Microstructural analysis of an HT9 fuel assembly duct irradiated in FFTF to 155dpa at 443°C

Sencer, Bulent H.; Kennedy, J. R.; Cole, James I.; Maloy, S.A.; Garner, F.А.

doi:10.1016/j.jnucmat.2009.06.010

Cited by 82 publications

(37 citation statements)

References 33 publications

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“…The model is based on the fuel/cladding/coolant volume fractions and oxide fuel composition (Table IV) Using the calculated one-group dpa cross section (about 0.024 barn·MeV) and assuming the displacement energy to be 40 eV, it is found that 4.0 dpa should be accumulated in the FFTF for every 10 22 neutrons (>0.1 MeV)/cm 2 . This is only slightly lower 32 Therefore, the approach of using dpa to assess radiation damage is consistent with the value reported for the FFTF. Nevertheless, the SFR community is widely using a fast neutron fluence of 4 × 10 23 neutrons (>0.1 MeV)/cm 2 as the radiation damage constraint 10 ; all the ANL SFR and GE S-PRISM core designs are based on this constraint.…”

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confidence: 86%

Advanced Burner Reactors with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

2017

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-This paper assesses the feasibility of designing seed-and-blanket (S&B) sodium-cooled fast reactor (SFR) cores to generate a significant fraction of the core power from radial thorium-fueled blankets that operate in the breed-and-burn (B&B) mode. The radiation damage on the cladding material in both seed and blanket does not exceed the presently acceptable constraint of 200 displacements per atom (dpa). The S&B core is designed to have an elongated seed (or driver) to maximize the fraction of neutrons that radially leak into the subcritical B&B blanket and reduce the neutron loss via axial leakage. A specific objective of this study is to maximize the fraction of core power generated by the B&B blanket that is proportional to the neutron leakage rate from the seed to the blanket. Since the blanket feed fuel is very inexpensive and requires no reprocessing and remote fuel fabrication, a larger fraction of power from the blanket will result in a lower fuel cycle cost per unit of electricity generated by the SFR core. It is found possible to design the seed of the S&B core to have a lower transuranics (TRU) conversion ratio (CR) than a conventional advanced burner reactor (ABR) core without deteriorating core safety. This is due to the unique synergism between a low CR seed and the B&B thorium blanket. The benefits of the synergism are maximized when using an annular seed surrounded by inner and outer thorium blankets. Two high-performance S&B cores are designed to benefit from the annular seed concept: (1) an ultra-long-cycle core having a CR = 0.5 seed and a cycle length of~7 effective full-power years (EFPYs) and (2) a high-transmutation core having a TRU CR of 0.0. The TRU transmutation rate of the latter core is comparable to that of the reference ABR with a CR of 0.5, and the thorium blanket can generate close to 60% of the core power. Because of the high blanket power fraction along with the high discharge burnup of the CR = 0 seed, the reprocessing capacity per unit of core power required by this S&B core is only approximately 1/6th that of the reference ABR core with a TRU CR of 0.5. Although the seed fuel CR is nearly zero, the burnup reactivity swing is low enough to enable a cycle length of more than 4 EFPYs. This is attributed to a combination of reactivity gain in the thorium blankets over the cycle and the relatively high heavy metal inventory. Moreover, despite the very low leakage, the S&B cores feature a less positive coolant reactivity coefficient and large enough negative Doppler coefficient even when using nonfertile fuel for the seed, because of the unique physics properties of the 233 U and Th in the thorium blankets. With the long cycles, the S&B SFR is expected to have a higher capacity factor, and therefore a lower cost of electricity, than conventional ABRs. The discharge burnup of the thorium blanket fuel is typically 70 MWd/kg such that the thorium fuel utilization is approximately 12 times that of natural uranium in light water reactors. A sensitivity study is subsequently un...

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confidence: 86%

Advanced Burner Reactors with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

2017

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“…The swelling-dose curve is plotted in Figure 3 and the steady state swelling rate was calculated to be 0.003%/dpa based upon these two conditions. This is similar to the swelling rate of 0.002% calculated by Sencer et al in a neutron irradiation of HT9, another ferritic-martensitic alloy, up to 155 dpa at 443 o C. [3] Further work will include irradiations to higher doses to map out the steady state swelling rate, as well as examining the effect of temperature on void nucleation and growth behavior. Additionally, other alloys including various heats and heat treatments of HT9, will be irradiated and examined, in order to determine an accurate method for modeling neutron damage with heavy ions.…”

