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Review of time-dependent fatigue behavior and life prediction for 2 1/4 Cr-1 Mo steel. [LMFBR]

Description: Available data on creep-fatigue life and fracture behavior of 2 1/4 Cr-1 Mo steel are reviewed. Whereas creep-fatigue interaction is important for Type 304 stainless steel, oxidation effects appear to dominate the time-dependent fatigue behavior of 2 1/4 Cr-1 Mo steel. Four of the currently available predictive methods - the Linear Damage Rule, Frequency Separation Equation, Strain Range Partitioning Equation, and Damage Rate Equation - are evaluated for their predictive capability. Variations in the parameters for the various predictive methods with temperature, heat of material, heat treatment, and environment are investigated. Relative trends in the lives predicted by the various methods as functions of test duration, waveshape, etc., are discussed. The predictive methods will need modification in order to account for oxidation and aging effects in the 2 1/4 Cr-1 Mo steel. Future tests that will emphasize the difference between the various predictive methods are proposed.
Date: January 1, 1982
Creator: Booker, M.K. & Majumdar, S.
Partner: UNT Libraries Government Documents Department

Analysis of an Internally Pressurized Prismatic Cell Can

Description: This report contains an elastic stress and displacement analysis of a prismatic cell can subjected to internal pressure. A computer program was written to perform the analysis. The results show that, for the geometry chosen, the thicknesses of the cell-can walls and the magnitude of the internal pressure are the most important parameters that determine the stresses and deformations of the cell can. Recommendations for future studies are included.
Date: April 1980
Creator: Majumdar, S.
Partner: UNT Libraries Government Documents Department

Creep-fatigue interactions in an austenitic stainless steel

Description: A phenomenological model of the interaction between creep and fatigue in Type 304 stainless steel at elevated temperatures is presented. The model is based on a crack-growth equation and an equation governing cavity growth, expressed in terms of current plastic strain and plastic strain rate. Failure is assumed to occur when a proposed interaction equation is satisfied. Various parameters of the equations can be obtained by correlation with continuously cycling fatigue and monotonic creep-rupture test data, without the use of any hold-time fatigue tests. Effects of various wave shapes such as tensile, compressive, and symmetrical hold on the low-cycle fatigue life can be computed by integrating the damage-rate equations along the appropriate loading path. Microstructural evidence in support of the proposed model is also discussed.
Date: January 1, 1978
Creator: Majumdar, S. & Maiya, P.S.
Partner: UNT Libraries Government Documents Department

Some stress-related issues in tokamak fusion reactor first walls

Description: Recent design studies of a tokamak fusion power reactor and of various blankets have envisioned surface heat fluxes on the first wall ranging from 0.1 to 1.0 MW/m/sup 2/, and end-of-life irradiation fluences ranging from 100 dpa for the austenitic stainless steels to as high as 250 dpa for postulated vanadium alloys. Some tokamak blankets, particularly those using helium or liquid metal as coolant/breeder, may have to operate at relatively high coolant pressures so that the first wall may be subjected to high primary stress in addition to high secondary stresses such as thermal stresses or stresses due to constrained swelling. The present paper focusses on the various problems that may arise in the first wall because of stress and high neutron fluence, and discusses some of the design solutions that have been proposed to overcome these problems.
Date: January 1, 1987
Creator: Majumdar, S.; Pai, B. & Ryder, R.H.
Partner: UNT Libraries Government Documents Department

Effects of high mean stress on the high-cycle fatigue behavior of PWA 1480

Description: PWA 1480 is a potential candidate material for use in the high-pressure fuel turbine blade of the Space Shuttle Main Engine. As an engine material it will be subjected to high-cycle fatigue loading superimposed on a high mean stress due to combined centrifugal and thermal loadings. This paper describes results obtained in an ongoing program to determine the effects of a high mean stress on the high-cycle fatigue behavior of this material.
Date: March 1, 1985
Creator: Majumdar, S.; Antolovich, S. & Milligan, W.
Partner: UNT Libraries Government Documents Department

Predictions of structural integrity of steam generator tubes under normal operating, accident, and severe accident conditions

