The next generation of fast reactors

Адамов, Е. О.; Orlov, V. V.; Filin, A. I.; Leonov, V. P.; Sila-Novitski, A.; Smirnov, V. S.; Tsikunov, V. S.

doi:10.1016/s0029-5493(97)00098-8

Cited by 45 publications

(13 citation statements)

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“…A power reactor design based on lead cooling and nitride fuel was developed in Russia under the acronym BREST [14]. This particular combination of fuel and coolant makes the design highly resistant to both UTOP and ULOF accidents, which will permit loading of a higher fraction of minor actinides and hence maximise transmutation rates.…”

Section: Lead Coolantmentioning

confidence: 99%

Physics of Americium Transmutation

Wallenius¹

2012

Nuclear Engineering and Technology

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Section: Lead Coolantmentioning

confidence: 99%

Physics of Americium Transmutation

Wallenius¹

2012

Nuclear Engineering and Technology

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“…[14][15][16][17] IPPE has also developed designs for 300 MWe and 1200 MWe Pb-cooled fast reactors: the BREST-300 and BREST-1200 reactors, respectively. 14,15,16 These nuclear-power-plant designs include the use of a pool-type reactor, a new system of refueling that permits a reduction of the overall dimensions of the central hall and buildings, a system of emergency cooling via tubes arranged directly in the Pb, and the use of emergency natural-circulation cooling, in case the circulating pump trips. Corrosion management was developed at the Laboratory for Heavy-metal Coolants Technology (IPPE-HMCT).…”

Section: Russia: Institute Of Physics and Power Engineeringmentioning

confidence: 99%

“…14,15 The selection and testing of high temperature structural materials, by using them in the presence of Pb and Pb-Bi at temperatures up to 1000 C. 22 Two facilities are of particular interest at IPPE:…”

mentioning

confidence: 99%

Idaho National Laboratory Lead or Lead-Bismuth Eutectic (LBE) Test Facility - R&D Requirements, Design Criteria, Design Concept, and Concept Guidance

Loewen¹,

Demkowicz²

2005

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SUMMARYThe Idaho National Laboratory Lead-Bismuth Eutectic Test Facility will advance the state of nuclear technology relative to heavy-metal coolants (primarily Pb and Pb-Bi), thereby allowing the U.S. to maintain the pre-eminent position in overseas markets and a future domestic market. The end results will be a better qualitative understanding and quantitative measure of the thermal physics and chemistry conditions in the molten metal systems for varied flow conditions (single and multiphase), flow regime transitions, heat input methods, pumping requirements for varied conditions and geometries, and corrosion performance. Furthering INL knowledge in these areas is crucial to sustaining a competitive global position.This fundamental heavy-metal research supports the National Energy Policy Development Group's stated need for energy systems to support electrical generation.

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“…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.…”

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confidence: 99%

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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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The next generation of fast reactors

Cited by 45 publications

References 0 publications

Physics of Americium Transmutation

Physics of Americium Transmutation

Idaho National Laboratory Lead or Lead-Bismuth Eutectic (LBE) Test Facility - R&D Requirements, Design Criteria, Design Concept, and Concept Guidance

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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