ML19261F046
| ML19261F046 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/06/1971 |
| From: | Ross D US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7910180663 | |
| Download: ML19261F046 (6) | |
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ATOMIC ENERGY COMMISSION j
p%ysg WASW NGTON. O.C. 20545 3
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July 6, 1971 i
R. C. DeYoung, Assistant Director for Pressurized Water Reactors, DRL,
i THRU: Charles G. Long, Chief, PWR Project Branch No. 2, DRL
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ROD EJECTION ACCIDENT ANALYSIS FOR THREE MILE ISLAND UNIT 1 (DOCKET NO. 50-289)
We have considered the dose conse'quences of the BSW rod-ejection accident in two separate ways:
1.
Consider the sole leakage path to be from the containment, with all of the released fission products distributed uniformly and instantaneously in the containment at time zero.
2.
Consider the leakage path to be through a prior leak in the steam generator, with all of the fission products distributed uniformly and instantaneously in the primary soolant.
On the basis of recent calculations by B&W, we can now evaluate a third model, which appea:s to be more realistic than the first two:
3.
Af ter the rod ejecti in occurs (this is a 2.76-in. primary system rupture) the RCS und rgoes a subcooled blowdown for 50 seconds.
Following this there I.s saturated blowdown. When the primary system pressure fallt below the setpcint of the OTSG secondary safety valves the sec ndary system can be isolated. Heat removal is accomplished by nr eup of cold water (2 HP pumps @ 500 gpm each) and release of hot ui ture to the containment.
ine primary pressure will hang up at about 1000 psi.
Fission products may leak through the steam generator path (prior leak) until the steam generator is isolated. Fission products can also escape via the containment path, at the design basis leak rate.
We have evaluated TMI-l on the basis of a;l three models, as described in Tables I-III respectively.
The third medel, in Table III, is recommended. For TMI-1, a relatively bad site, the 2-hour dose is 80 rad.
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R. C. DeYoung.
There are a few unanswered questions, such as:
1.
What is the quasi-equilibrium reactor coolant sys tem pressure, how is it calculated, and can the plant operator really isolate the steam generators? Would he know to do that? Would the operator really stop the emergency feedwater supply to one or both OTSCs?
2.
What is the appropriately conservative value of iodine release from the fuel?
3.
What are the consequences of using the No. 3 model for other PWRs?
This memo is intended to serve as an interim report on the development of an evaluation model.
Comments,and sugges tions will be appreciated.
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Demicad F. Ross PWR Project Branch No. 2 Division of Reactor Licensing cc:
Docket File (50-289)
PWR-2 Reading PWR Branch Chiefs H. Denton P. A. Morris J. Murphy D. Ross (3) 1486 152 me e-
-s TABLE I Release from Containment 7
1.
I-131 core inventory (TID-14844) a.
6.4 x 10 curies 13 rad b.
9.6 x 10 2.
Portion of Core that is affected 17.5% (Chap ter 14, T'!I-l FS AR)
/
2 3.
Break Area (2.76 in. hole) 0.n415 it 4.
Specifu Break Flow Rate ('f oo dy, '
2 approxir: ate value) 3000 lbc./sec-f t 5.
I'reak Flow rate for 2.76-in. hole a.
120 lba/sec.
b.
900 gpm (or less) 6.
Initial Primary Sys tem Inventory a.
11000 ft b.
484,000 lbs.
7.
Time to Release all of Primary System (assuming no mixing with incoming HP water) 4000 sec. or about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 8.
Assumed Time for Release of Fission Products to Containment 0
9.
Containment Leak Rate 0.2%/ day 10.
Leakage Fraction in Two Hours 1.67 x 10-(2 x 10' 12 )
x 11.
Airborne Iodine -131 Availabe for 11 Leakage at Time Zero 8.4 x 10 rad (9.6 x 1013). (0.175)
(0.10)
(.5)
(core total). (fraction). (releasi. ). (plateout) in DNB fraction 8
1.4 x 10 rad 12.
Leakage in Two Hours (8.4 x 1011).(1.67 x 10- )
1486 153 13.
Deple tion = (X/Q). (BR) 3.5 x 10 (10- ). ( 3. 5 x 10-4) 14.
Two llour Dose Without Sprays 49 rad (1.4 x 10 ) (3.5 x 10-7) 15.
Two-Hour Dose (f rom I-131).
11 rad With Sprays (DRF = 4.5) 16.
Approximate total dose from all iodines (11 x 1.8) 20 rad o
1486 L5'4
x TABLE II Leakage Throu ;h S team Generator, 1.
Primary-Secondary Leak Rate 10 gpm 2.
Total Carryover in Two IIours (assuming no pressure decrease) 1200 gpm l'
3.
Rad I-131 release to primary coolant 1.68 x 10 ~ rad (9.6 x 1013). (0.175).(O.10) 4.
Fraction of Primary C:.lant Carried
.l.36 x 10~
Over (4.53 m + 335 m )
10 5.
Rad release to secondary side of OTSG 2.28 x 10 rad (1.36 x 10-2) (1.68 x 1012) 6.
Plateout & partitioning 0.1 9
7.
Rad release to environs 2.28 x 10 8.
Depletion (as in Table I) 3.5 x 10-7 9.
Dose from I-131 S00 rad 10.
Approximate Total Iodine Do e 1440 rad (800 x 1.8) l 1486 155
t 1
TABLE III Combined Release "odel i
s 1.
- Primarf Leakage to Secondary, Prior to Isolation cf Steam Generator 50 gal.
2.
Dose Contribution From Primary-Secondary Leakage 60 rad
( 50 )
(1440) 1200 3.
Dose Contribution From Containment Leakage (no change)
- 20. rad 4.
Total Dose, All Iodines (Item 2 plus Item 3) 80 rad
- The 50 gallon figure was supplied by B&W.
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