• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.177-198
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• 2012

This paper reviews the fast reactor physics and computational methods. The basic reactor physics specific to fast spectrum reactors are briefly reviewed, focused on fissile material breeding and actinide burning. Design implications and reactivity feedback characteristics are compared between breeder and burner reactors. Some discussions are given to the distinct nuclear characteristics of fast reactors that make the assumptions employed in traditional LWR analysis methods not applicable. Reactor physics analysis codes used for the modeling of fast reactor designs in the U.S. are reviewed. This review covers cross-section generation capabilities, whole-core deterministic (diffusion and transport) and Monte Carlo calculation tools, depletion and fuel cycle analysis codes, perturbation theory codes for reactivity coefficient calculation and cross section sensitivity analysis, and uncertainty analysis codes.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.54-59
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• 2019

The main objective of nuclear reactor safety is to maintain the nuclear fuel in a thermally safe condition with enough safety margins during normal operation and anticipated operational occurrences. In this research, core flow bypass is studied under the conditions of the unavailability of safety systems. As core bypass occurs, the core flow rate is assumed to decrease exponentially with a time constant of 25 s to new steady state values of 20, 40, 60, and 80% of the nominal core flow rate. The thermal hydraulic code PARET is used through these calculations. Reactor thermal hydraulic stability is reported for all cases of core flow bypass.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.97-106
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• 2018

Improved load-following capability is one of the most important technical tasks of a pressurized water reactor. Controlling the nuclear reactor core during load-following operation leads to some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking: the core is subjected to sharp and large variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent the core power peaking. One of the important local power peaking components in nuclear reactors is axial power peaking, which continuously changes. The main challenge of nuclear reactor control during load-following operation is to maintain the AO within acceptable limits, at a certain reference target value. This article proposes a new robust approach to AO control of pressurized water reactors during load-following operation. This method uses robust feedback-linearization control based on the multipoint kinetics reactor model (neutronic and thermal-hydraulic). In this model, the reactor core is divided into four nodes along the reactor axis. Simulation results show that this method improves the reactor load-following capability in the presence of parameter uncertainty and disturbances and can use optimum control rod groups to maneuver with variable overlapping.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.896-904
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• 2017

Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1051-1059
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• 2018

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for ${\sim}400{\mu}s$, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.57-66
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• 2000

The RELAP5 code has been developed for best-estimate simulation of transients and accidents for pressurized water reactors and their associated systems, but it has not been fully assessed for those of CANDU reactors. However, a previous study suggested that the RELAP5 code could be applicable to simulate the transients and accidents for CANDU reactors. Nevertheless, it is indicated that there are some works to be resolved, such as modeling of headers and multi-channel simulation for the reactor core, etc. Therefore, this study has been initiated with an aim to identify the code applicability for all the postulated transients and accidents in CANDU reactors. In the present study, the small inlet header break experiment (B8604) in the RD-14 test facility was simulated with RELAP5/MOD3.2 code. The RELAP5 results were also compared with both experimental data and those of CATHENA analyses performed by AECL and the analyses demonstrated the code's capability to predict major . phenomena occurring in the transient with sufficient accuracy for both Qualitative and quantitative viewpoint However, some discrepancies in the depressurization of the primary heat transport system after the break and the consequent time delay of the major phenomena were also observed.

• Nuclear Engineering and Technology
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• v.37 no.2
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• pp.127-138
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• 2005

Approaches to several recent issues in the operation of nuclear power plants using computational intelligence are discussed. These issues include 1) noise analysis techniques, 2) on-line monitoring and sensor validation, 3) regularization of ill-posed surveillance and diagnostic measurements, 4) transient identification, 5) artificial intelligence-based core monitoring and diagnostic system, 6) continuous efficiency improvement of nuclear power plants, and 7) autonomous anticipatory control and intelligent-agents. Several changes to the focus of Computational Intelligence in Nuclear Engineering have occurred in the past few years. With earlier activities focusing on the development of condition monitoring and diagnostic techniques for current nuclear power plants, recent activities have focused on the implementation of those methods and the development of methods for next generation plants and space reactors. These advanced techniques are expected to become increasingly important as current generation nuclear power plants have their licenses extended to 60 years and next generation reactors are being designed to operate for extended fuel cycles (up to 25 years), with less operator oversight, and especially for nuclear plants operating in severe environments such as space or ice-bound locations.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.193-198
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• 1996

In this study the RETRAN-03 code was modified to analyze the thermal-hydraulic transients under three-dimensional ship motions for the application to the future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations are successfully simulated at various conditions.

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.595-604
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• 2014

SILER (Seismic-Initiated event risk mitigation in LEad-cooled Reactors) is a Collaborative Project, partially funded by the European Commission in the $7^{th}$ Framework Programme, aimed at studying the risk associated to seismic-initiated events in Generation IV Heavy Liquid Metal reactors, and developing adequate protection measures. The project started in October 2011, and will run for a duration of three years. The attention of SILER is focused on the evaluation of the effects of earthquakes, with particular regards to beyond-design seismic events, and to the identification of mitigation strategies, acting both on structures and components design. Special efforts are devoted to the development of seismic isolation devices and related interface components. Two reference designs, at the state of development available at the beginning of the project and coming from the $6^{th}$ Framework Programme, have been considered: ELSY (European Lead Fast Reactor) for the Lead Fast Reactors (LFR), and MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) for the Accelerator-Driven Systems (ADS). This paper describes the main activities and results obtained so far, paying particular attention to the development of seismic isolators, and the interface components which must be installed between the isolated reactor building and the non-isolated parts of the plant, such as the pipe expansion joints and the joint-cover of the seismic gap.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.1-17
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• 2018

This article reviews the multiscale nature of stress corrosion cracking (SCC) observed by high-resolution characterizations in austenite stainless steels and Ni-base superalloys in light water reactors (including boiling water reactors, pressurized water reactors, and supercritical water reactors) with related opinions. A new statistical summary and comparison of observed degradation phenomena at different length scales is included. The intrinsic causes of this multiscale nature of SCC are discussed based on existing evidence and related opinions, ranging from materials theory to practical processing technologies. Questions of interest are then discussed to improve bottom-up understanding of the intrinsic causes. Last, a multiscale modeling and simulation methodology is proposed as a promising interdisciplinary solution to understand the intrinsic causes of the multiscale nature of SCC in light water reactors, based on a review of related supporting application evidence.