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DOE OSTI2016

Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermal hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores due to hardened (or softened) spectrum. This study shows minimal impact of SiC-based cladding configurations on the transient response versus reference zirconium-based cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore the FeCrAl-based cases have a more rapid fuel thermal expansion rate and the resultant pellet-cladding interaction occurs more rapidly.

cladding mechanical performance↗

NASA NTRS2016

Design and Development of a 200-kW Turbo-Electric Distributed Propulsion Testbed

The National Aeronautics and Space Administration (NASA) Armstrong Flight Research Center (AFRC) (Edwards, California) is developing a Hybrid-Electric Integrated Systems Testbed (HEIST) Testbed as part of the HEIST Project, to study power management and transition complexities, modular architectures, and flight control laws for turbo-electric distributed propulsion technologies using representative hardware and piloted simulations. Capabilities are being developed to assess the flight readiness of hybrid electric and distributed electric vehicle architectures. Additionally, NASA will leverage experience gained and assets developed from HEIST to assist in flight-test proposal development, flight-test vehicle design, and evaluation of hybrid electric and distributed electric concept vehicles for flight safety. The HEIST test equipment will include three trailers supporting a distributed electric propulsion wing, a battery system and turbogenerator, dynamometers, and supporting power and communication infrastructure, all connected to the AFRC Core simulation. Plans call for 18 high performance electric motors that will be powered by batteries and the turbogenerator, and commanded by a piloted simulation. Flight control algorithms will be developed on the turbo-electric distributed propulsion system.

distributed↗

DOE OSTI2014

A high-performance workflow system for subsurface simulation

Subsurface modeling applications typically neglect uncertainty in the conceptual models, past or future scenarios, and attribute most or all uncertainty to errors in model parameters. In this contribution, uncertainty in technetium-99 transport in a heterogeneous, deep vadose zone is explored with respect to the conceptual model using a next generation user environment called Akuna. Akuna provides a range of tools to manage environmental modeling projects, from managing simulation data to visualizing results from high-performance computational simulators. Core toolsets accessible through the user interface include model setup, grid generation, parameter estimation, and uncertainty quantification. The BC Cribs site at Hanford in southeastern Washington State is used to demonstrate Akuna capabilities. At the BC Cribs site, conceptualization of the system is highly uncertain because only sparse information is available for the geologic conceptual model, the physical and chemical properties of the sediments, and the history of waste disposal operations. Using the Akuna toolset to perform an analysis of conservative solute transport, significant prediction uncertainty in simulated concentrations is demonstrated by conceptual model variation. This demonstrates that conceptual model uncertainty is an important consideration in sparse data environments such as BC Cribs. It is also demonstrated that Akuna and the underlying toolset provides an integrated modeling environment that streamlines model setup, parameter optimization, and uncertainty analyses for high-performance computing applications.

Akuna↗

NASA NTRS1998

Static Performance of a Wing-Mounted Thrust Reverser Concept

An experimental investigation was conducted in the Jet-Exit Test Facility at NASA Langley Research Center to study the static aerodynamic performance of a wing-mounted thrust reverser concept applicable to subsonic transport aircraft. This innovative engine powered thrust reverser system is designed to utilize wing-mounted flow deflectors to produce aircraft deceleration forces. Testing was conducted using a 7.9%-scale exhaust system model with a fan-to-core bypass ratio of approximately 9.0, a supercritical left-hand wing section attached via a pylon, and wing-mounted flow deflectors attached to the wing section. Geometric variations of key design parameters investigated for the wing-mounted thrust reverser concept included flow deflector angle and chord length, deflector edge fences, and the yaw mount angle of the deflector system (normal to the engine centerline or parallel to the wing trailing edge). All tests were conducted with no external flow and high pressure air was used to simulate core and fan engine exhaust flows. Test results indicate that the wing-mounted thrust reverser concept can achieve overall thrust reverser effectiveness levels competitive with (parallel mount), or better than (normal mount) a conventional cascade thrust reverser system. By removing the thrust reverser system from the nacelle, the wing-mounted concept offers the nacelle designer more options for improving nacelle aero dynamics and propulsion-airframe integration, simplifying nacelle structural designs, reducing nacelle weight, and improving engine maintenance access.

