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Corrosion of stainless and carbon steels in molten mixtures of industrial nitrates

Description: Corrosion behavior of two stainless steels and carbon steel in mixtures of NaNO{sub 3} and KNO{sub 3} was evaluated to determine if impurities found in commodity grades of alkali nitrates aggravate corrosivity as applicable to an advanced solar thermal energy system. Corrosion tests were conducted for 7000 hours with Types 304 and 316 stainless steels at 570C and A36 carbon steel at 316C in seven mixtures of NaNO{sub 3} and KNO{sub 3} containing variations in impurity concentrations. Corrosion tests were also conducted in a ternary mixture of NaNO{sub 3}, KNO{sub 3}, and Ca(NO{sub 3}){sub 2}. Corrosion rates were determined by descaled weight losses while oxidation products were examined by scanning electron microscopy, electron microprobe analysis, and X-ray diffraction. The nitrate mixtures were periodically analyzed for changes in impurity concentrations and for soluble corrosion products.
Date: March 1, 1994
Creator: Goods, S. H.; Bradshaw, R. W.; Prairie, M. R. & Chavez, J. M.
Partner: UNT Libraries Government Documents Department

304L stainless steel resistance to cesium chloride

Description: B and W Hanford Company have two Oak Ridge National Laboratory (ORNL) Type 4 canisters filled with cesium chloride (CsCl) originally produced at WESF (Waste Encapsulation and Storage Facility). These canisters are constructed of 304L stainless steel per drawing ORNL 970-294. Instead of removing the CsCl from the Type 4 canisters and repacking into an Inner Capsule, it is intended (for ALARA, schedule and cost purposes) that the Type 4 canisters be decontaminated (scrubbed) and placed [whole] inside a Type ``W`` overpack. The overpack is constructed from 316L stainless steel. Several tests have been run by Pacific Northwest National Laboratory (PNNL) over the. years documenting the corrosion compatibility of 316L SS with CsCl (Bryan 1989 and Fullam 1972). However, no information for 304L SS compatibility is readily available. This document estimates the corrosion resistance of 304L stainless steel in a WESF CsCl environment as it compares with that of 316L stainless steel.
Date: August 27, 1998
Creator: Graves, C. E.
Partner: UNT Libraries Government Documents Department

A study of a container for long term storage of high level waste using finite elements. Task B.1, Structural analysis and design: The Waste Package Project at The University of Nevada, Las Vegas

Description: Designs of selected components for a container are evaluated based on conservative loads and assumptions. The items selected for evaluation are a pintle and possible container topheads. An existing pintle for use in another application is evaluated under a more severe axial load. The results show that the pintle is adequate with the existing design to a safety factor of at least three. Improvements are suggested which raises this safety factor to approximately six. Several flat tophead designs are evaluated for stress under an annular load. The parameter selected for evaluation is the thickness of the circular plate. Results from a 1.5 inch thick plate are below a safety factor of three. Results with a 2 inch plate are improved, but marginal. Analysis results indicate that a thicker pintle or a stiffer plate will improve the design. Several curved tophead shapes are analyzed under an annular load. The ASME flanged and dished shape displays more desirable properties than others. A critical parameter is identified and changed to provide acceptable stress levels.
Date: June 1, 1992
Creator: Ladkany, S. G. & Kniss, B. R.
Partner: UNT Libraries Government Documents Department

Fatigue and environmentally assisted cracking in light water reactors

Description: Fatigue and stress corrosion cracking (SCC) for low-alloy steel used in piping and in steam generator and reactor pressure vessels have been investigated. Fatigue data were obtained on medium-sulfur-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor water, and in air. Analytical studies focused on the behavior of carbon steels in boiling water reactor (BWR) environments. Crack-growth rates of composite fracture-mechanics specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B steel were determined under small-amplitude cyclic loading in HP water with {approx}300 pbb dissolved oxygen. Radiation-induced segregation and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence also have been investigated. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain-rate tensile tests were conducted on tubular specimens in air and in simulated BWR water at 289{degrees}C.
Date: March 1, 1992
Creator: Kassner, T. F.; Ruther, W. E.; Chung, H. M.; Hicks, P. D.; Hins, A. G.; Park, J. Y. et al.
Partner: UNT Libraries Government Documents Department

