[en] The objective of this project was to determine if the local brittle zone (LBZ) problem, encountered in the testing of the heat-affected zone (HAZ) part of welds in offshore platform construction, can also be found in reactor pressure vessel (RPV) welds. Both structures have multipass welds and grain coarsening along the fusion line. Literature was obtained that described the metallurgical evidence and the type of research work performed on offshore structure welds

[en] Specimen size effects on K_Jc data scatter in the transition range of fracture toughness have been explained by external (weakest link) statistics. In this investigation, compact specimens of A 533 grade B steel were tested in sizes ranging from 1/2TC(T) to 4TC(T) with sufficient replication to obtain good three-parameter Weibull characterization of data distributions. The optimum fitting parameters for an assumed Weibull slope of 4 were calculated. External statistics analysis was applied to the 1/2TC(T) data to predict median K_Jc values for 1TC(T), 2TC(T), and 4TC(T) specimens. The distributions from experimentally developed 1TC(T), 2TC(T), and 4TC(T) data tended to confirm the predictions. However, the extremal prediction model does not work well at lower-shelf toughness. At -150 degree C the extremal model predicts a specimen size effect where in reality there is no size effect

[en] The surface crack embedded in the clad layer of a reactor pressure vessel (RPV) has been identified as a critical safety assessment condition relative to the pressurized thermal shock accident scenario. This project was initiated to determine the severity of such cracks experimentally. It is the first study to investigate irradiated, clad vessel steel, and to identify the material property and stress conditions in the local region of the crack that are significant to the analysis. Bend bar tests provided the experimental simulation of the subject RPV surface crack. This report covers analysis techniques used and presents the findings indicated by the experimental results for irradiated and unirradiated materials. 13 refs., 13 figs., 2 tabs

[en] The surface crack embedded in the clad layer of a reactor vessel has been identified as a critical fracture safety assessment condition relative to the pressurized thermal shock accident scenario. This project was initiated to determine the severity of such cracks experimentally, using irradiated material, and to identify the material property and stress conditions in the local region of the crack that are significant to the analysis. Bend bar tests provide the experimental simulation of the subject RPV surface crack. This report describes the initial investigation using unirradiated material, addresses the analysis techniques, and presents the findings indicated by the experimental results

[en] The materials that are present in the local region of the clad layer of RPV steel were evaluated for tensile properties and fracture toughness before and after irradiation damage. Residual stresses in the clad region were determined. The information described herein was used to understand the behavior of surface cracks embedded in the clad layer in beam tests conducted in another phase of this investigation. 7 refs.; 12 figs., 4 tabs

[en] The purpose of this task is to examine the effects of neutron irradiation on the fracture toughness (ductile and brittle) of the HAZ of welds and of A 302 grade B (A302B) plate materials typical of those used fabricating older RPVs. The initial plate material of emphasis will be A302B steel, not the A302B modified with nickel additions. This decision was made by the NRC following a survey of the materials of construction for RPBs in operating U.S. nuclear plants. Reference 1 was used for the preliminary survey, and the information from that report was revised by NRC staff based on information contained in the licensee responses to Generic Letter (GL) 92-01, open-quotes Reactor Vessel Structural Integrity, 10CFR50.54(f).close quotes The resulting survey showed a total of eight RPVs with A302B, ten with A302B (modified), and one with A302 grade A plate. Table 5.1 in the previous semiannual report provides a summary of that survey. For the HAZ portion of the program, the intent is to examine HAZ material in the A302B (i.e., with low nickel content) and in A302B (modified) or A533B-1 (i.e., with medium nickel content). During this reporting period, two specific plates were identified as being applicable to this task. One plate is A302B and the other is A302B (modified). The A302B plate (43 x 42 x 7 in.) will be prepared for welding, while the A302B (modified) plate already contains a commercially produced weld (heat 33A277, Linde 0091 flux). These plates were identified from a list of ten materials provided by Mr. E. Biemiller of Yankee Atomic Electric Company (YAEC). The materials have been requested from YAEC for use in this irradiation task, and arrangements are being made with YAEC for procurement of the plates mentioned above

