• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.31-44
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• 2019

This study simulated a control rod assembly (CRA), which is a part of reactor shutdown systems, in immersed and fluid flow conditions. The CRA was inserted into the reactor core within a predetermined time limit under normal and abnormal operating conditions, and the CRA (which consists of complex geometric shapes) drop behavior is numerically modeled for simulation. A full-scale prototype CRA drop test is established under room temperature and water-fluid conditions for verification and validation. This paper describes the details of the numerical modeling and analysis results of the several conditions. Results from the developed numerical simulation code are compared with the test results to verify the numerical model and developed computer code. The developed code is in very good agreement with the test results and this numerical analysis model and method may replace the experimental and CFD method to predict the drop behavior of CRA.

• BMB Reports
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• v.41 no.12
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• pp.868-874
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• 2008

Neurofilaments (NFs) are neuronal intermediate filaments composed of light (NF-L), middle (NF-M), and heavy (NF-H) subunits. NF-L self-assembles into a "core" filament with which NF-M or NF-H co-assembles to form the neuronal intermediate filament. Recent reports show that point mutations of the NF-L gene result in Charcot-Marie-Tooth disease (CMT). However, the most recently described rod domain mutant of human NF-L (A148V) has not been characterized in cellular level. We cloned human NF-L and used it to engineer the A148V. In phenotypic analysis using SW13 cells, A148V mutation completely abolished filament formation despite of presence of NF-M. Moreover, A148V mutation reduced the levels of in vitro self-assembly using GST-NF-L (H/R) fusion protein whereas control (A296T) mutant did not affect the filament formation. These results suggest that alanine at position 148 is essentially required for NF-L self-assembly leading to subsequent filament formation in neuronal cells.

• Journal of Magnetics
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• v.21 no.4
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• pp.586-592
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• 2016

The magnetic-jack type CRDM withdraws or inserts a control rod assembly from/to the reactor core to control the core reactivity. The CRDM housings form not only the path of the electromagnetic field but also the pressure boundary of a nuclear reactor, and a periodic in-service inspection should be carried out if there are welded or flange jointed parts on the pressure boundary. The in-service inspection is a time-consuming process during the reactor refueling, and moreover it is difficult to perform the inspection over the reactor head. A magnetic motor housing is applied for the current SMART CRDM and has several welding joints, however a nonmagnetic motor housing with fewer or no welding joints may improve the operational efficiency of the nuclear reactor by avoiding or simplifying the in-service inspection process. Prior to the development, the magnetic field transfer efficiency of the nonmagnetic housing was required to be assessed. It was verified and optimized by the electromagnetic analysis of the lifting force estimation. Magnetic flux rings were adopted to improve the efficiency. In this paper, the design and optimization process of a nonmagnetic motor housing with the magnetic flux rings for the SMART CRDM are introduced and the analyses results are discussed.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1110-1119
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• 2020

The Floating Absorber for Safety at Transient (FAST) is a safety device used in the innovative Sodium-cooled Fast Reactor (iSFR). The FAST insert negative reactivity under transient or accident conditions. However, behavior of the FAST is still unclear under transient conditions. Therefore, the existing Floating Absorber for Safety at Transient Analysis Code (FASTAC) is improved to analyze the FAST movement by considering the reactivity and temperature distribution within the reactor core. The current FAST system is simulated under a single control rod withdrawal accident condition. In this investigation, the reactor thermal power does not return to its initial thermal power even if the FAST inserts negative reactivity. Only a 9 K of coolant temperature margin, in the hottest fuel assembly at EOL, can lead to unnecessary insertion of the negative reactivity. On the other hand, the FASTs cannot contribute to controlling the reactivity when normalized radial power is less than 0.889 at BOL and 0.972 at EOL. These simulation results suggest that the current FAST design needs to be optimized depending on its installed location. Meanwhile, the FAST system keeps the fuel, cladding and coolant temperatures below their limit temperatures with given conditions.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.504-511
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• 2018

In Lithuania, all spent nuclear fuel (SNF) resulted from the operation of the Ignalina Nuclear Power Plant (INPP), which had two Russian Acronym for "Channelized Large Power Reactor"-type reactors. After the final shutdown, the total amount of SNF at the INPP was approximately 22,000 fuel assemblies. All these assemblies will be stored for about 50 years and disposed of after that. The decision to shut down and decommission both reactors in Lithuania before termination of design period raises a significant challenge for the treatment of accumulated SNF. Therefore, various techniques and technologies for SNF management were developed and justified for that specific case, and a set of special equipment was installed at the INPP, the effectiveness of which was demonstrated during its operation. This article presents unique techniques related to the management of SNF adopted and commissioned at the INPP after its operation shutdown, namely fuel rod cladding leak tightness control system and special equipment for collection of possible spillage during handling of SNF assembly in the hot cell. The operational experience and measurement results of fuel rod cladding leak tightness control system are presented.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1403-1415
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• 2021

