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Peterson, P. F.

Publications and source records attributed to Peterson, P. F..

8 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Progress in Development of Fluoride-salt-cooled High-Temperature Reactors (FHRs)

Fluoride-salt-cooled High-temperature Reactors (FHRs) are a new class of reactors that use graphite-matrix coated-particle fuel with a clean liquid salt coolant. The largest university FHR program is an Integrated Research Project (IRP) sponsored by the U.S. Department of Energy and others that is led by the Massachusetts Institute of Technology (MIT) and includes the University of California at Berkeley (UCB), the University of Wisconsin (UW), and the University of New Mexico (UNM). The year 2016 was the fifth year of this program. We report herein some of the major results of work done in 2016. The base-line FHR concept studied in this IRP uses (1) graphite-matrix coated-particle fuel with a pebble bed geometry and (2) a clean liquid salt coolant-flibe (a mixture of {sup 7}LiF and BeF{sub 2}). The combination of a very high-temperature fuel with failure temperatures above 1650 deg. C, a coolant boiling point above 1400 deg. C, and a coolant that traps most fission products implies the capability to assure no major fuel failure under severe accident conditions. There are a family of FHR concepts using different fuel geometries and different fluoride salts. The power conversion system is a Nuclear Air-Brayton Combined Cycle (NACC) with peak power provided by natural gas or a Firebrick Resistance Heated Energy Storage (FIRES) system. This enables a base-load reactor to provide variable electricity to the grid that substantially increases revenue relative to base-load nuclear reactors. Most of the results reported herein are applicable to most FHR concepts as well as liquid-fueled molten salt reactors (MSRs)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Design of a Supercritical Carbon Dioxide Corrosion Test Rig for Tube-to-Tubesheet Joint and High Temperature Insulation Testing

Supercritical carbon dioxide (S-CO{sub 2}) Brayton cycles are attractive for thermal power generation since they offer highly efficient means of power generation (thermal solar, coal, nuclear, etc.) with a compact footprint, resulting in potential cost savings. The University of California, Berkeley (UCB) is developing an S-CO{sub 2}- Brayton Cycle design for advanced reactors using sodium and molten salt coolants. Coiled tube gas heaters (CTGH) made from 316L stainless steel are presently foreseen to transfer heat from a sodium or molten salt coolant at atmospheric pressure (approximately 0.1 MPa) to an S-CO{sub 2} power conversion fluid at very high pressure (approximately 20 MPa or more). CTGHs operate at temperatures of 400 to 700 deg. C where time-dependent creep is an important design issue. Given the high external pressure conditions and compressive stress, the highest stresses and creep rates in CTGH are expected to occur at the tube-to-tubesheet joints. UCB has developed different diffusion bonding techniques to fabricate 316L stainless steel tube-to-tubesheet joint dummies representative of the final design: press fitting and diffusion bonding, tube expansion and diffusion bonding as well as hybrid friction diffusion bonding. High pressure/temperature testing of such joints in an S-CO{sub 2} environment is important prior to deployment and will be foreseen with the corrosion test rig introduced in this work. This summary provides a brief overview of an Innovative S-CO, creep corrosion test rig for tube-to-tube sheet joint testing currently under construction at UCB. In contrast to other comparable test rigs, the design (1) makes the samples an active part of the pressure barrier and (2) uses internal insulation to reduce the temperature at the vessel wall, allowing the vessel to be manufactured from commercial carbon steel. (authors)

42 ENGINEERING↗
DOE OSTI2016

Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis

The University of California, Berkeley (UCB) is performing thermal hydraulics safety analysis to develop the technical basis for design and licensing of fluoride-salt-cooled, high-temperature reactors (FHRs). FHR designs investigated by UCB use natural circulation for emergency, passive decay heat removal when normal decay heat removal systems fail. The FHR advanced natural circulation analysis (FANCY) code has been developed for assessment of passive decay heat removal capability and safety analysis of these innovative system designs. The FANCY code uses a one-dimensional, semi-implicit scheme to solve for pressure-linked mass, momentum and energy conservation equations. Graph theory is used to automatically generate a staggered mesh for complicated pipe network systems. Heat structure models have been implemented for three types of boundary conditions (Dirichlet, Neumann and Robin boundary conditions). Heat structures can be composed of several layers of different materials, and are used for simulation of heat structure temperature distribution and heat transfer rate. Control models are used to simulate sequences of events or trips of safety systems. A proportional-integral controller is also used to automatically make thermal hydraulic systems reach desired steady state conditions. A point kinetics model is used to model reactor kinetics behavior with temperature reactivity feedback. The underlying large sparse linear systems in these models are efficiently solved by using direct and iterative solvers provided by the SuperLU code on high performance machines. Input interfaces are designed to increase the flexibility of simulation for complicated thermal hydraulic systems. In conclusion, this paper mainly focuses on the methodology used to develop the FANCY code, and safety analysis of the Mark 1 pebble-bed FHR under development at UCB is performed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Pebble fuel design for the PB-FHR

This paper presents the results of parametric studies of pebble fuel that can guide the design of future PB-FHR cores. The pebble fuel designs are assessed using the following performance characteristics: burnup, reactivity feedback, transient response, timescale to reach equilibrium cycle, and protection of structural components. The performance of a thorium pebble blanket is assessed by comparing against a seed-only system and system that utilizes a graphite pebble reflector instead of a thorium blanket. This paper presents the functional requirements and a methodology to assess these fuel pebble designs. This paper identifies a feasible design space for low enriched uranium pebbles and selected a baseline pebble design for safe, economic energy generation. Furthermore, this study finds a thorium blanket does not increase the performance of the system significantly with respect to a graphite pebble reflector. Therefore, a graphite pebble reflector is recommended in the baseline full-core design to extend the lifetime of the outer solid graphite reflector to the life of plant. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2010

ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using factory prefabricated structural modules, for application to external event shell and base isolated structures.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