Irradiation creep and swelling of the US fusion heats of HT9 and 9Cr-1Mo to 208 dpa at ~ 400°C

Toloczko, Mychailo B.; Garner, F.А.; Eiholzer, C.R.

doi:10.1016/0022-3115(94)90131-7

Cited by 126 publications

(54 citation statements)

References 2 publications

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“…Although the elevated-temperature mechanical properties favor NF616, unfortunately, available data are lacking for this steel after neutron irradiation. In addition to the toughness advantage after irradiation of the modified 9Cr-1Mo steel over the HT9 steel, the modified 9Cr-1Mo steel has been shown to be more resistant to irradiation creep [9,10].…”

Section: Introductionmentioning

confidence: 98%

Mechanical properties and microstructural evolution of modified 9Cr-1Mo steel after long-term aging for 50,000 h

et al. 2009

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The mechanical properties and microstructural evolution of modified 9Cr-1Mo steel have been studied to investigate steel property changes after long-term isothermal aging at 600 o C for 50,000 h. The microhardness and strength were maintained constantly after aging but the impact energy was dramatically reduced by 62 % during the aging period. From the viewpoint of microstructural evolution after the aging process, Cr-enrichment and Fe-depletion took place within the M23C6-type precipitates in the as-aged steel and V-depletion also happened within the VX-type precipitates after aging. In addition, the precipitates of the M2Mo-type Laves phase and the segregation of the impurity atoms would be formed during the long-term aging period. It was considered that the sharp reduction of the impact energy could be related to the formation of the Laves phases and the impurity segregation after aging at 600 o C. The phase stability was also verified by the specific heat results up to 950 o C from a DSC test. It was concluded from this study that the modified 9Cr-1Mo steel would keep its microstructural stability at 600 o C during the long-term aging period of 50,000 h, which was equivalent to the in-service life of the SFR fuel cladding.

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Section: Introductionmentioning

confidence: 98%

Mechanical properties and microstructural evolution of modified 9Cr-1Mo steel after long-term aging for 50,000 h

et al. 2009

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“…In this study, three alternative interpretations of A for annular metallic fuel have been considered, defined as: (18) where ΔT f is the difference between the average fuel and average coolant temperatures.…”

Section: Introduction To the Methodsmentioning

confidence: 99%

“…This was achieved without excessive swelling or mechanical problems for HT9 steel in the Fast Flux Test Facility (FFTF) reactor. The corresponding number of displacements per atoms (DPA) was calculated to be ~208 [18]. Conservatively, 200 DPA or simply a fast fluence below 4 × 10 23 n/cm 2 is most commonly used as the upper irradiation damage constraint in fast reactor design.…”

Section: Irradiation Damage Limitsmentioning

confidence: 99%

Optimizing the Design of Small Fast Spectrum Battery-Type Nuclear Reactors

Qvist

2014

Energies

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This study is focused on defining and optimizing the design parameters of inherently safe "battery" type sodium-cooled metallic-fueled nuclear reactor cores that operate on a single stationary fuel loading at full power for 30 years. A total of 29 core designs were developed with varying power and flow conditions, including detailed thermal-hydraulic, structural-mechanical and neutronic analysis. Given set constraints for irradiation damage, primary cycle pressure drop and inherent safety considerations, the attainable power range and performance characteristics of the systems are defined. The optimum power level for a core with a coolant pressure drop limit of 100 kPa and an irradiation damage limit of 200 DPA (displacements per atom) is found to be 100 MWt/40 MWe. Raising the power level of an optimized core gives significantly higher attainable power densities and burnup, but severely decreases safety margins and increases the irradiation damage. A fully optimized inherently safe battery-type fast reactor core with an active height and diameter of 150 cm (2.6 m 3 ), a pressure drop limit of 100 kPa and an irradiation damage limit of 300 DPA can be designed to operate at 150 MWt/60 MWe for 30 years, reaching an average discharge burnup of 100 MWd/kg-actinide.

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“…The results presented in this article encompass the temperature range 305-700°C. The radiation creep of foreign-made ferrite-martensite steel is presented most fully in [14][15][16][17][18][19]. The investigations were performed with gas-fi lled ampoules, made primarily from NT-9 steel, in the range 400-605°C.…”

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confidence: 99%

Swelling and Radiation Creep of Ferrite-Martensite Steel Irradiated in the Bn-350 Reactor in a Wide Range of Temperature and Damaging Dose

et al. 2016

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The results of investigations of swelling and in-reactor creep of the ferrite-martensite class steels to damaging dose in the range 20-89 dpa are presented. The characteristics of radiation creep of three of the ferrite-martensite steels are close despite the differences of the chemical composition and heat-treatment. The relation B = 4.864·10 -2 + 1.45·10 -3 T describes the average modulus of radiation creep of the experimental steel in the interval 305-550°C.The transition of nuclear technologies to a new level and the development of next-generation reactor cores require structural materials for fuel elements and fuel assemblies that are capable of operating to damaging dose 150 dpa and higher. The development of such materials must be based on reliable data on the radiation damageability of prototype steel and research on the structural changes occurring under irradiation and the infl uence of these changes on the physical-mechanical properties. In this respect, ferrite-martensite class steels are the most promising materials for use in new nuclear plants. Several types of 12% chromium steel have been developed in our country for different purposes [1][2][3][4]. One of the most widely available steels is EP-450, which is used for fuel-assembly jackets and guide tubes of the safety-and-control system channels in BN-600 and -800 reactors [1, 5, 6]. The steels EI-852 and EP-823 were developed for facilities with lead-bismuth or lead coolant [2,7,8].The main advantages of ferrite-martensite steel are high resistance to vacancy swelling and low rate of radiation creep together with acceptable mechanical properties, which was shown in a post-irradiation investigation of this steel [9-13]. Nonetheless, there are not enough experimental data on radiation resistance; the greatest data defi ciency is observed for high damaging dose as well as low and high irradiation temperatures. Part of this data was obtained by irradiating the steel in BN-350. This reactor was widely used to irradiate samples of structural materials. A feature of this reactor was low sodium temperature at the core entry, which together with throttling of the assemblies or additional heating of the samples made it possible to conduct irradiation in a temperature range that is inaccessible for other reactors.Material-research assemblies irradiated in BN-350 were investigated in the hot laboratory at the FEI over a period of many years. In the present article, we present the results of investigations of swelling and radiation creep of irradiated

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Irradiation creep and swelling of the US fusion heats of HT9 and 9Cr-1Mo to 208 dpa at ~ 400°C

Cited by 126 publications

References 2 publications

Mechanical properties and microstructural evolution of modified 9Cr-1Mo steel after long-term aging for 50,000 h

Mechanical properties and microstructural evolution of modified 9Cr-1Mo steel after long-term aging for 50,000 h

Optimizing the Design of Small Fast Spectrum Battery-Type Nuclear Reactors

Swelling and Radiation Creep of Ferrite-Martensite Steel Irradiated in the Bn-350 Reactor in a Wide Range of Temperature and Damaging Dose

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