Full Core Research Articles

Criticality Analysis for BWR Spent Fuel Based on the Burnup Credit Evaluation from Full Core Simulations

This study performed criticality analysis for the GBC-68 storage cask loaded with boiling water reactor (BWR) spent fuel at the discharged burnups obtained from the full-core simulations. The analysis was conducted for: (1) different reloading scenarios; (2) target burnups; and (3) two fuel assembly types—GE14 and SVEA100—to estimate the impact each of the three factors has on the cask reactivity. The BWR spent fuel composition was estimated using the results of the nodal analysis for the advanced boiling water reactor (ABWR) core model developed in this study. The nodal calculations provided realistic operating data and axial burnup and coolant density profiles, for each fuel assembly in the reactor core. The estimated cask’s keff were compared with the fresh fuel and peak reactivity standards to identify the benefit of the burnup credit method applied to the BWR spent fuel at their potential discharge burnups. The analysis identified the significant cask criticality reduction from employing the burnup credit approach compared to the conventional fresh fuel approach. However, the criticality reduction was small compared to the peak reactivity approach, and could even disappear for low burnt fuel assemblies from non-optimal reloading patterns. In terms of cask manufacturing, the potential financial benefit from using the burnup credit approach was estimated to be USD 3.3 million per reactor cycle.

3D SN and Monte Carlo calculations of the Utah TRIGA reactor core using PENTRAN and MCNP6

We present a systematic and detailed approach to simulate the University of Utah TRIGA Reactor (UTR) in support of core criticality calculations to profile the entire reactor in detail. In performing this work, we utilized both a 3-D Cartesian SN deterministic code, PENTRAN, and a Monte Carlo code, MCNP6, to calculate complimentary, high accuracy 3-D transport derived neutron flux distributions at reactor full power. For deterministic models, our study shows that a 14-group CONTRIBUTON weighted multi-group cross-section library with up-scattering compared extremely well with continuous energy Monte Carlo results. We also modeled results to confirm the activity of an activated arsenic sample placed in the heavy water moderated thermal irradiation chamber of the UUTR (University of Utah TRIGA Reactor).A 138 pcm difference in the system eigenvalue for the full 3-D core PENTRAN discrete ordinates (SN) UUTR model utilizing 2×109 equations, in 14 energy groups, is compared to the reference case of a similar full core MCNP6 Monte Carlo calculation with continuous energy cross-sections. On average, deterministic and Monte Carlo core neutron group flux values differ by less than 1%, with some local maximum relative differences between 10% and 20% for the 3-D core flux distribution in some energy groups are observed. The differences of neutron fluxes from group 3/14 to group 10/14, with energies spanning from 1.11 eV to 13.8 MeV, are comparable to within 5% in the UUTR core active fuel region. For energy groups with energies greater than 13.8 MeV, MCNP6 is not able to resolve the flux distributions with reasonable statistical uncertainties due to the extremely low sampling probability for neutrons in those energies; orders of magnitude more histories would be required to do so. In contrast, the PENTRAN 3D SN calculation shows a very detailed flux distribution for energies larger than 13.8 MeV, highlighting the complimentary utility of applying both methods. Detailed approaches for the PENTRAN calculation, including energy group collapsing, homogenization of fuel + gap + cladding, and analysis are presented in the narrative.

Fuel Assembly Design Study for Modular Gas Cooled Fast Reactor using Monte Carlo Parallelization Method

As one of the advanced reactors new generation concept, Gas-cooled Fast Reactors can achieve higher electricity efficiency compare to the previous generation. The modular design is perfectly matched with Indonesia region that has many small islands. In this paper will be discussed about the design study of the fuel assembly for modular nuclear power plants using helium as cooling (GFR). The design reactors with hexagonal assembly and square assembly as a comparison to see which assembly configuration in GFR to get optimum results. Calculation for neutronic used Monte Carlo method with parallelization to accelerate computing time. Using OpenMC program code to build full three-dimension core simulation and neutronic calculation with nuclear data ENDF/B-VIII.b5 as a library. The neutronics results show the optimum design can be achieved.

