Engineering topics

Downar, T.

Publications and source records attributed to Downar, T..

17 records

DOE OSTI2022

Verification of the AGREE and Serpent for the steady state HTR-10 benchmark problems

This paper provides results from steady state HTR-10 benchmark calculations performed using AGREE and Serpent that are compared to experimental results as well as calculations performed by INET. The purpose of completing this benchmark is to validate AGREE and Serpent for the prediction of HTGR operation so that they may ultimately be used to support licensing and deployment efforts of advanced reactors. The benchmark consists of several problems ranging from control rod worth calculations to k{sub eff} calculations at various temperatures. Overall, AGREE and Serpent show good agreement with the reference solutions and are effectively able to predict HTGR operation for a variety of steady state cases. The largest difference from the reference was for the initial core single rod worth, which is possibly due to the larger core helium cavity causing inaccuracies in the diffusion calculation whereas the control rod worth determined by Serpent and AGREE is much closer to the reference result for the full core loading. Moreover, in every benchmark problem, increasing the number of energy groups in the cross sections results in improved agreement of the AGREE result with the reference, although it also causes an increase in computation time. For problems with experimental data available, the accuracy of results generated by Serpent and AGREE is comparable to results obtained by other benchmark participants. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Neutronics and thermal-hydraulics simulation of generic pebble-bed fluoride-salt-cooled high-temperature reactor (gFHR)

The fluoride-salt-cooled high-temperature reactor (FHR) is one type of the advanced reactors and has been attracting great interest from the research institutes and commercial companies in the recent years. However, currently the technology is relatively immature. To facilitate the design and safety analysis of FHRs, Kairos Power published the generic FHR (gFHR) benchmark. In this paper, a hybrid method combining the stochastic code and deterministic code is developed to simulate the gFHR benchmark. Serpent 2 is employed as a few-group cross section generator and the cross sections are applied to the finite difference neutron diffusion code AGREE. A good consistency with the gFHR benchmark is achieved. The agreement between the Serpent 2 results and the AGREE results shows that the hybrid method is applicable to FHRs. The thermal-hydraulics are coupled to neutronics for the steady-state calculations and good agreement between AGREE and SAM is achieved. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

Assessment of the 2D MOC solver in MPACT: Michigan parallel characteristics transport code

MPACT (Michigan Parallel Characteristics Transport Code) is a new reactor analysis tool being developed by researchers at the University of Michigan as an advanced pin-resolved transport capability within VERA (Virtual Environment for Reactor Analysis). VERA is the end-user reactor simulation tool being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL). The MPACT development project is itself unique for the way it is changing how students perform research to achieve the instructional and research goals of an academic institution, while providing immediate value to the industry. One of the major computational pieces in MPACT is the 2D MOC solver. It is critical that the 2D MOC solver provide an efficient, accurate, and robust solution over a broad range of reactor operating conditions. The C5G7 benchmark is first used to test the accuracy of the method with a fixed set of cross-sections. The VERA Core Physics Progression Problems are then used to compare the accuracy of both the 2D transport solver and also the cross-section treatments. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

Assessment and improvement of the 2D/1D method stability in DeCART

As part of ongoing work with Consortium for Advanced Simulation of Light Water Reactors (CASL), the 2D/1D code, DeCART, has demonstrated some of the advantages of the 2D/1D method with respect to realistic, full-core analysis, particularly over explicit 3D transport methods, which generally have higher memory and computation requirements. The 2D/1D method performs 2D-radial transport sweeps coupled with ID-axial diffusion calculations to provide a full 3D simulation. DeCART employs the 2D method of characteristics for the radial sweeps and ID one-node nodal diffusion for the axial sweeps, coupling the two methods with transverse leakages to ensure a more consistent representation of the transport equation. It has been observed that refinement of the axial plane thickness leads to instabilities in the calculation scheme. This work assesses the sources of these instabilities and the approaches to improve them, especially with respect to negative scattering cross sections and the tightness of the 2D-radial/ID-axial coupling schemes. Fourier analyses show that the existing iteration scheme is not unconditionally stable, suggesting a tighter coupling scheme is required. For this reason 3D-CMFD has been implemented, among other developments, to ensure more stable calculation. A matrix of test cases has been used to assess the convergence, with the primary parameter being the axial plane thickness, which has been refined down to 1 cm. These cases demonstrate the issues observed and how the modification improve the stability. However, it is apparent that more work is necessary to ensure unconditional stability. (authors)

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2013

Application of adaptive hierarchical sparse grid collocation to the uncertainty quantification of nuclear reactor simulators

