U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Engineering
Resource Information
The organization U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Engineering represents an institution, an association, or corporate body that is associated with resources found in Indiana State Library.
The Resource
U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Engineering
Resource Information
The organization U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Engineering represents an institution, an association, or corporate body that is associated with resources found in Indiana State Library.
- Label
- U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Engineering
- Authority link
- (EG-IN)1141770
- Subordinate unit
-
- Office of Nuclear Regulatory Research
- Division of Engineering
432 Items by the Organization U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Engineering
Context
Context of U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of EngineeringContributor of
No resources found
No enriched resources found
- A characterization of check valve degradation and failure experience in the nuclear power industry
- A damage mechanics based approach to structural deterioration and reliability
- A performance assessment methodology for high-level radioactive waste disposal in unsaturated, fractured tuff
- A pilot application of risk-based methods to establish in-service inspection priorities for nuclear components at Surry Unit 1 Nuclear Power Station
- A review of research conducted by Los Alamos National Laboratory for the NRC with emphasis on the Maxey Flats, KY, shallow land burial site
- A study of the use of crosslinked high-density polyethylene for low-level radioactive waste containers
- A summary of nuclear power plant fire safety research at Sandia National Laboratories, 1975-1987
- A system for generating long streamflow records for study of floods of long return period
- A user's guide to the NRC's piping fracture mechanics data base (PIFRAC)
- Accelerated irradiation test of Gundremmingen reactor vessel trepan material
- Acoustic emission system calibration at Watts Bar Unit 1 nuclear reactor
- Acoustic emission/flaw relationships for inservice monitoring of LWRs
- Age-related degradation of Westinghouse 480-volt circuit breakers
- Aging and qualification research on solenoid operated valves
- Aging and service wear of check valves used in engineered safety-feature systems of nuclear power plants
- Aging and service wear of control rod drive mechanisms for BWR nuclear plants
- Aging and service wear of electric motor-operated valves used in engineering safety-feature systems of nuclear power plants
- Aging assessment of BWR standby liquid control systems
- Aging assessment of bistables and switches in nuclear power plants
- Aging assessment of component cooling water systems in pressurized water reactors
- Aging assessment of essential HVAC chillers used in nuclear power plants
- Aging assessment of instrument air systems in nuclear power plants
- Aging assessment of nuclear air-treatment system HEPA filters and adsorbers
- Aging assessment of reactor instrumentation and protection system components : aging-related operating experiences
- Aging assessment of the Combustion Engineering and Babcock & Wilcox control rod drives
- Aging assessment of the Westinghouse PWR control rod drive system
- Aging data analysis and risk assessment : development and demonstration study
- Aging evaluation of class 1E batteries : seismic testing
- Aging mitigation and improved programs for nuclear service diesel generators
- Aging of cables, connections, and electrical penetration assemblies used in nuclear power plants
- Aging of control and service air compressors and dryers used in nuclear power plants
- Aging of nuclear plant resistance temperature detectors
- Aging of nuclear station diesel generators : evaluation of operating and expert experience
- Aging study of boiling water reactor residual heat removal system
- Aging, condition monitoring, and loss-of-coolant accident (LOCA) tests of class 1E electrical cables
- Alternate modal combination methods in response spectrum analysis
- Alternative modal combination methods in response spectrum analysis
- An analytical study of seismic threat to containmnent integrity
- An approach to the mathematical modelling of the uranium series redistribution within ore bodies
- An approach to the quantification of seismic margins in nuclear power plants : the importance of BWR plant systems and functions to seismic margins
- An experimental investigation of internally ignited fires in nuclear power plant control cabinets
- An exploratory study of element interactions and composition dependencies in radiation sensitivity development : final report
- An investigation of crack-tip stress field criteria for predicting cleavage-crack initiation
- An investigation of the effects of thermal aging on the fire damageability of electric cables
- An overview of the low-upper-shelf toughness safety margin issue
- Analyses and field tests of the hydraulic performance of cement grout borehole seals
- Analysis of H.B. Robinson PWR vessel fluence for cycle 10 utilizing partial length shield assemblies