Neutronics analysis of PWR core using DRAGON, TRIVAC, and DONJON computer codes

Operation of NPP has potential of nuclear hazard and radiation hazard. Therefore, all efforts are absolutely needed to be performed to ensure the safety of NPP operation. Neutronics analysis has important role in ensuring that safety aspect. In neutronics analysis, computer codes that could well simulate the neutronics condition in reactor core are needed. In this paper, neutronics analysis of 2652 MWt PWR core has been performed using DRAGON, TRIVAC, and DONJON computer codes. The purpose of simulation is to conduct the calculation of neutronics parameters which are important for safety. The neutronics analyses was conducted in two stages, i.e. cell calculation using DRAGON and full core calculation using TRIVAC and DONJON. For verification, the comparation between the calculation results of DRAGON, TRIVAC, and DONJON, and those of other computer codes results found in literature was performed. The effective multiplication factors calculated by DONJON for conditions where all control rods were fully up and fully down, and for ‘one stuck rod’ condition were 1.23399, 0.69705, and 0.91340, respectively. The core maximum power and average power were found to be 21166.1 kW and 16891.7 kW, respectively, with the power peaking factor of 1.2530. The maximum and average neutron fluxes in the core were 4.019038E+14 n/cm2s and 9.035312E+13 n/cm2s, respectively. The results of fuel cell calculation using DRAGON and the full core calculation using TRIVAC and DONJON have the differences of 4% and 1.3%, respectively, compared to the calculation results using other computer codes. This research shows that DRAGON, TRIVAC, and DONJON have capabilities to perform the calculation of neutronics parameters for NPP, specifically PWR type.

Burnup computations of multi-pass fuel loading scenarios in HTR-10 using a pre-generated fuel composition library

This study aims to precisely investigate the core characteristics of HTR-10 using an online-refueling fuel loading scenario through an innovative approach that incorporates a pre-generated fuel composition library. HTR-10 is an experimental high-temperaturegas- cooled pebble-bed reactor constructed in China. Applying multi-pass fuel loading scenario, the pebbles in HTR-10 pass through the core several times via onlinerefueling. In this study, we proposed a model that simulates the dynamic operation of HTR-10 by using continuously static calculations. The model was constructed by dividing the core into 3 flow channels with different amounts of horizontal zones according to the residence time of fuel pebbles. However, random distribution of the pebbles with various burnup values complicates the refueling process. In order to simulate online-refueling more conveniently, a comprehensive method based on multipass fuel loading scenarios was thus developed. Firstly, a pre-generated fuel composition library containing the content of both actinides and fission products of fuel pebbles as a function of burnup, was built by implementing full core depletion calculations. During the refueling processes, a specific ratio of discharged fuel pebbles to fresh fuel pebbles was reloaded into the top zones. Thus, the average burnup values in the top zones can be resulted by considering the supplement of fresh fuel pebbles. Eventually, the fuel composition in the top zones were obtained by interpolating the average burnup value from the pre-generated library. In this study, all the calculations were performed using the MCNP6 computer code together with the ENDF/B-VII continuous energy neutron data library. The results revealed that the core reached a state of equilibrium after seven refueling processes with an increased rate of volume fraction. Our findings correspond to those of other studies. Another model constructed by locating the actual distribution of fresh pebbles also confirms the verification of the refueling method. Eventually, the burnup characteristics as well as the power and flux distributions in the HTR-10 core were analyzed using the simulation method.

Analysis of the excess reactivity and control rod worth of RSG-GAS equilibrium silicide core using Continuous-Energy Monte Carlo Serpent2 code

This present work is a continuation of our previous work to verify the RSG-GAS equilibrium silicide core parameters. Earlier, the burnup of 22 fuel elements in the 88th equilibrium core has been measured, and it showed a good agreement between the experimental and calculation results. In this paper, several important neutronics parameters documented from the experiment at the same core, such as the excess reactivity, integral control rod worth, and total control rod worth, are confirmed using the full core calculation results of Monte Carlo Serpent2 code in conjunction with ENDF/B-VII.1 and the most recent ENDF/B-VIII.0 nuclear data libraries. The excess reactivity and integral control rod worth were measured by the positive and negative reactivity compensation method. It is worth mentioning that the measured integral control rod worth hasn’t been verified because the in-core fuel management code of RSG-GAS (BATAN-FUEL) solves only for a 2-dimensional configuration. The fuel element burnup at the end of cycle calculated by BATAN-FUEL and Serpent2 is also compared and discussed. The 3-D Serpent2 model of RSG-GAS first core is modified, and several improvements in the models are proposed, including the explicit modeling of 6 neutron beam tubes in the beryllium block reflector. It is found that the calculated integral control rod worth has a good agreement with the experiment, while Serpent2 overestimates the excess reactivity by a maximum of 661 pcm. The impact of nuclear data is insignificant for the selected core parameters.

