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Effect of helium on void swelling in vanadium

Description: Little difference in void microstructural swelling of vanadium is observed when helium is injected simultaneously with a 46- or 5-MeV nickel beam as compared to no helium injection, at least at high dose rates. At lower dose rates, a strong helium effect is seen when the helium is injected prior to heavy ion bombardment. The effect of the helium is shown to be a strong function of the overall displacement damage rate. (DLC)
Date: January 1, 1975
Creator: Brimhall, J.L. & Simonen, E.P.
Partner: UNT Libraries Government Documents Department

DATING: A computer code for determining allowable temperatures for dry storage of spent fuel in inert and nitrogen gases

Description: The DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) code can be used to calculate allowable initial temperatures for dry storage of light-water-reactor spent fuel. The calculations are based on the life fraction rule using both measured data and mechanistic equations as reported by Chin et al. (1986). The code is written in FORTRAN and utilizes an efficient numerical integration method for rapid calculations on IBM-compatible personal computers. This report documents the technical basis for the DATING calculations, describes the computational method and code statements, and includes a user's guide with examples. The software for the DATING code is available through the National Energy Software Center operated by Argonne National Laboratory, Argonne, Illinois 60439. 5 refs., 8 figs., 5 tabs.
Date: December 1, 1988
Creator: Simonen, E.P. & Gilbert, E.R.
Partner: UNT Libraries Government Documents Department

Implications of early stages in the growth of stress corrosion cracking on component reliability

Description: Environment-induced crack growth generally progresses through several stages prior to component failure. Crack initiation, short crack growth, and stage 1 growth are early stages in crack development that are summarized in this paper. The implications of these stages on component reliability, derive from the extended time that the crack exists in the early stages because crack velocity is slow. The duration of the early stages provides a greater opportunity for corrective action if cracks can be detected. Several important factors about the value of understanding short crack behavior include: (1) life prediction requires a knowledge of the total life cycle of the crack including the early stages, (2) greater reliability is possible if the transition between short and long crack behavior is known component life after this transition is short and (3) remedial actions are more effective for short than long cracks.
Date: April 1, 1995
Creator: Jones, R.H. & Simonen, E.P.
Partner: UNT Libraries Government Documents Department

Crack-tip chemistry modeling of stage I stress corrosion cracking

Description: Stage I stress corrosion cracking usually exhibits a very strong K dependence with Paris law exponents of up to 30. 2 Model calculations indicate that the crack velocity in this regime is controlled by transport through a salt film and that the K dependence results from crack opening controlled salt film dissolution. An ionic transport model that accounts for both electromigration through the resistive salt film and Fickian diffusion through the aqueous solution was used for these predictions. Predicted crack growth rates are in excellent agreement with measured values for Ni with P segregated to the grain boundaries and tested in IN H{sub 2}SO{sub 4} at +900 mV. This salt film dissolution may be applicable to stage I cracking of other materials.
Date: October 1, 1991
Creator: Jones, R.H. & Simonen, E.P.
Partner: UNT Libraries Government Documents Department

Effect of pulsed irradiation on void swelling in nickel

Description: This study has compared the void microstructure in nickel induced by a pulsed ion bombardment to that induced by a steady-state irradiation. Pulse cycles of 10 seconds on and 10 seconds off produced no measurable difference in the void growth and swelling in the temperature range 775 to 975/sup 0/K compared to continuous irradiation at the same instantaneous dose rate. Void annealing during the pulse annealing period was minimal due to the large void sizes which were obtained in these irradiations. Hence no measurable effect of pulsing on void growth was observed.
Date: July 1, 1981
Creator: Brimhall, J.L.; Charlot, L.A. & Simonen, E.P.
Partner: UNT Libraries Government Documents Department

