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At least 1,675 records · Page 93
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Assessment of PCMI Simulation Using the Multidimensional Multiphysics BISON Fuel Performance Code

Since 2008, the Idaho National Laboratory (INL) has been developing a next-generation nuclear fuel performance code called BISON. BISON is built using INL’s Multiphysics Object-Oriented Simulation Environment, or MOOSE. MOOSE is a massively parallel, finite element-based framework to solve systems of coupled non-linear partial differential equations using the Jacobian-FreeNewton Krylov (JFNK) method. MOOSE supports the use of complex two- and three-dimensional meshes and uses implicit time integration, which is important for the widely varied time scales in nuclear fuel simulation. MOOSE’s object-oriented architecture minimizes the programming required to add new physics models. BISON has been applied to various nuclear fuel problems to assess the accuracy of its 2D and 3D capabilities. The benchmark results used in this assessment range from simulation results from other fuel performance codes to measurements from well-known and documented reactor experiments. An example of a well-documented experiment used in this assessment is the Third Risø Fission Gas Project, referred to as “Bump Test GE7”, which was performed on rod ZX115. This experiment was chosen because it allows for an evaluation of several aspects of the code, including fully coupled thermo-mechanics, contact, and several nonlinear material models. Bump Test GE7 consists of a base-irradiation period of a full-length rod in the Quad-Cities-1 BWR for nearly 7 years to a burnup of 4.17% FIMA. The base irradiation test is followed by a “bump test” of a sub-section of the original rod. The bump test takes place in the test reactor DR3 at Risø in a water-cooled HP1 rig under BWR conditions where the power level is increased by about 50% over base irradiation levels in the span of several hours. During base irradiation, the axial power profile is flat. During the bump test, the axial power profile changes so that the bottom half of the rod is at approximately 50% higher power than at the base irradiation level, while the power at the top of the rod is at about 20% of the base irradiation power level. 2D BISON simulations of the Bump Test GE7 were run using both discrete and smeared pellet geometry. Comparisons between these calculations and experimental measurements are presented for clad diameter and elongation after the base irradiation and clad profile along the length of the test section after the bump test. Preliminary comparisons between calculations and measurements are favorable, supporting the use of BISON as an accurate multiphysics fuel simulation tool.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

A Complex-Geometry Validation Experiment for Advanced Neutron Transport Codes

The Idaho National Laboratory (INL) has initiated a focused effort to upgrade legacy computational reactor physics software tools and protocols used for support of core fuel management and experiment management in the Advanced Test Reactor (ATR) and its companion critical facility (ATRC) at the INL.. This will be accomplished through the introduction of modern high-fidelity computational software and protocols, with appropriate new Verification and Validation (V&V) protocols, over the next 12-18 months. Stochastic and deterministic transport theory based reactor physics codes and nuclear data packages that support this effort include MCNP5[1], SCALE/KENO6[2], HELIOS[3], SCALE/NEWT[2], and ATTILA[4]. Furthermore, a capability for sensitivity analysis and uncertainty quantification based on the TSUNAMI[5] system has also been implemented. Finally, we are also evaluating the Serpent[6] and MC21[7] codes, as additional verification tools in the near term as well as for possible applications to full three-dimensional Monte Carlo based fuel management modeling in the longer term. On the experimental side, several new benchmark-quality code validation measurements based on neutron activation spectrometry have been conducted using the ATRC. Results for the first four experiments, focused on neutron spectrum measurements within the Northwest Large In-Pile Tube (NW LIPT) and in the core fuel elements surrounding the NW LIPT and the diametrically opposite Southeast IPT have been reported [8,9]. A fifth, very recent, experiment focused on detailed measurements of the element-to-element core power distribution is summarized here and examples of the use of the measured data for validation of corresponding MCNP5, HELIOS, NEWT, and Serpent computational models using modern least-square adjustment methods are provided.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Modelling of LOCA Tests with the BISON Fuel Performance Code

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

BISON code↗
DOE OSTI2015

MODELLING OF FUEL BEHAVIOUR DURING LOSS-OF-COOLANT ACCIDENTS USING THE BISON CODE

This work presents recent developments to extend the BISON code to enable fuel performance analysis during LOCAs. This newly developed capability accounts for the main physical phenomena involved, as well as the interactions among them and with the global fuel rod thermo-mechanical analysis. Specifically, new multiphysics models are incorporated in the code to describe (1) transient fission gas behaviour, (2) rapid steam-cladding oxidation, (3) Zircaloy solid-solid phase transition, (4) hydrogen generation and transport through the cladding, and (5) Zircaloy high-temperature non-linear mechanical behaviour and failure. Basic model characteristics are described, and a demonstration BISON analysis of a LWR fuel rod undergoing a LOCA accident is presented. Also, as a first step of validation, the code with the new capability is applied to the simulation of experiments investigating cladding behaviour under LOCA conditions. The comparison of the results with the available experimental data of cladding failure due to burst is presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

