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At least 1,297 records · Page 72
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

TRITON11 - Westinghouse 11*11 BWR fuel design

Westinghouse has developed an 11*11 BWR fuel design referred to as the TRITON11 fuel design. Given the current economic pressure on the utilities, the key development objective was to significantly reduce the fuel cycle costs. Careful considerations were also made to meet the varying requirements between the different BWR utilities. In particular, the fuel was optimized for both short and long cycle operation, as well as for the up-rated cores and higher burnups. In addition to these advances, careful considerations were given to the overall reliability of the new design to further reduce the risk of fuel failures. Further improved fuel economy requires improved thermal margins since both SVEA-96 Optima2 and Optima3 are already highly optimized for reactivity performance at the fuel lattice level. It was realized that better thermal margins must come from adding more fuel rods, thereby reducing the linear heat generation rate and surface heat flux. This requires going beyond the 10*10 fuel lattice. An 11*11 fuel design was chosen as optimum, considering benefits versus manufacturing cost and added complexity. The numbers, dimensions and positioning of fuel rods and internal water channels (passing non-boiling water) were analyzed intensively. The resulting TRITON11 design has 109 fuel rods and 3 central water rods for improved moderation. The paper presents the TRITON11 fuel design. The development objectives are laid out and the resulting performance and reliability improvements are described. The development is in the final phase of testing and validation. Examples of recent tests are presented. The advances in BWR core analysis codes in support of TRITON11 fuel are briefly summarized. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

DNB evaluation of PWR steam-line break event using Westinghouse and CASL coupled code systems

ANCKVIPRE and the VERA-CS code systems have been successfully applied to predicting PWR core response with respect to departure from nucleate boiling (DNB) at the limiting time step of a postulated hot zero power (HZP) main steam-line break (MSLB) event with and without offsite power available. ANCKVIPRE links the ANC code with the VIPRE-W code. ANC is Westinghouse's multi-dimensional nodal diffusion code for all nuclear core design calculations. It predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. VIPRE-W is the Westinghouse version of a thermalhydraulic subchannel code developed for light water reactor core design applications. The reactor core simulation was based on input of the core boundary conditions from RETRAN system transient and STAR-CCM+ CFD core inlet distribution calculations. To account for the asymmetric power distribution in the reactor core due to the broken steam pipe in one loop and the stuck RCCA, both ANCKVIPRE and the VERA-CS models were set up for the whole reactor core. The advanced modeling and simulations confirmed that the HZP MSLB high-flow case with offsite power available is more limiting with respect to the DNB acceptance criterion, consistent with the limiting case analyzed in the 4-loop plant Safety Analysis Report (SAR). The results also demonstrated that the currently NRC-approved Westinghouse coupled code system such as ANCKVIPRE remains conservative.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

CASL COBRA-TF improvements for PWR DNB analysis

COBRA-TF (CTF) is a thermal hydraulic (T/H) subchannel code using either three-dimensional (3D) Cartesian or subchannel coordinate formulations for two-phase fluid flow and heat transfer solutions. CTF has been improved under the Consortium for Advanced Simulation of Light Water Reactors (CASL) program for Pressurized Water Reactor (PWR) applications, including software optimization, new closure models, pre- and post-processing and parallelization for modeling full reactor core T/H responses under normal operating and accident conditions. As a result of collaboration among CASL partners including the Westinghouse Electric Company, the Oak Ridge National Laboratory (ORNL), and the Sandia National Laboratories, additional modeling improvements were made to CTF specifically for PWR Departure from Nucleate Boiling (DNB) analysis, including a code option to evaluate fuel thermal margin in terms of DNB Ratio (DNBR) and an axial shape factor to account for effect of non-uniform axial power distribution on DNB. Multiple DNB correlations are now linked with CTF for different applications, including the Westinghouse proprietary WRB-1 correlation for fuel designs containing mixing vane grid spacers. The improved CTF code with the WRB-1 correlation (CTF/WRB-1) was validated using the DNB data from the PWR Subchannel Bundle Tests (PSBT). In addition to the comparison with the test data, the CTF/WRB-1 DNBR results and the associated local fluid conditions were compared to the results of the Westinghouse T/H design code, VIPRE-W, which is an enhanced version of the VIPRE-01 code originally developed by the Electric Power Research Institute (EPRI). The comparisons showed that CTF/WRB-1 DNBR predictions are in good agreement with the VIPRE-W results within the applicable range of the DNB correlation. A model sensitivity study was performed to confirm that the CTF void drift model had an insignificant effect on DNBR under the steam line break (SLB) low pressure condition. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Cold-cap temperature profile comparison between the laboratory and mathematical model - 15591

