A collision history-based approach to sensitivity/perturbation calculations in the continuous energy Monte Carlo code SERPENT

Aufiero, Manuele; Bidaud, A.; Hursin, Mathieu; Leppänen, Jaakko; Palmiotti, G.; Pelloni, S.; Rubiolo, P.

doi:10.1016/j.anucene.2015.05.008

Cited by 99 publications

(54 citation statements)

References 14 publications

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“…Several innovative mathematical methods were developed in the scope of the SUSD3D project, and some of them were later introduced also in other S/U computational sequences (e.g., TSUNAMI [24], SERPENT [25], and MCNP [26]) such as the following:…”

Section: Special Features Developed In Susd3dmentioning

confidence: 99%

SUSD3D Computer Code as Part of the XSUN-2017 Windows Interface Environment for Deterministic Radiation Transport and Cross-Section Sensitivity-Uncertainty Analysis

Kodeli

Slavic

2017

Science and Technology of Nuclear Installations

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A Windows interface XSUN-2017 facilitating the deterministic radiation transport and cross-section sensitivity-uncertainty (S/U) calculation is presented. The package was developed to assist the users in the preparation of input cards, rapid modification, and execution of the complete chain of codes including TRANSX, PARTISN, and SUSD3D, all available from the OECD/NEA Data Bank and RSICC. The objective of this work was to make the input and output handling for these codes as user-friendly as possible, passing information among codes internally. XSUN-2017 allows a user-friendly viewing of results obtained from the PARTISN and SUSD3D programs. The first version of the Windows interface XSUN-2013 was developed in 2013 and submitted to OECD/NEA Data Bank Computer Code Collection and RSICC in early 2014. An updated version, XSUN-2017, will be released in 2017. The package includes also the new version of the SUSD3D code. The XSUN-2017 and SUSD3D code systems and recent improvements and updates are described. Examples of the use and validation are presented, including the S/U intercomparison exercise using the SNEAK-7 benchmarks involving the XSUN-2017 code system comparison with the codes such as TSUNAMI, SERPENT, and MCNP6, and the S/U analysis of the eff and eff parameters for the MYRRHA accelerator driven system (ADS).

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Section: Special Features Developed In Susd3dmentioning

confidence: 99%

SUSD3D Computer Code as Part of the XSUN-2017 Windows Interface Environment for Deterministic Radiation Transport and Cross-Section Sensitivity-Uncertainty Analysis

Kodeli

Slavic

2017

Science and Technology of Nuclear Installations

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“…These rejected collisions were still recorded, and their effects were propagated to the multiplication factor using GPT. This methodology has been benchmarked against deterministic codes in Reference (6).…”

Section: Introductionmentioning

confidence: 99%

“…The sensitivity is then used with the covariance of the cross sections in order to quantify the uncertainty in the multiplication factor. 19 was modified to use GPT to assess the sensitivity of the multiplication factor to the cross sections of specified isotopes, as documented in Reference (6). To summarize, each of the cross sections were artificially inflated.…”

Section: Introductionmentioning

confidence: 99%

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Uncertainty Analysis of the TRU-Burning Thorium-Fueled RBWR Using Generalized Perturbation Theory

Bogetic

Gorman

Aufiero

et al. 2017

Volume 9: Student Paper Competition

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TitleUncertainty analysis of the tru-burning thorium-fueled rbwr using generalized perturbation theory ABSTRACTThe RBWR-TR is a thorium-based reduced moderation BWR (RBWR) with a high transuranic (TRU) consumption rate. It is charged with LWR TRU and thorium, and it recycles all actinides an unlimited number of times while discharging only fission products and trace amounts of actinides through reprocessing losses. This design is a variant of the Hitachi RBWR-TB2, which arranges its fuel in a hexagonal lattice, axially segregates seed and blanket regions, and fits within an ABWR pressure vessel. The RBWR-TR eliminates the internal axial blanket, eliminates absorbers from the upper reflector, and uses thorium rather than depleted uranium as the fertile makeup fuel. This design has been previously shown to perform comparably to the RBWR-TB2 in terms of TRU consumption rate and burnup while providing significantly more margin against critical heat flux.This study examines the uncertainty in key neutronics parameters due to nuclear data uncertainty. As most of the fissions are induced by epithermal neutrons and since the reactor uses higher actinides as well as thorium and 233 U, the cross sections have significantly more uncertainty than in typical LWRs. The sensitivity of the multiplication factor to the cross sections of many actinides is quantified using a modified version of Serpent 2

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“…The significant work by Rief on direct perturbation evaluations by Monte Carlo should be acknowledged [13,14]. Also some more recent work must be acknowledged [15]. The sampling of weighted quantities has been introduced for the evaluation of integral reactor parameters [16,17,18].…”

Section: Introductionmentioning

confidence: 99%

The adjoint neutron transport equation and the statistical approach for its solution

Saracco

Dulla²,

Ravetto³

2016

Eur. Phys. J. Plus

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The adjoint equation was introduced in the early days of neutron transport and its solution, the neutron importance, has ben used for several applications in neutronics. The work presents at first a critical review of the adjoint neutron transport equation. Afterwards, the adjont model is constructed for a reference physical situation, for which an analytical approach is viable, i.e. an infinite homogeneous scattering medium. This problem leads to an equation that is the adjoint of the slowing-down equation that is well-known in nuclear reactor physics. A general closed-form analytical solution to such adjoint equation is obtained by a procedure that can be used also to derive the classical Placzek functions. This solution constitutes a benchmark for any statistical or numerical approach to the adjoint equation. A sampling technique to evaluate the adjoint flux for the transport equation is then proposed and physically interpreted as a transport model for pseudo-particles. This can be done by introducing appropriate kernels describing the transfer of the pseudo-particles in phase space. This technique allows estimating the importance function by a standard Monte Carlo approach. The sampling scheme is validated by comparison with the analytical results previously obtained.

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A collision history-based approach to sensitivity/perturbation calculations in the continuous energy Monte Carlo code SERPENT

Cited by 99 publications

References 14 publications

SUSD3D Computer Code as Part of the XSUN-2017 Windows Interface Environment for Deterministic Radiation Transport and Cross-Section Sensitivity-Uncertainty Analysis

SUSD3D Computer Code as Part of the XSUN-2017 Windows Interface Environment for Deterministic Radiation Transport and Cross-Section Sensitivity-Uncertainty Analysis

Uncertainty Analysis of the TRU-Burning Thorium-Fueled RBWR Using Generalized Perturbation Theory

The adjoint neutron transport equation and the statistical approach for its solution

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