Judith M. Cuta

Vipre-02―a two-fluid thermal-hydraulics code for reactor core and vessel analysis : mathematical modeling and solution methods

This paper reports on the VIPRE-02 code which is a thermal-hydraulic analysis code designed to model steady-state conditions and operational transients in light water reactor cores and vessels. It uses a two-fluid representation of two-phase flow that solves conservation equations for mass, momentum, and energy for each phase. The code uses a subchannel formulation of the conservation equations but also contains an optional three-dimensional (r-[theta] coordinates) representation of the lower plenum for vessel modeling. The six-equation formulation is solved implicitly, by a modified Gauss-Seidel iteration procedure, and has no time step size limitation for stability. Models for phase interaction based on flow regime mapping are provided that use empirical models and correlations for heat and mass transfer at the interface and vapor generation. In addition, the code contains as an option a dynamic flow regime model, which uses an interfacial area transport equation to determine the phase interaction terms.

A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling, core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.

Dynamics of Crust Dissolution and Gas Release in Tank 241-SY-101

Due primarily to an increase in floating crust thickness, the waste level in Tank 241-SY-101 has grown appreciably and the flammable gas volume stored in the crust has become a potential hazard. To remediate gas retention in the crust and the potential for buoyant displacement gas releases from the nonconvective layer at the bottom of the tank, SY-101 will be diluted to dissolve a large fraction of the solids that allow the waste to retain gas. The plan is to transfer some waste out and back-dilute with water in several steps. In this work, mechanisms and rates of waste solids dissolution and gas releases are evaluated theoretically and experimentally. Particular emphasis is given to crust dissolution processes and associated gas releases, although dissolution and gas release from the mixed-slurry and nonconvective layers are also considered. The release of hydrogen gas to the tank domespace is modeled for a number of scenarios. Under the tank conditions expected at the time of back-dilution, no plausible continuous or sudden gas release scenarios resulting in flammable hydrogen concentrations were identified.

Thermal Analysis Capability of UNF-ST&DARDS

Abstract In the United States, approximately 2500 casks are loaded with commercial spent nuclear fuel (SNF) that has transitioned from wet storage (spent fuel pools) to dry storage. The number of loaded dry storage casks is increasing by approximately 200 each year. Over time, cask designs have evolved to enhance safety and to accommodate more fuel and higher heat loads. Also, higher burnup fuel is being transitioned into dry storage. The SNF is being stored in dry casks for longer times than specified in the original certification period. Several degradation mechanisms related to fuel assemblies and canisters are affected by temperature. For the cladding, temperature-dependent phenomena include creep and annealing, hydride reorientation and embrittlement, and the ductile-to-brittle transition. Temperature can also influence phenomena that affect the long-term integrity of the storage system, including deliquescence, corrosion, and stress-corrosion cracking. Therefore, accurate determination of the temperatures of various components is needed to evaluate potential safety-related issues during transportation after extended storage and to ensure SNF retrievability. The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) is being developed for the U.S. Department of Energy Office of Nuclear Energy to streamline analyses for the waste management system [Nucl. Technol., Vol. 195, p. 124 (2017)]. The thermal analysis capability within UNF-ST&DARDS and example results are discussed herein.

COBRA-SFS Thermal-Hydraulic Analysis Code for Spent-Fuel Storage and Transportation Casks: Models and Methods

Abstract COBRA-SFS, a thermal-hydraulic code developed for steady-state and transient analysis of multiassembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent-fuel-package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent-fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is the capability for detailed thermal radiation modeling within the fuel rod array.

Thermal Modeling of NUHOMS HSM-15 and HSM-1 Storage Modules at Calvert Cliffs Nuclear Power Station ISFSI

As part of the Used Fuel Disposition Campaign of the Department of Energy (DOE), visual inspections and temperature measurements were performed on two storage modules in the Calvert Cliffs Nuclear Power Station’s Independent Spent Fuel Storage Installation (ISFSI). Detailed thermal models models were developed to obtain realistic temperature predictions for actual storage systems, in contrast to conservative and bounding design basis calculations.