The transient capability of a SP₃ based pin-wise core analysis code SPHINCS is developed and verified through the analyses of the C5G7-TD benchmark. Spatial discretization is done by the fine mesh finite difference method (FDM) within the framework of the coarse mesh finite difference (CMFD) formulation. Pin size fine meshes are used in the radial fine mesh kernels. The time derivatives of the odd moments in the time-dependent SP₃ equations are neglected. The pin homogenized group constants and Super Homogenization (SPH) factors generated from the 2D single assembly calculations at the unrodded and rodded conditions are used in the transient calculations via proper interpolation involving the approximate flux weighting method for the cases that involve control rod movement. The simplifications and approximations introduced in SPHINCS are assessed and verified by solving all the problems of C5G7-TD and then by comparing with the results of the direct whole core calculation code nTRACER. It is demonstrated that SPHINCS yields accurate solutions in the transient behaviors of core power and reactivity.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.369-376
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• 2019

A complete solution for a soluble-boron-free (SBF) small modular reactor (SMR) is pursued with a new burnable absorber concept, namely centrally-shielded burnable absorber (CSBA). Neutronic flexibility of the CSBA design has been discussed with fuel assembly (FA) analyses. Major design parameters and goals of the SBF SMR are discussed in view of the reactor core design and three CSBA designs are introduced to achieve both a very low burnup reactivity swing (BRS) and minimal residual reactivity of the CSBA. It is demonstrated that the core achieves a long cycle length (~37 months) and high burnup (~30 GWd/tU), while the BRS is only about 1100 pcm and the radial power distribution is rather flat. This research also introduces a supplementary reactivity control mechanism using stainless steel as mechanical shim (MS) rod to obtain the criticality during normal operation. A further analysis is performed to investigate the local power peaking of the CSBA-loaded FA at MS-rodded condition. Moreover, a simple $B_4C$-based control rod arrangement is proposed to assure a sufficient shutdown margin even at the cold-zero-power condition. All calculations in this neutronic-thermal hydraulic coupled investigation of the 3D SBF SMR core are completed by a two-step Monte Carlo-diffusion hybrid methodology.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.817-825
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• 2022

A recent benchmarking effort, under the auspices of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA), has been made to evaluate the current state of modeling and simulation tools available to model fluoride salt-cooled high temperature reactors (FHRs). The FHR benchmarking effort considered in this work consists of several cases evaluating the neutronic parameters of a 2D prismatic FHR fuel assembly model using the participants' choice of simulation tools. Benchmark participants blindly submitted results for comparison with overall good agreement, except for some which significantly differed on cases utilizing a molybdenum-bearing control rod. Participants utilizing more recently updated explicit isotopic cross sections had consistent results, whereas those using elemental molybdenum cross sections observed reactivity differences on the order of thousands of pcm relative to their peers. Through a series of supporting tests, the authors attribute the differences as being nuclear data driven from using older legacy elemental molybdenum cross sections. Quantitative analysis is conducted on the control rod to identify spectral, reaction rate, and cross section phenomena responsible for the observed differences. Results confirm the observed differences are attributable to the use of elemental cross sections which overestimate the reaction rates in strong resonance channels.

• Journal of the Korean Society for Nondestructive Testing
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• v.18 no.6
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• pp.455-463
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• 1998

RCCA(rod cluster control assembly) End-Tip suffers from neutron irradiation and constant vibration due to high-speed internal flow of primary coolant during plant operation. Such operating conditions cause the RCCA end-tip crackings around tile circumferential weldment of the end-tip, and in some cases, the defective end-tips were completly broken loose. However, no reliable inspection techniques for end-tip crackings were developed in the past, although some techniques exist for inspecting RCCA control rod wears. Therefore, NDE group at KEPRI has developed an ECT technique for the detection and the sizing of the end-tip crackings. The technique uses a specially designed surface-riding probe that can detect size of circumferential crackings with an accuracy of ${\pm}5.31%$ RMS error. This paper describes the ECT instrumentation including the ECT probes, calibration bars, as well as technical approaches.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2712-2722
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• 2023

For an accurate and efficient time-dependent Monte Carlo (TDMC) neutron transport analysis, several advanced methods are newly developed and implemented in the Seoul National University Monte Carlo code, McCARD. For an efficient control of the neutron population, a dynamic weight window method is devised to adjust the weight bounds of the implicit capture in the time bin-by-bin TDMC simulations. A moving geometry module is developed to model a continuous insertion or withdrawal of a control rod. Especially, the history-based batch method for the TDMC calculations is developed to predict the unbiased variance of a bin-wise mean estimate. The developed methods are verified for three-dimensional problems in the C5G7-TD benchmark, showing good agreements with results from a deterministic neutron transport analysis code, nTRACER, within the statistical uncertainty bounds. In addition, the TDMC analysis capability implemented in McCARD is demonstrated to search the optimum detector positions for the pulsed-neutron-source experiments in the Kyoto University Critical Assembly and AGN201K.