U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Engineering
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- 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
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<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>
