• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.431-438
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• 2014

The effect of various pH agents on the corrosion behavior of carbon steel was investigated under a simulated secondary water chemistry condition of a pressurized water reactor (PWR) in a laboratory, and the steel's corrosion performance was compared with the field data obtained from Uljin NPP unit 2 reactor. All tests were carried out at temperatures of $50^{\circ}C-250^{\circ}C$and pH of 8.5 - 10. The pH at a given temperature was controlled by adding different agents. Laboratory data indicate that the corrosion rate of carbon steel decreased as the pH increased under the test conditions and the highest corrosion rate was measured at $150^{\circ}C$. This high corrosion rate may be related to high dissolution and instability of Fe oxide ($Fe_3O_4$) at $150^{\circ}C$. It was also found that an addition of ethanolamine (ETA) to ammonia was more effectivefor anticorrosion than ammonia alone, and that mixed treatment reduced 50% of iron or more at pHs of 9.5 or higher, especially in the steam generator (SG) and the moisture separator & re-heater (MSR).

• Journal of the Korea Academia-Industrial cooperation Society
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• v.9 no.6
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• pp.1768-1774
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• 2008

The particulate behaviors of nickel ferrite were investigated under the simulated PWR shutdown chemistry conditions. Temperature of the simulated water with concentration of 0.1 ppm Li and 2,000 ppm B was dropped from $300^{\circ}C$ to $150^{\circ}C$ with a rate of $0.625^{\circ}C/min$ and then constantly maintained at $150^{\circ}C$ under the pressure of 2,500 psi. The on-line particle counting and the concentration measurement of nickel dissolved were performed under 5, 15 and 25 cc/kg $H_2O$ dissolved hydrogen. Experimental results showed that total particle count in the simulated water was not greatly changed for three hydrogen concentrations as temperature was decreased. However, particles were smaller as temperature was decreased and then maintained constantly. The degree of variation in particle size distribution was greater at 15 cc/kg $H_2O$ dissolved hydrogen than any other dissolved hydrogen concentrations. Concentration of nickel ion was increased as temperature was decreased and was higher at 15 cc/kg $H_2O$ dissolved hydrogen than any other dissolved hydrogen concentrations. Theses results show that nickel ferrite is unstable with temperature variation and at dissolved hydrogen concentration of 15 cc/kg $H_2O$.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.187-198
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• 2009

In this paper, three power ramp tests performed on high burn-up Re-crystallized Zircaloy2 - UO2 BWR fuel rods (56 to 63 MWd/kgU) within the SCIP project are simulated with METEOR and ALCYONE 3D. Two of the ramp tests are of staircase type up to Linear Heat Rates of 420 and 520 W/cm and with long holding periods. Failure of the 420 W/cm fuel rod was observed after 40 minutes. The third ramp test consisted of a more standard ramp test with a constant power rate of 80 W/cm/min up to 410 W/cm with a short holding time. The tests were first simulated with the METEOR 1D fuel rod code, which gave accurate results in terms of profilometry and fission gas releases. The behaviour of a fuel pellet fragment and of the cladding piece on top of it was then investigated with ALCYONE 3D. The size and the main characteristics of the ridges after base irradiation and power ramp testing were recovered. Finally, the failure criteria validated for PWR conditions and fuel rods with low-to-medium burn-ups were used to analyze the failure probability of the KKL rodlets during ramp testing.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1173-1183
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• 2018

Transient fluid velocity and pressure fields in a pressurized water reactor (PWR) steam generator (SG) secondary side during the blowdown period of a feedwater line break (FWLB) accident were numerically simulated employing the saturated water flashing model. This model is based on the assumption that compressed water in the SG is saturated at the beginning and decompresses into the two-phase region where saturated vapor forms, creating a mixture of steam bubbles in water by bulk boiling. The numerical calculations were performed for two cases of which the outflow boundary conditions are different from each other; one is specified as the direct blowdown discharge to the atmosphere and the other is specified as the blowdown discharge to an extended calculation domain with atmospheric pressure on its boundary. The present simulation results obtained using the two different outflow boundary conditions were discussed through a comparison with the predictions using a simple non-flashing model neglecting the effects of phase change. In addition, the applicability of each of the non-flashing water discharge and saturated water flashing models for the confirmatory assessments of new SG designs was examined.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.125-133
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• 2004

Fission products of DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel, irradiated in HANARO research reactor with 61 ㎾/m of maximum linear power and 1,770 ㎿d/tU of average burn-up, was characterized by EPMA(Electron Probe Micro Analyzer). In order to find accurate characterization, the analysis results by EPMA of fresh simulated DUPIC fuel containing fission products as chemicals were compared with that of wet chemical analysis. The metallic precipitates observed at the center of the fresh simulated DUPIC fuel were about 1 $\mu\textrm{m}$ in size and their major components by EPMA were Mo-53.89 at.%, Ru-37.40 at.%, and Pd+Rh-8.71 at.%. Established procedure through the fresh simulated DUPIC fuel was applied to the irradiated DUPIC fuel. Observed size of metallic precipitates were 2∼2.5 $\mu\textrm{m}$ and their compositions were Mo-47.34 at.%, Ru-46 at.%, and Pd+Rh-6.65 at.%. What are uncommon things for this experiment, special treatment for improving the conductivity was attempted to the specimen and the conditions of exact irradiation of electron beam to small metallic precipitate were suggested.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.228-237
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• 2019

