ML20054F050
| ML20054F050 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/11/1982 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| REF-SSINS-6820 5211-82-145, IEB-79-01B, IEB-79-1B, NUDOCS 8206150170 | |
| Download: ML20054F050 (15) | |
Text
,
GPU Nuclear
[
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P.O. Box 480 Middletown, Pennsylvania 17057 717-944-7621 Writer's Direct Dial Number:
June 11, 1982 5211-82-145 Office of Nuclear Reactor Regulation Attn: John F. Stolz Operating Reactors Branch No. 4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Sir:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Containment Flood Level Calculations Enclosed please find our analysis of the containment flood level which is being submitted for NRC review in accordance with Partial Initial Decision (PID) of December 14, 1981 p. 302 and which is supportive of our responses to IE Bulletin 79-01B.
The analysis also addresses why the calculated flood level is lower than that experienced during the TMI-2 accident.
In the ASLB hearings there was some ambiguity as to the refinement of the max 3.tum flood level from 5.94 ft. to 5.66 ft.
This adjustment was warranted beciuse the original calculation had assumed design tank volumes rather that available volumes (down to the nozzles). Additionally, an adjustment was made for the dome at the bottom of the steam generators which, for cor.venience, was originally assumed to be a right cylindar (See Figure 4).
Af ter the above calculations were performed, two additional leakage criteria were evaluated:
For the normal leakage from other systems in the reactor building analysis, 300 valves of an average size of 3 inches were assumed to be leaking with an average leakage of 10 cc/hr/in. of diameter.
This results in 2.3 gph or 1656 gal. in 30 days or.02 ft increase in water level over a 30 day time period.
(Maximum Flood Level after 30 days - 5.68 ft.)
In conclusion, the maximum flood level following a LOCA is conservatively calculated to be 5.66 ft.
Finally, assuming normal leakage from other systems, the maximum flood level af ter 30 days would be 5.68 f t.
All required instruments located in the Reactor Building are above these calculated levels.
Sincerely, i
~
F206150170 820611
/
00l PDR ADOCK 05000289 I. D. 3 PDR Eukill cc:
R. C. Haynes Director, TMI-l R. Jacobs GPU Nuclear is a part of the General Pubhc Utilities System
L -
I.
INTRODUCTION As an integral part of the Environmental Qualification of Electrical Equipment Program it became necessary to review the topic of submergence as a potential harsh condition to which instruments inside the containment building were exposed.
As a result of this concern the maximum flood level was c alcula t ed f or design basis accidents (Large i
Break 10CA and Steamline/Feedline Break).
This report documents the i
me thod and the calculations employed and compares the conditions that resulted at IMI-2 during and following the accident at TMI-2 with the postulated conditions at IMI-1.
II.
METHODS In order to determine the maximum Reactor Building (RB) flood level, the amount of water which could be added to the building must be calculated.
The targe break LOCA sequence involves rupture of the Reactor Coolant System (RCS) followed by automatic actuation of the safety systems.
The High Pressure Inj ec tion (HPI) system will inject water from the Borated Water Storage Tank (BWST) upon system ac tuation.
As the RCS pressure continues to f all, the passive Core Flood Tanks (CFTs) will inject their
, contents and the Low Pressure Injection (LPI) system will begin to inject.
Coincident with this event, a high RB pressure will actuate the RB spray, which will also inject water from the BWST as well as the contents of the spray additive (NaOH) tank.
Feedwaterline (FL) or Steamline (SL) breaks inside the Reactor Building differ f r om a LOCA in that no rupture of the RCS occurs.
Thus, a water volume is retained within the RCS, the CFTs are not actuated and the LPI does not inject.
Additional water sources, however, must be considered.
The break initiates RB spray due to high RB pressure.
In addition, the inventory of the steam generators and f eed water addition before feed isolation (before feed pump trip and pump coast-down),
as well as FL water inventory between the isolation valve and the break (which enters the reactor building due to fluid flashing in the feed water piping),
must be considered.
The RCS does not rupture and theref ore, its volume l
is not included.
For bo th the LOCA and Feedwaterline/Steamline break it is conservatively assumed that the Borated Water Storage Tank (BWST) and sodium hydroxide l
(NaOH) tank are emptied due to HPI and RB Spray actuation.
The volumes i
used were available maximums to provide conservatism.
