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DOE OSTI2021

Material Interactions in Severe Accidents – Benchmarking the MELCOR V2.2 Eutectics Model for a BWR-3 MARK-I Station Blackout: Part I – Single Case Analysis

Here in this analysis, the two material interaction models available in the MELCOR code are benchmarked for a severe accident at a BWR under representative Fukushima Daiichi boundary conditions. This part of the benchmark investigates the impact of each material interaction model on accident progression through a detailed single case analysis. It is found that the eutectics model simulation exhibits more rapid accident progression for the duration of the accident. The slower accident progression exhibited by the interactive materials model simulation, however, allows for a greater degree of core material oxidation and hydrogen generation to occur, as well as elevated core temperatures during the ex-vessel accident phase. The eutectics model simulation exhibits more significant degradation of core components during the late in-vessel accident phase – more debris forms and relocates to the lower plenum before lower head failure. The larger debris bed observed in the eutectics model simulation also reaches higher temperatures, presenting a more significant thermal challenge to the lower head until its failure. At the end of the simulated accident scenario, however, core damage is comparable between both simulations due to significant core degradation that occurs during the ex-vessel phase in the interactive materials model simulation. A key difference between the two models’ performance is the maximum temperatures that can be reached in the core and therefore the maximum ΔT between any two components. When implementing the interactive materials model, users have the option to modify the liquefaction temperature of the ZrO 2 -interactive and UO 2 -interactive materials as a way to mimic early fuel rod failure due to material interactions. Through modification of the liquefaction of high melting point materials with significant mass, users may inadvertently limit maximum core temperatures for fuel, cladding, and debris components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI1996

Nuclear Safety [Vol. 37, No. 2, April-June 1996]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 97 The Nuclear Community and the Public: Cognitive and Cultural Influences on Thinking About Nuclear Risk, M. A. Meyer; 109 Twenty-Third Water Reactor Safety Information Meeting, D. A. Copinger; ACCIDENT ANALYSIS: 126 Analysis of a PWR LBLOCA Without SCRAM, Trevor N. Tyler, Rafael Macian-Juan and John H. Mahaffy; DESIGN FEATURES: 139 Vulnerability of Multiple-Barrier Systems, N. C. Lind; ENVIRONMENTAL EFFECTS: 149 A Study of Wet Catalytic Oxidation of Radioactive Spent Ion Exchange Resin by Hydrogen Peroxide, Xingchao Jian, Tianbao Wu, and Guichun Yun; 157 A Comparison Study and Resolution of Differences Between Emergency Response and Safety Analysis Codes Used at the Savannah River Site, A. A. Simpkins; OPERATING EXPERIENCES: 164 Reactor Shutdown Experience, Compiled by J. W. Cletcher; RECENT DEVELOPMENTS: 167 Reports, Standards, and Safety Guides, D. S. Queener; 172 Proposed Rule Changes as of Dec. 31,1995; ANNOUNCEMENTS: 178 American Nuclear Society 1997 Annual Meeting; 178 American Nuclear Society Nuclear Criticality and Safety Division Topical Meeting; 176 The Authors.

05 NUCLEAR FUELS↗

DOE OSTI2022

TEM Specimen Manufacturing to Support Analysis of IASCC Mechanisms in Irradiated Austenitic Steels – Light-Water– Relevant Materials

The present report describes the manufacturing of samples for advanced microstructure studies of mechanisms of irradiation-assisted stress corrosion cracking (IASCC) in the materials of light-water reactor (LWR) internal components. The samples were manufactured via FIB (focused ion beam milling) from 4-point-bend specimens subjected to oxidation and mechanical loading in a simulated LWR environment (primary water, PW). The corrosion tests were conducted at the University of Michigan (UM), and the tested specimens were transported to ORNL to perform FIB work. Finally, the manufactured FIB specimens were shipped back to the UM for microstructure characterization. Section 1 in the report provides a brief introduction; Section 2 describes materials, their in-service and irradiation history, and the geometry of specimens for IASCC testing. Section 3 documents typical work steps (specimen mapping) and reports the number of manufactured FIB objects. Section 4 briefly describes the preliminary results of microstructure analysis performed at the University of Michigan and highlights the key findings. The summary and conclusions section discusses future possible future steps and generalizes the obtained results, their unique character, and their importance for the nuclear industry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Elucidating the corrosion mechanism of Ni-based superalloys in the presence of uranium-containing chloride molten salt