The rate of waste vitrification in an electric melter is connected to the feed-to-glass conversion process, which occurs in the cold cap, a layer of reacting feed on top of molten glass. The cold cap consists of two layers: a low-temperature (∼100 deg. C - ∼800 deg. C) region of unconnected, reacting feed and a high-temperature (∼800 deg. C - ∼1100 deg. C) region of molten glass with gas bubbles and cavities. A recently developed mathematical model describes the effect of the cold cap on glass production. For verification of the mathematical model, a cold cap produced in a laboratory-scale melter was subsequently analyzed by correlating structural features of a polished region with those obtained in heat-treated feed samples to determine the temperature profile as a function of cold-cap position. After corrections were made for the diminished volume of bubbles and collapsed cavities that occurred in the cold-cap sample during cooling, the temperature gradients were alike in the model and the laboratory cold cap through the range from ∼500 deg. C to ∼1100 deg. C. The temperature profiles were different in the upper region of the cold cap because the radiant heat from the molten glass in the laboratory-scale melter resulted in dry and hot (∼400 deg. C) surface conditions whereas the model assumed a surface covered by boiling slurry. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

The hydrodynamics of the formation of microlayers beneath vapour bubbles growing on a heated substrate

Thin layers of liquid are known to be 'left behind' beneath essentially the entire extent of the (approximately) hemispherical bubbles that grow rapidly at low pressures on heated substrates. The formation of these 'microlayers' has been well established by modern sophisticated experimental observations. Whilst the rapid evaporation of these layers can make a dominant contribution to the formation and growth of these bubbles, their effect is incorporated into CFD boiling analysis via semiempirical 'sub-grid scale' modelling, and the physics of the formation and the prediction of the characteristics of the microlayers have received little attention. We present here an investigation into their formation via the solution of the full Navier Stokes equations for viscous flow. This investigation has allowed us to demonstrate that the overall extent of the micro-layer is a function of the balance of inertial and surface tension forces that control the macroscopic expansion of the bubble. The inertial effects inhibit the growth of the bubble away from the wall, and force it to adopt a more hemispherical shape, with a larger 'contact' area with the substrate. Beneath this 'contact' area a micro-layer is formed by liquid being left behind as the bubble grows. The depth of the micro-layer left behind seems to depend largely upon the viscous forces at work in the flow near the wall. (authors)

42 ENGINEERING↗
DOE OSTI2016

Separation of Zirconium from Zircaloy-2 for Spent Nuclear Fuel Cladding Applications using a Chloride Volatility Process

The United States produces a large amount of cladding material from used nuclear fuel, expected to approach 1,000 MT/year for the next 50 years. The recovery of zirconium from the cladding material would represent a ∼50% savings on the geologic waste disposal requirement, and savings of >$40 M/year due to the cost of hafnium-free zirconium. Previous simulation studies have shown that in the irradiated cladding, the major activation products found are {sup 125}Sb, {sup 60}Co, and {sup 93m}Nb, and the major fission product found is {sup 137}Cs. These radioisotopes make it impossible to treat the cladding as low level waste. In a chlorination method, zirconium cladding reacts with Cl{sub 2} at 350-380 deg. C to produce mainly ZrCl{sub 4}. The zirconium tetrachloride has a relatively low boiling point (sublimes at 330 deg. C) which allows its separation from other products, and so it can be treated as low level waste for disposal or recycling. This chlorination approach has been proposed for the reprocessing of zirconium. However, previous work has observed element impurities (Fe, Sn, Cr, Cs, Sb, and Nb) in the recovered ZrCl{sub 4} meaning that additional steps are necessary in a chlorination process previous to the reduction of ZrCl{sub 4} to the metallic form. The chlorination process for zirconium reprocessing has been highly encouraged as it has proven more efficient than previous mechanical processes, redox processes, and other chemical methods such as iodination. The overall goal of this project is to develop the road map for a chlorination process that will yield high purity zirconium. The current study focuses on the recovery of ZrCl{sub 4} from non-irradiated alloys used as nuclear fuel cladding. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Adsorption and Transport Properties of Zeolite SAPO-34 for Krypton/Xenon Separations

