The irradiation creep in reactor graphites for HTR applications

Veringa, H.J.; Blackstone, R.

doi:10.1016/0008-6223(76)90131-7

Cited by 13 publications

(4 citation statements)

References 11 publications

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“…Veringa and Blackstone [27] reported creep data for Gilso-carbon graphite and extruded pitch coke graphite irradiated at 650°C. The kE 0 values for the two graphite types were 2.1 Â 10 À21 cm 2 /n [E > 50 keV] (E 0 = 10.5 GPa) and 1.75 Â 10 À21 cm 2 /n [E > 50 keV] (E 0 = 10 GPa assumed here), respectively.…”

Section: Compressive Creepmentioning

confidence: 98%

Irradiation induced creep behavior of H-451 graphite

Burchell

2008

Journal of Nuclear Materials

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Section: Compressive Creepmentioning

confidence: 98%

Irradiation induced creep behavior of H-451 graphite

Burchell

2008

Journal of Nuclear Materials

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“…In general, AGC-2 creep coefficients agree remarkably well with the AGC-1 coefficients even with the large temperature variations between AGC-1 specimens. AGC-2 creep coefficients are compared to previous values calculated from historical creep studies 37,38,39,40,41,42,43,44,45 (Figure 27). This comparison demonstrates the AGC-2 creep coefficients are well within the historical values determined for the irradiation temperatures and dose levels of the AGC-2 test train.…”

Section: Creep and Creep Coefficient Analysis By Gradementioning

confidence: 99%

AGC-3 Irradiation Creep Strain Data Analysis

Windes

Rohrbaugh

Swank

2019

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This report documents the analysis of the creep strain data from the Advanced Graphite Creep (AGC)-2 graphite creep specimens. This is the second 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. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C, beginning with irradiation Cycle 149A on April 12, 2011 and ending with Cycle 151B on May 5, 2012, with a total received dose range of 1.3-4.7 dpa.AGC-2 was designed to provide irradiation conditions similar to AGC-1 (i.e, the same graphite grades, a nominal irradiation temperature of 600°C, and the same applied mechanical stress levels) but was irradiated for a shorter period of time to provide material property values for the graphite samples at lower dose levels than achieved in AGC-1. Material property and dimensional strain measurements were conducted on specimens from 15 nuclear graphite grades using a similar specimen assembly configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114.Significant modifications were made to the AGC-2 irradiation capsule to improve the specimen temperature issues that were encountered with AGC-1. While the AGC-2 irradiation temperature range (399-704°C) was much narrower than AGC-1 the range is still considered less than optimal. The centrally located creep specimens (specimens under an applied mechanical stress of 13.8, 17.2, or 20.7 MPa) and control specimens (specimens with no applied stress) were irradiated over even a narrower temperature range of 541-681°C with dose levels from 2.0-4.7 dpa.The data reported include specimen dimensions and hence the dimensional strain change upon irradiation. This allowed a comparison of these data for specimen-matched pairs yielding the dimensional and volumetric creep strain. The AGC-2 creep strain analysis methodology is similar to the AGC-1 analysis and the irradiation strain results from both capsules compare well for all grades. The derived creep coefficients have been calculated for each grade and are found to compare well to literature data, despite the larger than desired spread in specimen temperatures.

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“…Some results on the behaviour of graphite are available, e.g. dimensional changes [19,20], oxidation [21], irradiation creep [22], etc. SiC-based materials shall also be tested for creep behaviour and resistance [23].…”

Section: Introductionmentioning

confidence: 99%

New Experimental Loops for Research Reactor LVR-15

Ruzickova

Hájek

Vsolak

et al. 2008

MSF

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Two experimental loops for operation in research reactor LVR-15 in ÚJV Řež are recently under preparation: High Temperature Helium Loop (HTHL) and SuperCritical Water Loop (SCWL). Pure helium will be used as working medium in HTHL and its main physical parameters are: operating pressure 7MPa, max. temperature in the test section 900°C and flow rate 36kg/h. HTHL will include helium purification system, system for dosage of impurities (e.g. CO2, H2, H2O, O2, N2 etc.) and helium sampling. Helium purification experiments and testing of materials in simulated HTR conditions will take place in HTHL in the future. Main parameters of the SCWL are 25MPa, max. temperature in the test section 600°C, flow rate max. 200kg/h. SCWL will be used for corrosion tests of candidate materials, studies of water radiolysis at supercritical conditions and for testing of water chemistry suitable for operation. Both loops possess an irradiation channel with quite complicated internals design, whose complexity is imposed by current constraints on constructional materials of nuclear experimental devices, which limit the choice and maximum surface temperature of material of construction to 500°C for austenitic stainless steel. The working temperature will thus be attained only in a restricted volume of the test section. The channel internals will be briefly described. The mentioned loops will represent novel experimental devices, whose objective is to gain and extend knowledge on materials and environment performance under the influence of radiation.

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The irradiation creep in reactor graphites for HTR applications

Cited by 13 publications

References 11 publications

Irradiation induced creep behavior of H-451 graphite

Irradiation induced creep behavior of H-451 graphite

AGC-3 Irradiation Creep Strain Data Analysis

New Experimental Loops for Research Reactor LVR-15

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