• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2207-2220
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• 2021

Lead alloy is used as coolant in Lead-based cooled Fast Reactor (LFR). The natural characteristics of lead alloy are combined with the simple structural design of LFR. This constitutes the inherent safety characteristics of LFR. The main work of this paper is to take the main coolant pump (MCP) in the lead-cooled fast reactor (LFR) as the research object, and to study the flow pattern distribution of the internal flow field under the reverse rotation pump condition, the reverse rotation positive-flow braking condition and the reverse rotation negative-flow braking condition. In this paper, the double-outlet volute type and the space guide vane are selected as the potential designs of the CLEAR-I MCP. In this paper, the CFD method is used to study the potential reverse accident of the MCP. It is found that the highest flow velocity in the impeller appears at the impeller outlet, and the Q-H curves of the two design programs basically coincide. The space guide vane type MCP has better hydraulic performance under the reverse rotation positive-flow condition, the Q-H curves of the two designs gradually separate with increasing flow rate, and the maximum flow velocity inside the space guide vane type MCP is obviously lower than that of the double-outlet volute type. For the reverse rotation test of MCP, only the condition of the forward rotating pump of the main coolant pump is tested and verified. For the simulation of the MCP in LBE medium, it proved that the turbulence model and basic settings selected in the simulation are reliable.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3085-3094
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• 2022

The design concept of a Small High-flux Multipurpose LBE(Lead Bismuth Eutectic) cooled Fast Reactor (SHMLFR) was proposed in the paper. The primary cooling system of the reactor is forced circulation, and the fuel element form is arc-plate loaded high enrichment MOX fuel. The core is cylindrical with a flux trap set in the center of the core, which can be used as an irradiation channel. According to the requirements of the core physical design, a series of physical design criteria and constraints were given, and the steady and transient parameters of the reactor were calculated and analyzed. Regarding the thermal and hydraulic phenomena of the reactor, a simplified model was used to conduct a preliminary analysis of the fuel plates at special positions, and the temperature field distribution of the fuel plate with the highest power density under different coolant flow rates was simulated. The results show that the various parameters of SHMLFR meet the requirements and design criteria of the physical design of the core and the thermal design of the reactor. This implies that the conceptual design of SHMLFR is feasible.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1272-1278
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• 2019

A lumped kinetic modeling platform is developed to investigate the coupled nuclear/thermo-fluid features of the closed natural circulation loop in a low power lead cooled fast reactor. This coolant material serves a reliable choice with noticeable thermo-physical safety characteristics in terms of natural convection. Boussienesq approximation is resorted to appropriately reduce the governing partial differential equations (PDEs) for the fluid flow into a set of ordinary differential equations (ODEs). As a main contributing step, the coolant circulation speed is accordingly correlated to the loop operational power and temperature levels. Further temporal analysis and control synthesis activities may thus be carried out within a more consistent state space framework. Nyquist stability criterion is thereafter employed to carry out a sensitivity analysis for the system stability at various power and heat sink temperature levels and results confirm a widely stable natural circulation loop.

• Nuclear Engineering and Technology
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• 제38권2호
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• pp.109-118
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• 2006

Various indirect power cycle options for a helium cooled gas cooled fast reactor (GFR) with particular focus on a supercritical $CO_2(SCO_2)$ indirect cycle are investigated as an alternative to a helium cooled direct cycle GFR. The balance of plant (BOP) options include helium-nitrogen Brayton cycle, supercritical water Rankine cycle, and $SCO_2$ recompression Brayton power cycle in three versions: (1) basic design with turbine inlet temperature of $550^{\circ}C$, (2) advanced design with turbine inlet temperature of $650^{\circ}C$ and (3) advanced design with the same turbine inlet temperature and reduced compressor inlet temperature. The indirect $SCO_2$ recompression cycle is found attractive since in addition to easier BOP maintenance it allows significant reduction of core outlet temperature, making design of the primary system easier while achieving very attractive efficiencies comparable to or slightly lower than, the efficiency of the reference GFR direct cycle design. In addition, the indirect cycle arrangement allows significant reduction of the GFR &proximate-containment& and the BOP for the $SCO_2$ cycle is very compact. Both these factors will lead to reduced capital cost.

• Nuclear Engineering and Technology
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• 제55권6호
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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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• 제56권4호
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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).

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1834-1845
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• 2021

Liquid lead-bismuth cooled fast reactor is one of the most promising reactor types among the fourth-generation nuclear energy systems. The flow and heat transfer characteristics of lead-bismuth eutectic (LBE) are completely different from ordinary fluids due to its special thermal properties, causing that the traditional Reynolds analogy is no longer recommended and appropriate. More accurate turbulence flow and heat transfer model for the liquid metal lead-bismuth should be developed and applied in CFD simulation. In this paper, a specific CFD solver for simulating the flow and heat transfer of liquid lead-bismuth based on the k - 𝜀 - k_𝜃 - 𝜀_𝜃 model was developed based on the open source platform OpenFOAM. Then the advantage of proposed model was demonstrated and validated against a set of experimental data. Finally, the simulation of LBE turbulent flow and heat transfer in a 7-pin wire-wrapped rod bundle with the k - 𝜀 - k_𝜃 - 𝜀_𝜃 model was carried out. The influence of wire on the flow and heat transfer characteristics and the three-dimensional distribution of key thermal hydraulic parameters such as temperature, cross-flow velocity and Nusselt number were studied and presented. Compared with the traditional SED model with a constant Pr_t = 1.5 or 2.0, the k - 𝜀 - k_𝜃 - 𝜀_𝜃 model is more accurate on predicting the turbulence flow and heat transfer of liquid lead-bismuth. The average relative error of the k - 𝜀 - k_𝜃 - 𝜀_𝜃 model is reduced by 11.1% at most under the simulation conditions in this paper. This work is meaningful for the thermal hydraulic analysis and structure design of fuel assembly in the liquid lead-bismuth cooled fast reactor.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1460-1471
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• 2024

2D and 3D ultrasonic imaging has so far not been demonstrated in pure molten lead in the open literature. In this study the development of such an ultrasonic device for imaging is outlined and results from testing at 380 ℃ in lead are presented. The main difficulties were found to be achieving then maintaining suitable wetting while ensuring suitable durability of the device, both due to the harsh nature of molten lead and the elevated temperatures. The successful detection and imaging (2D and 3D), of differently shaped targets, where the features were above the size of the transmitted ultrasound beam was demonstrated.

• Nuclear Engineering and Technology
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• 제38권5호
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• pp.411-416
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• 2006

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the comer 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.323-334
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• 2013

During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.