Mechanistic issues for modeling radiation-induced segregation

Description: Model calculations of radiation-induced chromium depletion and radiation-induced nickel enrichment at grain boundaries are compared to measured depletions and enrichments. The model is calibrated to fit chromium depletion in commercial purity 304 stainless steel irradiated in boiling water reactor (BWR) environments. Predicted chromium depletion profiles and the dose dependence of chromium concentration at grain boundaries are in accord with measured trends. Evaluation of chromium and nickel profiles in three neutron, and two ion, irradiation environments reveal significant inconsistencies between measurements and predictions.
Date: March 1, 1993
Creator: Simonen, E.P. & Bruemmer, S.M.
Partner: UNT Libraries Government Documents Department

Radiation hardening and radiation-induced chromium depletion effects on intergranular stress corrosion cracking of austenitic stainless steels

Description: Available data on neutron-irradiated materials have been analyzed and correlations developed between fluence, yield strength, grain boundary chromium concentration and cracking susceptibility in high-temperature water environments. Large heat-to-heat differences in critical fluence (0.2 to 2.5 n/cm[sup 2]) for IGSCC are documented.In many cases, this variability is consistent with yield strength differences among irradiated materials. IGSCC correlated better to yield strength than to fluence for most heats suggesting a possible role of the radiation-induced hardening (and microstructure) on cracking. However, isolatedheats reveal a wide range of yield strengths from 450 to 800 MPa necessary to promote IGSCC which cannot be understood by strength effects alone. Grain boundary Cr depletion explain differences in IGSCC susceptibility for irradiated stainless steels. Cr contents versus SCC shows that all materials showing IG cracking have some grain boundary depletion ([ge]2%). Grain boundary Cr concentrations for cracking (below [approximately]16 wt %) are in good agreement with similar SCC tests on unirradiated 304 SS with controlled depletion profiles. Heats that prompt variability in the yield strength correlation, are accounted for bydifferences in their interfacial Cr contents. Certain stainless steels are more resistant to cracking even though they have significant radiation-induced Cr depletion. It is proposed that Cr depletion is required for IASCC, but observed susceptibility is modified by other microchemical and microstructural components.
Date: March 1, 1993
Creator: Bruemmer, S.M. & Simonen, E.P.
Partner: UNT Libraries Government Documents Department

Mechanistic issues for modeling radiation-induced segregation

Description: Model calculations of radiation-induced chromium depletion and radiation-induced nickel enrichment at grain boundaries are compared to measured depletions and enrichments. The model is calibrated to fit chromium depletion in commercial purity 304 stainless steel irradiated in boiling water reactor (BWR) environments. Predicted chromium depletion profiles and the dose dependence of chromium concentration at grain boundaries are in accord with measured trends. Evaluation of chromium and nickel profiles in three neutron, and two ion, irradiation environments reveal significant inconsistencies between measurements and predictions.
Date: March 1, 1993
Creator: Simonen, E. P. & Bruemmer, S. M.
Partner: UNT Libraries Government Documents Department

Radiation-induced segregation: A microchemical gauge to quantify fundamental defect parameters

Description: Defect Kinetic are evaluated for austenitic stainless alloys by comparing model predictions to measured responses for radiation-induced grain boundary segregation. Heavy-ions, neutrons and proton irradiations having substantial statistical bases are examined. The combined modeling and measurement approach is shown to be useful for quantifying fundamental defect parameters. The mechanism evaluation indicates vacancy, migration energies of 1.15 eV or less and a vacancy formation energy at grain boundaries of 1.5 eV. Damage efficiencies of about 0.03 were established for heavy-ions and for light-water reactor neutrons. Inferred proton damage efficiencies were about 0.15. Segregation measured in an advanced gas-cooled reactor component was much greater than expected using the above parameters.
Date: December 1, 1994
Creator: Simonen, E. P. & Bruemmer, S. M.
Partner: UNT Libraries Government Documents Department

Radiation hardening and radiation-induced chromium depletion effects on intergranular stress corrosion cracking of austenitic stainless steels

