Effects of Delta - Ferrite Content on the Integrity of Reactor Vessel Cladding

As the design life of new nuclear power plant increases, the austenitic stainless cladding integrity of reactor vessel becomes one of the new concerns. Since 1970’s, there have been some specific recommendations on delta ferrite content of austenitic cladding of reactor vessels and welds. It has been known that the delta ferrite is beneficial for reducing micro-fissure in welds, though the high delta ferrite content increases the probability of embrittlment of welds. In this study, the mechanical and microstructural properties of austenitic weld metals with the limit values of the recommended range (5 ∼ 18 FN) of the delta ferrite control on low alloy steels were characterized by using bending test and scanning electron microscopy. The base metal was ASME Code Sec. II specification SA 508 Gr. 3 Cl. 1 plate and weld materials were EQ308L and EQ309L strips. Four kinds of cladding were deposited with submerged arc welding process on SA508 cl.3 plates. The bending tests were performed through ASME code Sec. IX and the microstructure of fractured surfaces was analyzed by scanning electron microscopy (SEM). In bending tests, there were no fractures except the highest delta ferrite content specimens (28FN). From the SEM observation of fractured surfaces, cracks initiated from the interface between austenite and ferrites phases in the cladding layer and propagated through the continuous interfaces between two phases. For specimens without continuous interfaces of two phases, though the cracks were observed in the interface of phases, the propagation of cracks was not observed. From the test results, continuous interfaces between austenite matrix and ferrite phase provide the path for crack propagation. And the delta ferrite content affects the integrity of cladding of reactor vessel.

Rational Design of Novel Materials From Polymer Microrheology

The visco-elastic properties of entangled polymer liquids arise from molecular-scale topological interactions and stochastic fluctuations under flow. Here, the evolutions of individual entangled polymers were observed in rheologically relevant shear flow histories. We uncover a high degree of molecular individualism and broad conformational distributions resulting from incessant stretch-collapse cycles. The data and insights of the present study may lead to improved molecular-level models and constitutive equations. These tools, in turn, may enable the rational design of novel materials with properties tailored to accomplish specific tasks such as high-pressure vessels and piping with greater safety margins and cost-effectiveness.

Microstructural Changes Related to Through-Wall Attenuation of Neutron Irradiation Embrittlement

The through-wall attenuation of neutron fluence of reactor pressure vessel (RPV) steels is often expressed using an exponential decay function based on some estimate of displacements per atom (dpa). In order to verify this function, an irradiation project was performed in which 18 layers of Charpy specimens and one central temperatue control layer were stacked in a block to simulate a 190 mm thick RPV wall. Three western-type RPV steels (medium and low copper plates and a high copper Linde 80 flux weld) were irradiated in this project. Mechanical property tests of these materials have been performed under a consortium of EPRI, CRIEPI, NRI-Rez and ATI Consulting to fully characterize the mechanical properties in terms of Charpy transition temperature and upper-shelf energy, as well as reference fracture toughness using the Master Curve. Some results have been reported at previous PVP conferences. In this paper, we report the results of microstructural characterization using three-dimensional atom probe tomography (APT) of the medium copper plate and the high copper weld metal. The microstructures obtained by APT reasonably explain the changes in mechanical properties of these materials, and the difference in the response of these materials to irradiation was also identified. The mixed effect of fluence/flux/spectrum is discussed from the microstructural point of view.

Fast Neutron Reduction Techniques for Plant Lifetime Extension

A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in beltline region of a reactor vessel where a reduced fracture resistance exists due to neutron irradiation. Such an event may produce a flaw or cause the propagation of a flaw postulated to exist near the inner vessel wall surface, thereby potentially affecting the integrity of the vessel. In this paper fast neutron flux reduction techniques were implemented to reduce the potential risk of PTS due to the neutron irradiation on the pressure vessel beltline region. And the RTPTS value for the end of life of the plant was projected using the fast neutron fluence obtained by neutron transport calculations according to the various core loading pattern and reduction program possible for the future cycles.

The Application of Optimized Commissioning Schedule

The targeted 24.5 months of commissioning time for the ongoing SKN unit #1 project in Korea is a tough challenge in view of saving at least 7 months from the previous 32 months for UCN unit #3 project that has been a prior version of OPR1000 and also the reference plant of this unit. Such commissioning time-saving has been made possible by introducing renovated technology, optimization of commissioning technique, and adoption of lessons-learned from the previous projects. Our experiences are discussed herein with respect to the scheduling of commissioning operation of previous NPP projects on a milestone basis, including trouble areas that could cause schedule delay and our approaches adopted for solving the key issues. In addition, this paper will introduce a summarized 24 month commissioning schedule which has been optimized to the most efficient level by reflecting the prior project experience, and applying the new techniques and methods.

Mechanical Properties of SA508 Gr.4N Model Alloys as a High Strength RPV Steel

Demands of RPV materials with higher strength and toughness are rising to increase the power capacity and the operation life of nuclear power plants. The ASME SA508 Gr.4N specification can give a superior toughness and strength to the commercial low alloy steels such as SA508 Gr.3. However, the SA508-Gr.4N steels have not yet been used commercially due to a lack of information of the productivity and the age related properties. While the irradiation embrittlement studies are going-on, the current paper focused on the effects of alloying elements such as Ni, Cr and Mo on the fracture mechanical properties of the SA508 Gr.4N low alloy steels. Various model alloys were fabricated by changing the contents of alloying elements based on the composition range of the ASME specification. Tensile properties, Charpy impact toughness and fracture toughness of the model alloys were evaluated and those properties were discussed with the microstructural characteristics of each alloy. The strengths of the alloys were increased with increase of the Ni and Mo contents while there was no remarkable change of the yield strength with the Cr addition. The Charpy impact and fracture toughness were considerably improved with the increase of Ni, Cr contents. The Mo addition did not change the toughness properties significantly. The Cr contents were more effective on the fracture toughness through changing the carbides precipitation characteristics and the Ni contents were effective on the Charpy impact toughness through changing the effective grain size.

