ML20217Q428
| ML20217Q428 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 08/19/1997 |
| From: | Hosmer J COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9709030079 | |
| Download: ML20217Q428 (5) | |
Text
_ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
o.mmonwuio, i uon < omvan3 1 -saa Opm l'im thewm ts t nne,11 Mi% vol August 19,1997 U. S. Nuclear Regulatory Commlssion Washington, D. C. 20555 Attention:
Document Control Desk
Subject:
Response to Request for Additional Information Regarding the Revised Steam Generator Tube Rupture Analysis Byron Nuclear Power Station Facility Operating License NPF-37 and NPF-66 NRC Docket Numbers: 50-454 and 50-455 Braidwood Nuclear Power Station Facility Operating License NPF-72 and NPF-77 NRC Docket Numbers: 50-456 and 50-457
References:
1.
J. Hosmer Letter to USNRC, " Steam Generator Tube Rupture Analysis for Byron and Braidwood Generating Stations", dated November 13,1996.
2.
J. Hosmer Letter to USNRC, Response to Request for Additional Information Regarding the Revised Steam Generator Tube Rupture Analysis - Byron and Braidwood Stations, dated March 20,1997.
3.
J. Hosmer Letter to USNRC, Response to Request for
[],(
Additional Information Regarding the Revised Steam Generator Tube Rupture Analysis - Byron and Braidwood Stations, dated June 24,1997.
4.
USNRC Letter to Comed,- Request for Additional Information i
Regarding the Revised Steam Generator Tube Rupture I
I Analysis - Byron and Braldwood Stations, dated July 18, t
199'.
On November 13,1996, Commonwealth Edison Company (Comed) submitted its revised steam generator tube rupture analysis for Byron and Braidwood Stations (Reference 1). Comed provided responses to requests for additional information (RAI) on March 20 and June 24,1997. (References 2 and 3), Another request for additional information was transmitted on July 18,1997 (Reference 4). This document is Comed's response to the July,1997 request.
9709030079 970st9 PDR ADOCK 0500o454 h
,v k nla byrbwd sgttral3 doc:1
'iV d
_________A
e USNRC 2-August 19,1997 Please direct any questions to this office.
,dn 8/lrurm John B. Hosmer Vice President Engineering Attachment cc:
A. B. Beach - Regional Administrator, Rlli G. Dick, Jr. - Project Manager, NRR S. Burgess - Senior Resident inspector, Byron C. Phiilips - Senior Res.9nt inspector, Braidwood Office of Nuclear Safety, IDNS P
i
RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION REGARDING THE REVISED STEAM GENERATOR TUBE RUPTURE ANALYSIS BYRON STATION AND BRAIDWOOD STATION (DATED 7/18/97)
NRC Question #1:
The response to question B.3 in the licensee's March 20,1997, st.bmittal provided the sensitivity of operator responso times that exceeded the analysis response times. Please complete a similar analysis for Crew No.1. The analysis should include: (a) a case for the overall response time and other cases where Crew No.1 exceeded the analysis response times; and (b) a composite analysis of cases that exceeded the analysis response times.
Response #1:
The results of a sensitivity study which evaluates the significance of observed response times for Crew No.1 that exceed the revised analysis response times for the margin to overfill (MTO) case are included as Table 1. The replacement steam generator case is used as the base case since it results in the limiting MTO. These cases assume the maximum analysis response times for all segments of the evolution plus the change in response times for the particular segments being evaluated. The resulting response times exceed the totals assumed in the SGTR analysis and result in an apparent reduction in the MTO.
This reduction can be sufficient to cause an. pparent overfill condition as shown in Case 3 below.
Case 1 evaluates the impact of decreasing operator response time to isolate auxiliary feedwater (AFW) by 262 seconds. The impact is an increase in MTO of 731 ft'. This case is included to show the significance of AFW isolation.
