ML19330B874
| ML19330B874 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities  |
| Issue date: | 07/18/1980 |
| From: | Janecek R COMMONWEALTH EDISON CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| IEB-80-17, NUDOCS 8008070115 | |
| Download: ML19330B874 (9) | |
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one First Nation:1 Plaza, Chicago, Illinois Commonwealth Edison
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C /e Addrass R; ply to: P:st Off ca Box 767 f-
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Chicago. Illinois 60690 July 18, 1980 Mr. James G. Keppler, Director Directorate of Inspection and Enforcement - Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137
Subject:
Dresden Station Units 2 and 3 Quad Cities Station Units 1 and 2 Additional Response to IE Bulletin 80-17 NRC Docket Nos. 50-237/249 and 50-254/265 References (a):
J. G. Keppler letter to C. Reed dated July 3, 1980 (b):
R. F. Janecek letter to J.
G. Kepplce dated July 14, 1980 (c):
D. L. Peoples letter to J.
G. Keppler dated July ll, 1980
Dear Mr. Keppler:
This letter is to provide an additional response for l
Dresden Units 2 and 3 and Quad Cities 1 sad 2 to IE Bulletin 80-17, transmitted by Reference (a).
Reference (b) provided our original response to Item 7 of the bulletin.
In accordance with discussions with members of your i
Staf f and the Office of Nuclear Reactor Regulation, Attachment 1 to this letter contains an analysis for Dresden 2/3 and Quad Cities 1/2 l
of an MSIV closure event with full ATWS.
Note that this analysis is conservative in that the effects of the Isolation Condensers, RCIC, I
and HPCI systems have not-been included. to this letter provides the results of tests required by Items 2 and 3 of the bulletin for Quad Cities Unit 2.
The results of tests on Quad Cities Unit 1 which are referred to in were reported in Reference (c).
33t 2n 08i 8008 010 \\ \\ T5' l
k Mr. James G. Keppler, Director July 18, 1980 Page 2 Please address any questions concerning this matter to this office.
Very truly yours, Wlcw Robert F. Janecek Nuclear Licensing Administrator Boiling Water Reactors cc:
Director, Division of Reactor Operations Inspection SUBSCRIBED and SWORN to before Me this
/f T//, day of
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1980
{i wJ Notary Public 5364A
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Attachront 1 h'
ADDITIONAL INFORMATION 6
ATWS WITHOUT RPT FOR DRESDEN UNITS 2 AND 3 QUAD CITIES 1 AND 2 Introduction The purpose of this document is to provide additional information in the event of anticipated transients without scram (ATWS) and no recirculation pump trip (RPT).
The evaluation presented herein considers the very conservative case of MSIV closure with complete failure to scram.
The purpose of this analysis is to determine the reactor power level at which the peak vessel pressure reaches the assumed Service Level C limit of 1500 psig.
No credit is taken for operator action to mitigate the severity of the event.
Discussion MSIV Closure - No Scram The HSIV closure with a postulated failure to scram provides the most severe pressure increase.
The sequence of events begins with the closure of the MSIV's in 4 seconds.
With closure of the MSIV's, the pressure
~ immediately begins to rise, resulting in void collapse and a rapid increase in power.
The initial operating conditions and the assumptions used in the analysis are given in Table 1.
Normal reload 1.icensing basis values were employed to perform the analysis except for the'voia coefficient.
The most conservative nominal 3D dynamic void coefficient throughout the remainder of the cycle was used to provide a conservative result.
The analysis was performed at selected power levels so that the power which results in 1500 psig could be determined.
. For the typical case study at 80% power, the setpoint pressure of the relief valves is reached in 5.27 seconds and they open to arrest the pressure rise.
Pressure continues to rise until 48 secondr into the event when it peaks and bcgins to decrease.
The maximum pressure at the vessel bottom is 1465 psig.
The time response of the transient is shown in figure 1.
1 The results of the analyses are presented in Table 2.
Figure.2 provides a plot of peak vessel pressure versus initial power level.
As shown in Figure 2, the peak vessel pressure will be within the assumed Service Level C limit of 1500 psig for all power levels up to 81.5%
of rated power.
