John Paul Foster

VOID SWELLING AND IRRADIATION CREEP RELATIONSHIPS

Abstract Recent analytical and theoretical work on swelling enhanced irradiation creep and stress effects on swelling is reviewed. A proposed explanation for swelling enhanced irradiation creep involves consideration of the role of vacancy loops. Theoretical work leads to the development of a new relationship for swelling enhanced creep which predicts larger irradiation creep rates at high levels of swelling (>5%) than the original formulation. Consideration is given to an additional effect of stress on swelling which involves a stress effect on the incubation dose. A constitutive equation is presented to describe this phenomenon. Design related illustrations are presented for these high fluence irradiation induced phenomena.

316 stainless steel cavity swelling in a PWR

Abstract Swelling was measured on a cold-worked (CW) 316 stainless steel (SS) flux-thimble tube irradiated in a Pressurized Water Reactor (PWR) plant. The measured swelling was approximately 0.03% (in the range of 0.01 to 0.09%). The sample was irradiated at a temperature between 305 and 315°C to a calculated dose of about 35 dpa. Comparison with High Flux Isotope Reactor (HFIR) swelling data suggests that the observed cavity swelling is due to helium bubbles. The PWR datum was extrapolated to the approximate end-of-life reactor internals component dose of 100 dpa using HFIR swelling data. HFIR CW 316 SS swelling data indicate that swelling will be less than 3% at 100 dpa in the temperature range between 300°C and 440°C. Garner [1,2] has proposed that large amounts of swelling (> 10%) and severe swelling-embrittlement could possibly occur in PWR reactor internals during post shutdown handling. Since the PWR and HFIR end-of-life extrapolated swelling values are probably less than 3%, the proposed swelling and swelling-embrittlement is not considered to apply to PWR reactor internals fabricated with CW 316 SS and operating at temperatures between 300°C and 440°C.

Influence of final recrystallization heat treatment on Zircaloy-4 strip corrosion

Abstract Zircaloy-4 strip material was processed to the final size using standard fabrication techniques. The final size material was subjected to many different recrystallization heat treatments. The temperatures covered a range from 550 to 775°C, and post-beta-quench thermal processing, represented by the cumulative A parameter, ranged from 1.4 × 10−20 to 2.9 × 10−17 h. The uniform and nodular corrosion behavior was characterized using 400 and 500 °C autoclave tests. The results show that maximum uniform and nodular corrosion resistance may be obtained by control of the final recrystallization anneal. Maximum uniform and nodular corrosion resistance was obtained when the cumulativeA parameter was in the range of 4 × 10−19 to 7 × 10−18 h. Selected samples were examined using transmission electron microscopy. A relationship was observed between the mean diameter of the second phase particles and the corrosion behavior. The diameter increased with increasing cumulative A parameter. The particle structure changed from fcc to hexagonal as a result of alpha thermal processing after beta-quenching and may be related to the onset of uniform and nodular corrosion.

Residual stress behavior in fast neutron irradiated SA AISI 304l stainless steel cylindrical tubing

Abstract Residual stress measurements were made on solution-annealed (SA) AISI 304L stainless steel (SS) irradiated in EBR-II over a temperature range from 402 to 524°C by axially slitting short sections of tubing. The data were analyzed by using SA AISI 304 SS physical properties and SA AISI 304L SS swelling and irradiation creep empirical equations to calculate the slit width change (δ) versus fluence (φt) curve. At temperatures equal to and above 445°C, δ versus φt calculations indicate that the stress effect on swelling is sufficiently large to reduce the swelling rate temperature gradient, and consequently the on-power stress gradient, to zero. This behavior is confirmed by void volume gradient measurements. At lower temperatures, δ versus φt calculations indicate that stress affected swelling is smaller and does not relax the swelling rate temperature gradient. Void volume gradient measurements confirm the presence of a swelling gradient. Calculations of the δ versus φt curve were made with four different empirical swelling equation fluence dependencies, and the best agreement with the δ versus φt data was obtained with a power form type swelling equation. The equations fit the immersion density data ( ΔV V 0 versus φt ) within experimental scatter, but predict significantly different δ versus φt behavior. These results show that the slit tube results are very sensitive to the empirical swelling equation form.

Residual stress measurements in irradiated solution-annealed type 304 stainless steel tubing

Abstract Immersion density and residual stress measurements were made on solution-annealed type 304 stainless steel capsule tubing irradiated up to fluence levels of 9 × 10 22 n/cm 2 ( E > 0.1 MeV). The measured residual stress is dependent on the competition between differential swelling which builds up differential stresses, and irradiation creep which relaxes these stresses. The measurements were analyzed using a bilinear swelling equation formulated with swelling data from the same heat of material. The temperatures and fluence levels of the swelling and slit tube data were each calculated with the same computer code. At high fluence, when swelling was in the steady-state region, the effective irradiation creep rate increased by a factor of about three. Further analysis was made assuming that stress-enhanced swelling and swelling-enhanced irradiation creep were the enhanced relaxation mechanisms.

