• Solar Energy
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• v.14 no.2
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• pp.91-101
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• 1994

Heat release characteristics of a CaO packed bed reactor which is used for a chemical heat storage device has been studied. We employed Cu-plate fins to release the heat of reaction of the CaO packed bed inside the reactor fast and effectively. Two-dimensional analysis of unsteady state heat flow inside the bed was performed as a function of time and under various conditions of the Cu-plates. It is noted that the time required to release the heat of reaction with Cu fins is reduced more than twice fast compared to that without Cu fins. That was largely dependent upon the number of Cu-plate, as well.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2635-2649
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• 2022

CEFR is a small core-size sodium-cooled fast reactor (SFR) using high enrichment fuel with stainless-steel reflectors, which brings a significant challenge to the deterministic methodologies due to the strong spectral effect. The neutronic simulation of the start-up experiments conducted at the CEFR have been performed with a deterministic code system RAST-F, which is based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo (MC) code and a multi-group nodal diffusion solver. The RAST-F results were compared against the measurement data. Moreover, the characteristic of neutron spectrum in the fuel rings, and adjacent reflectors was evaluated using different models for generation of accurate nuclear libraries. The numerical solution of RAST-F system was verified against the full core MC solution MCS at all control rods fully inserted and withdrawn states. A good agreement between RAST-F and MCS solutions was observed with less than 120 pcm discrepancies and 1.2% root-mean-square error in terms of k_eff and power distribution, respectively. Meanwhile, the RAST-F result agreed well with the experimental values within two-sigma of experimental uncertainty. The good agreement of these results indicating that RAST-F can be used to neutronic steady-state simulations for small core-size SFR, which was challenged to deterministic code system.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.1954-1961
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• 2022

Lead-based fast reactor (LFR) has become one of the most promising reactors for Generation IV nuclear systems. A developing trend of LFR is high efficiency, along with operation temperatures up to 800 ℃ or even higher. One of key issues in the high-efficiency LFR is corrosion of cladding materials with lead at high temperatures. In this study, corrosion behavior of some refractory metals (Nb, Nb521, and Mo-0.5La) was investigated in static lead at 1000 ℃ for 1000 h. The results showed that Nb and Nb521 exhibited an intense dissolution corrosion with obvious lead penetration after corrosion, and lead penetration extended along the grain boundaries of the specimens. Furthermore, Nb521 showed a better corrosion resistance than that of Nb as a result of the elements of W and Mo included in Nb521. Mo-0.5La showed much better corrosion resistance than that of Nb and Nb521, and no lead penetration could be observed. However, an etched morphology appeared on the surface of Mo-0.5La, indicating the occurrence of corrosion to a certain degree. The results indicate that Mo-0.5La is compatible with lead up to 1000 ℃. While Nb and Nb alloys might be not compatible with lead for high-efficiency LFR at such high temperatures.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2697-2711
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• 2023

In this study, the validity of reducing the number of gas turbine stages designed for a nitrogen Brayton cycle coupled to a sodium-cooled fast reactor was assessed. The turbine performance was evaluated through computational fluid dynamics (CFD) simulations under different off-design conditions controlled by a reduced flow rate and reduced rotational speed. Two different multistage gas turbines designed to extract almost the same specific work were selected: two- and three-stage turbines (mid-span stage loading coefficient: 1.23 and 1.0, respectively). Real gas properties were considered in the CFD simulation in accordance with the Peng-Robinson's equation of state. According to the CFD results, the off-design performance of the two-stage turbine is comparable to that of the three-stage turbine. Moreover, compared to the three-stage turbine, the two-stage turbine generates less entropy across the shock wave. The results indicate that under both design and off-design conditions, increasing the stage loading coefficient for a fewer number of turbine stages is effective in terms of performance and size. Furthermore, the Ellipse law can be used to assess off-design performance and increasing exponent of the expansion ratio term better predicts the off-design performance with a few stages (two or three).