The Emergency
(
Feedwater System (EFW) vill continue to add water through the break until it is also isolated by manual operator ac tion.
The rate at which this i
addition can occur is dependent upon the number of EW pumps which are I
operating and the hydraulics of the EFW.
Cavitating venturis have been installed and will limit the EW flow to the af fected steam generator to approximately 600 GPM.
Based on this flow rate about 90 minutes of l
inj ec tion would be required before the RB water level from a FL/SL break would be expected to exceed that calculated for a LOCA.
Since operator l
action will occur in much less than 90 minutes, it is clear that LOCA is
(
the limiting case for RB flooding. Therefore, the pos t-LOCA conditions I
are the limiting case and these are used in this analysis.
i i
The floodable volume /per foot of elevation of the Reactor Building was calculated by determining the unobstructed volume of. the building and then subtracting from it the non-floodable volumes of major. equipment and structures.
III.
ASSUMPTIONS The assumptions and conservatism used in calculating volumes of the major equipment and structures were:
1.
The reactor vessel cavity is not floodable. (See Figure 1.)
Dcav = 2 X Reav = 2 X 16'6" = 33' 2
3 Vcav = f(Deav /4 = 855 ft / f t Reav = radius of reactor vessel cavity Dcav = diameter of reactor vessel cavity Vcav = volume of reactor vessel cavity 2.
The enclosed elevator shaf t in the lower elevation of the Reactor Building is not floodable.
(See Figure 1.)
Les = ll ft Wes = 9ft Ves = LW = 99 f t3/ft Les = length of elevator shaf t Wes = width of elevator shaf t 3.
Maj or concrete volumes and maj or equipment volumes were conservatively estimated and deducted from the R3 volume in calculating the free volume.
(See Figure 1.)
- a. "D" Ring Re = 23 ' 9", Ty =4 ', le = ( 24 ' 9" + 20 ' 3") X 2 = 90 f t Vdr = 2 RT + LT = 1007 f t3/ft R, = radius of D Ring To= thickness of D Ring l
Le = leng th of D Ring b. RCDT Compartment L = S1 + S2 + (3 + 14 + $5 + 16 L = 9'6" + 7' + 30' + 3' + 9'6" + 11' = 70', T = l'6" Vredt = LT = 105 ft 3/ f t l
L = length of wall T = wall thickness i
l l
c.
RB Tapered Zone Vt = 296 ft 3/ f t (see Figure 5)
- d. Let down Cubicle Wall i
3 3 + W tj + W} t5 Vid = WIT 1 + W t2 + W t 2
4 Vid = 12' X 2' + 13' X 3' + 10 X2
+8'X4'+6'X2'
'Vid = 127 ftVft.
W = length of wall t = thickness of wall The RB sump is initially full as is the trench.
e.
Vs = 1170 ft Vtr = L X W X D Ver = 30 ' X 10 ' X 1.5 ' = 450f t.3
- f. Steam Generator Lower Heads Vsgh 13 fd3 /ft (See Figure 4) 4 Make up tank volume was not assumed to be inj ected into the RB since the makeup tank is isolated during HPI.
Vmut = 507 ft 3 5.
Entire RCS volume was assumed to be spilled into RB. (Note that in case of a LOCA water will still be contained in the reactor vessel at all times up to the nozzles. This volume was assumed to be spilled in this analysis.
With the liquid level at the top of the core approximately 2020 ft remains in the vessel and 2980 f t remains in the loops.)
Vrv = reactor vessel volume Vsg = steam generator volume Vpzr = volume of pressurizer Vpsp = volume of pressurizer spray piping Vpsg = volume of pressurizer surge piping Vhl = volume of hot leg piping Vel - volume of cold leg piping Vap = volume of reactor coolant pump s Vres = Vrv + 2Vsg + Vpzr + Vpsp + Vpsg + Vhl + Ver + 4Vrep Vres=4g58ft3 + 2(2030 f t3) + 800 ft3 + 2ft3 + 20f t3+ 979 f t3+ 1102f t3
+ 4(56ft )
Vres = 11,245 ft.3 6.
No credit is taken for the water mass which may be held up as steam in the containment atmosphere or due to localized pocketing on elevated containment floors.
7.
Air inleakage f rom outside the Reactor Building is negligible (2 gallons in 30 days). - See Table 3.
8.
No leakage from containment to the outside is assumed.
1
9.
