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.

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].

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.

scholarly journals Friction Stir Welding and Preliminary Characterization of Irradiated 304 Stainless Steel

2019 ◽  
Author(s):  
Wei Tang ◽  
Maxim Gussev ◽  
Zhili Feng ◽  
Brian Gibson ◽  
Roger Miller ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Friction Stir Welding ◽  
Base Metal ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Weld Zone ◽  
Fusion Welding ◽  
Friction Stir

Abstract The mitigation of helium induced cracking in the heat affected zone (HAZ), a transition metallurgical zone between the weld zone and base metal, during repair welding is a great challenge in nuclear industry. Successful traditional fusion welding repairs are limited to metals with a maximum of a couple of atomic parts per million (appm) helium, and structural materials helium levels in operating nuclear power plants are generally exceed a couple of appm after years of operations. Therefore, fusion welding is very limited in nuclear power plants structural materials repairing. Friction stir welding (FSW) is a solid-state joining technology that reduces the drivers (temperature and tensile residual stress) for helium-induced cracking. This paper will detail initial procedural development of FSW weld trials on irradiated 304L stainless steel (304L SS) coupons utilizing a unique welding facility located at one of Oak Ridge National Laboratory’s hot cell facilities. The successful early results of FSW of an irradiated 304L SS coupon containing high helium are discussed. Helium induced cracking was not observed by scanning electron microscopy in the friction stir weld zone and the metallurgical zones between the weld zone and base metal, i.e. thermal mechanical affected zone (TMAZ) and HAZ. Characterization of the weld, TMAZ and HAZ regions are detailed in this paper.

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.

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.

Characterisation of Mechanical Properties by Small Punch Test

2014 ◽  
Vol 606 ◽  
pp. 15-18 ◽  
Author(s):  
Jan Siegl ◽  
Petr Haušild ◽  
Adam Janča ◽  
Radim Kopřiva ◽  
Miloš Kytka
Keyword(s):  
Mechanical Properties ◽  
Coming Out ◽  
Nuclear Power ◽  
Power Plants ◽  
Fracture Toughness Test ◽  
Small Punch Test ◽  
Charpy Test ◽  
Small Punch ◽  
Punch Test

The specific desired properties for structures and components working in critical environments (e.g. different structure parts of power plants) require current information about degradation processes coming out in materials. Obtaining of this information by the help of the classical tests of mechanical properties (tensile test, Charpy test, fracture toughness test, creep test etc.) is very limited namely in the case of nuclear power plants pressure vessel. Hence, the new innovative techniques based on miniaturized specimens have been developed for evaluation of mechanical properties and their changes. One of very promising techniques is Small Punch Test. Present paper deals with characterization of three different steels (15Ch2MFA, 10GN2MFA with different heat treatment and steel O8Ch18NT10 with various degree of deformation).

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.

Benchmark Accelerated Aging of Harvested Hypalon/EPR and CSPE/XLPE Power and I&C Cable in Nuclear Power Plants

2016 ◽  
Author(s):  
Robert C. Duckworth ◽  
Emily Frame ◽  
Leonard S. Fifield ◽  
Samuel W. Glass
Keyword(s):  
Mechanical Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Accelerated Aging ◽  
Functional Difference ◽  
Insulation Material ◽  
Cable Insulation ◽  
Control Rod ◽  
Oak Ridge

As part of the Light Water Reactor and Sustainability (LWRS) program in the U.S. Department of Energy (DOE) Office of Nuclear Energy, material aging and degradation research is currently geared to support the long-term operation of existing nuclear power plants (NPPs) as they move beyond their initial 40 year licenses. The goal of this research is to provide information so that NPPs can develop aging management programs (AMPs) to address replacement and monitoring needs as they look to operate for 20 years, and in some cases 40 years, beyond their initial, licensed operating lifetimes. For cable insulation and jacket materials that support instrument, control, and safety systems, accelerated aging data are needed to determine priorities in cable aging management programs. Before accelerated thermal and radiation aging of harvested, representative cable insulation and jacket materials, the benchmark performance of a new test capability at Oak Ridge National Laboratory (ORNL) was evaluated for temperatures between 70 and 135°C, dose rates between 100 and 500 Gy/h, and accumulated doses up to 200 kGy. Samples that were characterized and are representative of current materials in use were harvested from the Callaway NPP near Fulton, Missouri, and the San Onofre NPP north of San Diego, California. From the Callaway NPP, a multiconductor control rod cable manufactured by Boston Insulated Wire (BIW), with a Hypalon/ chlorosulfonated polyethylene (CSPE) jacket and ethylene-propylene rubber (EPR) insulation, was harvested from the auxiliary space during a planned outage in 2013. This cable was placed into service when the plant was started in 1984. From the San Onofre NPP, a Rockbestos Firewall III (FRIII) cable with a Hypalon/ CSPE jacket with cross-linked polyethylene (XLPE) insulation was harvested from an on-site, climate-controlled storage area. This conductor, which was never placed into service, was procured around 2007 in anticipation of future operation that did not occur. Benchmark aging for both jacket and insulation material was carried out in air at a temperature of 125°C or in a uniform 140 Gy/h gamma field over a period of 60 days. Their mechanical properties over the course of their exposures were compared with reference data from comparable cable jacket/insulation compositions and aging conditions. For both accelerated thermal and radiation aging, it was observed that the mechanical properties for the Callaway BIW control rod cable were consistent with those previously measured. However, for the San Onofre Rockbestos FRIII, there was an observable functional difference for accelerated thermal aging at 125°C. Details on possible sources for this difference and plans for resolving each source are given in this paper.

scholarly journals Process Optimization of Hexoloy SX-SiC Towards Improved Mechanical Properties

1993 ◽  
Author(s):  
G. V. Srinivasan ◽  
S. K. Lau ◽  
R. S. Storm ◽  
M. K. Ferber ◽  
M. G. Jenkins
Keyword(s):  
Mechanical Properties ◽  
High Temperature ◽  
Process Optimization ◽  
National Laboratory ◽  
Complete Characterization ◽  
Department Of Energy ◽  
High Temperature Application ◽  
Oak Ridge ◽  
Carbon Doped ◽  
Temperature Application

Hexoloy SX SiC materials, sintered with the addition of yttrium and aluminum containing compounds, have been demonstrated to possess higher toughness and strength than the boron and carbon doped Hexoloy SA [1]. Under a Department of Energy/Oak Ridge National Laboratory (DOE/ORNL) contract, a complete characterization was conducted on an SX composition selected for high temperature application. The particular composition selected for that study was Generation 1 SX-SiC (SX-G1), which contained about 2 wt% total additives.

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.

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