• 한국신재생에너지학회:학술대회논문집
• /
• 2008.05a
• /
• pp.456-459
• /
• 2008

The 10t/d pyrolysis demonstration plant for waste tire recycling have been constructed and operated for commissioning of the plant. The plant have the tube reactor with chain conveyer attached disk. The reactor temperature is 500$\sim$600deg.C and pressure is -80$\sim$-100mmHg. Non-condensable gas is used as fuel for pyrolysis heat source.

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

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1775-1782
• /
• 2023

Several reactor physics commissioning experiments were conducted to obtain the neutronic parameters at the beginning of the G.A. Siwabessy Multi-purpose Reactor (RSG-GAS) operation. These parameters are essential for the reactor to safety operate. Leveraging the experimental data, this study evaluated the calculated core reactivity, control rod reactivity worth, integral control rod reactivity curve, and fuel reactivity. Calculations were carried out with Serpent 2 code using the latest neutron cross-section data ENDF/B-VIII.0. The criticality calculations were carried out for the RSG-GAS first core up to the third core configuration, which has been done experimentally during these commissioning periods. The excess reactivity for the second and third cores showed a difference of 510.97 pcm and 253.23 pcm to the experiment data. The calculated integral reactivity of the control rod has an error of less than 1.0% compared to the experimental data. The calculated fuel reactivity value is consistent with the measured data, with a maximum error of 2.12%. Therefore, it can be concluded that the RSG-GAS reactor core model is in good agreement to reproduce excess reactivity, control rod worth, and fuel element reactivity.

• Nuclear Engineering and Technology
• /
• v.50 no.1
• /
• pp.203-210
• /
• 2018

As the modernization of the nuclear instrumentation system progresses, research reactors have adopted digital wide-range neutron power measurement (DWRNPM) systems. These systems typically monitor the neutron flux across a range of over 10 decades. Because neutron detectors only measure the local neutron flux at their position, the local neutron flux must be converted to total reactor power through calibration, which involves mapping the local neutron flux level to a reference reactor power. Conventionally, the neutron power range is divided into smaller subranges because the neutron detector signal characteristics and the reference reactor power estimation methods are different for each subrange. Therefore, many factors should be considered when preparing the calibration procedure for DWRNPM channels. The main purpose of this work is to serve as a reference for performing the calibration of DWRNPM systems in research reactors. This work provides a comprehensive overview of the calibration of DWRNPM channels by describing the configuration of the DWRNPM system and by summarizing the theories of operation and the reference power estimation methods with their associated calibration procedure. The calibration procedure was actually performed during the commissioning of an open-pool type research reactor, and the results and experience are documented herein.

• Nuclear Engineering and Technology
• /
• v.40 no.1
• /
• pp.93-98
• /
• 2008

Several adhesive tapes have been studied in terms of their suitability for securing gold wires into positions for neutron flux measurements in the reactor core and irradiation facilities surrounding the core of the Open Pool Australian Light water (OPAL) reactor. Gamma ray spectrometry has been performed on each irradiated tape in order to identify and quantify activated components. Numerous metallic impurities have been identified in all tapes. Calculations relating to both the effective neutron shielding properties of the tapes and the error in measurement of the $^{198}Au$ activity caused by superfluous activity due to residual tape have been made. The most important identified effects were the prolonged cooling times required before safe enough levels of radioactivity to allow handling were reached, and extra activity caused by residual tape when measured with an ionisation chamber. Knowledge of the most suitable tape can allow a minimal contribution due to these effects, and the use of gamma spectrometry in preference to ionisation chamber measurements of the flux wires is shown to make all systematic errors due to the tape completely negligible.