No credit is taken for absorption by materials such 'as concrete or puddling in the building or the buildup of film from spray.
(See Figures 2 and 3).
IV.
CALCULATIONS OF FLOOD LEVEL A.
Reactor Building Floodable Volume per foot of Height 2
3 Vrb = fr Drb = 13,273 f t /f t 4
l
.i Drb = diameter of Reactor Building = 130 ft Vrb = volume of Reactor Building Vfv = Vrb - (Vcav + Ves + Vdr + Vredr + Vid + Vt + Vsgh) i Vfv = 10,771 ft3 /ft Vfv = floodable volume j
l B.
Water Volume (LOCA & SL/FL)
Vloca = 60,962 ft3
= 455,966 gal (Table 1) 3 V s1/f1 = 51,822 f t = 387,629 gal (Table 2)
C.
Flood Level (LOCA & SL/FL)
Lloca = 60,962ft = 5.66 ft 10,771 ft3/ft (286 '8" elevation)
Ls1/f1 = 51,822 ft 3 = 4.81 f t (285' 9-3/4" elevation) 3 10,771 ft /gt V.
COMPARISON WITH TMI-2 ACCIDENT CONDITIONS In order to further evaluate the accuracy of the calculated post-accident flood level, the TMI-2 experience was evaluated as follows:
J The maximum TMI-2 Reactor Building water inventory occurred well into the accident and was approximately 625,000 gal.
Several water sources contributed to this volume which are not appropriate for consideration in i
the TMI-1 analysis.
First, 180,000 gal, were contributed from the Reactor Building Emergency Cooling System through a stuck open relief valve.
The 180,000 gal.
volume was verified by two independent methods:
First, the volume of water added to the RCS from the BWST and other make-up sources (boric e
acid mix tank, demineralized water, and RC bleed tanks) was subtracted f rom the total RB water volume.
Second, the quantity of unborated water required to produce the observed sump chemistry conditions accounting f or known boron additions was calculated.
The relief valve was set at 125 psig although the combined discharge pressure of the river water pumps (which had new impe11ers) exceeded this value.
This same situation does
not exist on TMI-1.
THI-l has separate cooling coils for normal and emergency cooling and each coil relief valve is set at 200 psig.
A similar influx of water in THI-l will not occur since the shut-off head of the emergency cooling water pumps is less than 200 psig.
Thus, this source of water is not appropriate for inclusion within this analysis.
Second, a management decision was made at TMI-2 not to recirculate RB sump water. As a result, additional outside water sources (boric acid mix tank, demineralized water, and reactor coolant bleed tanks) were injected into the RCS to keep up with the leakage.
This decision, and the water additions, were made because of the specifics of the TMI-2 event.
They represent, however, a deviation from the planned post-event sequence and, therefore, are not appropriate for inclusion in the TMI-l analysis.
Furthermore, the lessons learned from TMI-2 (i.e., cor.tainment isolation, high point vents, post-accident shielding, and leakage reduction) have been incorporated in TMI-1 per NUREG 0737.
Thus, a decision to not recirculate sump water is not expected.
VI.
VITAL INSTRUMENTATION LOCATION Steam generator and pressurizer water level measurements are key parameters used to monitor plant operation and accomplish safe shutdown.
Information provided by such measurements is important for maintenance of natural circulation cooling in the reactor coolant system.
The level transmitters for the steam generator and pressurizer level instruments are located in the lowest elevation of the Reactor Building ac a height above the floor that may make them vulnerable to flooding following certain postulated accidents.
These transmitters were subsequently raised above their initial location to position them above the water level resulting from the worst case design basis accident.
The lowest instrument elevation is 5.81 ft. above the RB floor as measured from the bottom of the transmitter housing.
The electronics are actually located 0.19 f t. above this point.
VII.
CONCLUSIONS Based on the above conservative evaluation and the assumptions stated in Section II, it is concluded that the maximum flood level for the Reactor Building is 5.66 ft above the floor in the immediate time frame following a LOCA.
The pressurizer and steam generator level instruments as installed are above the maximum water level that would occur.
The evert at TMI-2 is not directly comparable with the analysis for TMI-1.
It can be used for comparison only if appropriate corrections are applied.
Af ter subtracting those additions which are not applicable, and l
considering the specifics of the TMI-2 accident, it is concluded that the calculated flood level f or TMI-1 is conservative.
l
TABLE 1
$9 4 3 Sources of Water iq$
For a loss of Coolant c -
Accident (LOCA) 2 Water Sources Available Volumes - (ft. )
s 5
BWST 360,000 gal.
