AGC-2 Graphite Preirradiation Data Analysis Report

Windes, William E.; Swank, W.D.; Rohrbaugh, David; Lord, Joseph

doi:10.2172/1097190

Cited by 7 publications

(9 citation statements)

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“…The trend with temperature for both IG-110 and NBG-18 is a result of measurement differences between INL and ORNL discussed above. Based on previous data [19], the isotropy ratios for IG-110 and NBG-18 are ~1.01 and ~0.98, respectively, and are constant with temperature as are the ratios for PCEA and NBG-17. …”

Section: Thermal Expansionmentioning

confidence: 55%

AGC-3 Graphite Preirradiation Data Analysis Report

Windes¹,

Swank²,

Rohrbaugh³

et al. 2013

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SUMMARYThis report describes the specimen loading order and documents all preirradiation examination material property measurement data for the graphite specimens contained within the third Advanced Graphite Capsule (AGC-3) irradiation capsule. The AGC-3 capsule is third in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated, which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. The general design of the AGC-3 test capsule is similar to the AGC-2 test capsule. Material property tests were conducted on graphite specimens prior to loading into the AGC-3 irradiation assembly. However, the six major nuclear graphite grades in AGC-2 were modified, two previous graphite grades (IG-430 and H-451) were eliminated, and one graphite grade was added (Mersen's 2114 was added). Specimen testing from three graphite grades (PCEA, 2114, and NBG-17) was conducted at Idaho National Laboratory (INL) and specimen testing for two grades (IG-110 and NBG-18) was conducted at Oak Ridge National Laboratory (ORNL) from May 2011 to July 2013. This report also details the specimen loading methodology for the graphite specimens inside the AGC-3 irradiation capsule. The AGC-3 capsule design requires "matched pair" creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This experiment utilized the neutron flux profile calculated for the AGC-3 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide specimens to create a stacking order that can produce "matched pairs" of graphite specimens above and below the AGC-3 capsule elevation mid-point to provide specimens with similar neutron dose levels.

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Section: Thermal Expansionmentioning

confidence: 55%

AGC-3 Graphite Preirradiation Data Analysis Report

Windes¹,

Swank²,

Rohrbaugh³

et al. 2013

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“…The AGR-5/6/7 graphite holders are made of IG-430 nuclear-grade graphite. Material properties for unirradiated graphite IG-430 were determined as follows: specific heat values, as function of temperature, were taken from American Society for Testing and Materials (2014); density and expansion coefficients (measured at room temperature, 20℃) were taken from Windes et al (2017) and Swank et al (2012); and thermal diffusivities for temperature range 20-1000℃ were taken from Windes et al (2013). Unirradiated thermal conductivity as a function of temperature is calculated as the product of the diffusivity, specific heat, and density.…”

Section: Thermal Analysismentioning

confidence: 99%

AGR-5/6/7 Experiment Monitoring and Simulation Progress

Pham

Sterbentz

Hawkes

et al. 2019

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Advanced Gas Reactor (AGR)-5/6/7 is the last of a series of experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tri-structural isotropic (TRISO) low-enriched fuel for use in high-temperature gas-cooled reactors. The test train contains five separate capsules that are independently controlled and monitored. Each capsule contains multiple 12.51-mm-long compacts filled with low-enriched uranium carbide/oxide (UCO) TRISO fuel particles. The objectives of the AGR-5/6/7 experiment are to: 1. Irradiate reference-design fuel particles to support fuel qualification. 2. Establish operating margins for the fuel beyond normal operating conditions. 3. Provide irradiated-fuel performance data and irradiated-fuel samples for post-irradiation examination (PIE) and safety testing. vi Capsule 1 occurred from the fourth cycle (164B), and were rigorously mitigated to minimize fission product crosstalk between capsules. This gas line problem prevented fission product release measurement for Capsule 1 during the last three cycles and caused various fission gas leakage events into the other four capsules. By the end of the sixth cycle, 166A, a significant number of in-pile particle failures occurred in Capsule 1, causing a substantial increase in R/B values. In addition, numerous particle failures were observed in Capsule 3 and perhaps a small number of failures might have occurred in Capsule 2 during the last cycle, 168A. In contrast, no in-pile failures are likely in the top two capsules (4 and 5) based on absence of the typical spike in gross gamma counts and low failure estimates by the end of Cycle 168A using the AGR-3/4 R/B per exposed kernel model. Increased and unstable R/Bs in Capsules 4 and 5 can be contributed to fission gas leakage from Capsule 1.

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“…AGC-3 design documents and drawings pertinent to the AGC-3 graphite specimens have been reported in the previous AGC-3 graphite pre-irradiation data analysis report. 14 The technical and functional requirements for the AGC-3 test train are given in TFR-791. 5…”

Section: Agc-3 Test Train Capsulementioning

confidence: 99%

“…17 Pre-irradiation data for the AGC-3 specimens can be found in the AGC-3 Graphite Pre-Irradiation Data Analysis Report. 14 These initial comparisons provide an initial examination of the data for trends and correlations that are intuitive and logical. In this way, the physical and thermal post-irradiation property measurements are vetted.…”

Section: Testingmentioning

confidence: 99%

“…MLRF (Hemispherical shaped samples of SGL's NBG-17 grade coated with silicon carbide -SiC) b. SGL/SGL-SiC (Piggyback button-shaped samples of SGL's NBG-17 grade coated and uncoated with silicon carbide) c. TS (Piggyback button-shaped samples of a new experimental graphite grade from GrafTech, Inc.) d. PCIB-MG (Piggyback button-shaped samples of a special experimental formulation of GrafTech, Inc.'s PCIB graphite grade) 5. Single crystal graphite: HOPG.A more complete description of all of the graphite specimens included in the AGC-3 capsule is given in Table1 14.…”

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confidence: 99%

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AGC-3 Specimen Post-Irradiation Examination Data Package Report

Windes

Rohrbaugh

Swank

et al. 2017

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This report documents results of the post-irradiation examination material property testing from the third advanced graphite creep (AGC), AGC-3, capsule specimens. This is the third of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades to moderate dose levels (≤7 dpa). The AGC-3 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 820°C and to a peak dose of 3.7 dpa. Half of the AGC-3 specimens were subjected to compressive stresses to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. The data reported includes specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and dimensional data necessary to derive density, elastic constants (Young's modulus, shear modulus, and Poisson's ratio) from ultrasonic time of flight velocity measurements, Young's modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100 to 650°C.An abridged statistical analysis was performed on the irradiated data and a limited comparison between pre-and post-irradiation properties is presented to ensure that any property measurements exhibiting values significantly higher or lower than the average were not a measurement error. A more complete evaluation of trends in the material property changes, as well as irradiationinduced creep due to the irradiation environment and applied load on the specimens, will be discussed later in AGC-3 post-irradiation examination analysis reports.

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AGC-2 Graphite Preirradiation Data Analysis Report

Cited by 7 publications

References 0 publications

AGC-3 Graphite Preirradiation Data Analysis Report

AGC-3 Graphite Preirradiation Data Analysis Report

AGR-5/6/7 Experiment Monitoring and Simulation Progress

AGC-3 Specimen Post-Irradiation Examination Data Package Report

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