Fatigue of ferritic and austenitic steels. Final technical report, June 1, 1984--July 31, 1991

Description: One aim of this research program has been to increase our understanding of the mechanisms of fatigue crack growth as influenced by such parameters as the specific material, crack length, R-ratio, overloads, temperature and the environment. A second objective has been to utilize this understanding in the development of semi-empirical quantitative methods for the prediction of fatigue crack growth behavior. Although there are still many questions remaining concerning fatigue crack growth, we feel that some significant progress in meeting these two objectives has been made.
Date: December 31, 1991
Creator: McEvily, A. J.
Partner: UNT Libraries Government Documents Department

In-reactor corrosion: A paper presented at the 9th annual AEC Corrosion Symposium, Boston, Massachusetts, May 10--12, 1960

Description: Object of this paper is to present preliminary results of experiments in Hanford in-reactor loops to determine if exposure to neutrons will increase corrosion rates of Al alloys, Zy-2, and 304 stainless steel. Results were negligible or no corrosion.
Date: May 9, 1960
Creator: Larrick, A. P.
Partner: UNT Libraries Government Documents Department

Application of in situ x-ray absorption and fluorescence measurements to analyze solutions in a simulated pit

Description: X-ray energy-dispersive spectroscopy has been used to study the compositions of metal ions in solutions developed during localized corrosion. An electrochemical cell was designed to simulate a corrosion pit, maintaining one-dimensional diffusion and fulfilling the requirements for x-ray fluorescence measurements. The working electrode consisted of a dissolving thin foil of Type 304 stainless steel sealed between Mylar sheets through which the x-ray beam passed. Concentration gradients within the artificial pit were quantitatively determined.
Date: December 31, 1991
Creator: Isaacs, H. S.; Davenport, A. J.; Cho, J. H.; Rivers, M. L. & Sutton, S. R.
Partner: UNT Libraries Government Documents Department

Corrosion fatigue of iron-chromium-nickel alloys: Fracture mechanics, microstructure and chemistry. Progress report, January 1, 1992--December 31, 1992

Description: Phase transformation and cracking during RT aging of charged, high-purity Fe18Cr12Ni alloy and commerical 304 ss were examined; results show that {epsilon}* (hcp) hydride formed on Fe18Cr12Ni upon charging, and it decomposed rapidly to form first {epsilon} and then {alpha}` martensite. Morphology of fracture surfaces of Fe18Cr12Ni produced by corrosion fatigue in NaCl solutions and in hydrogen was found to be identical. Effort was made to examine the approaches and methodologies used in service life predictions and reliability analyses.
Date: January 25, 1993
Creator: Wei, R. P.
Partner: UNT Libraries Government Documents Department

Inelastic deformation and damage at high temperature. Progress report, April 1, 1991--March 31, 1992

Description: Combined experimental and theoretical investigations into the inelastic deformation and damage behavior of engineering alloys at elevated temperatures are being pursued. The analysis of previously performed strain rate change and relaxation tests on modified 9Cr-1Mo steel showed the need for inclusion of a recovery of state term in the growth laws for the state variables of the viscoplasticity theory based on overstress (VBO). Recovery of state terms were introduced and the experimental results were satisfactorily simulated. The finite deformation theory of VBO has been developed further to include a convected derivative rationale for the choice of the objective stress rate. The reversing direct current voltage drop measurements during low cycle fatigue at elevated temperature were improved. A passive filter bank and new positioning devices for the coils were installed. Tests at 650{degrees}C and lasting several days showed excessive, uncontrollable temperature changes. It was decided to drop the test temperature to 538{degrees}C which is close to the operating temperature of Type 304 Stainless Steel. The temperature fluctuations in torsion tests were within {plus_minus}3{degrees}C which was considered satisfactory. Testing will continue at 538{degrees}C.
Date: May 1, 1992
Creator: Krempl, E.
Partner: UNT Libraries Government Documents Department

