ML17298A472
| ML17298A472 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 07/25/1983 |
| From: | Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| ANPP-27402-WFQ, NUDOCS 8308010359 | |
| Download: ML17298A472 (12) | |
Text
REGULATORIINFORNATION OISTRISUTION +TEN (RIOR)
ACCESSION NBR 8308010359 OOC.DATE: 83/07/25 NOTARIZED; YES DOCKET FACIL:STN-50 528 Palo Verde Nuclear Stationi Unit 1N Arizona Publj 05000528 STN-50-529 Palo Verde Nuclear Statjoni Unit 2i Arjzona Publj 05000529
'STN 50-530 Palo Verde Nuclear Stationi Unit 3N Arizona Publi 05000530 AUTH NAME
-AUTHOR AFFILIATION.
VAN BRUNTNE.E.
AI izona Public Set vice Co ~
REGIP ~ NAME RECIPIENT AFFILIATION KNIGHTONNG ~
Licensina
-Br anch 3
SUBJECT:
Forwards r esoonse to 830617 reauest for addi infq >re FSAR Section 7.A.O concer ning mul title contr ol sys failure e.
DISTRIBU/ION CODE:
B001S COPIES RECEIVED:LTR ENCL SIZE:
-TITLE: Licensina Submittal:
PSAR/FSAR Amdts 8 Related Conrespondence NOTES:Stander dj zed nl ant.
.Standardized slants
- Standardized slants 05000528 05000529 05000530 RECIPIENT IO CODE/NAME NRR/DL/ADL NRR LB3 LA INTERNAL: ELD/HDS3 IE/DEPER/EPB 36 IE/DEQA/QAB 21 NRR/DE/CEB 11 NRR/DE/EQB 13 NRR/DE/t<EB 18 NRR/DE/SAB 2A NRR/DHFS/HFEB40 NRR/DHFS/PSRB NRR/DSI/AEB 26 NRR/DSI/CPB 10 NRR/DSI/ICSB 16 NRR/DSI/PSB 19 NRR/OSI/RSB 23 RGN5 EXTERNAL: 'AORS 41 OMB/DSS (AHDTS)
LPDR 03 NSIC 05 COPIES LTTR ENCL 1
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2 1
1 1
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01 IE FILE IE/DEPER/IRB 35 iVRR/DE/AEAB NRR/DE/EHEB NRR/OE/GB 28 NRR/OE/MTEB 17 NRR/DE/SGEB 25 NRR/DHFS/LQB '32 NRR/OL/SSPB NRR/OS I/ASB NRR/DS I/CSB 09 NRR/OSI/HETB 12 NRR RAB 22 EG FI E
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FEHA<<REP DIV 39 NRC 'POR 02 NTIS COPIES LTTR ENCL 1
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Arizona Public Service Company P.O. BOX 21666
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PHOENIX, ARIZONA 65036 July 25, 1983 ANPP-27402 WFQ/MSN Director of Nuclear Reactor Regulation Attention:
Mr. George Knighton, Chief Licensing Branch No.
3 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Units 1, 2 and 3
Docket Nos. STN-50-528/529/530 File:
83-056-026; G.l.01.10
Reference:
NRC letter from G.
W. Knighton to E. E. Van Brunt, Jr.,
APS, dated June 17, 1983.
Subject:
Request for Additional Information Palo Verde, Units 1, 2 and 3.
Dear Mr. Knighton:
Final Safety Analysis Report (FSAR), Section 7.A.4, provided our response to Question 222.04 concerning multiple control system failures.
The referenced letter requested additional information.
Attached is the additional quantitative analysis requested.
If you have any further questions, please contact me.
Very truly yours, E. E.
Van Brunt, Jr.
APS Vice President Nuclear Projects Management ANPP Project Director EEVB/MSN/sp Attachment cc:
E. A. Licitra A. C. Gehr T.
G.
Woods (w/a) 8308010359 830725 PDR ADOCK 05000528 A
July 25, 1983 ANPP-27402 - WFQ/HSN STATE OF ARIZONA
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) ss.