- Analysis of crack initiation and growth in the high level vibration test at Tadotsu
- Analysis of diffusion flame tests
- Analysis of shell-rupture failure due to hypothetical elevated-temperature pressurization of the Sequoyah Unit 1 steel containment building
- Analysis of the VENUS-3 experiments
- Analytical studies of transverse strain effects on fracture toughness for circumferentially oriented cracks
- Annual report of the TMI-2 EPICOR-II resin/liner investigation : Low-level Waste Data Base Development Program for fiscal year 1988
- Application of acoustic leak detection technology for the detection and location of leaks in light water reactors
- Application of reliability techniques to prioritize BWR recirculation loop welds for in-service inspection
- Application of stochastic methods to the simulation of large-scale unsaturated flow and transport
- Application of the J-integral and the modified J-integral to cases of large crack extension
- Applications of a new magnetic monitoring technique to in situ evaluation of fatigue damage in ferrous components
- Approaches for age-dependent probabilistic safety assessments with emphasis on prioritization and sensitivity studies
- Approaches to large scale unsaturated flow in heterogeneous, stratified, and fractured geologic media
- Approximate techniques for predicting size effects on cleavage fracture toughness (Jc)
- Assessment of leak detection systems for LWRs
- Assessment of seismic margin calculation methods
- Assessment of the impact of degraded shear wall stiffnesses on seismic plant risk and seismic design loads
- Assessment of uncertainties in measurement of pH in hostile environments characteristic of nuclear repositories
- Auxiliary feedwater system aging study
- Auxiliary feedwater system aging study
- BWR reactor water cleanup system flexible wedge gate isolation value qualification and high energy flow interruption test
- Basis for snubber aging research : nuclear plant aging research program
- Bentonite borehole plug flow testing with five water types
- Beta and gamma dose calculations for PWR and BWR containments
- Biaxial loading and shallow-flaw effects on crack-tip constraint and fracture toughness
- Biodegradation testing of TMI-2 EPICOR-II waste forms
- Boiling-water reactor internals aging degradation study : phase 1
- Bond strength of cement borehole plugs in salt
- Bond strength of cementitious borehole plugs in welded tuff
- Borehole closure in salt
- CARES (Computer Analysis for Rapid Evaluation of Structures) version 1.0 : seismic module
- CSNI Project for Fracture Analyses of Large-Scale International Reference Experiments (Project FALSIRE)
- Canadian seismic agreement
- Cavitation guide for control valves
- Characteristics of low-level radioactive waste
- Chemical decontamination and chemical cleaning of LWR components and possible interactions with metallurgical aging effects
- Cloride ion diffusion in low water-to-solid cement pastes
- Cobalt-60 simulation of LOCA radiation effects
- Collaborative study of NUPEC seismic field test data for NPP structures
- Comparative evaluation of selected continuum and discrete-fracture models : emphasis on dispersivity calculations for application to fractured geologic media, creston study area, eastern Washington
- Comparison of strongly heat-driven flow codes for unsaturated media
- Compendium and comparison of international practice for plugging, repair, and inspection of steam generator tubing
- Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors
- Component fragility research program : phase I component prioritization
- Component fragility research program : phase II development of seismic fragilities from high-level qualification data
- Comprehensive aging assessment of circuit breakers and relays
- Computer programs for eddy-current defect studies
- Conceptualization of a hypothetical high-level nuclear waste repository site in unsaturated, fractured tuff
- Constraint effects on fracture toughness for circumferentially oriented cracks in reactor pressure vessels
- Containment integrity research program plan
- Continuous AE crack monitoring of a dissimilar metal weldment at Limerick Unit 1
- Continuous cooling thermal cycle effects on sensitization in stainless steel
- Continuum and micromechanics treatment of constraint in fracture
- Control of water infiltration into near surface LLW disposal units
- Control of water infiltration into near surface LLW disposal units
- Correlation of irradiation-induced transition temperature increases from Cv and Kjc/Kic data : final report
- Correlations between power and test reactor data bases
- Crack-arrest behavior in SEN wide plates of low-upper-shelf base metal tested under nonisothermal conditions : WP-2 series
- Crack-arrest tests on two irradiated high-copper welds : phase II : results of duplex-type experiments
- Crack-speed relations inferred from large single-edge notched specimens of A 533 B steel
- Critical assessment of seismic and geomechanics literature related to a high-level nuclear waste underground repository
- Critical parameters for a high-level waste repository
- Crushed salt consolidation
