ML13317A848
| ML13317A848 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 02/02/1982 |
| From: | Baskin K Southern California Edison Co |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML13317A849 | List: |
| References | |
| NUDOCS 8202040250 | |
| Download: ML13317A848 (12) | |
Text
Southern California Edison Company P. 0. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD. CALIFORNIA 91770 K. P. BASKIN TELEPHONE MANAGER OF NUCLEAR ENGINEERING, (213) 572-1401 SAFETY, AND LICENSING February 2, 1982 Director, Office of Nuclear Reactor Regulation 9
Attention:
D. M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Gentlemen:
Subject:
Docket 50-206 Auxiliary Feedwater System Review San Onofre Nuclear Generating Station Unit 1 As a result of several telephone discussions with members of the NRC staff, we committed to provide information regarding the impact of initiating auxiliary feedwater system (AFWS) flow to a hot, dry steam generator.
In addition, it was requested that we provide the figures of the steam line break analysis performed to evaluate the impact of automatically initiated AFWS flow on this event. Accordingly, the above described information is provided as to this letter.
It should be noted that the determination of Enclosure 1 that the existing station procedures are adequate to resolve the concern regarding re-initiation of AFWS flow to a hot, dry steam generator, is dependent on the Westinghouse review of our procedures which is scheduled to be completed by February 12, 1982. If as a result of this ongoing review it is determined that additional steps need to be taken to resolve the concern, the NRC will be advised.
If you have any questions or desire additional information regarding this subject, please contact me.
Very truly yours, I
Enclosure E8202040250 820202 PDR ADOCK 05000206 P
ENCLOSURE 1 RESPONSES TO NRC QUESTIONS ON AUXILIARY FEEDWATER SYSTEM SAN ONOFRE UNIT 1, Item 1 What are the consequences of injecting cold auxiliary feedwater into a hot, dry steam generator?
Response
The analysis of a major feedwater line rupture is discussed in detail in Reference (1).
A feedline break results in the depressurization and blowdown of all three steam generators.
The Auxiliary Feedwater System (AFWS) is automatically actuated based on a low level signal from 2/3 steam generators to provide an assured source of feedwater to the steam generators for decay heat removal.
The AFWS consists of a motor-driven pump and a turbine-driven pump, feeding redundant headers which deliver flow to all three steam generators through normally open flow control valves.
Due to the depressurization of all three steam generators, the motor-driven AFW pump will trip on low discharge pressure, which provides pump runout protection. Operator action is assumed in accordance with current emergency operating procedures to isolate AFW to the faulted steam generator, restart the motor-driven AFW pump, and throttle the FCV's to deliver 250-300 gpm total flow to the two intact steam generators. The turbine-driven AFW pump is assumed to be unavailable due to loss of steam supply for secondary break events. The AFW flow is maintained until steam generator level is restored to the narrow range and then regulated to maintain an indicated narrow range water level.
The effect of injecting cold AFW into a hot, dry steam generator has been discussed with Westinghouse. Following TMI, Westinghouse developed preliminary recommendations for initiating and restoring auxiliary feedwater flow to a hot, dry steam generator under emergency and non-emergency conditions to minimize the thermal stress impacts to steam generator parts.
The Westinghouse recommendation for initiating AFW flow to a hot, dry steam generator under emergency conditions is to initiate AFW flow as quickly as possible without regard for thermal stress resulting from the injection of cold feedwater.
In an emergency situation, the first concern is to provide an adequate heat sink and return the plant to a stable condition. In such an event, the large thermal transient imposed on the dry steam generator by this operation could result in some permanent degradation (e.g., fatigue usage, flaw extension) to the steam generator. However, Westinghouse's best estimate judgment is that injection of cold AFW into a hot, dry steam generator can occur at a flow rate not exceeding 800 gpm without brittle fracture or ductile fatigue rupture of the primary pressure boundary or the secondary shell.
Limiting the AFW flow to any one steam generator to not more than 200 gpm will substantially decrease the probability of any steam generator damage by
-2 minimizing the thermal transient experienced when cold AFW is injected into a hot, dry steam generator. Tube leakage increases are possible, however, as a result of existing tube cracks being extended by this feedwater injection.
This is more prohable for units which have experienced the denting phenomena.
Following the occurrence of cold AFW injection into a hot, dry steam generator, the plant would not be returned to normal operation until the steam generator had been inspected.
As indicated above, the AFW flow for San Onofre Unit 1 for secondary side break events would be established at 250-300 gpm total flow to not less than two steam generators, or less than 150 gpm per steam generator. This AFW flow rate is below the Westinghouse recommendation guideline of 200 gpm per steam generator, which minimizes the probability of steam generator damage. Hence, the AFW system design and guidelines in current emergency operating procedure S01-1.2-1 (enclosed) are adequate to minimize the likelihood of thermal shock to the steam generator.
