• Journal of the Earthquake Engineering Society of Korea
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• v.18 no.6
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• pp.291-299
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• 2014

There has been an increasing demand for introducing a base isolation system to secure the seismic safety of a nuclear power plant. However, the design criteria and the safety assessment methodology of a base isolated nuclear facility are still being developed. A performance based design concept for the base isolation system needs to be added to the general seismic design procedures. For the base isolation system, the displacement responses of isolators excited by the extended design basis earthquake are important as well as the design displacement. The possible displacement response by the extended design basis earthquake should be limited less than the failure displacement of the isolator. The failure of isolators were investigated by an experimental test to define the ultimate strain level of rubber bearings. The uncertainty analysis, considering the variations of the mechanical properties of isolators and input ground motions, was performed to estimate the probabilistic distribution of the isolator displacement. The relationship of the displacement response by each ground motion level was compared in view of a period elongation and a reduction of damping. Finally, several examples of isolator parameters are calculated and the considerations for an acceptable isolation design is discussed.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.915-921
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• 2019

Radiation damage due to neutrons has been calculated in Ni-based alloys in Wolsong CANDU reactor environments. Two damage parameters are considered: displacement damage, and transmutation gas production. We used the SPECTER and SRIM computer codes in quantifying radiation damage. In addition, damage caused by Ni two-step reactions was considered. Estimations were made for the annulus spacers in a CANDU reactor that are located axially along a fuel channel and made of Inconel X-750. The calculation results indicate that the transmutation gas production from the Ni two-step reactions is predominant as the effective full power year increases. The displacement damage due to recoil atoms produced from Ni two-step reactions accounts for over 30% out of the total displacement damage.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.475-480
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• 2014

We developed a simple method to prepare the displacement damage cross section of SiC using NJOY and SRIM/TRIM. The number of displacements per atom (DPA) dependent on primary knock-on atom (PKA) energy was computed using SRIM/TRIM and it is directly used by NJOY/HEATR to compute the neutron energy dependent DPA cross sections which are required to estimate the accumulated DPA of nuclear material. SiC DPA cross section is published as a table in DeCART 47 energy group structure. Proposed methodology can be easily extended to other materials.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2003.11a
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• pp.41-49
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• 2003

Tube oscillation behaviour is experimentally investigated for the study on the fuel rod fretting that is caused by the flow-induced vibration in nuclear reactor. The experiment was conducted in all at room temperature. The specimen of tube assembly was supported by plate springs which simulated the spacer grids and fuel rods of a fuel assembly. To investigate the influence of contact condition between the grids and rods, normal load of 10 and 5 N, gaps of 0.1 and 0.3 mm were applied. The range of the oscillation at the center of the fuel rod specimen was varied as 0.2, 0.3 and 0.4 mm to simulate the fuel rod vibration due to flow. Displacements near the contact were measured with four displacement sensors during the tube oscillation. As results, the shape of oscillation (phase) varied depending on the contact condition. The oscillation displacement increased considerably from the contact to gap condition. The displacement increased further as the gap size increased. It is regarded that the spring shape influences the tube oscillation behaviour. Simple calculation showed that the slip displacement was very small. Therefore, cumulative damage concept is necessary for the fuel rod wear. The mechanism of plowing is thought required to explain the severe wear in the case of gap existence.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.497-505
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• 2003

This paper describes a computational approach to the quantification of primary damage under irradiation and demonstrates the effect of neutron energy spectra on the formation of the displacement cascade. The development of displacement cascades in ${\alpha}-Iron$ has been simulated using the MOLDY code - a molecular dynamics code for simulating radiation damage. The primary knock-on atom energy, key input to the MOLDY code, was determined from the SPECTER code calculation on two neutron spectra. The two neutron spectra include; (i) neutron spectrum in the instrumented irradiation capsule of the high-flux advanced neutron application reactor (HANARO), and (ii) neutron spectrum at the inner surface of the reactor pressure vessel steel for the Younggwang nuclear power plant No.5 (YG 5). Minor differences in the normalized neutron spectra between the two spectra produce similar values of PKA energy, which are 4.7 keV for HANARO and 5.3 keV for YG 5. This similarity implies that primary damage to the components of the commercial nuclear reactors should be well simulated by irradiation in the HANARO. Moreover, the application of the MD calculations corroborates this statement by comparing cascades simulation results.