Acceptance criteria for heat exchanger head staybolts

Description: Each of the six primary coolant loop systems of the Savannah River Site production reactors contains two parallel single-pass heat exchangers to transfer heat from the primary coolant (D{sub 2}O) to the secondary cooling water (H{sub 2}O). The configuration of the heat exchangers includes a plenary space defined by the heat exchanger tubesheet and the heat exchanger head at both the heat exchanger inlet and outlet to the primary piping. The primary restraint of the heat exchanger head (Type 304 stainless steel) is provided by 84 staybolts (Type 303 stainless steel) which attach to the tubesheet. The staybolts were cap seal-welded in the mid-1960`s and are immersed in moderator. Access to inspect the staybolts is limited to a recently-developed ultrasonic technique shooting a beam through the staybolt assembly. Acceptance Criteria to allow disposition of flaws detected by UT inspection have been developed. The structural adequacy to protect against collapse loading of the head is demonstrated by finite element analysis of the head assembly and fracture analysis of flaw postulates in the staybolts. Both normal operation and normal operation plus seismic loading conditions were considered. Several bounding cases containing various configurations of nonactive (exceeding critical flaw size) staybolts were analyzed. The model of the head assembly can be applied to evaluate any active staybolt configurations based on the results from future inspections. 9 refs.
Date: December 31, 1991
Creator: Sindelar, R. L.; Lam, P. S.; Barnes, D. M.; Placr, A. & Morrison, J. M.
Partner: UNT Libraries Government Documents Department

Defense Waste Processing Facility: Report of task force on options to mitigate the effect of nitrite on DWPF operations. Savannah River Site 200-S Area

Description: The possibility of accumulating ammonium nitrate (an explosive) as well as organic compounds in the DWPF Chemical Processing Cell Vent System was recently discovered. A task force was therefore organized to examine ways to avoid this potential hazard. Of thirty-two processing/engineering options screened, the task force recommended five options, deemed to have the highest technical certainty, for detailed development and evaluation: Radiolysis of nitrite in the tetraphenylborate precipitate slurry feed in a new corrosion-resistant facility. Construction of a Late Washing Facility for precipitate washing before transfer to the DWPF; ``Just-in-Time`` precipitation; Startup Workaround by radiolysis of nitrite in the existing corrosion-resistant Pump Pit tanks; Ammonia venting and organics separation in the DWPF; and, Estimated costs and schedules are included in this report.
Date: March 1, 1992
Creator: Randall, D. & Marek, J. C.
Partner: UNT Libraries Government Documents Department

Development of J-integral based UT flaw acceptance criteria for Savannah River reactor tanks

Description: The tank of a Savannah River Site reactor is a cylinder approximately 16 feet in diameter and 14 feet high and is not pressurized except for a 5 psig helium blanket gas in addition to the hydrostatic head of the heavy water (D{sub 2}O) moderator. The tank is made of Type 304 Stainless steel fabricated into cylindrical shells with 0.5 inch thick four to six wrought plates per vessel. The shell was made up in two flat half-sections for later rolling and welding. The tank bottom section containing the moderator effluent nozzles was welded to the shell in a T-Joint configuration. An ultrasonic (UT) in-service inspection program has been developed for the examination of these reactor tanks. Prior to the initiation of these inspections, criteria for the disposition of any indications that may be found were required. This paper describes the fracture mechanics evaluations that formed the technical bases for the flaw acceptance criteria. The fracture mechanics evaluation considered detailed finite element calculated stress states in the various regions of the tanks, measured irradiated fracture toughness properties (irradiated condition material J-R curves), fluence levels in the tanks, and intergranular stress corrosion cracking growth rates in the reactor environment. The irradiation program was conducted at the High Flux Isotope Reactor at Oak Ridge National Laboratory. Tensile and fracture toughness data were obtained from 36 compact tension, tensile and Charpy V-notch specimens. Through wall cracks are postulated in various regions of the tank and critical crack lengths are calculated by the elastic-plastic fracture mechanics based J-Integral/Tearing Modulus (J/T) approach. The applied values of J-integral were calculated by the well-known GE/EPRI estimation scheme. Acceptable crack lengths are then calculated following generally accepted safety factors based on the ASME Pressure Vessel Code. The acceptance criteria are then briefly described.
Date: December 31, 1991
Creator: Mehta, H. S.; Ranganath, S.; Awadalla, N. G.; Sindelar, R. L.; Caskey, G. R. Jr. & Daugherty, W. L.
Partner: UNT Libraries Government Documents Department