Case 2 evaluates the impact of increasing operator response time to initiate RCS depressurization by 15 seconds. The impact is a reduction in MTO of 11 ft*.
Case 3 evaluates the impact of increasing operator response time to terminate ECCS flow by 86 seconds. This case resulted in more than 60 ft' reduction in MTO and the ruptured steam generator becomes water solid (overfill).
Case 4 evaluates the impact of increasing operator response time to establish letdown by 25 seconds. This change has an insignificant impact on MTO.
Case 5 evaluates the impact of increasing operator response time to reopen pressurizer PORV by 43 seconds. The impact is a reduction in MTG af 30 ft'.-
=_
'+
oe Case 6 incorporates all the changes in operator response times from cases 1 to
- 5 and the results show an increase in MTO of 549 ft'. A composite analysis of only cases that exceeded the analysis response times (Cases 2 through 5) would result in an app;went overfill scenario.
Case 7 uses all of the actual Crew No.1 response times and shows an increase
-in MTO of 603 ft' from the case using the analysis times.
In conclusion, Crew No.1 is capable of successfully terminating break flow without filling the steam generator.
Table 1 : Sensitivity of Operator Response Time, Exceeding Analysis Response Times for Crew #1 Case Reference 1 Analysis Crew 1 Operator impact on Response Time Response Time Margin to Overfill (ft').
1 Isolate AFW to ruptured Isolate AFW to ruptured steam generator steam generator
+ 731 660 seconds 398 seconds 2
- depressurization depressurization
-11 120 seconds 135 seconds 3
. Terminate ECCS flow Terminate ECCS flow (Overfill) 120 seconds 206 seconds 4
Establish Letdown Establish Letdown -
0 180 seconds 205 seconds -
5 Reopen pressurizer
. Reopen pressurizer PORV PORV
- 30 240 seconds 283 seconds 6.
- Composite cases 1 to 5 Composite cases 1 to 5
+549-7 Analysis Time Overall response time
+ 603
=
=
_L_u__ _ _.. -
- s' NRC Question #2
The response to question B.4 in the licensee's March 20,1997, submittal provided a table (B.4) of observed operator times for vanous steps of the steam generator response effort. Please provide the margin to overfill for each cell in the table.
Response #2:
The requested data are provided in Table 2 (numbers in parentheses). The RSG case is used as the base case since it results in the limiting MTO. All four crews isolated AFW earlier than the assumed analysis time. The actual crew performances resulted in MTO much greater than tne analysis prediction.
Table 2 : Observed Operator Response Times and Associated MTO Response / Event Crew Crew Crew Crew Averaged Analysis
- 1
- 2
- 3
- 4 Time Time (sec)
(soc)
(sec)
(sec)
(sec)
(sec)
(ft')
(ft')
(ft')
(ft')
(ft')
Isolate AFW to ruptured 398 305 453 565 430 660 steam generator (2140)
(2360)
(2012)
(1751)
(2065)
(1535)
Initiate RCS cooldown 975 1212 1037 1113 1084 1080 (2043)
(2100)
(1860)
(1456)
(1911)
(1163)
Initiate RCS depressurization 135 122 100 37 99 120 (1278)
(1369)
(1264)
(886)
(1270)
(536)
Terminate ECCS flow (excepT 206 77 85 81 112 120 1 Centrifugal Charging Pump)
(1136)
(1301)
(1190)
(810)
(1183)
(444)
Establish charging 96 128 145 115 121 120 (1067)
(1237)
(1112)
(754)
(1112)
(371)
Establish letdown 205 120 193 206 181 180 (899)
(1162)
(980)
(626)
(986)
(243)
Reopen pressurizer PORV 283 150 237 279 237 240 (683)
(1066)
(821)
(450)
(822)
(76)
Total response tir le excluding 1900 1809 1797 1831 1834 1860 isolation of ruptured steam (663)
(1053)
(806)
(435)
(804)
(60) generator
_.. _ - _