The United States Department of Energy (DOE) is committed to the advancement of nuclear reactor technology through initiatives such as the Advanced Reactor Development Program (ARDP), in an effort to diversify the United States energy portfolio towards more sustainable energy options. The ARDP includes demonstration by industry partners of molten chloride fast reactors (MCFRs). Construction of MCFRs requires qualified nuclear structural materials. Unfortunately, there are no current materials that are fully qualified by the Nuclear Regulatory Commission for the construction of molten salt reactors, including MCFRs. Adapting current structural material qualifications requires expansion of our current knowledgebase on the property-performance relationships regarding corrosion performance. In this investigation, we assess microstructural changes in a Ni-based superalloy after exposure to a UCl3¬-containing chloride salt eutectic mixture through a correlated multi-modal approach combining several advanced characterization techniques, including scanning electron microscopy/focused ion beam (SEM/FIB) and transmission electron microscopy (TEM). SEM/FIB analysis will illustrate changes in elemental composition, microstructure, and isotopic information acquired from energy x-ray dispersive spectroscopy (EDS), electron backscatter diffraction (EBSD), and secondary ion mass spectroscopy (SIMS), respectively. This information will then aid in identifying localized regions to elucidate the corrosion mechanism with TEM through a combination of electron diffraction, electron energy loss spectroscopy (EELS), and additional EDS. The findings from this investigation will further expand our assessment of the corrosion performance of structural materials in molten salt chloride systems, aiding to developing fully qualified materials for construction of MCFRs.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Progress Report on Graphite-Salt Interaction Studies (FY21)

This report summarizes the activities performed in FY21 to investigate the interactions between nuclear graphite and molten salts, such as FLiNaK. First, building on last year’s achievements, the team improved the procedure for measurements on pressurized salt intrusion in nuclear graphite and performed new experiments at variable pressures with specimens of different sizes and shapes. Additionally, the team actively provided suggestions and comments for ASTM D8091, Standard Guide for Impregnation of Graphite with Molten Salts , which came up for periodic revision in 2021. Second, the team expanded its capabilities to further characterize salt-impregnated graphite by procuring and installing a laser-induced breakdown spectroscopy (LIBS) instrument capable of producing a 3D chemical composition of a surface and subsurface region of specimens. A methodology for accurately analyzing information was developed, and a manuscript was submitted for publication. The LIBS instrument sample cell is sealed in controlled inert atmosphere, which is an advantage for using humidity-sensitive samples and Be-contaminated samples. This aligns with the team’s effort to expand the capabilities of handling FLiBe and Becontaining materials in two newly installed glovebox units. Third, the team performed preliminary tribology tests for pebble graphite wear in contact with stainless steel in dry state and in molten salt at high temperature. These tests are needed to establish the baseline for the wear of pebble graphite by friction against metallic walls. The friction and wear of pebbles in the fluoride salt-cooled hightemperature reactor are expected to generate dust, which is a concern for other components’ properties and safe reactor operation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Characterization of advanced nuclear materials under extreme environments

The performance of materials under extreme environments poses important fundamental questions about the behavior of condensed matter under far-from-equilibrium conditions. These conditions create challenges in materials design, testing and evaluation. One important aspect of advancing nuclear power is the discovery and investigation of new classes of materials able to withstand the harsh environments in nuclear reactors. However, investigating and understanding the performance of these materials requires sophisticated tools and characterization techniques and skills. Furthermore, the goal of this special issue is to present recent research on the characterization of the response of candidate advanced nuclear materials to extreme environments.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Gamma-ray Spectroscopy in Low-Power Nuclear Research Reactors

