ML20236P100
| ML20236P100 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/08/1998 |
| From: | Salas P TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-327-98-01, 50-327-98-1, 50-328-98-01, 50-328-98-1, NUDOCS 9807160152 | |
| Download: ML20236P100 (6) | |
Text
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Tennessee Valley Authority, Post Office Box 2000, Soddy-Daisy, Tennessee 37379-2000 July 8, 1998 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20555 Gentlemen:
In the Matter of
)
Docket Nos. 50-327 Tennessee Valley Authority
)
50-328 SEQUOYAH NUCLEAR PLANT (SQN) - NRC INSPECTION REPORT 50-327, 328/98 RESPONSE TO UNRESOLVED ITEM (URI) 50-327,320/98-01-02 This letter provides our response to the subject URI as requested in the inspection report that was dated March 30, 1998.
The URI concerns Final Safety Analysis Report (FSAR)
-Chapter 15 accident analysis assumption for 10-minute operator action on a main feedwater piping rupture.
The FSAR i
analysis lists as an assumption a 10-minute operator action l
time to isolate Auxiliary Feedwater (AFW) to the affected I
Our. evaluation has concluded that the 10-minute operator action assumption referenced in the Technical Specification (TS) bases was not a requirement and has recommended that the statement be deleted from the TS bases.
The inspectors reviewed this evaluation and concluded i
the applicability of the FSAR assumption was not adequately resolved and that there were no licensing or procedural J :tions taken to assure that the assumption could be met or
-was not required.
Included in this response is our position l
that verification la net required for the de. sign basis assumption of 10-minute operator action.
The 10-minute operator action assumption is contained in the l
FSAR and Westinghouse Electr'.c Corporation analyses of a main l
feedwater piping rupture.
Tne 10-minute operator action gy\\
i.
time, for isolating the faulted steam generator (SG), is one of 18 bounding, conservative-initial assumptions for this
,y accident scenario. These 18 assumptions for the main 9907160152 980708 i
PDft ADOCK 05000327 l-0 PM Penne en seawart enor L
e U.S. Nuclear Regulatory Commission Page 2 July.9, 1998 feedwater piping rupture are described in the enclosure.
For the 10-minute operator action, the computer analysis assumes that no feedwater, via AFW or normal feedwater, is allowed into any of the SGs for 10 minutes.
By not allowing any feedwater flow to any of the SGs, the effects of the event are maximized.
Collectively, all 18 assumptions in this scenario are used to maximize coolant heatup rate and to minimize heat rejection to the secondary side.
This conservative method in modeling assumptions provides additional safety margin for the plant.
SON operators follow symptom-based technical guidance for response to emergency transients.
The emergency operating procedures provide a systematic approach for diagnosis and recovery from a broad spectrum of event sequences.
Operator action, in accordance with the procedures, ensures the timely and appropriate operator action is taken in the correct sequence.
In the event of a large main feedwater piping rupture (for example the double-ended rupture), it would be obvious.to the operators which SG was associated with the piping rupture due to the low-low SG level indication.
When presented with this indication, operators are trained to follow emergency operating procedures.
Relative to the 1995 problem evaluation report (PER) discussed in the URI, the simulator training runs did not l'
include all of the input assumptions of the FSAR Chapter 15 Accident Analysis.
The training scenarios performed were SG No. 3 feedline breaks (one case inside containment and one case outside containment), at 29 percent reactor power with the A-A motor-driven feedwater pump out for maintenance.
The simulated break in the feedwater line was not the double-ended break or maximum break area assumed in the FSAR, but a considerably smaller rupture in the feodwater piping.
Due to this smaller break area and resulting smaller change in the j
level of the SG, it would not be expected that the operators would J ocate and isolate this break as quickly as in the L
maximum break assumption scenario.
With the smaller break area, the SG takes longer to boil dry; therefore, the heatup of the RCS is initially less severe giving the operator more time to act.
The purpose of this scenario was to challenge the operator's diagnostic abilities and not to verify the 10-minute operator action scenario.
For the 10-minute operator 1
i
4 U.S. Nuclear Regulatory Commission Page 3 July 9, 1998 action to be considered meaningful in a training situation, the scenario used must consider all of the required analysis assumptions.
If a training scenario is used that does not set up the initial conditions like the FSAR analysis, then the evaluation of operator or plant performance should not be judged by the current Chapter 15 analysis results.
