Titanium and Aluminum Rod-Replacement Experiments in Fully-Reflected 6.90% Enriched UO2 Fuel Rod Lattices.
Abstract not provided.
Harms, Gary A.↗
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Titanium and Aluminum Rod-Replacement Experiments in Fully-Reflected 6.90% Enriched UO2 Fuel Rod Lattices.
Abstract not provided.
Harms, Gary A.↗
Demo Sister Rod Test Plan and ORNL Sister Rod Baseline Testing Progress.
Abstract not provided.
Saltzstein, Sylvia J.↗
The Study of Lightbridge Metallic Fuel Rods in MOOSE for Meshing Capabilities and Studying Lightbridge Metallic Fuel Rods with MOOSE
Supporting INL’s mission to transform our nation’s energy future, notably through innovative nuclear energy solutions, this project focuses on modeling accidental tolerant fuel. In collaboration with MIT, we modeled with MOOSE-based tools the Lightbridge metallic fuel design. Lightbridge fuel is envisioned to enable higher power output and larger safety margins for both current and advanced nuclear reactors. We present here the advanced meshing capabilities developed and preliminary physics results.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Measuring, Cutting and Sorting Irradiated Control Rods to Limit the Volume of Intermediate Level Long-lived Waste - 17001
During operation, the PWR control rods are either completely extracted (scram or safety rods) or partially inserted in the operating zone of the core (fine control or operating rods). At the nominal rating conditions or close to them, the operating zone is limited to some steps of insertion in the order of a few tens of cm. In fact, only the lower part of the control rods are subjected to a significant neutron flux with an impact on the neutron activation and so on the waste classification (Low Level Short-lived Waste versus Intermediate Level Long-lived Waste). Our calculative as well as experimental experience confirms the significant presence of radionuclides in the control rods, but only in the lower parts. Measuring, cutting and finally sorting control rod scraps can reduce the volume of Intermediate Level Long-lived Waste planned for our storage deposit facility. This paper presents the EDF/DP2D methodologies followed to determine the sort criteria between the Low Level Short-lived Waste versus the Intermediate Level Long-lived Waste zones for the Silver-Indium-Cadmium control rods. The irradiated control rods are stored in cases without their head. Each case can contain several control rods, and the cases are stored under water in the fuel assembly pool. To calculate the radioactive inventories, we have developed a calculation scheme which is composed of a 3-dimensional mapping of the neutron flux, a calculation of the evolution of the isotopic compositions under neutron flux and finally a comparison between calculated and measured values. The main hypotheses of the control rods calculation take into account: - 3-dimensional geometry and 3-dimensional neutron sources distribution, - the isotopic compositions including impurities, - the history of irradiation resulting mainly from the daily power production, the positions of the control rods in the loading patterns and the positions in the axial operating zone. In order to validate the calculations, measurements were made 20 meters under water in the fuel assembly pool. The experimental results show the presence of many gamma emitter radionuclides. Only Ag-108m was retained for validating the calculations. This is due to the fact that silver is considered as the 'standard' chemical element for our control rods and silver produces Ag-108m which is a long-lived gamma emitter radionuclide. The analysis of the calculated results and the measured values allow us to build a simplified physics model for the simulation of the control rods activation. Finally three sort criteria can be proposed: - Without measurement devices, the sorting is based on an empirical sort criterion. Cuttings must then be made on the 2 back-ends of the control rod pins. This option is due to the difficulty in knowing the orientation of the pins in the storage cases. The cutting lines are placed at approximately 1 m from the 2 back-ends of the pins. This approach does not allow a great optimization of the sorting, less than 50% of the length of the control rods can be saved from Intermediate Level Long-lived Waste storage. - With a simple dose rate measurement device, the knowledge of the orientation of the pins in the storage cases allows us to limit cuttings to the lower part of the control rods. Depending on the 'top to tail' cases, between approximately 50% and 75% of the length of the control rods can be saved from Intermediate Level Long-lived Waste storage. - With an accurate measurement device, the sorting can be based on a specific activity of Ag-108m. Over 4400 Bq/g of Ag-108m (corresponding to the Intermediate Level Long-lived Waste limit), the control rod pieces can be considered as a long-lived waste. The measurements must be made over the 2 back-ends of control rod pins to examine the cases of 'top to tail' storage. This approach allows a great optimization of the sorting. Depending on the history of irradiation, more than 75% of the length of the control rods can be saved from Intermediate Level Long-lived Waste storage. (author)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Development of a new control rod drive mechanism design for the ISU AGN-201M reactor
