Irradiation Performance Data Research Articles

Thermal Model of the AGR-5/6/7 Experiment with Offset Gas Gaps

Because of their high impact on calculated temperatures, the offset gas gaps between the graphite holder and capsule wall are evaluated for the thermal model of the Advanced Gas Reactor 5/6/7 (AGR-5/6/7) irradiation experiment. Fuel compact temperatures are a crucial factor in assessing the irradiation performance of tristructural isotropic (TRISO)-coated-particle fuel for use in high-temperature gas-cooled reactors. TRISO fuel irradiation performance is a core tenet of the U.S. Department of Energy’s Advanced Gas Reactor Fuel Development and Qualification Program, conducted in the Advanced Test Reactor at Idaho National Laboratory. Specifically, the irradiation experiments were intended to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support the development of fuel performance models and codes, and provide irradiated fuel and materials for postirradiation examination and safety testing. The thermal model used to predict the fuel compact temperatures was a three-dimensional finite element model that was subject to uncertainty. The most dominant factor in the uncertainty pertaining to calculated fuel temperatures is the gas gap uncertainty, which is caused by the graphite holder nub-to-capsule wall clearance. In AGR-5/6/7 capsules, the graphite holders can be shifted away from being perfectly centered inside the stainless steel capsule walls. This is due to irradiation-induced graphite shrinkage or to insufficient nubs on the graphite holder leading to azimuthally varying gas gaps during irradiation. A method was developed to find the best offset magnitude and direction at the top and bottom of the graphite holders in Capsules 1 and 2, which experienced higher temperatures than originally predicted among the five AGR-5/6/7 axially dispersed capsules irradiated between 2018 and 2020. The differences between the calculated and the measured temperatures of thermocouples (TCs) placed in the graphite holder near the fuel compacts were used as the basis for the holder offset determination. The best-fit offset assumes that the root-mean-square error between the measured and calculated TC temperatures is a minimum value. Typical offsets in the range of 0.002 to 0.004 in. increase the fuel compact temperatures in the range of 100°C to 200°C when compared to the zero offset for Capsules 1 and 2. Additionally, because of irradiation-induced changes in the capsule geometries, the holder offsets change throughout the irradiation period.

Status of the NGNP graphite creep experiments AGC-1 and AGC-2 irradiated in the advanced test reactor

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the next generation nuclear plant (NGNP) very high temperature gas reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have three different compressive loads applied to the top half of three diametrically opposite pairs of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment.

Status of the NGNP fuel experiment AGR-2 irradiated in the advanced test reactor

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also undergo on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

Design and status of the NGNP fuel experiment AGR-3/4 irradiated in the advanced test reactor

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) advanced gas reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the advanced test reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation.The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is significantly different from the first two experiments, though the control and monitoring systems are very similar. The purpose and design of this experiment will be discussed followed by its progress and status to date.

MOX Fuel Performance Experiment Specified for Japanese PWR Utilisation in the HBWR

In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pufissile enrichment of about 6 wt% have been irradiated in the HBWR. In-pile performance data of MOX have been obtained, and the peak burn-up of MOX pellet have reached to 66 GWd/tM as of October 2004. MOX fuel temperature is confirmed to have no significant difference compared to UO2, if taking into account adequately for thermal conductivity degradation due to PuO2 addition and burn-up development, and measured fuel temperature agrees well with HB-FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly larger than UO2 based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behaviour. MOX fuel swelling rate agrees well with solid swelling rate. Cladding elongation data shows onset of PCMI in high power region. Ramp test data from other experiment programs with various types of MOX fabrication route confirms superior PCI resistance of MOX compared to UO2, due to enhanced creep rate of MOX. The irradiation is expected to continue until achieving of 70 GWd/tM (MOX pellet peak).

UO2-PUO2–A Demonstrated Fuel for Plutonium Utilization in Thermal Reactors

Irradiation performance data available to date indicate that the behavior of Zircaloy-clad thermal reactor mixed-oxide fuels is entirely satisfactory. There do not appear to be any inherent performance limitations created by the addition of small amounts of PuO2 to UO2. The wide range of test conditions include peak linear heat ratings sufficient to cause fuel melting and peak burnups to over 50 000 MWd/MTM.Because small PuO2 additions to UO2 have an insignificant effect on those fuel properties which could affect irradiation behavior, thermal reactor mixed-oxide fuel performs similarly and equally as well as UO2-only fuel. The evidence indicates that the PuO2-rich regions that persist in nonrestructured mechanically mixed, mixed-oxide fuel have an insignificant effect on performance. The extent of plutonium migration which can occur during irradiation also has an insignificant effect on mixed-oxide fuel behavior. Satisfactory defect behavior of Zircaloy-clad UO2-PuO2 fuel under normal operating condition...