31 Matching Results

Search Results

Advanced search parameters have been applied.

Analysis of failed nuclear plant components

Description: Argonne National Laboratory has conducted analyses of failed components from nuclear power generating stations since 1974. The considerations involved in working with and analyzing radioactive components are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in service. The failures discussed are (a) intergranular stress corrosion cracking of core spray injection piping in a boiling water reactor, (b) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressure water reactor, (c) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (d) failure of pump seal wear rings by nickel leaching in a boiling water reactor.
Date: July 1, 1992
Creator: Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Structural materials for breeder reactor cores and coolant circuits

Description: The structural components of principal interest in LMFBR cores and cooling circuits include the reactor vessel, primary and secondary piping, intermediate heat exchanger (IHX), and steam generator. Load-bearing components inside the vessel, among these the fuel cladding and duct, are also included. The operating conditions present in a fast-breeder nuclear reactor impose a number of requirements on the mechanical, physical, and neutronic properties of the materials used to construct these components.
Date: February 1, 1984
Creator: Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Effect of decontamination on aging processes and considerations for life extension

Description: The basis for a recently initiated program on the chemical decontamination of nuclear reactor components and the possible impact of decontamination on extended-life service is described. The incentives for extending plant life beyond the present 40-year limit are discussed, and the possible aging degradation processes that may be accentuated in extended-life service are described. Chemical decontamination processes for nuclear plant primary systems are summarized with respect to their corrosive effects on structural alloys, particularly those in the aged condition. Available experience with chemical cleaning processes for the secondary side of PWR steam generators is also briefly considered. Overall, no severe materials corrosion problems have been found that would preclude the use of these chemical processes, but concerns have been raised in several areas, particularly with respect to corrosion-related problems that may develop during extended service.
Date: October 1, 1987
Creator: Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Analysis of cracked core spray injection line piping from the Quad Cities Units 1 and 2 boiling water reactors

Description: Elbow assemblies and adjacent piping from the loops A and B core spray injection lines of Quad Cities Units 1 and 2 Boiling Water Reactors have been examined in order to determine the nature and causes of coolant leakages and flaw indications detected during hydrostatic tests and subsequent ultrasonic inspections. The elbow assemblies were found to contain multiple intergranular cracks in the weld heat-affected zones. The cracking was predominantly axial in orientation in the forged elbow and wedge components, whereas mixed axial and circumferential cracking was seen in the wrought piping pieces. In at least two instances, axial cracks completely penetrated the circumferential weld joining adjacent components. Based upon the observations made in the present study, the failures were attributed to intergranular stress corrosion cracking caused by the weld-induced sensitized microstructure and residual stresses present; dissolved oxygen in the reactor coolant apparently served as the corrosive species. The predominantly axial orientation of the cracks present in the forged components is believed to be related to the banded microstructure present in these components. The metallographic studies reported are supplemented by x-radiography, chemical analysis and mechanical test results, determinations of the degree of sensitization present, and measurements of weld metal delta ferrite content.
Date: December 1, 1983
Creator: Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Analysis of cracked core spray piping from the Quad Cities Unit 2 boiling water reactor

Description: The results of a metallurgical analysis of leaking cracks detected in the core spray injection piping of Commonwealth Edison Company's Quad Cities Unit 2 Boiling Water Reactor are described. The cracks were present in a welded 105/sup 0/ elbow assembly in the line, and were found to be caused by intergranular stress corrosion cracking associated with the probable presence of dissolved oxygen in the reactor cooling water and the presence of grain boundary sensitization and local residual stresses induced by welding. The failure is unusual in several respects, including the very large number of cracks (approximately 40) present in the failed component, the axial orientation of the cracks, and the fact that at least one crack completely penetrated a circumferential weld. Virtually all of the cracking occurred in forged material, and the microstructural evidence presented suggests that the orientation of the cracks was influenced by the presence of axially banded delta ferrite in the microstructure of the forged components.
Date: September 1, 1982
Creator: Diercks, D.R. & Gaitonde, S.M.
Partner: UNT Libraries Government Documents Department

Analysis of Cracked Core Spray Injection Line Piping from the Quad Cities Units 1 and 2 Boiling Water Reactors

