ML19350D686
| ML19350D686 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/30/1981 |
| From: | Marston T ELECTRIC POWER RESEARCH INSTITUTE |
| To: | |
| Shared Package | |
| ML19350D684 | List: |
| References | |
| TASK-2.K.2.13, TASK-TM TAC-45202, NUDOCS 8105180372 | |
| Download: ML19350D686 (16) | |
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EPRI RESEARCH ON THE
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PERTINENT TO THE OVERCOOLING TRANSIENTS 4
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5 The significance of an overcooling transient (OT) on the III integrity of reactor pressure vessels is related to many factors, but the principal contributors are the thermal stresses associated with the transient and the degree of radiation embrittlement.
This memorandum reviews briefly the irradiated materials research sponsored by the Electric Power Research Institute (EPRI) germaine to the OT issue.
The work is divided into four areas:
the prediction of radiation damage, the measurement of fracture toughness of irradiated materials, the measurement of crack arrest conditions for irradiated materials, and thermal anneal treatment of embrittled reactor vessels.
The radiation damage prediction research is significant to the OT issue because it helps determine what reactor vessel materials and therefore which power plants are of interest and defines the period of plant life for which them transients are of negligible importance.
The fracture toughness research should help define what combinations of flaw size and transient severity should be of concern.
The crack arrest research helps define the behavior of a given defect during and after the transient.
The last research area, thermal annealing, is pertinent because this is the only method considered for the mitigation of radiation embrittlement.
Each of the areas are to be discussed below in greater detail.
A one-page summary of each research l
i project is appended to this memorandum.
A paper summarizing the thermal anneal research is attached for further information.
Radiation Damage Predictions The OT need be only considered when the radiation embrittlement is sufficient to place the reactor vessel materials into the less tough, transition temperature range during the transient.
Radiation exposure tends to increase the temperatures at which ferritic materials undergo a precipitous increase in fracture toughness (fracture resistance).
This temperature is referred to as the NDT or RT 3r FATT.
The degree to which NDT a material is affected by radiation is a function not only of the total exposure (fluence) but the chemical composition and microstructure of the material.
The currently recognized predictive methodology for radiation damage is the NRC Regulatory Guide 1.99.1 (ref. 1).
The 1.99.1 relationships are bounding in philosophy therefore are not accurate predictions for many materials (ref. 2, 3).
In the area of prediction development, EPRI is currently sponsoring four projects, RP1021-3, RP1240-1, RP1553-1, and RP1553-2.
The focus is on the treatment of surveillance data accounting for the possible contributing effects j
of thermal aging, in situ annealing and alloy as well as trace elmental compositions.
The work in this area should be completed by the third quarter of 1981.
The end result is hoped to be an accurate predictive methodology for radiation damage, i
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~3-1 Fracture Toughness of Irradiated Materials The prediction of reactor vessel integrity relies on the use of fracture mechanics that integrates the effects of the component stresses, the flaws present, and the materials' resistance to fracture.
With it use the critical combinations of the three variables can be defined.
The data base of fracture toughness (resistance to fracture) for irradiated materials is very sparse.
Research projects 886-1 and 886-2 are designed to develop fracture toughness data on a range of reactor pressure vessel materials at increasing exposure levels.
The materials included in the program represent weldments (principal focus), plates and forgings with the emphasis on the more embrittlement sensitive chemistries.
The projects develop toughness data throughout the transition and into the upper shelf temperature ranges.
The results should be directly applicable to the OT investigation.
The relationship between the toughness indicated by the Charpy impact test and the fracture mechanics test is to be developed and/or verified.
The research is scheduled to be completed by August of 1981.
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Crack Arrest Conditions for Irradiated Materials The research described above should help define the necessary conditions for the initiation of cracking but the conditions for the arrest of cracking under accident situations are also important.
The material property necessary to define the conditions (crack size, stress level, etc.) for the termination of cracking is called crack arrest toughness.
Research project 1326-1 is developing crack arrest data on irradiated materials.
The materials are removed from actual reactor vessels and include two weldments and two plates representing both old and new fabricating and processing procedures.
These test results will be used to verify or modify the ASME code procedure for predicting crack arrest toughness from Charpy data.
