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Thermal Properties of Structural Materials Found in Light Water Reactor Vessels

Description: High temperature material property data for structural materials used in existing Light Water Reactors (LWRs) are limited. Often, extrapolated values recommended in the literature differ significantly. To reduce such uncertainties, new data for SA533 Grade B, Class 1 (SA533B1) low alloy steel, Stainless Steel 304 (SS304), and Inconel 600, found in Light Water Reactor (LWR) vessels and penetrations, were acquired and tested using material property systems available at the High Temperature Test Laboratory (HTTL) at the Idaho National Laboratory (INL). Properties measured include thermal expansion, specific heat capacity, and thermal diffusivity for temperatures up to 1200 oC. From these results, thermal conductivity and density were calculated. Results show that, in some cases, previously recommended values for these material differ significantly from measured values at high temperatures. This is especially true for SA533B1, as previous data do not account for the phase transformation of this material between 740 oC and 840 oC.
Date: November 1, 2009
Creator: Daw, J. E.; Rempe, J. L. & Knudson, D. L.
Partner: UNT Libraries Government Documents Department

105 K-West isolation barrier leak recovery plan

Description: Leak testing is being performed in 105 KW to verify the performance of the isolation barriers which have been recently installed. When an 11 inch differential head is established between the main basin and the discharge chute, a leak-rate of approximately 30 - 35 gpm is observed. The leak-rate would be achieved by a 1.65`` - 2`` diameter hole (or equivalent). Analyses suggest that the flow is turbulent/laminar transitional (dominantly turbulent), which would be indicative of a single point leak, typical of a pipe or large opening. However, local vortex rotation is observed in the entry to the West transfer chute while no observable motion was seen in the East transfer chute: this may be an indication of seal leakage in the East isolation barrier. The potential for leakage had been considered during the design and field work planning stages. Review of potential leak detection technologies had been made; at the planning stage it was determined that location specific leak detection could be established relatively quickly, applying existing K Basins technology (dye or ultrasonics). The decision was made not to pre-stage leak detection since the equipment development is highly dependent on the nature and location of the leak, and the characteristics of the leak rate provides data which guides leak characterization technology. The expense could be deferred and potentially avoided without risk to critical path activity. Consistent with the above, a systematic recovery plan has been developed utilizing phased activities to provide for management discipline combined with timely diagnosis and correction. Because this activity is not critical path at this time, activities will be coordinated with other plant activity to optimize overall plant work. Particular care will be exercised in assuring that information gained from this recovery can be utilized in the more critical work in 105 KE.
Date: March 2, 1995
Creator: Wiborg, J.C.
Partner: UNT Libraries Government Documents Department

Biaxial loading effects on fracture toughness of reactor pressure vessel steel

Description: The preliminary phases of a program to develop and evaluate fracture methodologies for assessing crack-tip constraint effects on fracture toughness of reactor pressure vessel (RPV) steels have been completed by the Heavy-Section Steel Technology (HSST) Program. Objectives were to investigate effect of biaxial loading on fracture toughness, quantify this effect through existing stress-based, dual-parameter, fracture-toughness correlations, or propose and verify alternate correlations. A cruciform beam specimen with 2-D, shallow, through-thickness flaw and a special loading fixture was designed and fabricated. Tests were performed using biaxial loading ratios of 0:1 (uniaxial), 0.6:1, and 1:1 (equi-biaxial). Critical fracture-toughness values were calculated for each test. Biaxial loading of 0.6:1 resulted in a reduction in the lower bound fracture toughness of {approximately}12% as compared to that from the uniaxial tests. The biaxial loading of 1:1 yielded two subsets of toughness values; one agreed well with the uniaxial data, while one was reduced by {approximately}43% when compared to the uniaxial data. Results were evaluated using J-Q theory and Dodds-Anderson (D-A) micromechanical scaling model. The D-A model predicted no biaxial effect, while the J-Q method gave inconclusive results. When applied to the 1:1 biaxial data, these constraint methodologies failed to predict the observed reduction in fracture toughness obtained in one experiment. A strain-based constraint methodology that considers the relationship between applied biaxial load, the plastic zone width in the crack plane, and fracture toughness was formulated and applied successfully to the data. Evaluation of this dual-parameter strain-based model led to the conclusion that it has the capability of representing fracture behavior of RPV steels in the transition region, including the effects of out-of-plane loading on fracture toughness. This report is designated as HSST Report No. 150.
Date: March 1, 1995
Creator: McAfee, W.J.; Bass, B.R.; Bryson, J.W. Jr. & Pennell, W.E.
Partner: UNT Libraries Government Documents Department