Gamma-ray spectroscopy is an effective technique for radioactive material characterization, routine inventory verification, nuclear safeguards, health physics, and source search scenarios. Gamma-ray spectrometers typically cannot be operated in the immediate vicinity of nuclear reactors due to their high flux fields and their resulting inability to resolve individual pulses. Low-power reactor facilities offer the possibility to study reactor gamma-ray fields, a domain of experiments hitherto poorly explored. In this work, we present gamma-ray spectroscopy experiments performed with various detectors in two reactors: The EPFL zero-power research reactor CROCUS, and the neutron beam facility at the Ohio State University Research Reactor (OSURR). We employed inorganic scintillators (CeBr3), organic scintillators (trans-stilbene and organic glass), and high-purity germanium semiconductors (HPGe) to cover a range of typical—and new—instruments used in gamma-ray spectroscopy. The aim of this study is to provide a guideline for reactor users regarding detector performance, observed responses, and therefore available information in the reactor photon fields up to 2 MeV. The results indicate several future prospects, such as the online (at criticality) monitoring of fission products (like Xe, I, and La), dual-particle sensitive experiments, and code validation opportunities.

Pakari, Oskari V. (ORCID:0000000337048190)↗

DOE OSTI2021

Effect of thermal aging and irradiation on microstructure and crack growth response of Alloy 690

Nickel-base Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs reactors are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 690 during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni 2 Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 690 under accelerated thermal aging and irradiation conditions. The materials studied involved Alloy 690 from two sources: (i) Alloy 690 plate from an Alloy 152 dissimilar metal weld (DMW) joining the Alloy 690 plate to Alloy 533 LAS, aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service), and (ii) specimens neutron-irradiated in the BOR-60 reactor up to 40 dpa. For aged Alloy 690 specimens, hardness was found to increase with aging time, however, the microstructural characterization by means of synchrotron X-ray did not find evidence of LRO. The microstructural characterization of neutron-irradiated specimens by TEM found no evidence of LRO either. Testing in a primary water environment of Alloy 690 specimens aged to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. The SCC CGR response was also low. Overall, the two Alloy 690 heats investigated in this work, aged up to 60-year service equivalents or exposed to neutron irradiation up to 40 dpa, did not exhibit a deterioration in microstructure or performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Coated U3Si2 pellets with enhanced water and steam oxidation resistance

A method of forming a water resistant boundary on a fissile material for use in a water cooled nuclear reactor is described. The method comprises coating the fissile material, such as a pellet of U3Si2 and/or the grain boundaries, to a desired thickness with a suitable coating material, such as atomic layer deposition or a thermal spray process. The coating material may be any non-reactive material with a solubility at least as low as that of UO2. Exemplary coating materials include ZrSiO4, FeCrAl, Cr, Zr, Al—Cr, CrAl, ZrO2, CeO2, TiO2, SiO2, UO2, ZrB2, Na2O—B2O3—SiO2—Al2O3 glass, Al2O3, Cr2O3, carbon, and SiC, and combinations thereof. The water resistant layer may be overlayed with a burnable absorber layer, such as ZrB2 or B2O3—SiO2 glass.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Effect of Thermal Aging on Microstructure and Stress Corrosion Cracking Behavior of an Alloy 152 1st Layer Butter Weldment

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40- 80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is critical for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for different durations up to 75,000h (equivalent to 60 years of reactor service). The microstructural characterization by means of synchrotron X-ray conducted in small, 0.2 mm - step line scans in the high-deformation regions of the weld root – covering areas spanning from the weld heat affected zone (HAZ) in Alloy 690 to the weld and weld butter on LAS - did not show evidence of LRO in any of the three Alloy 152 heats aged at 370°C and 450°C to an equivalent of 60 years of service. However, the first weld butter layer has high levels of deformation and is highly susceptible to SCC even in its non-aged condition. Nanohardness testing confirmed the extreme hardening with aging (ΔHV ≅ 100) at two locations within this weldment. In absence of LRO, hardening is suspected to be due to thermally-induced Cr carbide precipitation and coarsening. Testing in a primary water environment of the 1st layer of Alloy 152 weld butter aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloys. Similarly, the SCC CGR response of the aged weld butter does not appear to show a deterioration in performance, however, the difficult-to-test weldment geometry may affect the test results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Investigation of interfacial strength in nacre-mimicking tungsten heavy alloys for nuclear fusion applications