Description: Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating, accident, and severe accident conditions are reviewed. Tests conducted in the past, though limited, tended to show that the earlier flow-stress model for part-through-wall axial cracks overestimated the damaging influence of deep cracks. This observation is confirmed by further tests at high temperatures as well as by finite element analysis. A modified correlation for deep cracks can correct this shortcoming of the model. Recent tests have shown that lateral restraint can significantly increase the failure pressure of tubes with unsymmetrical circumferential cracks. This observation is confirmed by finite element analysis. The rate-independent flow stress models that are successful at low temperatures cannot predict the rate sensitive failure behavior of steam generator tubes at high temperatures. Therefore, a creep rupture model for predicting failure is developed and validated by tests under varying temperature and pressure loading expected during severe accidents.
Date: September 1, 1996
Creator: Majumdar, S.
Partner: UNT Libraries Government Documents Department

Structural evaluation of electrosleeved tubes under severe accident transients.

Description: A flow stress model was developed for predicting failure of Electrosleeved PWR steam generator tubing under severe accident transients. The Electrosleeve, which is nanocrystalline pure nickel, loses its strength at temperatures greater than 400 C during severe accidents because of grain growth. A grain growth model and the Hall-Petch relationship were used to calculate the loss of flow stress as a function of time and temperature during the accident. Available tensile test data as well as high temperature failure tests on notched Electrosleeved tube specimens were used to derive the basic parameters of the failure model. The model was used to predict the failure temperatures of Electrosleeved tubes with axial cracks in the parent tube during postulated severe accident transients.
Date: November 12, 1999
Creator: Majumdar, S.
Partner: UNT Libraries Government Documents Department

Testing and analysis of structural integrity of electrosleeved tubes under severe accident transients

Description: The structural integrity of flawed steam generator tubing with Electrosleeves{trademark} under simulated severe accident transients was analyzed by analytical models that used available material properties data and results from high-temperature tests conducted on Electrosleeved tubes. The Electrosleeve material is almost pure Ni and derives its strength and other useful properties from its nanocrystalline microstructure, which is stable at reactor operating temperatures. However, it undergoes rapid grain growth, at the high temperatures expected during severe accidents, resulting in a loss of strength and a corresponding decrease in flow stress. The magnitude of this decrease depends on the time-temperature history during the accident. Failure tests were conducted at ANL and FTI on internally pressurized Electrosleeved tubes with 80% and 100% throughwall machined axial notches in tie parent tubes that were subjected to simulated severe accident temperature transients. The test results, together with the analytical model, were used to estimate the unaged flow stress curve of the Electrosleeved material at high temperatures. Failure temperatures for Electrosleeved tubes with throughwall and part-throughwall axial cracks of various lengths in the parent tubes were calculated for a postulated severe accident transient.
Date: December 10, 1999
Creator: Majumdar, S.
Partner: UNT Libraries Government Documents Department

Structural design criteria for high heat flux components.

Description: The high temperature design rules of the ITER Structural Design Criteria (ISDC), are applied to first wall designs with high heat flux. The maximum coolant pressure and surface heat flux capabilities are shown to be determined not only by the mechanical properties of the first wall material but also by the details of the blanket design. In a high power density self-cooled lithium blanket, the maximum primary stress in the first wall is controlled by many of the geometrical parameters of the blanket, such as, first wall span, first wall curvature, first wall thickness, side wall thickness, and second wall thickness. The creep ratcheting lifetime of the first wall is also shown to be controlled by many of the same geometrical parameters as well as the coolant temperature. According to most high temperature design codes, the time-dependent primary membrane stress allowable are based on the average temperature (ignoring thermal stress). Such a procedure may sometimes be unconservative, particularly for embrittled first walls with large temperature gradients. The effect of secondary (thermal) stresses on the accumulation of creep deformation is illustrated with a vanadium alloy flat plate first wall design.
Date: July 14, 1999
Creator: Majumdar, S.
Partner: UNT Libraries Government Documents Department

Development of structural design criteria for ITER.