COUNTY OF MARICOPA)
I, Edwin E. Van Brunt, Jr., represent that I am Vice President Nuclear Projects of Arizona Public Service Company, that the foregoing document has been signed by me on behalf of Arizona Public Service Company with full authority so to do, that I have read such document and know its contents, and that to the best of my knowledge and belief, the statements made therein are true.
Edwin E. Van Brunt, Jr.
Sworn to before me this~F4 day of
'-'t'2983-My Commission expires:
Mv Commission Expires April 6, 1987 Nota'"y "ubl-'c'-
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Frost M. Allen R. Bynum H. Wilson Shepherd Soteropoulos Ferguson Barnoski Mancuso W. Gross (w/a)
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ATTACHMENT PALO VERDE NUCLEAR GENERATING STATION MULTIPLE CONTROL SYSTEM FAILURES REQUEST FOR ADDITIONAL INFORMATION QUESTION f
By letter dated December 30,
- 1982, you provided information concerning multiple control system failures due to common power
- sources, common sensor or common instrument tap failures.
With one exception, these transient scenarios are bounded by the Chapter 15 analyses included in the Palo Verde FSAR.
For the exception, i.e.,
of failure of the E-NNN-Dll panel, you state that although this scenario is not bounded by the loss-of-feedwater event analyzed in Chapter 15 '(since the event did not include consideration of letdown isolation), the rate of RCS inventory addition is small and will not appreciably affect the peak RCS pressure or fuel performance aspects of this event.
In addition, you state that sufficient time exists for the operator to take action to prevent the pressurizer from filling.
We request that you provide a more quantitative analysis to support your conclusions relating to the failure of the E-NNN"Dll panel.
RESPONSE
The failure of distribution panel E-NNN-Dll will initiate a decrease in feedwater flow at the time the distribution panel loses power.
In
- addition, the PLCS will reduce letdown flow to 0
- gpm, and initiate charging flow from all three charging
- pumps, resulting in a
net mass addition to the primary system.
The SBCS and RPCS~will be unable to automatically respond to any challenges.
This event scenario is bound by the loss of feedwater flow event qualitatively presented in Section 15.2.7 of the CESSAR FSAR with respect to peak RCS pressure and fuel performance.
The loss of feedwater flow event is bound by the loss of condenser vacuum event which results in a rapid reactor trip.
Following the reactor trip the pressurizer level decreases to approximately its initial value.
The time dependent level increase caused by all three charging pumps turning on and remaining on and the letdown line isolating is calculated and added to the intitial volume to determine the time by which the 'operator must act to prevent the safety valve inlet nozzle from being covered.
In the same
- manner, the pre-trip pressurizer level increase due to the panel failure is calculated and added to the maximum transient pressurizer level.
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The calculation of the assumptions:
pressurizer'evel swell used the following 1)
The maximum initial pressurizer liquid level is 60%
2)
Each charging pump operates at 44 gpm 3)
The letdown is completely isolated due to the panel failure 4)
The charging flow is heated to RCS temperatures Using this data it can be calculated that the operator has at least 20 minutes to take action before the safety valve nozzles are submerged.
This is the same operator action response time used to set the initial pressurizer level technical specifications.
This calculation is conservative in that the RCS temperatures, and hence pressurizer
- level, are assumed to attain post trip values approximately equal to their initial values as predicted by the LOCV event.
The panel failure event will not result in the reactor coolant pump coastdown assumed in the LOCV event and therefore, will experience better heat transfer and a
lower post-trip pressurizer liquid level.
The reduced post-trip pressurizer level will result in additional margin for operator action.
The mass addition prior to reactor trip (approximately one minute) does not significantly increase the maximum pressurizer level (less than 30 ft. )
and will not affect maximum RCS pressure.
CESSAR PSAR figure 15B-5 illustrates this insensitivity of maximum RCS pressure to pre-trip RCS mass addition (analogous to an increase in the initial pressurizer water volume).
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