ML20135H702
| ML20135H702 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, Zion, 05000000 |
| Issue date: | 09/18/1985 |
| From: | George Alexander COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.3.05, TASK-TM 0654K, 654K, GL-85-12, NUDOCS 8509240144 | |
| Download: ML20135H702 (7) | |
Text
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) one First Nitionit Plaza. Chicago, Illinois Commonwealth Edison kT J Address Reply to: Post Othee Box 767 Chicago, lltinois 60690 j
September 18, 1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC' 20555
Subject:
Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 i
Zion Station Units 1 and 2 Response to Generic Letter 85-12 Implementation o f TMI _ Action Item II.K.3.5 Automatic ~ Trip of Reactor Coolant Pumps NRC Docket Nos.-50-295/304, 50-454/455 and 50-456/457 Reference (a):
Generic Letter 85-12 HL Thompson to all OLS and CPS with Westinghouse Nuclear Steam Supply Systems (b):
GL Alexander to HR Denton letter dated 08-22-85
Dear Mr. Denton:
Reference (a)-requested that applicants respond to questions in Section IV of the enclosed Safety Evaluation on reactor coolant pump trip or establish a. schedule for responding to the questions.
Reference (b) contained our initial response to Generic Letter i
85-12.
The purpose of this letter is to provide our response to question I VA.3 and to correct a typographical. error on the previous submittal.
An amended table 2 for Braidwood Station is included; the changes involve procedures whose prefix was designated Bw0A.
The correct designation should be BwCA and is shown on the amended page.
This letter completes our response for Byron and Braidwoo.d Stations.
Additional information+will be submitted for Zion Station according to the schedule in reference (b).
i i
k
. Please contact this of fice if you or your staf f have any further questions regarding Generic Letter 85-12.
Forty copies of this letter with attachments will be forwarded under separate cover.
Respectfully, G
Greg Alexander Nuclear Licensing Administrator
- atts, cc:
US NRC, Document Control Desk Washington DC 20555 JG Keppler - RIII RIII Inspectors - BY,8W and Z 0654K
BRAIDWOOD STATION UNITS -1 and 2 BYRON STATION-UNITS 1-and-2 ATTACHMENT 1 IV IMPLEMENTATION A.
Determination of RCP Trip Criteria.
3.
The LOFTRAN computer code was used to perform the alternate RCP trip criteria analyses.
Both Steam Generator Tube Rupture (SGTR) and non-LOCA event were simulated in these analyses.
Results from the SGTR analyses were used to obtain all but three of the trip parameters.
LOFTRAN is a Westinghouse licensed code used for FSAR SGTR and non-LOCA analyses.
The code has been validated against the January 1982 SGTR event at the Ginna plant.
The results of this validation show that LOFTRAN can accurately predict RCS pressure, RCS temperatures and secondary pressures especially in the first ten minutes of the transient.
This is the critical time period when minimum pressure and subcooling is determined.
The major causes of uncertainties and conservatism in the computer program results,(assuming no changes in the initial plant conditions i.e.
full power, pressurizer lever, all SI and AFW pumps run) are due to either models or inputs to LOFTRAN.
The followings are considered to have the most impact on the determination of the RCP trip criteria:
1.
Break flow 2.
SI flow 3.
Decay heat 4.
Auxiliary.feedwater flow The following sections provide an evaluation of-the uncertainties associated with each of these items.
To i
conservatively simulate a double ended tube rupture in safety analyses, the break flow.model used-in LOFTRAN includes substantial amount of conservatism (i.e.
predicts. higher break flow than actually expected).
Westlnghouse has performed analyses and developed a more -realistic break flow model that has been ' validated against the Ginna SGTR tube-rupture data.
The break flow model used in the WOG analyses has been shown to be. approximately 30% conservative when the e ffect.of the higher predicted break flow'is compared to the more realistic model. 'The consequence of the higher predlcted break flow is a lower than expected predicted minimum pressure.
w
i The SI flow inputs used was derived from best estimate calculations, assuming all SI trains operating.
An-evaluation of the calculational methodology shows that these inputs have a maximum uncertainty of 110%.
The decay heat model used in the WOG analyses _was based on the 1971 ANS 5.1 standard.
When compared with the more recent 1979 ANS 5.1 decay heat inputs, the values used in the WOG analyses is higher by about 5%.
To determine the effect of the uncertainty due to the decay heat-model, a sensitivity study was conducted for SGTR.
The results of this study show that a 20%
decrease in decay heat resulted in only a 1% decrease in RCS. pressure for the first 10 minutes of the transient.
Since RCS temperature is controlled by the steam dump, it is not affected by the decay heat model uncertainty.
The AFW flow rate input used in the WOG analyses are best estimate values, assuming that all auxiliary feed pumps are running, minimum pump start delay, and no throttling.
To evaluate the uncertainties with AFW flow rate, a sensitivity study was_ performed.
Results from the two loop plant study show that, a 64% increase in AFW flow resulted in only an 8% decrease in minimum RCS pressure, a 3% decrease in minimum RCS subcooling, and an 8% decrease in minimum pressure differential.
Results from the 3 loop plant study show that, a 27%
increase in AFW flow resulted in only a 3% decrease in minimum RCS pressure, a 2% decrease in minimum.RCS subcooling, and a 2% decrease in pressure. differential.
