Changheui Jang

Oxidation behaviors of wrought nickel-based superalloys in various high temperature environments

Abstract Oxidation characteristics of Alloy 617 and Haynes 230 at 900 °C in simulated helium environment, hot steam environment containing H 2 as well as in air and pure helium conditions were investigated. Compared to air condition, the oxidation rate of Alloy 617 was not significantly affected in helium and hot steam environments, while Haynes 230 showed lower oxidation rate in helium environment. On the other hand, the oxide morphology and structure of Alloy 617 were strongly affected by the environments, but those of Haynes 230 were less dependent on the environments. For Haynes 230, a Cr 2 O 3 inner layer and a protective MnCr 2 O 4 outer layer were formed in all environments, which contributed to the better oxidation resistance. As the mechanical properties, such as creep and tensile properties, were significantly affected by the oxidation behaviors, surface treatment methods to enhance oxidation resistance of these alloys should be developed.

ENVIRONMENTAL FATIGUE OF METALLIC MATERIALS IN NUCLEAR POWER PLANTS – A REVIEW OF KOREAN TEST PROGRAMS

Environmental fatigue of the metallic components in light water reactors has been the subject of extensive research and regulatory interest in Korea and abroad. Especially, it was one of the key domestic issues for the license renewal of operating reactors and licensing of advanced reactors during the early 2000s. To deal with the environmental fatigue issue domestically, a systematic test program has been initiated and is still underway. The materials tested were SA508 Gr.1a low alloy steels, 316LN stainless steels, cast stainless steels, and an Alloy 690 and 52M weld. Through tests and subsequent analysis, the mechanisms of reduced low cycle fatigue life have been investigated for those alloys. In addition, the effects of temperature, dissolved oxygen level, and dissolved hydrogen level on low cycle fatigue behaviors have been investigated. In this paper, the test results and key analysis results are briefly summarized. Finally, an on-going test program for hot-bending of 347 stainless steel is introduced.

DIAMETRAL CREEP PREDICTION OF THE PRESSURE TUBES IN CANDU REACTORS USING A BUNDLE POSITION-WISE LINEAR MODEL

The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.

Round robin analysis for probabilistic structural integrity of reactor pressure vessel under pressurized thermal shock

Performed here is a comparative assessment study for the probabilistic fracture mechanics approach of the pressurized thermal shock of the reactor pressure vessel A round robin consisting of one prerequisite deterministic study and five cases for probabilistic approaches is proposed, and all organizations interested are invited The problems are solved by the paiticipants and their results are compared to issue some recommendation of best practices and to assure an understanding of the key parameters in this type of approach, like transient description and frequency, material properties, defect type and distribution, fracture mechanics methodology etc, which will be useful in the justification through a probabilistic approach for the case of a plant over-passing the screening criteria Six participants from 3 organizations responded to the problem and their results are compiled and analyzed in this study

Fatigue life and crack growth mechanisms of the type 316LN austenitic stainless steel in 310°C deoxygenated water

The low cycle fatigue tests of the type 316LN stainless steel were conducted to investigate the cracking mechanisms in high-temperature water. The fatigue lives of the specimens tested in 310°C deoxygenated water were considerably shorter than those tested in air. For the specimens tested in 310°C deoxygenated water, the evidences for the metal dissolution such as the stream downed feature, the blunt crack shape, and the wider crack opening were observed but rather weakly. In the same specimens, the evidences for the hydrogen-induced cracking such as the coalescence of microvoids and the decrease of the dislocation spacing at the crack tip were observed rather clearly. Therefore, it is thought that the hydrogen-induced cracking is mainly responsible for the reduction in the fatigue life of the type 316LN stainless steel in 310°C deoxygenated water while the effect of metal dissolution is less significant. The hydrogen-induced cracking is more pronounced in the slower strain rates. This behavior is in accordance with the larger reduction in the fatigue life at the slower strain rates. Furthermore, the fatigue life and the dislocation spacing show the minimum value in the strain rate range from 0.008 to 0.04%/s, which indicates the existence of the critical strain rate.

