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].

Auxiliary Beam Stress Improved Laser Welding for Repair of Irradiated Light Water Reactor Components

2019 ◽  
Author(s):  
Jian Chen ◽  
Jonathan Tatman ◽  
Zhili Feng ◽  
Roger Miller ◽  
Wei Tang ◽  
...  
Keyword(s):  
Laser Welding ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
Department Of Energy ◽  
Light Water Reactor ◽  
Development Effort ◽  
Light Water ◽  
Water Reactor ◽  
Oak Ridge

Abstract The welding task focuses on development of advanced welding technologies for repair and maintenance of nuclear reactor structural components to safely and cost-effectively extend the service life of nuclear power reactors. This paper presents an integrated research and development effort by the Department of Energy Light Water Reactor Sustainability Program through the Oak Ridge National Laboratory (ORNL) and Electric Power Research Institute (EPRI) to develop a patent-pending technology, Auxiliary Beam Stress Improved Laser Welding Technique, that proactively manages the stresses during laser repair welding of highly irradiated reactor internals without helium induced cracking (HeIC). Finite element numerical simulations and in-situ temperature and strain experimental validation have been utilized to identify candidate welding conditions to achieve significant stress compression near the weld pool during cooling. Preliminary welding experiments were performed on irradiated stainless-steel plates (Type 304L). Post-weld characterization reveals that no macroscopic HeIC was observed.

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.

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.

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.

scholarly journals Improving Nuclear Power Plant Safety with FeCrAl Alloy Fuel Cladding

MRS Advances ◽  
2017 ◽  
Vol 2 (21-22) ◽  
pp. 1217-1224 ◽  
Author(s):  
Raul B. Rebak ◽  
Kurt A. Terrani ◽  
William P. Gassmann ◽  
John B. Williams ◽  
Kevin L. Ledford
Keyword(s):  
Nuclear Power ◽  
National Laboratory ◽  
Alloy System ◽  
Department Of Energy ◽  
Normal Operation ◽  
Oak Ridge ◽  
Plant Safety ◽  
Operation Conditions ◽  
Fecral Alloy ◽  
Nuclear Power Plant Safety

ABSTRACTThe US Department of Energy (DOE) is partnering with fuel vendors to develop enhanced accident tolerant nuclear fuels for Generation III water cooled reactors. In comparison with the standard current uranium dioxide and zirconium alloy system UO2-Zr), the proposed alternative accident tolerant fuel (ATF) should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General Electric, Oak Ridge National Laboratory and their partners have proposed to replace zirconium based alloy cladding in current commercial power reactors with an iron-chromium-aluminum (FeCrAl) alloy cladding such as APMT. The use of FeCrAl alloys will greatly reduce the risk of operating the power reactors to produce electricity.

scholarly journals Application of the FAVOR-OCI Fracture Mechanics Computer Program to ASME Code Section XI, IWB-3610 Flaw Acceptance Criteria Evaluations

2018 ◽  
Author(s):  
Terry L. Dickson ◽  
Paul T. Williams ◽  
B. Richard Bass ◽  
Hilda B. Klasky
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Critical Values ◽  
Initiation Toughness ◽  
Acceptance Criteria ◽  
Oak Ridge ◽  
The Us ◽  
Asme Code

