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DOE OSTI2017

Review of CTF's Fuel Rod Modeling Using FRAPCON-4.0's Centerline Temperature Predictions

Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF), or CTF1, is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF's fuel rod modeling is originally developed for VIPRE code. Its methodology is based on GAPCON and FRAP fuel performance codes, and material properties are included from MATPRO handbook. This work focuses on review of CTF's fuel rod modeling to address shortcomings in CTF's temperature predictions. CTF is compared to FRAPCON which is U.S. NRC's steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes in fuel rod variables and temperatures including the effects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite difference or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF's dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF's models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF's dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF's dynamic gap conductance model. Improving the CTF's dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Thermal Calculations to Determine Maximum Zircaloy Temperatures During Reactor Operations for Campaign 1 Fuel Assemblies

The H-Canyon Facility plans on processing bundles from Non-Aluminum Spent Nuclear Fuel (NASNF) Risk Group 1 (UO 2 core fuels). Uranium-zirconium (UZr) alloys and zirconium clad UO 2 fuel containing UZr intermetallic compound have the potential to cause explosions during dissolution. UZr intermetallic typically only forms when uranium and zirconium (or zirconium alloy) cladding are in contact at high temperatures. Thermal modeling was performed to determine the maximum cladding temperature for Campaign 1 Zircoloy cladded fuel assembles during reactor operation to help determine if the UZr intermetallic compound could have formed during reactor operations. The temperature of 350°C was chosen as a conservative minimum value for the possible formation of intermetallic UZr compound. The modeling approach, inputs, and results are outlined in this document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

RISMC Industry Application no. 1 (ECCS/LOCA) Core characterization automation: Reference PWR designs for IA no. 1

The U.S. NRC is currently proposing rulemaking '10 CFR 50.46c' to revise the loss-of-coolant-accident (LOCA) and emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burn-up on cladding performance. The key implications of this proposition are that both cladding and ECCS performance need to be considered in a coupled manner. This implies that LOCA methodologies will need to be updated for regulatory purposes. The Risk-Informed Safety Margin Characterization (RISMC) Pathway, as part of the DOE Light Water Reactor Sustainability (LWRS) research and development program initialed several programs at the Idaho National Laboratory (INL) to support the industry by offering potential solutions to this new LOCA/ECCS analysis. One of these programs is the Industry Application no. 1 (IA1). It's goal is to develop an integrated evaluation model called LOCA Toolkit for the US (LOTUS) to revisit how risks are evaluated in a LOCA under 10 CFR 50.46c. The focus of LOTUS is to establish the automation interfaces among the physics disciplines involved, namely the core design automation, the fuels/clad performance, the system analysis, the risk assessment and the core design optimization. The risk-informed margins management approach for IA1 will provide a means of quantifying the impact on the key LOCA analysis figures-of-merit like peak cladding temperature (PCT), equivalent cladding reacted (ECR), etc. of a change in LOCA analysis inputs. The information obtained can then be used for decision-making and margin management. The modular approach for LOTUS will enable plant owners/vendors to further customize the LOTUS framework for use within their established codes and methods. One of the LOTUS modules deals with the core design automation (CD-A). The purpose of this module is to supply the LOCA analysis with initial conditions, i.e. assembly/pin power histories, power shapes, etc. to be employed by the Fuels/Clad Performance (EP) and System Analysis (SA) modules. In the CD-A module, the plant owner/operator will characterize his core design with the tools he has. For the current LOTUS demonstration, the tools used are the already coupled codes PHISICS and RELAP5-3D2,3. Cross section generation will be done using the Studsvik lattice code HELIOS-2.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Post-Irradiation Examination for Conventional and Advanced Nuclear Fuel

