• Transactions of the Korean Society of Mechanical Engineers A
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• v.41 no.5
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• pp.375-380
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• 2017

This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the $B_{0.004}$ life would be guaranteed for the SFR fuel rods with 99％ confidence if no failure was observed at any of the contact surfaces of the five specimens.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.15-20
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• 1997

We present the results HELIOS verification against VENUS PWR critical experiments loaded with high plutonium content mixed oxides fuels. The effective multiplication factors are calculated to be slightly supercritical within an acceptable error bound. In the prediction of power shape, HELIOS results are in close agreement with the measured values. The RMS errors of re-normalized calculated fission rate distribution are less than 1.4 % with either explicit or implicit models or micro tubes/rods in each fuel assembly for both ALL-MOX and GD-MOX mock-up cores.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.225-230
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• 1996

A three-dimensional reactor core simulation code, MASTER has been developed as a part of ADONIS which is the Korean core design package in KAERI. CASMO-3 is used as a precedent lattice code for two-group microscopic cross section and heterogeneous formfunctions. The pin power reconstruction capability of CASMO-3/MASTER was evaluated for a validation and verification Five B&W critical experiments were selected as benchmark problems. These problems included two experiments for CE-type and three for WH-type fuel assemblies. Two of them contained gadolinia rods as burnable absorber. Comparison of the calculated pin power distributions with the measured ones demonstrate that CASMO-3/MASTER can predict the pin power distribution as well as CASMO-3/SIMULATE-3.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.7
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• pp.689-695
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• 2012

The friction factor in a rod bundle and the loss coefficient at a spacer grid were examined. As a test section, 25 smooth rods, 9.5 mm in diameter and 2000 mm in length, were prepared and installed in a $5{\times}5$ square array in a square channel. In this case, the P/D (Pitch-to-Diameter ratio) was 1.35. In this work, plain (i.e., no mixing vanes), split-vane, and hybrid-vane spacer grids were tested. In a bare rod bundle (i.e., no spacer grid), the measured friction factors were in good agreement with the previous correlations. Among the spacer grids tested, the hybrid-vane spacer grid presented the largest friction factor in the rod bundle and loss coefficient. This may be because of the flow pattern change induced by large relative plugging of the flow cross section and mixing vane geometry. At Re=$5{\times}10^5$, the predicted loss coefficients of plain, splitvane, and hybrid-vane spacer grids were approximately 0.79, 0.80, and 0.88, respectively.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.237-246
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• 1994

A high flow critical heat flux (CHF) correlation, based on the single-pin CHF experimental data for finned and unfinned heated rods, was developed for the thermal-hydraulic design and safety analysis of the Korea Multi-purpose Research Reactor (KMRR) core. The correlation consists of dimensionless parameters such as Reynolds number, thermodynamic equilibrium quality, liquid-to-vapor density ratio, and hydraulic equivalent diameter ratio. The fin effect was taken into account in the correlation by a finned-to-unfinned heated perimeter ratio. The effects of a cold wall and non-uniform axial power distribution ore discussed to verify the applicability of the single-pin based correlation to the KMRR fuel bundle. The correlation limit departure from nucleate boiling ratio (DNBR) was determined as 1.44 from the statistical analysis of the CHF data.

• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.342-352
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• 1998

A nuclear design feasibility of soluble boron free(SBF core for the medium-sized(600MWe) PWR was investigated. The result conformed that soluble boron free operation could be performed by using current PWR proven technologies. Westinghouse advanced reactor, AP-600 was chosen as a design prototype. Design modification was applied for the assembly design with burnable poison and control rod absorber material. In order to control excess reactivity, large amount of gadolinia integral burnable poison rods were used and B4C was used as a control rod absorber material. For control of bottom shift axial power shape due to high temperature feedback in SBF core, axial zoning of burnable poison was applied to the fuel assemblies design. The combination of enrichment and rod number zoning for burnable poison could make an excess reactivity swing flat within around 1% and these also led effective control on axial power offset and peak pin power, The safety assessment of the designed core was peformed by the calculation of MTC, FTC and shutdown margin. MTC in designed SBF core was greater around 6 times than one of Ulchin unit 3&4. Utilization of enriched BIO(up to 50w1o) in B4C shutdown control rods provided enough shutdown margin as well as subcriticality at cold refueling condition.

