A Foundation for Stressor-Based Prognostics for Next Generation Systems

2002 ◽  
Author(s):  
Don Jarrell ◽  
Daniel Sisk ◽  
Leonard Bond
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Department Of Energy ◽  
Residual Lifetime ◽  
Dynamic Force ◽  
Next Generation ◽  
Accurate Identification ◽  
Degradation Rates

Pacific Northwest National Laboratory (PNNL) scientists are performing research under the Department of Energy Nuclear Energy Research Initiative (NERI) program, to develop a methodology for accurate identification and prediction of equipment faults in critical machinery. The 3-year project, on-line intelligent self-diagnostic monitoring system (SDMS) for next generation nuclear power plants is scheduled for completion at the end of FY 2002. The research involves running machinery to failure in the Laboratory by the introduction of intentional faults. During testing, advanced diagnostic/prognostic sensors and analysis systems monitor the equipment stressor levels, correlate them with expected degradation rates, and predict the resulting machinery performance levels and residual lifetime. Application of a first principles physics-based approach is expected to produce prognostic methodologies of significantly higher accuracies than are currently available. This paper reviews the evolution and current state of the maintenance art. It presents a key measurement philosophy that results from the use of condition based maintenance (CBM) as a fundamental investigative precept, and explains how this approach impacts degradation and failure measurement and prediction accuracy. It then examines how this measurement approach is applied in sensing and correlating pump stressors with regard to degradation rate and time to equipment failure. The specifics are examined on how this approach is being applied at PNNL to cavitation and vibration phenomena in a centrifugal pump. Preliminary vibration analysis results show an excellent correspondence between the (laser) motor position indication, the vibration response, and the dynamic force loading on the bearings. Orbital harmonic vibratory motion of the pump and motor appear to be readily correlated through the FFTs of all three sensing systems.

Enhanced VVER-1000 Fuel Technology and Performance

2009 ◽  
Author(s):  
H. Shah ◽  
R. Latorre ◽  
G. Raspopin ◽  
J. Sparrow
Keyword(s):  
Pacific Northwest ◽  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
The United States ◽  
Department Of Energy ◽  
United States Department ◽  
Safety Level ◽  
And Performance

The United States Department of Energy, through the Pacific Northwest National Laboratory (PNNL), provides management and technical support for the International Nuclear Safety Program (INSP) to improve the safety level of VVER-1000 nuclear power plants in Central and Eastern Europe.

Machining Test Specimens From Harvested Zion RPV Segments

2015 ◽  
Author(s):  
Thomas M. Rosseel ◽  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
Compact Tension ◽  
Department Of Energy ◽  
Oak Ridge ◽  
Damage Models ◽  
The Us ◽  
Machining Test

The decommissioning of the Zion Nuclear Generating Station (NGS) in Zion, Illinois, presents a special and timely opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, an international nuclear services company, the selective procurement of materials, structures, components, and other items of interest from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), cutting these segments into blocks from the beltline and upper vertical welds and plate material and machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for microstructural (TEM, SEM, APT, SANS and nano indention) characterization. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models [1].

scholarly journals Current Status of the Characterization of RPV Materials Harvested From the Decommissioned Zion Unit 1 Nuclear Power Plant

2017 ◽  
Author(s):  
Thomas M. Rosseel ◽  
Mikhail A. Sokolov ◽  
Xiang Chen ◽  
Randy K. Nanstad
Keyword(s):  
Mechanical Properties ◽  
Base Metal ◽  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Department Of Energy ◽  
Current Status ◽  
Oak Ridge ◽  
Years Of Service