• Nuclear Engineering and Technology
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• v.33 no.6
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• pp.619-625
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• 2001

An axial-flow-induced vibration model was proposed for a rod supported by two translational springs at both ends. For developing the model, a one-mode approximation was made based on the assumption that the first mode was dominant in vibration behavior of the single span rod. The first natural frequency and mode shape functions for the flow-induced vibration, called the FIV model were derived by using Lagrange＇s method. The vibration displacements at reactor conditions were calculated by the proposed model for the spring-supported rod and by the previous model for the simple-supported(55) rod. As a result, the vibration displacement for the spring-supported rod was larger than that of the 55 rod, and the discrepancy between both displacements became much larger as flow velocity increased. The vibration displacement for the spring-supported rod appeared to decrease with the increase of the spring constant. AS flow velocity increased, the increase rate of vibration displacement was calculated to go linearly up, and that of the rod having the short span length was larger than that of the rod having the long span length although the displacement value itself of the long span rod was larger than that of the short one.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.336-343
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• 1998

An iteration method has been developed for determining crack growth and fracture resistance cure (J-R curve) from the load versus load-line displacement record only. In this method, the hardening curve, the load versus displacement curve at a given crack length, is assumed to be a power-law function, where the exponent varies with the crack length. The exponent is determined by an iterative calculation method with the assumption that the exponent varies linearly with the load-line displacement. The proposed method was applied to the static J-R tests using compact tension(CT) specimens, a three-point bend (TPB) specimen, and a cracked round bar (CRB) specimen as well as it was applied to the quasi-dynamic J-R tests using CT specimens. The J-R curves determined by the proposed method were compared with those obtained by the conventional testing methodologies. The results showed that the J-R curves could be determined directly by the proposed iteration method with sufficient accuracy in the specimens from SA508, SA533, and SA516 pressure vessel steels and SA312 Type 347 stainless steel.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3229-3240
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• 2023

Due to neutron radiation, atomic displacement has a significant effect on material in nuclear reactors. A range of secondary displacement models, including the Kinchin-Pease (K-P), Lindhard, Norgett-Robinson-Torrens (NRT), and athermal recombination-corrected displacement per atom (arc-dpa) have been suggested to calculate the number of displacement per atom (dpa). As neutron elastic interaction is the main cause of displacement damage, the focus of the current study is to calculate the atomic displacement caused by the neutron elastic interaction in order to estimate the exact amount of yield strength in a WWER-1000 reactor pressure vessel. To achieve this purpose, the reactor core is simulated by MCNPX code. In addition, a program is developed to calculate the elastic radiation damage induced by the incident neutron flux (RADIX) based on different models using Fortran programming language. Also, due to non-elastic interaction, the displacement damage is calculated by the HEATR module of the NJOY code. ASME E-693-01 standard, SPECTER, NJOY codes, and other pervious findings have been used to validate RADIX results. The results showed that the RADIX(arc-dpa)/HEATR outputs have appropriate accuracy. The relative error of the calculated dpa resulting from RADIX(arc-dpa)/HEATR is about 8% and 46% less than NJOY code, respectively in the ¼ and ¾ vessel wall.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.1
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• pp.93-103
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• 1997

Charpy V-notch impact tests were performed on the full-, half-and third-size specimens from two ferritic SA 508 Cl. 3 steels for nuclear pressure vessel. New normalization factors were proposed to predict the upper shelf energy(USE) and the ductile-brittle transition temperature(DBTT) of full-size specimens from the measured data on sub-size specimens. The factors for the USE and the DBTT are $(Bb^2/Kt); and; (Bb/R)^1/2/, $ respectively, where B the width, b the ligament size, $K_{t}$ the elastic stress concentration factor, and R the notch root radius. These correlations successfully estimated the USE and DBTT of the full-size specimens based on sub-size specimen data. In addition, the size effects were studied to develop the correlations among absorbed energy, lateral expansion(LE) and displacement. It was also found that the LE was able to be estimated from the displacement obtained by the instrumented impact test, and that the displacement would be used as a criterion for the toughness of the steels corresponding to change in their yield strength.h.

• Journal of Welding and Joining
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• v.27 no.1
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• pp.65-70
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• 2009

Ir-192 radiation sealed sources are widely employed to the therapeutic applications as well as the non-destructive testing. Production of Ir-192 sources requires a delicate but robust welding technique because it is employed in a high radioactive working environment. A GTA(Gas Tungsten Arc) welding technique is currently well established for this purpose. However, this welding method requires a frequent replacement of the electrode, which results in the delay of the production to take a preparatory action such as to isolate the radiation sources from the working place before getting access to the welding machine. Hence, a resistance welding technique is considered as an alternative method of the GTA welding technique. The advantages of resistance welding are high welding speed and high-rate production. Also it has very long life of electrode comparing to GTA welding. In this study, the resistance welding system and proper welding conditions were established for sealing Ir-192 source capsule. As a results of various experiments, it showed that electrode displacement can be employed as a indicator to predict welding quality. We proposed two mathematical models(linear and curvilinear) to estimate electrode displacement with process parameters such as applied force, welding current and welding time by using regression analysis method. Predicting results of both linear and curvilinear model were relatively good agreement with experiment.