During irradiation in a nuclear reactor, the fuel pellet experiences non-uniform fission rate and strong thermomechanical loads, which result in alterations and restructuring occurring non-homogeneously in the fuel volume. The irradiated fuel is characterized by a highly heterogeneous structure, with a correspondingly heterogeneous distribution of phases and properties. It is a big technical challenge for a hot cell facility to be able to cover effectively a broad variety of fuel types and compounds, characterized by different compositions, physico-chemical properties, geometries and configurations. The first challenge is to provide testing environments and procedures which do not alter or degrade the original irradiated fuel properties and are technically suitable for hot cell or shielded facility operation. For instance, non-oxide fuel compounds require protective atmosphere to avoid sample oxidation during storage, preparation and testing. Moreover, suitable adaptations of tools and measuring devices must be available to characterize fuel configurations which differ from the conventional cylindrical pellet-cladding geometries (e.g. spherical fuel elements for pebble bed HTR and TRISO coated fuel particles). Finally, adequate scientific methods of investigation must be available to measure properties and study the behaviour of highly heterogeneous irradiated fuel. In addition to basic techniques for non-destructive and destructive examination of nuclear fuel systems (e.g. covering fuel and cladding or fuel and coatings, or other configurations), 'in-depth' methods are applied to determine physical, thermomechanical and micro-analytical properties of specific phases/features present in irradiated fuel. This paper presents relevant trends and examples from fuel safety studies performed at JRC-Karlsruhe and in other laboratories which highlight the 'added value' obtainable by applying combined techniques, high resolution tools or 3D methods of analysis for the experimental 'multiscale' characterization of irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

SCALE Lattice Physics Code Assessments of Accident Tolerant Fuel

This report highlights accident-tolerant fuel (ATF) code assessment activities performed under Project NRCHQ- 60-17-T-0017, Lattice Physics Enhancements and Assessment. ATF covers a broad range of advanced fuel and clad designs for light water reactors (LWRs) to enhance performance under several accident conditions. Several ATF concepts are anticipated to be deployed as lead test rods (LTRs) or lead test assemblies (LTAs) within the next five years. The purpose of this work is to assess the predictive capabilities of NRC neutronics codes that underpin various licensing calculations. ATF designs use different fuel and clad materials compared to standard UO 2 fuel and zirconium-alloy claddings (hereafter UO 2 -Zry). These new materials and geometrical designs need to be assessed to quantify the impact of nuclear data uncertainties on quantities of interest (QOIs) in licensing calculations and the impact of modeling approximations which may be valid for UO 2 -Zry but not for ATF. This report outlines a systematic approach for ATF neutronics code assessment which includes sensitivity and uncertainty (S/U) analysis of nuclear data, identification of experimental benchmark and gaps for code validation, investigation of modeling approximations, and code-to-code comparisons of calculated QOIs against high-fidelity reference continuous energy (CE) Monte Carlo (MC) calculations. This report focuses on the assessment of the SCALE/Polaris lattice physics code for reactor safety analysis. Polaris lattice physics calculations generate few group (FG) cross sections for PARCS full-core calculations. (Full-core analysis, spent fuel analysis, and severe accident analysis will be performed in future work.) The selected ATF concepts for this report include Cr 2 O 3 and Al 2 O 3 -Cr 2 O 3 -doped UO 2 fuel, U 3 Si 2 fuel, FeCrAl cladding, SiC cladding, and Cr-coated cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Sister Rod Destructive Examinations (FY2021) Appendix J: Leaching of High Burnup Used Nuclear Fuel in De-ionized Water

As a part of the DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4, ZIRLO, and M5. The leaching experiment was a part of a larger study by Yadukrishnan Sasikumar in defending his doctoral thesis at The Open University, School of Engineering and Innovation [13]. It utilized post-CIRFT Sister Rod test specimens that were slated for disposal. The leaching experiment aims to understand the trends in the radiolytic dissolution of High Burnup Spent Nuclear Fuel (SNF) when exposed to water (e.g., in-reactor or in-pool cladding failures). Specimens from a baseline M5-clad rod and a heat-treated M5 rod were cut from the fractured CIRFT specimens and placed in 100 mL de-ionized water for a period of 128 days. Both radial and axial sections were cut to provide different surface areas of fuel in contact with the leachate. During the four-month exposure period, aliquot samples of the leachate were analyzed gamma spectroscopy and inductively coupled plasma mass spectrometry (ICP-MS). The analysis quantified the amount of fuel leached into the solutions and provided individual isotopic release fractions (of 30+ isotopes) which were compared as a function of time and surface area of fuel exposed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Uncertainty quantification and sensitivity analysis applications to fuel performance modeling