Materials performance in the atmospheric fluidized-bed cogeneration air heater experiment

Description: The Atmospheric Fluidized-Bed Cogeneration Air Heater Experiment (ACAHE) sponsored by the US Department of Energy (DOE) was initiated to assess the performance of various heat-exchanger materials to be used in fluidized-bed combustion air heater systems. Westinghouse Electric Corporation, through subcontracts with Babcock & Wilcox, Foster Wheeler, and ABB Combustion Engineering Systems, prepared specifications and hardware for the ACAHE tests. Argonne National Laboratory contracted with Rockwell International to conduct tests in the DOE atmospheric fluidized-bed combustion facility. This report presents an overview of the project, a description of the facility and the test hardware, the test operating conditions, a summary of the operation, and the results of analyzing specimens from several uncooled and cooled probes exposed in the facility. Extensive microstructural analyses of the base alloys, claddings, coatings, and weldments were performed on specimens exposed in several probes for different lengths of time. Alloy penetration data were determined for several of the materials as a function of specimen orientation and the exposure location in the combustor. Finally, the data were compared with earlier laboratory test data, and the long-term performance of candidate materials for air-heater applications was assessed.
Date: February 1, 1991
Creator: Natesan, K.; Podolski, W.; Wang, D. Y.; Teats, F. G.; Gerritsen, W.; Stewart, A. et al.
Partner: UNT Libraries Government Documents Department

Bowing-reactivity trends in EBR-II assuming zero-swelling ducts

Description: Predicted trends of duct-bowing reactivities for the Experimental Breeder Reactor II (EBR-II) are correlated with predicted row-wise duct deflections assuming use of idealized zero-void-swelling subassembly ducts. These assume no irradiation induced swellings of ducts but include estimates of the effects of irradiation-creep relaxation of thermally induced bowing stresses. The results illustrate the manners in which at-power creeps may affect subsequent duct deflections at zero power and thereby the trends of the bowing component of a subsequent power reactivity decrement.
Date: March 1, 1994
Creator: Meneghetti, D.
Partner: UNT Libraries Government Documents Department

Inspection indications, stress corrosion cracks and repair of process piping in nuclear materials production reactors

Description: Ultrasonic inspection of Schedule 40 Type 304 stainless steel piping in the process water system of the Savannah River Site reactors has provided indications of discontinuities in less than 10% of the weld heat affected zones. Pipe sections containing significant indications are replaced with Type 304L components. Post removal metallurgical evaluation showed that the indications resulted from stress corrosion cracking in weld heat-affected zones and that the overall weld quality was excellent. The evaluation also revealed weld fusion zone discontinuities such as incomplete penetration, incomplete fusion, inclusions, underfill at weld roots and hot cracks. Service induced extension of these discontinuities was generally not significant although stress corrosion cracking in one weld fusion zone was noted. One set of UT indications was caused by metallurgical discontinuities at the fusion boundary of an extra weld. This extra weld, not apparent on the outer pipe surface, was slightly overlapping and approximately parallel to the weld being inspected. This extra weld was made during a pipe repair, probably associated with initial construction processes. The two nearly parallel welds made accurate assessment of the UT signal difficult. The implications of these observations to the inspection and repair of process water systems of nuclear reactors is discussed.
Date: December 31, 1991
Creator: Louthan, M. R. Jr.; West, S. L. & Nelson, D. Z.
Partner: UNT Libraries Government Documents Department