Abstract Tungsten heavy alloys have been proposed as plasma facing material components in nuclear fusion reactors and require experimental investigation in their confirmation. For this purpose, a 90W–7Ni–3Fe alloy has been selected and microstructurally manipulated to present a multiphase brick-and-mortar structure of W-phase ‘bricks’ surrounded by a ductile ‘mortar’. This work draws inspiration from nature to artificially imitate the extraordinary combination of strength and stiffness exhibited by mollusks and produce a nacre-mimicking metal matrix composite capable of withstanding the extremely hostile environment of the reactor interior and maintaining structural integrity. The underlying mechanisms behind this integrity have been probed through high-resolution structural and chemical characterization techniques and have revealed chemically diffuse phase boundaries exhibiting unexpected lattice coherency. These features have been attributed to an increase in the energy required for interfacial decohesion in these systems and the simultaneous expression of high strength and toughness in tungsten heavy alloys.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Decarburization and Carburization Behavior of Grade 91 Steel in Sodium Environments

Two heats of G91 steel have been investigated in sodium at 550-650°C to understand its corrosion behaviour, microstructural evolution, and tensile properties. Sodium exposure tests of subsized tensile and coupon specimens of G91 base metal and weldment were conducted in Argonne’s forced convection sodium materials testing loops. Maximum exposure times of ~39,000 h at 550°C, ~37,000 h at 600°C, and ~20,000 h at 650°C have been achieved, respectively. Thermal aging experiments of G91 steel were conducted in parallel, and the data of the sodium-exposed specimens were compared with the thermal aging data of the same heat to isolate the thermal aging effect from the effect of sodium exposure. It was found that the tensile strength of G91 steel can be reduced by more than 50% after sodium exposures at 650°C. The effect of a sodium environment on tensile properties is largely dependent on the carburization/decarburization behavior of G91 steel exposed to sodium. The carbon concentrations in sodium in Argonne’s SMT-1 and SMT-2 loops were determined by the equilibration method. The carbon activity in sodium was established by equilibrating high-purity nickel in sodium, analyzing the material for carbon, and using the reported data on the carbon solubility for this material. The carbon activity results were combined with the carbon solubility data in sodium to establish the carbon concentration in sodium. The carbon concentrations in the SMT-1 and SMT-2 loop were estimated to be 1.23 and 0.25 ppm, respectively. With this initial success, we will continue to monitor the carbon contents in the two loops to correlate the microstructural and mechanical property changes with the sodium purity. Future experiments will also consider sodium exposure tests in controlled carbon activities. Online monitoring and control of carbon contents in sodium in nuclear reactors would also be of interest. Thermodynamic analysis was conducted to understand the carburization-decarburization behavior of G91 steel exposed to sodium in the SMT-1 and SMT-2 loops. The equilibrium carbon activity in the alloy was evaluated by the thermodynamics of M 23 C 6 carbides. The calculated carburization-decarburization boundary showed that G91 would undergo decarburization at 650°C in the SMT-1 loop and carburization at 600°C in the SMT-2 loop, which is consistent with our findings from the sodium exposure tests. Future work will include thermodynamic analysis of the carbon concentration – activity relationship including the effect of carbide-forming elements, vanadium and niobium in G91 steel. Kinetics of carbon diffusion and the effect on microstructure and mechanical performance under various service conditions will also be investigated. The ultimate goal is to understand the mechanisms and the kinetics of carbon transfer in the sodium-steel system and to provide a basis for predicting the effect of long-term decarburization-carburization process on the integrity of reactor components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

In-service Oxidation and IASCC in High Fluence Baffle-Former Bolts Retrieved from a Westinghouse PWR

Baffle-former bolts (BFBs) are subjected to significant mechanical stress and neutron irradiation from the reactor core during the plant operation. Over the long operation period, these conditions lead to potential degradation and reduced load-carrying capacity of the bolts, and life extension of existing pressurized water reactors (PWRs) would only cause more damage to the bolt material. To this end, the Light Water Reactor Sustainability (LWRS) Program Materials Research Pathway (MRP) successfully harvested two high fluence BFBs from a Westinghouse two-loop downflow type PWR in 2016. In the same year, the two BFBs were received at the Westinghouse Churchill facility for inspection and specimen fabrication. The fabrication was completed in 2017 with specimens shipped to Oak Ridge National Laboratory (ORNL) for further testing. The objective of this project is to provide information that is integral to evaluating end of life microstructure and properties as a benchmark of international models developed for predicting radiation-induced swelling, segregation, precipitation, mechanical property degradation, and with the latest findings, susceptibility to irradiation-assisted stress corrosion cracking (IASCC).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Characterization and Quantification of Radiation-Induced Clusters/Precipitates in RPV Steels Using STEM-EDS and Machine Learning