Fracture assessment of weld material from a full-thickness clad RPV shell segment

Description: Fracture analysis was applied to full-thickness clad beam specimens containing shallow cracks in material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPV) at beginning of life. The beam specimens were fabricated from a section of an RPV wall (removed from a canceled nuclear plant) that includes weld, plate, and clad material. Metallurgical factors potentially influencing fracture toughness for shallow cracks in the beam specimens include gradients of material properties and residual stresses due to welding and cladding applications. Fracture toughness estimates were obtained from load vs load-line displacement and load vs crack-mouth-opening displacement data using finite-element methods and estimation schemes based on the {eta}-factor method. One of the beams experienced a significant amount of precleavage stable ductile tearing. Effects of precleavage tearing on estimates of fracture toughness were investigated using continuum damage models. Fracture toughness results from the clad beam specimens were compared with other deep- and shallow-crack single-edge notch bend (SENB) data generated previously from A533 Grade B plate material. Range of scatter for the clad beam data is consistent with that from the laboratory-scale SENB specimens tested at the same temperature.
Date: July 1, 1996
Creator: Keeney, J. A.; Bass, B. R. & McAfee, W. J.
Partner: UNT Libraries Government Documents Department

Fracture behavior of shallow cracks in full-thickness clad beams from an RPV wall section

Description: A testing program is described that utilizes full-thickness clad beam specimens to quantify fracture toughness for shallow cracks in weld material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPVs). The beam specimens are fabricated from an RPV shell segment that includes weld, plate and clad material. Metallurgical factors potentially influencing fracture toughness for shallow cracks in the beam specimens include material gradients and material inhomogeneities in welded regions. The shallow-crack clad beam specimens showed a significant loss of constraint similar to that of other shallow-crack single-edge notch bend (SENB) specimens. The stress-based Dodds-Anderson scaling model appears to be effective in adjusting the test data to account for in-plane loss of constraint for uniaxially tested beams, but cannot predict the observed effects of out-of-plane biaxial loading on shallow-crack fracture toughness. A strain-based dual-parameter fracture toughness correlation (based on plastic zone width) performed acceptably when applied to the uniaxial and biaxial shallow-crack fracture toughness data.
Date: April 1, 1995
Creator: Keeney, J. A.; Bass, B. R. & McAfee, W. J.
Partner: UNT Libraries Government Documents Department

An interim report on shallow-flaw fracture technology development

Description: Shallow-flaw fracture technology is being developed for application to the safety assessment of radiation-embrittled nuclear reactor pressure vessels (RPVS) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT{sub NDT}) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) a strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness.
Date: June 1, 1995
Creator: Pennell, W.E.; Bass, B.R.; Bryson, J.W. & McAfee, W.J.
Partner: UNT Libraries Government Documents Department

Technical Issues Associated with Air Ingression During Core Degradation

Description: This paper has shown that it is possible to get significant air intrusion into a ruptured reactor vessel even from a reactor cavity with restricted access. This suggests that there is some importance to considering the consequences of air intrusion following vessel penetration by core debris. The consequences will depend on the nature of core degradation in air and other oxidizing gases. If, indeed, fuel becomes exposed to strongly oxidizing gases, significant releases of ruthenium and hexavalent urania can be expected. Hexavalent urania could alter the nature of cesium release and cesium revaporization from the reactor coolant system. Hexavalent urania could destabilize CSI and enhance the formation of gaseous iodine unless there are other materials that will react readily with atomic iodine along the flow path to the reactor containment.
Date: June 5, 2000
Partner: UNT Libraries Government Documents Department

