• Proceedings of the Korean Society of Precision Engineering Conference
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• 2010.05a
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• pp.1155-1156
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• 2010

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2477-2487
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• 2021

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3580-3596
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• 2021

An integral test facility, PILLAR, was commissioned, aiming to provide valuable experimental results which can be referenced by system and component designers and used for the performance demonstration of liquid-metal-cooled, passive small modular reactors (SMRs) toward their licensing. The setup was conceptualized by a scaling analysis which allows the vertical arrangements to be conserved from its prototypic reactor, scaled uniformly in the radial direction achieving a flow area reduction of 1/200. Its final design includes several heater rods which simulate the reactor core, and a single heat exchanger representing the steam generators in the prototype. The system behaviors were characterized by its data acquisition system implementing various instruments. In this paper, we present not only a detailed description of the facility components, but also selected experimental results of both steady-state and transient cases. The obtained steady-state test results were utilized for the benchmark of a system code, achieving a capability of accurate simulations with ±3% of maximum deviations. It was followed by qualitative comparisons on the transient test results which indicate that the integral system behaviors in passive LBE-cooled systems are able to be predicted by the code.

• International Journal of Safety
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• v.3 no.1
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• pp.53-58
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• 2004

SMART-P is a promising advanced small and medium category nuclear power reactor. It is an integral type reactor with a sensible mixture of new innovative design features and proven technologies aimed at achieving a highly enhanced safety and improved economics. The enhancement of the safety and reliability is realized by incorporating inherent safety improving features and reliable passive safety systems. The improvement in the economics is achieved through a system simplification, and component modularization. Preliminary safety analyses on selected limiting accidents confirm that the inherent safety improving design characteristics and the safety system of SMART-P ensure the reactor's safety. SMART-P is an advanced integral pressurized water reactor. The purpose of this study is for the safety analysis of the steam generator module pipe break for the SMART-P. The integrity of the fuel rod is the major criteria of this analysis. As a result of this analysis, the safety of the RCS and the secondary system is guaranteed against the module pipe break of a steam generator of the SMART-P.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.20 no.7
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• pp.658-664
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• 2010

A control element drive mechanism(CEDM) is a reactor regulating system, which inserts, withdraws or maintains a control rod containing a neutron absorbing material within a reactor core to control the reactivity of the core. The ball-screw type CEDM for the integral reactor has a spring-damper system to reduce the impact force due to the scram of the CEDM. This paper describes the experimental results to obtain the drop and damping characteristics of the CEDM. The drop tests are performed by using a drop test rig and a facility. A drop time and a displacement after an impact are measured using a LVDT. The influences of the rod weight, the drop height and the flow area of hydraulic damper on the drop and damping behavior are also estimated on the basis of test results. The drop time of the control element is within 4.5s to meet the design requirement, and the maximum displacement is measured as 15.6 mm. It is also found that the damping system using a spring-hydraulic damper plays a good damper role in the CEDM.

• 유체기계공업학회:학술대회논문집
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• 2003.12a
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• pp.161-166
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• 2003

It is the common features of the integral reactors that the main components of the RCS are installed within the reactor vessel, and so there are no any flow pipes connecting the coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the RCS of the integral reactors, and it also makes impossible measure the flow-rate of the reactor coolant. As a alternative method, the method by the measurement of coolant pump power has been introduced in this study. Up to now, we did not found out a precedent which the coolant pump power is used for the real-time flow-rate calculation at normal operation of the commercial nuclear power plants. The objective of the study is to embody the real-time flow-rate calculation method by the measurement of coolant pump power in an integral reactor. As a result of the study, we could theoretically reason that the capacity-head curve and capacity-shaft power curve around the rated capacity with the high specific-speeded axial flow pumps have each diagonally steep incline but show the similar shape. Also, we could confirm the above theoretical reasoning from the measured result of the pump motor inputs, So, it has been concluded that it is possible to calculate the real-time flow-rate by the measurement of pump motor inputs. In addition, the compensation for a above new method can be made by HBM being now used in the commercial nuclear power plants.

• Nuclear Engineering and Technology
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• v.33 no.1
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• pp.111-120
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• 2001

In this study, a dynamic analysis of the integral reactor SMART (System-integrated Modular Advanced ReacTor) under postulated seismic events is performed to review the response characteristics of the major components. To enhance the feasibility of an analysis model, a detailed finite element model is synchronized with the products of concurrent design activities. The artificial time history, which has been applied to the seismic analysis for the Korean Standard Nuclear Power Plant (KSNP), is chosen to envelop broad site specifics in Korea. Responses in the horizontal direction are found slightly amplified, while those in the vertical direction are suppressed. Since amplified response is monitored at the control element drive mechanism (CEDM), minor design provision is considered to enhance the integrity of the subsystem.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.877-885
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• 2018

Various controllers such as proportional-integral-derivative (PID) controllers have been designed and optimized for load-following issues in nuclear reactors. To achieve high performance, gain tuning is of great importance in PID controllers. In this work, gains of a PID controller are optimized for power-level control of a typical pressurized water reactor using particle swarm optimization (PSO) algorithm. The point kinetic is used as a reactor power model. In PSO, the objective (cost) function defined by decision variables including overshoot, settling time, and stabilization time (stability condition) must be minimized (optimized). Stability condition is guaranteed by Lyapunov synthesis. The simulation results demonstrated good stability and high performance of the closed-loop PSO-PID controller to response power demand.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.5 no.3
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• pp.227-231
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• 2004

It is the common features of the integral reactors that the main components of the RCS are installed within the reactor vessel, and so there are no any flow pipes connecting the steam generator or the pump whose type is the axial flow. Due to no any flow pipes, it is impossible to measure the differential pressure at the RCS of the integral reactors, and it also makes impossible measure the flow-rate of the reactor coolant. As a alternative method, the method by the measurement of the pump power of the axial flow pump has been introduced in this study. Up to now, we did not found out a precedent which the pump power is used for the flow-rate calculation at normal operation of the commercial nuclear power plants. The objective of the study is to embody the flow-rate calculation method by the measurement of the pump power in an integral reactor. As a result of the study, we could theoretically reason that the capacity-head curve and capacity-shaft power curve around the rated capacity with the high specific-speeded axial flow pumps have each diagonally steep incline but show the similar shape. Also, we could confirm the above theoretical reasoning from the measured result of the pump motor inputs. So, it has been concluded that it is possible to calculate the flow-rate by the measurement of the pump motor inputs.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2395-2407
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• 2022

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their "pre-INSPYRE" versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the "post-INSPYRE" code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.