The results may affect significantly the final depths of cracks following a transient and the potential for rupture if the vessel is repressurized.
The first data are being generated and the testing will continue into 1982.
Thermal Anneal of Reactor Vessels If the radiation embrittlement becomes too severe to insure vessel integrity, the only currently acknowledged procedure to j
reduce the accrued damage is thermal annealing.
The thermal l
I anneal procedure requires the vessel to be heated above its ww me-
~5-r normal operating temperature by a margin of 100*F to 300*F to permit the bake out of the radiation induced damage.
Although this technique has been demonstrated for a very small military I
reactor, it is untried on commercial scale vessels.
Research project 1021-1 is designed to determine the feasibility of and develop the methodology for annealing an embrittled reactor vessel.
The project is divided into two segments:
the development of annealing treatments and the assessments of the practicality of those treatments.
The first segment includes the n.ultiple irradiation of three embrittlement sensitive weldments to assess not only the initial damage and subsequent recovery, but the re-exposure sensitivity and the recovery following the re-exposure.
The second phase determines the procedures and evaluates the practicality and any possible detrimental side effects of the procedures.
It has already been determined that the so called
" wet anneal" (T s 650*F) is feasible, but may not be practical because of limited recovery (especially in transition temperature shift), required removal of the reactor fuel and lack of heat rejection capabilities.
This research is to be completed by July 1981.
Summary The memorandum reviews briefly the research on irradiated materials at EPRI that are pertinent to the OT issue.
The bulk
of the work should be completed by the end of 1981.
Although the preliminary results indicate some relief can be gained from improved material condition predictions, it appears prudent to avoid the conditions, if at all possible, that make the OT an issue to begin with.
That is the severe, or perceived to be severe, thermal down transients I)11 owed by significant repressurization.
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Radiation Damage Prediction 1 of 4
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4 RP1021-3 Steady State Radiation, Embrittl_eme,nt of Reactor Yessels Prime Contractor:
Westinghouse Electric Corporation (T. R. Mager)
Duration: March 1,1979 to May 30,1981 (to be extended to March 1982)
Project Cost:
$948,947 EPRI Project Manager:
T. U. Marston Objective: The objective is to verify that irradiation embrittlement in reactor pressure vessels reaches a steady-state or saturation 10 vel, and to provide the knowledge needed for revising and/or developing new irradiation damage trend curves.
Strategy: Through testing of reactor vessel surveillance capsules, analysis of test data, and microstructural and mechanistic studies, verify the consistency of a " steady-state" irradiation embrittlement which has been observed and determine the mechanisms involved in radiation damage saturation in pressure vessel steels.
Status: Nine capsules have been received, six capsules have been tested and two r,urveillance reports have been prepared.
Extensive metallurgical and ultrastructural investigations have been performed that indicate carbide alteration to be the principal embrittlement mechanism.
S. E. Yanichko et al., " Analysis of Capsule R From the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 2 Reactor Yessel Radiation Surveillance Program (WCA 9635)" EPRI RP1021-3 Topical Report, December 1979.
S. E. Yanichko et al., " Analysis of Capsule 125 From the Consumers Power Company Big Rock Point Nuclear Plant Reactor Yessel Radiation Surveillance Program (WCAP-9794)" EPRI RP1021-3 Topical Report, September 1980.
First Semiannual Technical Progress Report, October 1979.
1 Second Semiannual Technical Progress Report, April 1980.
Third Semiannual Technical Progress Report, October 1980.
T. R. Mager, " Analysis of Capsule R From the Wisconsin Public Service Corporation Kewaunee Nuchar Plant Reactor Yessel Radiation Surveillance Program (WCAP-9878)"
EPRI RP1021-3 Topical Report, March 1981.
T. R. Mager, " Analysis of the Maine Yankee Reactor Yessel Second Accelerated l
Surveillance Capsule (WCAP-9875)" EPRI RP1021-3 Topical Report, March 1981.
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Radiation Damage s
Prediction 2 of 4 RP1240-1 Collection and Evaluation of Data for Irradiated Pressure Yessel
_5 teel s Prime Contractor:
Fracture Control Corporation (R. A. Wu11aert)
Duration:
September 1,1978 to August 31, 1981 Project Cost:
$223,042 EPRI Project Manager:
T. U. Marston Objective: The primary goals of this program are to obtain and qualify the exist-ing data on mechanical properties and irradiation history for irradiated pressure vessel steels employed in commercial light water nuclear reactors, and to develop a weighting procedure whereby the " quality" of the data can be systematically assessed.