• The KSFM Journal of Fluid Machinery
• /
• v.13 no.5
• /
• pp.5-10
• /
• 2010

5MW, open pool type research reactor, is commonly used to education and experimental purpose. It is necessary to prepare a standardization of system designs for considering a demand. HANARO has prepared the standardization of 5MW research reactor system designs based on the design, installation, commissioning and operating experiences of HANARO. For maintaining an open pool type reactor safety, a primary cooling system (after below, PCS) should remove the heat generated by the reactor under a reactor normal operation condition and a reactor shutdown condition. For removing the heat generated by the reactor, the PCS should maintain a required coolant flow rate. For a verification of the required flow rate, a flow network analysis of the PCS was carried under a normal operating condition. Based on the flow network analysis result, this paper describes the PCS flow characteristics of a 5MW open pool type research reactor. Through the result, it was confirmed that the PCS met design requirements including design flow rate without cavitation.

• 한국신재생에너지학회:학술대회논문집
• /
• 2008.10a
• /
• pp.138-141
• /
• 2008

The 120t/d pyrolysis commercial plant for waste tire recycling have been constructed in Malaysia and is going to be operated. The plant have the tube reactor with chain conveyer attached disk developed in demonstration research stage. The reactor temperature for commercial plant is about 500deg.C and reactor inside pressure is -100$\sim$-120mmHg. Non-condensable gas is used as fuel for pyrolysis heat source, and the exhausted heat is recovered for cogeneration to produce steam and electric power of 600kw.

• Nuclear Engineering and Technology
• /
• v.52 no.1
• /
• pp.14-18
• /
• 2020

Korea Atomic Energy Research Institute (KAERI) has carried out various neutronics experiments in the commissioning stage of the Jordan Research and Training Reactor (JRTR), and this paper introduces the results of first criticality prediction and experiment for the JRTR. The Monte Carlo Code for Advanced Reactor Design and analysis (McCARD) with the ENDF/B-VII.0 nuclear library was used for prediction calculations in the process of the first criticality approach, which was performed to provide reference for the first criticality experiment. In the experiment, fuel loading was carried out by measuring the inverse multiplication factor (1/M) to predict the number of fuel assemblies at the first criticality, and the first critical was reached on April 25, 2016. Comparing the first criticality prediction and experiment, the calculated and measured CAR (Control Absorber Rod) heights for the first criticality were 575 mm and 570.5 mm, respectively, that is, the difference between the two results was approximately 5 mm. From this result, it was confirmed that JRTR manufacturing and various experiments had successfully progressed as designed.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05a
• /
• pp.163-168
• /
• 1996

Simulations of Wolsong-1 Phase-B commissioning measurements have been performed, as part of the program to validate WIMS-AECL lattice cell calculations for application to CANDU reactor simulations in RFSP. A required component of these simulations is the calculation of incremental cross sections representing reactivity control devices in the reactor. The incremental cross section properties of the Wolsong-1 adjusters, Mechanical Control Absorbers (MCA) and liquid Zone Control Units (ZCU) are based on the WIMS-AECL/MULTICELL modelling methods and the results are compared with those of WIMS-AECL/DRAGON-2 modelling methods.

• Nuclear Engineering and Technology
• /
• v.55 no.10
• /
• pp.3648-3658
• /
• 2023

The early localization of a fuel subassembly with a failed (wet rupture) fuel pin is very important in reactors to limit the associated radiological and operational consequences. This requires a fast and reliable system for failure detection and their localization in the core. In the Prototype Fast Breeder Reactor, the system specially designed for this purpose is Failed Fuel Location Modules (FFLM) housed in the control plug region. It identifies a failed sub-assembly by detecting the presence of delayed neutrons in the sodium from a failed sub-assembly. During the commissioning phase of PFBR, it is mandatory to demonstrate the FFLM effectiveness. The paper highlights the engineering and physics design aspects of FFLM and the integrated simulation towards its function demonstration with a source assembly containing a perforated metallic fuel pin. This test pin mimics a MOX pin of 1 cm² of geometrical defect area. At 10% power and 20% sodium flow rate, the counts rate in the BCCs of FFLM system range from 75 cps to 145 cps depending upon the position of DN source assembly. The model developed for the counts simulation is applicable to both metal and MOX pins with proper values of k-factor and escape coefficient.