Critical Buckling Generation of TCA Benchmark by the B1 Theory-Augmented Monte Carlo Calculation and Estimation of Uncertainties

The Korea Atomic Energy Research Institute (KAERI) has developed the DeCART2D 2-dimensional (2D) method of characteristics (MOC) transport code and the MASTER nodal diffusion code and has established its own two-step procedure. For design code licensing, KAERI prepared a critical experiment on the verification and validation (V&V) of the DeCART2D code. DeCART2D is able to perform the MOC calculation only for 2D nuclear fuel systems, such as the fuel assembly. Therefore, critical buckling in the vertical direction is essential for comparison between the results of experiments and DeCART2D. In this study, the B1 theory-augmented Monte Carlo (MC) method was adopted for the generation of critical buckling. To examine critical buckling using the B1 theory-augmented MC method, TCA critical experiment benchmark problems were considered. Based on the TCA benchmark results, it was confirmed that the DeCART2D code with the newly-generated critical buckling predicts the criticality very well. In addition, the critical buckling generated by the B1 theory-augmented MC method was bound to uncertainties. Therefore, utilizing basic equations (e.g., SNU S/U formulation) linking input uncertainties to output uncertainties, a new formulation to estimate the uncertainties of the newly generated critical buckling was derived. This was then used to compute the uncertainties of the critical buckling for a TCA critical experiment, under the assumption that nuclear cross-section data have uncertainties.

ANALYSIS OF THE OECD/NEA SFR BENCHMARK WITH ANTS REDUCED-ORDER NODAL DIFFUSION SOLVER AND THE SERPENT MONTE CARLO CODE

In order to meet modern industrial and scientific demands the Kraken multi-physics platform’s development was recently launched at VTT Technical Research Centre of Finland. The neutronic solver of the framework consists of two calculation chains, providing full core solutions by the Serpent high fidelity code (1) and the AFEN/FENM-based reduced-order diffusion solver called Ants (2) capable of handling square and hexagonal geometries in steady-state. Present work introduces the simulation of a large 3600 MWth Sodium-cooled Fast Reactor (SFR) described within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS) of OECD. Full-core 3D results were obtained by Serpent for carbide- and oxide-fuel cores, moreover group constants were generated for Ants utilizing 2D super-cell and single assembly infinite lattice models of Serpent. The continuous-energy Monte Carlo method provided the reference results for the verification of the reduced-order method. Implementing the spatially homogenized properties, 3D solutions were obtained by Ants as well for both core configurations. Comparison was made between the various core designs and codes based on reactivity feedbacks (Doppler constant, sodium voiding, control rod worth) considering power distributions. Regarding reactivity sensitivity on geometry, axial fuel- and radial core expansion coefficients were evaluated as well.

UNCERTAINTY CHARACTERIZATION FRAMEWORK FOR STEADY-STATE AND TRANSIENT NEUTRONICS SIMULATIONS OF A CANDU REACTOR

This paper reports on the development and testing of a comprehensive few-group cross section input uncertainty library for the NESTLE-C nodal diffusion-based nuclear reactor core simulator. This library represents the first milestone of a first-of-a-kind framework for the integrated characterization of uncertainties in steady-state and transient CANDU reactor simulations. The objective of this framework is to propagate, prioritize and devise a mapping capability for uncertainties in support of model validation of best-estimate calculations. A complete framework would factor both input and modeling uncertainty contributions. The scope of the present work is limited to the propagation of multi-group cross-section uncertainties through lattice physics calculations down to the few-group format, representing the input to the NESTLE-C core simulator, and finally to core responses of interest.

GENERATING ALBEDO BOUNDARY CONDITIONS FOR NODAL DIFFUSION CODES USING SERPENT 2

We use the Serpent Monte Carlo code to produce total and partial albedo boundary conditions that can be used to model the Loviisa NPP VVER-440 core with the nodal neutronics tools of Fortum. The albedo generation process is described in detail. The dependence of the generated albedos on boron content and water density is investigated and a clear distinction is noted in water density dependence between regions containing mostly water and those containing mostly structural materials. The Serpent generated albedos are currently used in production calculations for modeling the Loviisa reactors at Fortum.

