Dry storage systems (DSSs) store spent nuclear fuel at many operating and decommissioned power reactor sites in the United States. Carbon steel, low-alloy steel, and stainless steel components are commonly used to construct DSSs. Some of the components are exposed to outdoor air and sheltered environments in which the materials may be susceptible to degradation. Potential aging mechanisms include general corrosion, pitting and crevice corrosion, galvanic corrosion, microbiologically influenced corrosion, stress corrosion cracking (SCC), creep, fatigue, thermal aging, radiation embrittlement, stress relaxation, and wear. This paper presents an assessment of these aging mechanisms based on review of literature and operating experience from nuclear and nonnuclear applications and their long-term effects on the integrity of DSSs. The results of this study indicate that the following mechanisms are credible for safely storing spent nuclear fuel during a 60-year timeframe: (i) general corrosion, pitting and crevice corrosion, galvanic corrosion, stress relaxation, and wear of carbon steel and low-alloy steel; (ii) wear of stainless steel; (iii) galvanic corrosion of stainless steel in contact with graphite; (iv) SCC of stainless steel welds and the weld heat affected zone; and (v) pitting and crevice corrosion of stainless steel as a precursor, or initiation site, for SCC. For weld-free austenitic stainless steel components or regions away from welds, such as the storage canister body, atmospheric SCC could be a credible aging mechanism provided sufficient tensile stresses exist. The results of this work are being used to inform recommendations for monitoring, inspection, and other activities to manage the aging of DSSs.

During an inspection in 2002, control rod drive mechanism (CRDM) nozzle 3 of the original reactor pressure vessel head (RPVH-1) at Davis-Besse Nuclear Power Station, Unit 1 was found to be damaged due to a combination of pressurized water stress corrosion cracking (PWSCC) and boric acid corrosion. This nozzle was fabricated from UNS N06600, heat M3935. Due to the extent of damage to RPVH-1, it was replaced with an identical head from the cancelled Midland plant. The majority of CRDM housings in the replacement head (RPVH-2) were fabricated from UNS N06600, heat M7929. Operations resumed in 2004. During a 2010 inspection, cracking was detected in 24 of 69 CRDM housing of RPVH-2. Repairs were made to these nozzles. In 2012, RPVH-2 was replaced with a head containing nozzles fabricated from UNS N06690. This paper discusses crack growth rate and material characterization performed on samples removed from RPVH-2 and compares these results to previously conducted testing from RPVH-1. While crack growth rates of both materials were found to be high, substantial differences were observed between the two samples in both microstructure and grain boundary chemistry.

Spent nuclear fuel at a number of U.S. locations is stored at independent spent fuel storage installations in dry cask storage systems (DCSSs), which commonly consist of a welded austenitic stainless steel canister within a larger concrete vault or overpack. Stress corrosion cracking of welded stainless steel is considered a high-priority technical issue and functional monitoring has been identified as one of the top crosscutting issues concerning performance of various dry cask components. The work presented in this paper reviews and assesses the current state of technology for directly monitoring stress corrosion cracking, as well as the important environmental conditions—including temperature, humidity, and chloride concentration—that could affect this degradation mechanism. A variety of techniques were identified to be potentially suitable for application to extended storage, ranging from detecting crack initiation and measuring propagation from stress corrosion cracking to measuring the chloride concentration in deliquescent solutions. Some techniques are well developed and commercially available, but some require significant advancement to overcome limitations. Overall, because of geometry, space limitations, and the high ionizing radiation of DCSSs, all the monitoring methods must be modified and tailored for this application.

