On the use of predictor–corrector method for coupled Monte Carlo burnup codes

Kotlyar, Dan; Shwageraus, Eugene

doi:10.1016/j.anucene.2013.03.034

Cited by 36 publications

(22 citation statements)

References 12 publications

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“…2 below. The geometry and compositions applied in this model are presented in detail in the article by [7] and also very similar model has been applied in the study by [6]. Both descriptions are sufficient for reproducing our results and the model is representative for commercially applied PWR reactors.…”

Section: Test Model Of Pwr Assemblymentioning

confidence: 54%

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Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

2016

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Abstract.Monte Carlo methodology provides reference statistical solution of neutron transport criticality problems of nuclear systems. Estimated reaction rates can be applied as an input to Bateman equations that govern isotopic evolution of reactor materials. Because statistical solution of Boltzmann equation is computationally expensive, it is in practice applied to time steps of limited length. In this paper we show that simple staircase step model leads to underprediction of numerical fuel burnup (Fissions per Initial Metal Atom -FIMA). Theoretical considerations indicates that this error is inversely proportional to the length of the time step and origins from the variation of heating per source neutron. The bias can be diminished by application of predictor-corrector step model. A set of burnup simulations with various step length and coupling schemes has been performed. SERPENT code version 1.17 has been applied to the model of a typical fuel assembly from Pressurized Water Reactor. In reference case FIMA reaches 6.24% that is equivalent to about 60 GWD/t HM of industrial burnup. The discrepancies up to 1% have been observed depending on time step model and theoretical predictions are consistent with numerical results. Conclusions presented in this paper are important for research and development concerning nuclear fuel cycle also in the context of Gen4 systems.

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Section: Test Model Of Pwr Assemblymentioning

confidence: 54%

“…More advanced coupling schemes such as predictorcorrector variants (described by [6]) takes into account the variation of neutron source intensity and reaction rates. The simplest solution is application of an average values from the beginning and end of each step as an effective input for transmutation coefficients.…”

Section: Bias Of Numerical Burnupmentioning

confidence: 99%

Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

2016

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“…Previous studies proposed and investigated many coupling methods to evaluate these effects. Numerical instabilities were observed for example in various explicit approaches (Kotlyar and Shwageraus, 2013). Alternative methods (Kotlyar and Shwageraus, 2014) were suggested to address the stability issues.…”

Section: Discussionmentioning

confidence: 99%

“…Recent studies by Kotlyar and Shwageraus (2013) presented the effect of such coupling scheme choice on numerical stability and accuracy of the results. Therefore, new coupling methods have been developed for MC-burnup applications which also account for the dependence of reaction rates on thermal hydraulic conditions (Kotlyar and Shwageraus, 2014).…”

Section: Introductionmentioning

confidence: 99%

A perturbation-based susbtep method for coupled depletion Monte-Carlo codes

Kotlyar

Aufiero

Shwageraus

et al. 2017

Annals of Nuclear Energy

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a b s t r a c tCoupled Monte Carlo (MC) methods are becoming widely used in reactor physics analysis and design. Many research groups therefore, developed their own coupled MC depletion codes. Typically, in such coupled code systems, neutron fluxes and cross sections are provided to the depletion module by solving a static neutron transport problem. These fluxes and cross sections are representative only of a specific time-point. In reality however, both quantities would change through the depletion time interval. Recently, Generalized Perturbation Theory (GPT) equivalent method that relies on collision history approach was implemented in Serpent MC code. This method was used here to calculate the sensitivity of each nuclide and reaction cross section due to the change in concentration of every isotope in the system. The coupling method proposed in this study also uses the substep approach, which incorporates these sensitivity coefficients to account for temporal changes in cross sections. As a result, a notable improvement in time dependent cross section behavior was obtained. The method was implemented in a wrapper script that couples Serpent with an external depletion solver. The performance of this method was compared with other existing methods. The results indicate that the proposed method requires substantially less MC transport solutions to achieve the same accuracy.Published by Elsevier Ltd.

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“…The depletion solver calculates material changes over a time-step with fixed microscopic reaction rates; this part requires solving the Bateman depletion equations. These two parts are then combined to a burnup calculation and solved in a sequential manner using so-called predictor-corrector algorithm [12,13].…”

Section: Methodsmentioning

confidence: 99%

Neutronic Analysis of a Thorium-Fueled Reduced Moderation Boiling Water Reactor

Falaakh¹,

Agung²,

Harto³

2018

International Journal on Advanced Science, Engineering and Information Technology

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Abstract-A study about Resource Renewable Boiling Water Reactor (RBWR) core, a reduced moderation boiling water reactor that features the breeding ratio larger than 1 was conducted. This study focuses on the neutronic performances of the core and aims to investigate the core sustainability when using thorium as the main fertile fuel. A fuel-self-sustaining core with high burnup set as the design target. 233 U+Th were used as the initial fuel, and the impact of initial fissile ( 233 U) content in the core fissile zone on the core neutronic performances was evaluated. Parameters related to the neutronic performances such as the core burnup, fissile breeding, and fissile inventory ratio (FIR) are considered in this study. From these results, it was confirmed that it is feasible to create a selfsustaining fuel cycle system using thorium fueled RBWR. However, there was a trade-off between the core burnup and fissile breeding that can be a significant challenge in the development of this system. Evaluating the other design variables may be considered to address this challenge. The further study to analyze the safety performances of the core is required to arrive at a safe and reliable reactor system.

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On the use of predictor–corrector method for coupled Monte Carlo burnup codes

Cited by 36 publications

References 12 publications

Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

A perturbation-based susbtep method for coupled depletion Monte-Carlo codes

Neutronic Analysis of a Thorium-Fueled Reduced Moderation Boiling Water Reactor

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