ML19317E845
| ML19317E845 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Indian Point, Saint Lucie  |
| Issue date: | 05/30/1978 |
| From: | Liaw B Office of Nuclear Reactor Regulation |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19317E846 | List: |
| References | |
| NUDOCS 8001031005 | |
| Download: ML19317E845 (3) | |
Text
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HOTE TO:
D. G. Eisenhut, Assistant Director for Systems and Projects
\\p Division of Operating Reactors FROM:
B. D. Liaw, Acting Section Leader Engineering Branch I
Division of Operating Reactors THRU:
L. C. Shao, Chief 1
Engineering Branch Division of Operating Reactors i
SUBJECT:
REASSESSMENT OF STEAM GENERATOR TUBE LEAKAGE
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AT THE OCONEE HUCLEAR UNITS In a note dated May 3,1978 (enclosure 1) we informed you of a steam generator leak at Oconee Unit 1.
The unit was shut down on Frdidy, April 21 when the leakage rate exceeded the 0.3 GPM technical specification limit. Preliminary results from the ensuing steam generator inspection have been received. Three defects have been detected in steam generator 18. Tube 69-1, a peripheral tube seven rows from the open lane, was discovered to have a crack in the lower tube to tube sheet weld. The weld has been repaired.
In tube 74-2, located two rows from the open lane nad near the outside of the bundle, a 45* to 90* circumferential crack was detected near the y per tube sheet. A " ding like" indication was detected in tube 59-1, located in the periphery of the bundle.
All three tubes 69-1, 74-2, and 59-1 have been plugged. Duke Power Company reports that no significant increase in degradation has occurred in tubes that were inspected during this and previous inspections. More detailed infomation on the inspection and results are expected when Babcock and Wilcox (B&W) provides their final report to Duke Power.
With regard to fatigue related circumferential cracks, continued operation of the Oconec Units has been justified on the basis that these cracks would be quickly detected by a rapid increase in leak rate and that plant shutdown and corrective actions, in accordance with the technical specification, would occur in a timely fashion.
However, three seperate incidents indicate that fatigue related circumferential cracks can grow as large, as 90* circumferential cracks g
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circumferential cracks can grow as large as 90* circumferntially I
l and then stabalize at a leak rate below the 0.3 GPM technical specification limit.
In addition to the circumferentail crack j
discovered during this inspection, two similar defects have occurred in steam generators 28 and OS. Details of these defects i
were reported in a memo from L. C. Shao te J. G. Eisenhut dated February 1,1978 (enclosure 2). All of the circumferential fatigue i
cracks ovserved to date have occurred in the B steam generators within two rows of the open lane. Two of the tubes discussed above,
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77-2 and 74-2, are located near the outside of the tube bundle j
where lower cross flow velocities supply less energy for crack propagation. Tube 77-25 was located near the center of the open i
j tube land and it is postulated that the crack stablized in the tube j
because the plant was operated at reduced power levels resulting j
in lower flow velocities in the steam generators. These incidents indicate the potential for circumferential cracks existing in steam l
generator tubes during operation within the technical specification leakage limits.
Tube burst and collapse test were conducted by B&W and the results were submitted on the Oconee dochet on December 16, 1977. These tests included specimens with 120' and 360* circumferential slits ten mils wide and up to 70% through wall. None of the specimens reptured when subjected to internal pressures up to 5,000 PSI.
However, tubes withe 360* circumferential slits 50% and 70% through wall collapsed at 3500 psi and 2000 psi, respectively. These test results, although encouraging, do not specifically address the integrity of through wall cracks whichhmay respond differently under pressure loadings.
Futhermore, since evidence indicates that these type of defects might exist for long periods during nforal operation, the effects of loads.inposed during postulated accidents, should be analyzed.
It is futher noted that the tube testing program l
being conducted at Battelle - Pacific Northwest Laboratory oaes not,
l include through wall circumferential cracks.
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Based on the above discussion, a re-evaluation of the bases for continued operation of the Oconee Units ' appears to be warranted.
In addition to posing a safety concern, particularly under postulated ' W accident conditions, these types of defects are very difficult to i
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detect and have resulted in extensive down time at the Oconee Units..
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,9 N 0. G. Eisenhut It is our opinion that tests should be conducted to detemine the eesponse of tubes with through-wall circumferential cracks to Additionally, the alternate all postulated accident conditions.
preventative measure of plugging tubes located on the open tube land should be re-evaluated :onsidering the timelines of completing tests as suggested above and the benefit of minimizing downtime i
resulting from long, extensive inspections for circumferential Without this type fatigue cracks which are difficult to detect.
of action we will be in a difficult position to assure the integrity 1
of all tubes during accioent conditions.
