Peter Hosemann is professor in the Department for Nuclear Engineering at the University of California Berkeley and current department chair, head graduate adviser and UC Berkeley’s radiation safety chair. In 2017 Professor Hosemann was elected chair of the Nuclear Science User Facility user group an international network of institutions providing unique materials characterisation tools. Professor Hosemann received his PhD in Material Science from the Montanuniversität Leoben, Austria in 2008 while he conducted the research on lead bismuth eutectic corrosion, ion beam irradiations and microscale mechanical testing was carried out at Los Alamos National Laboratory. He continued his research at Los Alamos National Laboratory and joined the UC Berkeley faculty in 2010. Professor Hosemann has authored more than 160 per reviewed publications since 2008. In 2014 he won the best reviewer of the journal of nuclear materials award, the ANS literature award and in 2015 he won the TMS early career faculty fellow award and the AIME Robert Lansing Hardy award. While being dedicated to his research and teaching he also leads the UC Berkeley Bladesmithing team which won the title of “best example of a traditional blade” for UC Berkeley and is the lead faculty for the CalSol solar car racing team which won the American Solar challenge for Berkeley in 2017.
Professor Hosemann is also the current chair of the NSUF User Organisation
"A novel in-situ, lift-out, three-point bend technique to quantify the mechanical properties of an ex-service neutron irradiated inconel X-750 component"
Peter Hosemann, Cameron Howard, Colin Judge, David Poff, Stephen Parker, Malcolm Griffiths,
Journal of Nuclear Materials
A first of its kind small scale mechanical testing technique involving micro-three-point bending was invented, developed, and implemented on reactor irradiated, active Inconel X-750 components removed from service after approximately 53 and 67 dpa. These tests were performed at ambient room temperature in-situ using a scanning electron microscope in order to obtain live recordings of sample deformation and loading curves. Sample and testing apparatus preparation required novel lift-out and fabrication processes. Materials from two irradiation temperature regimes, low temperature (120-280 °C) and high temperature (300 ± 15 °C) were examined. Manufacturing and finishing (grinding) of this component create differences between its edge and center, so micro-specimens from both areas were extracted in order to study these differences. According to three-point beam bending theory, a 0.2% offset yield stress parameter is introduced and calculated for all specimens. Differences in mechanical properties due to irradiation temperature and dose effects were observed. Material irradiated at the higher temperature exhibited yield strength increases of ~540 MPa after 53 dpa and ~1000 MPa after 67 dpa. There was little difference (=310 MPa) in yield strength between materials irradiated at the lower temperature at 53 dpa and 67 dpa compared with non-irradiated material. Differences in yield strengths between the edge and center of the component are retained after irradiation. The difference in yield strengths for the edge and center regions was ~740 MPa for non-irradiated material. After irradiation to a dose of 67 dpa these differences were ~570 MPa for the lower irradiation temperature and ~710 MPa for higher irradiation temperature. There were no indications of grain boundary failures via cracking except for material irradiated to 67 dpa at low temperature.
|"An analytical method to extract irradiation hardening from nanoindentation hardness-depth curves" Anna Kareer, A. Prasitthipayong, David Krumwiede, D.M. Collins, Peter Hosemann, Steve Roberts, Journal of Nuclear Materials Vol. 498 2018 274-281 Link|
"Atom probe tomography analysis of high dose MA957 at selected irradiation temperatures"
Nathan Bailey, Peter Hosemann, Erich Stergar, Mychailo Toloczko,
Journal of Nuclear Materials
Oxide dispersion strengthened (ODS) alloys are meritable structural materials for nuclear reactor systems due to the exemplary resistance to radiation damage and high temperature creep. Summarized in this work are atom probe tomography (APT) investigations on a heat of MA957 that underwent irradiation in the form of in-reactor creep specimens in the Fast Flux Test Facility–Materials Open Test Assembly (FFTF–MOTA) for the Liquid Metal Fast Breeder Reactor (LMFBR) program. The oxide precipitates appear stable under irradiation at elevated temperature over extended periods of time. Nominally, the precipitate chemistry is unchanged by the accumulated dose; although, evidence suggests that ballistic dissolution and reformation processes are occurring at all irradiation temperatures. At 412 °C–109 dpa, chromium enrichments – consistent with the a' phase – appear between the oxide precipitates, indicating radiation induced segregation. Grain boundaries, enriched with several elements including nickel and titanium, are observed at all irradiation conditions. At 412 °C–109 dpa, the grain boundaries are also enriched in molecular titanium oxide (TiO).