Dissolution of Stainless Steel by Molten Aluminum and Aluminum Alloys - Final Report

Description: The purpose of this task was to investigate on a laboratory-scale the interactions of molten aluminum with stainless steel under hypothetical severe reactor accident conditions. This experimental effort provided data necessary to assess the susceptibility of the reactor vessel to breaching (general through-wall failure of vessel) in accident scenarios where contact of molten aluminum and stainless steel may occur. This report summarizes the results of the extensive experimental program.
Date: July 11, 2001
Creator: Marra, J.C.
Partner: UNT Libraries Government Documents Department

A Laser Metrology/Viewing System for ITER In-Vessel Inspection

Description: This paper identifies the requirements for a remotely operated precision laser ranging system for the International Thermonuclear Experimental Reactor. The inspection system is used for metrology and viewing, and must be capable of achieving submillimeter accuracy and operation in a reactor vessel that has high gamma radiation, high vacuum, elevated temperature, and magnetic field levels. A coherent, frequency modulated laser radar system is under development to meet these requirements. The metrology/viewing sensor consists of a compact laser-optic module linked through fiberoptics to the laser source and imaging units, located outside the harsh environment. The deployment mechanism is a remotely operated telescopic mast. Gamma irradiation up to 10{sup 7} Gy was conducted on critical sensor components with no significant impact to data transmission, and analysis indicates that critical sensor components can operate in a magnetic field with certain design modifications. Plans for testing key components in a magnetic field are underway.
Date: December 31, 1997
Creator: Spampinato, P.T.; Barry, R.E.; Chesser, J.B.; Menon, M.M.; Dagher, M.A. & Slotwinski, A.
Partner: UNT Libraries Government Documents Department

A Review of Large-Scale Fracture Experiments Relevant to Pressure Vessel Integrity Under Pressurized Thermal Shock Conditions

Description: Numerous large-scale fracture experiments have been performed over the past thirty years to advance fracture mechanics methodologies applicable to thick-wall pressure vessels. This report first identifies major factors important to nuclear reactor pressure vessel (RPV) integrity under pressurized thermal shock (PTS) conditions. It then covers 20 key experiments that have contributed to identifying fracture behavior of RPVs and to validating applicable assessment methodologies. The experiments are categorized according to four types of specimens: (1) cylindrical specimens, (2) pressurized vessels, (3) large plate specimens, and (4) thick beam specimens. These experiments were performed in laboratories in six different countries. This report serves as a summary of those experiments, and provides a guide to references for detailed information.
Date: January 29, 2001
Creator: Pugh, C. E.
Partner: UNT Libraries Government Documents Department

Effects of alloy chemistry, cold work, and water chemistry on corrosion fatigue and stress corrosion cracking of nickel alloys and welds.

Description: Reactor vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking (EAC). A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. The objective of this work is to evaluate and compare the resistance of Alloys 600 and 690 and their welds, such as Alloys 82, 182, 52, and 152, to EAC in simulated light water reactor environments. The existing crack growth rate (CGR) data for these alloys under cyclic and constant loads have been evaluated to establish the effects of alloy chemistry, cold work, and water chemistry. The experimental fatigue CGRs are compared with CGRs that would be expected in air under the same mechanical loading conditions to obtain a qualitative understanding of the degree and range of conditions for significant environmental enhancement in growth rates. The existing stress corrosion cracking (SCC) data on Alloys 600 and 690 and Alloy 82, 182, and 52 welds have been compiled and analyzed to determine the influence of key parameters on growth rates in simulated PWR and BWR environments. The SCC data for these alloys have been evaluated with correlations developed by Scott and by Ford and Andresen.
Date: April 1, 2001
Creator: Chopra, O. K.; Soppet, W. K.; Shack, W. J. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Fabrication Flaw Density and Distribution In Repairs to Reactor Pressure Vessel and Piping Welds

