• Korean Journal of Materials Research
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• v.25 no.4
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• pp.183-190
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• 2015

Fatigue crack growth rate tests were conducted as a function of temperature, dissolved hydrogen (DH) level, and frequency in a simulated PWR environment. Fatigue crack growth rates increased slightly with increasing temperature in air. However, the fatigue crack growth rate did not change with increasing temperature in PWR water conditions. The DH levels did not affect the measured crack growth rate under the given test conditions. At $316^{\circ}C$, oxides were observed on the fatigue crack surface, where the size of the oxide particles was about $0.2{\mu}m$ at 5 ppb. Fatigue crack growth rate increased slightly with decreasing frequency within the frequency range of 0.1 Hz and 10 Hz in PWR water conditions; however, crack growth rate increased considerably at 0.01 Hz. The decrease of the fatigue crack growth rate in PWR water condition is attributed to crack closure resulting from the formation of oxides near the crack tips at a rather fast loading frequency of 10 Hz.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.773-780
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• 2012

The article presented is concerned with an evaluation of the corrosion behavior of SA-508 low alloy steel (LAS) and Type 309L stainless steel (SS) cladding of a reactor pressure vessel under the simulated primary water chemistry of a pressurized water reactor (PWR). The uniform corrosion and galvanic corrosion rates of SA-508 LAS and Type 309L SS were measured in three different control conditions: power operation, shutdown, and power operation followed by shutdown. In all conditions, the dissimilar metal coupling of SA-508 LAS and Type 309L SS exhibited higher corrosion rates than the SA-508 base metal itself due to severe galvanic corrosion near the cladding interface, while the corrosion of Type 309L in the primary water environment was minimal. The galvanic corrosion rate of the SA-508 LAS and Type 309L SS couple measured under the simulated power operation condition was much lower than that measured in the simulated shutdown condition due to the formation of magnetite on the metal surface in a reducing environment. Based on the experimental results, the corrosion rate of SA-508 LAS clad with Type 309L SS was estimated as a function of operating cycle simulated for a typical PWR.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.851-859
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• 2021

Small PWR can be used for power generation and heating. Considering that small PWR has the characteristics of flexible operating conditions and complex operating environment, the controller designed based on single power level is difficult to achieve the ideal control of small PWR in the whole range of core power range. To solve this problem, a flexible switching controller based on fuzzy controller and LQG/LTR controller is designed. Firstly, a core fuzzy multi-model suitable for full power range is established. Then, T-S fuzzy rules are designed to realize the flexible switching between fuzzy controller and LQG/LTR controller. Finally, based on the core power feedback principle, the core flexible switching control system of small PWR is established and simulated. The results show that the flexible switching controller can effectively control the core power of small PWR and the control effect has the advantages of both fuzzy controller and LQG/LTR controller.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.752-759
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• 2017

The fatigue crack growth behavior of Type 347 (S347) and Type 347N (S347N) stainless steel was evaluated under the operating conditions of a pressurized water reactor (PWR). These two materials showed different fatigue crack growth rates (FCGRs) according to the changes in dissolved oxygen content and frequency. Under the simulated PWR conditions for normal operation, the FCGR of S347N was lower than that of S347 and insensitive to the changes in PWR water conditions. The higher yield strength and better corrosion resistance of the nitrogen-alloyed Type 347 stainless steel might be a main cause of slower FCGR and more stable properties against changes in environmental conditions.

• Journal of Surface Science and Engineering
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• v.57 no.5
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• pp.398-405
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• 2024

In this study, ZnO nanoparticle treatments were applied to stainless steel 304 to mitigate the generation of stress corrosion cracking (SCC) under pressurized water reactor (PWR)-simulated conditions, focusing on temperature and pressure (300℃, 150 bar), specifically simulating temperature and pressure. ZnO nanoparticles were synthesized via plasma discharge in an aqueous solution, with sizes ranging from 355 ± 142 nm to 25.7 ± 7.2 nm along the long axis, controlled by adjusting the voltage parameters. After treatment with 25 nm ZnO nanoparticle treatment, the surface of stainless steel 304 was analyzed using X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS), confirming the formation of a compact and dense ZnCr₂O₄ spinel oxide film with a thickness of approximately 65 nm. Corrosion potential tests conducted using a Potentiostat/Galvanostat revealed that corrosion resistance improved as ZnO nanoparticle size decreased. Additionally, U-bend tests under accelerated corrosion conditions showed significantly reduced SCC in samples treated with 25 nm ZnO nanoparticles. These findings suggest that ZnO nanoparticles synthesized via plasma discharge could be effectively applied for SCC mitigation in the nuclear industry.