Asbury, Scott C.↗

DOE OSTI2013

INL Results for Phases I and III of the OECD/NEA MHTGR-350 Benchmark

The Idaho National Laboratory (INL) Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Methods Core Simulation group led the construction of the Organization for Economic Cooperation and Development (OECD) Modular High Temperature Reactor (MHTGR) 350 MW benchmark for comparing and evaluating prismatic VHTR analysis codes. The benchmark is sponsored by the OECD's Nuclear Energy Agency (NEA), and the project will yield a set of reference steady-state, transient, and lattice depletion problems that can be used by the Department of Energy (DOE), the Nuclear Regulatory Commission (NRC), and vendors to assess their code suits. The Methods group is responsible for defining the benchmark specifications, leading the data collection and comparison activities, and chairing the annual technical workshops. This report summarizes the latest INL results for Phase I (steady state) and Phase III (lattice depletion) of the benchmark. The INSTANT, Pronghorn and RattleSnake codes were used for the standalone core neutronics modeling of Exercise 1, and the results obtained from these codes are compared in Section 4. Exercise 2 of Phase I requires the standalone steady-state thermal fluids modeling of the MHTGR-350 design, and the results for the systems code RELAP5-3D are discussed in Section 5. The coupled neutronics and thermal fluids steady-state solution for Exercise 3 are reported in Section 6, utilizing the newly developed Parallel and Highly Innovative Simulation for INL Code System (PHISICS)/RELAP5-3D code suit. Finally, the lattice depletion models and results obtained for Phase III are compared in Section 7. The MHTGR-350 benchmark proved to be a challenging simulation set of problems to model accurately, and even with the simplifications introduced in the benchmark specification this activity is an important step in the code-to-code verification of modern prismatic VHTR codes. A final OECD/NEA comparison report will compare the Phase I and III results of all other international participants in 2014, while the remaining Phase II transient case results will be reported in 2015.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Thermal Hydraulic Experimental Test Article - Status Report for FY2019 (Rev. 1)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in an SFR. High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. The design work for the facility has been completed and is currently in the procurement, manufacturing and initial testing stage. The THETA primary system, including the redan, inlet and outlet plenum, 38 kW electric core simulant heater, submerged flowmeter, pump, and all instrumentation has either been procured or is currently being tested at Argonne. The secondary sodium system for cooling the tube-side sodium of the intermediate heat exchanger has all been analyzed according to ASME code and is in the final development stages before procurement.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

SuPer-Homogenization (SPH) Corrected Cross Section Generation for High Temperature Reactor

The deterministic full core simulators require homogenized group constants covering the operating and transient conditions over the entire lifetime. Traditionally, the homogenized group constants are generated using lattice physics code over an assembly or block in the case of prismatic high temperature reactors (HTR). For the case of strong absorbers that causes strong local depressions on the flux profile require special techniques during homogenization over a large volume. Fuel blocks with burnable poisons or control rod blocks are example of such cases. Over past several decades, there have been a tremendous number of studies performed for improving the accuracy of full-core calculations through the homogenization procedure. However, those studies were mostly performed for light water reactor (LWR) analyses, thus, may not be directly applicable to advanced thermal reactors such as HTRs. This report presents the application of SuPer-Homogenization correction method to a hypothetical HTR core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

SCM overview & the EBR-II shutdown heat removal tests validation.

Pronghorn is an engineering-scale, coarse-mesh, thermal-hydraulics tool for supporting reactor-core simulations of advanced nuclear reactors. Most of the current efforts in Pronghorn have been devoted in developing porous finite-volume capabilities and adapting closure correlations for coarse-mesh thermal-hydraulics modeling. However, for liquid-metal reactors (LMRs) with wire-wrapped fuel pin assemblies, a pin-level thermal-hydraulic resolution is required for most safety case studies (pin rupture, channel blockage, etc.). For this purpose, a new Subchannel application is developed in MOOSE, which affords the required flow field resolution, while still preserving an engineering-scale approach. This new solver can be natively coupled to Pronghorn and other MOOSE objects to enable full-core, multi-physics, multi-scale engineering studies. This presentation presents the main features of the SCM code and demonstrates a validation case based on the EBR-II SHRT tests.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

DOE OSTI2021

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. Finally, 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.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Coolant density and control blade history effects in extended BWR burnup credit - 14494

Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k{sub eff}) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in various locations and at varying degrees during BWR operation based on the core loading pattern. When present during depletion, control blades harden the neutron spectrum locally because they displace the moderator and absorb thermal neutrons. The investigation of the effect of control blades on post operational cask reactivity is documented herein, as is the effect of multiple (continuous and intermittent) exposure periods with control blades inserted. The coupled effects of control blade presence on power density, void profile, or burnup profile will be addressed in future work. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Analysis of Reverse Flow Restriction Device to Prevent Dryout Fuel Damage during BWR Instability