The decommissioning of Units 1 and 2 of the Zion Nuclear Power Station in Zion, Illinois, after ∼ 15 effective full-power years of service presents a unique opportunity to characterize the degradation of in-service reactor pressure vessel (RPV) materials and to assess currently available models for predicting radiation embrittlement of RPV steels [1–3]. Moreover, through-wall thickness attenuation and property distributions are being obtained and the results to be compared with surveillance specimen test data. It is anticipated that these efforts will provide a better understanding of materials degradation associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service and subsequent license renewal. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the U.S. Department of Energy, Light Water Reactor Sustainability (LWRS) Program, coordinated procurement of materials, components, and other items of interest from the decommissioned Zion NPPs. In this report, harvesting, cutting sample blocks, machining test specimens, test plans, and the current status of materials characterization of the RPV from the decommissioned Zion NPP Unit 1 will be discussed. The primary foci are the circumferential, Linde 80 flux, wire heat 72105 (WF-70) beltline weld and the A533B base metal from the intermediate shell harvested from a region of peak fluence (0.7 × 1019 n/cm2, E > 1.0 MeV) on the internal surface of the Zion Unit 1 vessel. Following the determination of the through-thickness chemical composition, Charpy impact, fracture toughness, tensile, and hardness testing are being performed to characterize the through-thickness mechanical properties of base metal and beltline-weld materials. In addition to mechanical properties, microstructural characterizations are being performed using various microstructural techniques, including Atom Probe Tomography, Small Angle Neutron Scattering, and Positron Annihilation Spectroscopy.

Evaluation of Ultrasonic Phased-Array for Detection of Planar Flaws in High-Density Polyethylene (HDPE) Butt-Fusion Joints

2016 ◽  
Author(s):  
Matthew S. Prowant ◽  
Kayte M. Denslow ◽  
Traci L. Moran ◽  
Richard E. Jacob ◽  
Trenton S. Hartman ◽  
...  
Keyword(s):  
High Density Polyethylene ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Phased Array ◽  
National Laboratory ◽  
Cooling Water ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
High Density

The desire to use high-density polyethylene (HDPE) piping in buried Class 3 service and cooling water systems in nuclear power plants is primarily motivated by the material’s high resistance to corrosion relative to that of steel alloys. The rules for construction of Class 3 HDPE pressure piping systems were originally published as an alternative to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPVC) in ASME Code Case N-755 and were recently incorporated into the ASME BPVC Section III as Mandatory Appendix XXVI (2015 Edition). The requirements for HDPE examination are guided by criteria developed for metal pipe and are based on industry-led HDPE research and conservative calculations. Before HDPE piping will be generically approved for use in U.S. nuclear power plants, the U.S. Nuclear Regulatory Commission (NRC) must have independent verification of industry-led research used to develop ASME BPVC rules for HDPE piping. With regard to examination, the reliability of volumetric inspection techniques in detecting fusion joint fabrication flaws against Code requirements needs to be confirmed. As such, confirmatory research was performed at the Pacific Northwest National Laboratory (PNNL) from 2012 to 2015 to assess the ability of phased-array ultrasonic testing (PAUT) as a nondestructive evaluation (NDE) technique to detect planar flaws, represented by implanted stainless steel discs, within HDPE thermal butt-fusion joints. All HDPE material used in this study was commercially dedicated, 305 mm (12.0 in.) nominal diameter, dimension ratio (DR) 11, PE4710 pipe manufactured with Code-conforming resins, and fused by a qualified and experienced operator. Thermal butt-fusion joints were fabricated in accordance with or intentionally outside the standard fusing procedure specified in ASME BPVC. The implanted disc diameters ranged from 0.8–2.2 mm (0.03–0.09 in.) and the post-fabrication positions of the discs within the fusion joints were verified using normal- and angled-incidence X-ray radiography. Ultrasonic volumetric examinations were performed with the weld beads intact and the PA-UT probes operating in the standard transmit-receive longitudinal (TRL) configuration. The effects of probe aperture on the ability to detect the discs were evaluated using 128-, 64-, and 32-element PA-UT probe configurations. Results of the examinations for each of the three apertures used in this study will be discussed and compared based on disc detection using standard amplitude-based signal analysis that would typically be used with the ultrasonic volumetric examination methods found in ASME BPVC.

scholarly journals Risk-Informed Safety Margin Characterization

2009 ◽  
Author(s):  
Stephen M. Hess ◽  
Nam Dinh ◽  
John P. Gaertner ◽  
Ronaldo Szilard
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Safety Margin ◽  
The United States ◽  
Department Of Energy ◽  
United States Department ◽  
Plant Design ◽  
Time Horizons ◽  
Safety Margins

