Virtual Environment For Reactor Applications Research Articles

A study on the impact of using a subchannel resolution for modeling of large break loss of coolant accidents

The nuclear industry is investigating the feasibility of transitioning from 18- to 24-month fuel cycles because of the positive impact it would have on the operational costs for the current fleet of light-water reactors. A challenge to making this change is the increased risk of fuel fragmentation, relocation, and dispersal (FFRD) due to the known potential for ceramic fuel to pulverize into fine particles at the higher discharge burnups. Previous work has been performed by the Nuclear Energy Advanced Modeling and Simulation program to assess FFRD risk in high-burnup cores using the BISON fuel performance code and a coarse mesh thermal hydraulics (T/H) solution for a loss-of-coolant accident (LOCA) using the TRACE system T/H code. Because of the importance of the T/H solution for FFRD assessment, this study seeks to investigate the impact of using higher-fidelity subchannel techniques for modeling of the LOCA transient. CTF was used to model a subregion of a high-burnup core that was depleted by the Virtual Environment for Reactor Applications (VERA) multiphysics core simulator. Both coarse-mesh and pin-resolved models were created in CTF, and a consistent coarse-mesh TRACE model was also developed to allow for benchmarking the code results. A large-break loss-of-coolant accident (LBLOCA) reflood transient was simulated using these three models, and results were compared. Results showed some consistent differences between the CTF and TRACE coarse models, including a higher peak cladding temperature (PCT) prediction in CTF and later quenching in CTF; however, the transient clad temperature behavior was similar, and these differences are likely due to post-critical heat flux heat transfer modeling differences and minimum film boiling temperature model differences. The pin-resolved results indicate that the PCT in the lumped model is often under-predicted by as much as 70 °C and that PCT occurs at a different location than the high-power pin in the assembly. The lumped model predicts a difference of 10 °C or less between the average and hot pins in the assembly, whereas the pin-resolved model predicts a range of over 100 °C. These results indicate that higher-fidelity T/H results may have an impact on predicted core behavior during LOCA, which may be important to consider when assessing FFRD risk.

Multiphysics analysis of fuel Fragmentation, Relocation, and dispersal Susceptibility–Part 2: High-Burnup Steady-State operating and fuel performance conditions

The US nuclear industry is pursuing increased cycle lengths and increasing the peak rod-averaged burnup in an effort to increase the economic viability of the US nuclear fleet. Increasing burnup will afford economic viability by enabling utilities to optimize core designs to reduce the number of fresh fuel assemblies per cycle and allow nuclear power plants to operate for a longer period of time. Longer operating periods will also decrease the number of outages experienced by a nuclear power plants and, therefore, offer utilities significant operational savings. However, extending the peak rod-averaged burnup beyond 62 GWd/tU results in operating fuel rods to higher burnup under higher power conditions. This operating regime is expected to result in higher fuel temperatures, fission gas release (FGR), and rod internal pressures (RIPs) that may challenge historical safety basis and affect high-burnup (HBU) experimental testing. In particular, these conditions directly affect fuel fragmentation, relocation, and dispersal (FFRD) susceptibility, so understanding the pretransient operating conditions is critical for developing test plans that evaluate the FFRD and develop strategies to mitigate it. This paper evaluates the operating conditions and fuel performance of HBU (greater than62 GWd/tU rod average) fuel. Additionally, it investigates fuel performance sensitivities and discusses the effect on fuel performance.This work used two codes. Virtual Environment for Reactor Applications (VERA) was used to calculate steady-state power histories, identify HBU operating conditions using 10 different realistic HBU core designs, and down-select rods to a representative subset of fuel rods for subsequent BISON evaluation. The BISON fuel performance code was used to investigate steady-state HBU operating conditions and assess uncertainties associated with FGR and its effect on fuel temperatures and RIPs. The VERA and BISON results will provide direct input for HBU experimental testing and support subsequent TRACE and BISON transient fuel performance analyses.

Modeling and simulation of VERA core physics benchmark using OpenMC code

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.

Advanced two-phase subchannel method via non-linear iteration

A fast-running, robust two-phase flow, sub-channel model is presented based on non-linear solution of the steady-state subchannel fluid flow equations. The drift-flux model solves for conservation of liquid and vapor mass, mixture energy, and axial and transverse mixture momentum as part of an efficient planar marching scheme and nonlinear, nested outer and inner iteration. Models based on mechanistic subcooled boiling, two-phase turbulent void mixing, and drift are included. Solution verification and mesh convergence studies were performed for modern GE 10 × 10 fuel geometry and are shown to have excellent convergence behavior. Run time performance for a 50 axial mesh model showed 2.2 seconds on a single CPU core to tightly converge all 3D distributions (flow, void, pressure) for the GE 10 × 10 fuel geometry, supporting its efficient use within the Virtual Environment for Reactor Applications boiling water reactor framework.