Best-estimate fuel performance codes such as BISON currently under development at the Idaho National Laboratory, utilize empirical and mechanistic lower-length-scale informed correlations to predict fuel behavior under normal operating and accident reactor conditions. Traditionally, best-estimate results are presented using the correlations with no quantification of the uncertainty in the output metrics of interest. However, there are associated uncertainties in the input parameters and correlations used to determine the behavior of the fuel and cladding under irradiation. Therefore, it is important to perform uncertainty quantification and include confidence bounds on the output metrics that take into account the uncertainties in the inputs. In addition, sensitivity analyses can be performed to determine which input parameters have the greatest influence on the outputs. In this paper we couple the BISON fuel performance code to the DAKOTA uncertainty analysis software to analyze a representative fuel performance problem. The case studied in this paper is based upon rod 1 from the IFA-432 integral experiment performed at the Halden Reactor in Norway. The rodlet is representative of a BWR fuel rod. The input parameters uncertainties are broken into three separate categories including boundary condition uncertainties (e.g., power, coolant flow rate), manufacturing uncertainties (e.g., pellet diameter, cladding thickness), and model uncertainties (e.g., fuel thermal conductivity, fuel swelling). Utilizing DAKOTA, a variety of statistical analysis techniques are applied to quantify the uncertainty and sensitivity of the output metrics of interest. Specifically, we demonstrate the use of sampling methods, polynomial chaos expansions, surrogate models, and variance-based decomposition. The output metrics investigated in this study are the fuel centerline temperature, cladding surface temperature, fission gas released, and fuel rod diameter. The results highlighted the importance of the thermal conductivity of the fuel, the thermal conductivity of the cladding, the gap thickness, the fuel thermal expansion, and the system power factor as influential parameters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

High-burnup Experiments in Reactivity Initiated Accidents (HERA)

High-burnup Experiments in Reactivity-initiated Accidents (HERA) is a joint experimental program (JEEP) operating within the United States (U.S.) Nuclear Energy Agency’s (NEA’s) framework for irradiation experiments (FIDES). HERA is dedicated to the understanding of light water reactor (LWR) fuel performance at high burnup under reactivity-initiated accidents (RIA). In-pile RIA experiments have been performed on high-burnup fuels (e.g., above 60 gigawatt days per metric ton of uranium [GWd/MTU]) in the CABRI reactor at the Cadarache site in southern France, and the Nuclear Safety Research Reactor (NSRR) in Japan. However, most of these experiments have taken place with heavily corroded Zircaloy claddings with pulse widths that are narrower (e.g., 5 ms – 30 ms full-width-half-max [FWHM]) than what would be likely in a commercial LWR (e.g., 30 ms – 80 ms FWHM). A few tests were performed in the CABRI facility with FWHM up to 76 ms, though only one “blistered” rod experienced cladding failure. Heavy waterside corrosion and narrow pulse widths are both known to increase the vulnerability of LWR fuel to pellet cladding mechanical interaction (PCMI). The HERA proposal is designed to: (1) quantify the impact of pulse width on fuel performance, offering new insight into the applicability of existing data; (2) generate new data on high burnup fuel under pulse conditions prototypic of LWRs; (3) quantify the additional margin provided by modern cladding alloys to PCMI failure limits; and (4) offer improved data for modelers using specially designed tests that eliminate key uncertainties in high-burnup fuel tests.

(HERA)↗
DOE OSTI2016

A study on the assessment criteria of DHC in PWR spent fuel

One of the major degradation mechanisms of PWR spent fuel cladding is delayed hydride cracking (DHC), which can activate within a relatively low temperature range. Its maximum crack growth rate is estimated to be within the range of 10{sup -7} m/s, at which the cladding can rupture in 1 to 2 years during dry storage. For this reason, it is essential to prevent DHC crack growth itself, but most of the existing DHC studies have been focused on the crack growth rate. In this regard, the values and characteristics of the threshold stress intensity factor (K{sub IH}) for PWR Zircaloy-4 cladding were described. Based on the limited existing test results and analysis, the experimentally measured K{sub IH} were compared with the predicted K{sub IH}. The comparison shows that the predicted values are much lower than the measured ones and both theoretical estimation and measured values of K{sub IH} increase as the temperature increases. The possibilities of degradation by DHC were also discussed during long-term dry storage. For a more reliable analysis, additional information on a pre-existing crack of spent fuel is needed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Summary Report of Summer Work: High Purity Single Crystal Growth & Microstructure of Ferritic-Martensitic Steels

