• Nuclear Engineering and Technology
• /
• v.14 no.4
• /
• pp.163-171
• /
• 1982

An ABBN-type 26-group constant set, KAERI-26G, which can be reliably applicable to fast reactor calculations has been generated using the nuclear data of ENDF/B-IV or ENDL-78 and a processing code ETOX-K4. The KAERI-26G set was evaluated by analysing measured integral quantities such as effective multiplication factor, central reaction-rate ratio, and central reactivity coefficient for a variety of critical assemblies. All these calculated quantities were compared with results from other workers using similar-type sets.

• 한국전산유체공학회:학술대회논문집
• /
• 2001.10a
• /
• pp.118-125
• /
• 2001

The performance prediction of SMART MCP was performed using a computational fluid dynamics code. General capacity-head performance curve of MCP, which is provided to other design branches as design input, was obtained and it showed the typical type of axial pump performance curve. When four MCPs operate in parallel and one of them stops while the others continue to operate, SMART requires reduced power operation. A procedure for predicting the performance of SMART MCP for that case was developed and verified with available experimental data. An analysis based on the developed procedure was performed for two cases; the impeller of sloped MCP is fixed or free to rotate in reverse direction. According to the results, $73\%$ flow rate of normal operation enters the reactor core in the case of the locked impeller. In case of the impeller free rotation, the flow rate entering the reactor core is $62.8\%$.

• Nuclear Engineering and Technology
• /
• v.44 no.5
• /
• pp.501-506
• /
• 2012

Many innovative design features are employed in the reactor vessel internals of SMART, a small integral-type pressurized water reactor, one of which is the flow skirt, which uniformly distributes flow and horizontally restrains the lower part of the core support barrel. This new design requires a comprehensive investigation of vibration characteristics. Therefore, in this study, modal characteristics of flow skirts are investigated with finite element analysis. Specifically, we investigate how the presence of holes, the presence of three rings attached to the flow skirt, and the thickness of the lowest shell effect vibration characteristics. In addition, the fluid effect is addressed, since the flow skirt is submerged in the fluid.

• Nuclear Engineering and Technology
• /
• v.43 no.2
• /
• pp.167-174
• /
• 2011

A pressurizer in a small integral type pressurized water reactor is located inside the upper region of the reactor vessel, and uses a space between the upper head of the reactor vessel and the upper region of the upper guide structure which is partially filled with fluid depending on the operating power. This new design requires a comprehensive investigation of vibration characteristics. This study investigates the modal characteristics of a pressurizer which uses a simplified cylindrical shell model, focusing on how having fluid in the shell affects vibration and response characteristics. In addition, an analysis of sloshing is performed and the response characteristics are addressed.

• Nuclear Engineering and Technology
• /
• v.45 no.3
• /
• pp.347-360
• /
• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• Nuclear Engineering and Technology
• /
• v.46 no.4
• /
• pp.513-520
• /
• 2014

In this paper, three types of operational and industrial absorbers used at research reactors, including Ag-In-Cd alloy, $B_4C$, and Hf are selected for sensitivity analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, thermal neutron flux, power density distribution, and Power Peaking Factor (PPF). The IAEA 10 MW benchmark core is selected as the case study to verify calculations. A two-dimensional, three-group diffusion model is selected for core calculations. The well-known WIMS-D4 and CITATION reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the $B_4C$; also the lowest PPF is gained using the Ag-In-Cd alloy. The maximum point power densities belong to the inside fuel regions surrounding the central flux trap (irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the graphite reflectors. The greatest and least fluctuation of the point power densities are gained by using $B_4C$ and Ag-In-Cd alloy, respectively.

• Proceedings of the KSME Conference
• /
• 2004.04a
• /
• pp.72-77
• /
• 2004

In a primary reactor cooling system(RCS), a dissimilar weld zone exists between cast stainless steel(CF8M) in a pipe and low-alloy steel(SA508 cl.3) in a nozzle. Thermal aging is observed in CF8M as the RCS is exposed for a long period of time to a reactor operating temperature between 290 and $330^{\circ}C$, while no effect is observed in SA508 cl.3. The specimens are prepared by an artificially accelerated aging technique maintained for 300, 1800 and 3600 hrs at $430^{\circ}C$, respectively. The specimens for elastic-plastic fracture toughness tests are prepared one type, which notch is created in the heat affected zone(HAZ) of CF8M. And, the specimens for fatigue crack growth tests are prepared in three classes, which notches are created at the center of deposited zone, the HAZ of CF8M, and the HAZ of SA508 cl.3. From the experiments, the J-integral values with the increase of aging time decrease, and the differences of the fatigue crack growth behaviors are relatively small in the three classes specimens.

• Nuclear Engineering and Technology
• /
• v.55 no.3
• /
• pp.1125-1139
• /
• 2023

The system-integrated modular advanced reactor 100 (SMART100), an integral-type pressurized water small modular reactor, is based on a novel design concept for containment cooling and radioactive material reduction; it is known as the containment pressure and radioactivity suppression system (CPRSS). There is a passive cooling system using a condensation with non-condensable gas in the SMART CPRSS. When a design basis accident such as a small break loss of coolant accident (SBLOCA) occurs, the pressurized low containment area (LCA) of the SMART CPRSS leads to steam condensation in an incontainment refuelling water storage tank (IRWST). Additionally, the steam and non-condensable gas mixture passes through the CPRSS heat exchanger (CHX) submerged in the emergency cooldown tank (ECT) that can partially remove the residual heat. When the steam and non-condensable gas mixture passes through the CHX, the non-condensable gas can interrupt the condensation heat transfer in the CHX and it degrades CHX performance. In this study, condensation heat transfer experiments of steam and non-condensable gas mixture in the natural circulation loop were conducted. The pressure, temperature, and effects of the non-condensable gas were investigated according to the constant inlet steam flow rate with non-condensable gas injections in the loop.

• Tribology and Lubricants
• /
• v.21 no.6
• /
• pp.278-282
• /
• 2005

Deep groove ball bearing is installed in a control element of an integral nuclear reactor, where water is used as coolant and lubricant. This bearing is made of STS440C stainless steel for the raceways and the balls to use in radioactive environment and water. It is known that the retainer design affects ball bearing operability and endurance life, however there is no verified retainer design and material for water lubricated ball bearing. Four kinds of retainers are manufactured for the endurance test of water lubricated deep groove ball bearing. Three of them are commercially developed types and the other is designed for this research. It is verified that ball bearings with steel pressed and general plastic retainer can not survive to required life in the water, however bearings with machined type and cylinder type retainer can survive. This proves that one of the major design parameters for water lubricated ball bearing is retainer type and material. In this paper, experimental research of endurance test for water-lubricated ball bearing are reported.

• Nuclear Engineering and Technology
• /
• v.52 no.8
• /
• pp.1834-1846
• /
• 2020

The SMART (System-integrated Modular Advanced ReacTor) is an integral-type small modular reactor developed by KAERI (Korea Atomic Energy Research Institute). This paper discusses the development of a dynamic equivalent model of the SMART control rod drive mechanism that can be efficiently utilized for complicated analysis during the design of the SMART. A semi-empirical approach is used to develop the equivalent model; that is, the equivalent model is defined analytically and verified empirically. Two types of tests, dynamic characteristics tests and seismic loading tests, are conducted for the development and verification of the dynamic equivalent model, respectively. Acceleration response spectra from the seismic analysis based on the developed equivalent model show good agreement with those from the seismic loading tests.