Compatibility of FBR materials with sodium

Furukawa, Tomohiro; Kato, Shoichi; Yoshida, Eiichi

doi:10.1016/j.jnucmat.2009.03.003

Cited by 53 publications

(26 citation statements)

References 8 publications

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“…2) is similar to the distribution of the rate of corrosion of the walls of such a channel (Fig. 3) of steel 316 obtained in [5,6] and fi ts the data of [3,4], in accordance with which the rate of formation of deposits in the cold region of a sodium loop comprises 30-100% of the rate of corrosion of its walls in the hot region.…”

supporting

confidence: 87%

Mass Transfer of Corrosion Products in the Nonisothermal Sodium Loop of a Fast Reactor

Varseev

Alekseev

2014

J Eng Phys Thermophy

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The mass transfer of the products of corrosion of the steel surface of the sodium loop of a fast nuclear power reactor was investigated for the purpose of optimization of its parameters. The problem of deposition of the corrosion products on the surface of the heat-exchange unit of the indicated loop was considered. Experimental data on the rate of accumulation of deposits in the channel of this unit and results of the dispersion analysis of the suspensions contained in the sodium coolant are presented.Introduction. The mass transfer in nuclear power reactors with liquid-metal coolants represents a serious problem in the development and operation of these reactors because the mass transfer processes in them can have a number of negative consequences, among which is the corrosion of the material of the cooling loop of a reactor. As a result of the corrosion of the walls of this loop, they become thinner, which can lead to leakage of the coolant from the loop. The deposits formed on the cooled surface of the heat-exchange unit of the coolant loop reduces the heat transfer in it because of narrowing and blocking of its channel. Moreover, the corrosion products are activated when they pass through the active zone of the reactor, which increases the radiation level of its cold surface. All these factors adversely affect the reability and safety of the reactor [1].In this connection, the aim of the present work is to determine the rate of mass transfer of the products of corrosion of the metal walls of the sodium loop of a fast nuclear power reactor to the liquid-metal coolant in this loop and the rate of crystallization of the corrosion products in the cold zone of the loop by experimental methods and with the use of computational models.Experimental Procedure and Experimental Setup. The experiments were carried out in four separate runs in an nonisothermal sodium loop, namely, in its high-temperature section including a heat and mass transfer (HMTT) unit, a heater, a source of impurities, and a fi lter of suspensions. A diagram of the experimental setup is shown in Fig. 1.The HMTT unit represents a tube-in-tube heat exchanger with thermocouples positioned on its outer surface and hot-coolant and cold-coolant counterfl ows in the tubes of the unit. In the HMTT unit, the hot sodium was cooled by 250 o C and, in so doing, supersaturated with impurities that were accumulated on the walls of the inner HMTT.The impurity source is a stainless-steel vessel fi lled with Ch18N10T-steel shaves, the area of which was determined by the ratio between the surface of a small part of them to their mass. Samples of different materials were positioned at the inlet and outlet of this source.The fi lter of impurities represents a stainless-steel cylinder with a cylindrical orifi ce inserted into it. The inner tube is covered with several layers of a steel mesh of thickness ~100 μm. The design of the fi lter allows it to collect both the impurities accumulated on the mesh and the impurities separated from the mesh in the process of fi ltr...

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supporting

confidence: 87%

Mass Transfer of Corrosion Products in the Nonisothermal Sodium Loop of a Fast Reactor

Varseev

Alekseev

2014

J Eng Phys Thermophy

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“…In the present investigation, the reduction in the slope under flowing sodium environment with increasing temperature is less significant compared to that under air environment. As, flowing sodium environment has little effect on the creep behaviour of 316 stainless steel [25], the influence of creep in reducing fatigue life is the same in both environments. Oxidation plays a major role in reducing the fatigue life under air environment.…”

Section: Resultsmentioning

confidence: 91%

Evaluation of the Effect of Dynamic Sodium on the Low Cycle Fatigue Properties of 316L(N) Stainless Steel Base and Weld Joints

Ganesan

Kannan

Mariappan

et al. 2012

High Temperature Materials and Processes

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Low cycle fatigue (LCF) tests on 316L(N) austenitic stainless steel base and weld joints were at 823 K and 873 K at a constant strain rate of 3 10 3 s 1 with strain ranges varying from˙0:4% to˙1:0% in a servohydraulic fatigue test system under flowing sodium environment. The cyclic stress response exhibited a similar trend as that in air comprising of an initial rapid hardening, followed by a slight softening stage before saturation. The fatigue lives are significantly improved in sodium environment when compared to identical testing conditions in air environment. The lack of oxidation in sodium environment is attributed to the delayed crack initiation, reduced crack propagation rate and consequent increase in fatigue life. Comparison of the data evaluated in sodium with RCC-MR design code, derived on the basis of data obtained from air shows that the design based on air tests is conservative.

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“…From this experimental fact, it follows that the effect of high-temperature neutron irradiation would not be a main cause. The high-temperature flowing sodium corrosion tests of ferritic steels, including ODS ferritic steels, produced mass transfer of elements evinced in such behaviors as surface Cr depletion, Ni enrichment, and decarburization; however, it led to neither the peculiar microstructure change nor indications of its occurrence [15,18,19]. Therefore, the mass transfer produced by sodium flow also would not be a main cause.…”

Section: Selection Of Key Factor Leading To the Peculiar Irradiation mentioning

confidence: 99%

Investigation of the cause of peculiar irradiation behavior of 9Cr-ODS steel in BOR-60 irradiation tests

Ohtsuka

Kaito

Yano

et al. 2013

Journal of Nuclear Science and Technology

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Four experimental fuel assemblies (EFAs) containing 9Cr-ODS steel cladding fuel pins were previously irradiated in the BOR-60 to demonstrate the in-reactor performance of 9Cr-ODS steel for use as fuel cladding tubes. One of the EFAs achieved the best data, a peak burn-up of 11.9at% and a neutron dose of 51 dpa, without any microstructure instability or any fuel pin rupture. On the other hand, in another EFA (peak burn-up, 10.5at%; peak neutron dose, 44 dpa), peculiar irradiation behaviors, such as microstructure instability and fuel pin rupture, occurred. Investigations of the cause of these peculiar irradiation behaviors were carried out. The detection sensitivity in an ultrasonic inspection test was shown to be low for the metallic Cr and metallic Fe inclusions. The peculiar microstructure change reappeared with high-temperature thermal-aging of the 9Cr-ODS steel containing metallic Cr inclusions. The strength and ductility of the defective part containing metallic Cr inclusions were appreciably lower than those of a standard part without the inclusions. The combined effects of matrix Cr heterogeneity (presence of metallic Cr inclusions) and high-temperature irradiation were concluded to be the main cause of the peculiar microstructure change in 9Cr-ODS steel cladding tubes in the BOR-60 irradiation tests. They contributed to the fuel pin rupture.

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Compatibility of FBR materials with sodium

Cited by 53 publications

References 8 publications

Mass Transfer of Corrosion Products in the Nonisothermal Sodium Loop of a Fast Reactor

Mass Transfer of Corrosion Products in the Nonisothermal Sodium Loop of a Fast Reactor

Evaluation of the Effect of Dynamic Sodium on the Low Cycle Fatigue Properties of 316L(N) Stainless Steel Base and Weld Joints

Investigation of the cause of peculiar irradiation behavior of 9Cr-ODS steel in BOR-60 irradiation tests

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