• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2792-2800
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• 2022

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO₂ kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950-1100 ℃. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2112-2124
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• 2023

With the rise of economic and safety standards for nuclear reactors, new concepts of Gen-IV reactors and modular reactors showed more complex designs that challenge current tools for reactor physics analysis. A Monte Carlo (MC) two-step method was proposed in this work. This calculation scheme uses the continuous-energy MC method to generate multi-group cross-sections from heterogeneous models. The multi-group MC method, which can adapt locally-heterogeneous models, is used in the core calculation step. This calculation scheme is verified using a Gen-IV modular lead-based fast reactor (LFR) benchmark case. The influence of homogenized patterns, scatter approximations, flux separable approximation, and local heterogeneity in core calculation on simulation results are investigated. Results showed that the cross-sections generated using the 3D assembly model with a locally heterogeneous representation of control rods lead to an accurate estimation with less than 270 pcm bias in core reactivity, 0.5% bias in control rod worth, and 1.5% bias on power distribution. The study verified the applicability of multi-group cross-sections generated with the MC method for LFR analysis. The study also proved the feasibility of multi-group MC in core calculation with local heterogeneity, which saves 85% time compared to the continuous-energy MC.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.893-899
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• 2024

In a secondary cooling system of a sodium-cooled fast reactor (SFR), rapid detection of hydrogen due to sodium-water reaction (SWR) caused by water leakage from a heat exchanger tube of a steam generator (SG) is important in terms of safety and property protection of the SFR. For hydrogen detection, the hydrogen detectors using atomic transmission phenomenon of hydrogen within Ni-membrane were used in Japanese proto-type SFR "Monju". However, during the plant operation, detection signals of water leakage were observed even in the situation without SWR concerning temperature up and down in the cooling system. For this reason, the study of a new hydrogen detector has been carried out to improve stability, accuracy and reliability. In this research, the authors focus on the difference in composition of hydrogen and the difference between the background hydrogen under normal plant operation and the one generated by SWR and theoretically estimate the hydrogen behavior in liquid sodium by using ultra-accelerated quantum chemical molecular dynamics (UA-QCMD). Based on the estimation, dissolved H or NaH, rather than molecular hydrogen (H₂), is the predominant form of the background hydrogen in liquid sodium in terms of energetical stability. On the other hand, it was found that hydrogen molecules produced by the sodium-water reaction can exist stably as a form of a fine bubble concerning some confinement mechanism such as a NaH layer on their surface. At the same time, we observed experimentally that the fine H₂ bubbles exist stably in the liquid sodium, longer than previously expected. This paper describes the comparison between the theoretical estimation and experimental results based on hydrogen form in sodium in the development of the new hydrogen detector in Japan.

• Corrosion Science and Technology
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• v.4 no.3
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• pp.108-113
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• 2005

One of the recent safety issues in the pressurized heavy water reactor (PHWR) is the cracking of the feeder pipe. Because of the limited accessibility to the cracked region and a high dose of radiation exposure, it is difficult to inspect all the pipes with the conventional ultrasonic method. In order to solve this problem, a long-range guided wave technique has been developed. A computer program to calculate the dispersion curves in the pipe was developed and the dispersion curves for the feeder pipes in PHWR plants were determined. Several longitudinal and/or flexural modes were selected from the review of the dispersion curves and an actual experiment has been carried out with the specific alignment of the piezoelectric ultrasonic transducers. They were confirmed as L(0,1)) and/or flexural modes(F(m,2)) by the short time Fourier transformation(STFT) and were sensitive to the circumferential cracks, but not to the axial cracks in the pipe. An electromagnetic acoustic transducers(EMAT) was designed and fabricated for the generation and reception of the torsional guided wave. The axial cracks were detected by a torsional mode(T(0,1)) generated by the EMAT.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.3464-3466
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• 2007

Power uprate is the process of increasing the maximum power level at which a commercial nuclear power plant may operate. Power uprate applications(113 units) for NPPs(Nuclear Power Plants) were recently approved in the United States. Utilities have been using power uprates since the 1970s as a way of increasing the power output of their nuclear plants. To increase the power output of a reactor, typically more highly enriched uranium fuel and/or more fresh fuel is used. This enables the reactor to produce more thermal energy and therefore more steam, driving a turbine generator to produce electricity. In this paper, the propriety of power uprate is explained through the review on the power uprate method and the changes of the physical parameters due to power uprate. The analysis results showed that the CDF(Core Damage Frequency) and LERF(Large Early Release Frequency) are affected in the current probabilistic safety assessment (PSA) model.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1974-1982
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• 2020