NEUTRONIC BENCHMARK ON HOLOS-QUAD MICRO-REACTOR CONCEPT

The Holos-Quad micro-reactor concept is proposed by HolosGen LLC for civilian applications to generate 22 MWt with a lifetime of approximately 8 effective full power years (EFPYs). The design is based on a very innovative high-temperature gas-cooled reactor concept using four Subcritical Power Modules (SPMs) that fit into one commercial 40-foot transport ISO container. Neutronics benchmarks were developed based on a preliminary version of the Holos-Quad design to confirm feasibility of the neutronics design of this non-traditional high temperature gas-cooled (HTGR)-type micro-reactor concept. Calculations were performed using Monte Carlo codes (SERPENT and OpenMC) as well as the PROTEUS high-fidelity deterministic code for two exercises based on a unit cell model and a full core model. The results obtained showed good agreement with all of the evaluated parameters for the unit cell problem and the full core problem. SERPENT and OpenMC display consistently good agreement in eigenvalue within 150 pcm for the unit cell benchmark and less than 270 pcm for the full core benchmark. PROTEUS eigenvalues showed relatively larger differences but still reasonable agreement with the SERPENT solutions. For the depletion benchmark, the observed eigenvalue differences between SERPENT and OpenMC were within 300 pcm throughout the depletion up to 100 GWd/MT, ensuring that the two codes used equivalent input parameters such as the recoverable heat values for fission reactions. This benchmark exercise confirms the neutronic feasibility and core performance of the preliminary Holos-Quad design.

PIN POWER CALCULATION SCHEME FOR REACTOR PRESSURE VESSEL FAST NEUTRON FLUENCE ESTIMATION

The estimation of the neutron fluence at the Reactor Pressure Vessel (RPV) is classically carried out by a two-step approach. The first step is to estimate the full core neutron source term whether the second step of the calculation consists in the transport of neutrons from the core (source term) to the RPV using the neutron fission distribution determined in the previous step. For this purpose, the neutron fission distribution is to be accurately determined at the fuel pin level for the assemblies on the border of the core. To achieve this goal, two methods are evaluated in this study. The first method considered is a full core 2D Monte Carlo calculation using the MNCP6 code. The second method is based on a deterministic approach using the CASMO5 multi-segment option, allowing a full 2D transport calculation at the pin level with an expected accuracy similar to a stochastic method. The comparison of the two methods shows an overall good agreement with differences within the statistical uncertainty for different cores: homogeneous UOX core, mixed UOX-MOX loading and the effect of the hafnium rods used in the assemblies in the periphery of the core. The modelling limitation and the associated calculational time are discussed for the comparison of the two approaches.

NEUTRONICS ANALYSIS OF VVER-1000 CORE USING AT-FCM FUEL

Using the Fully Ceramic Microencapsulated (FCM) fuel in light water reactors has multiple advantages, as it is accident tolerant because of; no hydrogen generation due to the cladding interaction with steam at high temperature, better retention of fission fragments and proliferation resistant due to very small production of transuranic elements during the burnup as compared to the standard UO2 fuel. In this study neutronics analysis of AT-FCM fuel consisting of TRISO particles embedded in SiC matrix is performed for replacement in existing VVER-1000 reactors. Standard VVER-1000 fuel assembly is transformed to Accident Tolerant Fully Ceramic Microencapsulated (AT-FCM) fuel assembly based on hydraulic diameter of the VVER-1000 assembly, the number of fuel pins are decreased with increased diameter and enrichment to conserve the initial fissile loading in AT-FCM assembly. Fuel centerline temperature of the AT-FCM assembly is found to be lower than the reference UO2 fuel assembly at the same total power produced because of the much higher thermal conductivity. FCM-TRISO fuel assembly namely Array 15 with 169 pins is proposed and analyzed. Pin cell, assembly level and full core calculations have been performed with SERPENT code using implicit and explicit models. VVER-1000 full core is modelled using the transformed FCM assembly. The embedded TRISO particles in a SiC matrix and the use of FeCrAl cladding turns out to be the perfect case for accident tolerance. High burnup of AT-FCM core in terms of MWd/kgHM for the same number of EFPDs is observed as compared to reference UO2 core due to the small breeding of transuranic elements Pu-239, Pu-240 and Pu-241. Appreciable quantity of the power is produced due to the fission of transuranic elements in reference UO2 assembly so the burnup in MWd/kgHM remains smaller than the AT-FCM fuel. Comparatively more softening of spectrum is found in AT-FCM fuel cells and assemblies towards the middle of the cycle (MOC) and End of the Cycle (EOC), this softening of spectrum tends to increase the rate of U-235 depletion. Very small quantities of plutonium isotopes are produced in AT-FCM as compared to the reference UO2 assembly because of small loading of U-238 at the BOC. The neutronics performance of AT-FCM core with burnable poison consisting of Gd2O3 and Er2O3 turn out to be better than reference UO2 assembly as it exhibits smooth burnup. Fuel Temperature Coefficient (FTC) and Moderator Temperature Coefficient (MTC) of the AT-FCM assembly is negative for most part of the cycle however, towards the end of cycle it becomes less negative due to small quantities of resonance absorbers, softening of thermal flux and increased rate of fission absorption in UO2.

