Lisowyj B.,Omaha Public Power District OPPD |
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2010
After two decades of operation, austenitic stainless steel Control Element Drive Mechanism (CEDM) seal housings at a Pressurized Water Reactor (PWR) nuclear plant experienced Transgranular Stress Corrosion Cracking (TGSCC). In order to prevent the same cracking from occurring at the Fort Calhoun Nuclear Plant, a preventative program was initiated in 1999. All 37 CEDM seal housings have been inspected by using WesDyne Intraspect pancake and plus point eddy current probes. Examination of the eddy current data found that TGSCC was associated with localized areas of higher permeability (confirmed with a magnetometer). In order to quantitatively analyze the data, the normalized value from signal amplitude was defined as the arithmetic ratio between the absolute measurement of local permeability value (amplitude) and the eddy current signal value (amplitude) for the calibration standard axial notch. The data showed that in failed seal housings the normalized amplitudes were about three times greater than in non-cracked housings. Higher permeabilities were associated with cracked locations. The eddy current methodology therefore provides an empirical criterion to monitor when locally higher surface material permeability changes occur in order to determine the onset of TGSCC. © 2010 by ASME.
Gao Y.,Westinghouse |
Anderson V.K.,NEI |
Julka A.K.,700 Universe Blvd
PSAM 2014 - Probabilistic Safety Assessment and Management | Year: 2014
An overview and lessons learned of the latest 10 CFR50.65(a)(4) guidance update to include fire risk evaluations and its implementations in U.S. nuclear plants are presented. By December 1, 2013, all the US nuclear plants implemented this new NRC requirement for the Maintenance Rule (MR) (a)(4) program to include the fire risk evaluation and management actions as part of the existing at power MR (a)(4) program. This paper will introduce the background, the need, the requirement, the process and some of the implementation details of incorporating fire risk assessment within the existing MR (a)(4) program. This paper will also discuss some of the applicable program interactions within a nuclear plant, such as the interactions among Fire Protection, Appendix R PRA and Work Control programs and activities. All of these functions/programs are required to support a successful fire MR (a)(4) implementation. Some of the technical and implementation issues, such as use of safe shutdown analysis, qualitative and quantitative risk analysis application, equipment scoping and risk management actions are also discussed in the paper. Some of the lessons learned since the December 1, 2013 implementation of this new program are also presented.
Nilsson P.,European Spallation Source |
Lillberg E.,Westinghouse |
Wikstrom N.,FS Dynamics
Nuclear Engineering and Design | Year: 2012
This concerns Flow Induced Vibrations (FIV) in nuclear reactors and numerical analysis of such. Special attention is paid to structural excitation by sound generated remotely and turbulent flow around the structure. One hypothesis was that these phenomena can interact, so that the structure accumulates more energy from the flow if it also excited by sound from another source. In the studies, Fluid-Structure Interaction (FSI) is simulated with Large Eddy Simulations (LESs). It is shown possible to simulate excitation due to both acoustic and turbulence loads using the reported methods, at least qualitatively. The excitation levels are even of the right order of magnitude in some parts. However, there are some shortcomings in the modeling. The most important is perhaps the lack of non-reflecting boundary conditions. Another problem is the strong numerical damping in combination with demanding numerics for the selected solid solution methodology. Three cases are simulated, two for validation and one applied about steam dryers. For the applied case, it is concluded unlikely that excitation by the acoustic and turbulence loads can interact. The main reason is that the flow is controlled more by static geometrical factors, such as solid rotation sharp edges, than small deformations due to vibrations. © 2012 Published by Elsevier B.V.
Hall B.,Westinghouse |
Server W.,ATI Consulting |
Rosier B.,Westinghouse |
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2013
Uncertainty regarding radiation embrittlement at high fluence, indicative of extended operating life beyond 60 years for current operating pressurized water reactor (PWR) vessels, has been identified as a potential limiting degradation mechanism. There are limited U.S. power reactor surveillance data at fluences greater than about 4E19 n/cm2 (E > 1 MeV) currently available for comparison with existing embrittlement prediction models. Extended operating life to 80 years is projected to have vessel peak fluence approaching 1E20 n/cm2, for a small number of plants. The two current U.S. embrittlement models are contained in Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Revision 2, and the Code of Federal Regulations, 10 CFR 50.61a. This paper compares the latest available high fluence power reactor surveillance data to the predictions of these two models, and to another model that has been proposed as better for high fluence data based on combined test reactor and power reactor data from sources extending beyond the U.S. These comparisons indicate the fluence ranges and material groups where the different models deviate from the measured data. The results from these comparisons have been used to select materials for a proposed new PWR supplemental surveillance program (PSSP) that utilizes previously tested irradiated surveillance specimens reconstituted and subsequently re-irradiated to higher fluences. Copyright © 2013 by ASME.
IEEE MTT-S International Microwave Symposium Digest | Year: 2011
The Johns Hopkins University Radiation Laboratory was located in Baltimore, Maryland. Much of the early research at the Rad. Lab. was on proximity fuzes for surface based anti-aircraft artillery, evolving from World War II R&D. Pioneering work on millimeter wave propagation and component development was a major research effort during the early and mid 50's. The Director was Dr. Donald D. King. During his career Dr. King made many significant and documented contributions to the microwave technology. His career included the holding of many important positions in both the MTT-S and the IEEE culminating in being the President-Elect of the IEEE. © 2011 IEEE.
