Looking Back on My Experiences with Steam Generator Maintenance and Secondary Side Water Chemistry Control of PWR Power Plants

Author(s):  
Hiroshi TAKAMATSU
Author(s):  
G. Saji ◽  
V. A. Yurmanov ◽  
V. I. Baranenko ◽  
V. A. Fedorova ◽  
G. Karzov ◽  
...  

By focusing on NPPs of Western design (e.g. PWR and BWR), the first author (G. Saji) has established that ‘long cell action’ corrosion plays a pivotal role in practically all unresolved corrosion issues for all types of nuclear power plants as presented in a series of papers already published (1–9). The authors believe that a similar study of NPPs of Russian design, with their unique scientific and technological basis compared to Western plants, are important to illustrate that this mechanism can occur even with different materials, welding technology or operation (e.g. water chemistry control). Among all the differences, it is important to note that PWSCC per se does not seem to be occurring in VVER plants, although no specific reason has yet to be identified. In this paper, a detailed electrochemical assessment is first made on the behavior of ammonia-potassium water chemistry and structural materials at the normal operational temperature in the primary water of VVERs. The chemical and electrochemical characteristics of the ammonia in VVERS were found to be significantly different from those of PWRs which use the hydrogen water chemistry. However, the water chemistry of RBMK is not fundamentally different from that of the Western BWR and therefore the previous studies on SCC of BWRs are generally applicable. On the bases of these studies, various corrosion issues commonly experienced in NPPs of Russian design (VVER and RBMK) are briefly reviewed. They include: (i) pitting corrosion in un-clad VVER-440 RV; (ii) corrosion cracking at the transition welding joints of RV nozzles and piping; (iii) corrosion issues in PGV-440 steam generator collectors; (iv) steam generator tube and collector corrosion; (v) IGSCC in RBMK with austenitic steel piping; (vi) FAC (E-C) in the secondary system of VVERs; and (vii) Anomalous corrosion products sedimentation in the core region in some VVERs. Since the long cell action hypothesis does not seem to contradict the various corrosion activities being experienced in NPPs of Russian design, the first author invites further study on the potential involvement of this mechanism since this hypothesis provides new insight into many of the unresolved corrosion issues. More specifically, the VVERs’ ammonia-potassium water chemistry has theoretically been identified as playing a key role in the prevention of PWSCC, which is one of the most troublesome mechanism of corrosion degradation in many Western PWRs. In view of this significance, the authors proposed an urgent international joint initiative to prove or disprove this mechanism’s existence in nuclear power systems.


Author(s):  
John M. Pietralik ◽  
Kevin L. Heppner

Steam generator (SG) components are subjected to corrosive solutions in turbulent flow. Under such conditions, actual component lifetimes may be significantly reduced from their original design lifetimes. Premature replacement of steam generator components before their expected lifetime can be very expensive. Furthermore, degradation of essential components can reduce the steam generator efficiency, thus reducing net profits. Moreover, a SG failure can also be a safety issue. One of the degradation mechanisms affecting secondary-side SG internal structural elements, which are referred to as internals, is Flow-Accelerated Corrosion (FAC). The susceptibility to FAC depends on flow parameters, water chemistry, and materials. All SG internals made of carbon steel are susceptible to FAC to varying degrees. For FAC susceptibility prediction, flow velocity, pH, and oxygen distributions are needed. SG codes, including THIRST (Thermal Hydraulic analysis In STeam generators, a computer code developed by AECL), traditionally solve for thermalhydraulic parameters. A new chemistry module has been added to THIRST, which now makes this code useful for the prediction of local water chemistry parameters in the SG. The THIRST chemistry module is comprised of a multicomponent, multiphase mass transport model coupled with a multiphase chemical equilibrium model. As input, the module requires amine concentrations in the feedwater and reheater drains. The module predicts local distributions of amine concentration in the secondary side. The concentrations predicted by the module are used to compute the pH. The chemistry module was verified against results of other work in the literature and against station blowdown data. Flow and chemistry predictions of THIRST were used to predict FAC susceptibility for internals of a SG with an integral preheater and a SG without it. Ranking of SG locations in order of FAC susceptibility was estimated from an empirical, Kastner-Riedle model. The most susceptible internals are predicted to be those in the upper section of the hot side and those on the cold side that are near the SG centre, while SG lower regions, including the integral preheater, if one exists, are better protected.