[en] Proper identification of transition temperature and shape of the lower-bound (K_lc) fracture toughness curve in the transition range has been a long-term objective. A past practice has been to test a large number of specimens of varying sizes, from 1/2T to 8T compacts, in expectation that size effects and statistical variability of (K_jc) could be resolved empirically. Recently, statistical and constraint-based models have been developed that purport to explain much of what has been seen. Weakest-link theory has been successfully used to predict specimen size effects for the lower part of the transition curve. Constraint-based models of β_c -- β_lc and J_ssy (small-scale yield) also can model size effects, but these tend to conflict among themselves with regard to the prediction of full constraint K_jc. All lack potential for defining the absolute lower bound of fracture toughness. Statistically based models have the benefit of quantifying data scatter characteristics and provide a basis for making lower-bound toughness estimates with assigned error estimates. The K_jc, data are obtained from small specimens, the size of which is dictated by volume limitations of surveillance capsule size. A basis has been explored for establishing a lower-envelope curve from such data

[en] The goal of this work was to develop ductile fracture toughness data in the form of J-R curves for modified A302 grade B plate materials typical of those used in reactor pressure vessels. A previous experimental study on one heat of A302 grade B plate showed decreasing J-R curves with increased specimen thickness. This characteristic has not been observed in tests made on recent production materials of A533 grade B and A508 class 2 pressure vessel steels. It was unknown if the departure from norm for the material was a generic characteristic for all heats of A302 grade B steels or unique to that particular plate. Seven heats of modified A302 grade B steel and one heat of vintage A533 grade B steel were tested for chemical content, tensile properties, Charpy transition temperature curves, drop-weight nil-ductility transition (NDT) temperature, and J-R curves. Tensile tests were made in the three principal orientations and at four temperatures, ranging from room temperature to 550F. Charpy V-notch transition temperature curves were obtained in longitudinal, transverse, and short transverse orientations. J-R curves were made using four specimen sizes (1/2T, 1T, 2T, and 4T). The fracture mechanics-based evaluation method covered three test orientations and three test temperatures (80, 400, and 550F). However, the coverage of these variables was contingent upon the amount of material provided. Drop-weight NDT temperature was determined for the T-L orientation only. None of the heats of modified A302 grade B showed size effects of any consequence on the J-R curve behavior. Crack orientation effects were present, but none were severe enough to be reported as atypical. A test temperature increase from 180 to 550F produced the usual loss in J-R curve fracture toughness. Generic J-R curves and curve fits were generated to represent each heat of material. This volume deals with the evaluation of data and the discussion of technical findings. 8 refs., 18 figs., 8 tabs

[en] The purpose of this task is to examine the technical basis for the currently accepted methods for shifting fracture toughness curves to account for irradiation damage, and to work through national codes and standards bodies to revise those methods, if a change is warranted. During this reporting period, data from all the relevant HSSI Programs were acquired and stored in a database and evaluated. The results from that evaluation have been prepared in a draft letter report and are summarized here. A method employing Weibull statistics was applied to analyze fracture toughness properties of unirradiated and irradiated pressure vessel steels. Application of the concept of a master curve for irradiated materials was examined and used to measure shifts of fracture toughness transition curves. It was shown that the maximum likelihood approach gave good estimations of the reference temperature, T_o, determined by rank method and could be used for analyzing of data sets where application of the rank method did not prove to be feasible. It was shown that, on average, the fracture toughness shifts generally exceeded the Charpy 41-J shifts; a linear least-squares fit to the data set yielded a slope of 1.15. The observed dissimilarity was analyzed by taking into account differences in effects of irradiation on Charpy impact and fracture toughness properties. Based on these comparisons, a procedure to adjust Charpy 41-J shifts for achieving a more reliable correlation with the fracture toughness shifts was evaluated. An adjustment consists of multiplying the 41-J energy level by the ratio of unirradiated to irradiated Charpy upper shelves to determine an irradiated transition temperature, and then subtracting the unirradiated transition temperature determined at 41 J. For LUS welds, however, an unirradiated level of 20 J (15 ft-1b) was used for the corresponding adjustment for irradiated material

[en] Warm prestress (WPS) behavior is the term commonly used to describe an apparent increase in material toughness of pressure vessel steels resulting from previous loading at a higher temperature. Such load histories are of interest largely due to the fact that loss of coolant accident (LOCA) and pressurized thermal shock (PTS) related load histories are expected to result in WPS behavior. While previous experimental work has demonstrated WPS behavior, insufficient attention has been given to separating material toughness variability for the WPS effect. There also appears to be a basic lack of understanding of the mechanism by which WPS behavior occurs and as a result, there is no generally accepted model or fracture criterion for predicting WPS behavior. The objectives of this study were to develop WPS data for which the enhanced toughness due to WPS could be separated from the K/sub Ic/ variability of the virgin material and to evaluate several candidate WPS models. 33 refs., 17 figs., 5 tabs