In the United States, some spent nuclear fuel is stored in dry cask storage systems. Most dry storage systems use canisters fabricated from austenitic stainless steel. Canisters are exposed to the ambient atmosphere and literature data indicate that near regions of industrial, commercial, or agricultural activities, abundant airborne particulates may include sulfate, nitrate, and ammonium as well as a smaller amount of chloride. This study evaluated the stress corrosion cracking (SCC) susceptibility of as-received and sensitized UNS S30400 stainless steel exposed to ammonium nitrate mixed with sodium chloride salts. SCC tests were conducted at 45 °C-44% relative humidity using U-bend specimens coated with salt mixtures of ammonium nitrate and sodium chloride with mole ratios of 3 and 6. After a period of exposure of 1.5 months, examination of specimens showed that severe SCC occurred. The SCC severity is greater for specimens coated with ammonium nitrate and sodium chloride at mole ratio of 3. Previous work under this program showed that no SCC was observed on specimens exposed only to ammonium nitrate. The results indicate that chloride has a large effect on SCC initiation and nitrate did not inhibit SCC at the tested nitrate to chloride mixture mole ratios.

At a number of locations in the U.S., spent nuclear fuel (SNF) is maintained at independent spent fuel storage installations (ISFSIs). These ISFSIs, which include operating and decommissioned reactor sites, Department of Energy facilities in Idaho, and others, are licensed by the U.S. Nuclear Regulatory Commission (NRC) under Title 10 of the Code of Federal Regulations, Part 72. The SNF is stored in dry cask storage systems, which most commonly consist of a welded austenitic stainless steel canister within a larger concrete vault or overpack vented to the external atmosphere to allow airflow for cooling. Some ISFSIs are located in marine environments where there may be high concentrations of airborne chloride salts. If salts were to deposit on the canisters via the external vents, a chloride-rich brine could form by deliquescence. Austenitic stainless steels are susceptible to chloride-induced stress corrosion cracking (SCC), particularly in the presence of residual tensile stresses from welding or other fabrication processes. SCC could allow helium to leak out of a canister if the wall is breached or otherwise compromise its structural integrity. There is currently limited understanding of the conditions that will affect the SCC susceptibility of austenitic stainless steel exposed to marine salts. NRCmore » previously conducted a scoping study of this phenomenon, reported in NUREG/CR-7030 in 2010. Given apparent conservatisms and limitations in this study, NRC has sponsored a follow-on research program to more systematically investigate various factors that may affect SCC including temperature, humidity, salt concentration, and stress level. The activities within this research program include: (1) measurement of relative humidity (RH) for deliquescence of sea salt, (2) SCC testing within the range of natural absolute humidity, (3) SCC testing at elevated temperatures, (4) SCC testing at high humidity conditions, and (5) SCC testing with various applied stresses. Results to date indicate that the deliquescence RH for sea salt is close to that of MgCl{sub 2} pure salt. SCC is observed between 35 and 80 deg. C when the ambient (RH) is close to or higher than this level, even for a low surface salt concentration. (authors)« less

Some nuclear power plants operating in the United States store spent nuclear fuel on-site in dry cask storage systems. Most dry storage systems use canisters fabricated from austenitic stainless steel. This scoping study evaluated the stress corrosion cracking susceptibility of as-received, sensitized, and welded type 304 stainless steel exposed to ammonium nitrate, sulfate, and bisulfate salts representing the soluble chemical composition in non-coastal airborne particulates. Stress corrosion cracking tests were conducted at 45 °C-44% relative humidity (RH) and 35 °C-72% RH using type 304 stainless steel U-bend specimens coated with single salts and salt mixtures of ammonium sulfate, and ammonium nitrate with varying mole ratios representing the geographic, temporal, and diurnal variation of non-coastal salt composition. Examination of specimens from surface and cross section after exposure for less than two months showed that: (i) stress corrosion cracking had not occurred on any specimens, (ii) general corrosion occurred on the as-received and sensitized specimens deposited with NH4HSO4, (iii) a few shallow pits were evident on specimens deposited with a (NH4)2SO4 and NH4NO3 mixture, and (iii) the welded specimens exposed to NH4HSO4 had extensive general corrosion and grain boundary attack.