The concept of safety margins has served as a fundamental principle in the design and operation of commercial nuclear power plants (NPPs). Defined as the minimum distance between a system’s “loading” and its “capacity”, plant design and operation is predicated on ensuring an adequate safety margin for safety-significant parameters (e.g., fuel cladding temperature, containment pressure, etc.) is provided over the spectrum of anticipated plant operating, transient and accident conditions. To meet the anticipated challenges associated with extending the operational lifetimes of the current fleet of operating NPPs, the United States Department of Energy (USDOE), the Idaho National Laboratory (INL) and the Electric Power Research Institute (EPRI) have developed a collaboration to conduct coordinated research to identify and address the technological challenges and opportunities that likely would affect the safe and economic operation of the existing NPP fleet over the postulated long-term time horizons. In this paper we describe a framework for developing and implementing a Risk-Informed Safety Margin Characterization (RISMC) approach to evaluate and manage changes in plant safety margins over long time horizons.

Validation of Reanalysis-Based Offshore Wind Resource Characterization Using Lidar Buoy Observations

2020 ◽  
Vol 54 (6) ◽  
pp. 44-61
Author(s):  
Lindsay M. Sheridan ◽  
Raghavendra Krishnamurthy ◽  
Alicia M. Gorton ◽  
Will J. Shaw ◽  
Rob K. Newsom
Keyword(s):  
Pacific Northwest ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
The United States ◽  
Department Of Energy ◽  
Offshore Wind ◽  
Approximation Techniques ◽  
The North ◽  
Wind Resource

AbstractThe offshore wind industry in the United States is gaining strong momentum to achieve sustainable energy goals, and the need for observations to provide resource characterization and model validation is greater than ever. Pacific Northwest National Laboratory (PNNL) operates two lidar buoys for the U.S. Department of Energy (DOE) in order to collect hub height wind data and associated meteorological and oceanographic information near the surface in areas of interest for offshore wind development. This work evaluates the performance of commonly used reanalysis products and spatial approximation techniques using lidar buoy observations off the coast of New Jersey and Virginia, USA. Reanalysis products are essential tools for setting performance expectations and quantifying the wind resource variability at a given site. Long-term accurate observations at typical wind turbine hub heights have been lacking at offshore locations. Using wind speed observations from both lidar buoy deployments, biases and degrees of correspondence for the Modern-Era Retrospective Analysis for Research and Applications-2 (MERRA-2), the North American Regional Reanalysis (NARR), ERA5, and the analysis system of the Rapid Refresh (RAP) are examined both at hub height and near the surface. Results provide insights on the performance and uncertainty of using reanalysis products for long-term wind resource characterization. A slow bias is seen across the reanalyses at both deployment sites. Bias magnitudes near the surface are on the order of 0.5 m s−1 greater than their hub height counterparts. RAP and ERA5 produce the highest correlations with the observations, around 0.9, followed by MERRA-2 and NARR.

Code Acceptance of a New Joining Technology for Storage Containments

2009 ◽  
Author(s):  
Gary R. Cannell ◽  
Glenn J. Grant ◽  
Burton E. Hill
Keyword(s):  
Pacific Northwest ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Department Of Energy ◽  
Fusion Welding ◽  
Radioactive Materials ◽  
Friction Stir ◽  
Joining Technology ◽  
Consistent Basis ◽  
Remote Operations

