• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1479-1482
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• 2021

Research reactors are typically well-suited for outreach activities at different levels. However, unplanned seeking to increase the utilization of a research reactor may result in weakening the nuclear security of this facility. Research reactor staff might be in shortage of a functional nuclear security culture; specifically, there might be a conviction that the necessities of research can be given the priority over consistence with security procedural requirements. Research reactors are usually parts of bigger institutes or research labs of different activities. Moreover, the employments of research reactors are usually with the purpose that easy entry to the reactor premises is fundamental. So, they could be co-situated in places with different sorts of activities, mostly under similar security arrangements. The co-area of research reactor offices among different kinds of research labs introduces explicit security issues, the effects of which should be viewed as when building up a nuclear security framework. Notwithstanding potential security vulnerabilities presented in the design, research reactors frequently have devices kept promptly accessible to encourage research and education. The accessibility of these sorts of hardware could be used by an authorized person to commit an unapproved activity or cause harm. This paper aims to present insights to compromise contradicting conditions in new research reactors in which both enhancing utilization and ensuring security are satisfied.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2047-2052
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• 2023

Risk-informed approach has been widely applied in the safety design, regulation, and operation of nuclear reactors. It has been commonly accepted that risk-informed design optimization should be used in the innovative reactor designs to make nuclear system highly safe and reliable. In spite of the risk-informed approach has been used in some advanced nuclear reactors designs, such as Westinghouse IRIS, Gen-IV sodium fast reactors and lead-based fast reactors, the process of risk-informed design of nuclear reactors is hardly to carry out when passive system reliability should be integrated in the framework. A practical method for new passive safety reactors based on probabilistic safety assessment (PSA) and passive system reliability analyze linking is proposed in this paper. New three-dimension frequency-consequence curve based on risk concept with three variables is used in this method. The proposed method has been applied to the determination optimization of design options selection in a 10 MWth lead-based research reactor(LR) to obtain one optimized system design in conceptual design stage, using the integrated reliability and probabilistic safety assessment program RiskA, and the computation resources and time consumption in this process was demonstrated reasonable and acceptable.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.259-270
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• 2010

New generation nuclear reactors are designed using advanced safety analysis methods. A thorough understanding of different interacting physical phenomena is necessary to avoid underestimation and overestimation of consequences of off-normal transients in the reactor safety analysis results. This feature requires a multiphysics reactor simulation model. In this context, a coupled dynamics model based on a multiphysics formulation is developed indigenously for the transient analysis of large pressurized VVER reactors. Major simplifications are employed in the model by making several assumptions based on the physics of individual phenomenon. Space and time grids are optimized to minimize the computational bulk. The capability of the model is demonstrated by solving a series of international (AER) benchmark problems for VVER reactors. The developed model was used to analyze a number of reactivity transients that are likely to occur in VVER reactors.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1526-1533
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• 2022

Small modular reactors have garnered considerable attention in the recent years. Plate fuel elements exhibit a good application prospect in small modular pressurized water reactors for marine applications. Further, improved economic benefits can be achieved by extending the core lifetime of small modular reactors. However, it is necessary to realize a large initial residual reactivity for achieving a relatively long burnup depth finally. Thus, the selection of a suitable burnable poison (BP) is a crucial factor that should be considered in the design of small modular reactors. In this study, some candidate BPs are selected to realize the effective control of reactivity. The results show that ²³¹Pa₂O₃, ²⁴⁰Pu₂O₃, ¹⁶⁷Er₂O₃, PACS-J, and PACS-L are ideal candidates of BP, and since the characteristics of BP can increase the final burnup depth of assembly, the economic benefits are gained. Additionally, an optimal combination scheme of BPs is established. Specifically, it is proved that through a reasonable combination of BPs, a low reactivity fluctuation during the lifetime can be achieved, leading to a large final burnup depth.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.249-256
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• 2007

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modem fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with $380^{\circ}C$ core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around $500^{\circ}C$, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors.

• Carbon letters
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• v.4 no.1
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• pp.36-39
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• 2003

Carbon and carbon-based materials are used in nuclear reactors and there has recently been growing interest to develop graphite and carbon based materials for high temperature nuclear and fusion reactors. Efforts are underway to develop high density carbon materials as well as amorphous isotropic carbon for the application in thermal reactors. There has been research on coated nuclear fuel for high temperature reactor and research and development on coated fuels are now focused on fuel particles with high endurance during normal lifetime of the reactor. Since graphite as a moderator as well as structural material in high temperature reactors is one of the most favored choices, it is now felt to develop high density isotropic graphite with suitable coating for safe application of carbon based materials even in oxidizing or water vapor environment. Carboncarbon composite materials compared to conventional graphite materials are now being looked into as the promising materials for the fusion reactor due their ability to have high thermal conductivity and high thermal shock resistance. This paper deals with the application of carbon materials on various nuclear reactors related issues and addresses the current need for focused research on novel carbon materials for future new generation nuclear reactors.

• Nuclear Engineering and Technology
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• v.11 no.1
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• pp.29-45
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• 1979

In this paper. nuclear characteristics of TRIGA Mark-III has been analyzed in detail for six different fuel options. Presently, 70 w/o enriched FLIP fuels are adopted for TRIGA core to improve fuel lifetime. However, such highly enriched fuels are not easily obtained due to nonproliferation treaty. This research examines the possible substitution for FLIP fuels with high density fuels without reducing the nuclear performance. This work will provide long-time plan for TRIGA operation.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1120-1125
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• 2016

The neutron decay constant, ${\alpha}$, and effective delayed neutron fraction, ${\beta}_{eff}$, are important parameters for the control of the dynamic behavior of nuclear reactors. For the heavy water zero power reactor (HWZPR), this document describes the measurements of the neutron decay constant by noise analysis methods, including variance to mean (VTM) ratio and endogenous pulse source (EPS) methods. The measured ${\alpha}$ is successively used to determine the experimental value of the effective delayed neutron fraction as well. According to the experimental results, ${\beta}_{eff}$ of the HWZPR reactor under study is equal to 7.84e-3. This value is finally used to validate the calculation of the effective delayed neutron fraction by the Monte Carlo methods that are discussed in the document. Using the Monte Carlo N-Particle (MCNP)-4C code, a ${\beta}_{eff}$ value of 7.58e-3 was obtained for the reactor under study. Thus, the relative difference between the ${\beta}_{eff}$ values determined experimentally and by Monte Carlo methods was estimated to be < 4%.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.148-156
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• 2015

Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs) that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C) strategies for a large 1,000 MWe iPWR is described. Reactor system modeling-which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum-is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.47-58
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• 2019

The paper presents the results of a multi-criteria comparative evaluation of 12 feasible Russian nuclear energy deployment scenarios with thermal and fast reactors in a closed nuclear fuel cycle. The comparative evaluation was performed based on 6 performance indicators and 5 different MCDA methods (Simple Scoring Model, MAVT / MAUT, AHP, TOPSIS, PROMETHEE) in accordance with the recommendations elaborated by the IAEA/INPRO section. It is shown that the use of different MCDA methods to compare the nuclear energy deployment scenarios, despite some differences in the rankings, leads to well-coordinated and similar results. Taking into account the uncertainties in the weights within a multi-attribute model, it was possible to rank the scenarios in the absence of information regarding the relative importance of performance indicators and determine the preference probability for a certain nuclear energy deployment scenario. Based on the results of the uncertainty/sensitivity analysis and additional analysis of alternatives as well as the whole set of graphical and attribute data, it was possible to identify the most promising nuclear energy deployment scenario under the assumptions made.