Harnessing the power of the nuclear sciences for national security and to benefit others is one of Los Alamos National Laboratory’s missions. MST-8 focuses on manipulating and studying how the structure, processing, properties, and performance of materials interact at the atomic level under nuclear conditions. Within this group, single crystal scintillators contribute to the safety and reliability of weapons, provide global security safeguards, and build on scientific principles that carry over to medical fields for cancer detection. Improved cladding materials made of ferritic-martensitic alloys support the mission of DOE-NE’s Fuel Cycle Research and Development program to close the nuclear fuel cycle, aiming to solve nuclear waste management challenges and thereby increase the performance and safety of current and future reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Determination of the mechanical properties of blisters by nanoindentation

In light-water reactors, a zirconium oxide layer will be formed at the outer surface of the fuel cladding as a consequence of the chemical reaction which takes place between zirconium and the cooling water. This zirconium oxide layer is a good thermal insulator, but at some point, it may spall-out. As a consequence, a cold spot will be formed at the outer surface of the cladding. In this situation, the hydrogen will quickly migrate to this spot to form a highly-enriched hydride zone, known as blister. In this work, an experimental procedure has been devised to produce blisters with a similar shape to the ones reported in the literature for irradiated cladding with severe hydriding. Nano-indentation tests were performed on blisters prepared in laboratory. A numerical simulation by the finite element method was developed to fit the load versus penetration depth data from the aforementioned nano-indentation tests. By means of an inverse process which combines experimental data and numerical simulations, the constitutive equation of the blister was determined. The process has been used to calculate the plastic stress-strain curve of the blister under study at room temperature. The results show an almost perfect numerical fitting of the experimental data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Evaluation of effective-stress-function algorithm for nuclear fuel simulation

In a pressurized water reactor (PWR), the mechanical integrity of nuclear fuel is the most critical issue as it is an important barrier for fission products released into the environment. The integrity of zirconium cladding that surrounds uranium oxide can be threatened during off-normal operation owing to a pellet-cladding mechanical interaction (PCMI). To analyze the fuel and cladding behavior during off-operation, the fuel performance code should calculate an inelastic analysis in two - or three-dimensional calculations. In this paper, the effective stress function (ESF) algorithm based on a two-dimensional FE module has been implemented to simulate the inelastic behavior of the cladding with stability and accuracy. The ESF algorithm solves the governing equations of the inelastic constitutive behavior by calculating the zero of the appropriate effective-stress-function. To verify the accuracy of the ESF algorithm for an inelastic analysis, a code-to-code benchmark was performed using the commercial FE code, ANSYS 13.0. To demonstrate the stability and convergence of the implemented algorithm, the number of iterations in the ESF algorithm was compared with that in a sequential algorithm in the case of an inelastic problem. Consequently, the evaluation results demonstrate that the implemented ESF algorithm improves the efficiency of the computation without a loss of accuracy for an inelastic analysis. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Fabrication of Minor Actinide MOX Fuel for the AFC-2C and -2D Irradiations

Minor actinide bearing mixed oxide (MAMOX) fuels of nominal composition (U{sub 0.75}Pu{sub 0.20}Np{sub 0.02}Am{sub 0.03}) O{sub 2-x} were fabricated at Los Alamos National Laboratory (LANL) for the AFC-2C and AFC-2D irradiations that were conducted by the US-DOE under the Advanced Fuel Cycle program. The primary goal of this irradiation test series was to assess the irradiation performance behavior of MA-MOX fuel under prototypic fast reactor conditions. This paper describes the results of the MA-MOX fuel fabrication and pre-irradiation characterization effort and includes discussion of specific fresh fuel properties that affect irradiation performance. The as-fabricated MA-MOX compositions were (U{sub 0.77}Pu{sub 0.19}Np{sub 0.03}Am{sub 0.02})O{sub 2-x} with x=0.01 and 0.04. A comparison of fuel to clad interaction and cladding corrosion for fuels with O:M of 1.99 and 1.96 will be of particular interest during post-irradiation examination which is underway at the INL and discussed in a companion paper presented in this forum. The average fuel density was 87% of theoretical. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Advances in Westinghouse fuel rod design technology (PAD5)

