• Journal of Magnetics
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• v.15 no.4
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• pp.173-178
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• 2010

A new kind of magnetic rubber, Fe dispersed ethylene propylene monomer (EPM), was prepared by a conventional technique using a two roll mill. The magnetic fillers of Fe-nanoparicles were coated by low density polyethylene (LDPE). The purpose of surface treatment of nanoparticles by LDPE is to enhance wettability and lubricancy of the fillers in a polymer matrix. The mechanical strength and microstructure of the magnetic rubber were characterized by tensile strength test and scanning electron microscopy (SEM). Results revealed that the Fe nanoparticles were relatively well dispersed in an EPM matrix. It was found that the nano- Fe dispersed magnetic rubber showed higher coercivity and tensile strength than those of micron- Fe dispersed one.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2172-2194
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• 2023

A variational nodal method (VNM) with unstructured-mesh is presented for solving steady-state and dynamic neutron diffusion equations. Orthogonal polynomials are employed for spatial discretization, and the stiffness confinement method (SCM) is implemented for temporal discretization. Coordinate transformation relations are derived to map unstructured triangular nodes to a standard node. Methods for constructing triangular prism space trial functions and identifying unique nodes are elaborated. Additionally, the partitioned matrix (PM) and generalized partitioned matrix (GPM) methods are proposed to accelerate the within-group and power iterations. Neutron diffusion problems with different fuel assembly geometries validate the method. With less than 5 pcm eigenvalue (k_eff) error and 1% relative power error, the accuracy is comparable to reference methods. In addition, a test case based on the kilowatt heat pipe reactor, KRUSTY, is created, simulated, and evaluated to illustrate the method's precision and geometrical flexibility. The Dodds problem with a step transient perturbation proves that the SCM allows for sufficiently accurate power predictions even with a large time-step of approximately 0.1 s. In addition, combining the PM and GPM results in a speedup ratio of 2-3.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.282-287
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• 1996

We. develop in this study a wavelet transform method to apply to the flux reconstruction problem in reactor analysis. When we reconstruct pinwise heterogeneous flux by iterative methods, a difficulty arises due to the near singularity of the matrix as the mesh size becomes finer. Here we suggest a wavelet transform to tower the spectral radius of the near singular matrix and thus to converge by a standard iterative scheme. We find that the spectral radios becomes smatter than one after the wavelet transform is performed on sample problems.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.470-477
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• 2018

With a highly functional methyl vinyl silicone rubber (VMQ) matrix and filler materials of $B_4C$, PbO, and benzophenone (BP) and through powder surface modification, silicone rubber mixing, and vulcanized molding, a flexible radiation shielding and resistant composite was prepared in the study. The dispersion property of the powder in the matrix filler was improved by powder surface modification. BP was added into the matrix to enhance the radiation resistance performance of the composites. After irradiation, the tensile strength, elongation, and tear strength of the composites decreased, while the Shore hardness of the composites and the crosslinking density of the VMQ matrix increased. Moreover, the composites with BP showed better mechanical properties and smaller crosslinking density than those without BP after irradiation. The initial degradation temperatures of the composites containing BP before and after irradiation were $323.6^{\circ}C$ and $335.3^{\circ}C$, respectively. The transmission of neutrons for a 2-mm thick sample was only 0.12 for an Am-Be neutron source. The transmission of ${\gamma}$-rays with energies of 0.662, 1.173, and 1.332 MeV for 2-cm thick samples were 0.7, 0.782, and 0.795, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.95-106
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• 2019

The decommissioning of nuclear power plants should be prepared by quantitative and qualitative risk assessment. Radiological and non-radiological hazards arising during decommissioning activities must be assessed to ensure the safety of decommissioning workers and the public. Decommissioning experiences by U.S. operators have mainly focused on deterministic risk assessment, which is standardized by the U.S. Nuclear Regulatory commission (NRC) and focuses only on the consequences of risk. However, the International Atomic Energy Agency (IAEA) has suggested an alternative to the deterministic approach, called the risk matrix technique. The risk matrix technique considers both the consequence and likelihood of risk. In this study, decommissioning stages, processes, and activities are organized under a work breakdown structure. Potential accidents in the decommissioning process of NPPs are analyzed using the composite risk matrix to assess both radiological and non-radiological hazards. The levels of risk for all potential accidents considered by U.S. NPP operators who have performed decommissioning were estimated based on their consequences and likelihood of events.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2067-2076
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• 2022

The importance of safety culture has been emphasized to achieve a high level of safety. In this light, a systematic method to more properly deal with safety culture is necessary. Here, a decision-making tool that can apply a graded approach to the analysis of safety culture is proposed, called the F-D matrix, which determines the frequency and the difficulty of safety culture attributes recently defined by the IAEA. A hierarchical model of difficulty contributors was developed as a scoring standard, and its elements were weighted via expert evaluation using the analytic hierarchy process. The frequency of the attributes was derived by analyzing reported events from nuclear power plants in the Republic of Korea. Period-by-period comparisons with the F-D matrix can show trends in the change of the maturity level of an organization's safety culture and help to evaluate the effectiveness of previously implemented measures. In the evaluating the difficulty of the attributes in the recently developed harmonized safety culture model, the difficulties of Trending, Benchmarking, Resilience, and Documentation and Procedures were found to be relatively high, while the difficulties of Conflicts are Resolved, Ownership, Collaboration, and Respect is Evident were found to be relatively low. A case study was conducted with an analysis period of 10 years to attempt to reflect the many changes in safety culture that have been made following the Fukushima accident in March 2011. As a result of comparing two periods following the Fukushima accident, the overall frequency decreased by about 40%, providing evidence for the effects of the various improvements and measures taken following the increased emphasis on safety culture. The proposed F-D matrix provides a new analytical perspective and enables an in-depth analysis of safety culture.