U.S. Sodium Fast Reactor Codes and Methods: Current Capabilities and Path Forward

The United States has extensive experience with the design, construction, and operation of sodium cooled fast reactors (SFRs) over the last six decades. Despite the closure of various facilities, the U.S. continues to dedicate research and development (R&D) efforts to the design of innovative experimental, prototype, and commercial facilities. Accordingly, in support of the rich operating history and ongoing design efforts, the U.S. has been developing and maintaining a series of tools with capabilities that envelope all facets of SFR design and safety analyses. This paper provides an overview of the current U.S. SFR analysis toolset, including codes such as SAS4A/SASSYS-1, MC2-3, SE2-ANL, PERSENT, NUBOW-3D, and LIFE-METAL, as well as the higher-fidelity tools (e.g. PROTEUS) being integrated into the toolset. Current capabilities of the codes are described and key ongoing development efforts are highlighted for some codes.

safety codes↗
DOE OSTI2012

Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmark and with the results provided by PARCS code. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes

The Canadian SCWR has the potential to achieve the goals that the generation IV nuclear reactors must meet. As part of the optimization process for this design concept, lattice cell calculations are routinely performed using deterministic codes. In this study, the first step (self-shielding treatment) of the computation scheme developed with the deterministic code DRAGON for the Canadian SCWR has been validated. Some options available in the module responsible for the resonance self-shielding calculation in DRAGON 3.06 and different microscopic cross section libraries based on the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared to a reference calculation performed with the Monte Carlo code SERPENT under the same conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplication factor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use in this case of study. In addition, the 89 groups WIMS-AECL library for slight enriched uranium and the 172 groups WLUP library for a mixture of plutonium and thorium give the most consistent results with those of SERPENT. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Decay-ratio calculation in the frequency domain with the LAPUR code using 1D-kinetics

This paper deals with the problem of computing the Decay Ratio in the frequency domain codes as the LAPUR code. First, it is explained how to calculate the feedback reactivity in the frequency domain using slab-geometry i.e. 1D kinetics, also we show how to perform the coupling of the 1D kinetics with the thermal-hydraulic part of the LAPUR code in order to obtain the reactivity feedback coefficients for the different channels. In addition, we show how to obtain the reactivity variation in the complex domain by solving the eigenvalue equation in the frequency domain and we compare this result with the reactivity variation obtained in first order perturbation theory using the 1D neutron fluxes of the base case. Because LAPUR works in the linear regime, it is assumed that in general the perturbations are small. There is also a section devoted to the reactivity weighting factors used to couple the reactivity contribution from the different channels to the reactivity of the entire reactor core in point kinetics and 1D kinetics. Finally we analyze the effects of the different approaches on the DR value. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Status of verification and validation of AREVA's ARCADIA{sup R} code system for PWR applications

In March 2010 the submittal of Topical Reports for ARCADIA{sup R} and COBRA-FLX, the thermal-hydraulic module of ARCADIA{sup R}, to the U.S. Nuclear Regulatory Commission (NRC) concluded a major step in the development of AREVA's new code system for core design and safety analyses. This submittal was dedicated to the application of the code system to uranium fuel in pressurized water reactors. The submitted information comprised results for plants operated in the US (France)) and Germany and provided uncertainties for in-core measuring systems with traveling in-core detectors and for the aero-ball system of the EPR. A reduction of the uncertainties in the prediction of F{sub AH} and F{sub Q} of > 1 % (absolute) was derived compared to the current code systems. This paper extents the verification and validation base for uranium based fuel and demonstrates the basic capabilities of ARCADIA{sup R} of describing MOX. The achieved status of verification and validation is described in detail. All applications followed the same standard without any specific calibration. The paper gives also insight in the new capability of 3D full core steady-state and transient pin-by-pin/sub-channel-by-sub-channel calculations and the opportunities offered by this feature. The gain of margins with increasing detail of the representation is outlined. Currently, the strategies for worldwide implementation of ARCADIA{sup R} are developed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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

Steps towards verification and validation of the Fetch code for Level 2 analysis, design, and optimization of aqueous homogeneous reactors

Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' While AHRs have been modeled effectively using simplified 'Level 1' tools, the complex interactions between fluids, neutronics, and solid structures are important (but not necessarily safety significant). These interactions require a 'Level 2' modeling tool. Imperial College London (ICL) has developed such a tool: Finite Element Transient Criticality (FETCH). FETCH couples the radiation transport code EVENT with the computational fluid dynamics code (Fluidity), the result is a code capable of modeling sub-critical, critical, and super-critical solutions in both two-and three-dimensions. Using FETCH, ICL researchers and B and W engineers have studied many fissioning solution systems include the Tokaimura criticality accident, the Y12 accident, SILENE, TRACY, and SUPO. These modeling efforts will ultimately be incorporated into FETCH'S extensive automated verification and validation (V and V) test suite expanding FETCH'S area of applicability to include all relevant physics associated with AHRs. These efforts parallel B and W's engineering effort to design and optimize an AHR to produce Mo99. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Best-estimate plus uncertainty thermal-hydraulic stability analysis of BWRs using TRACG code

Over the last decade, Boiling Water Reactor (BWR) power up-rates have increased plant rated power output significantly. Subsequent projects have expanded flow domains (e.g. MELLLA+) for operation at these higher power levels. This has resulted in an increase in the power to flow ratio in regions susceptible to reactor thermal-hydraulic instabilities. Since BWRs are susceptible to coupled thermal-hydraulic/nuclear oscillations when operating at these conditions, such oscillations must be prevented or reliably detected and suppressed. The Detect and Suppress Solution - Confirmation Density (DSS-CD) is the most sophisticated GEH BWR instability protection system ever employed. DSS-CD implements algorithms that monitor closely-spaced groups of Local Power Range Monitor (LPRM) detectors to detect periodic behavior typical of reactor instability events. This system is able to detect small, localized power variations in the core, distinguish between true instabilities and plant noise, and trip/scram the reactor while maintaining adequate safety margins. The combination of hardware, software, and system setpoints provides protection against violation of the Safety Limit Minimum Critical Power Ratio (SLMCPR) for anticipated oscillations. To support DSS-CD implementation, the TRACG system code is used to simulate events to confirm the capability of the DSS-CD solution for early oscillation detection and suppression. TRACG is a GEH proprietary version of the Transient Reactor Analysis Code (TRAC). TRACG includes a multi-dimensional, two-fluid model for the reactor thermal-hydraulics and a three-dimensional reactor kinetics model. The models are qualified to simulate a large variety of tests and reactor configurations, including thermal-hydraulic stability events. These features allow for detailed, best-estimate simulation of a wide range of BWR phenomena. A set of integrated TRACG event simulations for reasonably limiting anticipated events can be used to calculate the effect on the Minimum Critical Power Ratio (MCPR) performance. The purpose of the DSS-CD TRACG analysis is to confirm the inherent MCPR margin afforded by the solution design. This paper presents the Best Estimate Plus Uncertainty (BEPU) DSS-CD TRACG methodology and its application to BWR Thermal-Hydraulic (T-H) stability analyses. The statistical Code Scaling, Applicability and Uncertainty (CSAU) methodology (defined in NUREG/CR-5249) is used to calculate the MCPR uncertainty. The TRACG simulation includes a full core individual bundle model in which each fuel bundle is modeled as an individual T-H channel. The complete CSAU analysis of full core individual bundle model is an innovative solution represents the state-of-the-art stability analysis of BWRs and is the first ever full statistical analysis for stability safety analyses. The adoption of BEPU methodologies for stability analyses advances the understanding of the associated physical phenomena and maintains the safety of reactor plant operation in expanded operation domain with up-rated power. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Self-shielding models of MICROX-2 code

The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. In the previous study, a new fine-group cross section library of the MICROX-2 was generated and tested against reference calculations and measurement data. In this study, existing physics models of the MICROX-2 are reviewed and updated to improve the physics calculation performance of the MICROX-2 code, including the resonance self-shielding model and spatial self-shielding factor. The updated self-shielding models have been verified through a series of benchmark calculations against the Monte Carlo code, using homogeneous and pin cell models selected for this study. The results have shown that the updates of the self-shielding factor calculation model are correct and improve the physics calculation accuracy even though the magnitude of error reduction is relatively small. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by approximately 0.1 % and 0.2% for the homogeneous and pin cell models, respectively, considered in this study. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

A mono-dimensional nuclear fuel performance analysis code, PUMA, development from a coupled approach