This chapter contains sections titled: Abstract Introduction PWSCC Historical Perspective Research NRC Observations Conclusions

Alloy 22 is the material preferred by the U.S. Department of Energy for the waste package outer container for geological disposal of high-level radioactive waste at the proposed site in Yucca Mountain, Nevada. Alloy 22 is considered to be extremely resistant to various modes of aqueous corrosion over broad ranges of temperature, pH, and concentration of anionic and oxidizing species. Uniform corrosion under passive dissolution conditions, localized corrosion in the form of crevice corrosion, and stress corrosion cracking are discussed on the basis of experimental results obtained with mill annealed, thermally treated, and welded specimens using electrochemical techniques. The approach developed for long-term performance prediction, including the use of empirically derived parameters for assessing localized corrosion and the modeling of the passive dissolution behavior, is described.

Common assumptions to extrapolate the lifetime of proposed high-level waste disposal containers made of Ni-Cr-Mo alloys, in the absence of environmental and electrochemical conditions leading to localized corrosion and stress corrosion cracking, are evaluated based on a mechanistic model for passive dissolution. The predominant charge conduction mechanism through the oxide film formed on Ni-Cr-Mo alloys is hypothesized to be interstitial transport of metal cations. Dissolution of the alloy and conduction of interstitial species through the film create vacancies in the alloy. The anodic current density under potentiostatic control decreases as a function of time, and potentiostatic decays in the current density are rationalized on the basis of vacancy accumulation at the metal-oxide interface. It is concluded that the dissolution process is regulated by vacancy-enhanced diffusion of the elements in the alloy. Long-term stoichiometric dissolution arises if the diffusion coefficients of the alloying elements are similar. No credible scenario is envisioned by which catastrophic failure may occur as a result of longterm passive dissolution.

Localized corrosion of high level nuclear waste containers is considered an important factor that will have a strong influence on the overall performance of the proposed repository at Yucca Mountain, NV. The present candidate container material, Alloy 22 [UNS N06022 (57Ni-22Cr-13.5Mo-3W-3Fe)], is highly resistant to localized corrosion. Assessing the performance of the HLW containers also requires an evaluation of localized corrosion resistance and determination of the passive corrosion rate of Alloy 22 weldments. The localized corrosion resistance of welded Alloy 22 specimens was evaluated by measuring the repassivation potential in chloride containing solutions whereas potentiostatic tests were used to determine the passive corrosion rate of the welded material. The results for the welded material are compared to those for the base metal.

In-situ coupled multielectrode array sensors were used to measure the non-uniform corrosion of carbon steel and stainless steel materials under KCl salt deposit in simulated dry repository environments. It was found that the initiation of non-uniform corrosion occurs at a relative humidity that is 14% lower than the deliquescence relative humidity of the chloride salt. It was found also that once significant corrosion had occurred, the non-uniform corrosion process for the carbon steel material under the salt deposit continues at relative humidities as low as 27%.

Many nuclear plants have begun or will have to start storing spent nuclear waste in onsite dry storage containers, because the room in the spent fuel pools is reaching capacity. Some of these plants are located close to the coast and are in chloride-rich environments. Because the dry storage containers may be made of austenitic stainless steel, a potential concern is their susceptibility to chloride induced stress corrosion cracking. The objective of this work is to evaluate the stress corrosion cracking susceptibility of austenitic type 304, 304L, and 316L stainless steel to chloride stress corrosion cracking. U-bend samples were created out of these alloys and assembled in an environmental chamber. The U-bend samples were heated to various temperatures and subjected to an accelerated corrosion tests, involving directly spraying the samples with simulated sea salt. Cracking was observed in all the samples except those where the temperature remained below a chloride stress corrosion cracking temperature limit. The test was overly conservative because of the direct spray, which is not expected to occur at the actual storage facilities. Secondly, the spray altered U-bend samples temperature making it more susceptible to SCC. In order to create a more realistic, but conservative accelerated test, the factors important to chloride stress corrosion cracking (i.e., relative humidity and temperature) were examined.