Validation of Light Water Reactor Ex-Core Calculations with VERA

The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) is a reactor simulation software. It offers unique capabilities by combining high-fidelity in-core radiation transport with temperature feedback by using MPACT (a deterministic neutron transport code) and COBRA-TF (a thermal-hydraulic code) with follow-on, fixed-source transport calculations using the Shift Monte Carlo code to calculate ex-core quantities of interest. In these coupled calculations, MPACT provides Shift with the fission source for follow-on ex-core calculations. These ex-core simulations can be set up to calculate detector responses, as well as the flux and fluence in ex-core regions of interest, such as the reactor pressure vessel, nozzle, and irradiated capsules. A Watts Bar Nuclear Plant Unit 1 (WBN1) ex-core model was developed, as described in this paper, and this model was used to perform coupon calculations. The results for the coupon flux calculations show close agreement with the reference values for cycle 1 produced by the two-dimensional Discrete Ordinates Transport (DORT) code and presented in a BWXT Services Inc. report. However, differences in the results (10%) seen in cycles 2 and 3 and the reasons for these differences are discussed in this paper. The VERA WBN1 model was also used to perform a vessel fluence calculation for cycle 1. Additionally, a collaboration between CASL and Duke Energy led to the first code-to-code validation of VERA for reactor ex-core applications that used a model for the Shearon Harris reactor. Results from this collaboration show excellent agreement between VERA and the Monte Carlo N-Particle Transport Code for the detector response calculations. The work performed under this collaboration is also detailed in this paper.

Implementation of two-phase gas transport into VERA for molten salt reactor analysis

Molten salt reactors (MSRs) are a class of next-generation nuclear reactors that have received recent industrial and research interest. A generalized species transport solver was implemented in the Virtual Environment for Reactor Applications (VERA) computing suite to extend this tool to analyze liquid-fueled MSRs. This core simulator has been extended to model the transport of fission product gases into a collection of circulating gas bubbles with the purpose of removing the gases. This paper presents the governing species transport equation, along with various nuclear source terms. Development of the source term for phase migration is discussed, along with a simplified interfacial area tracking method. Finally, a case study on a simplified MSR loop is presented in which modeling parameters were varied to assess their impact on gas removal. The steady state results show that parameters such as bubble diameter, gas injection rate and mass transfer coefficient have a low to moderate effect on the fraction of xenon in the core region. Removal efficiency has the greatest effect on the fraction in the core region. After the pump bowl, bubble diameter has a minor effect on the fraction of xenon in the gas void. These results point out that increasing parameters such as mass transfer coefficient, gas injection rate, and removal efficiency drives the xenon into the circulating gas void, while decreasing bubble diameter also drives xenon into the gas void by increasing interfacial area.

Optimizing the RMC Code Using the Decay Chain Method for Large-Scale Decay Calculations

Decay calculations play an important role in reactor physics simulations. In particular, the isotopic decay of the burned fuel during refueling is important for predicting the startup reactivity of the following burn-up cycle. In addition, there is a growing interest in high-fidelity simulations where the mesh in the burn-up region can involve millions of regions. However, existing models repeatedly solve the same Bateman equations for each region, which is a waste of calculational resources. RMC is a Monte Carlo neutron transport code developed for advanced reactor physics analysis including criticality calculations and burn-up calculations. This paper presents a decay calculation method named the Decay Chain Method (DCM) to optimize the RMC code for large-scale decay calculations. Unlike traditional methods, the Decay Chain Method solves the Bateman equations one decay chain at a time rather than one region at a time. The decay calculation in the burn-up mode then treats the decay steps as zero power burn-up steps with some optimized calculational methods to further reduce the calculational time. These methods were evaluated for a single pin example and for a Virtual Environment for Reactor Applications (VERA) full-core example. The calculations for the single pin example verify the accuracy of the decay step treatment in the burn-up mode and show the improved efficiency. The single pin is divided into 1–1,000,000 decay regions to study the efficiency differences between the Transmutation Trajectory Analysis (TTA) and DCM methods. Both methods have a linear complexity with respect to the number of regions but DCM costs just one-sixtieth of the TTA time. In the simplified VERA full core example, the DCM method reduces the decay calculation time to 0.32 min from 75.26 min while the accuracy remains unchanged.