The Westinghouse Performance Analysis and Design (PAD) code for light-water reactors has been recently updated to version PAD5 and submitted to NRC for review. PAD5 incorporates extensive updates in fuel performance and material property models, including fuel thermal conductivity, pellet fragment relocation, fission gas release, fission gas swelling, and cladding creep and growth. Advances in thermal models are mainly in pellet fragment relocation, radial power distribution within the pellet, and the changes in fuel thermal conductivity due to changes in temperature, burnup and fuel porosity. The PAD5 thermal fission gas release model is a phenomenological model based on the irradiation dependence of the gas release from mechanistic processes for the diffusion of the fission gas generated in the fuel grains to the grain boundaries, formation of fission gas bubbles on the grain boundaries and release of the fission gas when these grain boundary bubbles inter-link. The PAD5 cladding creep model is predominantly a combination of the following mechanisms: thermal creep, irradiation induced creep, and irradiation enhanced creep. The fuel performance database used for the development, calibration and testing of the PAD5 code covers a broad range of both commercial irradiation data and test rod data and requires continuously additional data focusing on the latest development of the PAD5 code. Test programs are also included in the database. These data are checked and stored in a hierarchical database structure that supports the automation of model development and calibration. New data, once incorporated into the database, can be quickly evaluated with respect to key fuel and cladding performance models to confirm expected trends or to evaluate potential model improvements. This automated calibration system has been integrated with state-of-the-art optimization packages in DAKOTA Analysis Kit to provide for advanced statistical assessments and model coefficient optimization. Calibration results are automatically tabulated and plotted to provide easy assessment of trends or new behaviors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Effect of Heat Treatment on Microstructure and Mechanical Property of 316L Stainless Steel Produced by Laser Powder Bed Fusion

The advanced non-light water reactor designs (Gen IV reactors), including molten salt/ very high temperature/ sodium-cooled and lead-cooled fast reactors, typically operate at higher temperatures and more extreme radiation conditions than light water reactors. An intrinsic part of the deployment and progress of Gen IV reactor designs is selecting the most suitable structural material for a specific application. Additive manufacturing (AM), a fairly new process of making physical, three-dimensional objects from a computer design file, is going to completely change the way of design, build and certify nuclear systems. It offers a range of opportunities to produce complex geometries from existing materials, offers new routes for processing of previously difficult to process materials, allows for design of new high-performance materials, and finally facilitates hybridization of dissimilar materials. This emerging technology has successfully produced cars, wind turbine blade molds and even live cells. It could also open up big opportunities for the nuclear industry to quickly deploy technologies at a fraction of the cost. So far, AM techniques have been preliminarily applied in the field of nuclear reactors, including the classical parts such as the pressure vessel of a small reactor with 508-III steel, the bottom nozzle of a fuel assembly with 304L steel, the fuel cladding with zirconium alloy and the integrated impeller of a pump and the multi-channel valve body with 316L steel [6,7]. The AM applications for operating nuclear reactors started in auxiliary plant components and have slowly migrated to metallic reactors and core components, but many of these are not safety critical components. Although many parts used for nuclear reactors have been fabricated by AM techniques, practical applications in engineering are still a long way off due to the uncertainty factors focused on the processing, material properties, analysis methods and application standards, which feeds the safety and life-cycle of the nuclear reactor. Due to rapid, repeated heating and cooling during production, a high dislocation density was present in the AM material. This microstructure feature is unstable at elevated temperature while high temperature is one of the typical operation environments for nuclear reactors. Thus, it is important to understand the thermal effect on the microstructure of AM material. The objectives of this study are to investigate the effect of heat treatment on the microstructure and mechanical properties of 316L stainless steel produced by laser powder bed fusion additive manufacturing, and to determine an appropriate heat treatment practice that will be applied to the lightweight AM lattice-structured material with the same chemistry. The heat treatment study consisted of annealing the samples at a temperature range of 800 to 1200 oC with a 50 oC increment for different times (1-24 hours), followed by vacuum or air cooling. Microstructural characterization was carried out by Scanning Electron Microscope (SEM). Grain size and crystallographic orientation were investigated by Electron Backscatter Diffraction (EBSD). Vickers hardness tests with a 0.5 kg load were employed to determine the hardness of samples after different heat treatments. After heat treatment, the random crystallographic orientation was preserved, and the volume fraction of high-angle grain boundaries (grain boundary misorientation =15 oC) remained the same. The dislocation density decreased with annealing temperature due to recovery. The fine subgrain structures in the as-printed specimen were quite stable up to 1200 oC. Minimal recrystallization was observed up to 1200 oC. Recrystallization initiated only after 8.5 hours at 1200 oC. The SEM images did not show obvious dependence of microstructure on cooling rate. The hardness of the specimens decreased with increasing annealing temperature as a result of the decrease in dislocation density. It is interesting to note that the AM material showed very similar hardness to the wrought material when annealing at similar temperature, although the microstructures are very different. Annealing at 1050 oC for 1 hour followed by air cooling was selected as the heat treatment procedure for the lattice designed lightweight AM 316L material.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Influence of pre-hydriding and quench on embrittlement of E110 alloy under LOCA conditions