• Corrosion Science and Technology
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• v.19 no.1
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• pp.1-7
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• 2020

Chalk River Unidentified Deposits (CRUDs) were collected from the Korean pressurized water reactor (PWR) plant (A, B, and C) where the axial offset anomaly (AOA) occurred. AOA, also known as a CRUD-induced power shift, is one of the key issues in maintaining stable PWR plant operations. CRUDs were sampled from spent nuclear fuel rods and analyzed using an electron probe micro-analyzer (EPMA). This paper describes the characteristics of boron-deposits from the CRUDs sampled from twice-burnt assemblies from the Korean PWR. The primary coolant of a PWR contains boron and lithium. It is known that boron deposition occurs in a thick CRUD layer under substantial sub-cooled nucleate boiling (SNB). The results of this study are summarized as follows. Boron was not found at the locations where the existence was confirmed in simulated CRUDs, in other words, the cladding and CRUD boundaries. Nevertheless, we clearly observed the presence of boron and confirmed that boron existed as a lump in crystalline form. In addition, the study confirmed that CRUD existed in a crystal form with a unique size of about 10 ㎛.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4329-4337
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• 2022

Jet impinging device is designed for decay heat removal on horizontal fuel rods in a low temperature heating reactor. An experimental system with a fuel rod simulator is established and experiments are performed to evaluate water film covering capacity, within 0.0287-0.0444 kg/ms mass flow rate, 0-164.1 kW/m² heating flux and 13.8-91.4℃ feeding water temperature. An effective method to obtain the film coverage rate by infrared equipment is proposed. Water film flowing patterns are recoded and the film coverage rates at different circumference angles are measured. It is found the film coverage rate decreases with heating flux during single-phase convection, while increases after onset of nucleate boiling. Besides, film coverage rate is found affected by Marangoni effect and film accelerating effect, and surface wetting is significantly facilitated by bubble behavior. Based on the observed phenomenon and physical mechanism, dry-out depth and initial dry-out rate are proposed to evaluate film covering potential on a heating surface. A model to predict film coverage rate is proposed based on the data. The findings would have reliable guide and important implications for further evaluation and design of decay heat removal system of new reactors, and could be helpful for passive containment cooling research.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1138-1147
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• 2018

Delayed hydride cracking (DHC) is an important failure mechanism for Zircaloy tubes in the demanding environment of nuclear reactors. The threshold stress intensity factor, $K_{IH}$, and critical hydride length, $l_C$, are important parameters to evaluate DHC. Theoretical models of them are developed for Zircaloy tubes undergoing non-homogenous temperature loading, with new stress distributions ahead of the crack tip and thermal stresses involved. A new stress distribution in the plastic zone ahead of the crack tip is proposed according to the fracture mechanics theory of second-order estimate of plastic zone size. The developed models with fewer fitting parameters are validated with the experimental results for $K_{IH}$ and $l_C$. The research results for radial cracking cases indicate that a better agreement for $K_{IH}$ can be achieved; the negative axial thermal stresses can lessen $K_{IH}$ and enlarge the critical hydride length, so its effect should be considered in the safety evaluation and constraint design for fuel rods; the critical hydride length $l_C$ changes slightly in a certain range of stress intensity factors, which interprets the phenomenon that the DHC velocity varies slowly in the steady crack growth stage. Besides, the sensitivity analysis of model parameters demonstrates that an increase in yield strength of zircaloy will result in a decrease in the critical hydride length $l_C$, and $K_{IH}$ will firstly decrease and then have a trend to increase with the yield strength of Zircaloy; higher fracture strength of hydrided zircaloy will lead to very high values of threshold stress intensity factor and critical hydride length at higher temperatures, which might be the main mechanism of crack arrest for some Zircaloy materials.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.109-116
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• 2003

The amount of radioactive waste is expected to be increased continuously because of the rapid growth of the domestic nuclear industry, full power operation of the HANARO reactor and the increased research activities of the nuclear fuel cycle. Accordingly the efforts are focused to achieve the handling of radioactive waste in safe and reduce the volume of radioactive waste. The PIEF is carrying out the PIE (post irradiation examination) of spent fuel rods related to the identification of cause defect and evaluation of integration safety. This study describes the technologies and experiences of compaction, shredding and cutting of the solid radioactive waste used in the PIE. The quantity of the high level waste was reduced by 1/12 using the 100-ton compressor installed in hot-cell. Also middle and low level waste was reduced by 1/8 using the 60-ton compressor installed in intervention area. Plastic drums were shredded by crusher to be compacted in the ratio of 1/5, used filters in the ratio of 1/6 and the number of drum is also reduced by cutting procedure for the non-volatile materials such as metal.