- Current applications of vibration monitoring and neutron noise analysis : detection and analysis strutural degradation of reactor vessel internals from operational aging
- DCM3D, a dual-continuum, three-dimensional, ground-water flow code for unsaturated, fractured, porous media
- Damping in building structures during earthquakes : test data and modeling
- Damping in low-aspect-ratio, reinforced concrete shear walls
- Dating ground water and the evaluation of repositories for radioactive waste
- Degradation of nuclear plant temperature sensors
- Demonstration of a performance assessment methodology for high-level radioactive waste disposal in basalt formations
- Design provisions for tangential shear in containment walls
- Detection and characterization of indications in segments of reactor pressure vessels
- Determination of the neutron and gamma flux distribution in the pressure vessel and cavity of a boiling water reactor
- Development of a mechanistic understanding of radiation embrittlement in reactor pressure vessel steels : final report
- Development of a testing and analysis methodology to determine the functional condition of solenoid operated valves
- Development of an engineering definition of the extent of J singularity controlled crack growth
- Development of an infiltration evaluation methodology for low-level waste shallow land burial sites
- Development of equipment parameter tolerances for the ultrasonic inspection of steel components
- Ductile to brittle toughness transition characterization of A533B steel
- Dynamic testing of a circular foundation and analyses of soil/structure interaction
- EPICOR-II resin/liner investigation : low-level waste data base development program for fiscal year 1990
- Effect of aging on response time of nuclear plant pressure sensors
- Effect of dynamic strain aging on the strength and toughness of nuclear ferritic piping at LWR temperatures
- Effect of pH on the release of radionuclides and chelating agents from cement-solidified decontamination ion-exchange resins collected from operating nuclear power stations
- Effect of prior deformation on sensitization development in stainless steel during continuous cooling
- Effect of welding conditions on transformation and properties of heat-affected zones in LWR vessel steels
- Effectiveness of fracture sealing with bentonite grouting
- Effects of manufacturing variables on performance of high-level waste low carbon steel containers : final report
- Effects of mineralogy on sorption of strontium and cesium onto Calico Hills tuff
- Effects of nonstandard heat treatment temperatures on tensile and charpy impact properties of carbon-steel casting repair welds
- Effects of tensile loading on upper shelf fracture toughness
- Elastic-plastic characterization of a cast stainless steel pipe elbow material
- Electrochemical evaluation of solid state pH sensors for nuclear waste containment
- Environmental effects on corrosion in the tuff repository
- Environmentally assisted cracking in light water reactors
- Environmentally assisted cracking in light water reactors
- Environmentally assisted cracking in light water reactors : semiannual report
- Estimation of fracture toughness of cast stainless steels during thermal aging in LWR systems
- Estimation of fracture toughness of cast stainless steels during thermal aging in LWR systems
- Evaluation and refinement of leak-rate estimation models
- Evaluation and refinement of leak-rate estimation models
- Evaluation of MHTGR fuel reliability
- Evaluation of behavior and the radial shear strength of a reinforced concrete containment structure
- Evaluation of crack pop-ins and the determination of their relevance to design considerations
- Evaluation of hypotheses for the cause of the 1886 Charleston earthquake : final report
- Evaluation of materials of construction for the reinforced concrete reactor containment model
- Evaluation of nuclear facility decommissioning projects : summary status report, Three Mile Island Unit 2
- Evaluation of rock joint models and computer code UDEC against experimental results
- Evaluation of safety implications of control systems in LWR nuclear power plants : technical findings related to unresolved safety issue A-47
- Evaluation of sampling plans for in-service inspection of steam generator tubes
- Evaluation of sampling plans for in-service inspection of steam generator tubes
- Evaluation of seals for mechanical penetrations of containment buildings
- Evaluation of the leakage behavior of inflatable seals subject to severe accident conditions
- Evaluation of uncertainties in irradiated hardware characterization
- Evaluations of core melt frequency effects due to component aging and maintenance
- Experiment investigation of sedimentation of LOCA-generated fibrous debris and sludge in BWR suppression pools
- Experimental and analytical investigation of the shallow-flaw effect in reactor pressure vessels
- Experimental assessment of damping in low aspect ratio, reinforced concrete shear wall structure
- Experimental assessment of the influence of dynamic loading on the permeability of wet and dried cement borehole seals