The crack growth responses of as-received and as-welded Alloy 600/182 and Alloy 690/152 welds to constant loading were measured by a direct current potential drop method using compact tension specimens in primary water at $325^{\circ}C$ simulating the normal operating conditions of a nuclear power plant. The as-received Alloy 600 showed crack growth rates (CGRs) between $9.6{\times}10^{-9}mm/s$ and $3.8{\times}10^{-8}mm/s$, and the as-welded Alloy 182 had CGRs between $7.9{\times}10^{-8}mm/s$ and $7.5{\times}10^{-7}mm/s$ within the range of the applied loadings. These results indicate that Alloys 600 and 182 are susceptible to cracking. The average CGR of the as-welded Alloy 152 was found to be $2.8{\times}10^{-9}mm/s$. Therefore, Alloy 152 was proven to be highly resistant to cracking. The as-received Alloy 690 showed no crack growth even with an inhomogeneous banded microstructure. The cracking mode of Alloys 600 and 182 was an intergranular cracking; however, Alloy 152 was revealed to have a mixed (intergranular + transgranular) cracking mode. It appears that the Cr concentration and the microstructural features significantly affect the cracking resistance and the cracking behavior of Ni-base alloys in PWR primary water.

• Corrosion Science and Technology
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• v.20 no.3
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• pp.158-168
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• 2021

Austenitic 316 stainless steel was irradiated with protons accelerated by an energy of 2 MeV at 360 ℃, the various defects induced by this proton irradiation were characterized with microscopic equipment. In our observations irradiation defects such as dislocations and micro-voids were clearly revealed. The typical irradiation defects observed differed according to depth, indicating the evolution of irradiation defects follows the characteristics of radiation damage profiles that depend on depth. Surface oxidation tests were conducted under the simulated primary water conditions of a pressurized water reactor (PWR) to understand the role irradiation defects play in surface oxidation behavior and also to investigate the resultant irradiation assisted stress corrosion cracking (IASCC) susceptibility that occurs after exposure to PWR primary water. We found that Cr and Fe became depleted while Ni was enriched at the grain boundary beneath the surface oxidation layer both in the non-irradiated and proton-irradiated specimens. However, the degree of Cr/Fe depletion and Ni enrichment was much higher in the proton-irradiated sample than in the non-irradiated one owing to radiation-induced segregation and the irradiation defects. The microstructural and microchemical changes induced by proton irradiation all appear to significantly increase the susceptibility of austenitic 316 stainless steel to IASCC.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.65 no.5
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• pp.836-842
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• 2016

When an accident occurs in the nuclear power plant, the faulted information might mislead to the high possibility of aggravating the accident. At the Fukushima accident, the operators misunderstood that there was no core exposure despite in the processing of core damage, because the instrument information of the reactor water level was provided to the operators optimistically other than the actual situation. Thus, this misunderstanding actually caused to much confusions on the rapid countermeasure on the accident, and then resulted in multiplying the accident propagation. It is necessary to be equipped with the function that informs operators the status of instrument integrity in real time. If plant operators verify that the instruments are working properly during accident conditions, they are able to make a decision more safely. In this study, we have performed various tests for the fault detection sensitivity of an data-driven empirical model to review the usability of the model in the accident conditions. The test was performed by using simulation data from the compact nuclear simulator that is numerically simulated to PWR type nuclear power plant. As a result of the test, the proposed model has shown good performance for detecting the specified instrument faults during normal plant conditions. Although the instrument fault detection sensitivity during plant accident conditions is lower than that during normal condition, the data-drive empirical model can be detected an instrument fault during early stage of plant accidents.

• Journal of Mechanical Science and Technology
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• v.14 no.8
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• pp.816-822
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• 2000

In the core of the nuclear power plant of PWR, several cases of fuel failure by unknown causes have been experienced for various fuel types. From the common features of the failure pattern, failure lead time, flow conditions, and flow induced vibration characteristics in nuclear fuel bundles, it is deduced that the fretting wear failure of the fuel rod at the spacer grid position is due to the fluidelastic vibration. In the past, fluidelastic vibration was simulated by quasi -static semi-analytical model, so called the static model, which could not account for the interaction between the rods within a bundle. To overcome this defect and to provide for more flexibilities applicable to the fuel bundle, Tanaka's unsteady model was modified to accomodate the geometrical differences and governing parameter changes during the operations such as the number of rods, pitch to diameter ratio (P/D), spring force, damping coefficient, etc. The critical velocity was calculated by solving the governing equations with the MATLAB code. A comparison between the estimated critical velocity and the test result shows a good agreement. Finally, the level of decrease of the critical velocity due to the reduction in the spring force and reduced damping coefficient due to the radiation exposure is also estimated.