Boiling Water Reactors (BWRs) are subjected to instabilities due to thermal hydraulics feedback or coupled Neutronics-thermal hydraulics feedback. Feedback is affected by many reactor parameters that could destabilize the core such as distribution of pressure drop across the core, axial power shape, subcooling level at the core inlet, radial power peaking factor, reactivity coefficients, and others. There are three common modes of instability in BWRs: single channel instability, core-wide instability, and regional instability. Each instability mode could damage the fuel if the power oscillations are left without an intervention like reactor scram. However, the instability problem becomes more dangerous if the automatic reactor scram is lost, like in Anticipated Transient without Scram (ATWS) accidents. A reverse flow restriction device (RFRD) would mitigate this problem. Core simulation using TRACE/PARCS code has been used to examine the potential effect of restricting the flow in the downward direction on mitigation of the power and flow oscillations to increase the safety of BWRs. The presented results demonstrate the ability of the device on limiting flow oscillation amplitude in the one bundle where it is used, on an adjacent bundle without it and reducing the power peak when the flow is low and power is high during BWR power oscillations. This exercise demonstrates the potential of previously described hardware device on preventing clad temperature excursion to keep the fuel safe from dryout damage and provide encouraging results ahead of the planned more detailed investigation where the device is to be used in more bundles up to full core. In the future work it is planned to examine the effect of device installed in a fraction of the bundles on the flow rate in other bundles. Also, the effect on the characteristics of dryout and rewetting and the resulting temperature excursions with and without the device will be given special attention. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k eff ) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in various locations and at varying degrees during BWR operation based on the core loading pattern. When present during depletion, control blades harden the neutron spectrum locally because they displace the moderator and absorb thermal neutrons. The investigation of the effect of control blades on post operational cask reactivity is documented herein, as is the effect of multiple (continuous and intermittent) exposure periods with control blades inserted. The coupled effects of control blade presence on power density, void profile, or burnup profile will be addressed in future work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Thermal Hydraulic Experimental Test Article: Status Report for FY2019

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in an SFR. High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. Figure 1 illustrates the main components of the primary system. THETA will be installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, the secondary system in the 18-inch Test Vessel #6, and its sodium-to-air heat exchanger will occupy the location for 18-inch Test Vessel #5 (Test Vessel #5 will not be installed), Figure 2. The design work for the facility has been completed and is currently in the procurement, manufacturing and initial testing stage. The THETA primary system, including the redan, inlet and outlet plenum, 38 kW electric core simulant heater, submerged flowmeter, pump, and all instrumentation has either been procured or is currently being tested at Argonne. The secondary sodium system for cooling the tube-side sodium of the intermediate heat exchanger has all been analyzed according to ASME code and is in the final development stages before procurement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Framework Application for Core Edge Transport Simulations (FACETS) Performance Analysis (Final Report)

The goal of the FACETS project (Framework Application for Core-Edge Transport Simulations) was to provide a multiphysics, parallel framework application (FACETS) that will enable whole-device modeling for the U.S. fusion program, to provide the modeling infrastructure needed for ITER, the next step fusion confinement device. Through use of modern computational methods, including component technology and object oriented design, FACETS is able to switch from one model to another for a given aspect of the physics in a flexible manner. This enables use of simplified models for rapid turnaround or high-fidelity models that can take advantage of the largest supercomputer hardware. FACETS does so in a heterogeneous parallel context, where different parts of the application execute in parallel by utilizing task farming, domain decomposition, and/or pipelining as needed and applicable. ParaTools, Inc. was tasked with supporting the performance analysis and tuning of the FACETS components and framework in order to achieve the parallel scaling goals of the project. The TAU Performance System® was used for instrumentation, measurement, archiving, and profile / tracing analysis. ParaTools, Inc. also assisted in FACETS performance engineering efforts. Through the use of the TAU Performance System, ParaTools provided instrumentation, measurement, analysis and archival support for the FACETS project. Performance optimization of key components has yielded significant performance speedups. TAU was integrated into the FACETS build for both the full coupled application and the UEDGE component. The performance database provided archival storage of the performance regression testing data generated by the project, and helped to track improvements in the software development.

97 MATHEMATICS AND COMPUTING↗

NASA NTRS1994

Electron acceleration at nearly perpendicular collisionless shocks. 3: Downstream distributions

Spacecraft observations at the Earth's bow shock and at interplanetary shocks have established that the largest fluxes of accelerated suprathermal electrons occur in so-called shock spike events immediately downstream of the shock ramp. Previous theoretical efforts have mainly focused on explaining upstream energetic electron beams. Here we investigate the general motion and acceleration of energetic electrons in a curved, nearly perpendicular shock by numerically integrating the orbits of solar wind halo electrons in shock fields generated by a hybrid simulation (core electron fluid and kinetic ions). Close to the angle Theta(sub Bn) = 90 degs between the upstream magnetic field and shock normal, the calculations result in a (perpendicular) temperature increase proportional to the magnetic field ratio and give the highest phase space densities in the overshoot. For a steep distribution, the temperature change can correspond to an enhancement of the distribution by several orders of magnitude. These results are in agreement with predictions from adiabatic mapping. With smaller angles Theta(sub Bn), the overshoot and downstream densities fall off quickly, because the adiabatic energy gain is less and fewer electrons transmit. The shock curvature also leads to an accumulation of electrons close to 90 degs. Without pitch angle scattering, energization is only significant within a few (approximately 5 to 10 degs) degrees of the point of tangency. However, shock spike events appear to be observed more easily and farther away from 90 degs. Given that over a region of several degrees around 90 degs the theory gives enhancements of up to approximately 4 orders of magnitude, such electrons could in principle account for the typically observed enhancements of 1 to 2 orders of magnitude, if they were distributed over Theta(sub Bn). To test the idea that scattering could efficiently redistribute the energetic electrons, we have conducted test particle simulations in which artificial pitch angle scattering is included. We find that such a process can indeed be very efficient and can explain observations of shock spike events far away from Theta(sub Bn) is approximately 90 degs. The scattering naturally leads to much higher phase space densities at smaller Theta(sub Bn) than what a local one-dimensional mapping would predict and thus can account for an observed discrepancy with adiabatic theory stated in the literature.