Description: The irradiation environment experienced by the in-vessel components of fusion reactors such as HER presents structural design challenges not envisioned in the development of existing structural design criteria such as the ASME Code or RCC-MR. From the standpoint of design criteria, the most significant issues stem from the irradiation-induced changes in material properties, specifically the reduction of ductility, strain hardening capability, and fracture toughness with neutron irradiation. Recently, Draft 7 of the interim ITER structural design criteria (ISDC), which provide new rules for guarding against such problems, was released for trial use by the ITER designers. The new rules, which were derived from a simple model based on the concept of elastic follow up factor, provide primary and secondary stress limits as functions of uniform elongation and ductility. The implication of these rules on the allowable surface heat flux on typical first walls made of type 316 stainless steel and vanadium alloys are discussed.
Date: June 22, 1998
Creator: Majumdar, S.
Partner: UNT Libraries Government Documents Department

Leakage analysis of the evolve first wall.

Description: Leakage of lithium through cracks in the first wall of EVOLVE was analyzed for two limiting cases, which are simplified versions of the real case, where the lithium enters the cracks as liquid and flashes to vapor phase within the first wall. Leakage rates were calculated for the cases of liquid lithium flow and lithium vapor flow. Inasmuch as the coolant pressure is close to the saturation pressure, the limiting case of lithium vapor flow should be closer to reality. The impact of lithium leakage on first-wall cooling and plasma contamination is discussed.
Date: March 8, 2002
Creator: Majumdar, S.
Partner: UNT Libraries Government Documents Department

Thermal stress and creep fatigue limitations in first wall design

Description: The thermal-hydraulic performance of a lithium cooled cylindrical first wall module has been analyzed as a function of the incident neutron wall loading. Three criteria were established for the purpose of defining the maximum wall loading allowable for modules constructed of Type 316 stainless steel and a vanadium alloy. Of the three, the maximum structural temperature criterion of 750/sup 0/C for vanadium resulted in the limiting wall loading value of 7 MW/m/sup 2/. The second criterion limited thermal stress levels to the yield strength of the alloy. This led to the lowest wall loading value for the Type 316 stainless steel wall (1.7 MW/m/sup 2/). The third criterion required that the creep-fatigue characteristics of the module allow a lifetime of 10 MW-yr/m/sup 2/. At wall temperatures of 600/sup 0/C, this lifetime could be achieved in a stainless steel module for wall loadings less than 3.2 MW/m/sup 2/, while the same lifetime could be achieved for much higher wall loadings in a vanadium module.
Date: January 1, 1977
Creator: Majumdar, S.; Misra, B. & Harkness, S.D.
Partner: UNT Libraries Government Documents Department

High power density self-cooled lithium-vanadium blanket.

Description: A self-cooled lithium-vanadium blanket concept capable of operating with 2 MW/m{sup 2} surface heat flux and 10 MW/m{sup 2} neutron wall loading has been developed. The blanket has liquid lithium as the tritium breeder and the coolant to alleviate issues of coolant breeder compatibility and reactivity. Vanadium alloy (V-4Cr-4Ti) is used as the structural material because it can accommodate high heat loads. Also, it has good mechanical properties at high temperatures, high neutron fluence capability, low degradation under neutron irradiation, good compatibility with the blanket materials, low decay heat, low waste disposal rating, and adequate strength to accommodate the electromagnetic loads during plasma disruption events. Self-healing electrical insulator (CaO) is utilized to reduce the MHD pressure drop. A poloidal coolant flow with high velocity at the first wall is used to reduce the peak temperature of the vanadium structure and to accommodate high surface heat flux. The blanket has a simple blanket configuration and low coolant pressure to reduce the fabrication cost, to improve the blanket reliability, and to increase confidence in the blanket performance. Spectral shifter, moderator, and reflector are utilized to improve the blanket shielding capability and energy multiplication, and to reduce the radial blanket thickness. Natural lithium is used to avoid extra cost related to the lithium enrichment process.
Date: July 1, 1999
Creator: Gohar, Y.; Majumdar, S. & Smith, D.
Partner: UNT Libraries Government Documents Department

Biaxial Creep-Fatigue Behavior of Type 316H Stainless Steel Tube

Description: Biaxial creep-fatigue test data for Type 316 stainless steel tubes at 1100*Y are presented. The specimens were subjected to constant internal pressure and fluctuating axial strain with and without hold times in tension as well as compress ion. The results show that internal pressure significantly affects diametral ratchetting and axial stress range. Axial tensile hold is found to he more damaging than axial compressive hold even cinder a biaxial state of stress.
Date: April 1979
Creator: Majumdar, S.
Partner: UNT Libraries Government Documents Department

ITER structural design criteria and their extension to advanced fusion reactor blankets.

