Stochastic semi-implicit substep method for coupled depletion Monte-Carlo codes

Coupled Monte Carlo burnup codes aim to evaluate the time evolution of different parameters, such as nuclide densities, for accurate modeling of the different reactor designs and associated fuel cycles. Recently a major deficiency in numerical stability of existing Monte Carlo coupling schemes was identified. Alternative, stable coupling schemes were derived, implemented and verified. These methods are iterative and rely on either the end- or middle-of-step (MOS) reaction rates to evaluate the end-of-step (EOS) nuclide densities. Here, we demonstrate that applying the EOS methods for realistic problems may lead to highly inaccurate results. Considerable improvement can be made by adopting MOS method but the accuracy may still be insufficient. The solution proposed in this work relies on the substep method that allows reducing the time discretization errors. The proposed and tested substep method also assumes that the reaction rates are linear functions of the logarithm of the nuclide densities. The method was implemented in BGCore code and subsequently used to perform a series of test case calculations. The results demonstrate that better accuracy and hence efficiency can be achieved with negligible additional computational burden.This is the author accepted manuscript. The final version is available from Elsevier via http://dx.doi.org/10.1016/j.anucene.2016.01.02

Sub-step methodology for coupled Monte Carlo depletion and thermal hydraulic codes

The governing procedure in coupled Monte Carlo (MC) codes relies on discretization of the simulation time into time steps. Typically, the MC transport solution at discrete points will generate reaction rates, which in most codes are assumed to be constant within the time step. This assumption can trigger numerical instabilities or result in a loss of accuracy, which, in turn, would require reducing the time steps size. This paper focuses on reducing the time discretization error without requiring additional MC transport solutions and hence with no major computational overhead. The sub-step method presented here accounts for the reaction rate variation due to the variation in nuclide densities and thermal hydraulic (TH) conditions. This is achieved by performing additional depletion and TH calculations within the analyzed time step. The method was implemented in BGCore code and subsequently used to analyze a series of test cases. The results indicate that computational speedup of up to a factor of 10 may be achieved over the existing coupling schemes.This is the author accepted manuscript. The final version is available from Elsevier via http://dx.doi.org/10.1016/j.anucene.2016.05.03

Disorder and Interaction in 2D: Exact diagonalization study of the Anderson-Hubbard-Mott model

We investigate, by numerically calculating the charge stiffness, the effects of random diagonal disorder and electron-electron interaction on the nature of the ground state in the 2D Hubbard model through the finite size exact diagonalization technique. By comparing with the corresponding 1D Hubbard model results and by using heuristic arguments we conclude that it is \QTR{it}{unlikely} that there is a 2D metal-insulator quantum phase transition although the effect of interaction in some range of parameters is to substantially enhance the non-interacting charge stiffness.Comment: 13 pages, 2 figures Revised version. Accepted for publication in Phys. Rev. Let

Thorium-based plutonium incineration in the I2^2S-LWR

This paper presents an analysis of a homogeneous thorium-plutonium fuel cycle developed for the Integral Inherently Safe LWR (I$^2$S-LWR). The I$^2$S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. Its baseline fuel and cladding materials are U$_3$Si$_2$ and advanced FeCrAl steel, respectively. The advanced steel cladding can withstand longer exposure periods with significantly lower degradation rates compared to traditional Zr-based alloys. However, longer fuel cycles would require higher fuel enrichment, and this is currently limited to 5$^w$% in the I$^2$S-LWR. Therefore, an alternative thorium-plutonium mixed oxide (TOX) fuel cycle is investigated. In principle, the TOX fuel cycle has no fissile content limitation and becomes even more attractive for long irradiation periods, due to the efficient build-up of $^{233}$U, which increases its cumulative energy share and hence decreases the initial Pu requirements per unit of energy produced by the fuel. Current Pu recycling practice in the form of U–Pu mixed oxide (MOX) fuel is not well-suited for Pu disposition due to continuous Pu production from $^{238}$U. This study compares the TOX and MOX cores in terms of efficiency of Pu disposition. The results show that the burnt Pu fraction in the TOX cycle is much higher, and could be further enhanced for longer irradiations (100 MWd/kg or more).Engineering and Physical Sciences Research Council (Grant ID: EP/K033611/1

Screening the design space for optimized plutonium incineration performance in the thorium-based I² S-LWR