Item 2 What is the effect of the cooldown of the primary system on reactor vessel integrity?
Response
Pressurized thermal shock to the San Onofre Unit 1 reactor pressure vessel due to loss of secondary cooling events has been addressed in References (2) and (3).
The results of the referenced analysis indicated that reactor vessel integrity will be maintained for the San Onofre Unit 1 vessel throughout its design lifetime.
Item 3 Provide the plots for the core response analysis during a main steamlime rupture with automatic initiation of the AFWS described in Reference (4).
Response
The requested plots are attached.
References (1) Letter from K. P. Baskin to D. M. Crutchfield dated March 6, 1981.
(2) Letter OG-66 from 0. D. Kingsley (Chairman Westinghouse Owners Group) to Harold R. Denton, dated December 30, 1981, transmitting WCAP-10019.
(3) Letter from K. P. Baskin to D. M. Crutchfield dated January 25, 1981.
(4) Letter from H. L. Ottoson to R. H. Engelken dated August 4, 1980.
CORE RESPONSE DURING STEAMLINE RUPTURE TRANSIENT SAN ONOFRE UNIT 1
CORE AVERAGE TEMPERATURE (DEC-F)
RCS PRESSURE (PSIA)
'<i
(
(A 0A (A
01 C) 0u A,
A J
0 no~
0-0D 0D 0D 0 CD 0D CDA C) Lfl C)0 C)
LA
- 0) 0D 0o CC)
C))
0 0
0 0
m :C C
D c
cD D
CD C
CD C
0 0.0 D
CD (D0 c-5 50.000 D
c
-a.
CD
(
(D 100.00 125.00-
3.0000 2.0000 8
1.0000
. 0.0
<t -1.0000
-2.0000 X
3.0000 I
oX 2.5000 cc CD 2.0000 35 2
00 1.5oo 1.0000 La
)
.50000 0:a o
0 a
0 o
TIME (SEC)
FIGURE 2 Base Case No Main or Auxiliary Feedwater Addition Hypothetical Steamline Break
2200.-0IIIII 2000.0 1750.0
- ~1500.0 w
1250.0 1000.00 cc.
La 00 O.00 I
at 550.00 QC 500.00 450.00 U6 400.00 LL 350.00 300.00 TIME (SEC)
FIGURE 3 Main and Auxiliary Feedwater Addition Hypothetical Steamline Break
3.0000 2.0000 1.0000
-0 t -1.0000 cc
-2.0000V4I 3.0000 o 1 2.5000 CD 2-0000 a0 25000 1.5ooo 1.0000
,- ~
.50000 0D 0.0 C
0 oD 0
0 TIME (SEC)
FIGURE 4 Main and Auxiliary Feedwater Addition Hypothetical Steamline Break
3.0000 1
I I
I I
I I
2.0000 e
1.0000 0.0
< -1.0000
-2.0000 I
I I
I I
I I c ~
3.0000I I
I I
I I
I I
I X
2.5000 4
CZ aD 2.0000 p
1.5000 1.0000 C.50000 a
0.0 0D 0
0 0
0 CD 0D 0D 0D 0D CD 0
0C0 0
oj m
~
4r U,
c a) 0 TIME (SEC)
FIGURE 5 Base Case No Main or Auxiliary Feedwater Addition Credible Steam Tne Break
2200.0 2000.0 z;;
1800.0
(
1600.0 1400.0 1200.0 1000.00I I
I I
I I
I I
600.00 I
575.00 550.00 525.00 Lj IL 4
500.00 475.00
'450.00 425.00 C
40 0.00 oo C
A*
0W 0
0 0M 0
00 TIME (SEC)
FIGURE 6 Base Case No Main or Auxiliary Feedwater Addition Credible Steamline Break
0 3.0000 z
2.0000 1.0000 0.0
< -1.0000 L"J
-2.0000 I
I I
3.0000 I
I c3 22.5000 4 z a
2.0000 CD X
Q CD 1.5000 1.0000 D.50000
<D 0.0 CD C>
0D 0C)0 TIME (SEC)
FIGURE 7 Main and Auxiliary Feedwater Addition Credible Steamline Break
2200.0 I
I.
I I
I II I
2 2000.0 1800.0 8
1800.0 0
1400.0 0:
1200 :
1000.00 800.00I I
I 5
5 50.00 L
L 00.00 a
SJL" 3p50-00 a300.00 o
0 a
0 0
0 a
0 a
on 0
00 0
07 0+
0
+:
0 0
TIME (SEC)
FIGURE 8 Main and Auxiliary Feedwater Addition Credible Steamline Break