• Corrosion Science and Technology
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• v.14 no.3
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• pp.113-119
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• 2015

Alloy 82/182 weld metals had been extensively used in joining the components of the PWR primary system. Unfortunately, there have been a number of incidents of cracking caused by PWSCC in Alloy 82/182 welds during the operation of PWR worldwide. To mitigate PWSCC, optimization of water-chemistry conditions, especially dissolved hydrogen (DH) and Zn contents, is considered as the most promising and effective remedial method. In this study, the PWSCC behaviours of Alloy 182 weld were investigated in simulated PWR environments with various DH content. Both in-situ and ex-situ oxide characterizations as well as PWSCC initiation tests were performed. The results showed that PWSCC crack initiation time was shortest in PWR water (DH: 30cc/kg). Also, high stress reduced crack initiation time. Oxide layer showed multi-layered structures consisted of the outer needle-like Ni-rich oxide layer, Fe-rich crystalline oxide, and inner Cr-rich inner oxide layers, which was not altered by the level of applied stress. To analyse the multi-layer structure of oxides, EIS measurement were fitted into an equivalent circuit model. Further analyses including TEM and EDS are underway to verify appropriateness of the equivalent circuit model.

• Nuclear Engineering and Technology
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• v.28 no.4
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• pp.405-414
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• 1996

In this paper, the unsteady 2-dimensional turbulent flow model for thermal stratification in a pressurizer surge line of PWR plant is proposed to numerically investigate the heat transfer and flow characteristics. The turbulence model is adapted to the low Reynolds number K-$\varepsilon$ model (Davidson model). The dimensionless governing equations are solved by using the SIMPLE (Semi-Implicit Method for Pressure Linked Equations) algorithm. The results are compared with simulated experimental results of TEMR Test. The time-dependent temperature profiles in the fluid and pipe nil are shown with the thermal stratification occurring in the horizontal section of the pipe. The corresponding thermal stresses are also presented. The numerical result for thermal stratification by the outsurge during heatup operation of PWR shows that the maximum dimensionless temperature difference is about 0.83 between hot and cold sections of pipe well and the maximum thermal stress is calculated about 322MPa at the dimensionless time 28.5 under given conditions.

• Nuclear Engineering and Technology
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• v.56 no.10
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• pp.4263-4279
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• 2024

The small modular PWR (SMPWR) usually adopts integral design. Under severe accident, the system responses are different from those large PWRs. It is necessary to study the severe accident behavior of the SMPWR. A MELCOR model is developed for SMPWR and its steady-state results are in good agreement with the design values. Severe accidents induced by secondary pipeline break accidents are simulated, and no pressure relief measures are taken to keep the primary loop under high pressure. The mitigation effects of passive containment air cooling system (PAS) and passive cavity injection system (PCIS) are evaluated under different cases. The results show that under high pressure conditions, PCIS can effectively cool the lower head. The earlier the PCIS operates, the more significant the mitigation effect can be. In addition, PAS can effectively reduce the peak pressure and temperature in the containment. This study can provide a reference for the formulation of severe accident management guidelines on SMPWRs.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2552-2564
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• 2020

The crevice stress corrosion cracking (SCC) susceptibility of austenitic stainless steels was evaluated in simulated pressurized water reactor (PWR) environments. To simulate the abnormal condition in temporary clamping devices on leaking small bore pipes, crevice bent beam (CBB) tests were performed in the oxygenated as well as hydrogenated conditions. No SCC cracks were found for SS316 in both conditions. SS304 also showed good resistance in the hydrogenated condition. However, all SS304 specimens showed SCC cracks in the oxygenated condition, indicating poor crevice SCC resistance. It was found that residual ferrites were selectively dissolved because of the galvanic corrosion coupled with the neigh-bouring austenite phase, resulting in SCC initiation in SS304. Crack morphologies were mostly transgranular assisted by the damaged δ-ferrite and deformation-induced slip bands.

• Nuclear Engineering and Technology
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• v.28 no.6
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• pp.542-550
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• 1996

Nuclear ship reactors have several features different from land-based PWR's. Especially, effects of ship motions on reactor thermal-hydraulics and good load following capability for abrupt load changes are essential characteristics of nuclear ship reactors. This study modified the RETRAN-03 to analyze the thermal-hydraulic transients under three-dimensional ship motions, named RETRAN-03/MOV in order to apply to future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. Calculations have been peformed under rolling, heaving and stationary inclination conditions during normal operation. Also, the natural circulation has been analyzed, which can provide the decay heat removal to ensure the passive safety of marine reactors. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations have been successfully simulated at various conditions.