Irradiation-Assisted Stress Corrosion Cracking (IASCC)
SIB-96-132, Rev. 1
Introduction
Operating experience and laboratory data show that highly neutron irradiated stainless steel components may develop intergranular stress corrosion cracks in the light water reactor (LWR) water environment. In addition, high levels of neutron irradiation modify the mechanical properties of these components making them less flaw tolerant. Limited periodic inspections of reactor internal components and fracture mechanics evaluations using estimated fluence levels and estimated mechanical properties, have been employed in an attempt to keep the problem in perspective. Additional operating experience and materials data continue to be obtained that assist in the understanding of IASCC and its effects on the structural integrity of reactor components.
Background
Irradiation-assisted stress corrosion cracking (IASCC) is not a new problem in boiling water reactors (BWRs). Type 304 stainless steel fuel cladding, used in early BWR designs, developed intergranular stress corrosion cracks not long after being placed in service. Neutron irradiation fluence levels at the time of failure were estimated to be on the order of 1022 n/cm2 (>1 MeV) and fabrication and fuel/cladding interaction stresses were believed to contribute to the problem. Other BWR component failures which were attributed to IASCC included neutron source holders and control rod absorber tubes. These components were also fabricated from Type 304 stainless steel and were exposed to fluences of about 1022 n/cm 2 (>1 MeV) at failure. Welding residual stresses were believed to contribute to the neutron source holder failures while swelling of B4C was the identified source of loading for the control rod absorber tubes failures.
However, the discovery of a cracked control blade handle at the Dresden 3 Nuclear Power Station suggested that IASCC may be possible with little or no stress. This component was fabricated from solution annealed Type 304 stainless steel and, except through the removal of the control blade, experienced no load.
More recently, cracking has been observed in the beltline region of the core shroud in foreign and domestic BWRs and has been observed in the top guide of one domestic BWR. For conservative analysis purposes, the cracking has been assumed to be due to IASCC. Fluence levels predicted at the end-of-life of the shroud can exceed 1021 n/cm2 (>1 MeV), and the end-of-life top guide fluences are on the order of 1022 n/cm2 (>1 MeV) at the highest fluence locations. These recent experiences combined with historical cracking events make IASCC an important current issue. IASCC Mechanism
 Like intergranular stress corrosion cracking (IGSCC) which has occurred in BWR stainless steel piping, IASCC is believed to be due to a critical combination of susceptible material condition, stress (or strain) and a suitably aggressive environment, as shown in the figure on the right.
For IGSCC of austenitic stainless steel piping, the susceptible material condition is sensitization (chromium depletion at the grain boundaries) due to the welding process. Stress results from piping loads and differential thermal expansion along with welding-induced residual stresses. The suitably aggressive environment is the oxygen-containing BWR water.
For IASCC, the stress corrosion cracking mechanism is not as well understood. Rather than classical chromium depletion sensiti- zation, the susceptible material condition is believed to be due to neutron-induced segregation of impurities (S, Si, P) to the grain boundaries. Although field experience suggests that with sufficient fluence, little external stress is needed to produce IASCC, high stress is still believed to be an accelerant. The oxidizing nature of the BWR environment, produced by radiolysis of the water by neutron and gamma irradiation in the core region, is even greater inside the reactor vessel where short-lived oxidizing species such as H2O2 also exist.
Mechanical Properties
In addition to contributing to IASCC of reactor internal components and structures, neutron irradiation also alters the mechanical properties of austenitic stainless steel. As part of a LWR Component Long-Term Integrity Study performed by Structural Integrity Associates (SI) for the Electric Power Research Institute (EPRI), the effects of neutron irradiation on the mechanical properties of Type 304 stainless steel were estimated from available data. Very significant increases in yield strength with corresponding decreases in elongation and reduction in areas were noted for fluences greater than 1021 n/cm2 (>1 MeV). For temperatures typical of LWR operation, yield strength is almost equivalent to the ultimate strength at fluences approaching 1022 n/cm2 (>1 MeV).
The reductions in toughness and increases in the ductile-to-brittle transition temperature are well known for ferritic materials such as low alloy reactor vessel steels. However, this effect on toughness has generally received less attention for austenitic materials which do not exhibit a classical ductile-to-brittle transition temperature. Although these austenitic materials are also exempt from ASME Code toughness testing requirements, several studies indicate that low levels of toughness result from high levels of irradiation; especially in irradiated weld metal. This phenomenon is shown in the accompanying table which gives lower bound estimates of fracture toughness (KIc) following neutron irradiation for Type 304, 316, and 308 (weld metal) stainless steels.
Significance of IASCC
In the 1980s, SI completed an evaluation for EPRI on the potential significance of IASCC of the BWR top guide structure. A schematic of the top guide structure with typical end-of-fluence estimates is shown in the accompanying figure. The study included the following:
- Analysis of appropriate materials data
- Definition and evaluation of top guide loads
- Calculation of top guide stresses
- Identification of critical crack locations and estimation of critical crack sizes
- Evaluation of possible failure consequences, and
- Identification of the types of additional information that were necessary to perform the required structural analysis.
The EPRI study showed that for the top guide configuration of a typical BWR/4, rather large IASCC cracks could develop before top guide beam failure would be predicted during an earthquake, the most significant loading event. Further, multiple beam failures would be required for top guide IASCC to become a significant safety issue. These conclusions were reached by conducting a finite element stress analysis of the top guide structure which realistically modeled load distributions; and a fracture mechanics analysis which included expected material fracture toughness gradients.
SI recently performed an analysis for a BWR which exhibited cracking in its top guide beams. Consideration was given to higher crack growth rates and a decrease in material toughness directly attributable to high fluence. The conclusions were essentially the same as those for the EPRI top guide study discussed above.
Utility-sponsored analyses of core shrouds containing cracks have shown that the shroud is also highly tolerant of flaws, even at the most highly irradiated, lowest toughness region. Although the cracking observed in core shrouds to date has not been a length sufficient to jeopardize the load-carrying ability of the shroud, very long cracks, apparently the result of IASCC, have been observed. Understanding the mechanism(s) of IASCC, the controlling parameters, and the expected crack growth rates are areas where SI continues to be an industry leader.
For additional information on IASCC or SI's capabilities in this area, please contact SI.
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