FR2538111A1 - METHOD FOR DETERMINING RESISTANCE RELATING TO CORROSION OF A ZIRCONIUM ALLOY - Google Patents

Abstract

IT CONSISTS OF: A. SUBMITTING A TEST OF A ZIRCONIUM ALLOY TO A WATER VAPOR ATMOSPHERE AT A TEMPERATURE OF 300 TO 420C FOR A PERIOD OF AT LEAST 5 HOURS; B. THEN SUBMIT THE SPECIMEN TO A WATER VAPOR ATMOSPHERE AT A TEMPERATURE OF 490 TO 520C FOR A PERIOD OF AT LEAST 12 HOURS; AND C. NOTE ANY FORMATION OF CORROSION ON THE WATER VAPOR SPECIMEN. APPLICATION TO ELEMENTS AND STRUCTURES OF WATER-COOLED NUCLEAR REACTORS.

Description

Zirconium alloys are widely used in components and

  core structures of water-cooled nuclear fission reactors because of their low

  neutron capture cross-section, among other things

  Suitable properties for such a service are, for example, US Pat. No. 4,212,686. Various zirconium alloy grades have been developed and placed on the market.

  mainly for applications to nuclear reactors

  Typical zirconium alloy compositions are the commercial alloys designated Zircaloy-2 and

  Zircaloy-4, including the alloys described in the

  U.S. Patent Nos. 2,772,964 and 3,148,056 disclose a zirconium alloy containing niobium for

in reactors.

  Zircaloys comprise alloy compositions containing at least about 95% by weight of zirconium and, in percent by weight, about 2.0% tin, up to about 0.5% iron, up to about 0.5% by weight. about 0.5% chromium

  and from 0 to about 0.15% nickel.

  The degree of sensitivity to corrosion of a material

  is a critical factor for its use or its

  performance in a water-cooled reactor

  of a reactor, the alloys of zirconium

  normally a superficial, dark and

  harmless, evenly distributed on the surface This so-called black oxide provides protection of the underlying metal and slowly thickens with increasing

  In the reactor, however, the zirconium alloys

  nium can also develop corrosion nodules

  harmful, sometimes referred to as "pustular corrosion".

  This nodular type of corrosion increases rapidly in size or area and depth on the surface of the alloy, which, under certain conditions, can adversely affect the integrity of the alloy.

  white oxide that can grow from several orders of magnitude

  faster than the black superficial oxide

  sif to produce a thick layer of white oxide

  singing among other things heat exchange.

  It has been found that the degree of nodular corrosion sensitivity of zirconium alloys subjected to the environment of a water-cooled reactor is dependent upon several or a combination of factors, including in particular the alloy composition, microstructure

  as well as the temperatures of the reactor in operation.

  For example, reference may be made to U UA patents.

  Nos. 3,150,972, 3,261,682 and 4,212,686.

  As indicated in US Pat. No. 4,238,251, there is an obvious correlation between the characteristics

  of the microstructure of a zirconium alloy composition

  nium and corrosion resistance of nodular type

in a reactor environment.

  It has been proposed in the art to manipulate the microstructure of zirconium alloys by annealing to improve the corrosion resistance and other essential properties of these alloys for use in reactors.

  Nos. 2,736,651, 2,894,866 and 3,884,728, for example, teaching

  the reformation of the microstructure of certain alloys

  of zirconium to increase their structural strength

  and their resistance to corrosion in service in reaction

  tors. However, it has been found that the microstructure of zirconium alloys, and in turn their susceptibility to corrosion, often vary.

  incomplete, and work hardening or manufacturing operations

  such as reduction, stretching, shaping or

  cutting, and welding can result in microstructures

  different or nonuniform tures in an alloy.

  As a result, there may be considerable latitude or uncertainty about the degree of sensitivity to corrosion

  for reactor components such as fuel ducts

  tible, the fuel channels that are made of a

zirconium alloy.

  The present invention relates to a method for determining the relative resistance to nodular corrosion of a zirconium alloy in the environment

  a water-cooled nuclear fission reactor.

  The method of distinction consists in submitting a test

  zirconium alloy to a high pressure water vapor atmosphere applied in a sequence of

  increasing temperatures, and consists of evaluating any changes

  appearing in the weight or surface appearance

  The invention is capable of establishing a

  the sensitivity of zirconium alloys to nodular corrosion in terms of weight gain by

  corrosion and also visual aspects.

  The present invention therefore aims:

  to provide means to determine the resistance

  relative corrosion rate of zirconium alloys; to distinguish the corrosion sensitivity of zirconium alloys for their use in

  water cooled nuclear fission reactors.

  The rest of the description refers to the drawing

  annexed which represents a curve of a correlation between the corrosion appearing in a reactor and that produced

by the process of the invention.

  According to the invention, it is possible to determine the sensitivity

  corrosion of zirconium alloys in the envi-

  reactor by subjecting a specimen of the alloy to high-pressure water vapor applied successively at two temperature levels, and then

  to evaluate any resulting physical changes.

  In the practice of the invention, a specimen of zirconium alloy or a suitable sample thereof is cleaned of any dirt and foreign matter, and its weight is precisely determined.

  cleaning can be done by conventional means such as

  taking an acid bath or "stripping", followed by rinsing in water. The test piece of alloy is then subjected to

  water vapor in an autoclave at a pressure of

  meter from about 70.3 kg / cm 2 to 105.5 kg / cm 2 The temperature of the vapor applied is brought to and maintained at an initial level of 300 to 420 ° C for a period of at least 5 hours, and then increased to a next level of from 490 to 5200 C for a period of at least 12 hours. Specific periods of application

  effective water vapor include, after warming

  temperature, 5 to 15 hours at the initial level.

  tial temperature of 300 to 420 'C for the initial phase, and 12 to 30 hours for the temperature level of 490 to

5200 C.

  A preferred embodiment for the implementation of the present invention comprises an initial temperature of the steam of the order of 410 ° C for a period of 8 to 10 hours followed by a subsequent temperature of the steam of the order from 5100 C for a duration of 16 to

24 hours.

