• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.132-140
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• 2024

In the event of a severe accident in Sodium Cooled Fast Reactors (SFR), the sodium combustion aerosols along with fission product aerosols would migrate to the environment through leak paths of the Reactor Containment Building (RCB) concrete wall under positive pressure. Understanding the characteristics of sodium aerosol transport through concrete leak paths is important as it governs the environmental source term. In this context, experiments are conducted to study the influence of various parameters like pressure, initial mass concentration, leak path diameter, humidity etc., on the transport and deposition of sodium aerosols in straight leak paths of concrete. The leak paths in concrete specimens are prepared by casting and the diameter of the leak path is measured using thermography technique. Aerosol transport experiments are conducted to measure the transported and plugged aerosol mass in the leak paths and corresponding plugging times. The values of differential pressure, aerosol concentration and relative humidity taken for the study are in the ranges 10-15 kPa, 0.65-3.04 g/m³ and 30-90% respectively. These observations are numerically simulated using 1-Dimensional transport equation. The simulated values are compared with the experimental results and reasonable agreement among them is observed. From the safety assessment view of reactor, the approach presented here is conservative as it is with straight leak paths.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1482-1494
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• 2022

Power conversion cycles (Subcritical Steam, Supercritical Steam, Open Air Brayton, Recuperated Air Brayton, Combined Cycle, Closed Brayton Supercritical CO₂ (sCO₂), and Stirling) are evaluated for land-based nuclear microreactors based on technical maturity, system efficiency, size, cost and maintainability, safety implications, and siting considerations. Based upon these criteria, Air Brayton systems were selected for further evaluation. A brief history of the development and applications of Brayton power systems is given, followed by a description of how these thermal-to-electrical energy conversion systems might be integrated with a nuclear microreactor. Modeling is performed for optimized cycles operating at 3 MW(e) with turbine inlet temperatures of 500 ℃, 650 ℃ and 850 ℃, corresponding to: a) sodium fast, b) molten salt or heat pipe, and c) helium or sodium thermal reactors, coupled with three types of Brayton power conversion units (PCUs): 1) simple open-cycle gas turbine, 2) recuperated open-cycle gas turbine, and 3) recuperated and intercooled open-cycle gas turbine. Aeroderivative turboshaft engines employing the simple Brayton cycle and two industrial gas turbine engines employing recuperated air Brayton cycles are also analyzed. These engines offer mature technology that can facilitate near-term deployment with a modest improvement in efficiency.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.731-744
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• 2013

Energy security is a topic of high importance to many countries throughout the world. Countries with access to vast energy supplies enjoy all of the economic and political benefits that come with controlling a highly sought after commodity. Given the desire to diversify away from fossil fuels due to rising environmental and economic concerns, there are limited technology options available for baseload electricity generation. Further complicating this issue is the desire for energy sources to be sustainable and globally scalable in addition to being economic and environmentally benign. Nuclear energy in its current form meets many but not all of these attributes. In order to address these limitations, TerraPower, LLC has developed the Traveling Wave Reactor (TWR) which is a near-term deployable and truly sustainable energy solution that is globally scalable for the indefinite future. The fast neutron spectrum allows up to a ~30-fold gain in fuel utilization efficiency when compared to conventional light water reactors utilizing enriched fuel. When compared to other fast reactors, TWRs represent the lowest cost alternative to enjoy the energy security benefits of an advanced nuclear fuel cycle without the associated proliferation concerns of chemical reprocessing. On a country level, this represents a significant savings in the energy generation infrastructure for several reasons 1) no reprocessing plants need to be built, 2) a reduced number of enrichment plants need to be built, 3) reduced waste production results in a lower repository capacity requirement and reduced waste transportation costs and 4) less uranium ore needs to be mined or purchased since natural or depleted uranium can be used directly as fuel. With advanced technological development and added cost, TWRs are also capable of reusing both their own used fuel and used fuel from LWRs, thereby eliminating the need for enrichment in the longer term and reducing the overall societal waste burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P) reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.359-367
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• 2018

Long-lived Small Modular Reactors are being promoted as an innovative way of catering to emerging markets and isolated regions. They can be operated continuously for decades without requiring additional fuel. A novel configuration of long-lived reactor core employs a mixed neutron spectrum, providing an improvement in nonproliferation metrics and in safety characteristics. Starting with a base sodium reactor design, moderating material is inserted in outer core assemblies to modify the fast spectrum. The assemblies are shuffled once during core lifetime to ensure that every fuel rod is exposed to the thermalized spectrum. The Mixed Spectrum Reactor is able to maintain a core lifetime over two decades while ensuring the plutonium it breeds is below the weapon-grade limit at the fuel discharge. The main drawbacks of the design are higher front-end fuel cycle costs and a 58% increase in core volume, although it is alleviated to some extent by a 48% higher power output.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.925-936
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• 2020

