ML19210A115
| ML19210A115 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/07/1978 |
| From: | Herbein J METROPOLITAN EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19210A116 | List: |
| References | |
| GQL-1154, NUDOCS 7910240833 | |
| Download: ML19210A115 (8) | |
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METROPOLIIAN EDISON COMPANY
~__.
POST OFFICE BOX 542 HE ADING, PENNSYLVANI A 19603 TELEPHONE 215 - 929 3601 July 7, 1978 G01 ll5h Director of Nuclear Reactor Regulations 1.ttn :
R. '4.
Reid, Chief Cperating Reactors Branch No. h U. S. Nuclear Regulatory Commission
'4ashington, D.C.
20555
Dear Si:
Three Mile Island duelear Station, Unit 1 ('D!I-1)
Cperating License No. EPR-50 Docket No. 50-289 In response to a request by Mr. J. J. ~4att during the Systenatic Evaluatica Program Review Team visit to TMI-l on June 26 - 28, 1978, enclosed please find plan and elevation drawings of the Reactor Building Su=p.
Also enclosed are copies of pages from Section 6 of the Final Safety Analysis Report with information on RB surp level following a LOCA and NPSH data for the DER pumps, and a page from Section 11 of the FSAR concerning the R3 sump grating.
It would be appreciated if a copy of the final report be forwarded to
atropolitan Edison.
Sincerely, j
J. G. Herbeir.
Vice President-Generation JGE:*4SS :tas Enclosures g;cr20015 0
1489 266 Kg
~
d.
Opening the e=ergency ecoling coil isolation valve en the outlet side of the coil. Inlet valves are normally open
(.
for leak =cnitering purposes.
e.
Closing the normal ecoling coil isolation valve.
6.L ENGINEERED SAFEGUARDS LEAKAGE AIID RADIATION CONSIDERATIONS
- 6..1 UTRODUCTION 7.te use of nor= ally cperating equip =ent for engineered safeguards func-tions and the location of sc=e of this equip =ent cutside the reactor building requires that consideraticn be given to direct radiation levels frc= the fluids circulating in these syste=s and the leakage from these syste=s after fission products have accu =ulated.
The shielding for cc=ponents of the engineered safeguards is designed to meet the following objectives in the event of a =aximum hypothetical acci-ient:
a.
To provide protection for personnel to perform all operations necessary for =itigatien of the consequencos of the accident, b.
To previde sufficient accessibility in all areas around the str. tion to permit safe continued operation of the required equip =ent.
6.
.2 SU:?fARY OF POST-ACCDE';T RECIRCULATION Telleving a postulated reactor coolant syste rupture, flow is initiated in the =akeup and purif.%Ation and decay heat removal syste=s frc= the b ratei vater storage tank o the reactor vessel. Flow is also initiated by the reactor building spra'/ system to the building spray headers. When the b0 rated water storage tank inventory is exhausted, recirculation frc=
the reacter building su=p is initiated by the operator for both the decay hea; re=cval flow and the reacter building sprays. The post-accident recirculation flow paths include all piping and equipment external to the reae::r building as shown on Figures 6-2 and 6-3 up to the valves leading to the berated water storage tank.
The CH E available to the decay heat and react'r building spray pu=ps o
during the pos LOCA recirculation phase has been calculated based on:
" issued for construction" piping drawings, a.
b.
pipe and fitting losses calculated using the infor ation in Crane Technical Paper No. L10, c.
total flow in a single string (i.e., consisting of one decay heat and one reactor building spray pu=p served by a single su=p suction line) is L500 gp: cr 3000 gp: to the decay heau pu=p and 1500 gp: to the rea::or building spra'y pump, 1489 267
(
@ @ g 6-13
^=. 23 c1_n_72)
d.
su=p water temperatures and reacter building pressures given in Table 6-11, water level io the reactor building sump is 23h ft 6 in.,
e.
f.