2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.


Author(s):  
Akber Pasha

In recent years the combined cycle has become a very attractive power plant arrangement because of its high cycle efficiency, short order-to-on-line time and flexibility in the sizing when compared to conventional steam power plants. However, optimization of the cycle and selection of combined cycle equipment has become more complex because the three major components, Gas Turbine, Heat Recovery Steam Generator and Steam Turbine, are often designed and built by different manufacturers. Heat Recovery Steam Generators are classified into two major categories — 1) Natural Circulation and 2) Forced Circulation. Both circulation designs have certain advantages, disadvantages and limitations. This paper analyzes various factors including; availability, start-up, gas turbine exhaust conditions, reliability, space requirements, etc., which are affected by the type of circulation and which in turn affect the design, price and performance of the Heat Recovery Steam Generator. Modern trends around the world are discussed and conclusions are drawn as to the best type of circulation for a Heat Recovery Steam Generator for combined cycle application.


2004 ◽  
Vol 41 (1) ◽  
pp. 44-54 ◽  
Author(s):  
Kazutoshi FUJIWARA ◽  
Hirotaka KAWAMURA ◽  
Hiromi KANBE ◽  
Hideo HIRANO ◽  
Hideki TAKIGUCHI ◽  
...  

Wear ◽  
2003 ◽  
Vol 255 (7-12) ◽  
pp. 1198-1208 ◽  
Author(s):  
Young-Ho Lee ◽  
Hyung-Kyu Kim ◽  
Hong-Deok Kim ◽  
Chi-Yong Park ◽  
In-Sup Kim

Author(s):  
Deok Hyun Lee ◽  
Do Haeng Hur ◽  
Myung Sik Choi ◽  
Kyung Mo Kim ◽  
Jung Ho Han ◽  
...  

Occurrences of a stress corrosion cracking in the steam generator tubes of operating nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, expansion transition, u-bend, ding, dent, bulge, etc. Therefore, information on the location, type and quantitative size of a geometric anomaly existing in a tube is a prerequisite to the activity of a non destructive inspection for an alert detection of an earlier crack and the prediction of a further crack evolution [1].


Author(s):  
Salim El Bouzidi ◽  
Marwan Hassan ◽  
Jovica Riznic

Nuclear steam generators are critical components of nuclear power plants. Flow-Induced Vibrations (FIV) are a major threat to the operation of nuclear steam generators. The two main manifestations of FIV in heat exchangers are turbulence and fluidelastic instability, which would add mechanical energy to the system resulting in great levels of vibrations. The consequences on the operation of steam generators are premature wear of the tubes, as well as development of cracks that may leak radioactive heavy water. This paper investigates the effect of tube support clearance on crack propagation. A crack growth model is used to simulate the growth of Surface Flaws and Through-Wall Cracks of various initial sizes due to a wide range of support clearances. Leakage rates are predicted using a two-phase flow leakage model. Non-linear finite element analysis is used to simulate a full U-bend subjected to fluidelastic and turbulence forces. Monte Carlo Simulations are then used to conduct a probabilistic assessment of steam generator life due to crack development.


2008 ◽  
Vol 50 (8) ◽  
pp. 506-510
Author(s):  
Shunsuke UCHIDA ◽  
Yosuke KATSUMURA ◽  
Motomasa FUSE ◽  
Takahiro SHIOKAWA ◽  
Hideki TAKIGUCHI

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