The Aerojet General Nucleonics (AGN) model 201-Modified, known as the AGN-201M reactor, plays an essential role in the educational and research activities at Idaho State University (ISU). The ISU AGN-201M's original Control Rod Drive Mechanism (CRDM) has been in operation for more than fifty years with no large-scale redesigns. The CRDM is required to eject the fuel rods within one second during a SCRAM event (also known as a 'reactor trip') and adjust the control rods' insertion speed, and keeps the rod insertion sequence correct. The existing control rod drive mechanisms meet these criteria but experience a few concerns due to aging. Concerns include complex maintenance and costly repairs for old electromechanical components, rod position feedback errors, and the impediment of the plate during a SCRAM due to binding of the lead screws of the existing mechanism. During a binding event, the drive mechanism becomes locked, preventing the control rod's magnetic plate from moving in or out under the reactor's normal and emergency operating conditions. Although, the binding has no effect on the ability of the rod to exit the core during the SCRAM. To counteract the concerns and issues with the current CRDM, a new design has been proposed using newer components and a simplified design. The new design utilizes more advanced electric and mechanical components that are commercially available. The new CRDM system is divided into four main aspects: (1) control rod movement design (motor, lead screw, guide rods), (2) control rod ejection (springs, electromagnet), (3) control rod position and feedback (position transducer, microswitches), and (4) material selection and structural analysis. The new design aims to reduce the overall complexity and probability of failure to improve the reactor's overall reliability. With proper material selection and improved structural design, the new drives are lighter with little to no change in structural integrity. The new control rod drive mechanism eliminates binding scenarios by using a single lead screw and implementing additional guide rods. An advanced linear position sensor and microswitches replace the existing and aging synchro system for accurate rod position feedback resulting in better reactivity control. The new design meets the reactor's operational limits by having an average reactivity insertion of 0.065% Δk/k per second, which corresponds to a total control rod insertion time of 19.23 s, while the control rod's ejection time remains less than one second during a SCRAM event. The new design ensures the reactor's long-term viability for educational and research activities by increasing the reliability and safety of operation for years to come.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Influence of Axial Power Distributions on Rod-to-wall Gap
In a scaled-down (3 x 3, 4 x 4, 5 x 5 or 6 x 6) rod bundle thermal hydraulic experiment, channel walls were often applied to surround the experimental rod bundle and provide a hydraulic flow boundary for the experiments. However, in the actual PWR (Pressurized Water Reactor), there are no walls surrounding the rod bundle fuel assembly. These non-representative walls can lead to non-prototypical conditions in the rod bundle experiments. There are several researches about the wall effects in rod bundle experiments. Based on hydraulic consideration, Zang developed a wall effect factor by subchannel analysis method to investigate the wall effects of different scale rod bundles. He did not pay attention to the wall effects in thermal aspects. Li studied the wall effects of different scale rod bundles of tight arrangement spiral-fin fuel assembly by CFD. The existing researches mostly focus on the wall effects of different scale rod bundles. There are very few, if any at all, studies focus on how to avoid or minimize wall effects through proper experimental design by adjusting the rod-to-wall gap (the gap between the channel wall and outer rod wall) for rod bundles of various scales. In addition, the axial power distribution is different in different periods of the life cycle in the PWR core. For example, it usually is outlet peak power distribution in the axial direction at the beginning of life span. However, it may turn to the inlet peak power distribution when the control rods are pulled out from the high burn up of fuel in the reactor core. Both flow dynamics and heat transfer