Description: The Pacific Northwest National Laboratory is developing a generalized fabrication flaw distribution for the population of nuclear reactor pressure vessels and for piping welds in U.S. operating reactors. The purpose of the generalized flaw distribution is to predict component-specific flaw densities. The estimates of fabrication flaws are intended for use in fracture mechanics structural integrity assessments. Structural integrity assessments, such as estimating the frequency of loss-of-coolant accidents, are performed by computer codes that require, as input, accurate estimates of flaw densities. Welds from four different reactor pressure vessels and a collection of archived pipes have been studied to develop empirical estimates of fabrication flaw densities. This report describes the fabrication flaw distribution and characterization in the repair weld metal of vessels and piping. This work indicates that large flaws occur in these repairs. These results show that repair flaws are complex in composition and sometimes include cracks on the ends of the repair cavities. Parametric analysis using an exponential fit is performed on the data. The relevance of construction records is established for describing fabrication processes and product forms. An analysis of these records shows there was a significant change in repair frequency over the years when these components were fabricated. A description of repair flaw morphology is provided with a discussion of fracture mechanics significance. Fabrication flaws in repairs are characterized using optimized-access, high-sensitivity nondestructive ultrasonic testing. Flaw characterizations are then validated by other nondestructive evaluation techniques and complemented by destructive testing.
Date: April 1, 2008
Creator: Schuster, G. J.; Simonen, F. A. & Doctor, S. R.
Partner: UNT Libraries Government Documents Department

Analytical modeling of the effect of crack depth, specimen size, and biaxial stress on the fracture toughness of reactor vessel steels

Description: Fracture, toughness values for A533-B reactor pressure vessel (RPV) steel obtained from test programs at Oak Ridge National Laboratory (ORNL) and University of Kansas (KU) are interpreted using the J-A{sub 2} analytical model. The analytical model is based on the critical stress concept and takes into consideration the constraint effect using the second parameter A{sub 2} in addition to the generally accepted first parameter J which represents the loading level. It is demonstrated that with the constraint level included in the model effects of crack depth (shallow vs deep), specimen size (small vs. large), and loading type (uniaxial vs biaxial) on the fracture toughness from the test programs can be interpreted and predicted.
Date: February 1, 1995
Creator: Chao, Yuh-Jin & Lam, Poh-Sang
Partner: UNT Libraries Government Documents Department

Impact induced response spectrum for the safety evaluation of the high flux isotope reactor

Description: The dynamic impact to the nearby HFIR reactor vessel caused by heavy load drop is analyzed. The impact calculation is carried out by applying the ABAQUS computer code. An impact-induced response spectrum is constructed in order to evaluate whether the HFIR vessel and the shutdown mechanism may be disabled. For the frequency range less than 10 Hz, the maximum spectral velocity of impact is approximately equal to that of the HFIR seismic design-basis spectrum. For the frequency range greater than 10 Hz, the impact-induced response spectrum is shown to cause no effect to the control rod and the shutdown mechanism. An earlier seismic safety assessment for the HFIR control and shutdown mechanism was made by EQE. Based on EQE modal solution that is combined with the impact-induced spectrum, it is concluded that the impact will not cause any damage to the shutdown mechanism, even while the reactor is in operation. The present method suggests a general approach for evaluating the impact induced damage to the reactor by applying the existing finite element modal solution that has been carried out for the seismic evaluation of the reactor.
Date: May 1, 1997
Creator: Chang, S.J.
Partner: UNT Libraries Government Documents Department