To ascertain the characteristics of pool sloshing behavior that might be encountered during a core disruptive accident of sodium-cooled fast reactors, in our earlier work several series of experiments were conducted under various scenarios including the condition with mono-sized solid particles. It is found that under the particle-bed condition, three typical flow regimes (namely the bubble-impulsion dominant regime, the transitional regime and the bed-inertia dominant regime) could be identified and a flow-regime model (base model) has been even successfully established to estimate the regime transition. In this study, aimed to further understand this behavior at more realistic particle-bed conditions, a series of simulated experiments is newly carried out using mixed-size particles. Through analyses, it is verified that for present scenario, by applying the area mean diameter, our previously-developed base model can provide the most appropriate predictive results among the various effective diameters. To predict the regime transition with a form of extension scheme, a correction factor which is based on the volume-mean diameter and the degree of convergence in particle-size distribution is suggested and validated. The conducted analyses in this work also indicate that under certain conditions, the potential separation between different particle components might exist during the sloshing process.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.34 no.11
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• pp.991-998
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• 2010

The effect of increasing the elevation of an IHX (intermediate heat exchanger) on the transient performance of the KALIMER-600 reactor pool during the early phase of a loss of normal heat sink accident was investigated. Three reactors equipped with IHXs that were elevated to different heights were designed, and the thermal-hydraulic analyses were carried out for the steady and transient state by using the COMMIX-1AR/P code. In order to analyze the effects of the elevation of an IHX between reactors, various thermal-hydraulic properties such as mass flow rate, core peak temperature, RmfQ (ratio of mass flow over Q) and initiation time of decay heat removal via DHX (decay heat exchanger) were evaluated. It was found that with an increase in the IHX elevation, the circulation flow rate increases and a steep rise in the core peak temperature under the same coastdown flow condition is prevented without a delay in the initiation of the second stage of cooling. The available coastdown flow range in the reactor could be increased by increasing the elevation of the IHX.

• Journal of Environmental Health Sciences
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• v.31 no.2 s.83
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• pp.99-106
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• 2005

The removals of organic matter, nitrogen and phosphate in wastewater were investigated with Sequencing Batch Reactor (SBR). Glucose and sodium acetate were Used for organic carbon source so as to know nutrient removal efficiency in proportion to MLSS concentration. In the case of glucose, the COD removal rate was $74\%,\;41\%\;and\;66\%$ in MLSS 5000, 3000 and 1000, respectively. On equal terms, the BOD was $57\%,\;21\%\;and\;38\%$, the T-N was $24\%,\;13\%\;and\;44\%$, and the T-P was $12\%,\;21\%\;and\;33\%$. As a result, the removal rate of organic materials showed the finest remove when MLSS was 5000, but the nutrient removal rate appeared as was best when MLSS was 1000. In the case of sodium acetate, the COD removal rate was $83\%,\;81\%\;and\;86\%$ in MLSS 5000, 3000 and 1000, respectively. On equal terms, the BOD was appeared by $76\%,\;82\%\;and\;92\%$, the T-N $57\%,\;42\%\;and\;78\%$, and the T-P $48\%,\;52\%\;and\;38\%$. As a result, organic and T-N removal rates were best when MLSS was 1000. But, the T-P removal rates were best when MLSS was 3000. Glucose was shown fast removal in reaction beginning, but screened by more efficient thing though sodium acetate removes organic matter, nitrogen and phosphate. Form of floc was ideal in all reactors regardless of carbon source and MLSS concentration. And its diameter was about $200\~500{\mu}m$.

• Transactions of the KSME C: Technology and Education
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• v.3 no.4
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• pp.257-264
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• 2015

To maintain sustainability of nuclear energy as an important energy source, both safety problem and Spent Nuclear Fuels(SNFs) problem should be solved. In case of Gen-IV reactors such as fast reactor, SNFs can be used as fuels by using fast neutrons. It can be a suitable treatment method of high-level waste in near future. Liquid metals such as Sodium or Lead-Bismuth Eutectic (LBE) can be possibly used as a coolant to use fast neutrons. In this paper, it was described that natural circulation parameter studies, design analyses, material selections and a completion of facilities. To develop a natural circulation facility, thermal hydraulic analyses were performed. Installation technique of liquid metal natural circulation were secured.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.915-922
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• 2018

Metallic fuel has been considered for sodium-cooled fast reactors because it can maximize the uranium resources. It generates rare earth elements as fission products, where it is reported by aggravating the fuel-cladding chemical interaction at the operating temperature. Rare earth elements form a multicomponent alloy (Ce-Nd-Pr-La-Sm-etc.) during reactor operation, where it shows a higher reaction thickness than a single element. Experiments have been carried out by simplifying multicomponent alloys for mono or binary systems because complex alloys have difficulty in the analysis. In previous experiments, xCe-yNd was fabricated with two elements, Ce and Nd, which have a major effect on the fuel-cladding chemical interaction, and the thickness of the reaction layer reached maximum when the rare earth elements ratio was 1:1. The objective of this study is to evaluate the effect and relationship of rare earth elements on such synergistic behavior. Single and binary rare earth model alloys were prepared by selecting five rare earth elements (Ce, Nd, Pr, La, and Sm). In the single system, Nd and Pr behaviors were close to diffusion, and Ce showed a eutectic reaction. In the binary system, Ce and Sm further increased the reaction layer, and La showed a non-synergy effect.

• 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.