"Ceramic composites: A review of toughening mechanisms and demonstration of micropillar compression for interface property extraction"
Christian Deck, Peter Hosemann, Yutai Katoh, Yevhen Zayachuk, Joey Kabel, David Armstrong, Takaaki Koyanagi,
Journal of Materials Research
Ceramic fiber–matrix composites (CFMCs) are exciting materials for engineering applications in extreme environments. By integrating ceramic fibers within a ceramic matrix, CFMCs allow an intrinsically brittle material to exhibit sufficient structural toughness for use in gas turbines and nuclear reactors. Chemical stability under high temperature and irradiation coupled with high specific strength make these materials unique and increasingly popular in extreme settings. This paper first offers a review of the importance and growing body of research on fiber–matrix interfaces as they relate to composite toughening mechanisms. Second, micropillar compression is explored experimentally as a high-fidelity method for extracting interface properties compared with traditional fiber push-out testing. Three significant interface properties that govern composite toughening were extracted. For a 50-nm-pyrolytic carbon interface, the following were observed: a fracture energy release rate of ~2.5 J/m2, an internal friction coefficient of 0.25 ± 0.04, and a debond shear strength of 266 ± 24 MPa. This research supports micromechanical evaluations as a unique bridge between theoretical physics models for microcrack propagation and empirically driven finite element models for bulk CFMCs.
|"Direct comparison of nanoindentation and tensile test results on reactor-irradiated materials" David Krumwiede, Takuya Yamamoto, Tarik Saleh, Stuart Maloy, G. Robert Odette, Peter Hosemann, Journal of Nuclear Materials Vol. 504 2018 135-143 Link|
|"Ion irradiation effects on commercial PH 13-8 Mo maraging steel Corrax" Ce Zheng, Ryan Schoell, Peter Hosemann, Djamel Kaoumi, Journal of Nuclear Materials Vol. 514 2019 255-265 Link|
"Mechanical characteristics of SiC coating layer in TRISO fuel particles"
Thak Sang Byun, David Frazer, Peter Hosemann, John Hunn, Maria Okuniewski, Kurt Terrani, Gokul Vasudevamurthy, J. N. Matros, Brian Jolly,
Journal of Nuclear Materials
Tristructural isotropic (TRISO) particles are considered as advanced fuel forms for a variety of fission platforms. While these fuel structures have been tested and deployed in reactors, the mechanical properties of these structures as a function of production parameters need to be investigated in order to ensure their reliability during service. Nanoindentation techniques, indentation crack testing, and half sphere crush testing were utilized in order to evaluate the integrity of the SiC coating layer that is meant to prevent fission product release in the coated particle fuel form. The results are complimented by scanning electron microscopy (SEM) of the grain structure that is subject to change as a function of processing parameters and can alter the mechanical properties such as hardness, elastic modulus, fracture toughness and fracture strength. Through utilization of these advanced techniques, subtle differences in mechanical properties that can be important for in-pile fuel performance can be distinguished and optimized in iteration with processing science of coated fuel particle production.
