About

Prof. Per F. Peterson is the William S. Floyd and Jean McCallum Floyd Chair in Engineering at the University of California, Berkeley. His research focuses on problems in energy and environmental systems, including high-temperature reactors, high-level nuclear waste processing, and nuclear materials management. He manages the UC Berkeley Thermal Hydraulics Research Laboratory, which conducts experiments related to reactor thermal hydraulics. Prof. Peterson teaches undergraduate and graduate courses in heat and mass transfer, fluid dynamics, reactor thermal hydraulics, and reactor safety, with a focus on nuclear applications. His research interests involve reactor safety and licensing, high-temperature reactor technology, nuclear air combined cycle power conversion, and security and safeguards technologies for nuclear materials and facilities. His ongoing projects include studies of heat transport and fluid mechanics in liquid-fluoride salt cooled reactors, gas-Brayton power conversion, performance-based regulation and licensing of advanced reactors, seismic base isolation, modular construction methods, and nuclear security. He has contributed to the development of safety standards for fluoride-salt-cooled high-temperature reactors and has participated in national advisory activities, including the Blue Ribbon Commission on America's Nuclear Future and the Generation IV Roadmap. His work in the 1990s influenced the passive safety systems used in modern reactor designs such as the GE ESBWR and Westinghouse AP-1000. Prof. Peterson's research primarily focuses on heat transfer, fluid mechanics, regulation, and licensing for high-temperature reactors using liquid fluoride salts as coolants.

Research topics

• Computer Science
• Mechanical engineering
• Engineering
• Geotechnical engineering
• Physics
• Manufacturing engineering
• Mechanics
• Software engineering
• Mathematics
• Reliability engineering
• Engineering management
• Systems engineering
• Simulation
• Geometry

Selected publications

• Modeling glass degradation and release of radionuclides from vitrified waste for performance assessment simulations

  Frontiers in Nuclear Engineering · 2026-01-05 · 1 citations

  articleOpen accessSenior author

  The release of radionuclides initially encapsulated in a slowly degrading solid waste form and contained in an eventually corroding canister defines the source term for numerical simulations for the assessment of a geologic repository for high-level radioactive waste. While the details of waste degradation, canister corrosion, and dissolution and mobilization of the radionuclides in pore water include complex chemical reaction and transport processes that are coupled to the thermal, hydrological, microbiological, and mechanical conditions in the repository, the source-term model suitable for use in a numerical performance assessment model should be a defensible abstraction of these mechanisms. We developed a radiological source-term model and implemented it into a non-isothermal flow and transport simulator. While the proposed source-term model is applicable to various waste forms, canister systems, and disposal concepts, we specifically considered radionuclide releases from vitrified high-level waste placed in a cylindrical canister disposed in a deep vertical borehole repository. In this model, waste degradation is a function of temperature, and it can be adjusted to evaluate the influence of and propagate uncertainties in pH, passivation reactions, and chemical conditions as well as geometrical factors. The time-dependent, congruent release of safety-relevant radionuclides present in the decaying inventory is then calculated. Finally, the radionuclides are mobilized by diffusive and advective transport according to the thermo-hydraulic conditions prevailing in the near field of the repository, from where they migrate through the geosphere to the accessible environment. We examine the influence of the source-term model’s parameters on performance assessment calculations through sensitivity and uncertainty propagation analyses, identifying influential factors and confirming the upper bound of their impact. These considerations align with the overarching goal of repository design, which is to demonstrate that engineered and natural barriers can collectively delay radionuclide migration for timescales far exceeding human planning, thereby providing multiple, redundant barriers against environmental contamination.

  PublisherOA PDFDOI

• Toward the performance assessment of advanced nuclear waste forms: temperature dependence of lanthanide borosilicate glass dissolution

  npj Materials Degradation · 2026-02-18

  articleOpen access

  Abstract Lanthanide borosilicate (LaBS) glasses are among the most promising waste forms for the immobilization of high-level radioactive waste generated from advanced nuclear fuel cycles. However, the temperature dependence of their dissolution kinetics remains poorly understood and constrained, limiting the integration of these materials into established performance assessment models. Here, we investigate the dissolution behavior of the legacy AmCm2-19 LaBS glass and the benchmark alkali aluminoborosilicate ISG-1 in deionized water between 50 °C and 250 °C using ASTM C1285 (Product Consistency Test-B) protocols. For AmCm2-19 LaBS glass, normalized elemental release rates for boron and silicon increase with temperature before plateauing near 150 °C, consistent with solubility-limited behavior. From data obtained at 50 °C and 100 °C, Arrhenius analysis yields activation energies of E a (B) = 24.8 ± 0.3 kJ mol⁻¹ and E a (Si) = 14.4 ± 0.2 kJ mol⁻¹, similar or slightly lower than those previously reported for two other compositions of LaBS glasses. No secondary phases or alteration layers were detected by SEM-EDX or pXRD. These results establish one of the first temperature-dependent kinetic datasets for LaBS glass dissolution, providing quantitative parameters to inform mechanistic corrosion models and predictive simulations of glass degradation in geological disposal environments.