Description: Elbow assemblies and adjacent piping from the loops A and B core spray injection lines of Quad Cities Units 1 and 2 Boiling Water Reactors have been examined in order to determine the nature and causes of coolant leakages and flaw indications detected during hydrostatic tests and subsequent ultrasonic inspections. The elbow assemblies were found to contain multiple intergranular cracks in the weld heat-affected zones. The cracking was predominantly axial in orientation in the forged elbow and wedge components, whereas mixed axial and circumferential cracking was seen in the wrought piping pieces. In at least two instances, axial cracks completely penetrated the circumferential weld joining adjacent components. Based upon the observations made in the present study, the failures were attributed to intergranular stress corrosion cracking caused by the weld-induced sensitized microstructure and residual stresses present; dissolved oxygen in the reactor coolant apparently served as the corrosive species. The predominantly axial orientation of the cracks present in the forged components is believed to be related to the banded microstructure present in these components. The metallographic studies reported are supplemented by x-radiography, chemical analysis and mechanical test results, determinations of the degree of sensitization present, and measurements of weld metal delta ferrite content.
Date: December 1983
Creator: Diercks, D. R.
Partner: UNT Libraries Government Documents Department

Failure analysis of cracked head spray piping from the Dresden Unit 2 Boiling Water Reactor

Description: Several sections of Type 304 stainless steel head spray piping, 6.25 cm (2.5 in.) in diameter, from the Dresden Unit 2 Boiling Water Reactor were examined to determine the nature and causes of coolant leakages detected during hydrostatic tests. Extensive pitting was observed on the outside surface of the piping, and three cracks, all located at a helical stripe apparently rubbed onto the outer surface of the piping, were also noted. Metallographic examination revealed that the cracking had initiated at the outer surface of the pipe, and showed it to be transgranular and highly branched, characteristic of chloride stress corrosion cracking. The surface pitting also appeared to have been caused by chlorides. A scanning electron microprobe x-ray analysis of the corrosion product in the cracks confirmed the presence of chlorides and also indicated the presence of calcium.
Date: July 1, 1983
Creator: Diercks, D.R. & Dragel, G.M.
Partner: UNT Libraries Government Documents Department

Stress-corrosion cracking susceptibility of V-15Cr-5Ti in pressurized water at 288/sup 0/C

Description: The stress-corrosion cracking susceptibility of V-15Cr-5Ti in pressurized water at 288/sup 0/C has been evaluated by means of constant extension rate tensile (CERT) tests in a refreshed autoclave system. The test environments included high-purity water as well as water containing SO/sub 4//sup 2 -/ and NO/sub 3//sup -/ impurities at a concentration of 10 wppM. Strain rates from 1 x 10/sup -6/ to 5 x 10/sup -8/ s/sup -1/ were employed, and dissolved oxygen levels ranged from <0.005 to 7.9 wppM. Test times were from 3.2 to 619 h. No stress corrosion cracking was observed under any of the test conditions. These results were analyzed using measured electrochemical potentials, available Pourbaix diagram information, and the observed oxidation behavior. 7 refs., 5 figs., 1 tab.
Date: July 1, 1987
Creator: Diercks, D.R. & Smith, D.L.
Partner: UNT Libraries Government Documents Department

Microstructural observations on time-dependent failure in weld-deposited Type 16-8-2 stainless steel

Description: Type 16-8-2 stainless steel weld metal is less resistant to rupture than Type 316 stainless steel base metal under monotonic creep conditions, but is substantially more resistant to failure under conditions of fatigue and combined creep and fatigue. Fractographic and metallographic analyses indicate that specimens from the two types of tests differ in failure modes. In monotonic-creep tests, the Type 16-8-2 weld-metal specimens failed by intergranular and interphase separation followed by microvoid growth and coalescence. Under creep-fatigue conditions, failure occurred by a mixture of matrix and interphase cracking, with substantial transgranular cracking present even for tests with a 1.8 x 10/sup 4 -/s (300-min) tension hold time per cycle. The present results indicate that the resistance of Type 16-8-2 weld metal to creep-fatigue failure is directly related to its resistance to grain-boundary and interphase cracking.
Date: January 1, 1978
Creator: Raske, D.T. & Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Failure Analysis of Cracked Head Spray Piping from the Dresden Unit 2 Boiling Water Reactor