Description: Available data on neutron-irradiated materials have been analyzed and correlations developed between fluence, yield strength, grain boundary chromium concentration and cracking susceptibility in high-temperature water environments. Large heat-to-heat differences in critical fluence (0.2 to 2.5 n/cm{sup 2}) for IGSCC are documented.In many cases, this variability is consistent with yield strength differences among irradiated materials. IGSCC correlated better to yield strength than to fluence for most heats suggesting a possible role of the radiation-induced hardening (and microstructure) on cracking. However, isolatedheats reveal a wide range of yield strengths from 450 to 800 MPa necessary to promote IGSCC which cannot be understood by strength effects alone. Grain boundary Cr depletion explain differences in IGSCC susceptibility for irradiated stainless steels. Cr contents versus SCC shows that all materials showing IG cracking have some grain boundary depletion ({ge}2%). Grain boundary Cr concentrations for cracking (below {approximately}16 wt %) are in good agreement with similar SCC tests on unirradiated 304 SS with controlled depletion profiles. Heats that prompt variability in the yield strength correlation, are accounted for bydifferences in their interfacial Cr contents. Certain stainless steels are more resistant to cracking even though they have significant radiation-induced Cr depletion. It is proposed that Cr depletion is required for IASCC, but observed susceptibility is modified by other microchemical and microstructural components.
Date: March 1, 1993
Creator: Bruemmer, S. M. & Simonen, E. P.
Partner: UNT Libraries Government Documents Department

Early stages in the development of stress corrosion cracks

Description: Processes in growth of short cracks and stage I of long stress corrosion cracks were identified and evaluated. There is evidence that electrochemical effects can cause short stress corrosion cracks to grow at rates faster or slower than long cracks. Short cracks can grow at faster rates than long cracks for a salt film dissolution growth mechanism or from reduced oxygen inhibition of hydrolytic acidification. An increasing crack growth rate with increasing crack length could result from a process of increasing crack tip concentration of a critical anion, such as Cl{sup {minus}}, with increasing crack length in a system where the crack velocity is dependent on the Cl{sup {minus}} or some other anion concentration. An increasing potential drop between crack tip and mouth would result in an increased anion concentration at the crack tip and hence an increasing crack velocity. Stage I behavior of long cracks is another early development stage in the life of a stress corrosion crack which is poorly understood. This stage can be described by da/dt = AK{sup m} where da/dt is crack velocity, A is a constant, K is stress intensity and m ranges from 2 to 24 for a variety of materials and environments. Only the salt film dissolution model was found to quantitatively describe this stage; however, the model was only tested on one material and its general applicability is unknown.
Date: December 1, 1993
Creator: Jones, R. H. & Simonen, E. P.
Partner: UNT Libraries Government Documents Department

Crack-tip chemistry modeling of stage I stress corrosion cracking

Description: Stage I stress corrosion cracking usually exhibits a very strong K dependence with Paris law exponents of up to 30. 2 Model calculations indicate that the crack velocity in this regime is controlled by transport through a salt film and that the K dependence results from crack opening controlled salt film dissolution. An ionic transport model that accounts for both electromigration through the resistive salt film and Fickian diffusion through the aqueous solution was used for these predictions. Predicted crack growth rates are in excellent agreement with measured values for Ni with P segregated to the grain boundaries and tested in IN H{sub 2}SO{sub 4} at +900 mV. This salt film dissolution may be applicable to stage I cracking of other materials.
Date: October 1, 1991
Creator: Jones, R. H. & Simonen, E. P.
Partner: UNT Libraries Government Documents Department

Irradiation-assisted stress corrosion cracking considerations at temperatures below 288{degree}C