- Experimental assessments of Gundremmingen RPV archive material for fluence rate effects studies
- Experimental results from containment piping bellows subjected to severe accident conditions
- Experimental results from pressure testing a 1:6-scale nuclear power plant containment
- Experimental results of tests to investigate flaw behavior of mechanically loaded stainless steel clad plates
- Extension and extrapolation of J-R curves and their application to the low upper shelf toughness issue
- Extrapolation of the J-R curve for predicting reactor vessel integrity
- Fatigue crack growth of part-through cracks in pressure vessel and piping steels : air environment results
- Fatigue life characterization of smooth and notched piping steel specimens in 288CÌŠ air environments
- Fatigue strength of smooth and notched specimens of ASME SA 106-B steel in PWR environments
- Feasibility of developing risk-based rankings of pressure boundary systems for inservice inspection
- Flow and transport at the Las Cruces trench site : experiments 1 and 2
- Fluid flow and solute transport modeling through three-dimensional networks of variably saturated discrete fractures
- Fracture evaluation of surface cracks embedded in reactor vessel cladding
- Fracture toughness characterization of nuclear piping steels
- Fracture-mechanics-based failure analysis
- Functional capability of piping systems
- Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors
- Further Neotectonic studies of earthquake zones in the eastern United States
- Generic analyses for evaluation of low Charpy upper-shelf energy effects on safety margins against fracture of reactor pressure vessels
- Generic issue 87 : flexible wedge gate valve test program : phase II results and analysis
- Geophysical investigations of the western Ohio-Indiana region
- Geophysical investigations of the western Ohio-Indiana region
- Georgia/Alabama Regional Seismographic Network
- Georgia/Alabama regional seismographic network
- Global positioning system measurements over a strain monitoring network in the eastern two-thirds of the United States
- Gradient study of a large weld joining two forged A 508 shells of the Midland reactor vessel
- Guide for preparing operating procedures for shipping packages
- Heavy-section steel irradiation program
- Heavy-section steel irradiation program
- Heavy-section steel technology program
- High-level seismic response and failure prediction methods for piping
- High-temperature crack-arrest behavior in 152-mm-thick SEN wide plates of quenched and tempered A 533 grade B class 1 steel
- High-temperature crack-arrest tests using 152-mm-thick SEN wide plates of low-upper-shelf base material : tests WP-2.2 and WP-2.6
- Hybrid digital signal processing and neural networks for automated diagnostics using NDE methods
- Hydrogeologic characterization of basalts : the northern rim of the Columbia Plateau physiographic province and of the Creston study area, eastern Washington
- Immersion studies on candidate container alloys for the tuff repository
- Impact of ENDF/B-VI cross-section data on H.B. Robinson cycle 9 dosimetry calculations
- Improved eddy-current inspection for steam generator tubing progress report for period January 1985 to December 1987
- Improved model for predicting J-R curves from Charpy data : phase I final report
- Improvements in motor operated gate valve design and prediction models for nuclear power plant systems : SBIR Phase I final report, September 1990-April 1991
- Inclusion of unstable ductile tearing and extrapolated crack-arrest toughness data in PWR vessel integrity assessment
- Influence of fluence rate on radiation-induced mechanical property changes in reactor pressure vessel steels : final report on exploratory experiments
- Initial assessment of the mechanisms and significance of low-temperature embrittlement of cast stainless steels in LWR systems
- Initial results of the influence of biaxial loading on fracture toughness
- Insights for aging management of light water reactor components
- Insights gained from aging research
- Interim fatigue design curves for carbon, low-alloy, and austenitic stainless steels in LWR environments
- Investigation of the Meers Fault in southwestern Oklahoma
- Investigations of irradiation-anneal-reirradiation (IAR) properties trends of RPV welds : phase 2 final report
- Irradiation effects on charpy impact and tensile properties of low upper-shelf welds, HSSI series 2 and 3
- Irradiation effects on fracture toughness of two high-copper submerged-arc welds, HSSI series 5
- Irradiation effects on strength and toughness of three-wire series-arc stainless steel weld overlay cladding
- Irradiation-anneal-reirradiation (IAR) studies of prototypic reactor vessel weldments
- J and CTOD estimation equations for shallow cracks in single edge notch bend specimens
- KEY analysis system user's guide : version 2.0
- Kansas-Nebraska seismicity studies using the Kansas-Nebraska microearthquake network : final report
- Laboratory analysis of fluid flow and solute transport through a variably saturated fracture embedded in porous tuff
- Laboratory testing of cement grouting of fractures in welded tuff