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,,e TABLE 1 Transient Input Parameters Power Level (mwt) 2527 6
Rated Core Flow (10 lb7hr) 98.0 6
Rated Steam Flow (10 lb/hr) 9.77 Steam Dome Pressure (psig) 1005 Turbine Bypass capacity (% rated steam flow) 40 Humber of Relief Valves 5
Setpoints (psig) 1125 Capacity (% rated steam flow at setpoint) 27.8 Humber of Safety Valves 8
Setpoint (psig) 1253 Capacity (% rated steam flow at setpoint) 50 Humber of Safety / Relief Valves N/A Setpoint (psig)
Capacity (% rated steam flow at setpoint)
Void Fraction (%)
34.5 Void Coefficient (-C/% Rg) 7.4 Doppler Coefficient (-C/ F) 0.31
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TABLE 2 MSIV Closure Without Scram Results Peak Peak Steamline Vessel Power Pressure Pressure Psig Psig
'84 1520 1562 80 1425 1466 o
70 1301 1341 e
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EEP LLL 5S m
0 CCCK AAA NT MSt EE l
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123 0
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C 0
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rigure 2 Peak Vessel Pressure Versus Initial Reactor Power Level
QUAD-CITIES STATION RESPONSE NRC IE BULLETIN 80-17 Item 2.
Using Temporary Procedure 1358, previously written and used during quad-Cities Unit One testing, data were collected as required on Unit Two.
Oh July 13, 1980, a manual scram at 250 MWt and an automatic scram at 101 MWt were performed to collect data. The automatic scram was done (as on Unit One) by down-ranging the IRMs, and thereby receiving a scram from IRM upscale.
a.
Thirty control rod scram insertion times were obtained utilizing a multi pen recorder.
From this information, an all-rod insert time can be estimated from the slowest of these rods.
TIMES:
3.33 sec Manual Scram 3.37 see Automatic Scram The identical rod was the slowest during both scrams.
The times listed above are comparable to those found during Hot Scram Timing Surveillance.
Computer scans of control rod positions also verified all rods were inserted past the 06 position.
A visual check was also done to verify all-rods had inserted.
b.
Voltage was measured across the scram solenoids while the scram signal was present. -The voltages for all four groups of both channels were found to-be zero. Also, the group lights on the 902-5 panel went out, which is a positive indicat!on of loss of voltage.
c.
An operator was stationed at the backup scram valves during both tests.
The valves operated correctly and air was vented as designed.
d.h.
The filling and venting of the instrument volume was 'onitored by m
attaching a multi pen recorder to detect the magnetrol level switch contact actuation. The following chronology for the manual and automatic scrams is provided to indicate the events as they occurred:
MANUAL AUTOMATIC t = 0 see Reactor Scram t = 0 sec 32 see SDV not drained alarm 35 sec 43 sec SDV High Level Rod Block 47 sec 34 sec SDV High Level Scram 37 see 100 sec Reset scram in Control Room 90 sec 2 min'34 sec SDV drain opened 3 min 39 sec 2 min 58 sec SDV not drained alarm reset 3 min 53 sec 2 min 59 sec SDV High Level Rod Block Re-4 min 11 sec set
MANUAL AUT0HATIC 3 min 08 see SDV High Level Scram Cleared 4 min 01 sec 5 min 11 sec SDV not drained alarm came 5 min 36 se.c up 5 min 49 sec SDV High Level Scram came up 6 min 22 sec 10 min 56 sec SDV High Level Scram Cleared 13 min 41 sec 11 min 04 sec SDV not drained alarm reset 13 min 49 sec 15 min 04 see All alarms cleared 16 min 58 sec The above data appear very similar to the data gathered during the Unit One tests. A discussion of the inconsistency of the data provided in the response to the Unit One testing (ref. letter NJ K-80-238).
A slow-fill test was not performed on Unit Two.
e,f.
Stroke times of the vent and drain valves were obtained between the two scram tests.
These times were as follows:
OPENING CLOSlHG SCRAM CLOSING TIME 2-302-22 Less than I sec 2.. sec 6.4 sec 2-302-21A Less than I sec 3 5 sec 6.5 sec 2-302-21B Less than I sec 3.7 sec 4.8 sec g.
A water sample was taken and analyzed for total suspended solids.
MANUAL SCRAM 10 ppm AUTOMATIC SCRAM Less than 1 ppm I.
As during the Unit One test, the instrument volume did not cool sufficiently between the two scrams to enable use of the procedure to check for water in the scram discharge volume.
During the unit startup, the test was successfully performed.
J.
A scram was not required to determine the scram reset delay times.
An identical procedure was used as previously outlined in the Unit One response.
Channel A Groups 1 & 4 - - - - 16 sec Channel A Groups 2 s 3 - - - - 16.5 sec Channel B Groups 1 & 4 - - - - 15 sec Channel B Groups 2 & 3 - - - - 15 sec k.
All data acquired have been reviewed, and are deemed acceptable.
This data are also comparable to that acquired on July 6,1980 for Unit One.
)
Itcm 3 At the conclusion of each of the two scrams, the vent valves were observed to open. This verification was done af ter the scram was cleared, the water sample taken, and the instrument volume draining was done. The scram dis-charge volumes were verified to be drained during the Unit Two startup.
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