Description: Irradiation-assisted stress corrosion cracking (IASCC) occurs above a critical neutron fluence in light-water reactor (LWR) water environments at 288 C, but very little information exists to indicate susceptibility as temperatures are reduced. Potential low-temperature behavior is assessed based on the temperature dependencies of intergranular (IG) SCC in the absence of irradiation, radiation-induced segregation (RIS) at grain boundaries and micromechanical deformation mechanisms. IGSCC of sensitized SS in the absence of irradiation exhibits high growth rates at temperatures down to 200 C under conditions of anodic dissolution control, while analysis of hydrogen-induced cracking suggests a peak crack growth rate near 100 C. Hence from environmental considerations, IASCC susceptibility appears to remain likely as water temperatures are decreased. Irradiation experiments and model predictions indicate that RIS also persists to low temperatures. Chromium depletion may be significant at temperatures below 100C for irradiation doses greater than 10 displacements per atom (dpa). Macromechanical effects of irradiation on strength and ductility are not strongly dependent on temperature below 288 C. However, temperature does significantly affect radiation effects on SS microstructure and micromechanical deformation mechanisms. The critical conditions for material susceptibility to IASCC at low temperatures may be controlled by radiation-induced grain boundary microchemistry, strain localization due to irradiation microstructure and irradiation creep processes. 39 refs.
Date: March 1, 1995
Creator: Simonen, E.P.; Jones, R.H. & Bruemmer, S.M.
Partner: UNT Libraries Government Documents Department

Measurement and modeling of radiation-induced grain boundary grain boundary segregation in stainless steels

Description: Grain boundary radiation-induced segregation (RIS) in Fe-Ni-Cr stainless alloys has been measured and modelled as a function of irradiation temperature and dose. Heavy-ion irradiation was used to produce damage levels from 1 to 20 displacements per atom (dpa) at temperatures from 175 to 550{degrees}C. Measured Fe, Ni, and Cr segregation increased sharply with irradiation dose (from 0 to 5 dpa) and temperature (from 175 to about 350{degrees}C). However, grain boundary concentrations did not change significantly as dose or temperatures were further increased. Impurity segregation (Si and P) was also measured, but only Si enrichment appeared to be radiation-induced. Grain boundary Si levels peaked at an intermediate temperature of {approximately}325{degrees}C reaching levels of {approximately}8 at. %. Equilibrium segregation of P was measured in the high-P alloys, but interfacial concentration did not increase with irradiation exposure. Examination of reported RIS in neutron-irradiated stainless steels revealed similar effects of irradiation dose on grain boundary compositional changes for both major alloying and impurity element`s. The Inverse Kirkendall model accurately predicted major alloying element RIS in ion- and neutron-irradiated alloys over the wide range of temperature and dose conditions. In addition, preliminary calculations indicate that the Johnson-Lam model can reasonably estimate grain boundary Si enrichment if back diffusion is enhanced.
Date: August 1, 1995
Creator: Bruemmer, S.M.; Charlot, L.A. & Simonen, E.P.
Partner: UNT Libraries Government Documents Department

Control of degradation of spent LWR (light-water reactor) fuel during dry storage in an inert atmosphere

Description: Dry storage of Zircaloy-clad spent fuel in inert gas (referred to as inerted dry storage or IDS) is being developed as an alternative to water pool storage of spent fuel. The objectives of the activities described in this report are to identify potential Zircaloy degradation mechanisms and evaluate their applicability to cladding breach during IDS, develop models of the dominant Zircaloy degradation mechanisms, and recommend cladding temperature limits during IDS to control Zircaloy degradation. The principal potential Zircaloy cladding breach mechanisms during IDS have been identified as creep rupture, stress corrosion cracking (SCC), and delayed hydride cracking (DHC). Creep rupture is concluded to be the primary cladding breach mechanism during IDS. Deformation and fracture maps based on creep rupture were developed for Zircaloy. These maps were then used as the basis for developing spent fuel cladding temperature limits that would prevent cladding breach during a 40-year IDS period. The probability of cladding breach for spent fuel stored at the temperature limit is less than 0.5% per spent fuel rod. 52 refs., 7 figs., 1 tab.
Date: October 1, 1987
Creator: Cunningham, M.E.; Simonen, E.P.; Allemann, R.T.; Levy, I.S. & Hazelton, R.F.
Partner: UNT Libraries Government Documents Department

