Corrosion Susceptibility in High Temperature Liquid/Vapor Environments of Materials for Tubing of Heavy-Water Collection System in CANDU Nuclear Power Plant

Author(s):  
Guangfu Li ◽  
Liang Zhao ◽  
Xinghong Yang

Corrosion especially pitting on the inner surface was regarded as the first step of the failure process of 316L and 304L stainless steel tubes of heavy-water collection systems in CANDU reactors of a power plant. In this work, four materials including the 316L, 304L, carbon steel A106B and Ni-based Alloy 690 were tested in 14 designed liquid/vapor LiOH-containing environments at both 250 and 150°C, to obtain basic information on the corrosion susceptibilities as a function of temperature, media state, Cl− and Li+ contents, for materials selection. Results showed that the corrosion susceptibility rank were A106B at the top, 690 at the bottom and the stainless steels between them. The corrosion susceptibility was lower in a liquid solution than in the vapor above the liquid, and was relatively lower at 150°C than at 250°C for the same media. Chloride promoted corrosion significantly but LiOH showed some protect effect. In the Cl-free solutions at 150°C, A106B exhibited almost no corrosion in the liquid but localized corrosion in the vapor above. It corroded apparently when adding 3%NaCl into the solution or raising temperature to 250°C. The stainless steels showed no corrosion in both the liquid and vapor states of Cl-free solutions at both 150 and 250°C, but exhibiting pitting in the vapor when 3%NaCl added to the solution. 690 always exhibited excellent corrosion resistance during long term tests in various environments.

CORROSION ◽  
10.5006/3319 ◽  
2019 ◽  
Vol 75 (11) ◽  
pp. 1276-1280
Author(s):  
Y. Emun ◽  
H.S. Zurob ◽  
J.R. Kish

This study compares the localized (exterior) corrosion susceptibility of chromized steel to bench-mark ferritic stainless steels for automotive exhaust applications. Continuous near-neutral salt fog exposure (ASTM B117) was used for this purpose. Corrosion susceptibility was determined using mass loss measurements coupled with a post exposure metallographic examination. Complementary potentiodynamic polarization measurements were made in the bulk salt solution to help interpret the relative performance. The elevated Cr content provided by the chromizing surface treatment provides comparable corrosion resistance to the more highly alloyed ferritic stainless steels studied. The major factor affecting localized corrosion susceptibility is the formation of rust deposits, which act as effective pit-like corrosion initiation sites.


2012 ◽  
Vol 2012 ◽  
pp. 1-17 ◽  
Author(s):  
Analia Bonelli ◽  
Oscar Mazzantini ◽  
Martin Sonnenkalb ◽  
Marcelo Caputo ◽  
Juan Matias García ◽  
...  

A description of the results for a Station Black-Out analysis for Atucha 2 Nuclear Power Plant is presented here. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be available. It is assumed that during the Station Black-Out sequence the first pressurizer safety valve fails stuck open after 3 cycles of water release, respectively, 17 cycles in total. During the transient, the water in the fuel channels evaporates first while the moderator tank is still partially full. The moderator tank inventory acts as a temporary heat sink for the decay heat, which is evacuated through conduction and radiation heat transfer, delaying core degradation. This feature, together with the large volume of the steel filler pieces in the lower plenum and a high primary system volume to thermal power ratio, derives in a very slow transient in which RPV failure time is four to five times larger than that of other German PWRs.


1998 ◽  
Vol 120 (1) ◽  
pp. 93-98 ◽  
Author(s):  
G. R. Reddy ◽  
H. S. Kushwaha ◽  
S. C. Mahajan ◽  
K. Suzuki

Generally, for the seismic analysis of nuclear power plant structures, requirement of coupling equipment is checked by applying USNRC decoupling criteria. This criteria is developed for the equipment connected to the structure at one location. In this paper, limitations of this criteria and modifications required for application to real life structures such as pressurized heavy water reactor building are discussed. In addition, the authors endeavor to present a decoupling model for multi-connected structure-equipment. The applicability of the model is demonstrated with pressurized heavy water reactor building internal structure and steam generator.


Author(s):  
Lei You ◽  
Fuchun Sun ◽  
Pan He ◽  
Hongkun Xu ◽  
Fang Fang

In this paper, we develop a monitoring system of reactor coolant pumps in nuclear power plant (CPS). The safe running of reactor coolant pump is important for nuclear power plant. Based on the Fourier transform (FT) and some algorithm, The data collected from the pump are analyzed. Once the accident happens, it would cause unimaginable outcome. The system will be jumped to failure process mode when the pump has something wrong. The advanced VXI and virtual instrument technology are applied to system, and the reactor coolant pump will be monitored overall so as to assure that the reactor coolant pump runs in safe, which has a significant value to secure the safe operation and reliability of the nuclear plant. The monitoring system will help the operators find fault of reactor coolant pump.


2016 ◽  
Vol 92 ◽  
pp. 284-288 ◽  
Author(s):  
Hocheol Shin ◽  
Changhoi Kim ◽  
Yongchil Seo ◽  
Kyungmin Jeong ◽  
Youngsoo Choi ◽  
...  

2007 ◽  
Vol 2007 ◽  
pp. 1-9
Author(s):  
O. Mazzantini ◽  
J. C. Ferreri ◽  
F. D'Auria ◽  
C. P. Camusso

A systematic study of natural circulation (NC) in a postulated, varying primary mass inventory scenario at residual power fractions has been performed for a nuclear power plant operating in Argentina. It is a pressurized heavy water reactor, cooled and moderated by heavy water. The analysis seems particularly relevant at present, because a second nuclear power plant (NPP), of similar design and nearly 745 MWe, is now under finalization. NRC-RELAP5/MOD3.3 was the code used to perform the simulations. Results obtained are presented in the form of natural circulation flow maps. The trends obtained fit in the expected limits for integral test facilities representative of PWRs. In addition, the validity of a simplified analysis to scale single and two-phase core flow has been verified. A set of constants has been obtained, which permits predicting NC core mass flow rate (CMFR) for this NPP. Results are partially validated, for single-phase NC flow, using a documented plant transient, showing reasonable agreement. Also, the effect of pressurizer size on the predicted evolution curve in the NC flow map (NCFM) is discussed.


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