Multiphysics Coupling Methods for Molten Salt Reactor Modeling and Simulation in VERA

New multiphysics coupling capabilities for molten salt reactor (MSR) analysis have been developed in the Virtual Environment for Reactor Applications (VERA). This development consisted of two main efforts. First, a generic species transport module was added in the CTF code, which is the thermal-hydraulics (TH) code for VERA. This module uses the velocity fields for which CTF solves during the TH calculation to transport species through the core and around the primary loop. Additionally, a gas sparging model has been added to CTF to model the movement of certain species, namely, fission products such as xenon gas, to transport between the molten salt and gas bubbles present in the salt. The second effort in this development was coupling this capability to VERA’s neutron transport code MPACT. This effort focused on coupling the detailed TH transport models in CTF to MPACT to account for feedback effects in the neutron transport calculations. Finally, the thermochemistry code Thermochimica has also been coupled to VERA. Thermochimica performs pointwise calculations for chemical potential and Gibbs free energy and determines what phases are produced by the temperature, pressure, and elemental concentrations at different locations in the primary loop. These capabilities are demonstrated using a model of the Molten Salt Reactor Experiment (MSRE). This reactor operated at Oak Ridge National Laboratory in the 1960s, providing sources of experimental data that were used to develop the model. Various combinations of species were modeled using VERA’s new multiphysics coupling capabilities. Species distributions and reactivity effects behaved as expected for the MSRE model, demonstrating that the coupling is behaving correctly and causing appropriate feedback. The results of these calculations show the potential for VERA to be used for a wide variety of MSR analyses.

Assessment of the Predictive Capability of VERA—CS for CASL Challenge Problems

Abstract This paper describes the process for assessing the predictive capability of the Consortium for the advanced simulation of light-water reactors (CASL) virtual environment for reactor applications code suite (VERA—CS) for different challenge problems. The assessment process is guided by the two qualitative frameworks, i.e., phenomena identification and ranking table (PIRT) and predictive capability maturity model (PCMM). The capability and credibility of VERA codes (individual and coupled simulation codes) are evaluated. Capability refers to evidence of required functionality for capturing phenomena of interest while credibility refers to the evidence that provides confidence in the calculated results. For this assessment, each challenge problem defines a set of phenomenological requirements (based on PIRT) against which the VERA software is evaluated. This approach, in turn, enables the focused assessment of only those capabilities that are relevant to the challenge problem. The credibility assessment using PCMM is based on different decision attributes that encompass verification, validation, and uncertainty quantification (VVUQ) of the CASL codes. For each attribute, a maturity score from zero to three is assigned to ascertain the acquired maturity level of the VERA codes with respect to the challenge problem. Credibility in the assessment is established by mapping relevant evidence obtained from VVUQ of codes to the corresponding PCMM attribute. The illustration of the proposed approach is presented using one of the CASL challenge problems called chalk river unidentified deposit (CRUD) induced power shift (CIPS). The assessment framework described in this paper can be considered applicable to other M & S code development efforts.

HIGH-FIDELITY EX-CORE CAPABILITIES IN VERA

High-fidelity nuclear reactor calculations have become increasingly important when considering lifetime extensions of the current nuclear fleet. The need for performing large calculations has led to development of advanced, novel methods for faster and more efficient computing system use. Under the Consortium for Advanced Simulation of Light Water Reactors (CASL), we have enabled the capability to perform high-fidelity ex-core calculations in the Virtual Environment for Reactor Applications (VERA) by coupling with the Shift Monte Carlo (MC) radiation transport package. The codes are coupled inmemory, allowing for the pin-by-pin fission source from the core simulator to be used by Shift. This unique capability allows the user to obtain both in-core and ex-core quantities by running a single simulation. Examples of ex-core calculations that can be performed include multicycle vessel fluence, detector response during reactor start-up and operation, and coupon fluence. One main advantage of the VERA ex-core capability is its flexibility and ease of use; users can select default settings with the standard VERA input for typical calculations or create their own ex-core geometry for specific cases. Also, Shift takes advantage of hybrid deterministic-MC methods to reduce variance and computational time. This paper details the full suite of VERA ex-core capabilities and provides input examples, simulation results, and computing resource use suggestions. These new capabilities have the potential to impact a wide user group in the nuclear community by enhancing and enabling high-fidelity light water reactor (LWR) ex-core calculations.

