ML20031F077
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UNITED STATES
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/*4 NUCLEAR REGULATORY COMMtssION g
WASHINGTON, D. C. 20666 e
Sg 151976 y
MEMORANDUM FOR:
S. Levine, Acting Director Office of Nuclear Regulatory Research FROM:
B. Rusche, Director, Office of Nuclear Reactor.7egulation
SUBJECT:
SUPPLEMENTARY REQUEST FOR CONFIRMATORY RESEARCH FO'R EXPERIMENTAL DETERMINATION OF. FAILURE FRESSURES OF STEAM GENERATOR TUBES (SUPPLEMENT l.to RSR 76-?)
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REFERENCE:
B. C. Rusche to H. J. Kouts "rMfimatory Research for Experimental Detemination of Failure Fmessuces of Steam Generator T.laes," dated April 8j 1976 (See Att:ct..:ent)
In the referenced memo NRR requested RES to expetimentally evaluat!
the margins to failure of d6 grad 5d PWR steam generator tubes under nomal operating and postulated accident conditions.
This task was proposed to be accomplished in two phases. The first phase consists of burst j..
and collapse tests of artifically defected tubing in a simulated o .
steam generator environment and the detemination.of lehkage rates of tubes with various size through wall cracks. Th results,-from the first phase are needed expeditiously to enable are indepe'ndent verification of test data provided by major PWR vendors and to resblve the ' apparent inconsistencies in their resu",s. NRR will utilize this' data in detemining acceptable design allowances required in processing reactor licensing applications, and ir foming bases for continued operation of plants with degraded steam generator tubes.
In conjunction with RES's planning to implement a research program to l
satisfy NRR's need, as stated above, a meeting was held on August 6, 1976 with representatives from various NRC organizations that have direct I
responsibility in reviewing the safety issues related to the integrity of steam generator tubing. A number of safety issues were identified in this meeting which reflected the latest operational experience with Westinghouse PWR steam genercors. These irges are associatad with the margin of safety as a result of a reduction in-steam generator tube dianttee or the phenomenon of " denting" during operation. Therefore.
as a supplement to the confimatory tests requssted in the April 8,1976 referenced mamo, confimatory tests are required to investigate the l
safety conseqi:ences of continued operation with dented tubes.
In addition, correlation of service induced tube degradation to that
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simulated in these tests should be provided through eddy current inspection and empirical techniques.
Status of Problem i '
NRC was infomed by Westinghouse in December 1975 that several of thair plants had experienced steam generator tube defomation in the fom of (e110390142 810827
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SEP 151973 -
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, a reduction in tube diameter.
In May 1976, Westinghcuse removed dented tube samples and se@nents of the tube support plate from Surry Unit 2 and Turkey Point Unit 3, and discovered that the tube support plates were also cracked. The preliminary laboratory reports indicated that the annulus between the tube and the tube support plates was filled with a hardened corrosion product that expanded volumetrically to exert sufficient forces to " dent" the tube diametrically and to crack the tube support plate ligaments between the tube holes and the water circulation flowholes. The safety analysis provided by Westinghouse did not indicate an innediate safety concern. However, the degree of denting has increased in the Surry and Turkey Point plants, and with the tubes tightly locked in their support plates there is a high probability that secondary t'
stresses resulting from nonnal plant heat-up and shutdown can cause fatigue cracks, or thennal ratcheting. There is no test data to indicate the margins of safety in terms of both the number of pressure and temperature cycles associated with nonnal heat-up/ shutdown operation, and the pressure transients during postulated accident conditions. The dented tube has also complicated the eddy current inspection procedure, because it can prevent the passage of inspection probes.
For the purpose of establishing criteria to evaluate the consequence of continued operation of steam generators with dented tubes, NRR needs to obtain data regarding the cyclic life of dented tubes and their pressure retaining capability.
Information Deficiencies and Proposed Supplementary Program Some work has been done on the mechanical integrity of the dented tubes i
at room temperature, but no work has been done at steam generator l
operating temperatures or with temperature cycles.