Krauss-Varban, D.↗

DOE OSTI2019

On the Neutrino Distributions in Phase Space for the Rotating Core-collapse Supernova Simulated with a Boltzmann-neutrino-radiation-hydrodynamics Code

With the Boltzmann-radiation-hydrodynamics code, which we have developed to solve numerically the Boltzmann equations for neutrino transfer, the Newtonian hydrodynamics equations, and the Newtonian self-gravity simultaneously and consistently, in this work we simulate the collapse of a rotating core of the progenitor with a zero-age- main-sequence mass of 11.2 M ⊙ and a shellular rotation of $1\,\mathrm{rad}\,{{\rm{s}}}^{-1}$ at the center. We pay particular attention in this paper to the neutrino distribution in phase space, which is affected by the rotation. By solving the Boltzmann equations directly, we can assess the rotation-induced distortion of the angular distribution in momentum space, which gives rise to the rotational component of the neutrino flux. We compare the Eddington tensors calculated both from the raw data and from the M1-closure approximation. We demonstrate that the Eddington tensor is determined by complicated interplays of the fluid velocity and the neutrino interactions and that the M1-closure, which assumes that the Eddington factor is determined by the flux factor, fails to fully capture this aspect, especially in the vicinity of the shock. We find that the error in the Eddington factor reaches ~20% in our simulation. This is due not to the resolution but to the different dependence of the Eddington and flux factors on the angular profile of the neutrino distribution function, and hence modification to the closure relation is needed.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2015

Systematics of the α-decay fine structure in even–even nuclei

We describe α-transitions to low-lying states in even–even nuclei with Z>50,N>82 using the Coherent State Model (CSM). We reproduce the energy levels and electromagnetic transition rates between the states of the ground band employing two parameters, namely the deformation parameter and the strength of the harmonic CSM Hamiltonian. The B(E2) values are described in terms of an effective charge which depends linearly on the deformation parameter. The treatment of the α-emission process is based on an α–daughter interaction containing a monopole component, calculated through a double folding procedure with a M3Y interaction plus a repulsive core simulating the Pauli principle, and a quadrupole–quadrupole (QQ) interaction. The decaying states are identified with the lowest narrow outgoing resonances obtained through the coupled channels method. The α-branching ratios to 2{sup +} states are reproduced by means of the QQ strength. This interaction strength can be fitted with a linear dependence on the deformation parameter, as predicted by the CSM. The theoretical intensities to 4{sup +} and 6{sup +} states are in reasonable agreement with available experimental data. Predictions are made for spherical, transitional and well deformed even–even α-emitters.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

Sensitivity analysis of VERA-CS and FRAPCON coupling in a multiphysics environment

A demonstration and description of the LOCA Toolkit for US light water reactors (LOTUS) is presented. Through LOTUS, the core simulator VERA-CS developed by CASL is coupled with the fuel performance code FRAPCON. The coupling is performed with consistent uncertainty propagation with all model inconsistencies being well-documented. Monte Carlo sampling is performed on a single 17 × 17 fuel assembly with a three cycle depletion case. Both uncertainty quantification (UQ) and sensitivity analysis (SA) are used at multiple states within the simulation to elucidate the behavior of minimum departure from nucleate boiling ratio (MDNBR), maximum fuel centerline temperature (MFCT), and gap conductance at peak power (GCPP). The SA metrics used are the Pearson correlation coefficient, Sobol sensitivity indices, and the density-based, delta moment independent measures. Results for MDNBR show consistency among all SA measures, as well for all states throughout the fuel lifecycle. MFCT results contain consistent rankings between SA measures, but show differences throughout the lifecycle. GCPP exhibits predominantly linear relations at low and high burnup, but highly nonlinear relations at intermediate burnup due to abrupt shifts between models. As a result, such behavior is largely undetectable to traditional regression or variance-based methods and demonstrates the utility of density-based methods.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