Analysis for the ARIANE BM1 and BM3 samples: nuclide inventory and decay heat

The Mixed Oxide samples (MOX) ARIANE Post Irradiation Examination samples BM1 and BM3 have been analyzed in this work, based on various two- and three-dimensional models. Calculated and measured nuclide inventories are compared based on CASMO5, SIMULATE and SNF simulations, and calculated values for the decay heat of the assembly containing the samples are also provided. For uncertainty propagation, the covariance information from three different nuclear data libraries are used. Uncertainties from manufacturing tolerances and operating conditions are also considered. The results from these two samples are compared with the ones from two UO2 samples, namely GU1 and GU3, also from the ARIANE program, applying the same calculation scheme and uncertainty assumptions. It is shown that a two-dimensional assembly model provides better agreement with the measurements than a two-dimensional single pin model, and that the full core three-dimensional model provides similar results compared to the assembly model, although no 148Nd normalization is applied for the full core model. For the MOX assembly decay heat, as expected, heavy actinides have a higher contribution compared to the cases with the UO2 samples; additionally, decay heat uncertainties are moderately smaller in the case of the MOX assembly.

AN IRSN CONTRIBUTION TO THE UAM PROJECT: THERMAL-HYDRAULIC AND NEUTRONIC UNCERTAINTIES PROPAGATION IN A ROD EJECTION, FIRST RESULTS

The paper presents our first results of the exercise III-I-2c from the OECD-NEA UAM-LWR benchmark intended to an elaboration of the methodology of uncertainty propagation. The considered case studied a full PWR core behavior in fast (~0.1 sec) rod ejection transient. According to the benchmark, the core represented a Hot Zero Power state. Authors used brute-force sampling propagating nuclear data and thermo-fluid uncertainties using 3D computational IRSN chain HEMERA. It couples the reactor physics code CRONOS and thermal-hydraulic core code FLICA4. The nuclear data uncertainties were represented in a form of cross sections standard deviations (in percentage of the mean cross sections values) supplied by the UAM team. In addition to the original benchmark, the study includes a case with an increased power peak by supplementary rod ejection, i.e. with higher reactivity. Both the results are similar to what we obtained in the mini-core rod ejection: the power standard deviation follows, in percentage of the mean power, the mean power curve. We split the variance with a direct calculation: once the cross sections are modified and the thermal-hydraulics inputs are kept constant, another time the contrary. The results show that uncertainties dues to nuclear data dominate over ones due to the thermal-flow area. Furthermore, the major contributors in peak-of-power variance lie in a fast group of cross sections.