HIGH-PERFORMANCE MULTI-BATCH FUEL MANAGEMENTS FOR THE ADVANCED SOLUBLE-BORON-FREE ATOM CORE

The autonomous transportable on-demand reactor module (ATOM), a 450 MWth PWR-type small modular reactor (SMR), is under development at Korea Advanced Institute of Science and Technology (KAIST). The ATOM core is designed for soluble-boron-free and passive autonomous load-following operations by utilizing successfully an advanced reactivity control technology, centrally-shielded burnable absorber (CSBA). To enhance the ATOM core safety, CrAl-coated Zircaloy-4 is adopted as an accident-tolerant-fuel cladding. For a long operational cycle, the reference ATOM core has primarily accomplished with a single-batch fuel management (FM). In this paper, for more flexible operation and enhanced fuel utilization, various multi-batch FMs are investigated while the core performance is maintained in terms of both neutronic and safety aspects. These aspects are refueling pattern, cycle length, burnup reactivity swing, discharge burnup, axial and radial power peaking factor (PPF), total PPF, and temperature coefficients. Several refueling types are examined: In-out (low leakage), out-in (flattened power), and randomly scattered schemes. In addition, new heavy reflector designs, ZrO2 and PbO, are introduced instead of stainless steel reflector for an improved core performance. Moreover, a new CSBA loading pattern is also proposed for an effective reactivity control of multi-batch FM strategy. Numerical results show that with a two-batch FM the cycle length can achieve above 2 years with an average discharge burnup of 40 GWd/tU, while the burnup reactivity swing remains less than 1,200 pcm. On top of that, the coolant and fuel temperature coefficients are highly negative at the beginning of cycle and power profile is comparable to that with the single-batch FM. All calculations in these multi-physics assessments of the ATOM core are performed using a Monte Carlo-diffusion hybrid code system based on Monte Carlo Serpent 2 and nodal diffusion COREDAX codes.

CORE DESIGN STUDY FOR A SMALL MODULAR BOILING WATER REACTOR

The hybrid Small Modular Boiling Water Reactor (SMBWR) is a new conceptual design of BWR-type SMR. The main features of SMBWR include a natural circulation loop in its coolant recirculation system and external superheaters system integrated into the steam cycle. A full core analysis of SMBWR is performed with the nodal diffusion code PANTHER using homogenised constant libraries generated by WIMS. The study compared a number of core geometry configurations and fuel management schemes to suppress excess reactivity throughout fuel depletion. Three options for SMBWR core aspect ratio using the same power density are investigated with the aim to assess the effect on the neutronic and thermal-hydraulic performance of the SMBWR. It is found that the thin and tall core configuration (192 fuel assemblies and 3.60 m) showed the least favourable performance out of the three options as it has the largest core pressure drop and thus requires taller chimney to develop natural circulation.

GENERATION OF NODAL CORE SIMULATOR UTILIZING VERA

Nuclear core simulators based upon few-group nodal diffusion method currently are widely used to predict light water reactor core behavior. Nodal parameters’ input, e.g. cross-sections, discontinuity factors, and pin form factors, are typically generated utilizing lattice physics codes. In doing so, a number of approximations are introduced related to using zero current boundary conditions, 2-D radial geometry, and uniform thermal conditions in coolant and fuel. Usage of full core models with prediction fidelity typical of lattice physics to predict nodal parameters would eliminate these approximations. The VERA code can serve as such a full core model and was so utilized in this work. Via subsequent processing of VERA predictions, for a range of state points, nodal parameters and their functionalization in terms of coolant density, fuel temperature, and soluble poison concentration, were obtained and input to the NESTLE nodal code. By processing VERA predictions using consistent nodal methodologies as used in NESTLE, when using nodal parameters after functionalization based upon All-Rods-Out (ARO) VERA state points, the maximum reactivity and pin power differences between VERA and NESTLE were 2 pcm and 0.003 for ARO core simulations. For rodded core simulations, these maximum differences grew to 58 pcm and 0.04. Increases in differences were determined to be attributed to usage of unrodded nodal parameters generated using the VERA ARO state points whether the core is partially rodded or not, consistent with lattice physics practice. Obtaining unrodded nodal parameters using the VERA rodded state points reduced maximum differences to 2 pcm and 0.003 in pin powers.