Description: Several sections of Type 304 stainless steel head spray piping, 6.25 cm (2.5 in.) in diameter, from the Dresden Unit 2 Boiling Water Reactor were examined to determine the nature and causes of coolant leakages detected during hydrostatic tests. Extensive pitting was observed on the outside surface of the piping, and three cracks, all located at a helical stripe apparently rubbed onto the outer surface of the piping, were also noted. Metallographic examination revealed that the cracking had initiated at the outer surface of the pipe, and showed it to be transgranular and highly branched, characteristic of chloride stress corrosion cracking. The surface pitting also appeared to have been caused by chlorides. A scanning electron microprobe x-ray analysis of the corrosion product in the cracks confirmed the presence of chlorides and also indicated the presence of calcium.
Date: July 1983
Creator: Diercks, D. R. & Dragel, Gabriel M.
Partner: UNT Libraries Government Documents Department

Vanadium-base alloys for fusion reactor applications

Description: Vanadium-base alloys offer potentially significant advantages over other candidate alloys as a structural material for fusion reactor first wall/blanket applications. Although the data base is more limited than that for the other leading candidate structural materials, viz., austenitic and ferritic steels, vanadium-base alloys exhibit several properties that make them particularly attractive for the fusion reactor environment. This paper presents a review of the structural material requirements, a summary of the materials data base for selected vanadium-base alloys, and a comparison of projected performance characteristics compared to other candidate alloys. Also, critical research and development (R and D) needs are defined.
Date: October 1, 1984
Creator: Smith, D.L.; Loomis, B.A. & Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Effect of heat treatment and impurity concentration on some mechanical properties V-15Cr-5Ti alloy

Description: The effects of heat treatment and O, N, C, Si, and S impurity level on the yield strength, ductility, and fracture mode for specimens from four different heats of the V-15Cr-5Ti alloy are presented. The heat treatments for the alloy consisted of annealing as-rolled material for one hour at either 950, 1050, 1125, or 1200/sup 0/C. The total oxygen, nitrogen, and carbon impurity concentration ranged from 400 to 1200 wppm. The Si concentration ranged from 300 to 1050 wppm, and the S concentration ranged from 440 to 1100 wppm. The yield strength and ductility for the alloy, regardless of impurity concentration, exhibited minimum and maximum values, respectively, for the 1125/sup 0/C anneal. The primary mode of failure for the tensile specimens was transgranular fracture.
Date: March 1, 1986
Creator: Loomis, B.A.; Kestel, B.J. & Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Steam generator tube integrity program. Semiannual report, August 1995--March 1996

Description: This report summarizes work performed by Argonne National Laboratory on the Steam Generator Tube Integrity Program from the inception of that program in August 1995 through March 1996. The program is divided into five tasks, namely (1) Assessment of Inspection Reliability, (2) Research on ISI (in-service-inspection) Technology, (3) Research on Degradation Modes and Integrity, (4) Development of Methodology and Technical Requirements for Current and Emerging Regulatory Issues, and (5) Program Management. Under Task 1, progress is reported on the preparation of and evaluation of nondestructive evaluation (NDE) techniques for inspecting a mock-up steam generator for round-robin testing, the development of better ways to correlate burst pressure and leak rate with eddy current (EC) signals, the inspection of sleeved tubes, workshop and training activities, and the evaluation of emerging NDE technology. Under Task 2, results are reported on closed-form solutions and finite element electromagnetic modeling of EC probe response for various probe designs and flaw characteristics. Under Task 3, facilities are being designed and built for the production of cracked tubes under aggressive and near-prototypical conditions and for the testing of flawed and unflawed tubes under normal operating, accident, and severe accident conditions. In addition, crack behavior and stability are being modeled to provide guidance on test facility design, to develop an improved understanding of the expected rupture behavior of tubes with circumferential cracks, and to predict the behavior of flawed and unflawed tubes under severe accident conditions. Task 4 is concerned with the cracking and failure of tubes that have been repaired by sleeving, and with a review of literature on this subject.
Date: April 1997
Creator: Diercks, D. R.; Bakhtiari, S. & Chopra, O. K.
Partner: UNT Libraries Government Documents Department