NUMA-AWARE DATA MANAGEMENT FOR NEUTRON CROSS SECTION DATA IN CONTINUOUS ENERGY MONTE CARLO NEUTRON TRANSPORT SIMULATION

The calculation of macroscopic neutron cross-sections is a fundamental part of the continuous-energy Monte Carlo (MC) neutron transport algorithm. MC simulations of full nuclear reactor cores are computationally expensive, making high-accuracy simulations impractical for most routine reactor analysis tasks because of their long time to solution. Thus, preparation of MC simulation algorithms for next generation supercomputers is extremely important as improvements in computational performance and efficiency will directly translate into improvements in achievable simulation accuracy. Due to the stochastic nature of the MC algorithm, cross-section data tables are accessed in a highly randomized manner, resulting in frequent cache misses and latency-bound memory accesses. Furthermore, contemporary and next generation non-uniform memory access (NUMA) computer architectures, featuring very high latencies and less cache space per core, will exacerbate this behaviour. The absence of a topology-aware allocation strategy in existing high-performance computing (HPC) programming models is a major source of performance problems in NUMA systems. Thus, to improve performance of the MC simulation algorithm, we propose a topology-aware data allocation strategies that allow full control over the location of data structures within a memory hierarchy. A new memory management library, known as AML, has recently been created to facilitate this mapping. To evaluate the usefulness of AML in the context of MC reactor simulations, we have converted two existing MC transport cross-section lookup “proxy-applications” (XSBench and RSBench) to utilize the AML allocation library. In this study, we use these proxy-applications to test several continuous-energy cross-section data lookup strategies (the nuclide grid, unionized grid, logarithmic hash grid, and multipole methods) with a number of AML allocation schemes on a variety of node architectures. We find that the AML library speeds up cross-section lookup performance up to 2x on current generation hardware (e.g., a dual-socket Skylake-based NUMA system) as compared with naive allocation. These exciting results also show a path forward for efficient performance on next-generation exascale supercomputer designs that feature even more complex NUMA memory hierarchies.

TOWARDS ZERO-VARIANCE SCHEMES FOR KINETIC MONTE-CARLO SIMULATIONS

The solution of the time-dependent transport problem for neutrons and precursors in a nuclear reactor is hard to treat in a naive Monte-Carlo framework because of the largely different time scales associated to the prompt-fission chains and to the decay of precursors. The increasing computer power and the development of variance-reduction techniques specific for reactor kinetics have recently unlocked the possibility to calculate reference solutions to the time-dependent transport problem. However, the application of time-dependent Monte Carlo to large systems (i.e., a full reactor core) is still stifled by the enormous computational requirements. In this paper, we formulate the construction of an optimal Monte-Carlo strategy (in the sense that it results in a zero-variance estimator) for a specific observable in time-dependent transport, in analogy with the existing schemes for stationary problems. As far as we are aware, zero-variance Monte-Carlo schemes for neutron-precursor kinetics have never been proposed before. We verify our construction with numerical calculations for a benchmark transport problem.

EFFECT OF THE UNIFORM FISSION SOURCE METHOD ON LOCAL POWER VARIANCE IN FULL CORE SERPENT CALCULATION

One of challenges of the Monte Carlo full core simulations is to obtain acceptable statistical variance of local parameters throughout the whole reactor core at a reasonable computation cost. The statistical variance tends to be larger in low-power regions. To tackle this problem, the Uniform-Fission-Site method was implemented in Monte Carlo code MC21 and its effectiveness was demonstrated on NEA Monte Carlo performance benchmark. The very similar method is also implemented in Monte Carlo code Serpent under the name Uniform Fission Source (UFS) method. In this work the effect of UFS method implemented in Serpent is studied on the BEAVRS benchmark which is based on a real PWR core with relatively flat radial power distribution and also on 3x3 PWR mini-core simulated with thermo-hydraulic and thermo-mechanic feedbacks. It is shown that the application of the Uniform Fission Source method has no significant effect on radial power variance but equalizes axial distribution of variance of local power.

THE McSAFE PROJECT - HIGH-PERFORMANCE MONTE CARLO BASED METHODS FOR SAFETY DEMONSTRATION: FROM PROOF OF CONCEPT TO INDUSTRY APPLICATIONS

The increasing use of Monte Carlo methods for core analysis is fostered by the huge and cheap computer power available nowadays e.g. in large HPC. Apart from the classical criticality calculations, the application of Monte Carlo methods for depletion analysis and cross section generation for diffusion and transport core simulators is also expanding. In addition, the development of multi-physics codes by coupling Monte Carlo solvers with thermal hydraulic codes (CFD, subchannel and system thermal hydraulics) to perform full core static core analysis at fuel assembly or pin level has progressed in the last decades. Finally, the extensions of the Monte Carlo codes to describe the behavior of prompt and delay neutrons, control rod movements, etc. has been started some years ago. Recent coupling of dynamic versions of Monte Carlo codes with subchannel codes make possible the analysis of transient e.g. rod ejection accidents and it paves the way for the simulation of any kind of design basis accidents as an alternative option to the use of diffusion and transport based deterministic solvers. The H2020 McSAFE Project is focused on the improvement of methods for depletion considering thermal hydraulic feedbacks, extension of the coupled neutronic/thermal hydraulic codes by the incorporation of a fuel performance solver, the development of dynamic Monte Carlo codes and the development of methods to handle large depletion problems and to reduce the statistical uncertainty. The validation of the multi-physics tools developed within McSAFE will be performed using plant data and unique tests e.g. the SPERT III E REA test. This paper will describe the main developments, solution approaches, and selected results.