Over the operational lifespan of a nuclear reactor, reactor pressure vessel (RPV) steels are subjected to significant neutron irradiation, resulting in complex microstructural changes and the consequent degradation of mechanical properties. Various physically motivated correlation models have been developed to predict neutron irradiation-induced embrittlement of RPVs under different irradiation conditions. However, the efficient and accurate characterizations and quantification of radiation-induced clusters in RPVs are still challenging, which will affect the precision of the predictive models for embrittlement of RPV components. In the DOE Visiting Faculty Program (VFP) research work at Oak Ridge National Lab (ORNL), I integrate machine learning to aid Scanning Transmission Electron Microscopy – Energy Dispersive X-ray Spectroscopy (STEM-EDS) analyses, which improve the characterization and quantification of radiation-induced clusters in RPV steels, thereby enabling more accurate predictions of material behavior under irradiation. The surveillance base- and welded- RPV steels were annealed at various temperatures of 340 °C, 450 °C and 500 °C for up to 168 hours, respectively. Afterwards, I have characterized radiation-induced clusters using advanced STEM-EDS techniques and subsequently applying machine learning algorithms to analyze and refine STEM-EDS datasets, enhancing the quantification of clusters compositions and distributions. In the end, an efficient workflow for integrating STEM-EDS data analysis with machine learning to address challenges including noise reduction has been developed. The completion of this VFP work will support bridge critical gaps in the accurate quantification of radiation-induced clusters in RPV steels using STEM-EDS and support the development of more precise models for predicting RPV embrittlement in the Light Water Reactor Sustainability program supported by Department of Energy and enhancing the collaboration between ORNL and Alred University. The outcome of the VFP project will leverage a few research papers submission to peer-reviewed journals in the relevant scientific field and a few oral presentations at national and international conferences.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

TEM Characterization of Nanostructured Grade 91 Neutron Irradiated to 6 DPA at 300 °C and 500 °C

Previously, samples of coarse grained (CG, grain diameter > 1 µm), ultra-fine grained (UFG, 1 µm> grain diameter> 100 nm), and nanocrystalline (NC, 100 nm> grain diameter) grade 91 steel were neutron irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Grade 91 is a ferritic/martensitic steel alloy with a nominal composition of Fe-9%Cr-1%Mo by weight, and it is being studied for applications in advanced nuclear reactor designs including as a fuel cladding material in gas cooled reactors. To assess the viability of this alloy for nuclear applications, the role of grain boundary size has not been sufficiently explored as a mechanism for developing radiation-tolerant materials. To achieve an UFG microstructure the material was processed using equal channel angular pressing (ECAP), and the NC microstructure was achieved through high pressure torsion (HPT). Such samples of various geometries for different types of characterization were irradiated to a dose of 6 displacements per atom (DPA) at separate temperature conditions of 300 °C and 500 °C. Thus, there were a total of 6 experimental conditions, three grain sizes (CG, UFG, and NC), each irradiated at two different temperature conditions (300 °C and 500 °C).

36 - MATERIALS SCIENCE↗

DOE OSTI2024

Irradiation And Post-Irradiation Examination Plan For A709

The Advanced Materials and Manufacturing Technologies program is proposing a neutron irradiation campaign for alloy 709 (A709) plate product developed by the Advanced Reactor Technologies program. This irradiation campaign will support the deployment of A709 (included welded components) by providing data of engineering importance, such as data that can be used for establishing end-of-life design parameters and knockdown factors for irradiated materials properties versus properties for non-irradiated material. It will also advance the scientific understanding of A709 irradiation behaviors and material property evolution, which is vital for deploying A709 components into reactors with different neutron spectra, neutron fluxes, and irradiation temperatures as well as accelerating the deployment of new product forms and weld procedures. This document presents the plan for the neutron irradiation and post-irradiation examination for A709 and the scientific and technical basis for the plan.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Assessment of irradiation embrittlement effect on fatigue life of a pressurized-water reactor pressure vessel using the fracture toughness master curve approach