This paper presents an optimization of a thorium-plutonium (Th-Pu) fuel cycle by screening various design options for the Integral Inherently Safe Light Water Reactor (I²S-LWR). The I²S-LWR is an advanced 2850 MWt integral pressurized water reactor with enhanced safety beyond that of Gen-III+ reactors. The features of this reactor, such as material choice, make it attractive for alternative fuel cycles including the use of thorium. Recently, the feasibility of the Th-Pu cycle was studied and the benefits associated with it were demonstrated. More specifically, the Pu incineration performance was enhanced by adopting multi-batch (i.e. more than 3-batch) schemes and extended burnup (above 100 MWd/kg). The optimized design with the most favorable loading pattern was obtained by applying the Simulated Annealing optimization technique. This paper demonstrates further plausible modifications to the Th-Pu cycle design that may enhance its performance considerably. The paper seeks to identify the contributory factors, such as cladding types, plutonium vectors and initial plutonium loadings, with major impact on the incineration performance. The postirradiation characteristics are also analyzed and suggest that such a cycle may simplify the design and operation of the waste repository

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

Coupled 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

Explicit decay heat calculation in the nodal diffusion code DYN3D

The residual radioactive decay heat plays an important role in some accident scenarios and, therefore, needs to be accurately calculated when performing accident analyses. The current reactor simulation codes used for accident analysis account for the residual decay heat by means of simplified models. Typically, these models rely on semi-empirical correlations which are defined over a limited range of burnup and fuel types. Therefore, the applicability of such correlations is limited and any deviation from the definition range may lead to high uncertainties, which is detrimental in evaluating safety margins. Reactor dynamic code DYN3D was originally developed for transient and accident analysis of LWRs. In DYN3D, the residual radioactive decay heat calculation is based on the German national standard DIN Norm 25463 model. The applicability of this model is limited to a low enriched uranium dioxide fuel for light water reactors. This paper describes a new general decay heat calculation model implemented in DYN3D. The radioactive decay rate of each nuclide in each spatial node is calculated by recently implemented depletion module and the cumulative released heat is used to obtain the spatial distribution of the decay power for every time step. Such explicit approach is based on first principles and is free from approximations and, thus, can be applied to any reactor system (e.g. thermal and fast) and fuel type. The proposed method is verified through code-to-code comparison with the Serpent 2 Monte Carlo code results

Hybrid microscopic depletion model in nodal code DYN3D

The paper presents a general hybrid method that combines the micro-depletion technique with correction of micro- and macro- diffusion parameters to account for the spectral history effects. The fuel in a core is subjected to time- and space-dependent operational conditions (e.g. coolant density), which cannot be predicted in advance. However, lattice codes assume some average conditions to generate cross sections (XS) for nodal diffusion codes such as DYN3D. Deviation of local operational history from average conditions leads to accumulation of errors in XS, which is referred as spectral history effects. Various methods to account for the spectral history effects, such as spectral index, burnup-averaged operational parameters and micro-depletion, were implemented in some nodal codes. Recently, an alternative method, which characterizes fuel depletion state by burnup and ²³⁹Pu concentration (denoted as Pu-correction) was proposed, implemented in nodal code DYN3D and verified for a wide range of history effects. The method is computationally efficient, however, it has applicability limitations. The current study seeks to improve the accuracy and applicability range of Pu-correction method. The proposed hybrid method combines the micro-depletion method with a XS characterization technique similar to the Pu-correction method. The method was implemented in DYN3D and verified on multiple test cases. The results obtained with DYN3D were compared to those obtained with Monte Carlo code Serpent, which was also used to generate the XS. The observed differences are within the statistical uncertainties.This is the author accepted manuscript. The final version is available from Elsevier via http://dx.doi.org/10.1016/j.anucene.2016.02.01

Long-lived fluorescence of homopolymeric guanine–cytosine DNA duplexes

International audienceThe fluorescence spectrum of the homopolymeric double helix poly(dG)·poly(dC) is dominated by emission decaying on the nanosecond time-scale, as previously reported for the alternating homologue poly(dGdC)·poly(dGdC). Thus, energy trapping over long periods of time is a common feature of GC duplexes which contrast with AT duplexes. The impact of such behaviour on DNA photodamage needs to be evaluated