- Leak and structural test of personnel airlock for LWR containments subjected to pressures and temperatures beyond design limits
- Life assessment procedures for major LWR components
- Light water reactor pressure isolation valve performance testing
- Loading rate effects on strength and fracture toughness of pipe steels used in task 1 of the IPIRG program
- Long term performance and aging characteristics of nuclear plant pressure transmitters
- Long-term embrittlement of cast duplex stainless steels in LWR systems
- Long-term embrittlement of cast duplex stainless steels in LWR systems : semiannual report, April-September 1987
- Long-term embrittlement of cast duplex stainless steels in LWR systems : semiannual report, October 1990-March 1991
- Low-level radioactive waste disposal facility closure
- Low-level radioactive waste research program plan
- Low-level waste shallow land disposal source term model : data input guides
- Low-level waste source term model development and testing
- Lower-bound initiation toughness with a modified-Charpy specimen
- Managing aging in nuclear power plants : insights from NRC maintenance team inspection reports
- Materials and design bases issues in ASME code case N-47
- Measurement and modeling of sensitization development in stainless steels as a function of thermomechanical processing
- Mechanical characterization of densely welded Apache Leap tuff
- Mechanical properties of thermally aged cast stainless steels from Shippingport reactor components
- Memphis area regional seismic network : final report, October 1986-September 1992
- Methodology for reliability based condition assessment : application to concrete structures in nuclear plants
- Methods used for the treatment of non-proportionally damped structural systems
- Microearthquakes in Kansas and Nebraska 1977-1989 : final report
- Mission survey for the Pressure Vessel Research User's Facility (PVRUF)
- Model validation at the Las Cruces trench site
- Modeling one-dimensional radionuclide transport under time-varying fluid-flow conditions
- Models for estimation of service life of concrete barriers in low-level radioactive waste disposal
- Models of transport processes in concrete
- Motor-operated valve research update
- Multivariable modeling of pressure vessel and piping J-R data
- NRC model simulations in support of the Hydrologic Code Intercomparison Study (HYDROCOIN), Level 1, Code verification
- NRC research program on plant aging : listing and abstracts of reports issued through February 1, 1989
- NRC research program on plant aging : listing and summaries of reports issued through July 1992
- NRC research program on plant aging : listing and summaries of reports issued through June 1991
- NRC research program on plant aging : listing and summaries of reports issued through May 1990
- NRC research program on plant aging : listing and summaries of reports issued through September 1993
- Near-surface neotectonic deformation associated with seismicity in the Northeastern United States
- Neutron exposure parameters for the metallurgical test specimens in the sixth heavy-section steel irradiation series
- New England seismotectonic study activities during fiscal year ...
- New York/New Jersey regional seismic network
- Nondestructive examination (NDE) reliability for inservice inspection of light water reactors
- Nuclear plant aging research (NPAR) program plan : components, systems, and structures
- Nuclear plant aging research (NPAR) program plan : status and accomplishments
- Nuclear plant aging research on high pressure injection systems
- Nuclear plant service water system aging degradation assessment
- Numerical modeling of ductile tearing effects on cleavage fracture toughness
- Oklahoma seismic network : final report
- PR-EDB, power reactor embrittlement data base, version 1 : program description
- Paleoliquefaction features along the Atlantic Seaboard
- Parametric study of pipe whip analysis
- Peer review of the Three Mile Island unit 2 vessel investigation project metallurgical examinations
- Pennsylvania seismic monitoring network and related tectonic studies : final report
- Performance of intact and partially degraded concrete barriers in limiting fluid flow
- Performance of intact and partially degraded concrete barriers in limiting mass transport
- Piedmont seismic reflection study : a program integrated with tectonics to probe the cause of eastern seismicity
- Piping benchmark problems for the General Electric advanced boiling water reactor
- Piping research program plan
- Piping system response during high-level simulated seismic tests at the Heissdampfreaktor facility (SHAM test facility)
- Platinum catalytic igniters for lean hydrogen-air mixtures
- Post-Pennsylvanian reactivation along the Washita Valley Fault, southern Oklahoma
- Post-irradiation fracture toughness characterization of four lab-melt plates
- Posttest analysis of a 1:6-scale reinforced concrete reactor containment building
- Posttest analysis of the NUPEC/NRC 1:4 scale prestressed concrete containment vessel model
- Posttest destructive examination of the steel liner in a 1:6-scale reactor containment model