In 2007, two severe transportation accidents, involving primarily long-haul tractor trailers, occurred in the State of California. In the first, which occurred in Oakland in the “MacArthur Maze” section of Interstate 580, a tractor trailer carrying gasoline impacted an overpass support column and burst into flames. The subsequent fire, which burned for over 2 hours, led to the collapse of the overpass onto the remains of the tractor trailer, due to the loss of strength in the steel exposed to the fire. The second incident was a chain-reaction accident involving several tractor trailers in the I-5 “Newhall Pass” truck bypass tunnel in Santa Clarita. This accident also involved an intense fire that damaged the tunnel and required the closing of the tunnel for repairs to the concrete walls. The US Nuclear Regulatory Commission is studying both these accidents to examine any potential regulatory implications related to the safe transport of radioactive materials and spent nuclear fuel in the United States. This paper will provide a summary of that effort.

In 2007, a severe transportation accident occurred in Oakland, California in what is commonly known as the "MacArthur Maze" section of Interstate 580 (I-580). The accident involved a tractor trailer carrying gasoline that impacted an overpass support column and burst into flames. The subsequent fire burned for over 2 hours and led to the collapse of the overpass due to the loss of strength in the structural steel that supported the overpass. The US Nuclear Regulatory Commission (NRC) studied this accident to examine any potential regulatory implications related to the safe transport of radioactive materials, including spent nuclear fuel. This paper will discuss the details of the NRC's MacArthur Maze fire investigation.

At present, corrosion protection of USMC steel components is being implemented through an initial abrasive blasting of the metal substrate followed by the application of solvent or water based coatings, which does not appear to provide sufficient long-term corrosion protection, especially if the coating is partially disrupted exposing the base metal. To address this issue, primer coatings, containing high concentrations of sacrificially available Zn, can be used for long-term protection of steel. Electrochemical tests were conducted to determine the corrosion performance of various commercially available Zn-rich coating products applied to steel immersed in simulated seawater for up to 2,000 hr. Initial findings show distinctive corrosion protection of the Zn-rich coatings, manifested by the difference in the open circuit potentials and corrosion rates between the coatings tested. For open circuit potentials more noble than -750 mV, electrochemical measurements indicated some corrosion in progress of steel and poor galvanic protection by the coating. At more negative potentials, enhanced galvanic protection by the surrounding coating was noted for some of the Zn-rich coatings tested, resulting in larger protective currents and prevention of steel corrosion for up to 2,000 hr of exposure.

Long-term corrosion performance of the waste package is among the key engineered barrier system attributes of a potential high-level waste repository at Yucca Mountain, Nevada. Waste package degradation processes are evaluated on the basis of independent investigations conducted at the Center for Nuclear Waste Regulatory Analyses. This paper summarizes the results of laboratory measurements and model analyses focused on uniform, localized, and microbially influenced corrosion, and stress corrosion cracking of Alloy 22 (UNS N06022).

Corrosion preventative compounds (CPCs) are used to slow corrosion of United States Marine Corps (USMC) assets. CPC coatings are temporary, and reapplication intervals are necessary to maintain the corrosion reducing ability of these compounds. Methods to perform condition based maintenance, including reapplication of CPCs as needed, are being sought in an effort to reduce the costs of USMC asset maintenance. To aid in this effort, coating degradation sensors integrated with a mote-based wireless network were deployed on actual USMC ground vehicles under a variety of storage options at Camp Lejeune, NC and Kaneohe Bay, HI. Results from each site were contrasted as a function of exposure conditions.

Corrosion of threaded fastener materials is an issue on multiple combat and combat support vehicles used by the US Marine Corps. Material selections for these components are based on numerous criteria such as strength, ballistics, and formability. When the protective paint systems are compromised, galvanic corrosion of the dissimilar materials is likely in environments with high relative humidity or in the presence of salt spray. Testing of fastener materials was conducted to determine compatibility with structural materials such as aluminum and steel. Electrochemical tests were used to measure corrosion rates and identify material compatibility.