Engineering topics

Kamerman, David W.

Publications and source records attributed to Kamerman, David W..

9 records

DOE OSTI2024

Investigation of Degradation Mechanisms of Cr-coated Zirconium Alloy Cladding in Reactivity-Initiated Accident

Reactivity-initiated accident testing has been performed on two baseline fuel rodlets which were fabricated from Zr-4 cladding and utilized fresh UO 2 fuel pellets. These experiments are part of a larger series of transient irradiations to explore Cr-coated zirconium alloy cladding under Reactivity-initiated accident conditions to evaluate ballooning and burst and sever oxidation characteristic of late-stage failure. The first two irradiations have been completed and post-transient examinations have been performed. In one rod, ballooning and burst occurred, along with oxidation. The second rod leaked and as a result did not balloon as much during the transient. This second rod also experienced increased oxidation compared to the first due to higher temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Enabling Digital Image Correlation for Deployment in the Hot Fuels Examination Facility

Digital image correlation (DIC) is a non-contact image-based technique for displacement and strain measurement. It is also a full-field method that can provide excellent spatial resolution. DIC has been widely used for measurements in many styles of mechanical tests and is desirable in part because of this flexibility to measure nearly any material across a range of time and length scales. Its ability to measure localized strain for tests and in environments not accessible to other strain measurement techniques makes it an ideal option for several mechanical test cases that exist in hot cell facilities such as the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). However, DIC has not been demonstrated at HFEF, and several barriers to implementation in the hot cell environment exist. This report outlines the possible paths for deploying DIC at HFEF, weighing the technical advantages and risks with each path. It also discusses the steps which must be taken to add this capability to the suite of post-irradiation examination (PIE) techniques available at HFEF. Finally, it concludes with recommendations for the best method to perform DIC at the facility.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Enhanced Stress Relaxation Behavior via Basal ⟨a⟩ Dislocation Activity in Zircaloy-4 Cladding

Herein this work evaluates the stress relaxation behavior of textured Zircaloy-4 cladding to understand how mechanical anisotropy influences pellet-cladding interactions. Uniaxial and biaxial stress relaxation tests are performed using full-tube axial tension and internal pressurization, respectively, aiming to achieve 0.25%, 1%, and 2% equivalent strains in the cladding samples at a temperature of 300°C. Internal pressure relaxation test results display enhanced stress relaxation compared to axial testing results, particularly for samples loaded beyond yield. Results of electron backscatter diffraction indicate increased deformation microstructure during loading and increased strain homogenization and recovery during relaxation for samples loaded via internal pressurization. Analysis indicates that the increased production and activity of basal ⟨a⟩ dislocations play a significant role in the enhanced relaxation measured in samples subjected to internal pressurization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Boiling-Water Reactor Testing Capability in the Advanced Test Reactor

I-Loop is an irradiation facility that is currently being installed at the Advanced Test Reactor. It is a two-loop test facility capable of performing Light Water Reactor (LWR) irradiations in prototypic coolant conditions. The two loops are being installed to be capable of both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) pressure, temperature, and chemistry environments. Each loop is nominally dedicated as a BWR or PWR for simplicity of operations. In-reactor water loop testing that an I-Loop provides is key to the deployment of new accident tolerant fuel technologies and other advanced LWR fuel concepts. Currently, pressurized water loops are the only testing facilities available to test BWR fuel concepts. Their test environments are non-prototypic at higher pressure/temperature and at single-phase fluid flow conditions. This void in the LWR test bed capabilities is one that the I-Loop is uniquely situated to provide. This report discusses the mechanical design, thermal hydraulic calculations, and neutronic calculations of a proposed standard experiment of accident tolerant BWR fuel concepts. Mechanical design examines the geometry and features of the main components. Thermal hydraulic calculations examine the modeling and results of the two-phase flow options available. Lastly, neutronic calculations examine the Monte Carlo analysis of enrichment, heat rates, and flux spectrum.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Evaluation of the ring tension test (RTT) for robust determination of material strengths

The ring tension test (RTT) is an experimental method for determining mechanical behavior in a material’s circumferential or hoop direction. It is a crucial test for testing anisotropic materials with tube geometry, such as nuclear fuel cladding or irradiated pipes. Several RTT configurations exist, each with their own advantages and disadvantages. However, this test is significantly more complex than traditional tensile testing and can be especially sensitive to small differences and inconsistencies in the test setup and geometry, ultimately affecting the derived mechanical properties. Previous research has focused on method development, and little work has been done on understanding the subtle differences between an ideal test and experiments, specifically when the tests are performed on highly irradiated materials in hot cells. In this work, a finite element-based investigation of the RTT is conducted to fill this gap. The two most used test configurations are investigated, comparing their ability to determine accurate material strengths through plastic deformation. Several non-ideal conditions and uncontrollable effects which are likely to occur during experimental testing such as machining tolerances, variations of specimen geometry from nominal dimensions, rotation of specimens and fixturing, and other test setup discrepancies are studied. The sensitivity of measured strengths to these conditions is presented. A mechanics-based approach to describing and correcting raw data to determine actual strengths is also included for one of the configurations, resulting in a robust correction method with highly accurate material strength measurements. Finally, based on these analyses, the hemicylindrical mandrel configuration is recommended with a gauge region oriented at a 45° angle.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

