Development of Repair Technologies for J-Groove Welds of Drain and Instrument Nozzle in Steam Generator

SG drain & instrument nozzles and their welds fabricated with alloy 600 and alloy 82/182 is susceptible to Primary Water Stress Corrosion Cracking (PWSCC). In Korea, the cracks due to PWSCC were discovered in the drain nozzle of Yongkwang units 3 & 4. Doosan has developed a system for steam generator to repair damaged drain nozzle & welds and to prevent further damage on the instrument nozzle & welds. The repair system consists of machining, welding equipment and installation tool for this equipment. The machining equipment is used to remove the nozzle and J-groove welds. The process is called mechanical machining and the main equipment is installed on steam generator’s outer wall. The welding equipment is designed for the machined J-groove welds and overlay. The auto welding equipment consists of welding head, controller, monitoring tool and Gas Tungsten Arc Welding (GTAW) power supply. Doosan has developed remote welding process using the monitoring tool. The installation tool consists of automatic installment tool for instrument nozzle and manual installment tool for drain nozzle. Doosan successfully completed a mockup test and field application for Yongkwang unit 3.

Construction Aspects of the Doel 1 and 2 SGR Projects

The Doel 1 and 2 PWR Nuclear Power Stations are the oldest commercially operating units in Belgium and the last to replace their two Steam Generators. The Doel 2 Steam Generators were replaced in 2004 and those of Doel 1 will be replaced late 2009. The replacement poses a particular challenge as these are the only stations in Belgium requiring the creation of primary and secondary containment opening for the SG exchange operation. Other construction challenges result from the a-typical SG support configuration which dates from the period well before the more or less standardized support configuration as used for later PWR units. The current paper discusses the construction approaches selected to facilitate the exchange operation and to minimize the outage duration and radiation worker exposure. The main particularities of the construction effort concern the secondary containment opening and closing using a structural formwork assembly, the use of containment platforms hanging inside the primary containment allowing for parallel primary and secondary containment reconstruction and the de-activation of some of the primary coolant piping and SG restraints following the licensing acceptance of the Leak Before Break concept for the primary piping. The specific construction options that made the Doel 2 replacement a success will be presented in this paper.

Modeling of Sodium Two-Phase Flow With the TRACE Code

In the framework of PSI’s FAST code system, the TRACE thermal-hydraulics code is being extended for representation of sodium two-phase flow. As the currently available version (v.5) is limited to the simulation of only single-phase sodium flow, its applicability range is not enough to study the behavior of a Sodium-cooled Fast Reactor (SFR) during a transient in which boiling is anticipated. The work reported here concerns the extension of the two-fluid models, which are available in TRACE for steam-water, to sodium two-phase flow simulation. The conventional correlations for ordinary gas-liquid flows are used as basis, with optional correlations specific to liquid metal when necessary. A number of new models for representation of the constitutive equations specific to sodium, with a particular emphasis on the interfacial transfer mechanisms, have been implemented and compared with the original closure models. As a first application, the extended TRACE code has been used to model experiments that simulate a loss-of-flow (LOF) accident in a SFR. The comparison of the computed results, with both the experimental data and SIMMER-III code predictions, has enabled validation of the capability of the modified TRACE code to predict sodium boiling onset, flow regimes, dryout, flow reversal, etc. The performed study is a first-of-a-kind application of the TRACE code to two-phase sodium flow. Other integral experiments are planned to be simulated to further develop and validate the two-phase sodium flow methodology.

