Robert L. Sindelar

LITERATURE SURVEY OF GASEOUS HYDROGEN EFFECTS ON THE MECHANICAL PROPERTIES OF CARBON AND LOW ALLOY STEELS

Literature survey has been performed for a compendium of mechanical properties of carbon and low alloy steels following hydrogen exposure. The property sets include yield strength, ultimate tensile strength, uniform elongation, reduction of area, threshold stress intensity factor, fracture toughness, and fatigue crack growth. These properties are drawn from literature sources under a variety of test methods and conditions. However, the collection of literature data is by no means complete, but the diversity of data and dependency of results in test method is sufficient to warrant a design and implementation of a thorough test program. The program would be needed to enable a defensible demonstration of structural integrity of a pressurized hydrogen system. It is essential that the environmental variables be well-defined (e.g., the applicable hydrogen gas pressure range and the test strain rate) and the specimen preparation be realistically consistent (such as the techniques to charge hydrogen and to maintain the hydrogen concentration in the specimens).

Flaw Stability in Mild Steel Tanks in the Upper-Shelf Ductile Range—Part II: J-Integral-Based Fracture Analysis

The J-integral fracture methodology was applied to evaluate the stability of postulated flaws in mild steel storage tanks. The material properties and the J-resistance (JR) curve were obtained from the archival A285 Grade B carbon steel test data. The J-integral calculation is based on the center-cracked panel (CCP) solution of Shih and Hutchinson (1976). A curvature correction was applied to account for the cylindrical shell configuration. A finite element analysis of an arbitrary flaw in the storage tank demonstrated that the curvature-corrected CCP solution is a close approximation. The maximum storage tank fluid level for a postulated flaw size can be established based on the J-integral flaw stability methodology.

A Framework to Develop Flaw Acceptance Criteria for Structural Integrity Assessment of Multipurpose Canisters for Extended Storage of Used Nuclear Fuel

A multipurpose canister (MPC) made of austenitic stainless steel is loaded with used nuclear fuel assemblies and is part of the transfer cask system to move the fuel from the spent fuel pool to prepare for storage, and is part of the storage cask system for on-site dry storage. This weld-sealed canister is also expected to be part of the transportation package following storage. The canister may be subject to service-induced degradation especially if exposed to aggressive environments during possible very long-term storage period if the permanent repository is yet to be identified and readied. Stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone because the construction of MPC does not require heat treatment for stress relief. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic Inservice Inspection. The external loading cases include thermal accident scenarios and cask drop conditions with the contribution from the welding residual stresses. The determination of acceptable flaw size is based on the procedure to evaluate flaw stability provided by American Petroleum Institute (API) 579 Fitness-for-Service (Second Edition). The material mechanical and fracture properties for base and weld metals and the stress analysis results are obtained from the open literature such as NUREG-1864. Subcritical crack growth from stress corrosion cracking (SCC), and its impact on inspection intervals and acceptance criteria, is not addressed.Copyright

Comparison of Ring Compression Testing to Three Point Bend Testing for Unirradiated ZIRLO Cladding

Safe shipment and storage of nuclear reactor discharged fuel requires an understanding of how the fuel may perform under the various conditions that can be encountered. One specific focus of concern is performance during a shipment drop accident. Tests at Savannah River National Laboratory (SRNL) are being performed to characterize the properties of fuel clad relative to a mechanical accident condition such as a container drop. Unirradiated ZIRLO tubing samples have been charged with a range of hydride levels to simulate actual fuel rod levels. Samples of the hydrogen charged tubes were exposed to a radial hydride growth treatment (RHGT) consisting of heating to 400°C, applying initial hoop stresses of 90 to 170 MPa with controlled cooling and producing hydride precipitates. Initial samples have been tested using both a) ring compression test (RCT) which is shown to be sensitive to radial hydride and b) three-point bend tests which are less sensitive to radial hydride effects.Hydrides are generated in Zirconium based fuel cladding as a result of coolant (water) oxidation of the clad, hydrogen release, and a portion of the released (nascent) hydrogen absorbed into the clad and eventually exceeding the hydrogen solubility limit. The orientation of the hydrides relative to the subsequent normal and accident strains has a significant impact on the failure susceptability. In this study the impacts of stress, temperature and hydrogen levels are evaluated in reference to the propensity for hydride reorientation from the circumferential to the radial orientation. In addition the effects of radial hydrides on the Quasi Ductile Brittle Transition Temperature (DBTT) were measured. The results suggest that a) the severity of the radial hydride impact is related to the hydrogen level-peak temperature combination (for example at a peak drying temperature of 400°C; 800 PPM hydrogen has less of an impact/ less radial hydride fraction than 200 PPM hydrogen for the same thermal history) and b) for critical strains in post drying handling, storage and accident conditions the 3 point bend strain tolerance is less affected by radial hydrides than the conventional ring compression test (the radial hydride related Quasi DBTT associated with a three point bend straining is lower (better) than that measured by the ring compression tests).Copyright

Experience With Damaged Spent Nuclear Fuel at U.S. DOE Facilities

This report summarizes some of the challenges encountered and solutions implemented to ensure safe storage and handling of damaged spent nuclear fuels (SNF). It includes a brief summary of some SNF storage environments and resulting SNF degradation, experience with handling and repackaging significantly degraded SNFs, and the associated lessons learned. This work provides useful insight and resolutions to many engineering challenges facing SNF handling and storage facilities. The context of this report is taken from a report produced at Idaho National Laboratory and further detailed information, such as equipment design and usage, can be found in the appendices to that report.Copyright

