Robin L. Jones

Threshold conditions for iodine-induced stress corrosion cracking of unirradiated zircaloy-4 tubing under internal pressurization

Abstract The results are presented of stress corrosion cracking (SCC tests in which nuclear power reactor grade zircaloy-4 tubing specimens were internally pressurized with a mixture of helium and iodine at ( 633 ± 5 ) K. Both as-received and artificially preflawed specimens were tested at an initial iodine availability of ~60 g/m2 zircaloy surface. It is shown that the failure times in these tests correlate more reliably with hoop stress than with nominal stress intensity or failure strain, and that a threshold hoop stress of ~295 MPa exists for SCC failure within test times up to 605 ks. The origin of this threshold stress is discussed and it is concluded that the observed behavior is consistent with either a critical stress or a critical strain rate being required for the formation of iodine-induced stress corrosion cracks in unirradiated zircaloy tubing.

Hydrogen water chemistry for BWRs

Abstract Intergranular stress corrosion cracking (IGSCC) has been responsible for more than 1,000 cases of cracking in austenitic stainless steel piping systems in boiling water reactors (BWRs). This paper presents the status of efforts in the United States to prevent IGSCC in BWRs during power operation by modifying the chemistry of the reactor water. The technical basis for this alternative water chemistry, called hydrogen water chemistry (HWC), is described and the results are presented of an ongoing in-plant program, at Commonwealth Edisons Dresden-2 plant, to verify the HWC concept and systematically assess the consequences of using it in an operating BWR. In addition, progress toward implementation of HWC at other U.S. plants is summarized.

Chemical aspects of iodine-induced stress corrosion cracking of Zircaloys

The thermodynamics of the zirconium-iodine system are summarized. Thermodynamic information for iodine chemisorbed on zirconium surfaces is also presented. These thermochemical results are used to analyze chemical behavior in situations related to stress corrosion cracking (SCC). Cracking initiation sites in commercial Zircaloy tubing were found to be associated with impurities (iron, aluminum, silicon, and chromium) in the Zircaloy surface in microsurface examination of failures produced in iodine-induced SCC tests. It is suggested that the impurity site may react with iodine to form a locally embrittled region that fails under stress and acts as a crack initiator.

Controlling stress corrosion cracking in boiling water reactors

Abstract Intergranular Stress Corrosion Cracking (IGSCC) adjacent to girth welds in austenitic stainless steel piping systems has been a serious problem in Boiling Water Reactor (BWR) plants in the U.S. for more than a decade. Recent observations suggest that SCC problems also may limit the service life of many reactor internals in BWRs. A major research and development program on BWR pipe cracking was cofunded by the Electric Power Research Institute (EPRI), the General Electric Company (GE), and the BWR Owners Group for IGSCC Research between 1979 and 1988 and a similar program on reactor internals and vessel attachments began in 1989. A brief description is presented of the pipe cracking remedies that were developed during the earlier program, and the prospects of adapting these remedies for the protection of internals and attachments are discussed.

Observations on Damage Accumulation During Iodine-Induced Stress Corrosion Cracking of Zircaloy Cladding

A predictive model has recently been proposed for pellet-cladding interaction (PCI) failures of water reactor fuel rods. The model is based on the assumption that PCI failures are attributable to stress corrosion cracking (SCC) of the Zircaloy cladding induced by fission product iodine. Because the kinetics of iodine-induced SCC of irradiated Zircaloys are not well-documented at present, the PCI model is based largely on failure-time data from constant stress tube pressurization experiments.

LWR water chemistry guidelines

Abstract This review describes the development of water chemistry guidelines to reduce corrosion damage and control radiation buildup in light water reactors. Three sets of guidelines are reviewed, covering BWR systems, PWR primary systems, and PWR secondary systems. The technical basis for the water chemistry parameters are discussed and the corrective actions to be taken if parameters exceed the recommended values are outlined.