  PublisherOA PDFDOI

• Graphite Waste Classification and Disposal Cost Estimation for High Temperature Gas and Salt Reactors

  SSRN Electronic Journal · 2025-01-01

  preprintOpen access
  PublisherDOI

• Graphite waste classification and disposal cost estimation for high temperature gas and salt reactors

  Annals of Nuclear Energy · 2025-08-29 · 1 citations

  article
  PublisherDOI

• An open-source Thermo-hydraulic Uniphase Advection and Convection Solver for Salt Flows (TUAS)

  International Journal of Advanced Nuclear Reactor Design and Technology · 2024-12-01

  articleOpen accessSenior author
  PublisherDOI

• An Open Source Thermo-Hydraulic Uniphase Solver for Advection and Convection in Salt Flows (Tuas)

  SSRN Electronic Journal · 2024-01-01

  preprintOpen accessSenior author
  PublisherDOI

• National Council on Radiation Protection (NCRP) 2024 annual meeting: advanced and small modular nuclear power reactors^*

  Journal of Radiological Protection · 2024-10-04 · 1 citations

  articleOpen access

  On 25-26 March 2023, the U.S. National Council on Radiation Protection and Measurements (NCRP) held its 2024 annual meeting in Bethesda, Maryland, USA. The NCRP dates from 1929, and this meeting celebrated the 60th anniversary of receiving a U.S. Congressional Charter. For this annual meeting the NCRP felt it was essential to provide a briefing about advanced and small modular nuclear reactors (SMRs). The Journal of Radiological Protection is delighted to publish the following synopsis of material presented at the U.S. NCRP meeting. This synopsis is divided into five sections. The first section provides an overview of the whole meeting together with summaries of two context setting overview papers. The following four sessions of this synopsis are specific to advanced and small modular nuclear power reactors. The meeting also included keynote presentations by three of NCRP annual award recipients. The meeting topical areas were Technology Overview and Critical Issues. The individual papers laid the groundwork to understanding reactor technologies, terminology, and the fundamental concepts and processes for electrical generation. The perspectives of the U.S. Environmental Protection Agency and states, through the Conference of Radiation Control Program Directors were provided. The papers included a discussion of diverse topics including potential emergency preparedness considerations, radiological survey requirements, an evaluation of the future of nuclear power, the economics of reactors (both large and small), and the critical issues identified by the recent National Academies of Sciences' study on advanced reactors. The summary papers were developed to briefly document the major points and concepts presented during the oral papers presented at the 2024 NCRP Annual Meeting. The meeting heralded the dawn of a new era for commercial nuclear power.

  PublisherOA PDFDOI

• Divertor Baffling and Biasing Experiments on DIII-D

  OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information) · 2024-04-01 · 6 citations

  paratextOpen access

  First results of divertor baffling and biasing experiments in Ohmic and low power beam-heated plasmas using the DIII-D Advanced Divertor are reported. The Advanced Divertor commissioned recently is designed to study baffling, plasma biasing, and dc helicity injection current drive. Using this unique baffle configuration in Ohmic and low powered beam-heated plasmas, we have obtained pressures under the baffle in excess of 5 mTorr and driven a poloidal scrape-off layer current in excess of 4 kA. Pressure under the baffle is strongly influenced by biasing. With the E_r x B drift in the direction of the baffle throat, the pressure under the baffle can be a factor of five higher than with the opposite polarity. Biasing increases the level of divertor region magnetic fluctuations.

  PublisherOA PDF

• Systems and methods for enhancing isolation of high-temperature reactor containments

  OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information) · 2023-01-23

  articleOpen access1st authorCorresponding

  A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

  PublisherOA PDF

• Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

  OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information) · 2023-04-27 · 1 citations

  articleOpen access1st authorCorresponding

  A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  Publisher

Frequent coauthors

• Haihua Zhao

  28 shared

• C.S. Debonnel

  Commissariat à l'Énergie Atomique et aux Énergies Alternatives

  24 shared

• V.E. Schrock

  22 shared

• W.R. Meier

  Xenobe Research Institute

  17 shared

• Charles Forsberg

  Massachusetts Institute of Technology

  17 shared

• S.S. Yu

  16 shared

• D. R. Welch

  14 shared

• Nicolas Zweibaum

  Kairos (United States)

  13 shared

Awards & honors

• Presidential Young Investigator Fellow, American Nuclear Soc…