One of the activities associated with cleanup throughout the Department of Energy (DOE) complex is packaging radioactive materials into storage containers. Much of this work will be performed in high-radiation environments requiring fully remote operations for which existing, proven systems do not currently exist. These conditions require a process that is capable of producing acceptable (defect-free) welds on a consistent basis; the need to perform weld repair, under fully-remote operations can be extremely costly and time consuming. Current closure-welding technologies (fusion welding) are not well suited for this application and will present risk to cleanup cost and schedule. To address this risk, Fluor and the Pacific Northwest National Laboratory (PNNL) are proposing that a new and emerging joining technology, Friction Stir Welding (FSW), be considered for this work. FSW technology has been demonstrated in other industries (aerospace and marine) to produce near flaw-free welds on a consistent basis. FSW is judged capable of providing the needed performance for fully-remote closure welding of containers for radioactive materials. The performance characteristics of FSW, i.e., high weld quality, simple machine-tool equipment and increased welding efficiency, suggest that this new technology should be considered for radioactive materials packaging campaigns. FSW technology will require some development/adaptation for this application, along with several activities needed for commercialization. One of these activities will be to obtain approval from the governing construction code to use the FSW technology. The American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME B&PVC) will govern this work; however, rules for the use of FSW are not currently addressed. A code case will be required to define appropriate process variables within prescribed limits for submittal to the Code for review/approval and incorporation.

Modeling Shock and Vibration on Used Nuclear Fuel During Normal Conditions of Transportation

2019 ◽  
Author(s):  
Nicholas Klymyshyn ◽  
Pavlo Ivanusa ◽  
Kevin Kadooka ◽  
Casey Spitz
Keyword(s):  
Nuclear Fuel ◽  
Numerical Models ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
The United States ◽  
Lessons Learned ◽  
Department Of Energy ◽  
United States Department ◽  
Fuel Cladding ◽  
Used Nuclear Fuel

Abstract In 2017, the United States Department of Energy (DOE) collaborated with Spanish and Korean organizations to perform a multimodal transportation test to measure shock and vibration loads imparted to used nuclear fuel (UNF) assemblies. This test used real fuel assembly components containing surrogate fuel mass to approximate the response characteristics of real, irradiated used nuclear fuel. Pacific Northwest National Laboratory was part of the test team and used the data collected during this test to validate numerical models needed to predict the response of real used nuclear fuel in other transportation configurations. This paper summarizes the modeling work and identifies lessons learned related to the modeling and analysis methodology. The modeling includes railcar dynamics using the NUCARS software code and explicit dynamic finite element modeling of used nuclear fuel cladding in LS-DYNA. The NUCARS models were validated against railcar dynamics data collected during captive track testing at the Federal Railroad Administration’s Transportation Technology Center in Pueblo, CO. The LS-DYNA models of the fuel cladding were validated against strain gage data collected throughout the test campaign. One of the key results of this work was an assessment of fuel cladding fatigue, and the methods used to calculate fatigue are detailed in this paper. The validated models and analysis methodologies described in this paper will be applied to evaluate future UNF transportation systems.

Quantitative and Qualitative Comparison of Light Water and Advanced Small Modular Reactors

2015 ◽  
Vol 1 (4) ◽  
Author(s):  
Fatih Aydogan ◽  
Geoffrey Black ◽  
Meredith A. Taylor Black ◽  
David Solan
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Department Of Energy ◽  
Passive Safety ◽  
Qualitative Comparison ◽  
Light Water ◽  
Small Modular Reactor ◽  
Safety Features ◽  
Physical Challenges ◽  
Integral Reactor

In recent years, several small modular reactor (SMR) designs have been developed. These nuclear power plants (NPPs) not only offer a small power size (less than 300 MWe), a reduced spatial footprint, and modularized compact designs fabricated in factories and transported to the intended sites, but also passive safety features. Some light water (LW)-SMRs have already been granted by Department of Energy: NuScale and mPower. New LW-SMRs are mainly inspired by the early LW-SMRs (such as process-inherent ultimate safety (PIUS), international reactor innovative and secure (IRIS), and safe integral reactor (SIR)). LW-SMRs employ significantly fewer components to decrease costs and increase simplicity of design. However, new physical challenges have appeared with these changes. At the same time, advanced SMR (ADV-SMR) designs (such as PBMR, MHR Antares, Prism, 4S, and Hyperion) are being developed that have improved passive safety and other features. This paper quantitatively and qualitatively compares most of the LW- and ADV-SMRs with respect to reactors, nuclear fuel, containment, reactor coolant systems, refueling, and emergency coolant systems. Economic and financing evaluations are also included in the paper. The detailed comparisons in this paper elucidate that one reactor is not superior to the others analyzed in this study, as each reactor is designed to meet different needs.

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