Ultimate Heat Sink Sizing Requirements on a Month-to-Month Basis

The Ultimate Heat Sink (UHS) of a nuclear power plant is a complex cooling water system which serves the plant during normal and accident conditions. For some next generation nuclear plants, the UHS sizing is a major design and licensing analysis task. The analysis involves detailed modeling of the transient heat loads and the selection of worst-case meteorological data for the plant site. The UHS sizing requirements for a representative next generation nuclear power plant are evaluated on a month-to-month basis. This paper assesses the UHS water requirement for each month of year. The UHS analysis for a representative next generation nuclear plant with mechanical draft cooling towers and a water basin is used to determine the maximum evaporation of the basin for the worst-case meteorological data on a month-to-month basis. To size the cooling tower basin, automated methods have been developed which determine the highest evaporative losses from the basin and highest basin temperature over a 30-day design basis accident period. This paper also evaluates the month-to-month basin temperature changes. This assessment is done for a representative next generation nuclear power plant and considers the monthly historical meteorological data over 45 years. The result of this assessment of monthly UHS water requirement is of interest in assessing the margin in the UHS design. This monthly assessment is also useful in demonstrating that the automated methods used to establish the limiting 30-day meteorological condition are indeed accurate. In addition, these results may be useful in helping to plan plant maintenance activities.

Experimental Measurements of Eddy Current Signal From SG Tubes of Fast Breeder Reactor Covered by a Thin Sodium Layer Using a SG Mock-Up

In Fast Breeder Reactors (FBR) which are sodium cooled, the steam generator (SG) heat exchanger tubes separate the low pressure sodium flowing in the SG vessel with the high pressure water-steam in tubes. During In-Service Inspection (ISI), sodium is first drained and then SG tubes are cooled down to the room temperature. After sodium draining, due to the high temperature (more than 500 °C), sodium adheres to SG tubes and structures around (SG support plates, welds) in a thin layer, filling eventually the gaps between SG support plates and tubes. During ISI, SG tubes are inspected for cracks and corrosions using differential eddy currents (EC) probes. Due to the high electrical conductivity of sodium adhering to the outer SG tube surface, the eddy current testing (ECT) signal modifies, in accord with sodium layer thickness or sodium deposits located on the outer SG tube surface. The sodium wetting properties depends on several factors as: material surface, temperature and sodium wetting time. The effect of sodium adhering to the outer SG tube on ECT signals were measured using a small mock-up tank (2 m high and 0.7 m in diameter) in which were introduced two SG tubes similar with the ones used in the Monju FBR (one tube is ferromagnetic and made of 2.25Cr–1Mo alloy, while the other one is made of SUS321 and is austenitic). Defects, SG support plates (on both helical and straight part of the tube) and welds were added to tubes and the ECT signal was measured before and after sodium draining. Variations in the sodium layer thickness and consequently its effect on ECT signals were measured by filling and draining the tank three times in order to recreate each time new layers of sodium. The paper describes the experimental conditions and the ECT results for both types of SG tubes by comparing the defects, SG support plates and weld signals before and after draining of sodium. Additionally, sodium structures were examined visually using a VideoScope camera, confirming the recorded ECT signals. The paper also presents details about sodium layer thickness measurements in several parts of SG tubes (near defect, SP, weld, bend, helical tube, straight tube) by scratching and collecting the sodium on a small area of 20mm×20mm. The volume of sodium drops is also estimated. The measurement results showed that there are significant differences in the sodium layer thickness depending on the SG tube material.

Investigation of the Beltline Welding Seam and Base Metal of the Greifswald WWER-440 Unit 1 Reactor Pressure Vessel

The investigation of reactor pressure vessel (RPV) materials from decommissioned NPPs offers the unique opportunity to scrutinize the irradiation behaviour under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterisation. The paper describes the investigation of trepans taken from the decommissioned WWER-440 first generation RPVs of the Greifswald NPP. Those RPVs represent different material conditions such as irradiated (I), irradiated and recovery annealed (IA) and irradiated, recovery annealed and re-irradiated (IAI). The working program is focussed on the characterisation of the RPV steels (base and weld metal) through the RPV wall. The key part of the testing is aimed at the determination of the reference temperature T0 following the ASTM Test Standard E1921-05 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step the trepans taken from the RPV Greifswald Unit 1 containing the X-butt multilayer submerged welding seam and from base metal ring 0.3.1 both located in the beltline region were investigated. Unit 1 represents the IAI condition. It is shown that the Master Curve approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T0 varies through the thickness of the welding seam. The lowest T0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. Beyond the welding root T0 shows a wavelike behaviour. The highest T0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on sub size Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do not represent the most conservative condition. Nevertheless, the Charpy transition temperature TT41J estimated with results of sub size specimens after the recovery annealing was confirmed by the testing of standard Charpy V-notch specimens. The evaluated Charpy-V TT41J shows a better accordance with the irradiation fluence along the wall thickness than the Master Curve reference temperature T0. The evaluated T0 from the trepan of base metal ring 0.3.1 varies through the thickness of the RPV wall. T0 increases from −120°C at the inner surface to −104°C at a distance of 33 mm from it and again to −115°C at the outer RPV wall. The KJc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during 2 campaigns operation can be assumed to be low for the weld and base metal.

