Reactor physics methods, models, and applications used to support the conceptual design of the Advanced Neutron Source

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This report summarizes the neutronics analysis performed during 1991 and 1992 in support of characterization of the conceptual design of the Advanced Neutron Source (ANS). The methods used in the analysis, parametric studies, and key results supporting the design and safety evaluations of the conceptual design are presented. The analysis approach used during the conceptual design phase followed the same approach used in early ANS evaluations: (1) a strong reliance on Monte Carlo theory for beginning-of-cycle reactor performance calculations and (2) a reliance on few-group diffusion theory for reactor fuel cycle analysis and for evaluation of reactor performance at specific ... continued below

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197 p.

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Gehin, J.C.; Worley, B.A.; Renier, J.P.; Wemple, C.A.; Jahshan, S.N. & Ryskammp, J.M. August 1, 1995.

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Description

This report summarizes the neutronics analysis performed during 1991 and 1992 in support of characterization of the conceptual design of the Advanced Neutron Source (ANS). The methods used in the analysis, parametric studies, and key results supporting the design and safety evaluations of the conceptual design are presented. The analysis approach used during the conceptual design phase followed the same approach used in early ANS evaluations: (1) a strong reliance on Monte Carlo theory for beginning-of-cycle reactor performance calculations and (2) a reliance on few-group diffusion theory for reactor fuel cycle analysis and for evaluation of reactor performance at specific time steps over the fuel cycle. The Monte Carlo analysis was carried out using the MCNP continuous-energy code, and the few- group diffusion theory calculations were performed using the VENTURE and PDQ code systems. The MCNP code was used primarily for its capability to model the reflector components in realistic geometries as well as the inherent circumvention of cross-section processing requirements and use of energy-collapsed cross sections. The MCNP code was used for evaluations of reflector component reactivity effects and of heat loads in these components. The code was also used as a benchmark comparison against the diffusion-theory estimates of key reactor parameters such as region fluxes, control rod worths, reactivity coefficients, and material worths. The VENTURE and PDQ codes were used to provide independent evaluations of burnup effects, power distributions, and small perturbation worths. The performance and safety calculations performed over the subject time period are summarized, and key results are provided. The key results include flux and power distributions over the fuel cycle, silicon production rates, fuel burnup rates, component reactivities, control rod worths, component heat loads, shutdown reactivity margins, reactivity coefficients, and isotope production rates.

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197 p.

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INIS; OSTI as DE96006234

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  • Other Information: PBD: Aug 1995

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  • Other: DE96006234
  • Report No.: ORNL/TM--12966
  • Grant Number: AC05-84OR21400
  • DOI: 10.2172/206382 | External Link
  • Office of Scientific & Technical Information Report Number: 206382
  • Archival Resource Key: ark:/67531/metadc664360

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  • August 1, 1995

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  • June 29, 2015, 9:42 p.m.

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  • Jan. 20, 2016, 4:16 p.m.

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Gehin, J.C.; Worley, B.A.; Renier, J.P.; Wemple, C.A.; Jahshan, S.N. & Ryskammp, J.M. Reactor physics methods, models, and applications used to support the conceptual design of the Advanced Neutron Source, report, August 1, 1995; Tennessee. (digital.library.unt.edu/ark:/67531/metadc664360/: accessed August 22, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.