It is apparent that experimental data on the integrity of dented tubes should be obtained in order for NRR to make an independent assessment of the margin of safety of all PWR steam generators that are affected with den'ed tubes.
NRR reconnends that the previously requested program as descrined in the referenced memo be expanded to include the following areas:
PHASE I Pre-test Inspection of Specimen All tube specimens should be inspected using an ultrasonic method for any pre-service defects such as ovality and non-concentricity of inner and outer surfaces. Amounts of tube ovality and non-concentricity should be quantified and their effects on the test results should be evaluated.
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1 SEP 151975
- Themal Ratcheting Data Pmvide an independent verification of the mechanical integrity of l.
severely dented Inconel 600 (ASME SB-163) tubes constrained in a simulated tube support plate hole that is diametrically closed and holds the tube in a fixed position.
It is recomended that the dented tube configurations be similar to those tested by Westinghuase for
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burst strength and flexing. NRR w111 supply the details of the t
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Westinghouse test configurations as needed. The dented tube should be pressure / temperature cycled concurrently for 400 cycles or to failure
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. whichever comes first at pressure and temperature ranges that occur l-e',
'with normal plant heat-up and shutdown conditions. Measurements of
'the tube wall defomation rates at both ends of the tube dent region -
I during the cyclic tests are desired. Examination for indications of crack initiation on both the ID and 00 tube surfaces of a few samples that sustained the cyclic test should be perfomed.
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Burst and Collapse Pressure Data
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The remaining dented or dented and wasted tube samples that sustained the pressure / temperature and cyclic tests and additional samples, as 1
I appropriate,'should tm tested to failure.
In addition, wastage and thinning of the tubes above the support plates should also be tested i
in conjunction with tube denting. Characterization of the wasted region should be provided by eddy current methods.
PHASE II Development of Empirical Models for the' Prediction of Steam Generator Tube Integrity 1
Actual test data parameters should be obtained in the test program to develop empirical relationships between the tube degradation charac-teristics and the failure pressures and leakage rates resulting from this test program. In addition, such empirical relationships should be correlated with either service induced (actual) flaws or environ-i mentally induced degradation. The empirical technique should be predictive in nature to determine the burst or collapse pressure given the inspection results of an actual steam generator field inspection, so that the margins of safety under nomal operating and accident conditions can be detennined.
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Interim Reporting Date for Supplementary Request l
In order to' meet licensing deadlines in late 1976 it would be desirable to obtain data on the mechanical integrity of dented Inconel 600 tubing in appmximately four (4) months.
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-4 SEP 151976
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m Licensing Impact The phenomena of steam generator tube " denting" with subsequent plant operation is of concern to NRR. NRR wishes to confinn that the integrity of dented tubes during normal operation and postulated accident conditions have adequate margins of safety to satisfy the requirements of Section III of the ASME Boiler and Pressure Vessel Code.
These concerns can only be satisfactorily answered utilizing experimental data obtained from independent tests perfomed under simulated operating or accident conditions. The infomation generated in this supplementary program will provide bases to realistically assess steam generator tube integrity in eleven (11) Westinghouse plants and one (1) Combustion Engineering plant that have experienced various degrees of tube denting.
_ Magnitude of Research Effort The Office of Nuclear Regulatory Research has estimated that the total program will be $350K. perimental tasks plus Phase I of the supplemental cost of the original ex I
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B.C.Rusche, director Office of Nuclear Reactor Regulation
Enclosure:
As stated cc:
V. Stello, Jr., OR R. Heineman, SS J. Reece. OR D. Eisenhut, OR J. Knight, SS L. Shao, OR l
W. Anderson, SD l
L. Frank, SD l
R. Stuart. OR W. Hazelton, OR L. Rubenstein, NRR T. Telford, OR M. Fairtile, OR W. Paulson, NRR l
F. A' meter, OR l
B. Liaw, OR l
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