This paper presents an overview of added features in a new version of the FAVOR (Fracture Analysis of Vessels Oak Ridge) computer code called FAVOR-OCI. The original FAVOR code was developed at the US Department of Energy’s Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission (NRC). FAVOR is applied by US and international nuclear power industries to perform deterministic and probabilistic fracture mechanics analyses of commercial nuclear reactor pressure vessels (RPVs). Applications of FAVOR are focused on insuring that the structural integrity of aging, and increasingly embrittled, RPVs is maintained throughout their licensed service life. Based on the final ORNL release of FAVOR, v16.1, FAVOR-OCI extends existing deterministic features of FAVOR while preserving all previously-existing probabilistic capabilities of FAVOR. The objective of this paper is to describe new deterministic features in FAVOR-OCI that can be applied to analytical evaluations of planar flaws. These evaluations are consistent with the acceptance criteria of ASME Code, Section XI, Article IWB-3610, including Subarticles IWB-3611 (acceptance based on flaw size) and IWB-3612 (acceptance based on applied stress intensity factor). The linear elastic fracture mechanics (LEFM) capabilities of FAVOR-OCI also incorporate the analytical procedures presented in the Nonmandatory Appendix A, Analysis of Flaws, Article A-3000, Method of KI Determination, for both surface and subsurface (embedded) flaws. The paper describes a computational methodology for determining critical values of fracture-related parameters that satisfy ASME Code Section XI acceptance criteria for flaws exposed to multiple thermal-hydraulic transients. These compute-intensive analyses can be carried out with a single execution of FAVOR-OCI. The new methodology determines critical values by solving for either the point of tangency or point of intersection between applied KI versus time histories and a user-selected cleavage initiation toughness material property (e.g., ASME KIc, FAVOR Weibull KIc, or Master Curve Weibull KJc) for surface or subsurface flaws. Situations where warm prestress conditions apply can also be addressed. The paper highlights a need for this new capability via applications to a recent independent review of safety cases for RPVs in two Belgian nuclear power plants (NPPs). That review required ASME Section XI assessments of several thousand embedded, quasi-laminar flaws in the wall of each RPV Analysis results provided by the new capability contributed to the technical bases compiled from several sources by the Belgian nuclear regulatory agency (FANC) and eventually used by FANC to justify the restart of these NPPs.

Codes, Standards and Regulatory Topics for Nuclear Facilities

2008 ◽  
Author(s):  
Taunia Wilde ◽  
Shannan Baker ◽  
Gary M. Sandquist
Keyword(s):  
Quality Control ◽  
Quality Assurance ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Quality Assurance Program ◽  
Nuclear Facilities ◽  
The Us ◽  
Regulatory Commission

The design, construction, operation, maintenance, and decommissioning and decontamination of nuclear infrastructure particularly nuclear power plants licensed in the US by the US Nuclear Regulatory Commission (NRC) or operated by the US Department of Energy (DOE) or the US Department of Defense (DOD) must be executed under a rigorous and documented quality assurance program that provides adequate quality control and oversight. Those codes, standards, and orders regulate, document and prescribe the essentials for quality assurance (QA) and quality control (QC) that frequently impact nuclear facilities operated in the US are reviewed and compared.

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.

scholarly journals Improved Heat Exchanger Lifecycle Prognostic Methods for Enhanced Light Water Reactor Sustainability

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Zachary Welz ◽  
Alan Nam ◽  
Michael Sharp ◽  
J. Wesley Hines ◽  
Belle R. Upadhyaya
Keyword(s):  
Heat Exchanger ◽  
Nuclear Power ◽  
Power Plants ◽  
Remaining Useful Life ◽  
Department Of Energy ◽  
Test Bed ◽  
Accelerated Degradation ◽  
The Us ◽  
Operational Information ◽  
Useful Life

As the licenses of many nuclear power plants in the US and abroad are being extended, accurate knowledge of system and component condition is becoming increasingly important. The US Department of Energy (DOE) has funded a project with the primary goal of developing lifecycle prognostic methods to generate accurate and continuous Remaining Useful Life (RUL) estimates as components transition through unique stages of the component lifecycle. Specific emphasis has been placed on creating and transitioning between three distinct stages of operational availability. These stages correspond to Beginning Of Life (BOL) where little or no operational information is available, early onset operations at various expected and observed stress levels where there is the onset of detectable degradation, and degradation towards the eventual End Of Life (EOL). This paper provides an application overview of a developed lifecycle prognostic approach and applies it to a heat exchanger fouling test bed under accelerated degradation conditions resulting in an increased understanding of system degradation. Bayesian and Bootstrap Aggregation methods are applied to show improvements in RUL predictions over traditional methods that do not utilize these methods, thereby improving thelifecycle prognostic model for the component. The analyses of results from applying these lifecycle prognostic algorithms to a heat exchanger fouling experiment are detailed.

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