- Potential change in flaw geometry of an initially shallow finite-length surface flaw during a pressurized-thermal-shock transient
- Potential safety-related pump loss : an assessment of industry data
- Power spectral density functions compatible with NRC regulatory guide 1.60 response spectra
- Prediction and mitigation of erosive-corrosive wear in secondary piping systems of nuclear power plants
- Prediction of check valve performance and degradation in nuclear power plant systems : final report
- Prediction of check valve performance and degradation in nuclear power plant systems : wear and impact tests : final report
- Preliminary assessment of the fracture behavior of weld material in full-thickness clad beams
- Preliminary structural evaluation of Trojan RCL subject to postulated RPV support failure
- Pressure vessel safety research for advanced reactors : semiannual progress report for December 1991-March 1992
- Pressurized thermal shock probabilistic fracture mechanics sensitivity analysis for Yankee Rowe reactor pressure vessel
- Pressurized-water reactor internals aging degradation study : phase I
- Pretest prediction analysis and posttest correlation of the Sizewell-B 1:10 scale prestressed concrete containment model test
- Prioritization of TIRGALEX, recommended components for further aging research
- Prioritization of reactor control components susceptible to fire damage as a consequence of aging
- Proceedings of the Fourth Workshop on Containment Integrity : held in Arlington, Virginia, June 14-17, 1988
- Proceedings of the Joint IAEA/CSNI Specialists' Meeting on Fracture Mechanics Verification by Large-Scale Testing : held at Pollard Auditorium, Oak Ridge, Tennessee
- Processing and interpretation of seismic reflection data near the Bane Dome in Bland County, Virginia
- Qualification process for ultrasonic testing in nuclear inservice inspection applications
- Quantitative measurement and modeling of sensitization development in stainless steel
- Radiation degradation in EPICOR-II ion exchange resins / prepared by J.W. McConnell, Jr., D.A. Johnson, R.D. Sanders, Sr
- Radiation embrittlement of the neutron shield tank from the Shippingport reactor
- Radionuclide accumulation by aquatic biota exposed to contaminated water in artificial ecosystems before and after its passage through the ground
- Radionuclide buildup in BWR reactor coolant recirculation piping
- Radionuclide distributions and migration mechanisms at shallow land burial sites : final report of PNL research investigations on the distribution, migration, and containment of radionuclides at Maxey Flats, Kentucky
- Recommendations for the shallow-crack fracture toughness testing task within the HSST program
- Regulatory analysis for USI A-40, "seismic design criteria" : draft report for comment
- Regulatory analysis for proposed resolution of USI A-47 : safety implications of control systems
- Relay test program : series I vibration tests
- Relay test program, series II tests : integral testing of relays and circuit breakers
- Report on annealing of the Novovoronezh unit 3 reactor vessel in the USSR
- Report on waste burial charges : escalation of decommissioning waste disposal costs at low-level waste burial facilities
- Results from the nuclear plant aging research program : their use in inspection activities
- Results of LWR snubber aging research
- Results of crack-arrest tests on two irradiated high-copper welds
- Review of ASME code criteria for control of primary loads on nuclear piping system branch connections and recommendations for additional development work
- Review of elastic stress and fatigue-to-failure data for branch connections and tees in relation to ASME design criteria for nuclear power piping systems
- Review of reactor pressure vessel evaluation report for Yankee Rowe nuclear power station (YAEC no. 1735)
- Review of structure damping values for elastic seismic analysis of nuclear power plants
- Review of the proposed materials of construction for the SBWR and AP600 advanced reactors
- Risk-based inspection : development of guidelines
- Round-Robin analysis of the behavior of a 1:6-scale reinforced concrete containment model pressurized to failure : posttest evaluations
- SANS investigation of low alloy steels in neutron irradiated, annealed, and reirradiated conditions
- SEN wide-plate crack-arrest tests using A 533 grade B class 1 material : WP-CE test series
- SHAG test series : seismic research on an aged United States gate valve and on a piping system in the decommissioned Heissdampfreaktor (HDR)
- Screening methods for developing internal pressure capacities for components in systems interfacing with nuclear power plant reactor coolant systems
- Second U.S. Nuclear Regulatory Commission international steam generator tube integrity research program : final project summary report
- Seismic fragility of nuclear power plant components (Phase II) : switchgear, I&C panels (NSSS) and relays
- Seismic fragility of nuclear power plant components (Phase II), Vol. 4, A fragility handbook on eighteen components
- Seismic investigations of the HDR Safety Program : summary report
- Seismic safety research program plan