Vessel support subsystem design description. Revision 1

Description: The Vessel Support Subsystem is one of three subsystems comprising the Vessel System of the Modular High Temperature Gas-Cooled Reactor 4 x 350 MW(t) Plant. The design of this subsystem has been developed by means of the Integrated Approach. This document establishes the functions and system design requirements of the Vessel Support Subsystem from the Functional Analysis, and includes institutional requirements from the Overall Plant Design Specification and the Vessel System Design Description. A description of the subsystem design which satisfies these requirements is presented. Lower-tier requirements at the subsystem level are next defined for the component design. This document also includes information on aspects of subsystem construction, operation, maintenance, and decommissioning.
Date: July 1, 1987
Creator: Perry, R.A. & Mehta, D.D.
Partner: UNT Libraries Government Documents Department

CSNI Project for Fracture Analyses of Large-Scale International Reference Experiments (FALSIRE II)

Description: A summary of Phase II of the Project for FALSIRE is presented. FALSIRE was created by the Fracture Assessment Group (FAG) of the OECD/NEA`s Committee on the Safety of Nuclear Installations (CNSI) Principal Working Group No. 3. FALSIRE I in 1988 assessed fracture methods through interpretive analyses of 6 large-scale fracture experiments in reactor pressure vessel (RPV) steels under pressurized- thermal-shock (PTS) loading. In FALSIRE II, experiments examined cleavage fracture in RPV steels for a wide range of materials, crack geometries, and constraint and loading conditions. The cracks were relatively shallow, in the transition temperature region. Included were cracks showing either unstable extension or two stages of extensions under transient thermal and mechanical loads. Crack initiation was also investigated in connection with clad surfaces and with biaxial load. Within FALSIRE II, comparative assessments were performed for 7 reference fracture experiments based on 45 analyses received from 22 organizations representing 12 countries. Temperature distributions in thermal shock loaded samples were approximated with high accuracy and small scatter bands. Structural response was predicted reasonably well; discrepancies could usually be traced to the assumed material models and approximated material properties. Almost all participants elected to use the finite element method.
Date: November 1, 1996
Creator: Bass, B. R.; Pugh, C. E.; Keeney, J.; Schulz, H. & Sievers, J.
Partner: UNT Libraries Government Documents Department

Report of material and equipment section`s activities at New York Shipbuilding Corporation during fabrication of AXC 167 1/2 starting May 18, 1951. Part 7, Section 3

Description: This document provides Part VII, Section III and Section IV of the report of the Material and Equipment Section`s activities at the New York Shipbuilding Corporation. The fabrication, inspection, and testing of reactor components is detailed.
Date: April 28, 1954
Creator: Stewart, J.R.
Partner: UNT Libraries Government Documents Department

Report of material and equipment section`s activities at New York Shipbuilding Corporation during fabrication of AXC 167 1/2 starting May 18, 1951. Part 3

Description: This report provides Part III through VI of the Material and Equipment Section`s activities at New York Shipbuilding Corporation. Fabrication, inspection, and testing of reactor components are detailed.
Date: May 26, 1954
Creator: Stewart, J.R.
Partner: UNT Libraries Government Documents Department


Description: Under the auspices of the ''Memorandum of Understanding for the Exchange of Technical Information and for Cooperation in the Field of Peaceful Uses of Nuclear Energy'' between the National Institute of Nuclear Research of Mexico (ININ) and the Los Alamos National Laboratory (LANL), scientists and engineers from ININ met and collaborated with scientists at LANL. The collaboration was sponsored by the US Department of Energy as part of its ''Sister Laboratories'' program. In this weeklong meeting, these scientists and engineers carried out mutual consultation and cooperative efforts in the field of computational research in nuclear power. Three main areas for technical collaboration were discussed: (a) establishment of electronic access to LANL open computational facilities and reactor physics codes from ININ, (b) calculation of radiation damage to BWR reactor vessels, and (c) calculation of BWR burnup for MOX fuel. These three tasks were successfully completed during the weeklong meeting between the laboratory scientists. The discussion, held at LANL in March 1999, involved ten LANL specialists and three ININ specialists. In addition, several computer technicians provided the necessary support for the utilization of the SUN computers, which were setup for the seminars. Discussions between team members occupied about half of the visit. Mixed Oxide (MOX) assembly models were developed and calculations made using HELIOS and MCNP the remainder of the time. As a result of the collaboration, the scientists from ININ returned to the institute and immediately began using the computational facilities at LANL for further MOX assembly calculations. As a result of the meeting, ININ is providing expert advice for the thermal hydraulic calculations for a similar cooperative program between Peruvian and LANL. The three areas of cooperation will be discussed in detail in this paper. Sample results of the MOX calculations at ININ will also be presented.
Date: July 1, 1999
Creator: PERRY, R.; CHARLTON, W. & AL, ET
Partner: UNT Libraries Government Documents Department