Overview of steam generator tube degradation and integrity issues

Description: The degradation of steam generator tubes in pressurized water nuclear reactors continues to be a serious problem. Primary water stress corrosion cracking is commonly observed at the roll transition zone at U-bends, at tube denting locations, and occasionally in plugs and sleeves. Outer-diameter stress corrosion cracking and intergranular attack commonly occur near the tube support plate crevice, near the tube sheet in crevices or under sludge piles, and occasionally in the free span. A particularly troubling recent trend has been the increasing occurrence of circumferential cracking at the RTZ on both the primary and secondary sides. Segmented axial cracking at the tubes support plate crevices is also becoming more common. Despite recent advances in in-service inspection technology, a clear need still exists for quantifying and improving the reliability of in- service inspection methods with respect to the probability of detection of the various types of flaws and their accurate sizing. Improved inspection technology and the increasing occurrence of such degradation modes as circumferential cracking, intergranular attack, and discontinuous axial cracking have led to the formulation of a new performance-based steam generator rule. This new rule would require the development and implementation of a steam generator management program that monitors tube condition against accepted performance criteria to ensure that the tubes perform the required safety function over the next operating cycle. The new steam generator rule will also be applied to severe accident conditions to determine the continued serviceability of a steam generator with degraded tubes in the event of a severe accident. Preliminary analyses are being performed for a hypothetical severe accident scenario to determine whether failure will occur first in the steam generator tubes, which would lead to containment bypass, or instead in the hot leg nozzle or surge line, which would not.
Date: October 1996
Creator: Diercks, D. R.; Shack, W. J. & Muscara, J.
Partner: UNT Libraries Government Documents Department

Crack growth behavior of candidate waste container materials in simulated underground water

Description: Fracture-mechanics crack growth tests were conducted on 25.4-mm-thick compact tension specimens of Types 304L and 316L Stainless steel and Incoloy 825 at 93{degrees}C and 1 atmosphere of pressure in simulated J-13 well water, which is representative of the groundwater at the Yucca Mountain site in Nevada that is proposed for a high-level nuclear waste repository. Crack growth rates were measured under various load conditions: load ratios of 0.2--1.0, frequencies of 2 {times} 10{sup {minus}4}{minus}1 Hz, rise times of 1--5000 s, and peak stress intensities of 25--40 MPa{center_dot}m{sup {1/2}}. The measured crack growthrates are bounded by the predicted rates from the current ASME Section 11 correlation for fatigue crack growth rates of austenitic stainless steel in air. Environmentally accelerated crack growth was not evident in any of the three materials under the test conditions investigated.
Date: December 31, 1992
Creator: Park, J.Y.; Shack, W.J. & Diercks, D.R.
Partner: UNT Libraries Government Documents Department

TMI-2 Vessel Investigation Project (VIP) Metallurgical Program

Description: The TMI-2 Vessel Investigation Project (VIP) Metallurgical Program is a part of the international TMI-2 Vessel Investigation Project being conducting jointly by the US Nuclear Regulatory Commission and the Organization for Economic Co-operation and Development (OECD). The overall project consists of three phases, namely (1) recovery of material samples from the lower head of the TMI-2 reactor, (2) examination and analysis of the lower head samples and the preparation and testing of archive material subjected to a similar thermal history, and (3) procurement, examination, and analysis of companion core material located adjacent to or near the lower head material. The specific objectives of the ANL Metallurgical Program, which comprises a major portion of Phase 2, are to prepare metallographic and mechanical test specimen blanks from the TMI-2 lower head material, prepare similar test specimen blanks from suitable archive material subjected to the appropriate thermal processing, determine the mechanical properties of the lower vessel head and archive materials under the conditions of the core-melt accident, and assess the lower head integrity and margin-to-failure during the accident. The ANL work consists of three tasks: (1) archive materials program, (2) fabrication of metallurgical and mechanical test specimens from the TMI-2 pressure vessel samples, and (3) mechanical property characterization of TMI-2 lower pressure vessel head and archive material.
Date: June 1, 1990
Creator: Diercks, D.R. & Neimark, L.A.
Partner: UNT Libraries Government Documents Department

Analysis of potential for jet-impingement erosion from leaking steam generator tubes during severe accidents.