The effects of the stainless steel cladding in pressurized thermal shock evaluation

Abstract Fracture mechanics analysis is the key element of the integrity evaluation of the nuclear reactor pressure vessel (RPV), such as the pressurized thermal shock (PTS) analysis and P–T limit curve construction. However, the existence of stainless steel cladding, with different thermal, physical, and mechanical property at the inner surface of reactor pressure vessel complicates the fracture mechanics analysis. In this paper, the simple analytical treatment schemes to calculate the stress and resulting stress intensity factor at the tip of the flaws in the RPV with stainless steel cladding are introduced. For a reference thermal–hydraulic boundary condition, the effects of cladding thermal conductivity and thermal expansion coefficients on the stress intensity factor of surface flaws were examined. Also, the effects of cladding plasticity and thickness were quantitatively examined. The analysis results showed that the existence of the stainless cladding had significant impacts on the RPV failure probabilities.

Analysis of oxidation behavior of Ni-base superalloys by laser-induced breakdown spectroscopy

The application of laser-induced breakdown spectroscopy (LIBS) to elemental depth profile analysis of high temperature oxidation behavior of Ni-based superalloys is presented. An oxidation test of two Ni-base superalloys, Alloy 617 and Haynes 230, to be considered as the most promising candidates for very high temperature reactor (VHTR), was carried out in dry air at 900 °C. A duplex external oxide layer of MnCr2O4 and Cr2O3 with internal Al2O3 oxides was mainly formed in both alloys. In addition, in Alloy 617, Ni and Ti enriched oxides were observed at the surface unlike in Haynes 230, and Alloy 617 was more susceptible to intergranular oxidation. Generally, the oxidation of Alloy 617 was more extensive than that of Haynes 230, i.e. Alloy 617 formed a thicker oxide layer of ∼8 μm, compared with Haynes 230 with an oxide layer of ∼5 μm in thickness after 1000 h. Depth profiles obtained by LIBS are found to be in broad agreement with those obtained by established techniques such as X-ray diffraction (XRD), scanning electron microscopy coupled to energy dispersive X-ray spectrometry (SEM-EDS), and secondary ion mass spectrometry (SIMS).

Pressurized thermal shock analyses of a reactor pressure vessel using critical crack depth diagrams

Evaluated in this study is the pressure vessel integrity under a pressurized thermal shock. Using transient histories such as temperature, pressure and heat transfer coefficient, the stress distribution is calculated and then stress intensity factors are obtained for a wide range of crack sizes. The stress intensity factors are compared with the fracture toughness to check if cracking is expected to occur during the transient. Critical crack depth diagrams are prepared for each transient which is expected to initiate a pressurized thermal shock accident. Plant-specific analyses of the most limiting plant in Korea are performed to assure the structural integrity of the reactor vessel and the results are discussed.

Tritium inventory prediction in a CANDU plant

Abstract The flow of tritium in a CANDU nuclear power plant was modeled to predict tritium activity build-up. Predictions were generally in good agreement with field measurements for the period 1983–1994. Fractional contributions of coolant and moderator systems to the environmental tritium release were calculated by least square analysis using field data from the Wolsong plant. From the analysis, it was found that: (1) about 94% of tritiated heavy water loss came from the coolant system; (2) however, about 64% of environmental tritium release came from the moderator system. Predictions of environmental tritium release were also in good agreement with field data from a few other CANDU plants. The model was used to calculate future tritium build-up and environmental tritium release at Wolsong site, Korea, where one unit is operating and three more units are under construction. The model predicts the tritium inventory at Wolsong site to increase steadily until it reaches the maximum of 66.3 MCi in the year 2026. The model also predicts the tritium release rate to reach a maximum of 79 KCi/yr in the year 2012. To reduce the tritium inventory at Wolsong site, construction of a tritium removal facility (TRF) is under consideration. The maximum needed TRF capacity of 8.7 MCi/yr was calculated to maintain tritium concentration effectively in CANDU reactors.

Development of a High Toughness Weld for Incoloy Alloy 908

Incoloy® alloy 908+ is a candidate conduit material for the large-scale Nb3Sn superconducting magnets of the International Thermonuclear Experimental Reactor (ITER).1 It is a nickel-iron base precipitation-hardening superalloy with a chemical composition that has been optimized for a low coefficient of thermal expansion, superior cryogenic structural properties, and phase stability during the Nb3Sn reaction heat treatment.2 The alloy precipitates γ′, Ni3(A1,Ti,Nb), as the primary strengthening phase and has demonstrated excellent mechanical properties at both 298 and 4 K.3,4 However, welds of the alloy have shown reduced fracture toughness. Since fabrication of the cablein-conduit conductors for ITER will require welding, the motive of the work has been to improve fracture toughness while maintaining adequate weld strength.