Hydrogen water chemistry for BWRs: a status report on the EPRI development program

Abstract Many boiling water reactors (BWRs) have experienced extensive intergranular stress corrosion cracking (IGSCC) in their austenitic stainless steel reactor coolant system piping, resulting in serious adverse impacts on plant capacity factors, O&M costs, and personnel radiation exposures. A major research program to provide remedies for BWR pipe cracking was co-funded by EPRI, GE, and the BWR Owners Group for IGSCC Research between 1979 and 1988. Results from this program show that the likelihood of IGSCC depends on reactor water chemistry (particularly on the concentrations of ionic impurities and oxidizing radiolysis products) as well as on material condition and the level of tensile stress. Tests have demonstrated that the concentration of oxidizing radiolysis products in the recirculating reactor water of a BWR can be reduced substantially by injecting hydrogen into the feedwater. Recent plant data show that the use of hydrogen injection can reduce the rate of IGSCC to insignificant levels if the concentration of ionic impurities in the reactor water is kept sufficiently low. This approach to the control of BWR pipe cracking is called hydrogen water chemistry (HWC). This paper presents a review of the results of EPRIs HWC development program from 1980 to the present. In addition, plans for additional work to investigate the feasibility of adapting HWC to protect the BWR vessel and major internal components from potential stress corrosion cracking problems are summarized.

Influence of surface condition on crack initiation in iodine-induced stress corrosion cracking of zircaloys

The effect of surface condition on the initiation of iodine-induced stress corrosion cracks in Zircaloy tubing was investigated. The internal surface of the Zircaloy tubing was given one of three surface treatments, namely etching, grit blasting, or shot blasting. Each of these treatments can readily be performed commercially on Zircaloy fuel cladding. Specimens of surface-treated tubing were locally stressed in an iodine environment at 590 K by indenting the outer surface of the tube wall with a small steel ball. The crack initiation pattern on the inner surface was examined in a scanning electron microscope. Crack initiation was found to be least developed for etched surfaces, was most developed for shot-blasted surfaces, and was developed to an intermediate degree for grit-blasted surfaces. Apparent anomalies between these crack initiation data and the time to failure data obtained previously in tube pressurization tests are rationalized on the basis of the cracking processes.

BWR pipe crack control using hydrogen water chemistry: Status report on Dresden-2 program

Abstract One of the proposed remedies for intergranular stress corrosion cracking of stainless steel piping in BWRs is an alternative water chemistry called hydrogen water chemistry (H 2 WC) that involves suppression of reactor water dissolved oxygen to ≤ 20 ppb via hydrogen injection to the feedwater in conjunction with control of conductivity to ≤ 0.3 μ mho/cm. A long-term verification program, over two or three 18 month fuel cycles, was started at Commonwealth Edisons Dresden-2 reactor in April 1983 (Cycle 9). This paper describes the results of the water chemistry changes, structural material and fuel evaluations, and plant radiation level changes during Cycle 9, which ended in October 1984. To date the results of the verification program are very encouraging. They indicate that the alternative water chemistry, based on hydrogen additions to the feedwater to suppress oxygen and low conductivity, can be maintained in a large operating BWR, and that it does mitigate IGSCC in stainless steel recirculation piping. Monitoring of fuel and plant materials will continue in Dresden-2 at least through Cycle 10 to confirm the absence of any unusual side effects of this remedy for IGSCC.

Applications of Fatigue and Fracture Tolerant Design Concepts in the Nuclear Power Industry

To assure the integrity of nuclear power plant components, fatigue and fracture tolerant design concepts have been incorporated in Sections III and XI of the ASME Code; these contain requirements for nuclear power plant design, construction, and in-service inspection. The methods used in the Code to design against fatigue and brittle fracture are described together with the fracture mechanics based procedure suggested in Sections XI for the evaluation of flaws detected by in-service inspections. Some aspects of the present Code methods that could probably be improved are identified. 19 refs.