EFFECT OF FISSION SOURCE SPECTRUM ON MONTE CARLO CALCULATION OF EX-CORE QUANTITIES

The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) offers unique capabilities to combine highfidelity in-core radiation transport with temperature feedback using MPACT and CTF with a follow-on fixed source transport calculation using the Shift Monte Carlo code to calculate ex-core quantities of interest. In these coupled calculations, MPACT provides a fission source to Shift for the follow-on radiation transport calculation. In past VERA releases, MPACT passed a spatially dependent source without the energy distribution to Shift. Shift then assumed a235U Watt spectrum to sample the neutron source energies. There were concerns that, in cases with burned or mixed oxide (MOX) fuel near the periphery of the core, the assumption of a235U Watt spectrum for the source neutron energies would not be accurate for studying ex-core quantities of interest, such as pressure vessel fluence or detector response. Therefore, two additional options were implemented in VERA for Shift to sample neutron source energies: (1) a nuclide-dependent Watt spectra for235U,238U,239Pu, and241Pu, and (2) to use the standard 51-energy group MPACT spectrum. Results show that the 51-group MPACT spectrum is not suitable for ex-core calculations because the groups have been fine-tuned for in-core calculations. Differences in relative detector response due to235U and nuclide-dependent Watt spectra sampling schemes were negligible; however, the use of nuclide-dependent Watt spectra for vessel fluence calculations was found to be important for fuel cycles with burned and fresh fuel.

RESULTS OF VERA APPLICATION TO THE AP1000® PWR STARTUP

This paper describes the application of the Virtual Environment for Reactor Applications, VERA, under development by the Consortium for Advanced Simulation of LWRs (CASL) to the core physics analysis of the AP1000 PWR. The AP1000 PWR features an advanced first core with radial and axial heterogeneities which provides significant enhancements compared to traditional first cores, allowing best fuel usage and short transition to the equilibrium cycle with subsequent fuel reloads. These core advanced features can pose some challenges to the core physics tools making application of VERA to the AP1000 PWR first core especially relevant to qualify VERA performance. This paper focuses on the qualification efforts at hot zero power conditions, where Monte-Carlo reference solutions have also been established. In particular, the paper focuses on the comparison of the predictions obtained with VERA for the four AP1000 units that recently started up with the measured values. It is shown that there is excellent agreement between VERA and the key reactor physics parameters measured during the AP1000 startup.

EFFECTS OF FUEL TEMPERATURE-SHAPING FUNCTIONS ON XENON OSCILLATIONS

In coupled multiphysics simulations, single pin-averaged values are typically used to describe the temperature, power, and burnup within a given fuel pin. However, since xenon oscillations are largely driven by fuel temperatures, radially dependent quantities have been implemented in the Virtual Environment for Reactor Applications. These radial-shaping functions are based on Zernike polynomial expansions and allow information to pass effectively between codes with differing spatial meshes. This work examines the effects of radial fuel temperature-shaping functions on the behavior of axial xenon oscillations. A test problem was developed from full-core, multi-cycle depletions using as-built fuel data. The center 25 assemblies of the full-core case were used to test the radial-shaping function by inducing an axial xenon oscillation using an instantaneous control rod movement. The test case was run with and without the radial shapes, and each component was also run individually. Including the shaping functions significantly impacted the xenon oscillations for this problem; the magnitude and period of the oscillations were altered, and large pin power and soluble boron differences were observed. Testing each component individually showed that the radial fuel temperature-shaping function had the largest effect.

Simulation of a NuScale core design with the CASL VERA code

Three-dimensional (3D) full-core calculations are an integral part of fuel reload design for today’s light water reactors (LWRs). The current approaches are typically based on the nodal diffusion codes that calculate criticality state points and power distributions as a function of burnup. The CASL VERA (Consortium for Advanced Simulation of Light Water Reactors, Virtual Environment for Reactor Applications) code represents one of the latest advancements in 3D full-core calculation analysis based on transport theory methods employing the Method of Characteristics (MOC) and coupled multi-physics. In this article, a publicly released version of the NuScale reactor core is analysed with the VERA code (version 3.9 and 4.1) and contrasted against some static Serpent and Polaris based simulations. The analysis shows an excellent agreement for the lattice-level calculations as well as with some of the 3D full-core models. However, larger deviations were found in cases with heavy reflector models, whereby the reflector composition was found to impact the differences between the VERA and Serpent results. In this analysis, it was determined that greater than 90% stainless steel content in the heavy reflector leads to higher deviations between the VERA results and the Serpent results. The burnup calculations showed that the presence of the heavy reflector extends the cycle length and also leads to a flatter power distribution in the core, which can generally be interpreted as more efficient.