"Micro-mechanical evaluation of SiC-SiC composite interphase properties and debond mechanisms"
Mehdi Balooch, Peter Hosemann, Cameron Howard, Yutai Katoh, Takaaki Koyanagi, Yong Yang, Joey Kabel, Kurt Terrani,
Composites Part B: Engineering
SiC-SiC composites exhibit exceptional high temperature strength and oxidation properties making them an advantageous choice for accident tolerant nuclear fuel cladding. In the present work, small scale mechanical testing along with AFM and TEM analysis were employed to evaluate PyC interphase properties that play a key role in the overall mechanical behavior of the composite. The Mohr-Coulomb formulation allowed for the extraction of the internal friction coefficient and debonding shear strength as a function of the PyC layer thickness, an additional parameter. These results have led to re-evaluation of the Mohr-Coulomb failure criterion and adjustment via a new phenomenological equation.SiC-SiC composites exhibit exceptional high temperature strength and oxidation properties making them an advantageous choice for accident tolerant nuclear fuel cladding. In the present work, small scale mechanical testing along with AFM and TEM analysis were employed to evaluate PyC interphase properties that play a key role in the overall mechanical behavior of the composite. The Mohr-Coulomb formulation allowed for the extraction of the internal friction coefficient and debonding shear strength as a function of the PyC layer thickness, an additional parameter. These results have led to re-evaluation of the Mohr-Coulomb failure criterion and adjustment via a new phenomenological equation.
|"Microstructural and nanomechanical characterization of in-situ He implanted and irradiated fcc materials" David Frazer, Peter Hosemann, Djamel Kaoumi, Ce Zheng, Microscopy & Microanalysis Vol. 23 (Suppl 1) 2017 756-757 Link|
"Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic"
Mehdi Balooch, Edgar Buck, Andy Casella, Peter Hosemann, Donald Olander, Dave Senor, Kurt Terrani,
Journal of Nuclear Materials
A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.
|"The crystal structure, orientation, relationships and interfaces of the nanoscale oxides in nanostructured ferritic alloys" Yuan Wu, Jim Criston, Stephan Kraemer, Nathan Bailey, G. Robert Odette, Peter Hosemann, Acta Materialia Vol. 111 2016 108-115 Link|
"Twin boundary-accelerated ferritization of austenitic stainless steels in liquid lead–bismuth eutectic"
David Frazer, Peter Hosemann, Konstantina Lambrinou, Erich Stergar,
Exposure of austenitic stainless steels to liquid lead-bismuth eutectic with low concentration of dissolved oxygen typically results in selective leaching of highly-soluble alloying elements and ferritization of the dissolution-affected zone. In this work, focused ion beam, transmission electron backscatter diffraction and scanning transmission electron microscopy were utilised to elucidate early-stage aspects of the dissolution corrosion process of cold-worked austenitic stainless steels exposed to static, oxygen-poor liquid lead-bismuth eutectic at 450C for 1000 hours. It was found that deformation-induced twin boundaries in the cold-worked steel bulk provide paths of accelerated penetration of the liquid metal into the steel bulk.
|"Effect of Thermal And Irradiation-induced Long Range Ordering in Ni-Cr Model Alloys" Fei Teng, David Sprouster, Peter Hosemann, Li-Jen Yu, Emmanuelle Marquis, Julie Tucker, NuMat October 14-18, (2018)|
|"Irradiation-induced microstructure and mechanical property evolution in an Fe-9Cr ODS alloy" Corey Dolph, Peter Hosemann, 8th International Conference on Processing & Manufacturing of Advanced Materials (THERMEC) [unknown]|
|"Manufacturing of nanostructured ODS steel cladding tubes for advanced nuclear reactors using cold spray technology" Mia Lenling, Hwasung Yeom, Ben Maier, Greg Johnson, David Hoelzer, Peter Hosemann, Stuart Maloy, Kumar Sridharan, Jeffrey Graham, The Minerals, Metals and Materials Society (TMS) 2019 March 10-14, (2019)|