Fracture toughness evaluation of a low upper-shelf weld metal from the Midland Reactor using the master curve

Description: The primary objective of the Heavy-Section Steel Irradiation (HSSI) Program Tenth Irradiation Series was to develop a fracture mechanics evaluation of weld metal WF-70, which was taken from the beltline and nozzle course girth weld joints of the Midland Reactor vessel. This material became available when Consumers Power Company of Midland, Michigan, decided to abort plans to operate their nuclear power plant. WF-70 is classified as a low upper-shelf steel primarily due to the Linde 80 flux that was used in the submerged-arc welding process. The master curve concept is introduced to model the transition range fracture toughness when the toughness is quantified in terms of K{sub Jc} values. K{sub Jc} is an elastic-plastic stress intensity factor calculated by conversion from J{sub c}; i.e., J-integral at onset of cleavage instability.
Date: March 1997
Creator: McCabe, D. E.; Sokolov, M. A. & Nanstad, R. K.
Partner: UNT Libraries Government Documents Department

Experiments on Corium Dispersion after Lower Head Failure at Moderate Pressure

Description: Concerning the mitigation of high pressure core melt scenarios, the design objective for future PWRS is to transfer high pressure core melt to low pressure core melt sequences, by means of pressure relief valves at the primary circuit, with such a discharge capacity to limit the pressure in the reactor coolant system to less than 20 bar. Studies have shown that in late in-vessel reflooding scenarios there may be a time window where the pressure is indeed in this range, at the moment of the reactor vessel rupture. It has to be verified that large quantities of corium released from the vessel after failure at pressures <20 bar cannot be carried out of the reactor pit, because the melt collecting and cooling concept of future PWRs would be rendered useless. Existing experiments investigated the melt dispersal phenomena in the context of the DCH resolution for existing power plants in the USA, most of them having cavities with large instrument tunnels leading into subcompartments. For such designs, breaches with small cross sections at high vessel failure pressures had been studied. However, some present and future European PWRs have an annular cavity design without a large pathway out of the cavity other than through the narrow annular gap between the RPV and the cavity wall. Therefore, an experimental program was launched, focusing on the annular cavity design and low pressure vessel failure. The first part of the program comprises two experiments which were performed with thermite melt steam and a prototypic atmosphere in the containment in a scale 1:10. The initial pressure in the RPV-model was 11 and 15 bars, and the breach was a hole at the center of the lower head with a scaled diameter of 100 cm and 40 cm, respectively. The main results were: 78% of melt mass ...
Date: September 21, 1999
Partner: UNT Libraries Government Documents Department

Evaluating the safety of aging nuclear reactor pressure vessels

Description: Regulatory requirements limit the permissible accumulation of irradiation damage in RPV material such that adequate fracture prevention margins are maintained throughout the licensed operating period of a nuclear plant. Experience with application of those requirements has identified a number of areas where they could be further refined to eliminate excess conservatism. Research is ongoin to provide the data required to support refinement of the regulatory requirements. Research programs are investigating theeffects of local brittle zones, shallow flaws, biaxial loading, and stainless steel cladding. Preliminary results from this research indicate a potential for beneficial changes in the P-T curve and PTS analysis rules.
Date: May 1, 1996
Creator: Pennell, W.E.
Partner: UNT Libraries Government Documents Department