ML19246B629
| ML19246B629 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/14/1979 |
| From: | Bosnak R Office of Nuclear Reactor Regulation |
| To: | Mattson R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7907180165 | |
| Download: ML19246B629 (4) | |
Text
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-ccen1rilJiTr JUN 141979 NRR Rdg. File - M. Graff DSS:MEB Rdg. File NOTE T0:
R. J. Mattson FROM:
R. J. Bosnak
SUBJECT:
THI-2 STEAM GENERATOR B
Reference:
Note to R. J. Bosnak from R. J. Mattson of June 1,1979 on THI-2 Steam Generators A baseline eddy current inspection of steam generator B at TMI-2 during Decerter 1977 revealed approximately 400 tubes with defects ranging from minor dents to 95% through wall penetrations.1 Thirty five tubes with defects greater than 40% of the wall thickness required plugging. A majority of the eddy current signals (ECT) were indicative of a din >ple (ding) 1.e., reduction of inside diarseter witnout detectable reduction in wall thickne:s. These defects prcsumably occurred during the fabrication process. Other ECT signals were indicative of scab type defects.
Circumferential cracks can initiate at such locations under conditions of high cycle fatigue. It is sumized that excessive flow incuced vibrations may have cau]ed high cycle fatigue failures at other B&W plants notably Oconee 1, 2, and 3.
Concern about excessive tuce vibration at TH1 had been raised af ter the baseline inspection in December 1977.2 A test was designed to investigate the reduction in alternating stress by installation of tube sleeves at two beations of concern and addition of internediate supports at two different locatiens at the upper most tube span.
It is possible tnat during the period between tne Dece@er 1977 baseline inspection and the accident at TMI-2 in March 1979, circumferential cracks may have been initf ated at locatians of high flow induced vibrations.
It is estimated the., a circureferential crack witn a depth which had progressed to greater than seventy percent inrough wall would be unable to withstand the pressure and themal loads imposed during the Marcn 1979 transients. Such a crack would tnen be expected to poo through the remaining wall resulting in primary to secondary leakage.
Contact:
J. R. Rajan, OSS::EB, X27538/72 7907180/d
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JUN 141979 R. J. Mattson The bending and thermal stresses in the tubes during pressure and thermal transients similar to those that occurrtd during the TMI-2 accident were er tiaated by B&W recently.3 In the analysis of the postulated trar.11ent, primary system pressure builds to the maximum value associated with the safety valve setpoint. Since primary flow is unavailacle, a steam environment exists on the primary side of die tubes. On the secondary side, the steam generator is completely depressurized, boiled dry. A temperatura differential of approximately 4500F may exist across the steas generator tube walls under such conditions.
Initiation of auxiliary feedwater flow to the steam generator at this stage results in a rapid cooling of the tunes which in turn produces significant tensile loads in the axial direction. The bending stress on the tube outer wall due to a temperature differential, AT, of 450oF across the tube wall can be as high as 80 ksi tension which will cause plastic deformation of a portion of the tube.
A circumferential crack of depth seventy percent through wall or acre located in this region is likely to penetrate the tube wall and the crack opening is likely to increase resultireg in a primary to secondary leak. During a decrease in AT across the tube wall and a consequent reduction in bending stress, the circumferential crack would tend to close up, resulting in either a decrease or complete stoppage of the primary to secondary leak, depending on the size of the crack.
Other possible sources of primary to secondary leakage were also examined. These include:
a.
Leakage due to failure of the welds, attachments or other modifications made in the TMI Unit 2 steam generators to install instrumentation for monitoring vibration flow and pressure data.4 b.
Leakage as a result of other design modifications made in the steam generators for example:
(1) Tube sleeving modifications, (2)
Lane flow blockers (3) Auxiliary feedwater nozzle modifications, and (4) Secondary side, lane tube stiffness.
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JUN 141979 R. J. fiattson 5-Failure of steam generator tube plugs installed after the initial c.
baseline inspection of 1977.
d.
Leakage of the tubes due to other types of damage viz. wear, stress-corrosion cracking or erosion / pitting.
While the possibility of leakage due to these mecnanisms cannot be ruled out, the most probable cause appears to be high cycle fatigue cracking discussed earlier.
R. J. Bosnak, Chief Mechanical Engineering Branch Division of Systems Safety cc:
F. Schroeder, DSS J. Knight, DSS W. Minners, 055 J. Rajan, DSS Ui9 l
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References:
1.
Letter from H. Silver to Metropolitan Edison Company - Summary of Meeting on Steam Generator Tube Inspection. June 30, 1978.
2.
Letter from Metropolitan Edison Company - Steam Generator Tube Sleeve Qualification Program, Dec. 22, 1977 to S. A. Varga, NRC.
3.
B&W Report of May 7, 1979, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant Appendix 2 (Steam Generator Tube Thermal Stress Evaluation) 4.
B&W Report of Dec. 22, 1972 "Once Through Steam Generator Instrumentation Program for Three Mile Island #2."
Report No. 773570139 4
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