Relationship between in-reactor stress relaxation and irradiation creep

Stress relaxation and irradiation creep data are available for 20% CW 316 SS samples fabricated from the same heat of material. The stress relaxation and irradiation creep tests were both performed in bending. The stress relaxation was calculated using an irradiation creep correlation formulated using the irradiation creep data. The calculations were in excellent agreement with the measurements and show that stress relaxation may be calculated using irradiation creep data when the proper methods are used. The methods for this calculation include the use of a transient exponential decay coefficient with a dose rate consistent with the stress relaxation application, the use of irradiation creep coefficients derived from tests with material consistent with the stress relaxation application and the use of irradiation creep data with the same stress state.

Correlation of irradiation creep data obtained in fast and thermal neutron spectra with displacement cross-sections

Abstract Irradiation creep data are available for 20% cold-worked M316 stainless steel, 20% cold-worked FV548 stainless steel, solution-annealed 304 stainless steel and aged Nimonic PE16 for the same heats of material and state of stress tested in thermal and fast neutron reactors. The transient and steady state irradiation creep coefficients were calculated using spectrum-averaged displacement cross-section dose units (i.e. displacements per atom). The transient irradiation creep coefficients appear to be subject to considerable scatter. The steady state irradiation creep coefficients were found to be a consistent factor of 2.5 larger in thermal than fast reactors. These results are in excellent agreement with the fast reactor dose rate dependence of the steady state irradiation creep rate observed by Lewthwaite and Mosedale. However, the dose rate effect-directly depends on the displacement cross-section, which is a calculated dependent parameter. The available information suggests that the displacement cross-section function uncertainty is the major effect responsible for the 2.5 factor in irradiation creep coefficients.

Temperature dependence of the 20% cold worked 316 stainless steel steady state irradiation creep rate

The irradiation creep data from four completed tests have been analysed to show that the steady state irradiation creep rate exhibits a moderate and complex temperature dependence. The irradiation creep tests were performed in the Experimental Breeder Reactor Number II (EBR-II) using beams and pressurized tubes, and in the Oak Ridge Reactor (ORR) and the High Flux Isotope Reactor (HFIR) using pressurized tubes. The data cover the temperature range from 200°C to 585°C, and show that from 200°C to 330°C, the steady state rate increases moderately with increasing temperature. At about 330°C, the steady state rate peaks and rapidly decreases with increasing temperature from 330°C to 370°C. From 370°C to 585°C the steady state rate moderately increases with increasing temperature.

ZIRLO TM Cladding Improvement

Improvements to fuel design have been made in recent years to meet the challenges of increases in fuel duty in terms of linear power and operating temperature. Improved cladding material is one of these design improvements. Specifically, Westinghouse has developed an improved version of ZIRLOTM called Optimized ZIRLO and denoted as OPT ZIRLO. The Sn level in this improved material is reduced from the nominal standard previous level of about 1 % to a range of 0.6 % to 0.8 %. The reduced Sn level has been optimized to produce a higher corrosion resistance and provide adequate in-reactor creep resistance. Out-reactor diameter creep tests have shown that decreasing Sn increases out-reactor creep, suggesting that decreasing Sn may also increase in-reactor creep. An in-reactor testing program in the Vogtle Unit 2 pressurized water reactor (PWR) was performed to confirm the predicted in-reactor behavior. The test samples were suspended as segmented rods inside fuel assembly thimble tubes. In-reactor diameter creep data confirmed that decreasing Sn increases in-reactor creep. As a result of the correlation between in-reactor and out-reactor creep, an extensive out-reactor diameter creep program was performed in order to develop methods to fabricate OPT ZIRLO with the same in-reactor creep properties as the currently used stress-relief annealed standard ZIRLO (denoted as SRA STD ZIRLO). The level of in-reactor diameter creep of SRA STD ZIRLO was achieved for OPT ZIRLO by two methods. One method involved changing the final microstructure from SRA to partially recrystallized (PRXA). The other method kept the final microstructure as SRA and changed the tube reduction sequence to decrease the final tube area reduction. In order to develop these methods, a series of material variation tests was performed. Some of the material variations included different final heat treatments, different tube reduction sequences, and different pre-charged hydrogen levels. These tests were performed for both out-reactor and in-reactor. In addition, the out-reactor and in-reactor creep were observed to directly correlate for OPT ZIRLO material fabricated with different final anneal temperatures. Thus, out-reactor creep may be used to predict in-reactor creep properties for different final anneal temperatures. These results show that fabrication changes may be used to control in-reactor creep. In this study, fabrication changes were used to compensate for the reduction in in-reactor creep strength associated with lower tin content in OPT ZIRLO.

Dependence of the non-swelling in-reactor steady-state creep component of austenitic phase alloys on the stacking fault energy

Steady-state irradiation creep data and stacking fault energy data are available for a wide composition range of austenitic phase alloys. The steady-state irradiation creep rate was found to increase with increasing stacking fault energy.