Neutron Irradiation Effects on the Corrosion of Zircaloy-4 in a PWR Environment

Transcription

1 Neutron Irradiation Effects on the Corrosion of Zircaloy-4 in a PWR Environment B.F. Kammenzind, J.A. Gruber, R. Bajaj, J.D. Smee Bechtel Marine Propulsion Corporation Bettis Laboratory Knolls Laboratory 18 th International Symposium on Zirconium in the Nuclear Industry Hilton Head, South Carolina May 15 19, 2016

2 Background The surface of zirconium alloy clad fuel elements corrode as a result of its exposure to high temperature, high pressure water in a light water reactor environment. The corrosion film is adherent to the cladding surface creating a thermally insulating layer The corrosion film is brittle, removing structural thickness from the cladding. Hydrogen absorbed into the base metal as a result of the corrosion reaction degrades the mechanical properties of the underlying base metal as well as potentially causing deformation. The amount of cladding corrosion that occurs can be life limiting requiring reliable predictive methods.

3 Background Studies into the corrosion of zirconium base alloys in PWR and BWR primary system environments have been ongoing since the beginning of the nuclear industry. Producing an engineering understanding of the corrosion kinetics of a number of zirconium alloys A general understanding of the basic mechanisms occurring in the formation of the adherent corrosion film on the surface of the zirconium alloys But detailed mechanistic understanding of how alloying elements improve or degrade the corrosion resistance of the alloy, or how irradiation in a PWR environment accelerates the corrosion rate of zirconium alloys still elude the industry.

4 Background Zirconium alloy corrosion This paper describes testing performed in the Advanced Test Reactor (ATR) at the Idaho National Engineering Lab (INEL) using non-fueled corrosion test coupons to examine two potential mechanisms for the in-reactor corrosion acceleration of Zircaloy-4: irradiation damage to the base metal irradiation damage to the developing oxide

5 Experimental Method Material Corrosion test specimens manufactured from rolled Zircaloy-4 sheet. Material chemistry is near the mean values specified in the Zircaloy-4 ASTM Specification. All final processing steps for the sheet materials were done in the alpha phase. The final plate received a recrystallization anneal following rolling The final annealing parameter A was approximately 1x10-16 A= t i exp (-40,000/T i ) Final texture is representative of hot rolled and alpha annealed plate product. Test specimens are machined from the final plate product

6 Experimental Method Exposure Conditions Exposed in flowing pressurized water test loops in the ATR PWR Water chemistry de-ionized water, degassed with an added hydrogen concentration of cc H 2 /Kg room temperature ph of from 10.0 to Water temperature in the loops controlled by heaters external to the reactor 270C, 310C, 330C, or 354C pressure maintained at MPa (2000 PSI) Placement of the specimens in the core determines the neutron flux to which they are exposed. Out-of-flux to ~1x10 14 neutrons/cm 2 (E>1 Mev) Examinations Periodic interim Visual, Weight gain Select end-of-life Metallography SEM, TEM

7 Irradiation Enhancement in Corrosion Rates Rate acceleration can be ~ 40 times out-of-reactor rates at 270C. Alpha-annealed Zircaloy-4 coupons exposed in-reactor at 270 C. Constant fast neutron flux is indicated in legend in units of n/cm 2 /s. Increasing flux

8 Irradiation Enhancement in Corrosion Rates Rate acceleration can be ~ 30 times out-of-reactor rates at 310C. Alpha-annealed Zircaloy-4 coupons exposed in-reactor at 310 C. Constant fast neutron flux is indicated in legend in units of n/cm 2 /s. Increasing flux In reactor Posttransition corrosion rate acceleration is a decreasing function of exposure temperature

9 Zirconium alloy corrosion (in-reactor) What is the cause of in-reactor corrosion rate acceleration in these alloys? Many discussed mechanisms are not specifically a result of irradiation. Lithium concentration due to heat flux and boiling High hydrogen concentrations at the corroding surface Neither is a factor in these ATR data Some proposed mechanisms are due to irradiation: Bulk radiolysis of the coolant Thick film hypothesis (heterogeneous radiolysis) Radiation damage to the oxide film Radiation damage to the base Zircaloy metal Additional Separate Effects ATR corrosion coupon tests are helping to further elucidate the accelerating mechanisms Likely more than one factor, but potentially several acting in synergy.

10 Irradiation Damage to the Passivating Oxide Layer Pre-transition corrosion behavior in-reactor is similar to that in an autoclave, irrespective of flux exposure. This sheds doubt on mechanisms that affect diffusion kinetics through the passivating boundary layer directly. Irradiation assisted diffusion Irradiation enhanced conductivity 40 alpha-annealed Zircaloy-4 coupons exposed in-reactor at 270 C. Weight gain (points) and fast neutron flux (lines) as indicated. Autoclave curve in blue.