Finally a methodology for estimating the irradiated fracture toughness including K3c, JIc and J-R is to be developed and verified.
Strateqv: All available surveillance data are to be assembled and qualified.
Statistical analysis of the data using state-of-the-art modeling and statistical techniques will be made to develop embrittlement trends.
Finally fracture tough-ness properties are to be integrated with the embrittlement data.
Status: A computer data base containing data from 81 surveillance reports repre-senting 53 reactors and 66 capsules is available and a surveillance manual is to be available by March 1981.
R. A. Wullaert, P. McConnel and W. Oldfield, " Status of Radiation Embrittlement Trend Curve Data Base Re-evaluation," Proceedings of the Third ASTM-Euratom Meeting on Reactor Dosimetry, Ispra, Italy, September 1979, pp. ___.
K. E. Stahlkopf, G. R. Odette and T. U. Marston, " Radiation Damage Saturation in Reactor Pressure Yessel S~cel:
Data and Preliminary Model," Proceedings, Fourth International Conference on Pressure Vessel Technology, London, May 1980, pp.
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Radiation Damage Pr: diction 3 of 4 RP1553-1 Evaluation of Irradiation Response of Reactor Pressure Vessel Material s Prime Contractor; Combustion Engineering, Inc. (J. J. Koziol)
Duration: July 19,1979 to May 18,1981 Project Cost:
3170,616 EPRI Project Manager:
T. U. Marston Objective: The objective is to provide a means for predicting the Charpy Y-notch toughness reponse of reactor pressure vessel materials to power reactor neutron exposures by taking into account material and operating variables not presently considered but suspected of contributing to the observed variability in current predictive procedures. This effort will utilize existing commercial power reactor pressure vessel Charpy surveillance data and all pertinent information associated with material fabrication and irradiation conditions available from these programs.
Stra tegy: Tasks I and II will compare the indices obtained by using various computer curve fitting techniques with reported indices obtained by manual curve fitting.
These comparisons will be conducted on pre-and post-irradiation data contained in the EPRI data bank. The proposed curve fitting techniques will be statistically evaluated with respect to their ability to produce curves which accurately represent the classical regions of Charpy transition curves and selected indices. Task III will statistically examine Charpy transition tempera-ture shift and decreases in upper shelf energy, using analytical models coupled with actual test data, in an attempt to establish predictive methods for these parameters.
Status:
A review and update of the EPRI surveillance data base was made in l
1980.
Statistical analysis of the data is progressing after final completion of Tasks I and II.
J. D. Varsik, S. T. Byrne, "An Empiricel Evaluation of the Irradiation Sensitivity of Reactor Pressure Yessel Materials," Effects of Radiation on Structural Materi-als, ASTM STP 683, 1979, pp. 252-266.
Radiation Damage Prediction 1
4of4 RP1553-2 Application of ALN Modeling to Radiation Embrittlement i
Prime Contractor:
Adaptronics, Inc.
Duration: June 25,1979 to March 31, 1981 Project Cost:
371,962 EPRI Project Manager:
T. U. Marston Objective: The obji cive is to establish the foundation for developing a statistically-based model for the dependence of toughness-test response on temper-ature and fluence, incorporating the irradiation time, impurity chemistry, neutron energy density, material type and mechanical properties for base, weld and heat-affected-zone nuclear reactor steels. A successful radiation embrittlement tough-ness response model will contribute significantly to the development of statistically-based trend curves.
Strategy: Univariate statistical analysis and cluster / pattern recognition methods will be used to characterize the EPRI irradiated data base prior to modeling.
The Adaptive Learning Network (ALN) modeling concept will be used to develop the required statistically-based model.
Status: The EPRI data base on irradiated steels has been received and inf talled on the contractor's computer.
Complete empirical analyses of the raw data are available. Analysis of the curve !!tved data is in process.