46,292
\\
y CFT (2) 2,080
,A-y*
R,,
NaOH Tank 12,750 gal.
1,345 N'; <
RCS Volume-11,245 s.,
1 455,966 gal.
TOTAL 60,962 ft.
=
l L
TABLE 2 Sources of Water Feedwater Line/Sterm Line Break (FL/SL)
Inside Containment Water Sources Available Volumes - ft.
BWST 46,292 NaOH Tank 1,345 Steam Generator 3,412 Feedwater flow (TOTAL) 205 Feedwater volume (TOTAL) 568 TOTAL 51,822'ft. = 387,629 gal.
1 1
1 4
1 i
i i
d
TABLE 3 Calculation of Water Addition to RB Sump Due to Air' Increase-a) Calculate the total weight of air inside the R.B.:
W = 5.3983 x 106 [
where W = Weight of air in RB, lb P = Partial pressure of air within the containment (psia)
Total - Water Vapor:: Total
=
T = Air Temperature at 80 F and - 2 psig (12.7 psia)
W = 1.2696 x 105 lb.
b)
Calculate wieght of air leakage for 30 days. Assuming outleakage is equal to inleakage at equal pressure differential:
at 2 psig max. allow. leakage = 0.02% wt. for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W air leakage for 30 days =
(W) (% allow, leakage / day) (30 days) = 761 lb air.
=
c) Calculate amount of moisture that will precipitate (assuming 80 % RH in the Reactor Bldg.
P l
RH = pw RH = relative humidity at)
Pw = vapor partial pressure Pwsat = Vapor partial pressure at saturation i
Pt = Total vapor pressure Amt. of moisture =
.622 x Pw l
Pt - Pw At 100 F & 100% RH outside air (14.7 psia)
Pw(sat) =.94924 psia (100% steam) lb H90 Amt. of moisture =
.622 x.94924 =.0429 l
14.7
.94924 lb dry air At 800F & 80% RH t
Pwsat =.50683 psia Pw =.8 x.50683 =.405464 psia
Amt. of moisture =-.622 x.405464 =.0176 lb H O 2
14.7
.405464 lb dry air Amt. of mositure condensation =
=(.0429
.0176) lb H O X 761 2
lb dry air lb dry air 30 days
= 19.25 lbs H O / 30 days 2
= 19.25 lbs/30 days
= 2.31 gal /30 days 8.33_1bs/ gal
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FIGURE 4 i
NON-FLOODABLE VOLUME BENEATH A STEAM GENERATOR f- ( VESSEL TAN.UNE OF LWR. HE AD 4 PR). FAC E CF TUP >E SHEST T
I q.
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. gT O\\4 SD
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4 STEAM i
GENERATOR OR SUPPORT LOWER HEAD Y
9 T
11'S' DIA.
The tangent line elevation is 9'-9" above the floor elevation per Babcock 6 Wilcox Drawing 131149E.
Thus the low point on the steam generator head will be 49.47 inches (4.12 ft.)
above the floor based on a head metal thickness of 8 inches (per Babcock & Wilcox Drawing 131120E).
l Tnus, the submergence will be 5.66 - 4.12 = 1.5 4 Ft.
The volume of this sphere segment is:
V = rr H ( R - )
H = 1.5 4 F t 77 R = 5.63 Ft W
/,")
3 3'
V=
38.1 Ft Total volume displaced by two steam gentrator heads is:
V=
76.2 Ft V
3 V
= - = 13.5 Ft /ft sg H
FIGURE 5 VOLUME OF REACTOR BUILDING TAPERED ZONE (REF. GAI DWG. E-421-030) 3/6' STEEL LINER P.,
1 Od 45'- 0" R To (
REACTOR BLD'Ca V].
SEE ENLARGED
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Assuming a final water height of H = 5.66 Ft r i.
I Approximate the shaded area as a Triangle f.j B = 2 TkNe 2
B = 1.46 Ft d
A=
2 4.13 Ft
=
Volume isrrDA D = Diameter of Center of l
Gravity of Area i
D = 130 Ft -
x 2
= 129 Ft f.,
n-V=
Tr(129)(4.13) = 1674 Ft V
3
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g = 296 Ft /Ft
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