We describe the establishment and characterization of a novel hepatoma cell line. This cell line, designated RBHF‐1, was established from a hepatocellular carcinoma of a 67‐yr‐old man with a history of genetic hemochromatosis. At this writing, the cells have been maintained in RPMI‐1640 tissue‐culture medium and fetal calf serum without any additional supplements for 30 mo. The cells form colonies on soft agar and are not tumorigenic in nude mice. The cell line is polymorphic and displays characteristics of mature hepatocytes by synthesizing albumin, α2‐macroglobulin, fibronectin and α‐fetoprotein. Cytogenetic analysis shows multiple chromosomal aberrations, with a consistent deletion in the long arm and deletions or rearrangements in the short arm of chromosome 1. There is no evidence for hepatitis B or hepatitis C virus infection of the cell line. The cells contain no detectable intracellular iron after staining with Perls' stain. Unlike other hepatoma cell lines, there is no detectable binding of epidermal growth factor to RBHF‐1 cells. This is the first cell line to be established from a patient with hemochromatosis, and it provides a potentially important model for the study of hepatocyte transformation in association with iron overload. (Hepatology 1994;20:74–81.)
This paper will provide the bases for the requirements in the Beyond Design Basis Events (BDBE) evaluation performed in accordance with the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (BPVC), Section XI, Code Case (CC). The CC provides rules to facilitate the affected components Return to Service (RTS) after a BDBE magnitude Earthquake. The paper describes the bases for the examination of Reactor Coolant Pressure Boundary (RCPB) Structures, Systems and Components (SSCs) for determining the actual seismic loading and magnitude. The implications of the examination data based on design allowable stress values. The pipe loads determined shall be used to calculate stresses on the pumps and valves. The paper also describes the methodology for seismic events lasting longer than 100 cycles. The Cumulative Usage Factor (U) due to the event is calculated from individual cycles as; U = U1 + U2 + U3 + ... + Un. Aftershocks are accounted for in the methodology. Fatigue usage from the event that increases the total U to a value greater than 0.8 shall be included as high risk location(s) in the ASME BPVC, Section XI, Inservice Inspection (ISI) Program.
During Refueling Outage 18 (RFO18, April 2011) Pilgrim Nuclear Power Station (PNPS) identified crack-like indications on the Steam Separator Lifting Lugs. A multi-disciplinary engineering effort was undertaken to determine the cause of the cracking and prepare the technical justification for long term operation of the lifting lugs. This approach focused on addressing the potential for future degradation due to the existing indications, and the resulting effects on the hardware and its function. A materials evaluation concluded that intergranular stress corrosion cracking (IGSCC), most likely associated with cold work present in the as-fabricated steam separator, was the cause of the indications found on the lifting lugs. To support long term operation of the lifting lugs at PNPS, a structural evaluation was completed using ANSI N14.6 and NUREG-0612 criteria with a conservative bounding configuration. Crack growth rates, based on BWRVIP-76 guidance of 5E-5 in/hr for length and 2.2E-5 in/hour for depth, were used in the analyses. The evaluation concluded that PNPS could continue with long term operation of the Steam Separator. Consistent with standard practice, a general heavy loads examination was performed in 2013 (RFO19), confirming no discernable changes. The general examination will be repeated in 2015 (RFO20), and an examination of the lifting lugs is planned for the 2017 Refueling Outage (RFO21) to confirm that the indication behavior is consistent with the evaluation results.
As plants prepare for license renewal, extending their original operating licenses from 40 years to 60 years of plant operation, various tasks are required to be performed. Some of the required tasks include validation of the current fatigue evaluations for ASME Class 1 components, evaluation of environmental affects for a subset of those components, and calculation of revised component life to demonstrate acceptability for the entire license renewal period. Validation of the current fatigue evaluations is a significant challenge for older-vintage plants because the documentation of transient parameters and event severity was often not rigorous, especially during the early years of plant operation. This is primarily because historical data gathered in the early years was difficult to obtain before the implementation of modern computer systems and was therefore only collected at limiting times and locations. As a result, transients were typically considered to be full design basis severity events. Also, the design basis transient definitions, as defined in the original Final Safety Analysis Report (FSAR), were limited to a simpler, bounding set to help simplify the process. Whereas these simplifications are typically conservative and make it easier for the staff implementing plant procedures, they can challenge the ability of utilities to be able to demonstrate long-term acceptability for license renewal where additional margins are required. The purpose of this paper is to provide an example of the efforts expended on this issue for the Pilgrim Nuclear Power Station (PNPS). The intent is to educate the reader on the efforts involved, the vast array of input that is needed, and the possible hurdles that may be encountered for the validation and re-evaluation of Class 1 component fatigue for license renewal. Prior to starting work on the fatigue analysis, PNPS requested that SI develop a design specification for the fatigue reevaluation project. The importance of this document is that it maintained consistency and documented methods to be used along with design inputs.
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