Nuclear power is an economical means of providing carbon-free energy, but requires effective ways to deal with the radioactive waste products. As part of the overall process for recycling used nuclear fuel (UNF), it is necessary to develop technologies to separate the off-gases formed during the fission of uranium and plutonium in nuclear power reactors. The off-gas contains {sup 129}I (radioisotope), NO, NO{sub 2}, CO{sub 2} containing {sup 14}C (radioisotope), {sup 3}H{sub 2}O (radioactive tritium oxide), {sup 85}Kr (radioisotope), and {sup 136}Xe. Separating and capturing each gas is essential to reduce the volume of radioactive waste for storage. Currently there exist economical technologies to separate {sup 129}I and {sup 14}CO{sub 2}. However, Kr and Xe separation is challenging because they are chemically inert and physically very similar. The Kr/Xe mixture is approximately 10/90 by volume and hence the separation of Kr would reduce the volume of radioactive {sup 85}Kr waste by a factor of 10. One way of separating them is by cryogenic distillation wherein they are condensed out as pure streams at very low temperature (Kr boiling point: -153 deg. C, Xe: -108 deg. C). Cryogenic distillation has intensive power requirements and a large system volume, resulting in high capital, operating, and disposal costs. Membrane-based separation using radiation-stable materials such as nanoporous zeolites is an attractive alternative, if membranes with sufficient Kr flux and selectivity can be fabricated. An ideal membrane process would permeate a stream of pure Kr for storage and also produce a retentate of pure Xe which has commercial value (> $8/g at 99+% purity from Airgas{sup R}). Zeolites are nanoporous crystalline aluminosilicate materials with pores that are capable of separating molecules by shape and size. SAPO-34 is a silicoaluminophosphate zeolite with pore size of 0.38 nm. The SAPO-34 is a molecular sieve composed of 3D networks with CHA frame structure containing uniform micropores. Since this pore size is intermediate between the kinetic diameters of krypton (0.36 nm) and xenon (0.396 nm), it was hypothesized that SAPO-34 membranes can separate krypton and xenon by diffusivity differences. Previously we fabricated SAPO-34 membranes and showed that they are capable of separating Kr from Xe. Here we describe a more detailed study of the fundamental adsorption and transport properties of SAPO34. In particular, the permeabilities of Kr and Xe in SAPO-34 are dependent on both their adsorption and diffusion characteristics in the nanoporous zeolite. Typically, nanoporous materials show stronger adsorption of Xe than Kr (due to stronger van der Waals forces) but faster diffusion of Kr (due to smaller kinetic diameter). Therefore the overall membrane selectivity towards Kr will be determined by competition between these two effects. A main objective of this study is to quantify and understand adsorption, diffusion, and permeation in detail and thereby obtain: (1) fundamental parameters useful for design calculations of a Kr/Xe separation membrane process, and (2) an experimental and modeling framework that can provide guidance on the expected limits of performance of SAPO-34 and other zeolite materials. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Nitride Fuel Potential for Power Uprates