Description: This report summarizes analytical evaluation of crack-opening areas and leak rates of superheated steam through flaws in steam generator tubes and erosion of neighboring tubes due to jet impingement of superheated steam with entrained particles from core debris created during severe accidents. An analytical model for calculating crack-opening area as a function of time and temperature was validated with tests on tubes with machined flaws. A three-dimensional computational fluid dynamics code was used to calculate the jet velocity impinging on neighboring tubes as a function of tube spacing and crack-opening area. Erosion tests were conducted in a high-temperature, high-velocity erosion rig at the University of Cincinnati, using micrometer-sized nickel particles mixed in with high-temperature gas from a burner. The erosion results, together with analytical models, were used to estimate the erosive effects of superheated steam with entrained aerosols from the core during severe accidents.
Date: May 1, 2002
Creator: Majumdar, S.; Diercks, D. R.; Shack, W. J. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Nuclear power plant Generic Aging Lessons Learned (GALL). Main report and appendix A

Description: The purpose of this generic aging lessons learned (GALL) review is to provide a systematic review of plant aging information in order to assess materials and component aging issues related to continued operation and license renewal of operating reactors. Literature on mechanical, structural, and thermal-hydraulic components and systems reviewed consisted of 97 Nuclear Plant Aging Research (NPAR) reports, 23 NRC Generic Letters, 154 Information Notices, 29 Licensee Event Reports (LERs), 4 Bulletins, and 9 Nuclear Management and Resources Council Industry Reports (NUMARC IRs) and literature on electrical components and systems reviewed consisted of 66 NPAR reports, 8 NRC Generic Letters, 111 Information Notices, 53 LERs, 1 Bulletin, and 1 NUMARC IR. More than 550 documents were reviewed. The results of these reviews were systematized using a standardized GALL tabular format and standardized definitions of aging-related degradation mechanisms and effects. The tables are included in volume s 1 and 2 of this report. A computerized data base has also been developed for all review tables and can be used to expedite the search for desired information on structures, components, and relevant aging effects. A survey of the GALL tables reveals that all ongoing significant component aging issues are currently being addressed by the regulatory process. However, the aging of what are termed passive components has been highlighted for continued scrutiny. This document is Volume 1, consisting of the executive summary, summary and observations, and an appendix listing the GALL literature review tables.
Date: December 1996
Creator: Kaza, K. E.; Diercks, D. R.; Holland, J. W. & Choi, S. U.
Partner: UNT Libraries Government Documents Department

Research perspectives on the evaluation of steam generator tube integrity.

Description: Industry effects have been largely successful in managing degradation of steam generator tubes due to wastage, pitting, and denting, but fretting, SCC and intergranular attack have proved more difficult to manage. Although steam generator replacements are proceeding there is substantial industry interest in operating with degraded steam generators, and significant numbers of plants will continue to do so. In most cases degradation of steam generator tubing by stress corrosion cracking is still managed by plug or repair on detection, because current NDE techniques for characterization of flaws are not accurate enough to permit continued operation. This paper reviews some of the historical background that underlies current steam generator degradation management strategies and outlines some of the additional research that must be done to provide more effective management of degradation in current generators and provide greater assurance of satisfactory performance in replacement steam generators.
Date: February 22, 2001
Creator: Muscara, J.; Diercks, D. R.; Majumdar, S.; Kupperman, D. S.; Bakhtiari, S. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Nuclear power plant Generic Aging Lessons Learned (GALL). Appendix B

