• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.227-239
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• 2015

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, doublefault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel-coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.111-120
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• 2021

During the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel and steel are solidified into debris particles, which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes the debris particles relocated and the bed flattened, which called debris bed relocation. Because the thickness of debris bed has great influence on the cooling ability of fuel debris in low plenum, it's very necessary to evaluate the transient changes of the shape and thickness in relocation behavior for CDA simulation analysis. To simulate relocation behavior, a large number of debris bed relocation experiments were carried out by improved bottom gas-injection experimental method in this paper. The effects of different experimental factors on the relocation process were studied from the experiments. The experimental data were also used to further evaluate a semi-empirical onset model for predicting relocation.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.41 no.5
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• pp.375-380
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• 2017

This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the $B_{0.004}$ life would be guaranteed for the SFR fuel rods with 99％ confidence if no failure was observed at any of the contact surfaces of the five specimens.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.572-594
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• 2016

As one of the Gen-IV nuclear energy systems, a sodium-cooled fast reactor (SFR) is being developed at the Korea Atomic Energy Research Institute. As a long-term national research project, advanced radiation resistant oxide dispersion strengthened steel (ARROS) is being developed as an in-core fuel cladding tube material for a SFR in the future. In this paper, the current status of ARROS development is reviewed and its future prospective is discussed.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.193-206
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• 2007

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1135-1144
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• 2024

With the increasing ratio of renewables in the grid, a low-carbon and stable base load source that also is capable of load tracking is in demand. Sodium cooled fast reactors (SFRs) coupled to thermal heat storage system (THS) is a strong candidate for the need. This research focuses on the designing and performance validation of a two-tank THS based on molten salt to integrate with a 280 MW_th sodium cooled fast reactor. Designing of the THS includes the vital component, sodium-to-salt heat exchanger which is a technology gap that needs to be filled, and designing and parameter selection of the tanks and related pumps. Modeling of the designed THS is conducted followed by the description of operation strategies and control logics of the THS. Finally, the dynamic simulation of the designed THS is conducted based on Fortran. Results show, the proposed power system meets the need of the design requirements to store heat for 18 h during a day and provide 500 MW_th for peak demand for the rest of the day.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.51-52
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• 2009

Recently, several industrial countries using the fission energy have given attention to the Gen-IV SFR (sodium-cooled fast reactor) for achieving sustainable nuclear energy systems. In this context, an SFR is currently developed at the design concepts study stage in the Republic of Korea [Kim & Hahn 200909]. The sustainability of systems means economic, environment-friendly, proliferation-resistant, and safer systems. More specifically, this sustainability can be accomplished in terms of resource recycling and radioactive waste reduction. In the present work, the objective of fuel cycle cost modeling is to identify the impact of various conceptual options as a cost reduction measure for the Gen-IV SFR at the design concepts study stage. It facilitates the selection of several reasonable fuel cycle pathways for the future Gen-IV SFR from an economic viewpoint.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.10
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• pp.1251-1259
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• 2013

In this study, high temperature design and creep-fatigue damage evaluation of a decay heat exchanger (DHX) in the decay heat removal systems of a sodium-cooled fast reactor (SFR) have been performed. Detail design and 3D finite element analysis have been conducted for the DHXs to be installed in active and passive decay heat removal systems in Korean Generation IV SFR, and the DHX installed in the STELLA-1(Sodium integral effect test loop for safety simulation and assessment) at KAERI (Korea Atomic Energy Research Institute). Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two Mod.9Cr-1Mo steel heat exchangers according to the elevated temperature design codes of ASME Section III Subsection NH and RCC-MR code. Code comparisons were made based on the creep-fatigue damage evaluation and issues on conservatisms of the design codes were discussed.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.306-314
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• 2015

This paper examines the difference in the value of the nuclear fuel cycle cost calculated by the deterministic and probabilistic methods on the basis of an equilibrium model. Calculating using the deterministic method, the direct disposal cost and Pyro-SFR (sodium-cooled fast reactor) nuclear fuel cycle cost, including the reactor cost, were found to be 66.41 mills/kWh and 77.82 mills/kWh, respectively (1 mill = one thousand of a dollar, i.e., $10^{-3}$ $). This is because the cost of SFR is considerably expensive. Calculating again using the probabilistic method, however, the direct disposal cost and Pyro-SFR nuclear fuel cycle cost, excluding the reactor cost, were found be 7.47 mills/kWh and 6.40 mills/kWh, respectively, on the basis of the most likely value. This is because the nuclear fuel cycle cost is significantly affected by the standard deviation and the mean of the unit cost that includes uncertainty. Thus, it is judged that not only the deterministic method, but also the probabilistic method, would also be necessary to evaluate the nuclear fuel cycle cost. By analyzing the sensitivity of the unit cost in each phase of the nuclear fuel cycle, it was found that the uranium unit price is the most influential factor in determining nuclear fuel cycle costs.

• Journal of the Korean Society of Mechanical Technology
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• v.13 no.2
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• pp.145-153
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• 2011

The irradiation tests of materials in HANARO have been performed usually at temperatures below $300^{\circ}C$ at which the RPV(Reactor Pressure Vessel) materials of the commercial reactors such as the light water reactor and CANDU are operated. As VHTR(Very High Temperature Reactor) and SFR(Sodium-cooled Fast Reactor) projects are being carried as a part of the present Gen-IV program in Korea, the requirements for irradiation of materials at temperatures higher than $500^{\circ}C$ are recently being gradually increased. To overcome the restriction in the use at high temperature of the existing Al thermal media, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated more advanced than the single thermal media capsule. At the irradiation test of the capsule, the temperature of the specimens successfully reached $700^{\circ}C$ and the integrity of Al, Ti and graphite material was maintained.