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Fracture Toughness 1 of 2 a
RP886-1 Analysis of Radiation Embrittleent Reference Tounhness Curves
'L Prime Contractor:
Fracture Control Corporation (R. A. Wu11aert) l Duration:
September 11, 1976 to March 31, 1980 Project Cost:
5642,076 EPRI Project Manager:
T. U. Marston Objective: The objective of this program is to develop a statistically valid radiation embrittlement data base for use in the critical evaluation of the proce-dures for predicting the fracture toughness of irradiated reactor pressure vessel steels as currently specified in NRC Regulatory Guide 1.99 Strategy: Task I involves the development of statistically-based reference tough-ness curves for irradiated materials. Task II is directed towards establishing the relationship between radiation embrittlement measured by the CNrpy V-notch test and several fracture mechanics tests. Task III is concerned with the acqui-sition, qualification, improvement and ultimate application of radiation exposure parameters and the modeling of radiation damage.
Status: All tasks have been completed and c final report has been prepared. A statistically-based reference toughness procedure has been submitted to ASME.
Several advanced concepts have been developed for measuring radiation embrittle-ment from Charpy surveillance specimens. The uncertainties associated with radia-tion analysis have been defined and a radiation damage model has been developed.
W. L. Server and W. Oldfield, " Nuclear Pressure Vessel Steel Data Base," EPRI NP-933, December 1978.
R. A. Wu11aert and W. L. Server, " Fracture Toughness of Nuclear Pressure Vessel Steels from Small Specimens," Transactions of the 5th International Conference on Structural Mechanics in Reactor Technology, Vol. G, paper G 2/1.
G. R. Odette, "A Quantitative Analysis of the Implications of the Accuracy of Dosimetry to Embrittlement Predictions:
Past, Present and Future," Proceedings Third ASTM-Euratom Meeting on Reactor Dosimetry, Ispra, Italy, September 1979.
W. L. Server, et al., " Analysis of Radiation Embrittlement Reference Toughness Curves," Final report EPRI NP-1661, Palo Alto, California, January 1981.
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Fractura Taughness 2 of 2 RP886-2 Evaluation anc Prediction of Neutron Embrittlement in Reactor Pressura Yessel Materials Prime Contractor:
Naval Research Laboratory -(J. R. Hawthorne)
Duration: January 10, 1977 to September 15, 1981 Project Cost:
$1,150,000 EPRI Project Manager:
T. U. Marston
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Objective: Objectives are to develop a data base for the evaluation of r.urrent radiation embrittlement project methods and for the development of improved proce-dures, to investigate the relationship between radiation effects measured by the Charpy-Y (Cy) test method and fracture mechanics tests methods, to determine the radiation embrittlement sensitivities of a broad range of reactor vesul materi-als, and to experimentally assess the effects of selected composition variations.
Strategy: Changes in the fracture resistance of representative reactor vessel materials are determined and compared with irradiation at 288'C in a nuclear test reactor. Three neutron fluence levels are employed.
Radiation sensitivity is judged from the degradation of Cy notch ductility and the degradation of fracture toughness determined by fatigue precracked Cy and compact toughness ( F ) tests.
Results test embrittlement projection methods for a wide range of composition and metallurgical variations.
Status: The irradiations are complete. All of the impact tests are complete and approximately 80% of the compact fracture specimens are tested.
Findings to date indicate that current embrittlement projection methods are inconsistent, accurate in some cases to highly conservative in others.
Numerous examples of J-resistance curves for irradiated materials are presented.
J. R. Hawthorne, Editor, "The NRL-EPRI Research Program (RP886-2), Evaluation and Prediction of Neutron Embrittlement in Reactor Pressure Yessel Materials, Annual Progress Report for CY 1979 Part I Dynamic C, PCCy Investigations," NRL Report y
4431, December 31, 1980.
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Crack Arrest i
1 of 1 RP1326-1 Development of a Crack Arrest Touahness Data Bank for Irradiated RPV tbterials Prime Contractor: Westinghouse Electric Corporation (T. R. Mager)
Duration:
February 16, 1979 to April 30, 1982 Project Cost: 31,050,843 EPRI Project Manager:
S. W. Tagart, Jr.
Objective: The overall objective is to determine the effects of fast neutron exposure on the crack arrest toughness of reactor pressure vessel materials.