Two-step chemical decontamination technology

Description: An improved two-step chemical decontamination technique was recently developed at INEL. This memorandum documents the addition of this technology to the SRTC arsenal of decontamination technology. A two-step process using NAOH, KMnO{sub 4} followed by HNO{sub 3} was used for cleaning doorstops (small casks) in the SRTC High Level Caves in 1967. Subsequently, more aggressive chemical techniques have been found to be much more effective for our applications. No further work on two-step technology is planned.
Date: August 14, 1992
Creator: Rankin, W. N.
Partner: UNT Libraries Government Documents Department

Corrosion of 304L and 316 in gadolinium nitrate neutron poison solutions

Description: Pitting corrosion has occurred on AISI Type 304L stainless steel (304L) conductivity probes used to monitor liquid levels of gadolinium nitrate neutron poison solutions (GPS). An electrochemical and immersion test program has led to a better understanding of the cause of corrosion of 304L probes. Results indicate that the alternating voltage applied to the probes to monitor contact with solution is the primary factor in the corrosion of the probes. A chloride-containing dye and low pH also contribute to the corrosion process, but appear to play a secondary role. AISI Type 316 stainless steel (316) was found to behave similarly to 304L in GPS, while nickel-based alloys such as Hastelloy G30, Hastelloy C22, and Inconel 625 were found to be more susceptible to corrosion as compared to 304L.
Date: December 31, 1991
Creator: Chandler, G. T. & Anderson, M. H.
Partner: UNT Libraries Government Documents Department

Tensile and burst tests in support of the cadmium safety rod failure evaluation

Description: The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of hypothetical LOCA event. Accordingly, an experimental safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. This report discusses confirmatory separate effects tests conducted to support the evaluation of failures observed in the safety rod thermal tests. As part of the failure evaluation, the potential for liquid metal embrittlement (LME) of the safety rod cladding by cadmium (Cd) -- aluminum (Al) solutions was examined. Based on the test conditions, literature data, and U-Bend tests, its was concluded that the SS304 safety rod cladding would not be subject to LME by liquid Cd-Al solutions under conditions relevant to the safety rod thermal tests or gamma heating accident. To confirm this conclusion, tensile tests on SS304 specimens were performed in both air and liquid Cd-Al solutions with the range of strain rates, temperatures, and loading conditions spanning the range relevant to the safety rod thermal tests and gamma heating accident.
Date: February 1, 1992
Creator: Thomas, J. K.
Partner: UNT Libraries Government Documents Department

Overview of the Savannah River reactor surveillance program

Description: Preliminary radiation effects studies, now nearing completion, are based upon short-term irradiations in the High Flux Isotope Reactor (HFIR) and University of Buffalo Reactor (UBR). The purpose of these studies is to assess the impact of radiation effects on the performance and service life of Savannah River Site (SRS) reactor components. The SRS reactor surveillance program was simultaneously established in order to validate the results of these short-term irradiations under conditions more applicable to the SRS reactor components. The main conditions of interest include: temperature, moderator chemistry and neutron spectrum. The materials employed in both the surveillance and short-term irradiation programs were taken from the SRS R-reactor process piping and included base, weld, and heat-effected zone metal samples in both the C-L and L-C orientations. Test specimen types include tensile, Charpy V-notch, compact tension, constant extension rate test, and wedge-opening-load. A program schedule has been established for the irradiation and testing of surveillance specimens. The surveillance capsules are currently undergoing irradiation near the center of the SRS K-rector core and will remain in this location until their total fast fluence is equal to the current maximum reactor tank wall exposure. The bulk of the specimens will then be moved to blanket capsules at the edge of the core in order to track the wall exposure. A small portion of the specimens will be removed and tested, and a third group of specimens will be removed from the central core position after they have accumulated a thermal fluence 50% greater than the current maximum tank wall exposure. Specimen groups will be removed from the blanket capsules and tested periodically. Additional capsules may be inserted in order to generate flow fluence or high thermal to fast flux ratio data. 17 refs.
Date: December 31, 1991
Creator: Sindelar, R. L.; Thomas, J. K. & Baumann, N. P.
Partner: UNT Libraries Government Documents Department

Effects of temperature and radiation on the nuclear waste glass product consistency leach test

