Flattening of the Power Distribution in the HTGR Core with Structured Control Rods

Control rods (CRs) have a significant influence on reactor performance. Withdrawal of a control rod leaves a region of the core significantly changed due to lack of absorber, leading to increased fission rate and later to Xe135 buildup. In this paper, an innovative concept of structured control rods made of tungsten is studied. It is demonstrated that the radial division of control rods made of tungsten can effectively compensate for the reactivity loss during the irradiation cycle of high-temperature gas-cooled reactors (HTGRs) with a prismatic core while flattening the core power distribution. Implementation of the radial division of control rods enables an operator to reduce this effect in terms of axial power because the absorber is not completely removed from a reactor region, but its amount is reduced. The results obtained from the characteristic evolution of the reactor core for CRs with a structured design in the burnup calculation using the refined timestep scheme show a very stable core evolution with a reasonably low deviation of the power density and Xe135 concentration from the average values. It is very important that all the distributions improve with burnup.

Preliminary Analysis on Burnup Calculation of Several Arrangement of TRISO and Pebble inside an MCNP Model of HTR Core

Several studies related to simplifying the modeling of pebble bed High-Temperature Reactor core (HTR) has been developed before. From some calculation on several MCNP models with a fueled pebble to dummy ratio 57:43, using a combination of several types of TRISO (TRi-structural ISOtropic particle fuel) unit and Pebble unit is modeled to achieve its first criticality. In this paper, some MCNP model that uses 27000 pebbles with a 57:43 ratio and 100% fueled pebble is created to be used on burnup calculation and to compare its k-eff and nuclide inventory. From this burnup calculation, it could be seen that SC (Simple Cubic) TRISO unit has faster calculation time followed by the HCP (Hexagonal Close Packed) TRISO unit and then the FCC (Face-Centered Cubic) TRISO unit. The BCC (Body-Centered Cubic) pebble unit had some consistent deviation from another pebble unit, and it still needs more study to know more about the reason behind it. It could be seen that if there are some dummy pebbles inside the reactor, then the deviation would be higher than if there is just fueled pebble inside the reactor. On the 57:43 ratio, the absolute average deviation of k-eff on burnup calculation is lower than 2% and 10% for nuclide inventory (mass). On 100% fueled pebble, it’s below 0.15% on k-eff absolute deviation and below 8% on nuclide inventory deviation.

Burnup calculation of the OECD VVER-1000 LEU benchmark assembly using MCNP6 and SRAC2006

The present work aims to perform burnup calculation of the OECD VVER-1000 LEU (lowenriched uranium) computational benchmark assembly using the Monte Carlo code MCNP6 and the deterministic code SRAC2006. The new depletion capability of MCNP6 was applied in the burnup calculation of the VVER-1000 LEU benchmark assembly. The OTF (on-the-fly) methodology of MCNP6, which involves high precision fitting of Doppler broadened cross sections over a wide temperature range, was utilized to handle temperature variation for heavy isotopes. The collision probability method based PIJ module of SRAC2006 was also used in this burnup calculation. The reactivity of the fuel assembly, the isotopic concentrations and the shielding effect due to the presenceof the gadolinium isotopes were determined with burnup using MCNP6 and SRAC2006 incomparison with the available published benchmark data. This study is therefore expected to reveal the capabilities of MCNP6 and SRAC2006 in burnup calculation of VVER-1000 fuel assemblies.