Description: The purpose of this generic aging lessons learned (GALL) review is to provide a systematic review of plant aging information in order to assess materials and component aging issues related to continued operation and license renewal of operating reactors. Literature on mechanical, structural, and thermal-hydraulic components and systems reviewed consisted of 97 Nuclear Plant Aging Research (NPAR) reports, 23 NRC Generic Letters, 154 Information Notices, 29 Licensee Event Reports (LERs), 4 Bulletins, and 9 Nuclear Management and Resources Council Industry Reports (NUMARC IRs) and literature on electrical components and systems reviewed consisted of 66 NPAR reports, 8 NRC Generic Letters, 111 Information Notices, 53 LERs, 1 Bulletin, and 1 NUMARC IR. More than 550 documents were reviewed. The results of these reviews were systematized using a standardized GALL tabular format and standardized definitions of aging-related degradation mechanisms and effects. The tables are included in volume s 1 and 2 of this report. A computerized data base has also been developed for all review tables and can be used to expedite the search for desired information on structures, components, and relevant aging effects. A survey of the GALL tables reveals that all ongoing significant component aging issues are currently being addressed by the regulatory process. However, the aging of what are termed passive components has been highlighted for continued scrutiny. This report consists of Volume 2, which consists of the GALL literature review tables for the NUMARC Industry Reports reviewed for the report.
Date: December 1996
Creator: Kasza, K. E.; Diercks, D. R.; Holland, J. W. & Choi, S. U.
Partner: UNT Libraries Government Documents Department

Failure behavior of internally pressurized flawed and unflawed steam generator tubing at high temperatures -- Experiments and comparison with model predictions

Description: This report summarizes experimental work performed at Argonne National Laboratory on the failure of internally pressurized steam generator tubing at high temperatures ({le} 700 C). A model was developed for predicting failure of flawed and unflawed steam generator tubes under internal pressure and temperature histories postulated to occur during severe accidents. The model was validated by failure tests on specimens with part-through-wall axial and circumferential flaws of various lengths and depths, conducted under various constant and ramped internal pressure and temperature conditions. The failure temperatures predicted by the model for two temperature and pressure histories, calculated for severe accidents initiated by a station blackout, agree very well with tests performed on both flawed and unflawed specimens.
Date: March 1998
Creator: Majumdar, S.; Shack, W. J.; Diercks, D. R.; Mruk, K.; Franklin, J. & Knoblich, L.
Partner: UNT Libraries Government Documents Department

Crack-growth-rate testing of candidate waste container materials

Description: Fracture-mechanics crack growth tests were conducted on 25.4-mm-thick compact tension specimens of Types 304L and 316L stainless steel (SS) and Incoloy 825 at 93{degree}C and 1 atmosphere of pressure is simulated J-13 well water, which is representative of the groundwater at the Yucca Mountain site in Nevada that is proposed for a high-level nuclear waste repository. Crack growth rates were measured under various load conditions: load ratios (R) of 0.5--1.0, frequencies of 10{sup {minus}3}{minus}1 Hz, rise times of 1--1000 s, and peak stress intensities of 25--40 MPa{center_dot}m{sup 1/2}. The measured crack growth rates are bounded by the predicted rates from the current ASME Section XI correlation for fatigue crack growth rates of austenitic stainless steel in air. Environmentally accelerated crack growth was not evident in any of the three materials under the test conditions investigated.
Date: December 31, 1991
Creator: Park, J.Y.; Shack, W.J. & Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Proceedings of the USNRC/EPRI/ANL heated crevice seminar.

Description: An international Heated Crevice Seminar, sponsored by the Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Argonne National Laboratory, and the Electric Power Research Institute, was held at Argonne National Laboratory on October 7-11, 2002. The objective of the seminar was to provide a working forum for the exchange of information by contributing experts on current issues related to corrosion in heated crevices, particularly as it relates to the integrity of PWR steam generator tubes. Forty-five persons from six countries attended the seminar, including representatives from government agencies, private industry and consultants, government research laboratories, nuclear vendors, and electrical utilities. The seminar opened with keynote talks on secondary-side crevice environments associated with IGA and IGSCC of mill-annealed Alloy 600 steam generator tubes and the submodes of corrosion in heat transfer crevices. This was followed by technical sessions on (1) Corrosion in Crevice Geometries, (2) Experimental Methods, (3) Results from Experimental Studies, and (4) Modeling. The seminar concluded with a panel discussion on the present understanding of corrosive processes in heated crevices and future research needs.
Date: August 31, 2003
Creator: Park, J. Y.; Fruzzetti, K.; Muscara, J.; Diercks, D. R.; Technology, Energy; EPRI et al.
Partner: UNT Libraries Government Documents Department