• The Korean Journal of Nuclear Medicine Technology
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• v.24 no.1
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• pp.27-32
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• 2020

Purpose Glomerular Filtration Rate(GFR) is an important indicator for evaluating renal function and monitoring the progress of renal disease. Currently, the method of measuring GFR in clinical trials by using serum creatinine value and ^99mTc-DTPA(diethylenetriamine pentaacetic acid) renal dynamic scan is still useful. After the Gates method of formula was announced, when ^99mTc-DTPA Renal dynamic scan is taken, it is applied the GFR is measured using a gamma camera. The purpose of this paper is to measure the GFR by applying the Gates method of formula. It is according to effect activity and matrix size that is related in the GFR. Materials and Methods Data from 5 adult patients (patient age = 62 ± 5, 3 males, 2 females) who had been examined ^99mTc-DTPA Renal dynamic scan were analyzed. A dynamic image was obtained for 21 minutes after instantaneous injection of ^99mTc-DTPA 15 mCi into the patient's vein. To evaluate the glomerular filtration rate according to changes in activity and matrix size, total counts were measured after setting regions of interest in both kidneys and tissues in 2-3 minutes. The distance from detector to the table was maintained at 30cm, and the capacity of the pre-syringe (PR) was set to 15, 20, 25, 30 mCi, and each the capacity of post-syringe (PO) was 1, 5, 10, 15 mCi is set to evaluate the activity change. And then, each matrix size was changed to 32 × 32, 64 × 64, 128 × 128, 256 × 256, 512 × 512, and 1024 × 1024 to compare and to evaluate the values. Results As the activity increased in matrix size, the difference in GFR gradually decreased from 52.95% at the maximum to 16.67% at the minimum. The GFR value according to the change of matrix size was similar to 2.4%, 0.2%, 0.2% of difference when changing from 128 to 256, 256 to 512, and 512 to 1024, but 54.3% of difference when changing from 32 to 64 and 39.43% of difference when changing from 64 to 128. Finally, based on the presently used protocol, 256 × 256, PR 15 mCi and PO 1 mCi, the GFR value was the largest difference with 82% in PR 15 mCi and PO 1 mCi. conditions, and at the least difference is 0.2% in the conditions of PR 30 mCi and PO 15 mCi. Conclusion Through this paper, it was confirmed that when measuring the GFR using the gate method in the ^99mTc-DTPA renal dynamic scan. The GFR was affected by activity and matrix size changes. Therefore, it is considered that when taking the ^99mTc-DTPA renal dynamic scan, is should be careful by applying appropriate parameters when calculating GFR in the every hospital.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3095-3105
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• 2022

Solid Microencapsulated Fuel (SMF) is a type of solid fuel rod design that disperses TRISO coated fuel particles directly into a kind of matrix. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study focused on the neutronics and some of the fuel cycle characteristics of SMF by using OpenMC. Two kinds of SMFs have been designed and evaluated - fuel particles dispersed into a silicon carbide matrix and fuel particles dispersed into a zirconium matrix. A 7×7 fuel assembly with increased rod diameter transformed from the standard NHR200-II 9×9 array was also introduced to increase the heavy metal inventory. A preliminary study of two kinds of burnable poisons (Erbia & Gadolinia) in two forms (BISO and QUADRISO particles) was also included. This study found that SMF requires about 12% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in NHR200-II, which is about 7% for SMF with increased rod diameter. Feedback coefficients are less negative through the life of SMF than the reference. And it is estimated that the average center temperature of fuel kernel at fuel rod centerline is about 60 K below that of reference in this paper.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3732-3753
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• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2422-2430
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• 2020

In order to improve calculational efficiency of the CAPP code in the analysis of the hexagonal reactor core, we have tried to implement a refined AFEN method with transverse gradient basis functions and interface flux moments in the hexagonal geometry. The numerical scheme for the refined AFEN method adopted here is the response matrix method that uses the interface partial currents as nodal unknowns instead of the interface fluxes used in the original AFEN method. Since the response matrix method is single-node based, it has good properties such as good calculational efficiency and parallel computing affinity. Because a refined AFEN method equivalent nonlinear FDM response matrix method tried first could not provide a numerically stable solution, a direct formulation of the refined AFEN response matrix were developed. To show the numerical performance of this response matrix method against the original AFEN method, the numerical error analyses were performed for several benchmark problems including the VVER-440 LWR benchmark problem and the MHTGR-350 HTGR benchmark problem. The results showed a more than three times speedup in computing time for the LWR and HTGR benchmark problems due to good convergence and excellent calculational efficiency of the refined AFEN response matrix method.