It is our position that the 10-minute operator action assumption to isolate the AFW system from the faulted SG is not an operator action time that must be validated.
- Rather, the 10-minute operator action time, as required by the NRC Standard Review Plan, is one of several bounding assumptions, which have been conservatively chosen to maximize the RCS heatup rate.
The systematic approach of the emergency procedures ensures that timely and appropriate operator action is taken in the correct sequence.
The discussion provided above is relevant to the main feedwater line rupture event.
In the broader perspective, we are cognizant and actively participating in the ongoing industry /NRC issues relative to required operator action times assumed in the accident analysis and actual times documented in operator training scenarios.
We have participated in the Westinghouse Owners Grov- (WOG) Operator Response Time Program.
We are evaluating, f'r applicability, the resulting WOG suggested methods for improving or reducing operator action times to ensure that, for those accidents which operator action is required, the time to respond is minimized.
Please direct questions concerning this issue to mc at (423) 843-7170 or J.
D. Smith at (423) 843-6672.
Sincerely, f
2 f
alas Lihensing and Industry Affairs Manager Enclosure cc:
See page 4
U.S. Nuclear Regulatory Commission Page 4 July 9, 1998 a
cc (Enclosures):
Mr. R. W.
Hernan, Project Manager Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 NRC Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road 4
Soddy-Daisy, Tennessee 37379-3624 Regional Administrator l
U.S. Nuclear Regulatory Commission Region II Atlanta Federal Center i
61 Forsyth St.,
SW, Suite 23T85 Atlanta, Georgia 30303-3415 l
P I
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i ENCLOSURE TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLP R PLANT (SQN)
UNITS 1 AND 2 NRC INSPECTION REPORT 50-327, 328/98 1 RESPONSE TO UNRESOLVED ITEM (URI) 50-327,328/98-01-02 1.
The plant is initially operating at 102 percent of the engineered safeguards design rating.
2.
Initial reactor coolant average temperature is 5.5 degrees Fahrenheit above nominal value, and the initial pressurizer pressure is 30 psi above its nominal value.
1 3.
No credit is taken for pressurizer spray.
l 4.
No credit is taken for the high pressurizer pressure reactor trip.
Note:
This assumption is made for calculational l
convenience.
Pressurizer power-operated relief valves and spray could act to delay the high pressure trip.
Assumptions 3 and 4 permit i
evaluation of one hypothetical, limiting case rather than two possible cases:
one with a high pressure trip and no pressure controls and one with a pressure control but no high pressure trip.
5.
Main feed to all SGs is assumed to stop at the time the break occurs.
6.
Saturated liquid discharge (no steam) is assumed from the affected SG through the feedline rupture.
This assumption minimizes energy removal from the nuclear steam supply system during blowdown.
7.
No credit is taken for the low-low water level trip on the affected SG until the SG level reaches O percent of narrow range span and after the expiration of any applicable delays in the TTD System.
8.
The worst possible break area is assumed, i.e.,
one l
that ensures the initial reactor coolant system
t (RCS) depressurization is maximized.
This assumption minimizes subcooling margin during the post-trip reactor coolant heatup period.
9.
No credit is taken for heat energy deposited in RCS metal during RCS heatup.
10.
No credit is taken for charging or letdown.
11.
Loss of offsite electrical power (LOEP) is assumed after the reactor trip and reactor coolant flow decreases to natural circulation.
12.
SG heat transfer area is assumed to decrease as the shells.'de liquid inventory decreases.
13.
Conservative core residual heat generation is assumed based upon long-term operation at initial power level (102 percent) preceding the trip.
14.
The AFW is actuated by the low-low SG water level signal.
Assuming a feedline rupture to Loop 4, the following assumptions were made:
SGs 1 and 2 receive no AFW from the A-A motor driven auxiliary feedwater pump (MDAFWP)
(single failure).
(15)
SGs 3 and 4 receive no AFW from the B-B e
MDAFWP (all flow out break).
(16)
Turbine driven auxiliary feedwater pump e
(TDAFWP) flow spills out of the SG 4 break, and thereby all unfaulted SGs are starved.
(17)
Operator action isolates the AFW line e
from the TDAFWP and B-B MDAFWP.
Operator action is taken credit for isolating the AFW system from the faulted SG 10 minutes after the low-low SG water level signal.
(18)
A 60-second delay was assumed following the low-low SG level signal to allow time for startup of the emergency diesel generators and the AFW pumps.
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