This work presents the methodology and current results for design analysis of grid supported pressurized water reactor cores using uranium nitride (UN) fuel in square arrays. Uranium nitride has a higher theoretical density and higher thermal conductivity than uranium oxide, allowing for lower fuel temperatures, longer cycle length and higher power densities with similar enrichment. In order to avoid the UN and high temperature water reaction, silicon carbide cladding has been proposed in place of Zircaloy cladding. The analysis aims at finding the maximum thermal power over a range of bundle geometries (rod diameter and pitch). The same methodology is applied to the design of UO{sub 2} oxide fueled cores to provide a fair comparison of the achievable power rating between the two fuel types. Steady-state limits are considered, including fuel bundle pressure drop, minimum departure from nucleate boiling ratio (MDNBR), fuel temperature, fuel rod vibrations and pin fretting and sliding wear. Transients such as loss of flow and overpower accident will be considered. A steady state analysis was undertaken to determine the maximum achievable power that can be sustained by square array fuel assemblies loaded with nitride and oxide fuels with same or better safety performances as the reference design. Steady-state design limits were applied including MDNBR, fuel bundle pressure drop, fuel temperature, cumulative sliding and fretting wear. The major conclusions of this study are as follows: 1) It is possible to achieve significant power increase for UN and UO{sub 2} compared to the reference core power output of 3400 MWth with smaller but more numerous rods and higher pitch over diameter ratios. A power uprate of 25% can be reached according to the VIPRE-MATLAB interface for full core analysis with the same number of spacer grids, and 32% according to the new MATLAB script for single hot channel analysis, with five more spacer grids. For this latter core, a higher enrichment would be required. 2) The nitride fuel temperature constraint from dissociation doesn't influence steady state performances as other design constraints such as MDNBR and pressure loss are more restrictive. The oxide fuel temperature constraint from fission gas release is reached in part of the study domain, but not in the region that allows for possible uprates. 3) As a result, the thermal hydraulic performance of UN and UO{sub 2} are similar. Differences in possible power uprates will arise from fretting and sliding wear. Wear depends of the time spent by the fuel in the core, which depends on burnup and heavy metal inventory. UN being denser, it can allow for power uprates or increased cycle length. There is a trade off between power uprates and cycle length with regards to fuel pin wear. Reducing fuel pin wear by reducing cycle length can allow for higher thermal power. This affects the plants' economics as well. This work will be completed by a transient analysis using CASMO4E, SIMULATE3 and SIMULATE3k to model the neutronics of a nitride fueled core. The transient performances will be used to determine the maximum power uprates while performing as well or better during transients as the reference design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Sensitivity Study of Void Coefficients in the SHEBA-II Critical Assembly using MCNP6

SHEBA-II was an aqueous critical assembly experiment that was operated at the Los Alamos Critical Experiments Facility. It is a bare assembly that is fueled with an aqueous solution of ∼5% enriched nonstoichiometric uranyl fluoride. The composition of uranyl fluoride changes over time due to chemical and nuclear reactions taking place within the assembly. This study investigates how changes in the solution temperature, void volume, and void location effect criticality, and to provide information about potential hazards in off-normal situations. This study had extended previous work on the SHEBA-II CAV to include temperature dependence, void volume dependence, and the thermochemical properties of the solution (density). The overall results indicate that a critical boiling solution cannot be reached. The solution height needed within the tank to become critical would overfill the tank. For future work, these results will be used to calculate the temperature effects on void reactivity coefficients. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Plant-wide simulation model for transient studies on the XE-100 - 18464

This paper gives an overview of the plant-wide simulation model that is used by X-energy LLC for developing their plant design and control philosophy, as well as assessing plant upset conditions. The simulation model was created in Flownex{sup R} thermal-fluid system simulation software, which solves mass, momentum and energy equations for compressible, incompressible and homogeneous two-phase flow. The plant model encompasses all major components, from reactor to condenser, including main control loop elements. The paper pays particular attention to the validation of the two-phase flow framework used for the secondary steam cycle. Two steady-state validation cases are done using operational data from the THTR. Validation of transient operation is done by simulating density wave oscillations, a type of system-level boiling phenomenon. This is regarded as a thorough test of transient capability because of the challenges this phenomenon poses to several aspects of the calculation and to the consistency of the code as a whole. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Simulation and control of a passively cooled small modular reactor Samet - 142