The aims of present work were to investigate the corrosion behavior and residual ductility of pre-hydrided E110 alloy under LOCA conditions including the influence of cooling rate. The fragments of unirradiated and pre-hydrided cladding (9,5*8,33 mm with a 10 mm length) were used as samples. The claddings were made from an E110 sponge based alloy (finish polishing). The hydrogen concentration varied from 5 (as received) to 800 wppm. As a result of pre-hydriding δ- hydrides of zirconium have been formed into cladding material. Corrosion tests were carried out at temperatures 1100 and 1200 C. degrees in a steam (80 mg/s) and argon (25 ml/s) mixture flow. The samples were cooled by 2 regimes: the first one is cooling in the cold zone of test facility with rate about 20 C. degrees/s, the second is cooling in the cold zone of test facility with rate about 20 C. degrees/s from test temperature to 800 C. degrees and quench by dropping in to the water. The change of a sample mass was determined as a difference between initial sample mass and its mass after oxidation. The mechanical tests of oxidized claddings in a diametrical direction were carried out with the use of a universal tension testing machine at a steady deformation rate of 1 mm/min at 135 C. degrees. The investigations have established the following. The initial concentration of hydrogen has no effects on the oxidation kinetics, while the oxidation kinetics are parabolic and the breakaway oxidation is not observed. The oxide films on cladding surfaces are black and shining. There are no cracks, visual spots and peelings. The residual ductility of oxidized samples decreases with the increase of hydrogen concentration. The residual ductility of slow cooled claddings is higher than residual ductility of quenched claddings. The limiting value of ECR (Equivalent Cladding Reacted) corresponding to ductile-brittle transition as function of initial hydrogen content was evaluated. The value of residual ductility 2% was used as a ductile-brittle threshold. The limiting value of ECR for as-received E110 alloy is 18% and it decreases slowly to 15% with hydrogen content rising to 350 wppm. At hydrogen content of about 350 - 400 wppm there is a rapid change of limiting value of ECR from 15% to 8 %. At hydrogen content from 400 to 800 wppm the limiting value of ECR decreases from 8% to 3%. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Modeling used fuel response to normal conditions of transportation package drops to access geometric sensitivities

The US Department of Energy Used Fuel Disposition program is performing research to determine the loading conditions that used nuclear fuel (UNF) experiences during normal conditions of transportation (NCT). This loading environment includes mechanical shock and vibration as the UNF is transported by rail or truck, and it also includes potential package free drop conditions, which are the topic of this study. NCT drop test requirements for UNF package designs are defined in 10 CFR 71.71. The free drop test height is 30 cm for the UNF packages considered in this study. Explicit dynamic finite element models of generic rail and truck UNF packages containing a highly resolved 17x17 PWR fuel assembly are used to estimate the potential range of loading experienced by UNF during NCT horizontal package drops onto an unyielding surface. There are many variables and uncertainties that can affect the loading of UNF during this type of free drop scenario. This study focuses on geometric variables, such as gap conditions and small adjustments to the angle of impact, in a number of analysis cases. The LS-DYNA finite element analyses calculate detailed results for all the components of the fuel assembly. The results show that the peak localized axial strain in the fuel cladding is estimated to be 0.006421 (nominal) with an upper bound of 0.014511, based on the peak instantaneous bending moment of 63.65 N-m and other uncertainty factors. For the assumed low burnup condition of the UNF, the cladding yield strain is 0.011027 and the failure strain is 0.015574, which means that the calculated upper bound strain approaches the cladding failure limit without exceeding it. The peak shear force on the cladding was calculated to be 7879 N. This could be significant to cladding integrity if applied to the UNF in a situation where the fuel pellets do not support the shear transfer. Further evaluation is needed to determine if this is a credible threat to cladding integrity for these 30 cm free drop scenarios.

42 ENGINEERING↗