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Session Chair: Hasan Charkas, Electric Power Research Institute-EPRI, United States of America
Effect of Ni interlayer on Diffusion-Bond Joint Efficiency of Austenitic Fe-base Alloys
Ji-Hwan Cha, Sung Hwan Kim, Changheui Jang
Micro-channel heat exchangers with high heat transfer area such as the printed circuit heat exchanger (PCHE) are likely to be adopted to next generation nuclear reactors such as Small Modular Reactor (SMR) and Sodium-cooled Fast Reactor (SFR) to ensure high thermal efficiencies and compact size. Diffusion bonding is a key technology for fabrication of such heat exchangers. Meanwhile, cases of premature failure along the bond-line of diffusion-bonded specimens have been reported during tensile testing, especially at higher temperatures above 550 oC. This was attributed to formation of precipitates along the bonding interface during diffusion bonding process, resulting in a planar bond-line. In this study, Ni-foil interlayer was inserted into joint region in order to remove precipitates and induce grain boundary migration at the bond line to obtain better bond quality. Furthermore, application of post bond heat treatments on the diffusion-bonded specimens resulted in improved tensile properties and uniformity of Ni content across the bonding interface. The mechanical integrity of the diffusion-bond joints was evaluated by tensile testing at room temperature, 550 oC and 650 oC. Moreover, the mechanical integrity of the diffusion-bond joints will be discussed in view of the microstructural features at the joint region.
Development of thermal annealing parameters for the long-term operation of WWER 440 reactor internals assurance
Ivana Eliasova1, Radim Kopriva1, Petra Klatovska1, Ales Materna2
1UJV Rez; 2Czech Technical University
The reactor pressure vessel internals are not monitored by surveillance program. The degree of material degradation reactor pressure vessel internals is estimated on literature data from tests of similar materials, or from predictive relationships reported in literature, or calculation procedures and standards. The materials of reactor pressure vessel internals are exposed to significantly more intense neutron flux than the reactor pressure vessel. The thermal annealing can be one of the possible solutions to re-establish initial mechanical properties of reactor pressure vessel internals and thus can contribute to long term operation of nuclear power plants. This paper includes results obtained within the project TH02020565 “Assurance of Safe and Long-Term Operation of Nuclear Pressure Vessel Internals” which is realized by ÚJV Řež, a. s. in cooperation with Czech Technical University in Prague, FNSPE, in the period from 2017 to 2020 with the support of Technology Agency of the Czech Republic.
PWSCC Initiation of Thermally-Aged or Cold-Rolled Alloy 182 Welds
Dayu Fajrul Falaakh, Jae Phil Park, Chi Bum Bahn
Pusan National University
In this study, we investigated the effects of long-term thermal aging and cold work on the PWSCC initiation of Alloy 182. The Alloy 182 bulk specimen was fabricated by the weld deposit method on the 316 L stainless steel plate. A loop type testing facility simulating the primary water conditions of PWRs was used to carry out the PWSCC initiation test. The PWSCC initiation tests were done with the U-bend specimens treated under the following four conditions: 1) as-welded, 2) 15-year equivalent thermal aging, 3) 30-year equivalent thermal aging and 4) cold-rolled. The thermal aging conditions corresponding to 15 and 30 years of operation in typical NPPs at 320℃ were simulated by performing heat treatments for 1713 and 3427 hours in the 400°C Ar furnace, respectively. The PWSCC initiation testing up to ~10,000 hrs revealed that the thermal aging decreases the susceptibility of PWSCC initiation, and the cold work increases the susceptibility as expected. The weld residual stress appears to play a key role in this behavior. Based on the experimental data, a two-parameter Weibull distribution was adopted to create a probabilistic model for the prediction of the PWSCC initiation for Alloy 182. The Weibull parameters were estimated using MLE method and it was found that both the Weibull scale and shape parameters are affected by the thermal aging and cold working conditions.
Role of grain boundary carbides in PWSCC of stainless steels - Microstructural changes in 347H stainless steel
1Khalifa University of Science and Technology; 2Nuclear Technology Center; 3Korea Advanced Institute of Science and Technology
The stress corrosion cracking (SCC) phenomenon of stainless steels and Ni-base alloys in the primary water (PW) of pressurized water reactors (PWRs) is one of the issues threatening the safety of nuclear power plant. It has been well known that the PWSCC of stainless steels in the primary water environment is suppressed by the sensitization heat treatment. However, it is still unclear how the SCC is suppressed by the microstructural changes after the sensitization treatment. In order to identify the main role of carbides precipitated by the treatment, a project plan has developed. Stage 1) Heat-treatments of 347H (Nb stabilized) stainless steel and Stage 2) PWSCC tests using samples with and without grain boundary carbides. In 347H stainless steel Cr depletion does not happen after the sensitization treatment since NbC is formed instead of Cr carbides. Therefore, it is expected that heat-treated samples will show the role of grain boundary carbides without Cr-depleted zone on the PWSCC. In this paper, the microstructural changes in 347H stainless steel by heat-treatments (results from Stage 1) will be presented. 347H stainless steel was first solution-annealed and then heat-treated at different temperature and time conditions. Heat-treated samples were characterized using SEM (scanning electron microscope) and TEM (transmission electron microscope). The microstructural changes by the heat-treatments are discussed in the aspects of the effect of temperature and exposure time.
Reducing the capital cost of nuclear power plants using seismic isolation
Kaivalya Lal1, Sai Sharath Parsi1, Hasan Charkas2, Koroush Shirvan3, Michael Cohen4, Paul Kirchman5, Benjamin Kosbab6, Andrew Whittaker1
1Univeristy at Buffalo; 2Electric Power Research Institute-EPRI; 3Massachusetts Institute of Technology; 4TerraPower LLC; 5X-energy; 6Simpson, Gumpertz and Heger
Seismic isolation is an effective means to reduce seismic demands on building framing and equipment in nuclear power plants, thereby increasing safety (and reducing risk). No modern cost data exists to aid accurately quantify the possible savings. The Electric Power Research Institute (EPRI) and the Advanced Research Projects Agency – Energy (ARPA-E) are currently funding projects to characterize possible reductions in overnight capital cost of new build plants, with a focus on the financial impact of the seismic load case. Two strategies for seismic protection are being looked at: 1) base isolation of reactor buildings, addressed by the EPRI-funded study and 2) equipment-based isolation, addressed by the ARPA-E MEITNER project.
Two generic reactor buildings were designed, one for a molten chloride fast reactor (MCFR) and the other for a high-temperature gas reactor (HTGR). Each building was populated with three pieces of equipment: a reactor vessel, a steam generator and a housing for a control rod drive mechanism (CRDM). The two reactor buildings were assumed to be sited at the Idaho National Laboratory, in Idaho Falls, ID. Response-history analysis was performed to generate datasets on equipment weight and lateral accelerations (for design of internals) as a function of incremented levels of earthquake shaking. Base isolating the two buildings resulted in reduced thicknesses of vessel walls and lateral accelerations of vessel internals and CRDM housings. Analysis results were used to draft a questionnaire, to enable industry experts to identify the cost impacts of the seismic load case. The questionnaire sought to characterize the cost of analysis, design, qualification, and fabrication of equipment for incremented levels of earthquake shaking. The questionnaire was transmitted to consultants, equipment fabricators and suppliers, nuclear steam system suppliers, and operators of commercial reactors.