Recent efforts in the application of uncertainty quantification to nuclear systems have utilized methods based on generalized perturbation theory and stochastic sampling. While these methods have proven to be effective they both have major drawbacks that may impede further progress. A relatively new approach based on spectral elements for uncertainty quantification is applied in this paper to several problems in reactor simulation. Spectral methods based on collocation attempt to couple the approximation free nature of stochastic sampling methods with the determinism of generalized perturbation theory. The specific spectral method used in this paper employs both the Smolyak algorithm and adaptivity by using Newton-Cotes collocation points along with linear hat basis functions. Using this approach, a surrogate model for the outputs of a computer code is constructed hierarchically by adaptively refining the collocation grid until the interpolant is converged to a user-defined threshold. The method inherently fits into the framework of parallel computing and allows for the extraction of meaningful statistics and data that are not within reach of stochastic sampling and generalized perturbation theory. This paper aims to demonstrate the advantages of spectral methods-especially when compared to current methods used in reactor physics for uncertainty quantification-and to illustrate their full potential. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Critical node treatment in the analytic function expansion method for Pin Power Reconstruction

Pin Power Reconstruction (PPR) was implemented in PARCS using the eight term analytic function expansion method (AFEN). This method has been demonstrated to be both accurate and efficient. However, similar to all the methods involving analytic functions, such as the analytic node method (ANM) and AFEN for nodal solution, the use of AFEN for PPR also has potential numerical issue with critical nodes. The conventional analytic functions are trigonometric or hyperbolic sine or cosine functions with an angular frequency proportional to buckling. For a critic al node the buckling is zero and the sine functions becomes zero, and the cosine function become unity. In this case, the eight terms of the analytic functions are no longer distinguishable from ea ch other which makes their corresponding coefficients can no longer be determined uniquely. The mode flux distribution of critical node can be linear while the conventional analytic functions can only express a uniform distribution. If there is critical or near critical node in a plane, the reconstructed pin power distribution is often be shown negative or very large values using the conventional method. In this paper, we propose a new method to avoid the numerical problem wit h critical nodes which uses modified trigonometric or hyperbolic sine functions which are the ratio of trigonometric or hyperbolic sine and its angular frequency. If there are no critical or near critical nodes present, the new pin power reconstruction method with modified analytic functions are equivalent to the conventional analytic functions. The new method is demonstrated using the L336C5 benchmark problem. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

The method of modular characteristic direction probabilities in MPACT

The method of characteristic direction probabilities (CDP) is based on a modular ray tracing technique which combines the benefits of the collision probability method (CPM) and the method of characteristics (MOC). This past year CDP was implemented in the transport code MPACT for 2-D and 3-D transport calculations. By only coupling the fine mesh regions passed by the characteristic rays in the particular direction, the scale of the probabilities matrix is much smaller compared to the CPM. At the same time, the CDP has the same capacity of dealing with the complicated geometries with the MOC, because the same modular ray tracing techniques are used. Results from the C5G7 benchmark problems are given for different cases to show the accuracy and efficiency of the CDP compared to MOC. For the cases examined, the CDP and MOC methods were seen to differ in k{sub eff} by about 1-20 pcm, and the computational efficiency of the CDP appears to be better than the MOC for some problems. However, in other problems, particularly when the CDP matrices have to be recomputed from changing cross sections, the CDP does not perform as well. This indicates an area of future work. (authors)

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2012

Post-refinement multiscale method for pin power reconstruction

The ability to accurately predict local pin powers in nuclear reactors is necessary to understand the mechanisms that cause fuel pin failure during steady state and transient operation. In the research presented here, methods are developed to improve the local solution using high order methods with boundary conditions from a low order global solution. Several different core configurations were tested to determine the improvement in the local pin powers compared to the standard techniques based on diffusion theory and pin power reconstruction (PPR). The post-refinement multiscale methods use the global solution to determine boundary conditions for the local solution. The local solution is solved using either a fixed boundary source or an albedo boundary condition; this solution is 'post-refinement' and thus has no impact on the global solution. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Equilibrium cycle pin by pin transport depletion calculations with DeCART

As the Advanced Fuel Cycle Initiative (AFCI) program has matured it has become more important to utilize more advanced simulation methods. The work reported here was performed as part of the AFCI fellowship program to develop and demonstrate the capability of performing high fidelity equilibrium cycle calculations. As part of the work here, a new multi-cycle analysis capability was implemented in the DeCART code which included modifying the depletion modules to perform nuclide decay calculations, implementing an assembly shuffling pattern description, and modifying iteration schemes. During the work, stability issues were uncovered with respect to converging simultaneously the neutron flux, isotopics, and fluid density and temperature distributions in 3-D. Relaxation factors were implemented which considerably improved the stability of the convergence. To demonstrate the capability two core designs were utilized, a reference UOX core and a CORAIL core. Full core equilibrium cycle calculations were performed on both cores and the discharge isotopics were compared. From this comparison it was noted that the improved modeling capability was not drastically different in its prediction of the discharge isotopics when compared to 2-D single assembly or 2-D core models. For fissile isotopes such as U-235, Pu-239, and Pu-241 the relative differences were 1.91%, 1.88%, and 0.59%), respectively. While this difference may not seem large it translates to mass differences on the order of tens of grams per assembly, which may be significant for the purposes of accounting of special nuclear material. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Coupled full core neutron transport/CFD simulations of pressurized water reactors