mechanism change depending on various axial power distributions. Subsequently, it is important to study the impacts of the wall effects on flow and heat transfer with different axial power distributions in rod bundle experiments. That is, in the evaluation of rod-to-wall gap in rod bundle experiments, the influence of axial power distributions should be examined. In this paper, the rod-to-wall gap was calculated in axially uniform and non-uniform power, including inlet peak, cosine, and exit peak heat flux distributions. The rod-to-wall gap calculation results were then compared with respect to two different criterion: the average fluid temperature and the maximum wall temperature criterion, to study the influence of axial heat flux profile on radial thermal distribution.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Sister Rod Destructive Test Results (FY19)
This report documents work performed under the Spent Fuel and Waste Disposition, Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 2 Milestone M2SF-19ORO010201026 within work package SF-19OR01020102 and is an update to the work reported in M2SF-19OR010201028 and M3SF-19OR010201027. The document incorporates comments received with respect to the previous reports and incorporates additional test results. As a part of the US Department of Energy (DOE) Office of Nuclear Energy (NE) High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 gigawatt days per metric ton uranium) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called the “sister rods,” are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zircaloy-4 (LT Zirc-4), ZIRLO ® , and M5 ® . The DEs are being conducted to obtain a baseline of the HBU rod’s condition prior to dry storage and are focused on understanding overall SNF rod strength and durability. Both composite fuel and empty cladding will be tested to derive material properties. While the data generated can be used for multiple purposes, a primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and consolidation activities. This report documents the status of the ORNL Phase I DEs of 8 sister rods and outlines the DE tasks performed and the data collected to date, as guided by the sister rod test plans. The DEs are performed using a phased approach, and the Phase 1 DEs being performed at ORNL include: • full-length rod heat treatments (FHT) of 3 selected sister rods to examine the effects of temperatures reached during dry storage preparation, • rod internal pressure and void volume measurements of the 3 FHT rods and 3 corresponding baseline rods, as well as 2 additional rods selected for depressurization/gas transmission tests and fatigue lifetime tests, • fission gas sampling and analysis, • depressurization and gas transmission tests, • rough segmentation of the selected rods for mechanical tests and rod characterization, • fuel sampling and burnup analysis, • metallography, • cladding total hydrogen measurements, • mechanical testing, including fatigue lifetime (CIRFT), four-point bend, axial tension, microhardness, ring compression, and burst tests. Table S-1 provides a summary of the DE status. The mechanical testing will be performed using fueled segments and is expected to complement previous and current mechanical test results using defueled cladding segments.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Rod internal pressure quantification and distribution analysis using Frapcon
This report documents work performed supporting the Department of Energy (DOE) Office of Nuclear Energy (NE) Fuel Cycle Technologies Used Fuel Disposition Campaign (UFDC) under work breakdown structure element 1.02.08.10, ST Analysis. In particular, this report fulfills the M4 milestone M4FT- 15OR0810036, Quantify effects of power uncertainty on fuel assembly characteristics, within work package FT-15OR081003 ST Analysis-ORNL. This research was also supported by the Consortium for Advanced Simulation of Light Water Reactors (http://www.casl.gov), an Energy Innovation Hub (http://www.energy.gov/hubs) for Modeling and Simulation of Nuclear Reactors under U.S. Department of Energy Contract No. DE-AC05-00OR22725. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle assembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each evaluated WBN1 fuel rod. An alternate model for the amount of helium released from the zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layer is derived and applied to FRAPCON output data to quantify the RIP and CHS for these types of fuel rods. SCALE/Polaris is used to quantify fuel rodspecific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rod without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel pellets. The primary contributor to elevated RIP predictions at burnups less than and greater than 30 GWd/MTU is determined to be the total fuel rod void volume and the amount of released fission gas in the fuel rod, respectively. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP and CHS predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceed a specified RIP or CHS limit. Results are separated into IFBA and standard rods so that the two groups may be analyzed individually. FRAPCON results are provided in sufficient detail to enable the recalculation of the RIP while considering any desired plenum gas temperature, total void volume, or total amount of gas present in the void volume. A method to predict the CHS from a determined or assumed RIP is also proposed, which is based on the approximately linear relationship between the CHS and the RIP. Finally, improvements to the computational methodology of FRAPCON are proposed.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Statistical and sensitivity analysis of failing rods in EPR LBLOCA
A statistical fuel failure analysis procedure has been developed, and is now applied in full scale to a large break loss-of-coolant accident (LBLOCA) scenario in EPR-type nuclear power plant. The goal is to statistically determine the number of failing fuel rods. It is also important to bring out the underlying causes for rod failures, and therefore a sensitivity analysis procedure for the preceding complex calculation chain is outlined and adopted. In order to produce a statistically reliable estimation on the number of failing rods, a large number of simulations are required. In the analysis, single rods are simulated with the coupled fuel performance - thermal hydraulics code FRAPTRAN-GENFLO. The statistically varied factors are divided into global and local by their range of influence. Transient thermal hydraulic and power history boundary conditions and model parameters are global and affect all the rods. Fuel rod manufacturing parameters and other rod position related factors are local and are related to a certain rod. The number of global scenarios is 59 as dictated by the Wilks' formula, and 1000 randomly sampled rods are simulated in each scenario. The thermal hydraulic and transient power boundary conditions for the fuel performance code are calculated with a statistical version of the system code APROS. The steady-state irradiation histories are simulated with FRAPCON, and the steady-state power histories used as boundary conditions for FRAPCON are obtained from SIMULATE calculations. Thus, a multistage calculation chain is required to consolidate the procedure. As an outcome, in the worst global scenario, 1.4% of the simulated rods failed. It can be concluded that the Finnish safety regulations, i.e. max. 10% of the rods allowed to fail, are met. In order to find out which input parameters have significance to the outcome of the analysis, a sensitivity analysis is done. At the moment, only the worst global scenario is considered but later on it is possible to include also the sensitivities of the global factors. Due to complexity of the existing data, first the relevant input parameters for the sensitivity analysis have to be specified. Data visualization with a cobweb graph is used for the screening. Then, selected sensitivity measures are calculated between the chosen input and output parameters. The sensitivity indices calculated are the Borgonovo's delta measure, the first order Sobol sensitivity index, and squared Pearson correlation coefficients. The first mentioned is a novelty in this context. As an outcome, the most relevant parameters with respect to the cladding integrity were determined to be the decay heat power during the transient, the thermal hydraulic conditions in the rod's location in the reactor, and the steady-state irradiation history of the rod as represented in this analysis by the rod burnup. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Modeling and simulation performance of sucker rod beam pump
Artificial lift is a mechanism to lift hydrocarbon, generally petroleum, from a well to surface. This is used in the case that the natural pressure from the reservoir has significantly decreased. Sucker rod beam pumping is a method of artificial lift. Sucker rod beam pump is modeled in this research as a function of geometry of the surface part, the size of sucker rod string, and fluid properties. Besides its length, sucker rod string also classified into tapered and un-tapered. At the beginning of this research, for easy modeling, the sucker rod string was assumed as un-tapered. The assumption proved non-realistic to use. Therefore, the tapered sucker rod string modeling needs building. The numerical solution of this sucker rod beam pump model is computed using finite difference method. The numerical result shows that the peak of polished rod load for sucker rod beam pump unit C-456-D-256-120, for non-tapered sucker rod string is 38504.2 lb, while for tapered rod string is 25723.3 lb. For that reason, to avoid the sucker rod string breaks due to the overload, the use of tapered sucker rod beam string is suggested in this research.