- Seismological investigation of earthquakes in the New Madrid seismic zone : final report, September 1986-December 1992
- Seismological investigation of earthquakes in the New Madrid seismic zone and the northeastern extent of the New Madrid seismic zone : final report
- Selected fault testing of electronic isolation devices used in nuclear power plant operation
- Selection of earthquake resistant design criteria for nuclear power plants : methodology and technical cases
- Selection of siliceous aggregate for concrete
- Self-monitoring surveillance system for prestressing tendons : phase I small business innovation research
- Service life of concrete
- Service life of concrete
- Severe accident testing of electrical penetration assemblies
- Shear wall ultimate drift limits
- Shippingport station aging evaluation
- Short cracks in piping and piping welds
- Short cracks in piping and piping welds : semiannual report
- Simulation of liquid and vapor movement in unsaturated fractured rock at the Apache Leap Tuff Site : models and strategies
- Size effects on J-R curves for a 302-B plate
- Slow strain rate testing of a cyclically stabilized A 516 Gr. 70 piping steel in PWR conditions
- Soil characterization methods for unsaturated low-level waste sites
- Soil physical properties at the Las Cruces trench site
- Southern Appalachian regional seismic network
- Stability of cracked pipe under inertial stresses
- Stable isotopes of authigenic minerals in variably-saturated fractured tuff
- Static and simulated seismic testing of the TRG-7 through -16 shear wall structures
- Static load cycle testing of a low-aspect-ratio four-inch wall, TRG-type structure, TRG-5-4 (1.0, 0.56)
- Static load cycle testing of a low-aspect-ratio six-inch wall, TRG-type structure, TRG-4-6 (1.0, 0.25)
- Static load cycle testing of a very low-aspect-ratio six-inch wall : TRG-type structure, TRG-6-6 (0.27, 0.50)
- Statistically based reevaluation of PISC-II round robin test data
- Steam Generator Tube Integrity Program/Steam Generator Group Project : final project summary report
- Steam generator group project : task 13 final report : nondestructive examination validation
- Steam generator group project : task 7 final report : post-service baseline eddy current examination
- Steam generator group project : task 9 final report : nondestructive evaluation round robin
- Steam generator tube integrity program : phase II : final report
- Stiffness and damping properties of a low aspect ration shear wall building based on recorded earthquake responses
- Stiffness of low-aspect-ratio, reinforced concrete shear walls
- Structural Aging Program technical progress for period January-December 1992
- Structural integrity of water reactor pressure boundary components
- Studies of the pattern and ages of post-metamorphic faults in the Piedmont of Virginia and North Carolina
- Submergence and high temperature steam testing of class 1E electrical cables
- Sulfate-attack resistance and gamma-irradiation resistance of some Portland cement based mortars
- Summary of work completed under the environmental and dynamic equipment qualification research program (EDQP)
- Summary, analysis, and response to public comments on proposed amendments to 10 CFR parts 30, 40, 50, 51, 70, and 72 : decommissioning criteria for nuclear facilities
- Survey of PWR water chemistry
- System performance of high-level waste package components
- TMI-2 EPICOR-II resin/liner investigation : low-level waste data base development program for fiscal year 1989 : annual report
- TMI-2 vessel investigation project (VIP) metallurgical program
- TMI-2 vessel investigation project (VIP) metallurgical program
- TR-EDB, test reactor embrittlement data base, version 1
- Technical basis for evaluating electromagnetic and radio-frequency interference in safety-related I&C systems
- Technical findings document for generic issue 51 : improving the reliability of open-cycle service-water systems
- Technology, safety, and costs of decommissioning a reference boiling water reactor power station : comparison of two decommissioning cost estimates developed for the same commercial nuclear power station
- Technology, safety, and costs of decommissioning a reference pressurized water reactor power station : technical support for decommissioning matters related to preparation of the final decommissioning rule
- Technology, safety, and costs of decommissioning reference non-fuel-cycle nuclear facilities : compendium of current information
- Tectonic deformation revealed in baldcypress trees at Reelfoot Lake, Tennessee
- Tensile and J-R curve characterization of thermally aged cast stainless steels
- Tensile-property characterization of thermally aged cast stainless steels
- Tensile-property characterization of thermally aged cast stainless steels
- The Impact of LWR decontaminations on solidification, waste disposal and associated occupational exposure, annual report
- The Influence of precompression on the lower-bound initiation toughness of A 533 B reactor-grade steel
- The SAFT-UT real-time inspection system : operational principles and implementation
- The Seismic category I structures program results for FY ...
