• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.887-895
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• 2017

Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR) system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits "self-regulating" behavior, minimizing the need for external controller action for load-following maneuvers.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.504-516
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• 2017

The Jordan Research and Training Reactor (JRTR) is the first research reactor in Jordan, the commissioning of which is ongoing. The reactor is a 5-MWth, open-pool type, light-water-moderated, and cooled reactor with a heavy water reflector system. The neutron measurement system (NMS) applied to the JRTR employs a wide-range fission chamber that can cover from source range to power range. A high-sensitivity boron trifluoride counter was added to obtain more accurate measurements of the neutron signals and to calibrate the log power signals; the NMS has a major role in the entire commissioning stage. However, few case studies exist concerning the application of the NMS to a research reactor. This study introduces the features of the NMS and the boron trifluoride counter in the JRTR and shares valuable experiences from lessons learned from the system installation to its early commissioning. In particular, the background noise relative to the signal-to-noise ratio and the NMS signal interlock are elaborated. The results of the count rates with the neutron source and the effects of the discriminator threshold are summarized.

• 유체기계공업학회:학술대회논문집
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• 2005.12a
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• pp.503-508
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• 2005

An advanced Research Reactor (ARR) consists of an open-tank-type reactor assembly within a light water pool and generates thermal power of 20 MW. The thermal power is including a fission heat in the core, a fuel generated heat temporary stored in the pool, a circulating pumps generated heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the primary cooling system will be installed. In this study, the conceptual design of the primary cooling system has been carried out using a design methodology of HANARO within a permissible range of safety. As results, it has been established that the conceptual design of the primary cooling system including design requirements, performance requirements, design restrictions, system descriptions and system operation to maintain the system functions.

• Transactions of the KSME C: Technology and Education
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• v.5 no.2
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• pp.135-143
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• 2017

The domestic unique research reactor, HANARO (Hi-flux Advanced Neutron Application ReactOr), has been constructed with the open-pool, the core is submerged in, for the multi-purpose neutron application. The reactor has a primary cooling system to remove the fission heat from the core and its connected fluidic systems. Since the works are required at the reactor pool top as a characteristic of the research reactor, the radiation shall be minimized with the operation of the hot water layer system to avoid unnecessary radiation exposure on the workers during work at the pool top. Moreover, the pool water management system is connected to the reactor pool to maintain the pool temperature below $50^{\circ}C$ to minimize the uprising radioactive gas or impurity from the colder pool bottom. For the efficient flow rate of the PWMS, the thermal capacity of heat exchanger is selected with 260 kW in the normal operation condition. In this paper, the modeling is formulated to figure out whether or not each pool temperature maintains below the temperature limit and the calculation results show that the designed PWMS heat exchanger has enough capacity with the design margin regardless of the reactor operation mode.

• 한국전산유체공학회:학술대회논문집
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• 2004.03a
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• pp.214-219
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• 2004

The HANARO, a multi-purpose research reactor of 30 MWth, open-tank-in-pool type, has been under normal operation since its initial criticality in February, 1995. The HANARO is composed of inlet plenum, grid plate, core channel with flow tubes and chimney. The reactor core channel is located at about twelve m (12 m) depth of the reactor pool and cold by the upward flow that the coolant enters the lower inlet of the plenum, rises up through the grid plate and the core channel and exit through the outlet of chimney. A guide tube is extended from the reactor core to the top of the reactor chimney for easily un/loading a target under the reactor normal operation. But active coolant through the core can be Quickly raised up to the top of the chimney through the guide tube by jet flow. This paper is described an analytical analysis to study the flow behavior through the guide tube under reactor normal operation and unloading the target. As results, it was conformed through the analysis results that the flow rate, about fourteen kilogram per second (14 kg/s) suppressed the guide tube jet and met the design cooling flow rate in a circular flow tube, and that the fission moly target cooling flow rate met the minimum flow rate to cool the target.

