• 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.33 no.2
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• pp.132-144
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• 2001

Fracture mechanics is one of the major areas of the pressurized thermal shock (PTS) evaluation. To evaluate the reactor pressure vessel integrity associated with PTS, PFM methodology demands precise calculation of temperature, stress, and stress intensity factor for the variety of PTS transients. 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, treatment schemes to evaluate stress and resulting stress intensity factor for RPV with stainless steel clad are introduced. For a reference transient, the effects of clad thermal conductivity and thermal expansion coefficients on deterministic fracture mechanics analysis are examined.

• Proceedings of the KWS Conference
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• 2001.05a
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• pp.228-231
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• 2001

Elastic-plastic fracture toughness $J_{1c}$ can be used as an effective design criterion in elastic plastic fracture mechanics. Most of these systems are operated at high temperature and $J_{1c}$ values are affected by temperature. therefore, the $J_{1c}$ valuse at high temperature must be determined for use of integrity evaluation and designing of such systems. Elastic-plastic fracture toughness $J_{1c}$ tests were performed on SA516 carbon steel plate and test results were analyzed according to ASTM E 813-8, ASTM 1813-89. Safety and integrity are required for reactor pressure vessels vecause pthey are operated in high temperature. there are single specimen method, which used as evaluation of safety and integrity for reactor pressure vessels. In this study, elastic-plastic fracture toughness$(J_{1c})$ and $J-\Delta{a}$ of SA 516/70 steel used as reactor pressure vessel steel are measured and evaluated at room Temperature, $150^{\circ}C$, $250^{\circ}C$ and $370^{\circ}C$ according to unloading compliance method.

• Journal of Ocean Engineering and Technology
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• v.15 no.2
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• pp.99-104
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• 2001

Elastic-plastic fracture toughness $J_{lc}$ can be used as an effective design criterion in elastic plastic fracture mechanics. Most of these systems are$J_{lc}$ $J_{lc}$ value at high temperature must be determined for use of integrity evaluation and designing of such systems. Elastic-plastic fracture toughness $J_{lc}$ tests were performed on SA516/70 carbon steel plate and test results were analyzed according to ASTM E 813-87, ASTM E 813-89 and ASTM E 1152-87.safety and integrity are required for reactor pressure vessels because, they are operated in high temperature. There are single specimen method, which used as evaluation of safety and integrity for reactor pressure vessels. In this study, elastic-plastic fracture toughness($J_{lc}$) and J-$\Delta$a of SA 516/70 steel used as reactor pressure vessel steel are measured and evaluated at room temperature, 150$^{\circ}C $, 250$^{\circ}C $ and 370$^{\circ}C $ according to unloading compliance method.

• Corrosion Science and Technology
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• v.12 no.2
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• pp.85-92
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• 2013

This work is concerned with an evaluation of dynamic boric acid corrosion (BAC) of low alloy steel for reactor pressure vessel of a pressurized water reactor (PWR). Mockup test method was newly established to investigate dynamic BAC of the low alloy steel under various conditions simulating a primary water leakage incident. The average corrosion rate was measured from the weight loss of the low alloy steel specimen, and the maximum corrosion rate was obtained by the surface profilometry after the mockup test. The corrosion rates increased with the rise of the leakage rate of the primary water containing boric acid, and the presence of oxygen dissolved in the primary water also accelerated the corrosion. From the specimen surface analysis, it was found that typical flow-accelerated corrosion and jet-impingement occurred under two-phase fluid of water droplet and steam environment. The maximum corrosion rate was determined as 5.97 mm/year at the leakage rate of 20 cc/min of the primary water with a saturated content of oxygen within the range of experimental condition of this work.

• Journal of Ocean Engineering and Technology
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• v.14 no.1
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• pp.44-51
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• 2000

Safety and integrity are required for reactor pressure vessels because they are operated in high temperature. There are single specimen method multiple specimen method and load ratio analysis method which used as evaluation of safety and integrity for reactor pressure vessels. In this study the fracture resistance curve(J-R curve) elastic-plastic fracture toughness($J_{IC}$) and material tearing modulus ($T_{mat}$) of SA 508 class 3 alloy steel used as reactor pressure vessel steel are measured and evaluated at room temperature 20$0^{\circ}C$ and 30$0^{\circ}C$ according to unloading compliance method and load ration analysis method. And then the comparison with experimental $J_{IC}$ and theoretical$J_{IC}$ by local fracture strain is managed.

