• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4022-4032
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• 2021

In this study, machine learning (ML) techniques were used to model surveillance test data of nuclear power plants from an international database of the ASTM E10.02 committee. Regression modeling was conducted using various techniques, including Cubist, XGBoost, and a support vector machine. The root mean square deviation of each ML model for the baseline dataset was less than that of the ASTM E900-15 nonlinear regression model. With respect to the interpolation, the ML methods provided excellent predictions with relatively few computations when applied to the given data range. The effect of the explanatory variables on the transition temperature shift (TTS) for the ML methods was analyzed, and the trends were slightly different from those for the ASTM E900-15 model. ML methods showed some weakness in the extrapolation of the fluence in comparison to the ASTM E900-15, while the Cubist method achieved an extrapolation to a certain extent. To achieve a more reliable prediction of the TTS, it was confirmed that advanced techniques should be considered for extrapolation when applying ML modeling.

• Journal of Information Technology Services
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• v.8 no.2
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• pp.147-156
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• 2009

As the result of the rapid development of IT technology, an on-line diagnostic system using the field bus communication network coupled with a smart sensor module will be widely used at the nuclear power plant in the near future. The smart sensor system is very useful for the prompt understanding of abnormal state of the key equipments installed in the nuclear power plant. In this paper, it is assumed that a smart sensor system based on the fieldbus communication network for the surveillance and diagnostics of safety-critical equipments will be installed in the harsh-environment of the nuclear power plant. It means that the key components of fieldbus communication system including microprocessor, FPGA, and ASIC devices, are to be installed in the RPV (reactor pressure vessel) and the RCS (reactor coolant system) area, which is the area of a high dose-rate gamma irradiation fields. Gamma radiation constraints for the DBA (design basis accident) qualification of the RTD sensor installed in the harsh environment of nuclear power plant, are typically on the order of 4 kGy/h. In order to use a field bus communication network as an ad-hoc diagnostics sensor network in the vicinity of the RCS pump area of the nuclear power plant, the robust survivability of IT-based micro-electronic components in such intense gamma-radiation fields therefore should be verified. An intelligent CCD camera system, which are composed of advanced micro-electronics devices based on IT technology, have been gamma irradiated at the dose rate of about 4.2kGy/h during an hour UP to a total dose of 4kGy. The degradation performance of the gamma irradiated CCD camera system is explained.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.340-350
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• 2015

This paper proposes strain-based acceptance criteria for assessing plastic instability of the safety class 1 nuclear components made of ferritic steel during level D service loads. The strain-based criteria were proposed with two approaches: (1) a section average approach and (2) a critical location approach. Both approaches were based on the damage initiation point corresponding to the maximum load-carrying capability point instead of the fracture point via tensile tests and finite element analysis (FEA) for the notched specimen under uni-axial tensile loading. The two proposed criteria were reviewed from the viewpoint of design practice and philosophy to select a more appropriate criterion. As a result of the review, it was found that the section average approach is more appropriate than the critical location approach from the viewpoint of design practice and philosophy. Finally, the criterion based on the section average approach was applied to a simplified reactor pressure vessel (RPV) outlet nozzle subject to SSE loads. The application shows that the strain-based acceptance criteria can consider cumulative damages caused by the sequential loads unlike the stress-based acceptance criteria and can reduce the overconservatism of the stress-based acceptance criteria, which often occurs for level D service loads.

• Journal of the Korean Society for Nondestructive Testing
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• v.18 no.6
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• pp.477-483
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• 1998

Hysteresis loop, Barkhausen noise(BN), and hardness were measured in the neutron irradiated RPV steel for various fluence, irradiated dose up to $10^{18}n/cm^2$. The coercivity, remanence and maximum induction of neutron irradiated samples did not change significantly, but the BNA and BNE were decreased as the neutron irradiation increased. The changes of BNE and BNA were characterized by three stages with respect to neutron dose. The BNA and BNE were decreased with an increase of neutron dose to $10^{12}n/cm^2$, and remained nearly constant up to $10^{16}n/cm^2$, then were decreased rapidly with an increase of the neutron dose above $10^{16}n/cm^2$. On the other hand, the hardness was observed revesely with the change of BNA and BNE.

