Harold E. Adkins

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.

Deposition Velocities of Newtonian and Non-Newtonian Slurries in Pipelines

The WTP pipe plugging issue, as stated by the External Flowsheet Review Team (EFRT) Executive Summary, is as follows: “Piping that transports slurries will plug unless it is properly designed to minimize this risk. This design approach has not been followed consistently, which will lead to frequent shutdowns due to line plugging.” A strategy was employed to perform critical-velocity tests on several physical simulants. Critical velocity is defined as the point where a stationary bed of particles deposits on the bottom of a straight horizontal pipe during slurry transport operations. Results from the critical velocity testing provide an indication of slurry stability as a function of fluid rheological properties and transport conditions. The experimental results are compared to the WTP design guide on slurry transport velocity in an effort to confirm minimum waste velocity and flushing velocity requirements as established by calculations and critical line velocity correlations in the design guide. The major findings of this testing is discussed below. Experimental results indicate that the use of the Oroskar and Turian (1980) correlation in the design guide is conservative—Slurry viscosity has a greater affect on particles with a large surface area to mass ratio. The increased viscous forces on these particles result in a decrease in predicted critical velocities from this traditional industry derived equations that focus on particles large than 100 m in size. Since the Hanford slurry particles generally have large surface area to mass ratios, the reliance on such equations in the Hall (2006) design guide is conservative. Additionally, the use of the 95% percentile particle size as an input to this equation is conservative. However, test results indicate that the use of an average particle density as an input to the equation is not conservative. Particle density has a large influence on the overall result returned by the correlation. Lastly, the viscosity correlation used in the WTP design guide has been shown to be inaccurate for Hanford waste feed materials. The use of the Thomas (1979) correlation in the design guide is not conservative—In cases where 100% of the particles are smaller than 74 m or particles are considered to be homogeneous due to yield stress forces suspending the particles the homogeneous fraction of the slurry can be set to 100%. In such cases, the predicted critical velocity based on the conservative Oroskar and Turian (1980) correlation is reduced to zero and the design guide returns a value from the Thomas (1979) correlation. The measured data in this report show that the Thomas (1979) correlation predictions often fall below that measured experimental values. A non-Newtonian deposition velocity design guide should be developed for the WTP— Since the WTP design guide is limited to Newtonian fluids and the WTP expects to process large quantities of such materials, the existing design guide should be modified address such systems. A central experimental finding of this testing is that the flow velocity required to reach turbulent flow increases with slurry rheological properties due to viscous forces dampening the formation of turbulent eddies. The flow becomes dominated by viscous forces rather than turbulent eddies. Since the turbulent eddies necessary for particle transport are not present, the particles will settle when crossing this boundary called the transitional deposition boundary. This deposition mechanism should be expected and designed for in the WTP.

Predictive Bias and Sensitivity in NRC Fuel Performance Codes

The latest versions of the fuel performance codes, FRAPCON-3 and FRAPTRAN were examined to determine if the codes are intrinsically conservative. Each individual model and type of code prediction was examined and compared to the data that was used to develop the model. In addition, a brief literature search was performed to determine if more recent data have become available since the original model development for model comparison.

Spent Nuclear Fuel Structural Response when Subject to an End Impact Accident

The U.S. Nuclear Regulatory Commission (USNRC) is tasked with licensing safe spent fuel storage and transportation systems. A subset of this responsibility is to investigate and understand the structural performance of these systems. In a joint effort between staff at the Pacific Northwest National Laboratory (PNNL) and the USNRC, computational studies were performed to predict the structural response of spent nuclear fuel when subject to an end impact accident. In this study the structural performance of a typical pressurized water reactor fuel assembly is evaluated using the ANSYS® /LS-DYNA® finite element analysis code.Copyright

Test Loop Demonstration and Evaluation of Slurry Transfer Line Critical Velocity Measurement Instruments

This report presents the results of the evaluation of three ultrasonic sensors for detecting critical velocity during slurry transfer between the Hanford tank farms and the WTP.

FUEL ASSEMBLY SHAKER TEST SIMULATION

This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through direct comparison of model results to recorded test results. This does not offer validation for the fuel assembly model in all conceivable cases, such as high kinetic energy shock cases where the fuel assembly might lift off the basket floor to strike to basket ceiling. This type of nonlinear behavior was not witnessed in testing, so the model does not have test data to be validated against.a basis for validation in cases that substantially alter the fuel assembly response range. This leads to a gap in knowledge that is identified through this modeling study. The SNL shaker testing loaded a surrogate fuel assembly with a certain set of artificially-generated time histories. One thing all the shock cases had in common was an elimination of low frequency components, which reduces the rigid body dynamic response of the system. It is not known if the SNL test cases effectively bound all highway transportation scenarios, or if significantly greater rigid body motion than was tested is credible. This knowledge gap could be filled through modeling the vehicle dynamics of a used fuel conveyance, or by collecting acceleration time history data from an actual conveyance under highway conditions.

Hanford Tank Farms Waste Feed Flow Loop Phase VI: PulseEcho System Performance Evaluation

This document presents the visual and ultrasonic PulseEcho critical velocity test results obtained from the System Performance test campaign that was completed in September 2012 with the Remote Sampler Demonstration (RSD)/Waste Feed Flow Loop cold-test platform located at the Monarch test facility in Pasco, Washington. This report is intended to complement and accompany the report that will be developed by WRPS on the design of the System Performance simulant matrix, the analysis of the slurry test sample concentration and particle size distribution (PSD) data, and the design and construction of the RSD/Waste Feed Flow Loop cold-test platform.

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.

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.

THERMAL PERFORMANCE SENSITIVITY STUDIES IN SUPPORT OF MATERIAL MODELING FOR EXTENDED STORAGE OF USED NUCLEAR FUEL

The work reported here is an investigation of the sensitivity of component temperatures of a storage system, including fuel cladding temperatures, in response to age-related changes that could degrade the design-basis thermal behavior of the system. Three specific areas of interest were identified for this study. • degradation of the canister backfill gas from pure helium to a mixture of air and helium, resulting from postulated leakage due to stress corrosion cracking (SCC) of canister welds • changes in surface emissivity of system components, resulting from corrosion or other aging mechanisms, which could cause potentially significant changes in temperatures and temperature distributions, due to the effect on thermal radiation exchange between components • changes in fuel and basket temperatures due to changes in fuel assembly position within the basket cells in the canister The purpose of these sensitivity studies is to provide a realistic example of how changes in the physical properties or configuration of the storage system components can affect temperatures and temperature distributions. The magnitudes of these sensitivities can provide guidance for identifying appropriate modeling assumptions for thermal evaluations extending long term storage out beyond 50, 100, 200, and 300 years.