• Corrosion Science and Technology
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• v.18 no.5
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• pp.190-195
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• 2019

SA508 low-alloy steel for a reactor vessel was exposed to primary water in a pressurized water reactor (PWR) plant because the cladding layer of type 309 stainless steel for the RPV was removed, due to an accident in which the detachment of the thermal sleeve occurred. The major advantage of the electrochemical deposition (ECD) Ni plating technique is that the reactor pressure vessel can be repaired without significant thermal effects, and Ni has solid corrosion resistance that can withstand boric acid. The corrosion rate assessment of the damaged part was performed, and its trend was analyzed. Essential variables of the Ni plating for repair of the damaged part were derived. These conditions are applicable variables for the repair plating device, and have been carefully adjusted using the repair plating device. The process for establishing ASME technical standards called Code Case N-840 is described. The process of developing Ni-plating devices, and the electroplating procedure specification (EPS) are described.

• Nuclear Engineering and Technology
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• v.36 no.3
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• pp.229-236
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• 2004

Radiation hardening is a multiscale phenomenon involving various processes over a wide range of time and length. We present a multiscale model for estimating the amount of radiation hardening in pressure vessel steel in the environment of a light water reactor. The model comprises two main parts: molecular dynamics (MD) simulation and a point defect cluster (PDC) model. The MD simulation was used to investigate the primary damage caused by displacement cascades. The PDC model mathematically formulates interactions between point defects and their clusters, which explains the evolution of microstructures. We then used a dislocation barrier model to calculate the hardening due to the PDCs. The key input for this multiscale model is a neutron spectrum at the inner surface of reactor pressure vessel steel of the Younggwang Nuclear Power Plant No.5. A combined calculation from the MD simulation and the PDC model provides a convenient tool for estimating the amount of radiation hardening.

• Journal of Ocean Engineering and Technology
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• v.9 no.1
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• pp.111-119
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• 1995

The elastic-plastic fracture toughness($J_{IC}$) and fracture resistance (J-R curve) of SA508-3 alloy steel used for nuclear reactor pressure vessel are investigated by using CT-type specimens. Fracture toughness tests are conducted by unloading compliance method and multiple specimen method at room temperature, -2$0^{\circ}C$ and 20$0^{\circ}C$. The apparent negative crack growth phenomenon which usually arises in partial unloading compliance test is well known. The negative crack growth phenomenon in determining J sub(IC) or J-R cure from partial unloading compliance experiments may be eliminated by the offset technique. In this study, the evaluation of $J_{IC}$ multiple specimen method recommended by the JSME gives the most reliable results by using half-size CT(similar-type) specimens.

• Journal of Welding and Joining
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• v.1 no.2
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• pp.45-52
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• 1983

Many pressure vessels for the hot H$\sub$2//H$\sub$2/S service are made of 2+1/4Cr-1Mo steel with austenitic stainless steel overlay to combat agressive corrosion due to hydrogen sulfide. Hydrogen dissolves in to materials during operation, and sometimes gives rise to unfore-seeable damages. Appropriate precautions must, therefore, be taken to avoid the hydrogen induced damages in the design, fabrication and operation stage of such reactor vessels. Recently, hydrogeninduced cracking (or Disbonding) was found at the interface between base metal and stainless weld overlay of a desulfurizing reactor. Since the stainless steel overlay weld metal is subjected to thermal and internal-pressure loads in reactor operation, it is desirable for the overlay weld metal to have high strength and ductility from the stand point of structural safety. In section III of ASME Boiler and Pressure Vessel Code, Post-Weld Heat Treatment(PWHT) of more than one hour per inch at over 1100.deg. F(593.deg. C) is required for the weld joints of low alloy pressure vessel steels. This heat treatment to relieve stresses in the welded joint during construction of the pressure vessel is considered to cause sensitization of the overlay weld metal. The present study was carried out to make clear the diffusion of carbon migration by PWHT in dissimilar metal welded joint. The main conclusion reached from this study are as follows: 1) The theoretical analysis for diffusion of carbon in stainless steel overlay weld metal does not agree with Fick's 2nd law but the general law of molecular diffusion phenomenon by thermodynamic chemical potential. 2) In the stainless steel overlay welded joint, the PWHT at 720.deg. C for 10 hours causes a diffusion of carbon atoms from ferritic steel into austenitic steel according to the theoretical analysis for carbon migration and its experiment. 3) In case of PWHT at 720.deg. C for 10 hours, the micro-hardness of stainless steel weld metal in bonded zone increase very highly in the carburized layer with remarkable hardening than that of weld metal.