11 Irradiation Damage to the Zircaloy Metal Previous published data have shown that effects of in-reactor irradiation are carried forward in post irradiation autoclave testing H. Stehle et al. in ASTM STP 824 (1984) In-reactor specimens transferred to 280C autoclaves with oxide intact. Specimens initially corrode in autoclave at in-reactor rates. Over time, rate slows down to within a factor of approximately two of normal autoclave rate. B. Cheng et al. in ASTM STP 1245, (1994) In-Reactor exposed structural materials transferred to autoclave at 316C water most without oxide. Most specimens showed only a small acceleration over nonirradiated specimens. Two weld specimens showed large and sustained acceleration. Memory attributed to precipitate amorphization in the metal.

12 Irradiation Damage to the Zircaloy Metal Like the Adamson/ Cheng work. ATR coupons were removed from test after in-reactor exposure, oxide films removed, surfaces re-pickled, but samples were put back in-reactor Weight gain after oxide removal only Shifted to reflect prior weight gain and exposure time Corrosion weight gain for 310 C alpha-annealed Zircaloy-4 coupons after oxide removal. Prior exposure as shown. Weight gain after reinsertion (left) and shifted to reflect prior weight gain and exposure time (right). Corrosion kinetics proceed upon reinsertion as if original oxide film was still in-place

13 Irradiation Damage to the Zircaloy Metal Similar oxide removal experiment at 270C Weight gain after oxide removal only Shifted to reflect prior weight gain and exposure time Corrosion weight gain for 270 C alpha-annealed Zircaloy-4 coupons after oxide removal. Prior exposure as shown. Weight gain after reinsertion (left) and shifted to reflect prior weight gain and exposure time (right). Irradiation Damage to the base metal is a dominant factor in the in-reactor corrosion acceleration of Zircaloy-4. Occurs at Fluences associated with second phase particle degradation (Not point defect and loop formation) Mechanism acts synergistically with irradiation, as the accelerations are much greater than observed in post irradiation autoclave testing. Acceleration appears to saturate near film thicknesses of several microns

14 Going in-reactor with a post transition, porous outer layer in the corrosion film leads to an immediate in-reactor acceleration. Acceleration is not a strong function of film thickness, once greater than ~ 4-5 microns Acceleration does increase with increasing fluence, independent of film thickness (once greater than ~ 4-5 ) Irradiation Acceleration with Post Transition Films Corrosion weight gain for alpha-annealed Zircaloy-4 coupons at 330 C. Average behavior for each group is shown as solid lines. Corrosion rate as a function of fast neutron fluence for alpha-annealed Zircaloy-4 coupons at 330 C. Average behavior for each group is shown as solid lines. Suggests synergy between processes occurring in the thicker porous corrosion films and Irradiation assisted dissolution of the second phase particle.

15 Further environmental shifting experiments (downshift from 350 C to 270 C) Presence of thick film alone does not produce large in-reactor acceleration at 270 C. Appears to be synergy between cumulative irradiation effects at lower temperature and thick film Irradiation Acceleration with Post Transition Films Const. 350 C Const. 350 C Const. 270 C Const. 270 C Alpha-annealed Zircaloy-4 corrosion coupons showing effect of post-transition temperature downshift from 350 C to 270 C. Fast neutron flux is shown as dashed lines.

16 Microstructural Evidence for the Effect of Irradiation Damaged Precipitates on the Corrosion Process Paper describes microstructural examinations of out of reactor and inreactor grown corrosion films for evidence of the in-reactor acceleration mechanisms Both FEG SEM and ATEM examinations of corrosion film cross sections Not described in the presentation due to lack of time In most aspects, the corrosion films forming in the autoclave environment were found to be very similar to those forming in the radiation environment and similar to characteristics already reported in the literature The evolutionary change of the second phase particles was one of the more obvious differences between films that grew in an autoclave environment vs those grown in a radiation environment Mechanistically, not obvious from the FEG SEM and ATEM how this results in the large increase in post-transition kinetics that occurs under irradiation at high fluences But Interconnected porosity appeared closer to the O/M interface in the in-reactor grown films.

17 Conclusions The long time post-transition corrosion rate of Zircaloy-4 is significantly accelerated in a PWR radiation environment The radiation environment has little effect on the pretransition corrosion rates even after extended time in the environment. Pre-irradiation of the Zircaloy-4 metal to high fluences prior to corrosion testing, significantly decreases subsequent in-reactor exposure time to achieve the high in-reactor post-transition corrosion rate accelerations The maximum sustained corrosion rate acceleration of highly irradiated Zircaloy-4 seen in-reactor (30 to 40x) is much greater than the sustained corrosion rate acceleration seen with highly irradiated Zircaloy-4 tested out-ofreactor (~2x) Indicating the radiation environment is acting synergistically with the corrosion film formed from the neutron damaged metal to enhance the corrosion rates in-reactor. The presence of thicker prefilms on non-irradiated material, in and of itself, accelerates the in-reactor corrosion rate of Zircaloy Suggesting interactions of the fossil film with the radiation environment alone may accelerate Zircaloy corrosion rates in-reactor. It is suggested in the film microstructure that the water environment does have closer access to the oxide metal interface in the in-reactor grown corrosion films Not obvious from the microstructural examinations how the incorporation of the irradiation damaged Zircaloy precipitates into the corrosion film might be facilitating that. Has led us to consider in more detail heterogeneous radiolysis effects within the oxide along with the semiconducting nature of the zirconium oxide film and potential photo dissolution as a mechanism.