructive evaluation technologies) have been performed during scheduled outages in NPPs. Using the thickness measurement data, wall thinning rates of each component are determined conservatively according to several evaluation methods developed by the United States Electric Power Research Institute. However, little is known about the conservativeness or reliability of the evaluation methods because of a lack of understanding of the measurement error. In this study, quantitative models for UT thickness measurement deviations of nuclear pipes and fittings were developed as the first step for establishing an optimized thinning evaluation procedure considering measurement error. In order to understand the characteristics of UT thickness measurement errors of nuclear pipes and fittings, round robin test results, which were obtained by previous researchers under laboratory conditions, were analyzed. Then, based on a large dataset of actual plant data from four NPPs, a quantitative model for UT thickness measurement deviation is proposed for plant conditions.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.426-433
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• 2017

This paper describes the development process, the innovative techniques used and insights gained from the latest integrated, full scope, multistate Level 2 PSA analysis conducted at the Leibstadt Nuclear Power Plant (KKL), Switzerland. KKL is a modern single-unit General Electric Boiling Water Reactor (BWR/6) with Mark III Containment, and a power output of $3600MW_{th}/1200MW_e$, the highest among the five operating reactors in Switzerland. A Level 2 Probabilistic Safety Assessment (PSA) analyses accident phenomena in nuclear power plants, identifies ways in which radioactive releases from plants can occur and estimates release pathways, magnitude and frequency. This paper attempts to give an overview of the advanced modeling techniques that have been developed and implemented for the recent KKL Level 2 PSA update, with the aim of systematizing the analysis and modeling processes, as well as complying with the relatively prescriptive Swiss requirements for PSA. The analysis provides significant insights into the absolute and relative importances of risk contributors and accident prevention and mitigation measures. Thanks to several newly developed techniques and an integrated approach, the KKL Level 2 PSA report exhibits a high degree of reviewability and maintainability, and transparently highlights the most important risk contributors to Large Early Release Frequency (LERF) with respect to initiating events, components, operator actions or seismic component failure probabilities (fragilities).

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.979-988
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• 2009

Tensioned metastable fluids provide a powerful means for low-cost, efficient detection of a wide range of nuclear particles with spectroscopic capabilities. Past work in this field has relied on one-component liquids. Pure liquids may provide very good detection capability in some aspects, such as low thresholds or large radiation interaction cross sections, but it is rare to find a liquid that is a perfect candidate on both counts. It was hypothesized that liquid mixtures could offer optimal benefits and present more options for advancement. However, not much is known about radiation-induced thermal-hydraulics involving destabilization of mixtures of tensioned metastable fluids. This paper presents results of experiments that assess key thermophysical properties of liquid mixtures governing fast neutron radiation-induced cavitation in liquid mixtures. Experiments were conducted by placing liquid mixtures of various proportions in tension metastable states using Purdue's centrifugally-tensioned metastable fluid detector (CTMFD) apparatus. Liquids chosen for this study covered a good representation of both thermal and fast neutron interaction cross sections, a range of cavitation onset thresholds and a range of thermophysical properties. Experiments were devised to measure the effective liquid mixture viscosity and surface tension. Neutron-induced tension metastability thresholds were found to vary non-linearly with mixture concentration; these thresholds varied linearly with surface tension and inversely with mixture vapor pressure (on a semi-log scale), and no visible trend with mixture viscosity nor with latent heat of vaporization.

• Journal of Applied Biological Chemistry
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• v.59 no.1
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• pp.5-7
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• 2016

One of oriental medicinal plants, Potentilla chinensis, has been used for anti-inflammation, hemostatic, decryption, and antipyretic. Especially, a root of Potentilla chinensis was used as important material for oriental medication. Although several kinds of bioactive component of Potentilla chinensis extract from stems and leaves were identified, the major component of Potentilla chinensis from roots is not well established. In this study, the root of Potentilla chinensis was extracted in different solvent system and analyzed by high performance liquid chromatography (HPLC). According to HPLC analysis, a major component was isolated and its physicochemical properties were evaluated by mass spectrometry and nuclear magnetic resonance. Based on these results, isolated compound was identified as 2,3,8-Tri-O-methylellagic acid. And quantification of 2,3,8-Tri-O-methylellagic acid with different extraction solvent system was performed for industrial application.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3910-3917
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• 2021

Reactor fuel's power distribution plays a vital role in designing the new generation thermionic Space Reactor Power Systems (SRPS). In this paper, the 1/12th SPACE-R's full reactor core was numerically analyzed with two kinds of different axial power distribution, to identify their impacts on thermal-hydraulic and thermoelectric characteristics. In the benchmark study, the maximum error between numerical results and existing data or design values ranged from 0.2 to 2.2%. Four main conclusions were obtained in the numerical analysis: a) The axial power distribution has less impact on coolant temperature. b) Axial power distribution influenced the emitter temperature distribution a lot, when the core power was cosine distributed, the maximum temperature of the emitter was 194 K higher than that of the uniform power distribution. c) Comparing to the cosine axial power distribution, the uniform axial power distribution would make the maximum temperature in each component of the reactor core much lower, reducing the requirements for core fuel material. d) Voltage and current distribution were similar to the axial electrode temperature distribution, and the axial power distribution has little effect on the output power.

• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.219-230
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• 2010

In nuclear power plants (NPPs), periodic sensor calibrations are required to assure that sensors are operating correctly. By checking the sensor's operating status at every fuel outage, faulty sensors may remain undetected for periods of up to 24 months. Moreover, typically, only a few faulty sensors are found to be calibrated. For the safe operation of NPP and the reduction of unnecessary calibration, on-line instrument calibration monitoring is needed. In this study, principal component-based auto-associative support vector regression (PCSVR) using response surface methodology (RSM) is proposed for the sensor signal validation of NPPs. This paper describes the design of a PCSVR-based sensor validation system for a power generation system. RSM is employed to determine the optimal values of SVR hyperparameters and is compared to the genetic algorithm (GA). The proposed PCSVR model is confirmed with the actual plant data of Kori Nuclear Power Plant Unit 3 and is compared with the Auto-Associative support vector regression (AASVR) and the auto-associative neural network (AANN) model. The auto-sensitivity of AASVR is improved by around six times by using a PCA, resulting in good detection of sensor drift. Compared to AANN, accuracy and cross-sensitivity are better while the auto-sensitivity is almost the same. Meanwhile, the proposed RSM for the optimization of the PCSVR algorithm performs even better in terms of accuracy, auto-sensitivity, and averaged maximum error, except in averaged RMS error, and this method is much more time efficient compared to the conventional GA method.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2720-2727
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• 2022

The Generation IV Lead-cooled fast reactor (LFR) take the liquid lead or lead-bismuth eutectic alloy (LBE) as the coolant of the primary cooling circuit. Combined with the natural characteristics of lead alloy and the design features of LFR, the system is the simplest and the number of equipment is the least, which reflects the inherent safety characteristics of LFR. The nuclear main coolant pump (MCP) is the only power component and the only rotating component in the primary circuit of the reactor, so the various operating characteristics of the MCP are directly related to the safety of the nuclear reactor. In this paper, various working conditions that may occur in the normal rotation (positive rotating) of the MCP and the corresponding internal flow characteristics are analyzed and studied, including the normal pump condition, the positive-flow braking condition and the negative-flow braking condition. Since the corrosiveness of LBE is proportional to the fluid velocity, the distribution of flow velocity in the pump channel will be the focus of this study. It is found that under the normal pump condition and positive-flow braking conditions, the high velocity region of the impeller domain appears at the inlet and outlet of the blade. At the same radius, the pressure surface is lower than the back surface, and with the increase of flow rate, the flow separation phenomenon is obvious, and the turbulent kinetic energy distribution in impeller and diffuser domain shows obvious near-wall property. Under the negative-flow braking condition, there is obvious flow separation in the impeller channel.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.373-379
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• 2017

Seismic probabilistic safety assessments are used to help understand the impact potential seismic events can have on the operation of a nuclear power plant. An important component to seismic probabilistic safety assessment is the seismic hazard curve which shows the frequency of seismic events. However, these hazard curves are estimated assuming a normal distribution of the seismic events. This may not be a strong assumption given the number of recorded events at each source-to-site distance. The use of a normal distribution makes the calculations significantly easier but may underestimate or overestimate the more rare events, which is of concern to nuclear power plants. This paper shows a preliminary exploration into the effect of using a distribution that perhaps more represents the distribution of events, such as the t-distribution to describe data. The integration of a probability distribution with potentially larger tails basically pushes the hazard curves outward, suggesting a different range of frequencies for use in seismic probabilistic safety assessments. Therefore the use of a more realistic distribution results in an increase in the frequency calculations suggesting rare events are less rare than thought in terms of seismic probabilistic safety assessment. However, the opposite was observed with the ground motion prediction equation considered.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.110-115
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• 2014

The nuclear power plant was launched by Kori unit 1 in 1978 years. Currently, 23 nuclear power plants have been operating in Korea since 1978 years. The localization was completed for most of the reactor facility from Hanbit(Youngkwang) unit 3&4. However, RCP(Reactor Coolant Pump) and MMIS(Man Machine Interface System) is an important technology that has been excluded from the scope of the technical transfer has been dependent on a specific overseas vendor. Recent success in RCP development through co-operation with government and industries. Developed RCP will be applied to Shin-Hanul unit 1&2 nuclear power plants. The RCP operates in high speed and high pressure condition and only rotating component in the NSSS(Nuclear Steam Supply System). Therefore, the problem of vibration has arisen caused by the hydraulic forces of the working fluid. These forces can influence on the stability characteristics for entire RCS(Reactor Coolant System) loop, and can act as significant destabilizing forces. In this study, vibration evaluation of the pump shaft of development RCP estimated under normal operation and over speed conditions. In order to predict the vibration characteristics and dynamic behavior, modal analysis, critical speed analysis and unbalance response spectrum analysis were performed.