The deformation and burst behavior of Zircaloy-4 cladding tubes with hydride rim features subject to internal pressure loads

This study is motivated by the desire to conduct a separate effects test replicating specific aspects of an in-pile pulse reactor test on high exposure light water reactor (LWR) fuel rods. The principal purpose of the tests was to determine whether temperature and or hydrogen concentration thresholds exist in assessing the vulnerability of a stress relieved zirconium alloy cladding tubes to low strain ruptures. A secondary objective was to determine whether observations of the cladding stress during deformation could aid in the understanding of the crack formation and propagation in cladding tubes with hydride rim features. Here, the study utilized artificially hydrided Zircaloy-4 cladding tubes and an isothermal, internal pressure test that was controlled with an in-situ strain measurement. The test results show a clear transition in failure vulnerability between room temperature and 150 °C. Below 150 °C, cladding tubes rupture at very low uniform plastic strain—even at relatively modest cladding hydrogen levels. At and above 150 °C, cladding tubes only rupture at low uniform plastic strain if the cladding hydrogen content is greater than 500 ppm and usually experience at least some modest amount of plastic deformation prior to rupture. Post-test metallographs from non-ruptured cladding tubes show numerous primary cracks forming in the hydride rim and arresting at the metal/hydride interface. This observation combined with the decrease in bulk yielding of hydrided cladding tubes when compared to fresh non-hydrided tubes leads to the conclusion that the fracture stress of the hydride rims is related to and bounded by the yield stress of the zirconium metal. Some evidence is presented for a change in crack propagation mechanism from one of void formation on hydride plates and linkage at temperatures below 150 °C to more classical ductile shearing beneath the primary crack tip at temperatures above 150 °C. This apparent change in crack propagation mechanism may help explain the observed temperature and hydrogen concentration thresholds.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Formation and characterization of hydride rim structures in Zircaloy-4 nuclear fuel cladding tubes

Zirconium alloy tubes used as nuclear fuel cladding are subject to oxidation and subsequent hydrogen pickup during their long service in commercial light water reactors. The hydrogen picked up in the cladding can precipitate as a brittle hydride rim feature on the cladding outer surface. Here, to better understand the effect of hydride rims on the fracture behavior of stress-relieved zirconium alloy cladding tubes, a procedure to produce these rim-like structures has been developed and is described herein. Extensive characterization of the ‘as hydrided’ tubes is performed. The hydrogen charging apparatus consists of a tube furnace with a quartz chamber that is connected to a vacuum pump as well as a bottle of pure hydrogen gas. The charging station uses a static charge of hydrogen as opposed to a flowing gas. The chief advantage of this approach is the ability to monitor the pressure drop in the hydriding chamber and correlate this pressure drop to a known rate of hydrogen pickup in the cladding tube. It was found that the hydrogen partial pressure, metal temperature, and surface treatment all clearly played a role in whether a hydride rim was formed. Extensive characterization of the hydride rims shows they consist of needle like platelets of δ phase hydrides (ZrH1.66) oriented in the circumferential direction with a radial spacing of several microns in a sandwich-like structure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Resumption of water capsule reactivity-initiated accident testing at TREAT

A series of integral reactivity-initiated accident (RIA) commissioning experiments were completed in a new static water capsule in the Transient Reactor Test Facility (TREAT), marking the first such tests in the U.S. in more than 40 years. The test campaign included a verification test followed by five tests in the Static Environment Rodlet Transient Test Apparatus (SERTTA) capsule. The capsule initial conditions varied from room temperature and pressure up to 200°C and 2.5 MPa, with energy depositions varying between ~500–1100 J/gUO 2 . The series of tests allowed for a number of instrumentation qualifications and demonstrations, including cladding thermometry, rodlet plenum pressure, cladding elongation, and an electro-impedance boiling detector. Post transient examinations such as gamma emission spectroscopy, profilometry, and microscopy were performed to document the end state of the fuel rods. Furthermore, the results from the experiments show that the fuel rodlets behaved very similar to historical tests under similar energy depositions. This paper documents the design of the capsule and highlights some results from the commissioning tests and post-transient examination.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Accident Tolerant Fuel Test (ATF-2c) Irradiation Test Report

A new fully prototypic testing platform for testing new accident tolerant fuel (ATF) designs for light water reactors has been established in the center flux trap of Idaho National Laboratory’s advanced test reactor (ATR). The irradiation experiment named ATF-2C has completed its first cycle of prototypic steady state irradiation. Irradiation conditions have been maintained via the loop 2a pressurized water coolant loop and have been monitored through in-situ instrumentation during each cycle. Power and fast neutron flux histories of each of the 24 test pins has been calculated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