Multidimensional-multi-physical phenomena in nuclear fuels are treated as a set of mono-dimensional-coupled problems which encompass heat, displacement, fuel constituent redistribution, and fission gas release. Rather than uncoupling these coupled equations as in conventional fuel performance analysis codes, efforts are put into to obtain fully coupled solutions by relying on the recent advances of numerical analysis. Through this approach, a new SFR metal fuel performance analysis code, called PUMA (Performance of Uranium Metal fuel rod Analysis code) is under development. Although coupling between temperature and fuel constituent was made easily, the coupling between the mechanical equilibrium equation and a set of stiff kinetics equations for fission gas release is accomplished by introducing one-level Newton scheme through backward differentiation formula. Displacement equations from 1D finite element formulation of the mechanical equilibrium equation are solved simultaneously with stress equation, creep equation, swelling equation, and FGR equations. Calculations was made successfully such that the swelling and the hydrostatic pressure are interrelated each other. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Development of a new flux map processing code for moveable detector system in PWR

This paper presents an introduction to the development of the flux map processing code MAPLE developed by China Nuclear Power Technology Research Institute (CNPPJ), China Guangdong Nuclear Power Group (CGN). The method to get the three-dimensional 'measured' power distribution according to measurement signal has also been described. Three methods, namely, Weight Coefficient Method (WCM), Polynomial Expand Method (PEM) and Thin Plane Spline (TPS) method, have been applied to fit the deviation between measured and predicted results for two-dimensional radial plane. The measured flux map data of the LINGAO nuclear power plant (NPP) is processed using MAPLE as a test case to compare the effectiveness of the three methods, combined with a 3D neutronics code COCO. Assembly power distribution results show that MAPLE results are reasonable and satisfied. More verification and validation of the MAPLE code will be carried out in future. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Monte Carlo burnup code acceleration with the correlated sampling method. Preliminary test on an UOX cell with TRIPOLI-4{sup R}

For several years, Monte Carlo burnup/depletion codes have appeared, which couple a Monte Carlo code to simulate the neutron transport to a deterministic method that computes the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3 dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the time-expensive Monte Carlo solver called at each time step. Therefore, great improvements in term of calculation time could be expected if one could get rid of Monte Carlo transport sequences. For example, it may seem interesting to run an initial Monte Carlo simulation only once, for the first time/burnup step, and then to use the concentration perturbation capability of the Monte Carlo code to replace the other time/burnup steps (the different burnup steps are seen like perturbations of the concentrations of the initial burnup step). This paper presents some advantages and limitations of this technique and preliminary results in terms of speed up and figure of merit. Finally, we will detail different possible calculation scheme based on that method. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Validation Study in SAS4A Code in Simulated Mild TOP Condition

Model improvements and validation efforts for SAS4A code, which is the most advanced computer code for simulating the the primary phase of the Core Disruptive Accident (CDA) of MOX-fueled Sodium-cooled Fast Reactors, have been performed utilizing the experimental data from the CABRI programs by JAEA (Japan Atomic Energy Agency), IRSN (Institut de Radioprotection et de Surete Nucleaire), CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives) and KIT (Karlsruhe Institute of Technology). In order to confirm validity of improved models of SAS4A, a systematic and comprehensive validation plan was defined. It was based on the PIRT (Phenomena Identification and Ranking Table) result for CDA of MOX-fueled fast reactors. The fuel failure position and time are the most important parameters, which dominate the analytical results in a reactor calculation. The failure position will affect the reactivity feedback due to the fuel distribution after failure and consequently would have an effect on the energy release in the initiating phase. This paper describes the preliminary result of validation study for SAS4A code in simulated Transient Over-Power (TOP) condition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Validation of continuous energy sensitivity coefficients calculations in the MORET 5 code - 14269

A continuous-energy sensitivity coefficients calculation capability has been developed in the MORET 5 Monte Carlo code recently. As discussed in this paper, it relies on the differential operator method, an adjoint weighting model (to avoid calculating the fission source derivatives) and a constrained sensitivity coefficient model for sensitivity profiles to distributions such as, for instance, the fission spectrum. For now, the coefficients are provided per medium, per isotope, per reaction on an energy binning defined by the user. The sensitivity coefficients are currently available for the following quantities: total, elastic scattering, total inelastic scattering, fission, total capture and radiative capture cross sections, average neutron produced by fission and fission spectrum. This paper deals with the validation of this new feature in the MORET code on several benchmarks and applications with various energy spectra and geometrical configurations. This validation is performed through the UACSA expert group and the Jezebel ICSBEP benchmarks, but also thanks to calculations performed in the frame of experimental program design in the STACY facility. Comparisons are made using the MCNP6.1 code and the SCALE6.1 package as references. Preliminary results show a pretty good agreement between the different software. (authors)

42 ENGINEERING↗