71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
SU-E-I-29: Generalized Relative Object Detectability (G-ROD): A Metric for Comparing X-Ray Imaging Detector Systems
Purpose: To determine the relative task-specific performance of two x-ray imaging detectors including the effect of other components of the imaging chain. Methods: The relative object detectability (ROD) was defined as the ratio of the integral of the detective-quantum efficiency (DQE) of a detector weighted by the square of the Fourier Transform (or frequency spectrum) of the object to the same integral for a second detector. The generalized ROD (G-ROD) is calculated by replacing DQE by the generalized DQE (GDQE) which includes geometric unsharpness and scatter to evaluate the total imaging chain. The G-ROD was calculated for the micro-angiographic fluoroscope (MAF) relative to a flat-panel detector (FPD) for low (1.05) and high (1.2) magnifications with two focal-spot sizes, small (0.3 mm) and medium (0.5 mm), and for the same value of scatter fraction. Solid spheres from 50 to 600 microns were simulated as the objects for this evaluation. The G-ROD results were compared with those of the ROD which considers the detectors alone. Results: For 50-micron spheres, the ROD is 7.93 and, for the small [medium] focus, the G-ROD is 6.74 [5.22] at low magnification and 3.03 [1.66] at high magnification, indicating superior performance of the MAF particularly at low magnification. Both ROD and G-ROD decrease monotonically with size and remain around one above 400-microns. The G-ROD decreases more dramatically with low compared to high magnification for both focal spots. Conclusion: In all cases, the MAF performs better than the FPD and the decreasing trend of the G-ROD indicates that the performance of both detectors becomes similar as the object size increases and becomes nearly equal above 400-micron diameter sphere size. Finally, increased magnification and focal-spot size decrease the G-ROD, indicting that they degrade the performance of the MAF more than that of the FPD for a small-object viewing task. Supported by NIH Grant: 2R01EB002873 and an equipment grant from Toshiba Medical Systems Corporation.
60 APPLIED LIFE SCIENCES↗
Tantalum Experiments in Fully-Reflected Water-Moderated Triangular-Pitched U(6.90)O 2 Fuel Rod Lattices (1.02 cm Pitch)
The United States (US) Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories (SNL). The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5 % 235 U. The 7uPCX assembly is a water moderated and — reflected array of aluminum-clad U(6.90 %)O 2 fuel rods. Other critical experiments performed in the 7uPCX assembly are documented in LEU-COMP-THERM-078, LEU-COMP-THERM-080, LEU-COMP-THERM 096, LEU- COMP-THERM-097, LEU-COMP-THERM-101, LEU-COMP-THERM-102, and LEU-COMP-THERM-111. Applied interests in neutron-absorbing elements at epithermal and intermediate neutron energy ranges at DOE’s Hanford Site and Savannah River Site were identified through DOE Office of Environmental Management (EM) workshops and needs surveys performed in conjunction with the National Nuclear Security Administration/Nuclear Criticality Safety Program. In response, a team lead by Oak Ridge National Laboratory (ORNL) performed a study evaluating the use of 7uPCX for testing the epithermal cross sections of materials at SNL. This marked the inception of the set of experiments described here. The purpose of these experiments was to measure the effects of tantalum in nearly-critical systems. The tantalum was introduced into the fuel arrays as experiment rods within a central test region. The central test region was designed to target the epithermal neutron energy range by providing a dry cylindrical cavity constructed of aluminum and lined with 0.04 in (0.1016 cm) thick cadmium sheet. The central test region has an outer diameter of 3.75 in (9.525 cm) length of 31 in (78.74 cm) and can hold 85 tantalum rods nominally 0.25 in (0.635 cm) outside diameter and 31.25 in (79.375 cm) long. The critical experiments were done using a set of triangular-pitched grid plates fabricated for these experiments. The grid plate set accommodated a fuel array of a total of 2016 fuel rod positions on a pitch of 0.4 in (1.016 cm) in a series of 21 hexagonal rings positioned around the central test region. The fuel used in these experiments was fabricated using unirradiated 