Description: Application of the new low-temperature-design rules of the ITER Structural Design Criteria (ISDC) is illustrated by considering copper alloys that, according to recent data, are particularly prone to irradiation embrittlement at relatively low fluences at certain temperatures, Allowable stresses are derived and the impact of the embrittlement on allowable surface heat flux of a simple first-wall/limiter design is demonstrated. High-temperature-design rules of ISDC are applied to EVOLVE (Evaporation Of Lithium and Vapor Extraction), a blanket design concept currently being investigated under the U.S. APEX (Advanced Power Extraction) program. One version of this concept envisions the use of a series of parallel tungsten tubes (first-wall) that are cooled internally by lithium vapor, typically. at 1200 C. A single tungsten first-wall tube is considered for thermal and stress analyses by finite-element method.
Date: September 3, 1999
Creator: Kalnin, G. & Majumdar, S.
Partner: UNT Libraries Government Documents Department

Picotron 100 streak tubes as a 150-channel photometer

Description: The characterization of a streak camera based upon Picotron 100 tube types is given. Both a large (30 cm 1 x 10 cm dia.) and a small (18 cm 1 x 5 cm dia.) version of this design has been tested. Over 150 channels of information are simultaneously time resolved with system S.N.R. of 3 at 100 picosecond time resolution without post intensification. Absolute photometric evaluation is given in the dynamic mode, i.e. while operating in the picosecond time domain. Such quantitative data has been lacking in the past, particularly for multiple channel applications.
Date: January 1, 1982
Creator: Majumdar, S.; Weiss, P.B. & Black, J.P.
Partner: UNT Libraries Government Documents Department

First wall and limiter lifetime in pulsed tokamak reactors

Description: This study concentrates on the structural integrity of certain reactor subsystems under cyclic operation to answer the question: how long a burn pulse is needed to achieve the benefits of steady-state operation. Component lifetime in the steady-state is limited by three effects: radiation damage, disruptions, and sputtering erosion. Cyclic operation modifies one of these (the number of disruptions may increase with the number of burn cycles) and introduces a fourth life limit, thermal fatigue. Our design strategy is to determine the structure and coating thicknesses which maximize component lifetime against all life limitations. After calculating disruption damage (vaporization, melting) for candidate materials we present the lifetime analysis for different structures.
Date: December 1, 1983
Creator: Ehst, D.; Majumdar, S.; Cha, Y. & Hassanein, A.
Partner: UNT Libraries Government Documents Department

Assessment of thermal storage systems and thermomechanical effects for pulsed reactors

Description: Pulsed operation of fusion power plants has severe impact on all major reactor components. This analysis focuses on the sensitivity of one subsystem, the breeding blanket, to pulsed operation in terms of thermal storage requirements and thermomechanical effects. For analysis, a water-cooled Li/sub 2/O breeding blanket (400 MWth, 3.45 MW/m/sup 2/ neutron wall loading) was chosen. With the operating temperature window, 800/410/sup 0/C for Li/sub 2/O, thermal analysis shows that for the coolant-in-tube design (STARFIRE) there would be 10 rows of coolant tubes in the radial direction of the blanket. Since the thermal inertia of the blanket is larger further away from the first wall, the mixed mean temperature of coolant from all regions will dictate the design requirements for the thermal storage system. Three representative blanket regions were analyzed under four burn scenarios (startup/shutdown time = 10 s, steady-state time = 3600 s, and dwell time = 0, 30, 90, and 200 s) to estimate the thermal storage requirements. The size of the thermal storage system is dictated primarily by the energy deficiency that occurs during the dwell/startup and shutdown phase, although time/temperature response of the heat transfer fluid is critical to the design. Only pressurized water/steam and hot sodium thermal storage systems are considered for this study, since alternative systems are not attractive for heat storage of the order of 11 MWh to 230 MWh.
Date: December 1, 1983
Creator: Misra, B.; Stevens, H.; Majumdar, S. & Ehst, D.
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 4: High-Temperature Materials PIRTs