Numerical prediction of fission product plateout and circulating coolant activities under normal operating conditions is crucial in the design of a high temperature gas-cooled reactor (HTGR). The results are used for the maintenance and repair of the components as well as the safety analysis regarding early source terms under loss of coolant accident scenarios. In this work, a new computer code named POSCA (Plate-Out Surface and Circulating Activities) was developed based on a one-dimensional model to evaluate fission product plateout and circulating coolant activities within the primary circuit of a HTGR. The verification and validation of study for the POSCA code was done using available analytical results and two in-pile experiments (i.e., OGL-1 and VAMPYR-1). The results of the POSCA calculations show that POSCA is able to simulate plateout and circulating coolant activities in a HTGR with fast computation and reasonable accuracy.

• The KSFM Journal of Fluid Machinery
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• v.3 no.2 s.7
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• pp.24-30
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• 2000

The structural analysis of a reactor coolant pump(RCP) of a nuclear power plant is very important for the safety assessment of the plant. Accurate boundary conditions for the heat transfer coefficient are required for reliable thermal stress analysis of the pump casing, especially in transient operations of the pump since the coolant properties are largely dependent on operational conditions. In the present study, a 3D mixed flow type coolant pump was modeled from the RCP drawings and analyzed in the steady state and number of transient flow conditions by using a commercial code STAR-CD. From the result of the computation, it is seen that the average heat transfer coefficients for the cases considered are found to be the suggested values of the manufacturer, Westinghouse Energy System. The unevenness in local heat transfer coefficients, however, is found to be considerable so that the use of average heat transfer coefficients in all boundaries might not give reliable thermal stress predictions.

• Journal of computational fluids engineering
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• v.15 no.1
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• pp.1-8
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• 2010

RCCS is a passive safety-related system that removes the decay heat of VHTR when normal decay heat removal systems are in failure. Understanding thermo-hydraulics of RCCS is important to design a safer VHTR. RCCS consists of 292 cooling panels, which are placed in the reactor cavity. The layout of RCCS gives an idea that, for CFD simulations, cooling panels can be assumed to be one annulus tube. This assumption can reduce significantly the computational time, especially for the unsteady simulation. To simulate RCCS in an axisymmetric manner, three models were suggested and compared. Each model has (1) the same outer radius, (2) the same cross-sectional area (3) the same pressure drop, respectively, as the RCCS cooling panels. The steady-state simulation was conducted with these three models and the DO radiation model. It is found that over 90% of the heat from the outer wall of the reactor pressure vessel is transported to the RCCS by radiative heat transfer. The simulation with the third model, which has the same pressure drop as the design, estimates the closest wall temperature profiles to a thermo-hydraulic code, GAMMA+, result.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2002.05a
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• pp.135-141
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• 2002

An optimization of KTF thorium fuel assembly design was performed on the basis of the design parameter studies. Optimization goals ware to make the core have both proliferation resistance and fuel cycle economics. Four kinds of proliferation resistance indexes were used; SNS, TG, BCM, Toxicity. A new index, FEI was regarded as a limiting index for the maximization of fuel cycle economics. Optimized thorium fuel design was applied for APR-1400 reactor core. Nuclear core design procedures were examined to solve the thorium fuel reactor problems. It was shown that heterogeneous thorium fuel core option is acceptable in safety and economics aspects.

• Journal of Energy Engineering
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• v.21 no.1
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• pp.75-80
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• 2012

In this paper a realistic bounding method for flooding analysis following rupture of large size of thanks and piping is defined. Mass and energy release during main feedwater line break accident is analyzed with RETRAN code. It is modeled that the main feed water control valve is closed in 5.0 seconds after reactor trip. In result of the analysis, largest mass and energy is discharged at 70% reactor power. The flood sources for main feedwater room are calculated when piping failure occurs in the high energy line and medium energy line. Based on the result of flood level (1.43m), it is investigated that all of the safety-related environmental qualification equipments are well located above the flood level.