Nuclear Engineering Division

Irradiation Performance

Light Water Reactor Materials

 

The Section’s activities in the fission reactor area are focused principally on light water reactors (LWRs). The principal ongoing and upcoming activities in this research are described below.

High-Burnup Cladding Performance

The program, "High-Burnup Cladding Performance", is a major long-range research effort in the Section with participation by staff from NE’s Corrosion Section. Funding comes from the NRC Office of Nuclear Regulatory Research (NRC-RES). Recently, NRC renamed the program: Advanced Fuel Cladding Response to Limiting Conditions. The purposes of this program are to determine the behavior of LWR fuel rods under conditions relevant to loss-of-coolant accidents (LOCA) and to establish a mechanical properties data base for high-burnup (>60 MWd/kg) Zircaloy and advanced-alloy cladding materials. These data are needed for the analysis of various transients, including LOCA, reactivity-initiated accidents (RIA), and anticipated transients without scram (ATWS), and are important in licensing-safety analyses and acceptance criteria. The mechanical properties data are also needed for assessment of spent-fuel cladding behavior during transfer, transportation, dry-cask storage, transport to the repository and repository storage. To accomplish these objectives, the project is divided into three major tasks: Fuel Characterization, LOCA-Related Tests, and Mechanical Properties Tests.

Fuel Selection and Characterization

Under special agreements between the Electric Power Research Institute (EPRI) and NRC-RES, EPRI has supplied us with segments of seven Zircaloy-2-clad fuel rods from a boiling water reactor (BWR), twelve Zircaloy-4-clad fuel rods from a pressurized water reactor (PWR) and 4 advanced-alloy M5 fuel rods from a PWR. The PWR fuel rods had undergone high burnup (>60 MWd/kg), while the BWR fuel rods had undergone about 56 MWd/kg. Zircaloy-2 and -4 fuel and cladding have been characterized by the IPS in the Alpha-Gamma Hot-Cell Facility (AGHCF). The high-burnup BWR rod segments were received at Argonne in May 2000, the high-burnup Zircaloy-4 PWR fuel rods were received in May 2001, and the M5 PWR rods arrived in March 2007. In addition, three lower-burnup PWR rods (≈36 MWd/kg), which had been stored in dry casks for 15 years, were received at Argonne in February 2001. Early in the program, PWR fuel rod segments (≈50 MWd/kg) from the Three Mile Island-1 (TMI-1) Reactor were used to demonstrate our ability to characterize fuel and cladding and to validate the experimental procedures for LOCA and mechanical properties testing.

LOCA-Related Tests

This task involves steam-oxidation-kinetics studies of traditional-Zircaloy and advanced-alloy (ZIRLO and M5) cladding at high temperatures (1000-1200°C), breakaway oxidation tests at lower temperatures (800-1000°C), LOCA-integral tests, and post- LOCA-quench ductility tests. The oxidation kinetics studies have been completed on nonirradiated and high-burnup Zry-2 at 1000-1200°C. Similar work is in progress for nonirradiated and high-burnup Zry-4. In parallel to this effort, we are comparing the oxidation kinetics and post-quench ductility of advanced cladding alloys (ZIRLO and M5, as well as the Russian E110 alloy) to the behavior of Zry-2 and Zry-4 tested under the same conditions and analyzed by a common methodology. Samples of these alloys are oxidized at temperatures in the range of 1000-1200°C, cooled to 800°C, and water-quenched prior to subjecting them to ring-compression ductility tests. Extensive data have been generated for as-fabricated Zry-2, Zry-4, M5, ZIRLO and E110), for pre-hydrided Zry-4 and for high-burnup Zry-4. The results have been documented in an NRC NUREG report and will be used by NRC-RES to formulate a Research Information Letter to NRC-NRR (licensing) recommending changes to the current LOCA acceptance criteria. Argonne is the only lab with permission to test these cladding alloys and to publish the results in the open literature. Efforts are in progress to complete the testing for pre-hydrided and high-burnup M5 and ZIRLO (supplied by Studsvik in cooperation with NRC).

The LOCA-integral tests involve ramping (at 5°C/s) the temperature of pressurized, fuel-cladding segments in a flowing-steam environment through ballooning and burst (at ≈750°C) to 1204°C, holding for various times to generate oxidation levels consistent with current NRC criteria (17% equivalent cladding reacted), slow cooling (at 3°C/s) to 800°C, and rapid cooling through water quenching. Behavior of the cladding during, and following, the water quench determines whether the current NRC criteria for LOCA events are conservative for high-burnup fuel rods. Four tests with high-burnup BWR fuel segments have been completed: ramp-to-burst to study ballooning and burst behavior; ramp-to-1204°C, hold in flowing steam for 5 min, and slow cool to study fuel behavior and hydrogen and oxygen pickup; and the full LOCA integral sequence, including quench. Argonne is the only lab to have conducted these tests with high-burnup fuel samples. Baseline tests have also been performed with pressurized nonirradiated Zry-2.

high-burnup BWR Zry-2(a)

nonirradiated BWR Zry-2(b)

LOCA integral results for ramp-to-burst tests: (a) high-burnup BWR Zry-2 and (b) nonirradiated BWR Zry-2.