Impact of Radial Reflector Fidelity on Neutronics and Vessel Fluence Simulations

The Consortium for Advanced Simulation of Light Water Reactors is developing the Virtual Environment for Reactor Applications (VERA), and the MPACT code, which is the primary deterministic neutron transport solver in VERA, provides sub-pin level flux and power distributions as part of full-scale cycle depletion and analysis. In such calculations, an important aspect is the radial reflector treatment. To improve the fidelity of the radial reflector treatment, MPACT was extended to approximate the modeling of the reactor’s structural components such as the core shroud, barrel, neutron pads, and vessel. This work explores several modeling configurations with varying levels of fidelity and computational burden and assesses the importance of modeling fidelity on the eigenvalue and pin power distribution. Two two-dimensional (2-D) problems were analyzed to assess the impact on eigenvalue and pin power distributions with low-fidelity, coarse square cell reflector representations: (1) a Watts Bar Nuclear Plant Unit 1 (WBN1) quarter-core slice with depletion and (2) an AP1000 quarter-core slice. The analyses showed that the effect on eigenvalue is fairly small, but the effect on pin power is more pronounced, especially locally in the assemblies closest to the periphery, where the maximum pin power difference is nearly 3.5% in the AP1000 case. Two additional 2-D problems were used to assess the comparison between the low-fidelity coarse square cell treatment and a high-fidelity geometric representation that uses subpin material specification: (1) the same WBN1 quarter-core slice and (2) a representative model of the NuScale small modular reactor (SMR), which features a solid reflector design with moderator holes. These results demonstrate that even a coarse, low-fidelity representation adequately captures the necessary simulation characteristics. Last, these capabilities were applied to the 2-D WBN1 quarter-core depletion to assess the impact on vessel fluence using VeraShift. From adjoint calculations, pins along the periphery were observed to be of highest importance for fluence calculation, so the impact of the reflector representation in MPACT could theoretically substantially affect the predicted result. However, it was observed that the change in pin powers along the periphery minimally impacts the maximum vessel fluence with a difference within the statistical uncertainty but provides terrific insight on the sensitivity of the peripheral pins.

Assessment of CASL VERA for BWR analysis and application to accident tolerant SiC/SiC channel box

Application of the Virtual Environment for Reactor Applications (VERA) to BWR analysis is assessed in this paper by comparing results to those calculated using other widely-used modeling tools, namely the U.S. Nuclear Regulatory Commission’s PARCS/PATHS and the Serpent Monte Carlo particle transport code. Additionally, VERA is used to calculate 3-D temperature and fast neutron flux distributions in silicon carbide (SiC) fiber-reinforced, SiC matrix composite (SiC/SiC) BWR channel boxes, which are being studied as an Accident Tolerant Fuel core structural material concept. The code-to-code comparisons were favorable, and the SiC/SiC channel box evaluation demonstrates the many advanced modeling features VERA provides while also highlighting the non-uniformity in fast neutron flux distributions that can play a role in potential SiC/SiC channel box deformation. Traditional BWR analysis tools do not have the calculation fidelity necessary for coupled assessment of flux and temperature gradients in a SiC/SiC channel box. VERA is a state-of-the-art modeling environment that was developed to increase the safety and economic competitiveness of nuclear power through improved modeling accuracy. While VERA has already been deployed in the nuclear industry for PWR applications, the current study is a vital initial step in the extensive development, validation, and verification that VERA must go through to be useful for BWR applications.

Simulation of the BEAVRS benchmark using VERA

The Virtual Environment for Reactor Applications (VERA) is being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL). VERA is used to perform the Benchmark for Evaluation and Validation of Reactor Simulations (BEAVRS), which provides two cycles’ worth of operating power history, along with a full, detailed description of the geometry, and measured data. Cycle 1 and 2 are simulated with VERA and the results are compared to the measured zero power physics tests, critical boron concentration, and flux maps.