Strategy: Specimens fabricated from four materials characteristic of current reactor vessel materials will be irradiated in the University of Virginia Research Reactor to a fluence of about 1 x 1019 n/cr8(E>1.0MeV). The irradiated specimens sill be tested at Battelle Columbus Laboratories and Westinghouse Research Laboratories. Test data will be evaluated and compared to test data for unirradiated materials of the same types.
Statistical analyses will be performed to determine the effect of irradiation on crack arrest toughness.
Status:
Due to extension of the irradiation schedule at the University of Virginia reactor, this project is being extended over a longer time period.
The first irradiation cycle has been completed and testing is in progress.
The second cycle is currently in progress and the third cycle will begin in April 1981. Two semiannual progress reports have been received and reviewed by EPRI.
G. T. Hahn, et al., " Critical Experiments, Measurements, and Analyses to Establish a Crack Arrest, Methodology for Naclear Pressure Vessel Steels," BMI-2025,1979.
T. R. Mager, "Developnent of a Crack Arrest Toughness Data Bank for Irradiated RPV Materials, semiannual technical progress reports 1 and 2 through March 1980.
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Thermal Ann::aling 1 of 2 RP1021-1 Feasibility of and Methodology for Thermal Annealing an Embrittled Reactor Vessel Prime Contractor:
Westinghouse Electric Corporation (T. R. Mager)
Duration: June 20,1977 to May 31, IS?1 Project Cost:
$1,370,614 EPRI Project Manager:
T. U. Marston Objective: The overall objective is the development of an optimal in-situ thermal annealing methodology for reactor vessels which maximizes fracture toughness recovery, minimizes reexposure sensitivity, and minimizes downtime.
Strategy:
Through irradiation, annealing, and testing of specimens fabricated from material typical of reactor pressure vessel weldments, and analysis and evaluation of the resulting data, determine fracture toughness recovery as a function of annealing time and temperature,. determine post-anneal embrittlement sensitivity, and the kinetics of embrittlement, thermal annealing, post-anneal embrittlement, and subsequent reannealing for the materials. Based on the results, establish optimal themal annealing procedures for field application.
Status:
Irradiation of the three sets of capsules at the University of Virginia Research Reactor has been completed. heing of irradiated specimens and analysis of test results have been nearly completed Tor all irradiations. System evalua-tion for high temperature (dry) anneal are underway.
J. S. Schlonski, " Feasibility of Operating the Reactor Coolant System at Design Temperature f'or Reactor Vessel Annealing," EPRI RP1021-1 Topical Report,1978.
S. L. Anderson, "Characterizationn of the University of Virginia Research Reactor Radiation Envirorent,' EPRI RP1021-1 Topical Report,1979.
T. A. Meyer, "Desig:' and Fabrication of Specimen Irradiation Capsules for the Feasibility of and Methodology for Thermal Annealing an Embrittled Reactor Yessel Program," EPRI RP1021-1 Topical Report, 1979.
T. R. Mager, et al., " Feasibility of and Methodology for Thermal Annealing an Embrittled Reactor Vessel" Semiannual Technical Progress Reports No.1-6.
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Thermal Annesi...g r,..**
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RP1021-2 Feasibility and Mathndolocy for Thermal Annealing an Embrittled Reactor Vessel Prime Contractor:
University of Virginia (B. Shriver)
Duration: March 1, 1978 to June 30, 1982 Project Cost:
3584,103 EPRI Project Manager:
T. U. Marston Objective: The primary' objectives of this project are to determine the feasibil-ity of, and develop the methodology f,or tneraal annealing an embrittled reactor pressure vessel. This annealing treatment is designed to restore the mechanical properties reduced by neutron irradiation.
Strategy: The strategy of the project is as follows:
determine the extent of fracture toughness for ~ radiation ser.sitive materials as a function of thermal treatment; determine the post anneal embrittlement, anneal, reembrittlement and reanr.eal; if significant residual recovery is possible, establish optimal thermal anneal procedures for field application and evaluate technical risks of the pro-posed anneal procedures.
Status: The irradiations are being conducted under this contract at the Univer-s.ty of Virginia test reactor for contracts RP1021-1, RP1326-1 and RP1325-4.
Various UVAR semiannual reports prepared by Professor B. Shriver concerning 6
reactor operation; no specimen irradiation.
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