|"Micro-Cantilever Testing of Environmental Barrier Coatings on CVD SiC" Peter Hosemann, Yutai Katoh, ANS Annual Meeting 2018 June 18-22, (2018)|
|"On the Opportunities and Challanges for Using Test REactor and Charged Particle Irradiations to Help Predict Inaccessible In-service Neutron Irradiations Effects" Peter Hosemann, Peter Wells, Takuya Yamamoto, TMS 2014 February 16-20, (2014)|
|"Post Irradiation Examination of Fast Neutron Irradiated 14YWT Tubes at Nuclear Science User Facilities" Eda Aydogan, Peter Hosemann, David Krumwiede, Stuart Maloy, Tarik Saleh, 2017 ANS Annual Meeting [unknown]|
|"Program Review Presentation Entitled "Development of Low Temperature Spray Process for Manufacturing Fuel Cladding and Surface Modification of Reactor Components"" Mia Lenling, Kumar Sridharan, Hwasung Yeom, Peter Hosemann, David Hoelzer, Jeffrey Graham, Stuart Maloy, Advanced Methods for Manufacturing Program Review December 4-6, (2018)|
|"The Role of Ordering Phase Transformations in Micro-mechanics of Ni-Cr Alloys for Nuclear Applications" Fei Teng, Stephanie Pitts, Hi Vo, Peter Hosemann, Julie Tucker, TMS March 11-14, (2018)|
|From the Users Group: Nominations Open for Users Group Executive Committee - Open positions include secretary/chair elect, committee members and student members The goal of the Users Group is to provide a formal and clear channel from the NSUF users to the NSUF program office as well as advocate for the experimental activities at NSUF, and the Executive Committee helps facilitate this. Wednesday, August 28, 2019 - Newsletter, Users Group|
|From the Users Group: Vote for Your New Executive Committee - Election closes 10/31 The goal of the Users Group is to provide a formal and clear channel from the NSUF users to the NSUF program office as well as advocate for the experimental activities at NSUF, and the Executive Committee helps facilitate this. Wednesday, October 9, 2019 - Users Group, Newsletter|
|Users Organization Meeting Presentations Now Available - Wednesday, March 25, 2020 - Newsletter, Users Group|
|2020 NSUF Annual Review - Presentations The 2020 NSUF Annual Review presentations are now available online Tuesday, December 15, 2020 - DOE, Annual Review, Presentations|
|NSUF awards 24 Rapid Turnaround Experiment proposals - Approximately $1.42M has been awarded. Wednesday, February 8, 2023 - Calls and Awards|
This NSUF Profile is 85
Authored 10+ NSUF-supported publications
Presented an NSUF-supported publication
Submitted 3+ RTE Proposals to NSUF
Awarded 3+ RTE Proposals
Collaborated on 3+ RTE Proposals
Reviewed 10+ RTE Proposals
Characterization of CANDU Core Internals via Small Scale Mechanical Testing - FY 2015 RTE 1st Call, #530
Characterization of reactor irradiated ODS materials using Local Electrode Atom Probe Tomography. - FY 2012 RTE Solicitation, #361
Installation of a novel high throughput micro and macro scale machining capability for pre and post irradiation examination - University General Scientific Infrastructure FY18, #1345
Mesoscale irradiation of HT-9 - FY 2023 RTE 1st Call, #4534
A Bootstrapping Approach to Optimizing the Fidelity of Ion Versus Neutron Irradiations - FY 2015 RTE 3rd Call, #595
Atom probe tomography on highly irradiated F/M steels - FY 2011 RTE Solicitation, #306
Characterization of Grain Boundary Damage in Highly Irradiated Specimens Exposed to Irradiation Assisted Stress Corrosion Cracking. - FY 2017 RTE 3rd Call, #1076
Effects of dose and temperature on microstructural evolution of Zircaloy-4 alloys during proton irradiation - FY 2019 RTE 1st Call, #1681
Hardness profiling of ion irradiated 15-15Ti cladding steel using CSM nano-indentation - FY 2017 RTE 3rd Call, #1023
Microstructure Analysis of High Dose Neutron Irradiated Microstructures - FY 2016 RTE 1st Call, #604
Nanohardness Measurements on Neutron Irradiated Steel Samples for Next Generation Reactors3 - FY 2015 RTE 3rd Call, #585
Nanoscale analysis of Mn-Si-Ni phase in neutron irradiated T91 at 320C - FY 2019 RTE 2nd Call, #1721
Synchrotron X-ray Diffraction Measurements of Spatially Resolved Strain Fields in Nuclear Fuel Plates - FY 2010 APS, #225