Regulatory Experience of Leak-Before-Break (LBB) Technology to the High Energy Piping Systems in Korea

The application of the leak-before-break (LBB) technology to the newly constructed pressurized water reactors (PWRs) has been approved for the several high energy piping systems inside containment in Korea. The main purpose of the LBB application for these systems at the design stage is the removal of the dynamic effects associated with the postulated double-ended guillotine break (DEGB) from design basis loads, as well as to the elimination of the pipe whip restraints and jet impingement barriers so as to increase the access the inspections. LBB technology is based on the low probability of pipe ruptures in the candidate piping systems using fracture mechanics and the insights from the state-of-the-art technology including operating experience. The procedures for LBB application is fundamentally based on the Unite States Nuclear Regulatory Commission (US NRC) requirements as detailed in the standard review plan (SRP) 3.6.3. However, a number of the additional requirements and issues are not specified in the review procedure during regulatory review were imposed and addressed during the review process. The regulatory review is focused on the confirmation on the methods for the elements in the screening criteria and several technical concerns on the determination of material properties, the validation of crack evaluation methods and leak rate estimation in the LBB evaluation considering the adequate margin. Although the application of the LBB has been approved by the safety authority for some high energy systems, the validation of LBB is continuously maintained in consideration of operating experience. In this paper, the regulatory positions for LBB application are described for the areas of screening criteria, leak rate estimation including the capability of leak detection system, material properties, load combination, crack stability methods, and margins in the crack stability evaluations. The issues encountered during the regulatory review such as the dynamic fracture test to consider the dynamic strain aging (DSA) of carbon and low alloy steel, thermal stratification and striping in the pressurizer surge line, water/steam hammer in main steam lines, and estimation of the crack opening area at the pipe-to-nozzle interface considering the asymmetry are also introduced. In addition, several regulatory actions to improve the reliability in the capability of leak detection systems and to clarify the screening criteria such as the corrosion resistance is provided.

Characterizing the Stability of Carbon Nanotube Enhanced Water as a Heat Transfer Nanofluid

Heat dissipation is a major challenge for many technologies. Possible solutions include thermal energy transfer via coolant fluid to a phase change material (PCM), with higher thermal conductivity a design goal. In recent years, heat transfer nanofluids (fluids with suspended nanoparticles) have received attention based on their potential for improving thermal conductivity. Carbon nanotubes (CNTs) are an attractive additive due to their enhanced thermal conductivity and ability to remain suspended over long times. However, characterizing their potential is difficult due to the many design variables and the need for repeated thermal conductivity tests for comparison. Since thermal conductivity enhancement is dependent on a dispersed nanotube network, the electrical conductivity of CNTs can be exploited to monitor the stability of such nanofluids, as such testing is quick and simple. The aim of this research was to evaluate electrical conductivity testing as a means to monitor stability of CNT-enhanced distilled water as a PCM, with varying CNT size, type, and concentration; and various other processing variables. The prepared nanofluids were tested after repeated phase change cycles. Results indicate that electrical conductivity testing is a practical means of monitoring the nanofluid stability, and CNT-based nanofluids show both promise and limitations as a PCM.

Verification of the Output Signal Change of Control Rod Position Indicator at the End of Its Life Time

The system-integrated reactor currently under development at the Korea Atomic Energy Research Institute is being designed with a soluble boron-free operation and the use of a nuclear heating for the reactor start-up. The Control Element Drive Mechanism (CEDM) for the integral reactor is designed to raise and lower the control rod in steps of 2mm in order to satisfy the design features of the integral reactor which are the soluble boron-free operation and the use of a nuclear heating for the reactor start-up. The actual position of the control rod could be achieved to sense the magnet connected to the control rod by the position indicator around the upper pressure housing of CEDM. It is sufficient that the actual position information of the control rod at a 20mm interval from the position indicator is used for the core safety analysis. As the magnet moves upward along the position indicator assembly from the bottom to the top in the upper pressure housing, the output voltage increases linearly step-wise at 0.2VDC increments. Between every step, there are transient areas which occur by a contact closing of three reed switches, which is the 2-3-2 contact closing sequence. In this paper the output voltage signal corresponding to the position of the control rod is expected as the 2-1-2 contact closing sequence due to the aging of the magnet at the end of its life time. We verify that the output signal change at the end of the life time of the position indicator could not affect the operational performance of the position indicator. Also it is found that the voltage divider network for the output signal of the position indicator is designed to be robust and reliable during the aging process.

The IMS Development Plan in Construction Phase for Nuclear Power Plants

The implementation of advanced information technology for nuclear power plant construction industry in the Republic of Korea is a challenging task to improve the competitive edge for APR1400 across the global nuclear business market. This paper will briefly describe the phases of development of an information management system (hereinafter IMS) in nuclear power construction during the last 40 years in Korea. Additional development task which strengthen the IMS capability will be described from our experience. It will analyze the IMS development and implementation stages for Korean nuclear power plant construction projects such as YGN 3&4, UCN 3&4, YGN 5&6 and UCN 5&6, and the successful application case of the UCN 5&6 Radwaste Building in which 3D CAD technology was implemented for the first time. By reviewing lessons learned, this paper will define the Information Technology advancements resulting the reduction of project costs and construction schedule both by project execution procedures and IT systems including 3D CAD application. The future plan will include integrating project management systems based on data-centric approach and handover strategy for better O&M phase through configuration management technology. In this report, we address the functions to be developed and added in the new IMS.