MODELING HEAT TRANSFER IN SPENT FUEL TRANSFER CASK NEUTRON SHIELDS – A CHALLENGING PROBLEM IN NATURAL CONVECTION

In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from the spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 10–15 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions. However, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper proposes that there may be reliable CFD approaches to the transfer cask problem, specifically coupled steady-state solvers or unsteady simulations; however, both of these solutions take significant computational effort. Segregated (uncoupled) steady state solvers that were tested did not accurately capture the flow field and heat transfer distribution in this application. Mesh resolution, turbulence modeling, and the tradeoff between steady state and transient solutions are addressed. Because of the critical nature of this application, the need for new experiments at representative scales is clearly demonstrated.Copyright

Buoyancy and Dissolution of the Floating Crust Layer in Tank 241-SY-101 During Transfer and Back-Dilution

To remediate gas retention in the floating crust layer and the potential for buoyant displacement gas releases from below the crust, waste will be transferred out of Hanford Tank 241-SY-101 (SY-101) in the fall of 1999 and back-diluted with water in several steps of about 100,000 gallons each. To evaluate the effects of back-dilution on the crust a static buoyancy model is derived that predicts crust and liquid surface elevations as a function of mixing efficiency and volume of water added during transfer and back-dilution. Experimental results are presented that demonstrate the basic physics involved and verify the operation of the models. A dissolution model is also developed to evaluate the effects of dissolution of solids on crust flotation. The model includes dissolution of solids suspended in the slurry as well as in the crust layers. The inventory and location of insoluble solids after dissolution of the soluble fraction are also tracked. The buoyancy model is applied to predict the crust behavior for the first back-dilution step in SY-101. Specific concerns addressed include conditions that could cause the crust to sink and back-dilution requirements that keep the base of the crust well above the mixer pump inlet.

Validation of COBRA-SFS with Measured Temperature Data from Spent-Fuel Storage Casks

Abstract The COBRA-SFS thermal-hydraulic code has been incorporated into the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System tool as a module devoted to spent-fuel-package thermal analysis. COBRA-SFS has been extensively validated and widely applied to thermal-hydraulic analysis of a large range of spent-fuel storage systems. Instead of recapping that long and detailed history, this paper summarizes the most significant and unique verification and validation of COBRA-SFS, which consists of comparison of code temperature predictions to experimental data obtained in the Test Area North Facility at the Idaho National Laboratory in the 1980s and early 1990s. These data were obtained as part of a program undertaken by the U.S. Department of Energy Office of Civilian Radioactive Waste Management for thermal performance testing of commercial spent-fuel storage cask designs. In total, four casks were tested, and all tests were performed with Westinghouse 15×15 pressurized water reactor spent fuel from the Surry or Turkey Point reactors. COBRA-SFS code results and experimental data comparisons are shown only for the CASTOR-V/21 and the TN-24P casks. CASTOR-V/21 was loaded with the highest decay heat load tested in this program, with individual assembly decay heat values up to 1.83 kW. This effectively bounds storage conditions currently contemplated for high-heat-load systems with test conditions reaching fuel cladding temperatures that approached and in some cases exceeded 400°C, the current regulatory limit for peak cladding temperature in dry storage. TN-24P, with a decay heat load of 20.5 kW, provides comparisons with experimental data that represent a realistic upper bound on typical dry storage initial conditions in independent spent fuel storage installations around the country. The consistency and accuracy of the COBRA-SFS temperature predictions in comparison to the measured data from these casks show that the code appropriately predicts the thermal-hydraulic and heat transfer behavior of these systems. The results presented here provide an excellent illustration of the capability of the COBRA-SFS code to correctly capture all three modes of heat transfer (thermal radiation, conduction, and convection) and the internal circulation of the backfill gas within a spent-fuel package in horizontal or vertical orientation.

Preliminary Thermal Modeling of Hi-Storm 100S-218 Version B Storage Modules at Hope Creek Nuclear Power Station ISFSI

This report fulfills the M3 milestone M3FT-13PN0810022, “Report on Inspection 1”, under Work Package FT-13PN081002. Thermal analysis is being undertaken at Pacific Northwest National Laboratory (PNNL) in support of inspections of selected storage modules at various locations around the United States, as part of the Used Fuel Disposition Campaign of the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) Fuel Cycle Research and Development. This report documents pre-inspection predictions of temperatures for four modules at the Hope Creek Nuclear Generating Station ISFSI that have been identified as candidates for inspection in late summer or early fall/winter of 2013. These are HI-STORM 100S-218 Version B modules storing BWR 8x8 fuel in MPC-68 canisters. The temperature predictions reported in this document were obtained with detailed COBRA-SFS models of these four storage systems, with the following boundary conditions and assumptions.