ADVANTG An Automated Variance Reduction Parameter Generator

Mosher, Scott W.; Bevill, Aaron M.; Johnson, S. R.; Ibrahim, Ahmad; Daily, Charles R.; Evans, Thomas M.; Wagner, John C.; Johnson, J.O.

doi:10.2172/1105937

Cited by 80 publications

(74 citation statements)

References 14 publications

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“…Details of the FW-CADIS method can be found in the references. In this study we apply the FW-CADIS technique in the ADVANTG code [7] to develop weight windows and consistently biased sources for use in the MCNP radiation transport code [8]. We then use MCNP to calculate neutron and photon dpa rates in the "standard" dosimetry locations and subsequently for an axial section of the entire reactor vessel using a detailed cylindrical mesh tally.…”

Section: Epj Web Of Conferencesmentioning

confidence: 99%

Analysis of dpa Rates in the HFIR Reactor Vessel using a Hybrid Monte Carlo/Deterministic Method*

Risner

Blakeman

2016

EPJ Web of Conferences

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Abstract. The Oak Ridge High Flux Isotope Reactor (HFIR), which began full-power operation in 1966, provides one of the highest steady-state neutron flux levels of any research reactor in the world. An ongoing vessel integrity analysis program to assess radiation-induced embrittlement of the HFIR reactor vessel requires the calculation of neutron and gamma displacements per atom (dpa), particularly at locations near the beam tube nozzles, where radiation streaming effects are most pronounced. In this study we apply the Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) technique in the ADVANTG code to develop variance reduction parameters for use in the MCNP radiation transport code. We initially evaluated dpa rates for dosimetry capsule locations, regions in the vicinity of the HB-2 beamline, and the vessel beltline region. We then extended the study to provide dpa rate maps using three-dimensional cylindrical mesh tallies that extend from approximately 12 in. below to approximately 12 in. above the height of the core. The mesh tally structures contain over 15,000 mesh cells, providing a detailed spatial map of neutron and photon dpa rates at all locations of interest. Relative errors in the mesh tally cells are typically less than 1%.

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Section: Epj Web Of Conferencesmentioning

confidence: 99%

Analysis of dpa Rates in the HFIR Reactor Vessel using a Hybrid Monte Carlo/Deterministic Method*

Risner

Blakeman

2016

EPJ Web of Conferences

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“…The MCNP ex-core model includes ex-core structures, such as: reactor vesel, ex-core neutron detectors and all the surrounding concrete structures. Using hybrid code ADVANTG [10] weight windows for neutron transport will be generated and used in the MCNP simulation to study the ex-core detector response and neutron dose field in the containment.…”

Section: Calculation Proceduresmentioning

confidence: 99%

Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

et al. 2020

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An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

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“…The synthesis method remains the de facto state-of-the-art method used in nuclear industry for calculating neutron fluxes for RPV surveillance programs. However, a 3-D deterministic transport codes (Rhoades (1997) [10]; Evans et al (2010) [11]) or hybrid deterministic Monte Carlo methods (Mosher et al (2013) [12]) are available and could be used in future analyses.…”

Section: Radiation Transport Simulationsmentioning

confidence: 99%

Characterization of Radiation Fields in Biological Shields of Nuclear Power Plants for Assessing Concrete Degradation*

Remec

Rosseel

Field

et al. 2016

EPJ Web of Conferences

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Abstract. Life extensions of nuclear power plants to 60 and potentially 80 years of operation have renewed interest in long-term material degradation. One material being considered is concrete, with a particular focus on radiation-induced effects. Based on the projected neutron fluence values (E > 0.1 MeV) in the concrete biological shields of the US pressurized water reactor fleet and the available data on radiation effects on concrete, some decrease in mechanical properties of concrete cannot be ruled out during extended operation beyond 60 years. An expansion of the irradiated concrete database and a reliable determination of relevant neutron fluence energy cutoff value are necessary to ensure reliable risk assessment for extended operation of nuclear power plants.

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ADVANTG An Automated Variance Reduction Parameter Generator

Cited by 80 publications

References 14 publications

Analysis of dpa Rates in the HFIR Reactor Vessel using a Hybrid Monte Carlo/Deterministic Method*

Analysis of dpa Rates in the HFIR Reactor Vessel using a Hybrid Monte Carlo/Deterministic Method*

Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

Characterization of Radiation Fields in Biological Shields of Nuclear Power Plants for Assessing Concrete Degradation*

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