This work regulates a small modular reactor (SMR) with traditional control methodologies and analyzes its response to typical disturbances. An analytical model for the passively cooled NuScale SMR is developed to represent the essential dynamics. The MATLAB/Simulink based model consists of reactor core, steam generator, pressurizer, hot leg riser and downcomer, and turbine representations. The point kinetics equations with a single combined neutron precursor group and the lumped parameter representations with single-phase natural circulation account for the neutronics and thermohydraulics in the reactor core. A lumped parameter, moving-boundary approach forms the basis of the nonlinear model adopted for the once-through helical-coil heat exchangers in which boundaries between regions of different fluid states (i.e., subcooled, boiling, and superheated) vary over time. A proportional-integral-derivative controller employing a sliding-average temperature strategy commonly used for pressurized water reactors is implemented. The responses of the SMR to two different disturbances are analyzed and discussed in detail. For the first scenario, a 5% step increase in the load is applied to the system, which results in a change in the steam valve opening. For comparison, two different simulations under this same disturbance are run: one including and one excluding the control systems. The second scenario tests the effectiveness of the control system to increase (or decrease) the reactor thermal power to a certain level within a desired time period. The simulation results show that important system variables are kept at the desired values by the proposed control strategy. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Thermal Hydraulics Safety Analyses of TRIGA Reactor Upon Fuel Reconfiguration

Safety analyses of the Training, Research, Isotope, General Atomics (TRIGA) reactor at Kansas State University(KSU) is performed using thermal hydraulic simulations based on changes to reactivity or the volumetric heat generation rate due to changes in the positioning of the enriched fuel elements in the reactor core. If the reactivity or neutronics parameters are known, then they can be used to simulate the thermal energy state of the material which is crucial to assess the safety of the nuclear fuel inside a nuclear reactor. Several thermal hydraulics models and codes have been developed in the past to assess reactor safety. RELAP (Reactor Excursion and Leak Analysis Program) is one such tool widely used throughout the world for safety analyses. Although RELAP is considered a reliable tool by regulators, vendors, and operators, it lacks the numerical fidelity provided by 3-D computational fluid dynamics (CFD) models. This work involves the critical task of assessing safety of a research reactor upon fuel reconfiguration based on models developed in RELAP and comparing them with a commercially available 3-D CFD program. The TRIGA Mark-II reactor, designed by General Atomics and located at Kansas State University, is a class of small pool type inherently safe research reactor licensed to operate at steady state power of up to 1.25 MW. It has a cylindrical core measuring 0.45 meters diametrically and 0.38 meters axially positioned in a tank measuring 1.98 meters diametrically and 6.25 meters axially. It has uranium zirconium hydride (UZrH) fuel rods housed in stainless steel cladding material arranged in five concentric rings (B,C,D,E and F rings, in that order, moving outwards from the center). TRIGA fuel is nominally 20% enriched, but the reactivity of the fuel can be changed by using burnable poisons (not used at KSU), or by varying the total weight percent of uranium in the UZrH fuel meat. The TRIGA reactor is designed to be an inherently safe reactor and is designed to be cooled by natural convection. Thus it is of paramount importance to avoid the film boiling regime; failing to do so will subject the fuel to severe conditions, possibly resulting in its failure due to fuel/clad interactions. A high DNBR ratio is a primary consideration in the reactor core design and thus any modification in the core should ensure that this criteria is always met. In order to fully utilize the capabilities of KSU TRIGA reactor facility, it should be operated at its full capacity, i.e. 1.25 MW, which requires increasing the fuel loading in some of the E- and F-ring elements from 8.5% U (nominal) to 12.0% U (nominal). Only four elements are to be switched to 12% loading, causing local power peaking in these elements. The degree of power peaking to be expected in these 12%-loaded fuel elements was calculated based on a detailed MCNP (Monte Carlo N-Particle code) model of the reactor core. Some preliminary results of the calculations are presented in this paper based on changes in peaking factors and power levels. To predict reactor thermal behavior, axial changes in the peaking factors obtained from MCNP models were used to calculate volumetric energy density functions for a single channel one dimensional RELAP model and a three dimensional COMSOL model of the TRIGA MARK-II reactor. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Benchmark of Spacer Grid Models used in Nuclear Power Plants

