• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4652-4659
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• 2022

Among the detector materials available at room temperature, thallium bromide (TlBr), which has a relatively high atomic number and density, is widely used for gamma camera imaging. This study aimed to verify the usefulness of TlBr through quantitative evaluation by modeling detectors of various compound types using Monte Carlo simulations. The Geant4 application for tomographic emission was used for simulation, and detectors based on cadmium zinc telluride and cadmium telluride materials were selected as a comparison group. A pixel-matched parallel-hole collimator with proven excellent performance was modeled, and phantoms used for quality control in nuclear medicine were used. The signal-to-noise ratio (SNR), contrast to noise ratio (CNR), sensitivity, and full width at half maximum (FWHM) were used for quantitative analysis to evaluate the image quality. The SNR, CNR, sensitivity, and FWHM for the TlBr detector material were approximately 1.05, 1.04, 1.41, and 1.02 times, respectively, higher than those of the other detector materials. The SNR, CNR and sensitivity increased with increasing detector thickness, but the spatial resolution in terms of FWHM decreased. Thus, we demonstrated the feasibility and possibility of using the TlBr detector material in comparison with commercial detector materials.

• 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.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.795-798
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• 2016

Transmission X-ray tubes based on carbon nanotube have attracted significant attention recently. In most of these tubes, tungsten is used as the target material. In this article, the well-known simulator Geant4 was used to obtain some of the tungsten target parameters. The optimal thickness for maximum production of usable X-rays when the target is exposed to electron beams of different energies was obtained. The linear variation of optimal thickness of the target for different electron energies was also obtained. The data obtained in this study can be used to design X-ray tubes. A beryllium window was considered for the X-ray tube. The X-ray energy spectra at the moment of production and after passing through the target and window for different electron energies in the 30-110 keV range were also obtained. The results obtained show that with a specific thickness, the target material itself can act as filter, which enables generation of X-rays with a limited energy.

• 제어로봇시스템학회:학술대회논문집
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• 1990.10a
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• pp.602-607
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• 1990

Spent nuclear fuel is very dangerous substance emitting strong ionizing radiation which is harmful to human body. The remote handling of spent nuclear fuel is essential because people cannot access this substance without protecting radiation. To handle highly radioactive material or nuclear waste, many kinds of teleoperators such as master slave manipulator, electro mechanical manipulator, servo manipulator, mobile robot was developed. The distance and dimension of target object cannot be measured easily when highly radioactive material is handled by teleoperator because one should use lead glass or TV camera and monitor to protect radiation and see target object. During experiments on the remote handling of spent nuclear fuel by electro mechanical manipulator, we often felt that a distance and dimension measurement system is necessary to handle the objects which is in the highly radioactive environment, so we developed a system which is appropriate for this purpose.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.6
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• pp.1015-1022
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• 2001

The objective of this paper is to develop an expert system called NPIES for nuclear piping integrity. This paper describes the structure and the development strategy of the NPIES system. The NPIES system consists of 3 part; the data input part, the analysis part and the output part. The data input part consists of the material properties database module and the suer interface module. The analysis part consists of the LEFM, CDFD, J/T, limit load modules and the 12 analysis routines for different cracks and loading conditions are provided respectively. Analysis results are presented to screen, printer and text file in the output part. Several case studies on circumferentially cracked piping were performed to evaluate the accuracy and the usefulness of the code. Maximum piping loads predicted by the NPIES system agreed well with those by the 3-dimensional finite element analysis. In addition, even if the material properties were not fully given, the NPIES system provided reasonable evaluation results with the predicted material properties inferred from the material properties database module.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.91-100
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• 2016

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.343-354
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• 2012

Primary water stress corrosion cracking (PWSCC) is a major safety concern in the nuclear power industry worldwide. PWSCC is known to initiate only in the condition in which sufficiently high tensile stress is applied to alloy 600 tube material or alloy 82/182 weld material in pressurized water reactor operating environments. However, it is still uncertain how much tensile stress is re-quired to generate PWSCC or what causes such high tensile stress. This study was performed to pre-dict the magnitude of weld residual stress and operating stress and compare it with previous experi-mental results for PWSCC initiation. For the study, a pressurizer safety nozzle was selected because it is reported to be vulnerable to PWSCC in overseas plants. The assessment was conducted by nu-merical analysis. Before performing stress analysis for plant conditions, a preliminary mock-up ana-lysis was done. The result of the preliminary analysis was validated by residual stress measurement in the mock-up. After verification of the analysis methodology, an analysis under plant conditions was conducted. The analysis results show that the stress level is not high enough to initiate PWSCC. If a plant is properly welded and operated, PWSCC is not likely to occur in the pressurizer safety nozzle.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.895-906
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• 2015

Background: Mitigation of primary water stress corrosion cracking (PWSCC) is a significant issue in the nuclear industry. Advanced nickel-based alloys with lower susceptibility have been adopted, although they do not seem to be entirely immune from PWSCC during normal operation. With regard to structural integrity assessments of the relevant components, an accurate evaluation of crack growth rate (CGR) is important. Methods: For the present study, the extended finite element method was adopted from among diverse meshless methods because of its advantages in arbitrary crack analysis. A user-subroutine based on the strain rate damage model was developed and incorporated into the crack growth evaluation. Results: The proposed method was verified by using the well-known Alloy 600 material with a reference CGR curve. The analyzed CGR curve of the alternative Alloy 690 material was then newly estimated by applying the proven method over a practical range of stress intensity factors. Conclusion: Reliable CGR curves were obtained without complex environmental facilities or a high degree of experimental effort. The proposed method may be used to assess the PWSCC resistance of nuclear components subjected to high residual stresses such as those resulting from dissimilar metal welding parts.

• Nuclear Engineering and Technology
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• v.46 no.1
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• pp.73-80
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• 2014

A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches [1], those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1060-1068
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• 2019

This work aims to establish a model of a primary water stress corrosion crack growth rate of Alloy 690 material for the head penetration nozzles of Korean pressurized water reactors. The test material had an inhomogeneous microstructure with bands of fine-grains and intragranular carbides in the matrix of coarse-grains, which was similar to the archive materials of the head penetration nozzles. The crack growth rate was measured from the strain-hardened materials as a function of the stress intensity factor in simulated primary water at various temperatures and dissolved hydrogen contents. The effects of strain-hardening, temperature, and dissolved hydrogen on the crack growth rate were analyzed independently, and were then introduced as normalizing factors in the crack growth rate model. The crack growth rate model proposed in this work provides a key element of the tools needed to assess the progress of a stress corrosion crack when detected in thick-wall Alloy 690 components in Korean reactors.