  After removing the autoclave and cooling

  at ambient conditions, the test piece of

  binding treated with water vapor and any increase in

  The test specimen can also be examined.

  visually to ensure corrosion

nodular on its surface.

  An increase in the weight of the specimen

  The process of the above-mentioned method substantially greater than about 300 to 400 mg / dm 2 of surface area indicates that a zirconium alloy such as Zircaloy-2 may be sensitive to nodular corrosion. Figure 1 illustrates this point by comparing the resistance. Zircaloy-2 tube corrosion in the laboratory and in a reactor The figure shows a correlation between nodular corrosion in the reactor and the weight gains resulting from the laboratory steam test method according to the invention. (carried out at 410 ° C. and 5100 ° C.) with Zircaloy-2 bars.

  any nodular corrosion formation attributed to

  the above process and covering more than 20 to 30% of the total area of the test

  the alloy may be susceptible to corro-

nodular growth.

  A specimen was cut into a tubular container for nuclear fuel made of an alloy

  zirconium, and it has been debarked and cleaned.

  worm, if any, surface oxide using abrasive paper Cleaning includes etching in an acid solution containing, for example, 2.5 to 5.0% by volume of concentrated hydrofluoric acid (HF ), 45% by volume of concentrated nitric acid (HNO 3), the balance being

distilled water.

  After the attack, the test tube is washed,

dry and weigh to the nearest 0.2 mg.

  The test piece thus prepared is then suspended in an autoclave, steam is applied and the system is brought to equilibrium at 410 ° C. and at a pressure of

  , 5 kg / cm 2 This temperature balance is maintained

  pressure of the steam atmosphere for about 8 hours

  for the initial phase, after which the temperature is increased

for the next phase.

  When a temperature of 510 ° C. is reached, the system is again brought to equilibrium and maintained at 5100 ° C. and at a pressure of approximately 105.5 kg / cm 2

  about 16 hours during this subsequent phase.

  At the end of the steam treatment periods at the two temperature levels, the autoclave is returned to ambient conditions, the specimen is removed, dried and weighed and examined visually. of the test piece, and can be examined visually for the existence of nodular corrosion.

Claims (9)

  1 Method for determining the relative resistance to corrosion of a zirconium alloy in the environment of a water-cooled nuclear fission reactor, characterized in that it consists in: a) subjecting a specimen of an alloy of   zirconium to an atmosphere of water vapor at a temperature of   300 to 420 ° C for a period of at least 5 hours;   b) then subject the test piece to an atmosphere of   of water vapor at a temperature of 490 to 520 ° C   for a period of not less than 12 hours; and (c) note any corrosion formation on   the test piece subjected to water vapor.   Process according to Claim 1, characterized in that the test piece is subjected to the vapor atmosphere   of water from step a) for a period of 5 to 15 hours.   3 Process according to claim 1, characterized in that the test piece is subjected to the vapor atmosphere   of water from step b) for a period of 12 to 30 hours.   Process according to claim 1, characterized in that the specimen is subjected to a steam atmosphere in an autoclave at a pressure of 70.3 to 5 kg / cm 2 and a temperature of 300 to 420 ° C. C for a period of 5 to 10 hours, then at a temperature of   490 to 5200 C for a period of 16 to 24 hours.   A method for determining the relative corrosion resistance of a zirconium alloy in the environment of a water-cooled nuclear fission reactor, characterized in that it comprises subjecting a specimen of a zirconium alloy to a steam atmosphere in an autoclave and at a pressure of about 3 to 105.5 kg / cm 2 for a period of 5 to hours at a temperature of 300 to 4200 C, and then for a period of 12 to 30 hours at a temperature of 490 to 5200 C, then note any increase in weight   the specimen subjected to water vapor.   Process according to Claim 5, characterized in that the test piece is subjected to steam at a temperature of 300 to 420 ° C for a period of about 5 to 10 hours and then at a temperature of 490 to 520 ° C. 'C.   during a period of 16 to 24 hours.   Process according to Claim 5, characterized in that the test piece is subjected to water vapor at   a temperature of 400 ° C for a period of 5 to 15 hours   and then at a temperature of about 500 ° C a period of 12 to 30 hours.   8 Method for determining the relative corrosion resistance of a zirconium alloy in the environment of a nuclear fission reactor core   water cooled, characterized in that it consists of   put a sample of a zirconium alloy in a steam atmosphere in an autoclave according to the sequence of conditions below:   (a) an atmosphere of water vapor at a temperature   at 300 to 420 ° C at a pressure of 70.3 to 105.5 kg / cm 2 for a period of 5 to 10 hours; b) thereafter a steam atmosphere at a temperature of 490-520 ° C at a pressure of 70.3-5.5 kg / cm 2 for a period of about 16-24 hours; (c) finally to note any increase in weight of   the test piece subjected to water vapor.   9 Process according to claim 8, characterized in that the test piece is subjected to a steam atmosphere at about 4000 C for a period of 8 to 10 hours and then to a water vapor at 5000 C approximately   during a period of 16 to 24 hours.   Process according to Claim 8, characterized in that the test piece is a zirconium alloy composed of at least 95% by weight of zirconium and comprising, in percent by weight, up to 0.2% tin, up to at about 0.5%   iron, up to 0.5% chromium and from 0 to 0.15% nickel.