Characteristics of the New Embrittlement Correlation Method for the Japanese Reactor Pressure Vessel Steels

Neutron irradiation embrittlement of reactor pressure vessel steels is an important ageing issue for the long term operation of light water reactors. A new embrittlement correlation method was developed by CRIEPI and the Japanese electric utilities in 2007. This method is primarily based on the fundamental understandings on the embrittlement mechanisms: i.e. microstructural changes were modeled by the mathematical form of rate equations, and the predicted microstructural changes were further correlated with the mechanical property changes in transition temperature region. The coefficients of the rate equations were optimized using the Japanese surveillance data of RPV embrittlement. This method was adopted as the revision of the Japanese code, JEAC 4201–2007, in 2007. In this paper, after a brief explanation on the new correlation method, the predictions of the new method will be investigated through comparisons with the previous correlation, JEAC4201–2004, and the US surveillance data in order to identify the characteristics of the new method.

The Influence of Chrome in the Oxidation Process of Steels in Flowing LBE

The role of the alloying element has been analyzed during the oxidation process of stainless steels in the flowing lead bismuth eutectic (LBE) environment. The growth of the protective oxide film of steels has been studied at a mesoscopic scale. The influence of chrome has been studied using the developed stochastic cellular automaton mesoscopic oxidation model, considering the formation of the chrome oxide and magnetite. During the oxidation process, the scale removal effect has been taken into account as well.

Study of the Influence of Surface Finishing on the Corrosion/Oxidation Behaviour of Stainless Steels in Stagnant Lead Bismuth Eutectic

Lead-bismuth eutectic (LBE) is of interest as a coolant in the design of fast reactors and also as both a coolant and a spallation target in proposed transmutation schemes for radioactive waste. However, liquid metal corrosion to the structural materials can proceed via various processes: species dissolution, formation of the inter-metallic compounds at the steels/liquid metal interface... It is known that the formation of an oxide scale on a steel surface can protect it dissolution attack by liquid LBE. The variables that play an important role on the feasibility of the formation of an oxide layer and on its protective characteristics for steels in contact with LBE are oxygen content, temperature, materials composition and evolution with time, but also surface state of steels prior to testing or weld joints, being these parameters not very widely studied. For the study of the influence of the surface finishing, specimens with different surface states were prepared (as-received, grinded, grinded and polished and electrolitically polished). These treatments gave to the materials a different degree of cold working, the higher for the mechanised samples and the lowest for the electrolitically polished. Tests were carried out at 500 and 550°C from 100 to 2000 hours with two different H2/H2O ratios: 3 and 0.03. The general conclusion is that the effect of surface finishing on the corrosion/protection processes is not significant for the conditions tested.

Designing Mass Exchangers for Control of Oxygen Content in Pb-Bi (Pb) Coolants in Various Research Facilities

The solid-phase method designed by the State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering (SSC RF - IPPE) for the oxygen thermodynamic activity control [1, 2] implies the use of oxygen source, i.e. dissolvable lead oxide spheroids hold within limited section of the circuit (reaction vessel of mass exchanger) connected by the pipeline to the main circuit. PbO spheroids being in contact with flowing lead-bismuth (lead) are dissolved with generation of oxygen, which is transported throughout the circuit by the coolant flow. Technological implementation of solid phase control method is based on the use of specially designed mass exchangers (MX) being an important component of heavy liquid metal coolant technology.