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confidence: 80%

Void Swelling in Self-Ion Irradiated Ferritic-Martensitic Alloy T91

Beckett¹,

Jiao²,

Wang³

et al. 2013

Microsc Microanal

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Determining the microstructural behavior of body centered cubic (BCC) ferritic-martensitic alloys is important for predicting the safety and structural integrity of fast spectrum reactors. Of particular interest is the phenomenon of radiation-induced void swelling, which could potentially cause dimensional changes in key structural components in reactors. Irradiations performed with heavy ion irradiations can be used to model neutron irradiations with the benefits of accelerated dose rate, decreased irradiation time required and excellent temperature control of irradiated samples.Self-ion irradiation experiments have been performed on ferritic-martensitic alloy T91 to determine swelling behavior at 440°C to doses of 280 dpa and 375 dpa, in order to examine the effect of irradiation dose on void size, density and swelling. T91 is a modified 9Cr-1Mo alloy, which has demonstrated swelling resistance in previous neutron irradiation studies.[1] Samples were preimplanted with 100±10 atom parts per million (appm) helium at a depth of 300-1000 nm from the sample surface. Irradiations were performed with 5 MeV Fe++ ions on samples using raster scanning on a Tandetron accelerator at the Michigan Ion Beam Laboratory. The effect of dose on void swelling was determined by examining the void distribution using transmission electron microscopy (TEM). TEM samples were prepared using the liftout technique on a FEI Quanta Focused Ion Beam (FIB) system.The irradiated microstructure at 280 dpa is shown in Figure 2. The microstructure includes many dislocation loops, lines and precipitates, which is characteristic of ferritic-martensitic alloys. The damage curve and the pre-implanted helium distribution were superimposed over the sample irradiated to 375 dpa as shown in Figure 1. A comparison of void data at both doses is given in Table 1. Voids preferentially nucleated over the area pre-implanted with helium. The additional 100 dpa of dose greatly increased void diameter, density and swelling, which indicates that the sample is still in the nucleating regime at 280 dpa but is entering the steady state growth regime by 375 dpa. Despite the large increase in void number, density and swelling at 375 dpa, T91 demonstrates superior swelling resistance to high doses, relative to stainless steel alloys which reach swelling rates as high as 1%/dpa. [2] Though the amount of neutron data is limited, T91 irradiated in the Fast Flux Test Facility at 420 o C to 203 dpa showed swelling of 1.76%. [1] This, when compared to the self-ion irradiated data, indicates that self-ion irradiations may include a longer nucleation regime before entering the growth regime. The swelling-dose curve is plotted in Figure 3 and the steady state swelling rate was calculated to be 0.003%/dpa based upon these two conditions. This is similar to the swelling rate of 0.002% calculated by Sencer et al. in a neutron irradiation of HT9, another ferritic-martensitic alloy, up to 155 dpa at 443 o C. [3] Further work will include irradiations to higher doses to map out the s...

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“…Experimental observations of neutron irradiated FM steel HT-9 duct in FFTF revealed glissile a o /2h1 1 1i and sessile a o h1 0 0i interstitial loops following irradiation [12].…”

Section: Dislocation Loop Naturementioning

confidence: 90%

Anisotropic dislocation loop distribution in alloy T91 during irradiation creep

Was

2014

Journal of Nuclear Materials

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a b s t r a c tFerritic-martensitic (FM) steel T91 was subjected to irradiation with 3 MeV protons while under load at stresses of 100 MPa, 180 MPa, and 200 MPa. Transmission electron microscopy of crept T91 samples was conducted to determine the nature of the loops and to quantify the edge-on dislocation loop density as a function of their orientation to the tensile axis. Loops were determined to be interstitial in nature of the type a o h1 0 0i. The distribution of a o h1 0 0i loops was found to be strongly anisotropic in which loops preferentially formed with their normal in the tensile direction. An empirical correlation was developed to describe the relationship between the orientation of the loop normal to the tensile axis and the local dislocation loop density. Loop anisotropy was found to increase as a function of the applied stress and their strain contributions relative to total creep strain were calculated at 4-11% of the total creep strain. Results from this study provide a means for quantifying empirically measured anisotropic microstructure and including it in irradiation creep mechanisms for FM steels.Published by Elsevier B.V.

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Microstructural analysis of an HT9 fuel assembly duct irradiated in FFTF to 155dpa at 443°C

Cited by 82 publications

References 33 publications

Advanced Burner Reactors with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

Advanced Burner Reactors with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

Void Swelling in Self-Ion Irradiated Ferritic-Martensitic Alloy T91

Anisotropic dislocation loop distribution in alloy T91 during irradiation creep

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