IEEE MTT-S International Microwave Symposium Digest | Year: 2011
A radar is designed to meet customer specifications of range performance, angle accuracy, etc. However, an airborne radar must operate in an environment which may be different from the assumptions used in writing the performance specification. In an aircraft installation, the radar antenna is protected from the atmospheric environment by a radome which may introduce distortions and reflections of the radar energy. In some installations, such as AWACS, the aircraft may intrude into the near-field of the antenna, with subsequent distortions and reflections. The atmosphere between radar and target introduces diffraction effects which frequently cause fluctuation of the received target signal. The surface of the Earth acts as a large, complex target. Modern pulse Doppler radars use range-Doppler processing to separate airborne targets from the backscatter from the Earth (dubbed clutter). The antenna is designed with low side-lobes to minimize "side-lobe clutter" and with adequate system frequency stability to handle "main-beam clutter." The forward scatter from the Earth frequently causes the radar to see a mirror reflection of the desired target, apparently below the Earth's surface. This may cause undesired effects in tracking the target. This paper discusses these environmental effects revealed in flight testing of two Westinghouse radars. © 2011 IEEE.
Olofsson F.,Lloyd's Register |
Knochenhauer M.,Lloyd's Register |
Ohlin T.,Westinghouse |
Adolfsson Y.,Lloyd's Register
11th International Probabilistic Safety Assessment and Management Conference and the Annual European Safety and Reliability Conference 2012, PSAM11 ESREL 2012 | Year: 2012
As a consequence of the Fukushima nuclear power plant accident, the safety of all EU nuclear plants is reviewed with the focus on evaluating the response of nuclear power plants to beyond design basis events. The Nordic Owners Group has initiated an analysis of different aspects related to the stress tests with the aim to make sure that aspects related to the capacity of the NPPs to withstand extreme external events have been properly addressed. The project covers potential effects from external flooding and other extreme natural events on Swedish and Finnish BWR plants. A review of existing analyses was performed with the intention to identify and reevaluate assumptions and simplifications in previous analyses which may conceal important aspects of events. A previous NOG project dealing with methodologies for analyzing certain external events was used as a reference, and defined types of initiating events with fast developments are used as benchmarks. Additional analyses have been performed with different approaches and can be utilized as benchmarks to existing analyses. Three sub-projects were defined within the project: Extreme water levels in the Baltic Sea and Skagerrak Historically observed events with extreme high sea levels in the Baltic Sea and Skagerrak and the development over time of such events were analysed. The objective was to evaluate if the development over time allows initiation of preventive measures. The work also included a survey of the potential impact from other factors affecting extreme water levels, such as climate change, ocean waves and a number of other phenomena. Methodology for the analysis of multiple external events A screening methodology for identification of potential combinations of external events has previously been developed mainly for appliance within Probabilistic Safety Assessment (PSA). Using this methodology as a benchmark, a screening methodology applicable for deterministic analyses was been suggested. Main heat sink accessibility Based on the status of the safety documentation for the nuclear facilities, areas where development of further guidance is needed were identified. The objective was also to suggest measures to prevent different causes of events causing unavailability of the main heat sink. Proposals requiring limited effort and a more comprehensive solution for diversification were presented. The implications for such measures were discussed. It was concluded that the work from the earlier NOG-project can still be used as a basis for further work although some questions require further analyses. Copyright © (2012) by IAPSAM & ESRA.
Charest J.A.,Altran GmbH |
Wood W.T.,Westinghouse |
Rodenberger G.,Spiniello Corporation
American Society of Mechanical Engineers, Power Division (Publication) POWER | Year: 2011
The use of insitu methods to rehabilitate buried or inaccessible piping systems is an emerging technology which has received recognition as an ASME XI Code approved alternate to traditional repair practices. This type repair requires the insertion of a carrier tube, containing thermosetting resins and reinforcement fillers into the host pipe. The length of the carrier tube is only limited by the pot life of the resin and restrictions of piping geometry. Once the insertion is completed, the resin is cured using hot water, air or steam, which is circulated through the host pipe. After curing, the resultant product is a form-fitted structurally reinforced resin pipe within the existing host piping. The Cured-In-Place Pipe (CIPP) has mechanical properties similar to those of fiberglass piping. The savings of this type repair, in both system downtime and replacement cost, is substantial. This methodology provides utilities with a technically sound, economically feasible solution to buried piping repairs and the ASME Section XI Code Case N-589 provides the requirements for materials, design and installation. Copyright © 2011 by ASME.
Dammann A.,ITER Organization |
Antola L.,Amec Foster Wheeler |
Beaudoin V.,ITER Organization |
Dremel C.,Westinghouse |
And 5 more authors.
Fusion Engineering and Design | Year: 2015
Internal components of the ITER Tokamak are replaced and transferred to the Hot Cell by remote handling equipment. These components include port plugs, cryopumps, divertor cassettes, blanket modules, etc. They are brought to the refurbishment area of the ITER Hot Cell Building for cleaning and maintenance, using remote handling techniques. The ITER refurbishment area will be unique in the world, when considering combination of size, quantity of complex component to refurbish in presence of radiation, activated dust and tritium. The refurbishment process to integrate covers a number of workstations to perform specific remote operations fully covered by a mast on crane system. This paper describes the integration of the Refurbishment Area, explaining the functions, the methodology followed, some illustrations of trade-off and safety improvements. © 2015.
Agency: GTR | Branch: Innovate UK | Program: | Phase: Innovation Voucher | Award Amount: 5.00K | Year: 2013
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