ML19322C515
| ML19322C515 | |
| Person / Time | |
|---|---|
| Site: | Crane, Davis Besse  |
| Issue date: | 11/07/1977 |
| From: | Lauer J BABCOCK & WILCOX CO. |
| To: | Domeck C TOLEDO EDISON CO. |
| References | |
| TASK-TF, TASK-TMR BWT-1561, NUDOCS 8001170844 | |
| Download: ML19322C515 (3) | |
Text
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o TELEC0PY babcock &WilCOX
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P.o. Bax 12r4. Lyn-neur. Va. 24 A* S November 7,1977 Tmm m04 2ns1:2 BWT-1561 O
Q File: T1.2/12B p
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P.. Faist AHL/ DST JAL Records Center T1.2 Mr. C. R. Domeck R. C. Luken Nuclear Project Engineer W. H. Spangler Toledo Edison Company.
A. H. McBride Power Engineering & Construction G. A. Meyer 300 Madison Avenue C. W. Bruny, Mt.V.
Toledo, Ohio 43652*
J. P. Jones R. B. Davis
Subject:
Toledo Edison comoany E. Kane REPORT ON DEPRESSUP.I4ATION r
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Davis-Betse Unit 1
~T Be'rchin B&W Reference NSS-14
Dear Hr. Domeck:
Our letter BWT-1589 dated November 1,1977 foivarded input for a TECO repo'rt to NRC regarding the depressurization event on September 24.
To supplement that letter, you will find attached another writeup which evaluates the reactor coolant components.
It may be inserted into the previous report as Attachment B.
Very truly yours.
A. H.. Lazar Senior Project Manager W
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A. Lauer Attachant ject Manager cc:
J. D. Lenardson w/a J. C. Lewis g r. C D. J. DeLaCroix 1! O 4
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Babcock &Wilcox AUACHMENT B REACTOR C00 LAM COMPONEtiTS
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s B&W has comoleted their evaluation of the September 24 incident at Davis-Besse and found no harmful effects were incurred in the reactor vessel steam generator pres-surizer and primary piping pressure boundary.
During this rapid depressurization event the reactor coolant pressure dropped from 2300 psig to 930 psig in 7-1/2 minutes and gradually recovered to 1800 psig in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
During the firgt 7-1/2 ginutes the reactor outlet. temperature dropped at varying rates from 580 F to 533 F. For this evaluation it is assgmed that the total temperature drop occurred at the initial rate. This results in a 49 tercerature drop over a 6 minute period. Approximately 30 minutes after the fnitiai temperature drop a second slower and smaller temperature drop f~0m 540 F to 505 F occurred over a 21 minute period.
Following this second temperature ramp, the temperature gradually increased over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period to 528 F.
The reactor inlet temperature ramps and durations were 0
the same as for the reactor outlet temperature.
The secondary side pressure in steam generator No.1 reached a maximum of 1050 psig and decreased to 850 psig within 15 minutes and remained at that level. The secondary side pressure in steam generatoFNo. 2 reached a maximum pressure of 980 psig decreased to a minimum of 610 psig in 14 minutes, and returned to 850 psis in 2 minutes. Twenty minutes later the pressure again dropped to 610 psig and recovered gradually over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period.
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The Design Specificatien for the Davis-Besse I plant required evaluation of 40 cycles of a rapid depressurization event which included a drop in the reactor coolant pressure from 2200 pgi to 800 psi, a drop in the reactor coolant system average g
temperature from 553 F to 500 F in 15 minutes, and a drop of secondary pressure from 1050 psi to 640 psi.
The major difference between the actual transient and the design transient is the rate of the temperature drop in the reactor coolant system. The actual raie of temperature drop was twice the rate of the design transient but the total temperature change was only 78% of that of the design transient. The net result is that the fatigue usage of this one rapid depressurization is about the same as that predicted for one cycle.of the design transient.
As a more direct comparison, the transient event identified was analyzed for the reactor vessel shell and comoared to the design transient.
The results were that the range l
in thermal radial gradient stress for the actual transient was 5400 psi and the range of thermal radial gradient stress for the design transient was 6600 psi. TMs comparison would be representative of other thicknesses throughout the pressure boundary.
l The conclusions of the analysis are:
1.
Stresses in the pressureboundary did not exceed those already calculated on a design basis. This is verified by the actual pressure not exceeding 2500 psi and the themal transient being less severe than a combination of design transients for l
a rapid depressurization and a reactor trip.
A p.i h.-
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U i
r-l B-2 Babcock (Milcox Fatigue life of the reactor coolant components is not affected if one cycle of 2.
the reactor trip design transient and two cycles of a raoid depressurization design transient are considered to be used for this transient.
Two cycles of the rapid depressurization transient are necessary because the H?! syster was actuated twice 1
during the event and two cycles are necessary to reflect tne therr.a1 transient in the high pressure injection nozzle.
The effect of the intire event on the fatigue l;ife of the steam generators can be i
accounted for by using one cycle of the design transient for rapid depressurization i
j and one cycle of the design transient for loss of feedwater to one generator, I
3.
The effect of the change in water level on the pressurizer has a very minor effect i
on the pressurizer shell stresses. The pressurizer has been previously analyzed for the thermal effect of water-steam interface and the enange of level does not j
l affect that analysis.
4.
No significant thermal shock shnuld occur to the heaters since the heaters were de-activated due to a low water level sensor and not reactivated till the level re-covered.
i
- 5. 'No_ dynamic effects were caused by the rapid pressure decrease. No specific analysis was done but a dynamic response of the shells would require a large pressure drop in the order of milliseconds and the actual change was on the scale of minutes.
The reduced feedwater Y10w to steam generatot No. 2 was not sufficient to raintain a -
water level during the first five minutes of the event and this steam. generator The primary concern with a dry generator is the tube to.she,11 temperature boiled dry.
difference.
In this event a water level was established before the systein cooldown was started and acceptable tube to shell temperature differences were maintained.
This condition is similar to the loss of feedwater design transient followed by restart of a dry pressurized generator using the auxiliary feedwater system.
The burst rupture disc of the pressurizer quench tank resulted in a stream of steam and water imoinging on steam generator No. 2.
This stream removed a section of insulation 10' hign and 20' wide from the lower sh:11 of the generator and impinged The temperature of the impinging water was assumed directly gn the generator shell.A conservative evaluation of the rapid temperature change in th to be 212 F.
region of the vessel shell was perfor=ed.
The results of this evaluation indicate that this one event used less than it of the total fatigue life of the vessel. The predicted fatigue usage factor for the 40 year design life of the vessel in this area was less than 0.10. This jet impingerent did not significantly reduce the fatigue life of the generator.
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