TOWARDS CAD-BASED GEOMETRY MODELLING WITH THE RANDOM RAY METHOD

The Advanced Random Ray Code (ARRC) is a high performance computing application capable of high-fidelity simulations of full core nuclear reactor models. ARRC leverages a recently developed stochastic method for neutron transport, known as The Random Ray Method (TRRM), which offers a variety of computational and numerical advantages as compared to existing methods. In particular, TRRM has been shown to be capable of efficient simulation of explicit three dimensional geometry representations without assumptions about axial homogeneity. To date, ARRC has utilized Constructive Solid Geometry (CSG) combined with a nested lattice geometry which works well for typical pressurized water reactors, but is not performant for the general case featuring arbitrary geometries. To facilitate simulation of arbitrarily complex geometries in ARRC efficiently, we propose performing transport directly on Computer-Aided Design (CAD) models of the geometry. In this study, we utilize the Direct-Accelerated Geometry Monte Carlo (DAGMC) toolkit which tracks particles on tessellated CAD geometries using a bounding volume hierarchy to accelerate the process, as a replacement for ARRC’s current lattice-based accelerations. Additionally, we present a method for automatically subdividing the large CAD regions in the DAGMC model into smaller mesh cells required by random ray to achieve high accuracy. We test the new DAGMC geometry implementation in ARRC on several test problems, including a 3D pincells, 3D assemblies, and an axial section of the Advanced Test Reactor. We show that DAGMC allows for simulation of complex geometries in ARRC that would otherwise not be possible using the traditional approach while maintaining solution accuracy.

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.

A SIMPLIFIED SMAHTR BENCHMARK PROBLEM SET

The Small, Modular Advanced High Temperature Reactor (SmAHTR) is a preconceptual design for a fluoride salt-cooled small modular reactor (SMR) [1]. In this paper, a stylized 2D benchmark problem set has been created based on SmAHTR. Certain gaps and considerations in burnable poison and control rod content were unspecified/undetermined in the preconceptual design, but those gaps were filled for the stylized problem set. With those features, this problem set could then be used for benchmarking neutron transport methods as well as its low order methods in 2D single assembly and full core configurations. For this benchmark set, continuous energy Monte Carlo calculations were performed. Those calculations provided keff values of 0.9459 (± 11 pcm) and 1.1436 (± 12 pcm) in the full core configuration with all the control rods fully inserted and withdrawn, respectively. The single assembly calculations yielded an eigenvalue, kinf, of 0.9987 (± 15 pcm) and 1.2117 (± 15 pcm) with all of the control rods either inserted or removed, respectively. In the full core configuration, the worth of all the control rods and burnable poison particles were determined to be 197.6 (± 0.16) mk and 311.6 (± 0.23) mk, respectively. The corresponding results in the single assembly configurations are 213 (± 0.21) mk and 337.4 (± 0.20) mk, respectively. A near-critical configuration was also determined for the reactor by inserting control rods in some assemblies, thus providing a case with a keff value of 0.9909 (± 12 pcm).

HIGH FIDELITY AND REDUCED ORDER SOLUTIONS TO AN SMR-LEVEL PROGRESSION PROBLEM WITH THE KRAKEN COMPUTATIONAL FRAMEWORK

Kraken is a new computational reactor analysis framework under development at VTT Technical Research Center of Finland Ltd. The framework builds heavily on the new generation of Finnish simulation codes such as the Serpent Monte Carlo code, the Ants nodal neutronics solver and the FINIX fuel behavior module. This paper describes the application of Kraken to its first realistic full core problem, a small modular reactor (SMR) core in its fresh state in order to evaluate control rod worths, shutdown margins and reactivity coefficients for the core.

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.