6.90 % enriched UO 2 fuel pellets from fuel elements designed to be used in the internal nuclear superheater section of the Pathfinder boiling water reactor operated in South Dakota by the Northern States Power Company in the 1960s. The fuel elements were obtained from The Pennsylvania State University where they had been stored for many years. The fuel pellets in those fuel elements were removed from the original Incoloy cladding and reclad in 3003 aluminum tubes and end caps for use in the experiments reported here. The eight cases in this experiment series were performed in 2024 in the Sandia Critical Experiments (SCX) at the Sandia Pulsed Reactor Facility by varying the number of fuel rods in the critical assembly. The personnel involved in conducting these experiments included David Ames, Elijah Lutz, Beth Hanson, Patrick Ward, Jason Soares, and Augie Chapa from SNL, as well as Mathieu Dupont and B.J. Marshall from ORNL. Case 1 had no tantalum experiment rods. Case 2 had seven tantalum rods in the center of the central test region. Case 3 had 18 tantalum rods in the third hexagonal ring of the central test region. Case 4 had 19 tantalum rods in the center of the central test region. Case 5 had 30 tantalum rods in the outer most positions of the central test region. Case 6 had 37 tantalum rods in the center of the central test region. Case 7 had 61 tantalum rods in the center of the central test region. Case 8 had 85 tantalum rods filling all positions in the central test region. All eight critical experiments are judged to be acceptable as benchmark experiments.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Acoustic sensors for fission gas characterization: R and D skills devoted to innovative instrumentation in MTR, non-destructive devices in hot lab facilities and specific transducers for measurements of LWR rods in nuclear plants
First of all, we will present the main principle of the method. A piezoelectric transducer, driven by a pulse generator, generates the acoustic waves in a cavity that may be the fuel rod or a chamber connected to an instrumented rod. The composition determination consists in measuring the time of flight of the acoustic signal emitted. The pressure can be estimated by a calibration process, above the measurement of the amplitude of the signal. Two projects will then be detailed. The first project consists in the development of advanced instrumentation for in-pile experiments in Material Testing Reactor. It constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). As a first step of the development program, we performed in-pile tests on the most sensitive component, i.e., the piezoelectric transducer. For this purpose, the active part of this sensor has been qualified on gamma and neutron radiations and at high temperature. Various industrial piezo-ceramics were exposed to a high activity Cobalt source for few days. The cumulated dose was ranged from 50 kGy up to 2 MGy. Next, these devices were placed inside a Material Test Reactor to investigate their reliability towards neutron fluence. The final fluence after 150 days of irradiation was up to 1.6.10{sup 21}n/cm{sup 2} (for thermal neutron). Irreversible variations have been measured. Next, a specific sensor has been implemented on an instrumented fuel rod and tested in the frame of a REMORA 3 Irradiation test. It was the first experiment under high mixed, temperature neutron and gamma flux. A first irradiation phase took place in March 2010 in the OSIRIS reactor and in November 2010 for the second step of the irradiation. The instrumented fuel rod incorporating the ultrasonic gas composition sensor was finally irradiated during 2 weeks in nominal conditions. Neutronics calculation will be performed in order to calculate the thermal and fast neutron fluence and the gamma dose absorbed by acoustic sensor. A first evaluation gives a thermal fluence about 4,5.10{sup 19} n/cm{sup 2}, a fast neutrons fluence about 4,5.1018 n/cm{sup 2} and a total gamma dose up to 0,25 MGy The maximal temperature during the irradiation test was about 150 C. Although the ultrasonic sensor appears to be damaged, the optimization of the electrical attack parameters and the development of a new signal processing maintain the measurement feasibility up the end of the irradiation campaign. It was the first time that the composition of fission gas has been