The fuel bundles of both PWR and BWR reactors require spacers to maintain the integrity and spacing of the slender fuel rods. These spacers can provide benefits such as increased mixing and turbulence, which increases heat distribution, leading to better Departure from Nucleate Boiling (DNB) conditions in PWRs. They also enhance the critical power ratios by increasing the thickness of the liquid layer in BWRs. However, the spacers also cause a pressure drop which must be engineered into the performance and operation. The flow within a rod bundle involves a number of complicated factors, such as cross flow, (de)entrainment of liquid layer, bubble or droplet break up/join. This requires specialized 3 dimensional computational fluid dynamics for the complete flow, however, most of the newly developed models have been accomplished by industry, so they are proprietary and in addition require extensive programming and cost to implement. In the public domain, a number of 1 dimensional equations have been developed that could benefit in initial design and decision making, if the equations are valid in the new and different applications the models are used. In 2002, an OECD/NRC BFBT benchmark workshop determined the variables used in the development of models, to predict flow within nuclear fuel bundles, had not been updated with technological advances to replace general theory and characteristics. If those properties could be identified and modeled, they could be applied to a wider range of conditions. This study utilizes NUPEC BFBT tests' measured data, on an average spacer over a range of pressure and flows, which was determined to be the most detailed measured data available. The focus is to describe the mathematical models of spacer grids and validate them with experimental results. This study consists of compiling and comparing available and established single phase equations and models computations. While these models may not have the desired degree of accuracy for final decisions, they may be accurate enough for the initial phases of design. No conclusion on a universal equation that fits all boundary conditions accurately or detailed effect of properties was determined. However, by using the BFBT experimental data, boundary conditions where the spacer grid models can be useful were identified. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Experimental Study of Molten Metallic Fuel Relocation in Sodium-Filled Fuel Structures

The Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) is being developed by Korea Atomic Energy Research Institute (KAERI). While their design incorporates inherent safety features demonstrated in the landmark Shutdown Heat Removal Test (SHRT) conducted at EBR-II including a large pool configuration and uranium alloy fuel, it is still necessary to evaluate metal fuel relocation and theirs consequences in case of postulated core disruptive accidents. Extensive data regarding fuel relocation in sodium from both in-pile and out-of-pile tests were obtained for oxide fuels in the 1960's - 1970's, while few studies of metallic fuel behavior were performed because of concerns that metallic fuel could not achieve high burnup due to the irradiation induced swelling problem. Following new metal fuel design development that mitigated the swelling problem, experiments regarding interaction of molten metal fuels and liquid sodium have been conducted by several researchers. Most studies that were conducted using metal simulants in various liquids concluded that the molten metals were heavily fragmented after liquid contact and the dominant factors governing the fragmentation were superheating and latent heat of fusion of the molten metal fuel. Gabor et al. conducted experiments using kilogram quantities of various molten uranium alloys in an open sodium pool configuration. They found that the metal fuel fragments were basically in forms of filaments and sheets with a high bed voidage in the order of 0.9. Their calculation indicated that the debris beds formed by relocated metal fuels in an open sodium pool would be largely coolable by conduction heat transfer; and even if deep beds were to form, convective heat transfer and boiling heat transfer could preclude further melt penetration. Although Gabor's tests are useful in evaluating the coolability of relocated metal fuels during a postulated accident in a SFR, they focused on late phase behavior in which the core material relocates downward through the core support plate into the inlet plenum where the structures are more open and thus consistent with the open pool configuration. There is still a knowledge gap related to nature of metal fuel fragmentation and relocation within the fuel assembly structures. Argonne National Laboratory (ANL) is planning to conduct experiments regarding metal fuel relocation in the core structures which mock up the PGSFR's fuel assembly design. In the present study, preliminary tests for demonstration of melting and injecting the metal fuel were conducted using copper. Metal fuel relocation experiments with single fuel pin test section are planned to be conducted for understanding the flow and freezing behavior of the molten metal fuel in a narrow sodium channel and for evaluating the coolability of the relocated fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Improvement of Wall Drag Modeling Capabilities of CTF

Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF), or CTF, is a nuclear thermal hydraulic subchannel code used throughout academia and industry. It solves discretized forms of nine coupled equations for the mass, momentum, and energy equations for the liquid and vapor fields; the mass and momentum equation for the droplet field; and the energy equation for solid structures. This work focuses on improving the wall drag modeling capabilities of CTF by adding new friction models that account for surface roughness. Prior to this work, CTF used simple correlations that assumed that surfaces were smooth. In reality, different materials in the reactor have varying degrees of roughness that act to increase drag and frictional pressure drop. The frictional pressure drop is important, as it partly defines the driving head necessary for achieving certain flow conditions. This is particularly important in passively cooled systems that rely on natural circulation. Therefore, the effects of surface roughness cannot be neglected, as it affects flow rate and distribution, heat transfer and turbulence. (authors)

42 ENGINEERING↗
DOE OSTI2016

Advances in Fuel Rod Modeling in Subchannel Thermal-Hydraulic Codes

Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF), or CTF, is a nuclear thermal hydraulic subchannel code used throughout academia and industry. It solves discretized forms of nine coupled equations for the mass, momentum, and energy equations for the liquid and vapor fields; the mass and momentum equation for the droplet field; and the energy equation for solid structures. This work focuses on improving the temperature predictions of CTF by adding new fuel thermal conductivity models that account for burnup, gadolinia content in addition to temperature. Prior to this work, CTF used a temperature dependent fuel thermal conductivity model. In reality, the fuel thermal conductivity is affected by fission product buildup and radiation damage under irradiation. This results in fuel thermal conductivity degradation, and correspondingly an increase in fuel centerline temperature. Accounting of the irradiation effects is particularly important for correct prediction of fuel temperature profile for the Doppler feedback calculation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Uncertainty Analysis of CTF for the Uncertainty Analysis in Modeling Benchmark

This work describes the results of a quantitative sensitivity and uncertainty analysis of the thermal hydraulic subchannel code, Coolant Boiling in Rod Arrays - Three Field (COBRA-TF). Four steady state cases from Phase II, Exercise 3 of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Uncertainty Analysis in Modeling (OECD/NEA UAM) Benchmark are analyzed using the statistical analysis tool, Design Analysis Kit for Optimization and Tera-scale Applications (Dakota). This study supports the larger goal of the UAM Benchmark, which is to assess the uncertainty inherent in nuclear codes in a straightforward framework. The benchmark first focuses on assessing the uncertainty in stand-alone codes and then uses these results to propagate those uncertainties to coupled calculations. The work presented here adds to the existing results for this benchmark. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Critical Heat Flux Model Improvement in CTF for Natural Circulation Type Reactors

The objective of this study was to enhance the departure from nucleate boiling (DNB) modeling capabilities of the subchannel analysis code CTF. The intent was to include consideration for PWR-type reactors which are characterized by relatively low subchannel mass flux and high inlet subcooling during normal operation. Identifying a CHF model appropriate for these type of conditions is required for an accurate thermal-hydraulics performance analysis. Several CHF models that are available in open literature, including those already implemented in CTF, were studied for their applicability. This includes the CHF models by Westinghouse (W-3), Electric Power Research Institute (EPRI-1), Biasi, Bowring, and Groeneveld. Since the models were developed within a certain validity range of operating conditions (mass flux, equilibrium quality, pressure, hydraulic diameter, and others), these ranges were used as a measure of acceptability. It was found in these considerations that the Groeneveld and Bowring models were the most appropriate. Therefore, the models by Bowring and Groeneveld were selected for implementation in CTF. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