Mechanical Properties Tests

Tensile tests with high and low strain rates for RIA and LOCA analyses are in progress to determine the mechanical properties of BWR and PWR Zircaloy cladding in both the hoop and axial directions. In addition, biaxial tests are in progress using the plane-strain specimen geometry developed by Penn State University (PSU).

Extensive finite-element analyses have been conducted by the NE Corrosion and Mechanics of Materials Section to determine the optimal specimen geometry and loading inserts for these tests and to interpret the results of the ring hoop-tensile tests. TMI-1 PWR cladding was used to demonstrate IPS capabilities of performing hoop tensile tests with irradiated cladding ring samples. However, the demonstration test, which was performed out-of-cell, proved to deposit too much contamination on the loading grips to continue testing. As a result, a radiological glove box was constructed to house the testing machine and samples. A smaller glove box was also built for installation of an automatic microhardness indenter. The changes in spacing of the microhardness indents are used to determine local strains directly following the test. Axial, hoop, and PSU plane-strain tests of high-burnup cladding are in progress.

Dry Cask Storage

The IPS has been investigating spent-fuel behavior during dry cask storage and transportation-accident conditions. Issues related to low-to-intermediate burnup fuel have been resolved. However, data are needed for high-burnup cladding before transport casks can be licensed. High-burnup degradation - mainly hydrogen pickup and orientation of hydrides - affects cladding integrity during storage and transport. Improvements in the mechanical properties data base are needed to provide a technical basis for licensing dry-cask transportation of high-burnup spent nuclear fuel.

For intermediate-burnup fuel rods (≤45 MWd/kg), IPS completed its program in characterizing and testing (axial tensile properties and long-term thermal creep) of PWR cladding (from the Surry Reactor) that had been irradiated to 36 MWd/kg and stored in a helium-filled cask for 15 years. The results of these studies were used by the NRC Spent Fuel Project Office to formulate staff guidance (ISG-11, Rev. 2 and Rev. 3) regarding criteria for vacuum-drying and transfer, transport to interim storage, dry storage, and transport to a permanent repository. In particular, the thermal creep data generated by IPS at 400°C were used to justify the 400°C limit imposed on these operations. Even after 15 years of dry storage the Surry cladding had a creep ductility > 5%. The Surry work was funded jointly by NRC, EPRI and DOE.

For the high-burnup portion of this work, several H. B. Robinson PWR rods (67 MWd/kg) have been used in the NRC-sponsored program to generate data on fuel isotopic content, fuel and cladding metallography, cladding hydrogen analysis, cladding tensile properties, and cladding creep properties. Argonne’s Chemical Sciences and Engineering Division conducted the fuel isotopic and burnup analyses for this program. Thermal creep tests were conducted using high-burnup PWR cladding samples with ≈600 wppm hydrogen. These samples have more than twice the hydrogen content and neutron exposure - both considered to decrease creep rate and creep ductility - of the Surry Reactor samples. There were two surprising results from this investigation. First, the Robinson cladding exhibited a higher creep strain rate than the Surry cladding with both materials tested at 400°C and 190 MPa hoop stress.

Thermal creep of Surry cladding at 400°C after 15 years of dry cask storageThermal creep of Surry cladding at 400°C after 15 years of dry cask storage. Click on image to view larger image.

nonirradiated BWR Zry-2Thermal creep behavior of high-burnup Robinson and low-burnup Surry cladding at 400°C and 190 MPa. Click on image to view larger image.

The second surprising result came after the last run of Robinson sample C15. It was cooled under full pressure, much the same as two of the Surry samples had been cooled, to study the reorientation of hydrides from the circumferential to radial direction. The presence of radial hydrides can significantly reduce the ductility of the cladding in response to hoop-stress loading. After the sample cooled to ≈200°C at a hoop stress of 190 MPa, it failed rather dramatically, even though Surry samples had survived such cooling under higher internal pressures and stresses. Although this failure occurred in the end-cap weld region failure, it resulted in a redirection of the program to investigate the conditions under which hydrides reorient to the radial direction and the consequences of such reorientation on cladding response to hoop loading under high-strain-rate and impact conditions. DOE-RW, independently of NRC, funded IPS to study the effects of radial hydrides on high-burnup cladding performance because of concerns regarding cladding response to transportation of the fuel to permanent repositories.

Having the commercial high-burnup fuel at Argonne has proven to be a great asset. Through a DOE contract with Sandia National Laboratories to investigate the vulnerability of fuel in dry casks to transportation accidents and sabotage, IPS has fabricated test rodlets for Sandia from the high-burnup Robinson PWR fuel segments, as well as from the dry-cask-stored Surry rod segments. This effort was initiated late in FY2003 and will continue through FY2008.

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Last Modified: Thu, April 21, 2016 5:02 AM

 

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Irradiation Performance Section
Sect. Manager: Michael C. Billone
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