Radiation-induced grain boundary segregation in austenitic stainless steels

Description: Radiation-induced segregation (RIS) to grain boundaries in Fe-Ni-Cr-Si stainless alloys has been measured as a function of irradiation temperature and dose. Heavy-ion irradiation was used to produce damage levels from 1 to 20 displacements per atom (dpa) at temperatures from 175 to 550{degrees}C. Measured Fe, Ni, and Cr segregation increased sharply with irradiation dose (from G to 5 dpa) and temperature (from 175 to about 350{degrees}C). However, grain boundary concentrations did not change significantly as dose or temperatures were further increased. Although interfacial compositions were similar, the width of radiation-induced enrichment or depletion profiles increased consistently with increasing dose or temperature. Impurity segregation (Si and P) was also measured, but only Si enrichment appeared to be radiation-induced. Grain boundary Si peaked at levels approaching 10 at% after irradiation doses to 10 dpa at an intermediate temperature of 325{degrees}C. No evidence of grain boundary silicide precipitation was detected after irradiation at any temperature. Equilibrium segregation of P was measured in the high-P alloys, but interfacial concentration did not increase with irradiation exposure. Comparisons to reported RIS in neutron-irradiated stainless steels revealed similar grain boundary compositional changes for both major alloying and impurity elements.
Date: November 1, 1994
Creator: Bruemmer, S. M.; Charlot, L. A.; Vetrano, J. S. & Simonen, E. P.
Partner: UNT Libraries Government Documents Department

Defect-solute interactions near irradiation grain boundaries

Description: Defect-solute interactions control radiation-induced segregation (RIS) to interfacial sinks, such as grain boundaries, in metallic materials. The best studied system in this regard has been austenitic stainless steels. Measurements of grain boundary composition indicate that RIS of major alloying elements are in reasonable agreement with inverse-Kirkendall predictions. The steep and narrow composition profiles are shown to result from limited back diffusion near the boundary. Subsequently, defect-solute interactions that affect the near boundary defect concentrations strongly affect RIS. The variability in measured RIS may in part be caused by grain boundary characteristics.
Date: November 1, 1993
Creator: Simonen, E. P.; Vetrano, J. S.; Heinisch, H. L. & Bruemmer, S. M.
Partner: UNT Libraries Government Documents Department

Tritium/hydrogen barrier development

Description: A review of hydrogen permeation barriers that can be applied to structural metals used in fusion power plants is presented. Both implanted and chemically available hydrogen isotopes must be controlled in fusion plants. The need for permeation barriers appears strongest in Li17-Pb blanket designs, although barriers also appear necessary for other blanket and coolant systems. Barriers that provide greater than a 1000 fold reduction in the permeation of structural metals are desired. In laboratory experiments, aluminide and titanium ceramic coatings provide permeation reduction factors, PRFS, from 1000 to over 100,000 with a wide range of scatter. The rate-controlling mechanism for hydrogen permeation through these barriers may be related to the number and type of defects in the barriers. Although these barriers appear robust and resistant to liquid metal corrosion, irradiation tests which simulate blanket environments result in very low PRFs in comparison to laboratory experiments, i.e., <150. It is anticipated from fundamental research activities that the REID enhancement of hydrogen diffusion in oxides may contribute to the lower permeation reduction factors during in-reactor experiments.
Date: June 1, 1994
Creator: Hollenberg, G. W.; Simonen, E. P.; Kalinen, G. & Terlain, A.
Partner: UNT Libraries Government Documents Department

