Early Flight Fission Test Facilities (EFF-TF) To Support Near-Term Space Fission Systems

Dyke, Melissa Van

doi:10.1063/1.1649634

Cited by 3 publications

(1 citation statement)

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“…To address near term issues with an early hardware program, the Marshall Space Flight Center initiated an effort to examine a heat pipe system (Van Dyke, 2001). One of these systems, referred to as the Safe Affordable Fission Engine (SAFE), is a 100-kWt Los AIamos National Laboratory design that makes use of 61 stainless steelisodium heat pipe modules operating at a nominal temperature of 973 K (Houts, 2003) (Van Dyke, 2004). 1nG;astructure has been established to fabricate, fill, process, and evaluate the heat pipe modules at a component level (Martin, 2004a).…”

Section: Introductionmentioning

confidence: 99%

Multiple Restart Testing of a Stainless Steel Sodium Heat Pipe Module

Martin

Mireles²,

Reid³

2005

AIP Conference Proceedings

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A heat pipe cooled reactor is one of several candidate reactor cores being considered for space power and propulsion systems to support firture space exploration activities. Long life heat pipe modules. with designs verified through a combination of theoretical analysis and experimental evaluations. would be necessary to establish the viability of this option. A hardware-based program was initiated to begin experimental testing of components to verify c.omp:iao~e ofproposed ies;gtlj. TG this snd. a n u m b of stairlie55 a'b&'w&unl fr& pipe modules haw 'ken designed and fabrat& to support experimental testing of a Safe Affordable Fission Engine (SAFE) project, a 100-kWt a r e design pursued jointly by the Marshall Space Flight Center and the Los Alamos National Laboratory. One of the SAFE heat pipe modules was successllly subjected to over 200 restarts. examining the behavior of multiple passive 5eeze'thiw opesations Typical operation included a I-hour startup to an average evaporator temperature of IO00 K followed by a 15minute hold at tempaature. Nominal maximum input power during the hold period was 1.9 kW. Between heating cycles the module was cooled to less than 325 K, returning the sodium to a frozen state in preparation fop the next stiutup cycle.

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Section: Introductionmentioning

confidence: 99%

Multiple Restart Testing of a Stainless Steel Sodium Heat Pipe Module

Martin

Mireles²,

Reid³

2005

AIP Conference Proceedings

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Dynamic Response Testing in an Electrically Heated Reactor Test Facility

Bragg-Sitton

James²

2006

AIP Conference Proceedings

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Abstract. Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment.Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). Reactivity feedback calculations were then based on a bulk reactivity coeficient and measured average core temperature. Similar dynamic test techniques will be applied to a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. DDG testing will utilize a higher fidelity point kinetics model that incorporates the complete six delayed neutron groups to control core power transients. Additionally, reactivity feedback will be based on localized reactivity feedback coefficients and several independent temperature measurements taken within the core block. The neutron generation time and individual temperature feedback coefficients will be provided as model inputs. This paper will discuss the methodology that will be implemented in DDG testing and will assess the additional instrumentation needs to implement high fidelity dynamic testing. Dynamic testing is expected to commence early in FY06.

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