Coupled fuel performance calculations in VERA and demonstration on Watts Bar unit 1, cycle 1

As the core simulator capabilities in the Virtual Environment for Reactor Applications (VERA) have become more mature and stable, increased attention has been focused on coupling Bison to provide fuel performance simulations. This technique has been a very important driver for the pellet-clad interaction challenge problem being addressed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). In this article, two coupling approaches are demonstrated on quarter core problems based on Watts Bar Nuclear Plant Unit 1, Cycle 1: (1) an inline approach in which a one-way coupling is used between neutronics/thermal hydraulics (through MPACT/CTF) and fuel performance (through Bison) but no fuel temperature information is passed back to MPACT/CTF, and (2) a two-way approach (coupled) in which the fuel temperature is passed from Bison to MPACT/CTF. In both approaches, power and temperature distributions from MPACT/CTF are used to inform the Bison simulations for each rod in the core.The demonstrations presented here are the first integrated fuel performance simulations in VERA, which opens many possibilities for future work, including applications to accident-tolerant fuel efforts and transient simulations, which are of critical importance to CASL. These demonstrations also highlight the potential to move away from the current Bison-informed fuel temperature lookup table approach, which is the default in MPACT/CTF simulations, if performance improvements are made in the near future.

RMC/CTF multiphysics solutions to VERA core physics benchmark problem #8

The Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark Progression problems #1 to #10 based on the 17 × 17 Pressurized Water Reactor (PWR) Watts Bar Unit 1 designed by Westinghouse have been proposed by the Consortium for Advanced Simulations of Light Water Reactors (CASL) as core physics benchmarks. These problems were selected to assist nuclear software and methods developers to develop the capabilities needed to accurately model nuclear power reactors. These problems have 10 different conditions ranging from a simple 2D pin cell to the full core time-dependent depletion. VERA Problem #8, Physical Reactor Flux Maps, describes the first 48 h of a reactor startup with the power gradually increasing from 0% to 100% including the control rod movements which must be considered. This problem was solved in this study using the continuous energy Monte Carlo code RMC coupled with the subchannel thermal-hydraulics code CTF in a hybrid computational strategy. The effects of axial-layering on the solutions were studied with the model divided into 10 layers in the axial direction with the analyses considering both the cost and the accuracy with VERA Problem #8 simulated on the Tianhe II supercomputer. A large number of particles was simulated to get a reliable solution with the mean uncertainty in the power distribution reduced to about 2.6% for all the 526,640 burnup regions for each iteration between RMC and CTF. The eigenvalue trajectory of the critical boron concentration during the startup was provided with the predicted keff which differs from the common continuous decline. The RMC/CTF code also gives a 3D view of the relative power density distribution and the relative uncertainty distribution at three typical time points and the subchannel outlet temperatures whose peak locations are consistent with the radial power distribution. The RMC/CTF solution is the first published coupled Monte Carlo neutronics/subchannel thermal hydraulics solutions for Problem #8.

MCS/TH1D analysis of VERA whole-core multi-cycle depletion problems

This paper presents the verification and validation elements of the UNIST in-house Monte Carlo code, MCS, for the multi-cycle and multi-physics analyses of high-fidelity, large-scale commercial pressurized water reactors (PWRs). Analysis on the neutronic performance with thermal/hydraulic (T/H) feedback is the key to detecting the complex behavior of an operating nuclear power reactor. The MCS solutions with T/H feedback, TH1D, of the consortium for advanced simulation of light water reactors (CASL) virtual environment for reactor applications (VERA) core physics benchmark progression problems 6 and 7 showed excellent agreement in eigenvalues, temperature and power profiles with the MC21/COBRA-IE, MC21/CTF and VERA-CS solutions for the single assembly and whole core of Watts Bar Nuclear 1 (WBN1) Cycle 1 under beginning-of-cycle and hot-full-power condition. Furthermore, the core depletion analysis is one of the most compelling advances for reactor analysis. Therefore, this work is significantly focused on the nuclide depletion simulation coupled with TH1D of the first two cycles of WBN1 to address the VERA core physics benchmark problems 9 and 10. MCS is one of the few Monte Carlo codes that have the capability of depletion calculation for both WBN1 Cycles 1 and 2. The accuracy of MCS simulation of WBN1 Cycle 1 is within 40 ppm and 30 ppm in critical boron concentration (CBC) for all burnup points, compared with the measured data and VERA-CS solutions, respectively. To demonstrate the multi-cycle refueling capability of MCS, the WBN1 Cycle 2 is simulated and compared with the solutions of VERA-CS only, because of the lack of available measured data. MCS shows excellent agreement compared with VERA-CS within 30 ppm in CBC, and the average bias for the entire Cycle 2 is approximately 20 ppm. These results provide confidence in MCS’s capability in high-fidelity, multi-cycle calculations of the practical PWR core.