• Journal of computational fluids engineering
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• v.10 no.2
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• pp.1-6
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• 2005

The HANARO, a multi-purpose research reactor of 30 MWth, open-tank-in-pool type, has been under normal operation since its initial criticality in February, 1995. The HANARO is composed af inlet plenum, grid plate, core channel with flow tubes and chimney. The reactor core channel is located at about twelve meters (12 m) depth of the reactor pool and cooled by the upward flow that the coolant enters the lower inlet of the plenum, rises up through the grid plate and the core channel and comes out from the outlet of chimney. A fission moly guide tube is extended from the reactor core to the top of the reactor chimney for easily loading a fission moly target under the reactor normal operation. But active coolant through the core can be quickly raised up to the top of the chimney through the guide tube by jet flow. This paper describes an analytical analysis that is the study of the flow behavior through the guide tube under reactor normal operation and unloading the target. As results, it was conformed through the analysis results that the flow rate, reduced to about fourteen kilogram per second (14 kg/s) from the original flow rate of sixteen point three kilogram per second (16.3 kg/s) did not show the guide tube jet.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.1835-1840
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• 2004

The HANARO, a multi-purpose research reactor of 30 MWth, open-tank-in-pool type, has been under normal operation since its initial criticality in February, 1995. The HANARO is composed of inlet plenum, grid plate, core channel with flow tubes and chimney. The reactor core channel is located at about twelve meters (12 m) depth of the reactor pool and cooled by the upward flow that the coolant enters the lower inlet of the plenum,. rises up through the grid plate and the core channel and comes out from the outlet of chimney. A guide tube is extended from the reactor core to the top of the reactor chimney for easily un/loading a target under the reactor normal operation. But active coolant through the core can be quickly raised up to the top of the chimney through the guide tube by a jet flow. This paper describes an analytical analysis that is the study of the flow behavior through the guide tube under reactor normal operation and unloading the target. As results, it was conformed through the analysis results that the guide jet is suppressed under the top of the chimney after modifying the orifice diameter of 37.5 mm to 31 mm.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.811-816
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• 2022

This work presents the methodology used to expand the capabilities of the Monte Carlo code OpenMC for the calculation of reactor kinetic parameters: effective delayed neutron fraction β_eff and neutron generation time L. The modified code, OpenMC(Time-Dependent) or OpenMC(TD), was then used to calculate the effective delayed neutron fraction by using the prompt method, while the neutron generation time was estimated using the pulsed method, fitting Λ to the decay of the neutron population. OpenMC(TD) is intended to serve as an alternative for the estimation of kinetic parameters when licensed codes are not available. The results obtained are compared to experimental data and MCNP calculated values for 18 benchmark configurations.

• Nuclear Engineering and Technology
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• v.42 no.6
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• pp.608-619
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• 2010

A multi-scale analysis of water-cooled reactor thermalhydraulics can be used to take advantage of increased computer power and improved simulation tools, including Direct Numerical Simulation (DNS), Computational Fluid Dynamics (CFD) (in both open and porous mediums), and system thermalhydraulic codes. This paper presents a general strategy for this procedure for various thermalhydraulic scales. A short state of the art is given for each scale, and the role of the scale in the overall multi-scale analysis process is defined. System thermalhydraulic codes will remain a privileged tool for many investigations related to safety. CFD in porous medium is already being frequently used for core thermalhydraulics, either in 3D modules of system codes or in component codes. CFD in open medium allows zooming on some reactor components in specific situations, and may be coupled to the system and component scales. Various modeling approaches exist in the domain from DNS to CFD which may be used to improve the understanding of flow processes, and as a basis for developing more physically based models for macroscopic tools. A few examples are given to illustrate the multi-scale approach. Perspectives for the future are drawn from the present state of the art and directions for future research and development are given.

• Proceedings of the KIEE Conference
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• 1997.07b
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• pp.600-602
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• 1997

The kalman filter which has good estimating capabilities by means of the recursive computation from the previously known of obtained data is usually used for the system estimation in the case of not being directly measurable. The best estimating technique is still open issues on the PWR reactor control system to increase operating contingencies and to predict the safety margins for safer reactor operation. This paper addressed its estimating technique using kalman filter for the more flexible reactor control and showed the reasonable approach for discretization of the continuos-time system for reduction of computation errors.