monitored all along an irradiation experiment in a MTR, giving access to the gas release kinetics. New researches involve thick film transducers produced by screen-printing process in order to propose piezoelectric structures for harsh temperature and irradiation measurements. The second project consists in the development of a non-destructive device that can be directly applied on a LWR fuel rod. The problem to be solved relates to the measurement of the fission gas pressure and composition in a fuel rod using a non-destructive method. Fuel rod internal pressure is one of the safety criteria applied in nuclear power analyses. This criterion must be verified in order to avoid any fuel-cladding gap reopening risk and therefore any local clad ballooning. Apart from the safety implications, this parameter is also a fuel behaviour indicator and reflects the overall fuel performance in operation, but also during shipping and long-term storage. Rod internal pressure is one criterion amongst others, like cladding corrosion, against which the acceptable fuel burn-up limit is set. A sensor has been achieved in 2007. A full-scale hot cell test of the internal gas pressure and composition measurement by an acoustic sensor was conducted successfully between 2008 and 2010 on 5 high burn-up MOX fuel rods and 2 very high burn-up UO{sub 2} fuel rods in LECA Facility at Cadarache Centre. An improvement of this sensor has been proposed, allowing us to divide by two the uncertainty on the pressure measurement. In the case of hot-cell measurements, viscous liquid can be used to couple the sensor with the rod. For gas content with a pressure exceeding 15 bars and a 10% Xe/Kr ratio, such coupling may reduce relative acoustic method accuracy by ±7% for pressure measurement result and ±0.25 % for the assessment of gas composition. These results make it possible to demonstrate the feasibility of the technique on LWR fuel rods. The transducer and the associated methodology are now operational for non-destructive measurements in hot lab facilities and allow characterising the fission gas without puncturing the fuel rods. Up to now, any other non-destructive method can be proposed. A next step will be the development of an industrial application in a fuel storage pool in order to perform a large number of measurements on a fuel assembly in nuclear plants.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NON-DESTRUCTIVE POST IRRADIATION EXAMINATION OF FIRST CYCLE ACCIDENT TOLERANT AND ADVANCED ZIRCONIUM ALLOY HIGH BURNUP FUEL RODS
Post irradiation examination (PIE) of accident tolerant fuel concepts and high burnup uranium dioxide advanced zirconium alloy clad fuel contribute to near term nuclear industry goals of enhancing light water reactor safety and economics. Westinghouse is developing chromium coated zirconium alloy cladding as an accident tolerant fuel option for current pressurized water reactors. Examination of high burnup fuel further augments the technical foundation for extending the peak rod average burnup limits beyond the current regulatory limit of 62 MWd/kgU. Seven fuel rods were received at the Oak Ridge National Laboratory Irradiated Fuel Examination Facility. This includes three chromium coated zirconium alloy fuel rods and four high burnup advanced zirconium alloy fuel rods. Six of the seven rods were fueled with standard uranium dioxide while one of the chromium coated rods contained Westinghouse’s ADOPT fuel. The overall hot-cell PIE plan and some PIE results have been presented previously. This paper focuses on the additional non-destructive examination and fission gas release. Highlights from visual examination of these rods are presented. The chromium coating showed little to no visual change following irradiation. Observations of the axial burnup trends in the fuel rods are derived from full rod axial gamma-ray scans with additional observations related to local fission product migration. The axial dimensional changes of the seven rods are evaluated. As expected, creep down had not yet occurred in the chromium coated rods, and the maximum diameter strain in the high burnup advanced zirconium alloy cladding was less than 0.5%. Finally, the steady state fission gas release for these rods was measured ranging from 4.1% to 17.6% which matched expectation from literature. These fuel rods now await further destructive examination.