Description: Previous leach studies carried out with monolithic glass samples have shown that glass dissolution rates increase with increasing temperature and may or may not increase on exposure to external gamma radiolysis. In this study we have investigated the effects of temperature (70--1200{degrees}C) and radiation on the dissolution of simulated radioactive waste glasses using the Product Consistency Test (PCT). The PCT is a seven day, crushed glass leach test in deionized water that is carried out at 9OO{degrees}C. To date our results indicate no significant effect of external Co--60 gamma radiation when testing various simulated waste glasses at 90{degrees}C in a wellinsulated compartment within a Gammacell 220 irradiation unit. The temperature dependence for glass dissolution clearly exhibits Arrheniustype behavior for two of the three glasses tested. For the third glass the dissolution decreases at the higher temperatures, probably due to saturation effects. Actual radioactive waste glasses will be investigated later as part of this study.
Date: April 1, 1993
Creator: Crawford, C. L. & Bibler, N. E.
Partner: UNT Libraries Government Documents Department

Double-ended break probability estimate for the 304 stainless steel main circulation piping of a production reactor

Description: The large break frequency resulting from intergranular stress corrosion cracking in the main circulation piping of the Savannah River Site (SRS) production reactors has been estimated. Four factors are developed to describe the likelihood that a crack exists that is not identified by ultrasonic inspection, and that grows to instability prior to growing through-wall and being detected by the ensuing leakage. The estimated large break frequency is 3.4 {times} 10{sup {minus}8} per reactor-year.
Date: 1991
Creator: Mehta, H. S.; Daugherty, W. L.; Awadalla, N. G. & Sindelar, R. L.
Partner: UNT Libraries Government Documents Department

The effects of tritium and decay helium on the fracture toughness properties of stainless steels

Description: J-integral fracture mechanics techniques and scanning electron microscopy observations were used to investigate the effects of tritium and its decay product, helium-3, on Types 304L, 316L, 21-6-9, A286, and JBK-75 (Modified A286) stainless steels. Tritium-exposed samples of each steel had lower fracture toughness values and less resistance to stable crack growth than control samples. Type 316L stainless steel was more resistant to the embrittling effects of tritium and decay helium than the other steels.
Date: December 31, 1991
Creator: Morgan, M. J.
Partner: UNT Libraries Government Documents Department

Growth of IGSC cracks in Type 304 stainless steel at 100{degrees}C in an aqueous environment

Description: Intergranular stress corrosion (IGSC) cracking has been observed in the primary coolant system of the Savannah River Site Reactors. There have been several cases during the over one hundred reactor-years of plant operating experience when IGSC cracks have grown through-wall and minor leaks have occurred. Approximately 7% of the heat affected zones of pipe-to-pipe butt welds show indications of IGSC cracking during ultrasonic testing (UT). Other piping and component areas, sensitized by flame washing or hot forming, have also developed IGSC cracks. The entire system was fabricated in the 1950`s from Type 304 stainless steel. All joining was by the metal inert gas welding process. IGSC crack growth rates have been measured on compact tension specimens under controlled environmental conditions that encompass the observed conditions in the SRS reactor primary coolant systems. Growth rates were measured extending from less than 10{sup {minus}9} to approximately 10{sup {minus}5} millimeter per second. These growth rates bound the growth rates that have been inferred from a statistical analysis of UT indications. The UT data were collected since 1984 from weld heat affected zones in pipe-to-pipe butt welds in the SRS reactor primary coolant piping. Chloride and sulfate anions, dissolved oxygen, and peroxide have been identified as the water impurities that influence IGSC cracking in the SRS reactor primary coolant systems. A quantitative relationship has been established for susceptibility to IGSC cracking in terms of concentrations of these impurities and temperature. The heavy water reactor moderator and coolant is acidified with nitric acid to a pH of 4.7 to minimize corrosion of the aluminum cladding on the fuel elements.
Date: December 31, 1990
Creator: Caskey, G. R.; Stoner, K. J.; Daugherty, W. L.; Ondrejcin, R. S. & Postles, R. L.
Partner: UNT Libraries Government Documents Department