R.B. spray pu=p shaft center line at elevation 262 ft 3 in.,
g.
decay heat pu=p shaft center line at elevation 263 ft 7 in.,
The NPSH's available to each pump, calculated frem the above data, are given in Table 6-11.
The minimum calculated NFSH's available to each pump, assuming a saturated su=p, are compared belov vith the NPSH's required.
Pu=o Item Decay Heat R. 3. Stray Flev Rate, gpm 3,000 1,500 Maximum Head Loss in Suction Piping, ft H 0 T.5 T.9 2
NPSH, ft H O 2
Available 13.h lk.h Required 10 5 13.8 The required NPSH's indicated above reflect measured test values. The certified perfor ance test hta for R.3. spray pump 3S-Pla are shown in Figure 6-8A.
In additien, the pump performance vill be verified prior to fuel loading (see page 13A-5 for Decay Heat Removal System Functional Test and page 13A-7 fcr Reactor Building Spray System Functiona'. Test).
(
An the event that the available NPSH is considered insufficiert while operating on recirculation from the reacter building su=p, the decay heat flev rate can be reduced from the control rec = by throttling the valves OH-VhA and DH-VL3 and/or the reacter building spray flev rate can be reduced frc= the control roc = by throttling valves ES-VlA and SS-V13.
There is "jeg" close valve control in the control rocs for these four valves. If finer centrol or 1cng term flow throttling of the decay heat flow rate is desired, manually operated globe valves DE-Vl9A and DH-V193 (normally vide cpen) may be utiliced to achieve this. Extension handles on these two valves permit their operation fz - the 305 ft f1cer elevation in the auxiliary building.
6-13a Am. 27 (8-15-72) a 1489 268
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O Table 6-11 Reactor Building Conditions for the Design Accident With Two Reactor Building Air Coolers and 1500 GPM Spray m
8uW Beactor Beactor Building th11 ding DK Pump AS Pump DH Pump BS Pump Total Steam Vapsr Sump NPSd NPSH NPSK Available g,g NPSH Available g,g Time.
Pressure.
Pressure.
Temp.
Temp.
Available.
Available.
Saturated Sump.
Saturate 4 Sump.
Period seconde psig pela F
F ft ft ft ft Initial (Before Acci-0 0
0 110 dent)
.s beginning of Recircu-b.200 3.9 27 135 210 24.1 25.1 13 4 14.6 lation Q
g Maximum Recirculation 8 koo 5.4 3.9 152 218 22.1 23.0 13.4 14.4
'g Sump Temperature N
Maximum Recirculation 11.800 5.7 4.1 154 213 26.b 27.4 13.4 14.4 M
Vapor Temperature n
[
(* Reactor building total pressure assumed to be equal to the saturation H
pressure of the sump water (boiling sump assumption).
O I-4 M
8 v
500 J 400 E
E 300 N
200 m
100 0
30 J
20 E NPSH
- J m
600 10 $
W n
(-
~
Au-1 0
BHP cd 200
- 5 22 0
=
0 1
2 3
4 5
Capacity, gpm x 10-3 148D1 a'D 0 DECAY HEAT REMOVAL PUMP CHARACTERISTICS THREE.\\flLE ISLAND NL' CLEAR STATION UNIT I N..) M l k q=.
,,4 FIGURE 6-6
- ce-( AM. 25 4-10-72)
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'.T L. P-S M 150 75 construction and :'LA93 I fen s+t e A'as t lie ry and %el veMii n.* Pul ld t n.rs cor.cre. pit
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y Heat Ps-tanaer Vault / collect s areiand Wt t. P-lA 20 35 mt considered part of liquid waste water seape.re and water from mine, floor S.S. I t - *1 WI.L-P- lin 20 li system, water norestiv collecteJ and equipment drains witt.in 'fX vault coce ret. pi t
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weste systen. Remot e pot ent is t tective continer 2F 2*-6" for conteetnet ton.
l TifrAL LIQt'ID CAFACITY Iff LIQUID WASTE SYSTD4 St*'PS - 2. 7% fi EE50
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