Harp, Jason↗
Key results from examinations of seven high burnup pressurized water reactor spent nuclear fuel rods
At present, spent nuclear fuel (SNF) assemblies discharged from US commercial power plants are placed into dry storage following a short cooling time (<10 years) in the plant’s spent fuel pool. The process of packaging the spent fuel into dry-storage canisters includes a drying step to remove residual water from the canister. During the drying process, the fuel rod cladding may reach temperatures as high as 400°C. Oak Ridge National Laboratory (ORNL) is performing destructive examinations of high burnup (HBU) (>45 GWd/MTU) SNF rods to address knowledge and data gaps related to extended interim storage and eventual transportation for disposal. The rods examined include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. Three rods were subjected to a thermal transient to assess the effects of decay-heat-driven high temperatures expected during vacuum drying of the fuel as it is prepared for interim dry storage. The examinations focus on the composite fuel rod performance, as compared with the performance of defueled rod cladding, and establish the baseline mechanical properties of a fuel rod before interim dry storage. The key results of these examinations are presented, including the measured mechanical and fatigue properties, observations of cladding hydrogen pickup and hydride reorientation effects on rod performance, effects of the simulated drying temperatures on rod performance, and general conclusions of SNF performance in extended interim dry storage and transport. The rods were found to be strong and durable in the expected loading conditions, even considering the formation of radial hydrides associated with vacuum drying. The combined testing provides a broad body of data supporting extended interim storage and transportation performance of HBU spent fuel.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Key Results from Detailed Nondestructive Examinations of 25 Pressurized Water Reactor High Burnup Spent Nuclear Fuel Rods
The High Burnup Spent Fuel Data Project, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is focused on understanding the effects of long-term storage and transportation on high burnup (HBU) (>45 GW days per tonne uranium) light water reactor fuel. The project includes 32 HBU spent nuclear fuel (SNF) assemblies (the project assemblies) that are stored in a typical independent spent fuel storage installation (ISFSI) and 25 “sister rods”—9 SNF rods that were removed from the fuel assemblies prior to insertion to the ISFSI and 16 SNF rods removed from similar HBU assemblies. The sister rods provide a baseline of the condition of the HBU rods before loading, drying, and long-term dry storage. The project assemblies will be inspected after 10 years, and the physical state of the stored rods will be compared with the condition of the sister rods to identify any changes in physical properties during the dry storage period. Our report focuses on key results from the nondestructive postirradiation examinations of the sister rods and summarizes the results of detailed visual examinations, gamma scans, dimensional measurements, and eddy current liftoff measurements of the combined Chalk River unidentified deposits and oxide layer on the waterside surface of the rod. Additionally the data are used to calculate fuel rod and pellet stack growth rates, estimated remaining fuel rod plenum volumes, and the percentage change in fuel rod cladding diameter.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Experimental Study on Surrogate Nuclear Fuel Rods under Reversed Cyclic Bending
The mechanical behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending or bending fatigue must be understood to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup fuels (>45 GWd/MTU), which have the potential for increased structural damage. It has been demonstrated that the bending fatigue of SNF rods can be effectively studied using surrogate rods. In this investigation, surrogate rods made of stainless steel (SS) 304 cladding and aluminum oxide pellets were tested under load or moment control at a variety of amplitude levels at 5 Hz using the Cyclic Integrated Reversible-Bending Fatigue Tester developed at Oak Ridge National Laboratory. The behavior of the rods was further characterized using flexural rigidity and hysteresis data, and fractography was performed on the failed rods. The proposed surrogate rods captured many of the characteristics of deformation and failure mode observed in SNF, including the linear-to-nonlinear deformation transition and large residual curvature in static tests, PPI and PCMI failure mechanisms, and large variation in the initial structural condition. Rod degradation was measured and characterized by measuring the flexural rigidity; the degradation of the rigidity depended on both the moment amplitude applied and the initial structural condition of the rods. It was also shown that a cracking initiation site can be located on the internal surface or the external surface of cladding. Finally, fatigue damage to the bending rods can be described in terms of flexural rigidity, and the fatigue life of rods can be predicted once damage model parameters are properly evaluated. The developed experimental approach, test protocol, and analysis method can be used to study the vibration integrity of SNF rods in the future.
cladding failure↗