Stress Corrosion Cracking of Carbon Steel Weldments

An experiment was conducted to investigate the role of weld residual stress on stress corrosion cracking in welded carbon steel plates prototypic to those used for nuclear waste storage tanks. Carbon steel specimen plates were butt-joined with Gas Metal Arc Welding technique. Initial cracks (seed cracks) were machined across the weld and in the heat affected zone. These specimen plates were then submerged in a simulated high level radioactive waste chemistry environment. Stress corrosion cracking occurred in the as-welded plate but not in the stress-relieved duplicate. A detailed finite element analysis to simulate exactly the welding process was carried out, and the resulting temperature history was used to calculate the residual stress distribution in the plate for characterizing the observed stress corrosion cracking. It was shown that the cracking can be predicted for the through-thickness cracks perpendicular to the weld by comparing the experimental KISCC to the calculated stress intensity factors due to the welding residual stress. The predicted crack lengths agree reasonably well with the test data. The final crack lengths appear to be dependent on the details of welding and the sequence of machining the seed cracks, consistent with the prediction.

Flaw Stability Considering Residual Stress for Aging Management of Spent Nuclear Fuel Multiple-Purpose Canisters

A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. Because heat treatment for stress relief is not required for the construction of the MPC, the canister is susceptible to stress corrosion cracking in the weld or heat affected zone regions under long-term storage conditions. Logic for flaw acceptance is developed should crack-like flaws be detected by Inservice Inspection. The procedure recommended by API 579-1/ASME FFS-1, Fitness-for-Service, is used to calculate the instability crack length or depth by failure assessment diagram. It is demonstrated that the welding residual stress has a strong influence on the results.

Diffusion, Thermal Properties and Chemical Compatibilities of Select MAX Phases with Materials For Advanced Nuclear Systems

The demands of Gen IV nuclear power plants for long service life under neutron irradiation at high temperature are severe. Advanced materials that would withstand high temperatures (up to 1000+ oC) to high doses in a neutron field would be ideal for reactor internal structures and would add to the long service life and reliability of the reactors. The objective of this work is to investigate the chemical compatibility of select MAX with potential materials that are important for nuclear energy, as well as to measure the thermal transport properties as a function of neutron irradiation. The chemical counterparts chosen for this work are: pyrolytic carbon, SiC, U, Pd, FLiBe, Pb-Bi and Na, the latter 3 in the molten state. The thermal conductivities and heat capacities of non-irradiated MAX phases will be measured.

Chloride-Induced Stress Corrosion Crack Growth Under Dry Salt Conditions: Application to Evaluate Growth Rates in Multipurpose Canisters

Many dry cask storage systems for spent nuclear fuel consist of a dry shielded canister (DSC) design that includes a welded construction (and weld-sealed) austenitic stainless steel multipurpose canister that is placed within a concrete overpack and stored on an outside pad. The present regulatory basis for dry cask storage is 60 years (20-year initial and up to 40-year relicense). Aging of the materials and structures of Dry Cask Storage Systems (DCSS) are considered in the demonstration that the safety functions are maintained throughout the license period. The sealed stainless steel canister provides a confinement function in a DCSS. Stress corrosion cracking (SCC) may occur when chloride-bearing salts and/or dust deliquesce on the external surface of the spent nuclear fuel (SNF) canister at weld residual stress regions. An SCC growth rate test was developed using instrumented bolt-load compact tension specimens (ASTM E1681) with experimental apparatus that allows an initially dried salt to deliquesce and infuse naturally to the crack front under temperature and humidity parameters relevant to the canister storage environmental conditions. The shakedown tests were conducted over a range of relative humidity controlled by the guidance in ASTM E104 at 50 °C with salt assemblages of (1) mixture of artificial dust and deliquescent salts (2) a mixture of artificial dust and ASTM simulated sea salt. After five months exposure the specimens were examined for evidence of CISCC and observations are reported for both salt/dust mixtures. The test specimen and apparatus designs will be modified to enhance the interaction between the deliquescing salt and the crack front for more accurate characterization of the crack growth rate as a function of stress intensity factor, which is an essential input to the determination of in-service inspection frequency of SNF canisters.

Performance of bolted closure joint elastomers under cask aging conditions

The bolted closure joint of a bare spent fuel cask is susceptible to age-related degradation and potential loss of confinement function under long-term storage conditions. Elastomeric seals, a component of the joint typically used to facilitate leak testing of the primary seal that includes the metallic seal and bolting, is susceptible to degradation over time by several mechanisms, principally via thermo-oxidation, stress-relaxation, and radiolytic degradation under time and temperature condition. Irradiation and thermal exposure testing and evaluation of an ethylene-propylene diene monomer (EPDM) elastomeric seal material similar to that used in the CASTOR® V/21 cask for a matrix of temperature and radiation exposure conditions relevant to the cask extended storage conditions, and development of semiempirical predictive models for loss of sealing force is in progress. A special insert was developed to allow Compressive Stress Relaxation (CSR) measurements before and after the irradiation and/or thermal exposure without unloading the elastomer. A condition of the loss of sealing force for the onset of leakage was suggested. The experimentation and modeling being performed could enable acquisition of extensive coupled aging data as well as an estimation of the timeframe when loss of sealing function under aging (temperature/radiation) conditions may occur.