Irradiation-assisted stress corrosion cracking of fusion reactor material

Description: Irradiation-assisted stress-corrosion cracking (IASCC) is a phenomenon produced by radiation-induced alterations in the material and environment. These alternations include radiation-induced segregation and depletion of specific elements at grain boundaries, radiation creep and hardening and radiolytic effects induced in the aqueous environment. This phenomenon has been clearly identified as an active crack growth mechanism for in-core components in fission reactor must be considered as a potential crack growth mechanism for water-cooled fusion reactors such as ITER or power reactors. The potential for IASCC phenomenon occurring in ITER structural materials is being evaluated by modeling and experiment. Results from modeling calculations for impurity segregation at ITER-relevant temperatures have been completed and suggest that this phenomenon is not likely to induce IASCC during the ITER design life. If a fusion power reactor is water cooled, IASCC is a definite concern for austenitic stainless steels. It has been clearly demonstrated with modeling and experimental measurements that Cr depletion occurs within about 1 dpa. Phosphorus and Si grain boundary segregation can also occur at this same dose and temperature but their effect on IASCC appears to be secondary to Cr depletion. Also, irradiation creep-induced crack tip strain appears to be a secondary effect. However, there are a number of unexplained observations in the literature on IASCC which may be caused by radiation damage effects other than Cr depletion or impurity segregation.
Date: August 1, 1990
Creator: Jones, R.H.; Simonen, E.P. & Bruemmer, S.M.
Partner: UNT Libraries Government Documents Department

Recommended temperature limits for dry storage of spent light water reactor Zircaloy-clad fuel rods in inert gas

Description: It is concluded that the recommendation of a single-valued temperature limit of 380/sup 0/C should be replaced by multiple limits to account for variations in fuel design, burnup level, spent fuel age, and storage cask design. A single-valued limit to account for these factors would, in some situations, impose unnecessary conservatisms and, potentially, economic penalties for utilities and storage cask vendors. The technical validity and conservatism of the CSFM model should assure acceptance by the NRC for utility and cask vendor use.
Date: May 1, 1987
Creator: Levy, I.S.; Chin, B.A.; Simonen, E.P.; Beyer, C.E.; Gilbert, E.R. & Johnson, A.B. Jr.
Partner: UNT Libraries Government Documents Department

PNL Technical Review of Pressurized Thermal Shock Issues Supplement 1: Technical Critique of the NRC Near-Term Screening Criteria

Description: Pacific Northwest Laboratory (PNL) provided a technical critique of the draft report, NRC Staff Evaluation of Pressurized Thermal Shock, dated September 13, 1982. This report provided the basis for the NRC near-term regulatory position on pressurized thermal shock {PTS) and recommended a generic screening criteria for welds in the vessel beltline region. The PNL staff concluded that the screening criteria were adequate to meet the intent of the NRC safety goal and to retain past predictions of vessel reliability. The conclusion was based on selecting the plant-specific nilductility transition reference temperature (RT{sub NDT}) in the conservative manner described within the staff report. Conservative and unconservative factors were mentioned throughout the NRC staff report. The PNL staff has listed these factors together with unknown (may be either conservative or unconservative) factors and estimated, where possible, the range in °F RT{sub NDT}. The unknown factors were so widespread that the PNL staff recommended that specific conservatisms not be reduced until the unknowns are further resolved.
Date: May 1, 1983
Creator: Pederson, L. T.; Apley, W. J.; Bian, S. H.; Pelto, P. J.; Simonen, E. P.; Simonen, F. A. et al.
Partner: UNT Libraries Government Documents Department

PNL technical review of pressurized thermal-shock issues. [PWR]

Description: Pacific Northwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the results of supporting research and concluded that none of the eight reactors would undergo vessel failure from a PTS event before several more years of operation. Operator actions, however, were often required to terminate a PTS event before it deteriorated to the point where failure could occur. Therefore, the near-term (less than one year) recommendation is to upgrade, on a site-specific basis, operational procedures, training, and control room instrumentation. Also, uniform criteria should be developed by NRC for use during future licensee analyses. Finally, it was recommended that NRC upgrade nondestructive inspection techniques used during vessel examinations and become more involved in the evaluation of annealing requirements.
Date: July 1, 1982
Creator: Pedersen, L.T.; Apley, W.J.; Bian, S.H.; Defferding, L.J.; Morgenstern, M.H.; Pelto, P.J. et al.
Partner: UNT Libraries Government Documents Department