B. D. Liaw, Acting Section B Leader i
Engineering Branch
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Division of Operating Reactors j
Enclosure:
As stated 1
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D. Eisenhut, Assistant Director fiOTE T0:
for Systems & Projects Division of Operating Reactors i
L. C. Shao, Chief FR0!4:
Engineering Branch Division of Operating Reactors STATUS OF STEAli GEllERATOR OPERATION AT OC ST. LUCIE UillT 1, Af1D lilDIA!! P01riT UllIT 2 e,
SUBJECT:
APRIL 27,1978 s
t generators DATE:
The following is a summary of the current status of the s eam Unit 2.
at Oconec Unit 1, St. Lucie Unit 1, and Indian Point OC0 flee UllIT l_
t Oconee Unit 1 A primary to secondary steam generator leak was detecte The lenk in-20, 1970 at 4:30 A.!1.25 GPl4 at 9:00 P.14.
on the inorning of Thursday, Aidil creased steadily until it leveled off at a rate of 0.1 at 10:15 A.f4.
It then fluctuated just above 0.25 GR4 until Friday, April 2 ification when the leak rate exceeded the unit's 0.3 GP!4 technical s i
I leakage rate limit and the unit uns shut down. Inspection of the steam at 100*: power up to the time of shut down.
April 26 because generator for the leak did not begin until Wednr* day, t!o results from j
of excessive airborne reactivity in the containment.
Duke Power Company the on-going inspection are available at this time.lieve it is in g
has not postulated the nature of the leak a "B" and because of the sudden indication and re a l tively J.'l!
it is a leaking rariid increase in leak rate they do not believe thatDu i
they intend I
in establishing an inspection program and has indicated t ith previously tube plug.
to perform tube inspections possibly including 311 tubes w Although the inspection program has i
t known degradations. He are being continuously informed on the deve
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officially established i
as it becomes of this situation and will provide additional informat on would be prudent.
available.
ST LUCIE UtilT 1 ducting St. Lucie Unit 1 (Florida Power and Light Co.) is current
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The stecm gencrcine iur.pecticns c'uring their first ref uel ng ou l
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D. Eisenhut 21., 1976 has Decemberin design to unit which began commercial operation on l
Combustion Engineering Steam Generators simi ar ith AVT secondary water Millstone Unit 2 and has operated continually wThe in t leg tubes The average dent magnitude is in and 110 cold leg tubes in one steam generator, chemistry.
m dent magnitude observed indicated 55 tubes with dent signals.the rang following i
being 0.004".from florida Power and Light Company:
ations of the dented l
a tube map indicating the locations and e evi g the f (1) the number of condenser tube failures dur ndenser tub
- tubes, (ii) operation /the measured chloride levels following con (iii) i to look for signs of support plate crack ng.f the insp and (iv) 20, 1978.
We have also requested that a detailed report t rt date of May be IllDIAll P01f1T UflIT 2 d a section of the fio. 1 Con Edison (licensee) remove team generator flo. 23 support plate containing the Inconel tubes from sThe su On April 25, 1978 The an area - 14" by 5".
l ding the Inconel tubes by i
at Indian Point 2.
rows of tubes in columns 3 thru 13 occupy the EDM process and removal was through a the tube sheet area.
d hole, the support plate broke
- 38 inches below the fio.1 support plate near y from the support plate, guiding the support plate through the hanresu Thirteen tubes remained intact ed c3cked with magretite.
and three other tubes were loose. support plate in llowing:
Preliminary visual examination revealed the fo b s in the support Corrosion products appear to be holding the tu e 1.
l t closure on the hot plate.
Measurement of the flow slots indicate d on the cold side.
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an eliptical Flow holes between rows 1 and 2 were d i or dicieters is shepe where the difference in the cajor and m lot.
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- 1/4 inch.
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t as distorted
- 4. Flow holes between the second and third rows w I
as those between rows one and two.
distorted Flow holes beyond the 3rd row of tubes do not appear
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by boroscope observations.
holes and Cracks in' the support plate appeared between,the flow These cracks do the tube holes (ligaments) with missing tubes.pport plate removal 6.
not appear to have occurred during su quested to The licensee is preparing a report on this matter and has meet with the staff on May 5,1978.and start-up is exp 1978.
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'L. C. Shao, Chief Engineering Branch Division of Operating Reactors V. Stcllo, Jr.
cc:
J. Reece B. K. Grimes A. Schwencer P. Erickson T. Wambach M. Fairtile B. D. Liau F. M. Almeter D. Dilanni R. G. LaGrange J. R. Strosnider R.'Mattson J. P. Knight D
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