- The Shoreline environment atmospheric dispersion experiment (SEADEX)
- The behavior of shallow flaws in reactor pressure vessels : status report
- The effect of electric discharge machined notches on the fracture toughness of several structural alloys
- The effects of aging on BWR core isolation cooling systems
- The effects of solar-geomagnetically induced currents on electrical systems in nuclear power stations
- The high level vibration test program : final report
- The impact of thermal aging on the flammability of electric cables
- The leachability of decontamination ion-exchange resins solidified in cement at operating nuclear power plants
- The sealing performance of bentonite/crushed basalt borehole plugs
- The tensorial nature of effective porosity and large-scale dispersion coefficients : application to the Creston study area, eastern Washington
- Thermal overload protection for electric motors on safety-related motor-operated valves : Generic issue II.E.6.1
- Thermodynamics of technetium related to nuclear waste disposal : solubilities of Tc(IV) oxides and the electrode potential of the Tc(VII)/Tc(IV)-oxide couple
- Transport calculations of neutron transmission through steel using ENDF/B-V, revised ENDF/B-V, and ENDF/B-VI iron evaluations
- Transport calculations of radiation exposure to vessel support structures in the Trojan reactor
- Two new NDT techniques for inspection of containment welds beneath coatings : final report, October 1989-March 1990
- Two-parameter fracture mechanics : theory and applications
- U.S./French joint research program regarding the behavior of polymer base materials subjected to beta radiation
- Ultrasonic inspection reliability for intergranular stress corrosion cracks : a round robin study of the effects of personnel, procedures, equipment and crack characteristics
- Unirradiated material properties of Midland weld WF-70
- Unsaturated flow and transport through fractured rock related to high-level waste repositories : final report--phase II
- Unsaturated flow and transport through fractured rock related to high-level waste repositories : final report--phase III
- Unsaturated fractured rock characterization methods and data sets at the Apache Leap Tuff Site
- Use of linear reduced-stiffness analytical models to predict seismic response of damaged concrete structures
- Use of thickness reduction to estimate values of K
- User's manual for the NEFTRAN II computer code
- Utility financial stability and the availability of funds for decommissioning : an analysis of internal and external funding
- VAM2D--variably saturated analysis model in two dimensions : version 5.0 with hysteresis and chained decay transport : documentation and user's guide
- Validation of seismic probabilistic risk assessment of nuclear power plants
- Value/impact assessment of jet impingement loads and pipe-to-pipe impact damage : revised methods and criteria
- Verification of nonlinear piping response calculation with data from seismic testing of an in-plant piping system
- Verification of piping response calculation of SMACS code with data from seismic testing of an in-plant piping system
- Virginia regional seismic network : final report (1986-1992)
- Viscoplastic stress-strain characterization of A533 grade B class 1 steel
- Warm prestress modeling : comparison of models and experimental results
- Wellfield installation and investigations, Creston study area, eastern Washington
- pH sensors based on iridium oxide
Embed
Settings
Select options that apply then copy and paste the RDF/HTML data fragment to include in your application
Embed this data in a secure (HTTPS) page:
Layout options:
Include data citation:
<div class="citation" vocab="http://schema.org/"><i class="fa fa-external-link-square fa-fw"></i> Data from <span resource="http://link.library.in.gov/resource/87z6QJW7QK8/" typeof="Organization http://bibfra.me/vocab/lite/Organization"><span property="name http://bibfra.me/vocab/lite/label"><a href="http://link.library.in.gov/resource/87z6QJW7QK8/">U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Engineering</a></span> - <span property="potentialAction" typeOf="OrganizeAction"><span property="agent" typeof="LibrarySystem http://library.link/vocab/LibrarySystem" resource="http://link.library.in.gov/"><span property="name http://bibfra.me/vocab/lite/label"><a property="url" href="http://link.library.in.gov/">Indiana State Library</a></span></span></span></span></div>
Note: Adjust the width and height settings defined in the RDF/HTML code fragment to best match your requirements
Preview
Cite Data - Experimental
Data Citation of the Organization U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Engineering
Copy and paste the following RDF/HTML data fragment to cite this resource
<div class="citation" vocab="http://schema.org/"><i class="fa fa-external-link-square fa-fw"></i> Data from <span resource="http://link.library.in.gov/resource/87z6QJW7QK8/" typeof="Organization http://bibfra.me/vocab/lite/Organization"><span property="name http://bibfra.me/vocab/lite/label"><a href="http://link.library.in.gov/resource/87z6QJW7QK8/">U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Engineering</a></span> - <span property="potentialAction" typeOf="OrganizeAction"><span property="agent" typeof="LibrarySystem http://library.link/vocab/LibrarySystem" resource="http://link.library.in.gov/"><span property="name http://bibfra.me/vocab/lite/label"><a property="url" href="http://link.library.in.gov/">Indiana State Library</a></span></span></span></span></div>