• Korean Journal of Metals and Materials
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• v.49 no.8
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• pp.628-634
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• 2011

SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are higher than in commercial SA508 Gr.3 Mn-Mo-Ni low alloy steels, may be a candidate reactor pressure vessel (RPV) material with higher strength and toughness from its tempered martensitic microstructure. The inner surface of the RPV is weld-cladded with stainless steels to prevent corrosion. The goal of this study is to evaluate the microstructural properties of the clad interface between Ni-Cr-Mo low alloy steel and stainless weldment, and the effects of post weld heat treatment (PWHT) on the properties. The properties of the clad interface were compared with those of commercial Mn-Mo-Ni low alloy steel. Multi-layer welding of model alloys with ER308L and ER309L stainless steel by the SAW method was performed, and then PWHT was conducted at $610^{\circ}C$ for 30 h. The microstructural changes of the clad interface were analyzed using OM, SEM and TEM, and micro-Vickers hardness tests were performed. Before PWHT, the heat affected zone (HAZ) showed higher hardness than base and weld metals due to formation of martensite after welding in both steels. In addition, the hardness of the HAZ in Ni-Cr-Mo low alloy steel was higher than that in Mn-Mo-Ni low alloy steel due to a comparatively high martensite fraction. The hardness of the HAZ decreased after PWHT in both steels, but the dark region was formed near the fusion line in which the hardness was locally high. In the case of Mn-Mo-Ni low alloy steel, formation of fine Cr-carbides in the weld region near the fusion line by diffusion of C from the base metal resulted in locally high hardness in the dark region. However, the precipitates of the region in the Ni-Cr-Mo low alloy steel were similar to that in the base metal, and the hardness in the region was not greatly different from that in the base metal.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.3
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• pp.165-172
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• 2016

An experimental study of the thermal-hydraulic characteristics of passive safety systems (PSSs) was conducted using a system-integrated modular advanced reactor-integral test loop (SMART-ITL). The present passive safety injection system for the SMART-ITL consists of one train with the core makeup tank (CMT), the safety injection tank, and the automatic depressurization system. The objective of this study is to investigate the injection effect of the PSS on the small-break loss-of-coolant accident (SBLOCA) scenario for a 0.4 inch line break in the safety-injection system (SIS). The steady-state condition was maintained for 746 seconds before the break. When the major parameters of the target value and test results were compared, most of the thermal-hydraulic parameters agreed closely with each other. The water level of the reactor pressure vessel (RPV) was maintained higher than that of the fuel assembly plate during the transient, for the present CMT and safety injection tank (SIT) flow rate conditions. It can be seen that the capability of an emergency core cooling system is sufficient during the transient with SMART passive SISs.

• Nuclear Engineering and Technology
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• v.22 no.4
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• pp.337-350
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• 1990

A post-irradiation annealing study was conducted with use of reactor pressure vessel(RPV) steel A533B Cl.1 base metal irradiated to a dose of 4.84$\times$10$^{18}$ n/$\textrm{cm}^2$ at about 38$0^{\circ}C$. Microhardness and positron annihilation (PA) methods were used to obtain better understanding of the recovery of radiation hardening. Isochronal anneal experiments indicated that two recovery processes occur during annealing of irradiated specimens. The first recovery process occurs in the temperature range of 280-3O5$^{\circ}C$, Michrohardness and positron annihilation (PA) methods were used to obtain better understanding of the recovery of radiation hardening. Isochronal anneal experiments indicated that two recovery processes occur during annealing of irradiated specimens. The first recovery process occurrs in the temperature range of 280-305$^{\circ}C$. The variations of Ip, Iw and R parameters indicated that the formation of vacancy clusters by vacancy agglomeration and the annihilation of monovacancies are the first recovery process. The second recovery process occurs in the range of 405-49$0^{\circ}C$ and positron annihilation parameters measured indicated that the dissolution of carbon atoms decorated around vacancy-type defects and possible precipitates, and the annihilation of monovacancies give rise to the second recovery process. It was further indicated that radiation anneal hardening (RAH) in the range of 305-405$^{\circ}C$ between the temperature ranges for the two processes occurs due to the formation of carbon-decorated vacancy clusters and precipitates. The activation energies, orders of reaction and other characteristics of recovery processes were determined by the Meechan-Brinkman method. The activation energy for the first recovery process was determined as 1.76 eV and that for the second recovery process as 2.00eV. These values are lower than those obtained by other workers. This difference may be attributed to the lower copper content of the RPV steel used in the present study. The order of reaction for the first recovery process was determined as 1.78, while that for the second recovery process as 1.67 Non-integer orders of reaction for recovery processes seem to be attributed to the fact that several mechanisms for the first order and the second order of reaction are compounded in one process. This result also supports for the above conclusions from measurements of PA parameters.