The effects of all these uncertainties with the models and input parameters were evaluated and it was concluded that the contributions from the break flow conservatism and the SI uncertainty dominate.
The calculated overall uncertainty in the.WOG analyses as a result of these. considerations for the Byron and Braidwood units is - 150 psig to +150 psig for the RCS pressure RCP trip ~setpoint.
Due to the minimal effects from the decay-heat model and AFW input, these results include only the effects of the uncertainties due to the break flow model and SI flow inputs.
'0654K w-
Braidwood-Station Table 2 Summary of-RCP Trip and RCP Restart -Steps
- in 'the ERG ERG.
RCP-Trip RCP-Restart 1BwEP
.T 1BwEP ES-0.1 R
33wEP ES-0.2 R
10wEP ES-0.3 R
.lVwEP ES-0.4 R
18wEP 1 T
1BwEP ES-1.1 R
1BWEP ES-1.2 TA,TI R
1BwEP 3 T,TA R
18wEP ES-3.1 TI 18wEP ES-3.2 TI 1BwEP ES-3.3 TI 18wCA 1.1 TI 18wCA.2.1 T
R 18wCA_3.1 TA,TI R
18wCA 3.2 TA,TI R
'lBwCA 3.3 TI R
18wFR C.1 ST SR 18wFR C.2 ST 1BwFR H.1 ST-18wFR-P.1
'R 18wFR I.3 R
- T.-
RCP trip criteria' steps
~
i TA Trip-all but'one RCP steps TI - No.-1 Seal RCP-trip steps ST - Other special trip steps R -- Restart criteria _and support conditions required.
SR Special restart.without support conditions required 4
iO654K.
ZION STATION UNITS 1 and 2 ATTACHMENT 2 IV IMPLEMENTATION A.
Determination of RCP Trip Criteria 3.
The LOFTRAN computer code was used to perform the a'1 ternate RCP trip criteria analyses.
Both Steam Generator Tube Rupture (SGTR) and non-LOCA event were simulated in these analyses.
Results from the SGTR analyses were used to obtain all but three of the trip parameters.
LOFTRAN is a Westinghouse licensed code used for FSAR SGTR and non-LOCA analyses.
The code has been validated against the January 1982 SGTR event at the Ginna plant.
The results of this validation show that LOFTRAN can accurately predict RCS pressure, RCS temperatures and secondary pressures especially in the first ten minutes of the. transient.
This is the critical time period when minimum pressure and subcooling is determined.
The major causes of uncertainties and conservatism in the computer program results, assuming no changes in the initial plant conditions (i.e. full power, pressurizer lever, all SI and AFW pumps run) are due to either models or inputs to LOFTRAN.
The followings are considered to have the most impact on the determination of the RCP trip criteria:
1.
Break flow 2.
SI flow 3.
Decay heat 4.
Auxiliary feedwater flow The following sections provide an evaluation of the uncertainties associated with each of these items.
To conservatively simulate a double ended tube rupture in safety analyses, the break flow model used in LOFTRAN includes substantial amount of conservatism (i.e.
predicts higher break flow than actually expected).
Westinghouse has performed analyses and developed a more realistic break flow model that has been validated against the Ginna SGTR tube rupture data.
The break flow model used in the WOG_ analyses has been shown to be approximately 30% conservative when the effect of the higher predicted break flow is compared to the more realistic model.
The consequence of the higher predicted break flow is a lower than expected predicted minimum pressure.
The SI ilow inputs used was derived from best estimate calculations, assuming all SI trains operating.
An evaluation of the calculational methodology shows that these inputs have a maximum uncertainty of 110%.
The decay heat model used in the WOG analyses was based on the 1971 ANS 5.1 standard.
When compared with the more.recent 1979 ANS 5.1 decay heat inputs, the values used in the.WOG analyses is higher by about 5%.
To determine the ef fect of the uncertainty 'due to the decay heat model, a sensitivity study was conducted for SGTR.
The results of this study show that a 20%
decrease in decay heat resulted in only a 1% decrease in RCS pressure for the first 10 minutes of the transient.
Since RCS temperature is controlled by the steam dump, it is not affected by the decay heat model uncertainty.
The AFW flow rate input used in the WOG analyses are best estimate values, assuming that.all auxiliary feed pumps are running, minimum pump start delay, and no throttling.
To evaluate the uncertainties with AFW flow rate, a sensitivity study was performed.
Results l
from the two loop plant. study show that, a 64% increase in AFW flow resulted-in only an 8% decrease in minimum RCS pressure, a 3% decrease.in minimum RCS subcooling, and an 8% decrease in minimum pressure differential.
Results from'the 3 loop plant study show that, a 27%
increase in AFW flow resulted in only a 3% decrease in minimum RCS pressure, a 2% decrease in minimum RCS subcooling, and a 2%. decrease in pressure. differential.
.The effects of all these uncertainties with the models and input parameters were evaluated and it was concluded that the contributions from the break flow conservatism and the SI uncertainty dominate.
The calculated overall uncertainty in the'WOG analyses as a result of these considerations for the Zion unit is
+30 to.+200 psig for the RCS pressure RCP trip setpoint.
Due to the minimal effects from the decay
~ heat model and AFW input, these results. include only the effects of the uncertainties oue to the break flow model and SI flow inputs.
0654K'