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Verification and Demonstration of One-Dimensional Freezing Model in SAM for Salt-Cooled Reactor Analysis Applications
This work presented the development and implementation of the one-dimensional freezing model in system analysis code, SAM, as well as code verification, and code demonstration during a postulated overcooling transient, for fluoride salt-cooled high-temperature reactor (FHR) system and safety analysis applications. The paper at first summarized the freezing model, finite element numerical method, and special numerical treatment for handling phase appearance/disappearance. Analytical solutions were derived for two cases (with and without solid walls) for code verifications purpose. As expected, numerical results predicted by the SAM code agreed very well with the analytical solution. A code demonstration was then performed on a postulated protected overcooling event transient of a generic reference PB-FHR design. The code was found to successfully predict salt freezing during such a postulated event. However, due to lack of salt freezing testing data, code validation has not been performed in this work, which will be pursued in later studies when such data becomes available.
High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion
For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.
Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator
The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.
Gamma-spectrometry measurements of filtration media from an Advanced Gas-cooled Reactor
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Defining the performance envelope of reactivity-initiated accidents in a high-temperature gas-cooled reactor
Not Available
Environmental degradation of structural materials in liquid lead- and lead-bismuth eutectic-cooled reactors
Not provided.
Identifying and quantifying a complete set of full-power initiating events during early design stages of high-temperature gas-cooled reactors
Not Available
Assessment of Fission Product Release in High-Temperature Gas-Cooled Reactors using TRISO Fuel for Radiological Consequence Assessment
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MELCOR Best Practices for High Temperature Gas-cooled Reactor Modeling.
Abstract not provided.
Modular High Temperature Gas-Cooled Reactor: Severe Accident Phenomena and Safety Criteria [Slides]
Radionuclide Retention within Modular HTGR Fuel
Fabrication and testing of tungsten heat pipes for heat pipe cooled reactors
Fabrication and tests of tungsten reactor pipe and its heating characteristics up to 1900 K
Fabrication and testing of tungsten heat pipes for heat pipe cooled reactors
Li-filled vacuum-deposited W heat pipes for efficient heat extraction from nuclear reactors, discussing design, fabrication and testing
Fabrication and testing of tungsten heat pipes for heat pipe cooled reactors.
The choice of chemically vapor deposited (CVD) tungsten as material for the heat pipes is based on studies reported by Busse et al. (1970). Aspects of pipe construction are discussed together with wick construction, and heat pipe operation. An auxiliary heater was used to overcome problems during the start of operation in the case of a thermally loaded pipe. Two heat pipes were operated at temperatures up to 1850 K. Annular and channeled wicks were investigated. The design of a cross flow heat exchanger is presented together with the design of a reactor heat pipe life test station.
Simulator test to study hot-flow problems related to a gas cooled reactor
An advance study of materials, fuel injection, and hot flow problems related to the gas core nuclear rocket is reported. The first task was to test a previously constructed induction heated plasma GCNR simulator above 300 kW. A number of tests are reported operating in the range of 300 kW at 10,000 cps. A second simulator was designed but not constructed for cold-hot visualization studies using louvered walls. A third task was a paper investigation of practical uranium feed systems, including a detailed discussion of related problems. The last assignment resulted in two designs for plasma nozzle test devices that could be operated at 200 atm on hydrogen.
Cooling Channel Optimization in Additively Manufactured Gas Cooled Reactor Core
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Analysis of Graphite Matrix Kinetics and Burnoff Products under Off-Normal High-Temperature Gas-Cooled Reactors Conditions
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