Recently as part of the CASL project, a capability to perform 3D whole-core coupled neutron transport and computational fluid dynamics (CFD) calculations was demonstrated. This work uses the 2D/1D transport code DeCART and the commercial CFD code STAR-CCM+. It builds on previous CASL work demonstrating coupling for smaller spatial domains. The coupling methodology is described along with the problem simulated and results are presented for fresh hot full power conditions. An additional comparison is made to an equivalent model that uses lower order T/H feedback to assess the importance and cost of high fidelity feedback to the neutronics problem. A simulation of a quarter core Combustion Engineering (CE) PWR core was performed with the coupled codes using a Fixed Point Gauss-Seidel iteration technique. The total approximate calculation requirements are nearly 10,000 CPU hours and 1 TB of memory. The problem took 6 coupled iterations to converge. The CFD coupled model and low order T/H feedback model compared well for global solution parameters, with a difference in the critical boron concentration and average outlet temperature of 14 ppm B and 0.94 deg. C, respectively. Differences in the power distribution were more significant with maximum relative differences in the core-wide pin peaking factor (Fq) of 5.37% and average relative differences in flat flux region power of 11.54%. Future work will focus on analyzing problems more relevant to CASL using models with less approximations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Modeling prismatic HTGRs with U.S. N.R.C advanced gas reactor evaluator (AGREE)

A core fluids and heat transfer model has been developed for the prismatic high temperature gas reactor in support of the US NRC Next Generation Nuclear Plant (NGNP) evaluation model. The core fluids modeling relies on a subchannel approach in which the primary coolant flow path through the core region and vertical in-core and ex-core gaps can be modeled as individual subchannels. These subchannels are connected together to represent a three dimensional reactor. An initial validation calculation for the core fluids model has been performed using data available in literature for bypass flow. The predicted bypass flow was within 2.6% of the value reported in the literature. The core level heat transfer model is based on a triangular finite volume method, where the base triangle is one sixth of the prismatic block. In order to improve the spatial accuracy at this level, a triangular refinement method was also implemented. The fuel compact temperature is calculated by a cylindrical conduction model which is implicitly coupled to the triangular core level model. The preliminary verification of the model was performed by comparing AGREE to a finite element code COMSOL by analyzing the MHTGR core heat transfer. Further verification and validation is currently an ongoing effort. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

TRACE/PARCS analysis of the OECD/NEA Oskarshamn-2 BWR stability benchmark

On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations with a decay ratio of about 1.4. The event was successfully modeled by the TRACE/PARCS coupled code system, and further analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations (decay ratio) and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validation for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal-hydraulics (TH), and TH/NK. coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one. Interested participants are invited to contact authors for more information. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Convergence analysis of a CMFD method based on generalized equivalence theory

CMFD acceleration methods have been successful in reducing the computational burden for steady-state and transient reactor calculations. However, recent work on a complex coupled code BWR ATWS event has exposed possible issues with the stability of the CMFD method when standard CMFD methods are used. During the simulation of the ATWS boron injection event in the BWR, the PARCS code failed to converge with the existing CMFD method. A new CMFD method based on generalized equivalence theory was developed and the PARCS solution converged for the same ATWS event. This paper presents the new method and a detailed analytic and numerical convergence analysis. The results show that the new CMFD converges for all possible cross sections combinations anticipated in Light Water Reactor simulation and unlike existing CMFD methods, it is very robust even when the initial guess is far from final true solution. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Comparison of XSUSA and 'two-step' approaches for full-core uncertainty quantification

While there are multiple sources of error that are introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily for quantifying the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the 'Two-Step' method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the Two-Step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor and in the core power distribution will be examined in the framework of phase 1-3 of the UAM Benchmark. Using the TMI core as a base model for analysis, the XSUSA and Two-Step methods are applied with certain limitations and the results are compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions are made for which method is currently a more viable option for computing uncertainties in burnup and transient calculations. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)

The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Convergence analysis of a CMFD method based on generalized equivalence theory

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Validation of the U.S. NRC NGNP evaluation model with the HTTR

The High Temperature Test Reactor (HTTR) was modeled with TRITON/PARCS. Traditional light water reactor (LWR) homogenization methods rely on the short mean free paths of neutrons in LWR. In gas-cooled, graphite-moderated reactors like the HTTR neutrons have much longer mean free paths and penetrate further into neighboring assemblies than in LWRs. Because of this, conventional lattice calculations with a single assembly may not be valid. In addition to difficulties caused by the longer mean free paths, the HTTR presents unique axial and radial heterogeneities that require additional modifications to the single assembly homogenization method. To handle these challenges, the homogenization domain is decreased while the computational domain is increased. Instead of homogenizing a single hexagonal fuel assembly, the assembly is split into six triangles on the radial plane and five blocks axially in order to account for the placement of burnable poisons. Furthermore, the radial domain is increased beyond a single fuel assembly to account for spectrum effects from neighboring fuel, reflector, and control rod assemblies. A series of five two-dimensional cases, each closer to the full core, were calculated to evaluate the effectiveness of the homogenization method and cross-sections. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