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3571-3580
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• 2023

The application of the induction bending process to pipe systems in various industrial fields is increasing. Recently, efforts have also been made to apply this bending process to nuclear power plants because it can innovatively reduce welded parts of the curved pipes, such as elbows. However, there have been no cases of the application of induction bending to the piping of nuclear power plants. In this study, the applicability of the P91 induction bending piping for the sodium-cooled fast reactor PGSFR was validated through high temperature low cycle fatigue tests and creep tests using P91 induction bending pipe specimens. The tests confirmed that the materials sufficiently satisfied the fatigue life and the creep rupture life requirements for P91 steel at 550 ℃ in the ASME B&PV Code, Sec. III, Div. 5. The results show that the effects of heating and bending by the induction bending process on the material properties were not significant and the induction bending process could be applicable to piping system of PGSFR well.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.873-879
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• 2024

A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1472-1479
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• 2024

Cyber-Physical Energy Systems (CPESs) integrate cyber and hardware components to ensure a reliable and safe physical power production and supply. Renewable Energy Sources (RESs) add uncertainty to energy demand that can be dealt with flexible operation (e.g., load-following) of CPES; at the same time, scenarios that could result in severe consequences due to both component stochastic failures and aging of the cyber system of CPES (commonly overlooked) must be accounted for Operation & Maintenance (O&M) planning. In this paper, we make use of Deep Reinforcement Learning (DRL) to search for the optimal O&M strategy that, not only considers the actual system hardware components health conditions and their Remaining Useful Life (RUL), but also the possible accident scenarios caused by the failures and the aging of the hardware and the cyber components, respectively. The novelty of the work lies in embedding the cyber aging model into the CPES model of production planning and failure process; this model is used to help the RL agent, trained with Proximal Policy Optimization (PPO) and Imitation Learning (IL), finding the proper rejuvenation timing for the cyber system accounting for the uncertainty of the cyber system aging process. An application is provided, with regards to the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED).

• Archives of Metallurgy and Materials
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• v.64 no.3
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• pp.953-957
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• 2019

U-10wt.%Zr metallic fuel slugs containing rare-earth (RE: a rare-earth alloy comprising 53% Nd, 25% Ce, 16% Pr and 6% La) elements for a sodium-cooled fast reactor were fabricated by modified injection casting as an alternative method. The distribution, size and composition of the RE inclusions in the metallic fuel slugs were investigated according to the content of the RE inclusions. There were no observed casting defects, such as shrunk pipes, micro-shrinkage or hot tears formed during solidification, in the metallic fuel slugs fabricated by modified injection casting. Scanning electron micrographs and energy-dispersive X-ray spectroscopy (SEM-EDS) showed that the Zr and RE inclusions were uniformly distributed in the matrix and the composition of the RE inclusions was similar to that of a charged RE element. The content and the size of the RE inclusions increased slightly according to the charge content of the RE elements. RE inclusions in U-Zr alloys will have a positive effect on fuel performance due to their micro-size and high degree of distribution.

• Journal of Environmental Science International
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• v.20 no.8
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• pp.929-942
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• 2011

A novel conductivity technique was developed to detect penetration depth of the vessel fluid into the feedpipe. For a given reactor geometry, critical agitator speeds were experimentally determined at the onset of feedpipe backmixing using Rushton 6 bladed disk turbine (6BD) and high efficiency axial flow type 3 bladed (HE-3) impellers. The ratio of the feedpipe velocity to the critical agitator speed ($v_f/v_t$) was constant for either laminar or turbulent feedpipe flow regimes. Compared to the results of fast competitive reaction, feedpipe backmixing had to penetrate at least one feedpipe diameter into the feedpipe to significantly influence the yield of the side product. However, higher $v_f/v_t$ than that for L/d = 0 (position at the feedpipe end) of the conductivity technique is recommended to completely eliminate feedpipe backmixing in conservative design criteria. The conductivity technique was successful in all feedpipe flow conditions of laminar, transitional and turbulent flow regimes.

• Nuclear Engineering and Technology
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• v.17 no.1
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• pp.16-24
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• 1985

A computer model is developed capable of simulating the transient behavior of a pool-type liquid metal-cooled fast breeder reactor (LMFBR). The model, SIMFARP, is a fast running computer code which may be used to simulate the loss of power to any pump(s), a complete loss-of-forced cooling, and the natural circulation behavior. Eight governing equations are derived and a Runge-Kutta algorithm is applied to integrate the eight differential equations. The developed computer program is applied to two cases; loss of electric power to any pump(s), and loss of all external electric supply power without scram in Super-Phenix-I.