Preliminary safety evaluation of the advanced burner test reactor.

Dunn, F.E.; Fanning, Thomas G.; Cahalan, J.E.

doi:10.2172/924683

Cited by 12 publications

(10 citation statements)

References 2 publications

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“…The peak cladding temperature rises to 660°C slightly after the end of pump coast down. The peak clad temperature increase is very mild and within the design limit, which can be compared with the maximum 750°C peak clad temperature predicted for the ABTR design 19 . The second peak of the peak cladding temperature arrives around half an hour later at much lower value.…”

Section: Iid Lofc Analysis Resultsmentioning

confidence: 64%

Improving SFR economics through innovations from thermal design and analysis aspects

Zhao

Zhang

Mousseau

et al. 2009

Nuclear Engineering and Design

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Section: Iid Lofc Analysis Resultsmentioning

confidence: 64%

Improving SFR economics through innovations from thermal design and analysis aspects

Zhao

Zhang

Mousseau

et al. 2009

Nuclear Engineering and Design

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“…Before the safety analyses of SLFFR, verification tests of the MUSA code were performed by a code-to-code comparison with the well-validated safety analysis code SAS4A/SASSYS-1. The steady-state operating conditions and the ULOF accident scenario of the advanced burner test reactor (ABTR) were analyzed using the MUSA code, and the results were compared with the reference solutions obtained with the SAS4A/SASSYS-1 code system (Dunn, et al, 2006).…”

Section: Verification Tests Of Musa Codementioning

confidence: 99%

Stationary liquid fuel fast reactor SLFFR — Part II: Safety analysis

Tian¹,

Jung²,

Yang³

2016

Nuclear Engineering and Design

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Safety characteristics have been evaluated for the stationary liquid fuel fast reactor (SLFFR) proposed for effective burning of hazardous TRU elements of used nuclear fuel. In order to model the geometrical configuration and reactivity feedback mechanisms unique to SLFFR, a multi-channel safety analysis code named MUSA was developed. MUSA solves the timedependent coupled neutronics and thermal-fluidic problems. The thermal-fluidic behavior of the core is described by representing the core with one-dimensional parallel channels. The primary heat transport system is modeled by connecting compressible volumes by liquid segments. A point kinetics model with six delayed neutron groups is used to represent the fission power transients. The reactivity feedback is estimated by combining the temperature and density variations of liquid fuel, structural material and sodium coolant with the corresponding axial distributions of reactivity worth in each individual thermal-fluidic channel. Preliminary verification tests with a conventional solid fuel reactor agreed well with the reference solutions obtained with the SAS4A/SASSYS-1 code. Transient analyses of SLFFR were performed for unprotected transient overpower (UTOP), unprotected loss of heat sink (ULOHS) and unprotected loss of flow (ULOF) accidents. The results showed that the thermal expansion of liquid fuel provides sufficiently large negative feedback reactivity for passive shutdown of UTOP and ULOHS. The ULOF transient is also terminated passively with the negative reactivity introduced by the gas expansion modules installed at the core periphery. The maximum coolant temperatures have sufficient margins to the sodium boiling point for all three unprotected transient scenarios.

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“…The lack of general thermal mixing and stratification models in those codes severely limits their application and accuracy for safety analysis, especially for passively safe advanced light-water reactors (ALWRs), where the primary system and containments are more strongly coupled (Zhao and Peterson, 2010). The SASSYS code developed by argonne national laboratory (ANL), only provides lumped-volume-based 0-D models that can only give very approximate results and can only handle simple cases with one mixing source (Dunn et al, 2006). COMMIX code developed by ANL uses CFD tools to analyze simple configuration small-scale thermal stratification problems and achieved limited success (Chang and Bottoni, 1994;Kasza et al, 2007).…”

Section: Thermal Stratification In Large Poolsmentioning

confidence: 99%