Evaluation of Serpent Capabilities for Hyperfidelity Depletion of Pebble Bed Cores

Accurate burnup calculation in pebble bed reactor cores is today necessary but challenging. The continuous advancement in computing capabilities make the use of Monte Carlo transport codes possible to efficiently study individual pebbles depletion without making strong assumptions. The purpose is to eliminate unnecessary typical assumptions made in existing codes, while being flexible and suitable for commonly available computing machines. Among the available codes, Serpent 2 provides extremely useful tools to make pebble beds modeling and simulation efficient. The explicit stochastic geometry definition handles irregular pebble beds with comparable performances to regular lattices. Optimization modes controlling the use of unionized energy grids, cross-sections pre-calculation and flux calculation through spectrum collapse or direct tally lead to high flexibility and optimal memory usage while limiting calculation time. Automated burnable materials division is a useful tool to lower the memory requirements while quickly generating the geometry and materials. Finally, parallelization and domain decomposition allow for decreasing unreasonable memory constraints for large cores. This work thus explores the possibilities of Serpent 2 when applying depletion in pebble beds, compares the optimization modes and quantifies the simulation time and memory usage depending on the conditions of the calculation. Overall, the results show that Serpent 2 is well adapted to the use of small to large cores calculations with commonly available resources.

Cross Section and Fission Yields Induced Uncertainty in the Vver-440 Burnup Calculation

The properties of nuclear fuel depend on the actual isotopic composition which develops during a reactor operation. In practice, the prediction accuracy of burnup calculations serves as the basis for the future precise estimation of a core lifetime and other safety-based core characteristics. The present study quantifies nuclear data induced uncertainties of nuclide concentrations and multiplication factors in VVER-440 fuel depletion analysis. The well-known SCALE system and the TRITON sequence are used with the NEWT deterministic solver in the SAMPLER module that implements stochastic techniques to assess the uncertainty in computed results. The propagation of uncertainties in neutron cross section and fission yields is studied through the depletion calculation of 2D heterogeneous VVER-440 fuel assembly with an average enrichment of 4.87 wt % of 235U and six gadolinium rods with 3.35 % of Gd2O3. In the paper, fixed nominal depletion conditions are based on the real operational data of the Slovak NPP Bohunice unit 4 during cycle 30. In total 250 cases with uncertain parameters are computed and the results are evaluated by an auxiliary tool.

MPACT VERIFICATION WITH MAGNOX REACTOR NEUTRONICS PROGRESSION PROBLEMS

MPACT is a state-of-the-art core simulator designed to perform high-fidelity analysis using whole-core, three-dimensional, pin-resolved neutron transport calculations on modern parallel computing hardware. MPACT was originally developed to model light water reactors, and its capabilities are being extended to simulate gas-cooled, graphite-moderated cores such as Magnox reactors. To verify MPACT’s performance in this new application, the code is being formally benchmarked using representative problems. Progression problems are a series of example models that increase in complexity designed to test a code’s performance. The progression problems include both beginning-of-cycle and depletion calculations. Reference solutions for each progression problem have been generated using Serpent 2, a continuous-energy Monte Carlo reactor physics burnup calculation code. Using the neutron multiplication eigenvalue ke_ as a metric, MPACT’s performance is assessed on each of the progression problems. Initial results showed that MPACT’s multigroup cross section libraries, originally developed for pressurized water reactor problems, were not sufficient to accurately solve Magnox problems. MPACT’s improved performance on the progression problems is demonstrated using this new optimized cross section library.

FISSION YIELD UNCERTAINTY PROPAGATION IN MULTI-PASS REFUELING PEBBLE-BED HTGR

Multi-pass refueling scheme is a highlighted feature of pebble bed HTGR which spatially mixes the burnup calculation inside core. Such refueling scheme relate burnup calculation in one region of the core to others and thus affects the uncertainty propagation of nuclear data, e.g. fission product yield. In this work, thermal neutron induced U-235 fission product yield uncertainties are propagated in HTR-PM models with various refueling schemes in V.S.O.P. code. And the effect of multi-pass refueling scheme is studied. Bayesian method is applied to estimate the covariance of fission product yield based on ENDF/B-VII.1 fission yield sub-library. Uncertainty quantification is performed with stochastic sampling method and log-normal based correlated sampling method is used to generate reasonable and self-consistent fission product yield samples. The analyzed results indicate that multi-pass refueling scheme could affect the uncertainty propagation of reactor local responses.