The reactor pressure vessel (RPV) is a critical structural component in pressurized-water reactors, and it is designed to withstand extreme conditions, such as high pressures, elevated temperatures, and prolonged radiation exposure. Ensuring RPV integrity is essential for the safe and reliable long-term operation of nuclear power plants, especially as aging mechanisms such as fatigue and irradiation embrittlement pose increased risks. Fatigue, caused by cyclic thermal and mechanical loading, can lead to crack initiation in localized high-stress regions. Simultaneously, neutron irradiation, particularly in the beltline region, progressively reduces fracture toughness, increasing susceptibility to brittle fracture. These combined effects of fatigue and irradiation embrittlement potentially impact the RPV structural integrity, necessitating fitness-for-service assessments. This study applies the fracture toughness master curve approach to evaluate the impact of irradiation embrittlement on RPV fatigue life. A coupled thermo-mechanical stress analysis identifies critical stress locations under normal service transients, pinpointing regions most vulnerable to fatigue crack initiation and growth. Stress intensity factors for postulated flaws at these locations are calculated, enabling an assessment of fatigue life under irradiated and unirradiated conditions. The results indicate that neutron irradiation embrittlement accelerates the conditions in which a critical crack can form and lead to failure, particularly at lower temperatures. The failure occurs where reduced fracture toughness limits the material’s resistance to crack growth. Axial cracks at nozzle corners are the most life limiting without irradiation, while circumferential cracks demonstrate longer fatigue lives. The findings highlight the importance of incorporating irradiation effects into fatigue life predictions to ensure the long-term structural integrity of RPVs.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Radiation damage effects in beryllium for next generation neutrino beam targetry (Final Technical Report)

Current and future high-power accelerators put severe requirements on materials used for target and beam windows and target facilities have been recognized as a critical challenge in development of future particle accelerators. In accelerators, window and target materials are exposed to extreme conditions, which include bombardment with very high energy protons (1- 100 GeV) and thermomechanical shock waves. Radiation can cause direct damage in the material, and it leads to production of transmutation products (especially helium), both phenomena having a potential adverse effect on the stability and durability of the target/window material. At high enough temperatures, He can aggregate to form gas bubbles, which in turn cause significant dimensional changes (swelling), enable easy crack propagation, and eventually cause failure by fracture. On the other hand, if the temperature is too low, radiation damage accumulates in the form of internal defects (e.g., dislocations), leading to hardening and a decreased ductility of the material. In this project, we will focus on beryllium since it is considered to be one of the candidate materials for beam windows and targets in the next-generation proton accelerators, e.g., the Long Baseline Neutrino Facility (LBNF). Radiation effects in Be have been studied in the context of nuclear fusion reactor applications. However, key differences exist between reactor and accelerator conditions, including neutron vs. proton irradiation, continuous vs. pulsed beam flux, much higher energies of bombarding particles in accelerators, and higher operating temperatures for typical reactors. For example, the impact of beam pulsing on the radiation damage and the He bubble kinetics is largely unknown. While results obtained on Be from fusion research might not be directly transferrable to understanding target materials, there is an opportunity to bring state-of-the-art tools from materials research in nuclear reactors to aid design of target and beam window materials in high-power accelerators. To this end, the overarching goal of this project are to develop an experimentally-validated computational framework capable of predicting radiation damage evolution in beryllium relevant to beam window and target conditions, focusing on He bubble formation and growth as a function of irradiation temperature. Our model will be based on the cluster dynamics formalism, where size distribution of defects and He bubbles is simulated as a function of time, temperature, and radiation dose. Parameters for the model will be taken from published experiments and from high-fidelity atomistic simulations proposed in this project. In addition, we will carry out a series of targeted ex-situ and in-situ dual-beam experiments using low-energy protons to provide critical data for validation of the model on the effects of radiation on He clustering, He bubble distribution, and dislocation loop density/size in proton irradiated Be.

36 MATERIALS SCIENCE↗