Description: The Phenomena Identification and Ranking Table (PIRT) technique was used to identify safety-relevant/safety-significant phenomena and assess the importance and related knowledge base of high-temperature structural materials issues for the Next Generation Nuclear Plant (NGNP), a very high temperature gas-cooled reactor (VHTR). The major aspects of materials degradation phenomena that may give rise to regulatory safety concern for the NGNP were evaluated for major structural components and the materials comprising them, including metallic and nonmetallic materials for control rods, other reactor internals, and primary circuit components; metallic alloys for very high-temperature service for heat exchangers and turbomachinery, metallic alloys for high-temperature service for the reactor pressure vessel (RPV), other pressure vessels and components in the primary and secondary circuits; and metallic alloys for secondary heat transfer circuits and the balance of plant. These materials phenomena were primarily evaluated with regard to their potential for contributing to fission product release at the site boundary under a variety of event scenarios covering normal operation, anticipated transients, and accidents. Of all the high-temperature metallic components, the one most likely to be heavily challenged in the NGNP will be the intermediate heat exchanger (IHX). Its thin, internal sections must be able to withstand the stresses associated with thermal loading and pressure drops between the primary and secondary loops under the environments and temperatures of interest. Several important materials-related phenomena related to the IHX were identified, including crack initiation and propagation; the lack of experience of primary boundary design methodology limitations for new IHX structures; and manufacturing phenomena for new designs. Specific issues were also identified for RPVs that will likely be too large for shop fabrication and transportation. Validated procedures for on-site welding, post-weld heat treatment (PWHT), and inspections will be required for the materials of construction. High-importance phenomena related to the RPV include crack ...
Date: March 1, 2008
Creator: Corwin, William R; Ballinger, R.; Majumdar, S. & Weaver, K. D.
Partner: UNT Libraries Government Documents Department

Thermal and structural limitations for impurity-control components in FED/INTOR

Description: The successful operation of the impurity-control system of the FED/INTOR will depend to a large extent on the ability of its various components to withstand the imposed thermal and mechanical loads. The present paper explores the thermal and stress analyses aspects of the limiter and divertor operation of the FED/INTOR in its reference configuration. Three basic limitations governing the design of the limiter and the divertor are the maximum allowable metal temperature, the maximum allowable stress intensity and the allowable fatigue life of the structural material. Other important design limitations stemming from sputtering, evaporation, melting during disruptions, etc. are not considered in the present paper. The materials considered in the present analysis are a copper and a vanadium alloy for the structural material and graphite, beryllium, beryllium oxide, tungsten and silicon carbide for the coating or tile material.
Date: February 1, 1983
Creator: Majumdar, S.; Cha, Y.; Mattas, R.; Abdou, M.; Cramer, B. & Haines, J.
Partner: UNT Libraries Government Documents Department

Fracture behavior of advanced ceramic hot-gas filters

Description: We have evaluated the microstructural/mechanical, and thermal shock/fatigue behavior and have conducted stress analyses of hot-gas candle filters made by various manufacturers. These filters include both monolithic and composite ceramics. Mechanical-property measurement of the composite filters included diametral compression testing with O-ring specimens and burst testing of short filter segments using rubber plug. In general, strength values obtained by burst testing were lower than those obtained by O-ring compression testing. During single-cycle thermal-shock tests, the composite filters showed little or no strength degradation when quenched from temperatures between 900 and 1000{degrees}C. At higher quenching temperatures, slow strength degradation was observed. The monolithic SiC filters showed no strength degradation when quenched from temperatures of up to {approx}700-900{degrees}C, but displayed decreased strength at a relatively sharp rate when quenched from higher temperatures. On the other hand, a recrystallized monolithic SiC filter showed higher initial strength and retained this strength to higher quenching temperatures than did regular SiC filters. This may be related to the difference in strength of grain boundary phases in the two cases. In thermal cycles between room temperature and 800- 1000{degrees}C, both monolithic and composite filters show a small strength degradation up to three cycles, beyond which the strength remained unchanged. Results of rubber-plug burst testing on composite filters were analyzed to determine the anisotropic elastic constants of the composite in the hoop direction. When these results are combined with axial elastic constants determined from axial tensile tests, the composite can be analyzed for stress due to mechanical (e. g., internal pressure) or thermal loading (thermal shock during pulse cleaning). The stresses can be compared with the strength of the composite to predict filter performance.
Date: May 1, 1996
Creator: Singh, J.P.; Majumdar, S.; Sutaria, M. & Bielke, W.
Partner: UNT Libraries Government Documents Department