ML18019A396
| ML18019A396 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 09/26/1985 |
| From: | Zimmerman S CAROLINA POWER & LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NLS-85-344, NUDOCS 8510010204 | |
| Download: ML18019A396 (17) | |
Text
V SUBJECT!
Forwaf ds addi infoiper SER Open Item,8 re fire.protectiong including discussion of high/low, pressure interfaceidouble fusing.
8 br eaker coordination testing per Auxiliary Sys r
B anch reviewer questions.
DISTRIBUTION CODEo BOO?D COPIES RECEIVED:LTR "ENCL
'IZE!
TITLEi*Licensing Submittal: Fire Protection NOTES!
REGUL~AlY INFORNATION DISTRIBUTIOQYSTSN (RIBS)
J I ACCESSION NBR!8510010204 DOC ~ DATE: 85/09/26 NOTARIZED NO DOCKET FACIL:50 400 Shearon Har r is Nuclear Power Plant'i Uni t 1i 'ar ol ina 05000400 "AUTH,NAME AUTHOR AFFILIATION
'ZIMMERMAN/SORY Carolina. Power 8 Light: Co, RECIP ~ NAME".
RECIPIENT AFFILIATION DENTON~H.R ~
OfficeI of Nucl ear Reactor Regul ation'ir ector RECIPIENT ID CODE/NAME)
NRR LB3 BC COPIES'TTR. ENCLI 1
RECIPIENT ID CODE/NAME BUCKLEYiB 01 COPIES'TTR KNCLt 1 ~
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TOTAL'UMBER OF COPIES REQUIRED N L'TTR 19 ENCL'7-
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CNK Carolina Power & Light Company gp 8 6,1S85 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO. I DOCKET NO.50-000 RESPONSE TO REQUEST FOR INFORMATION FIRE PROTECTION SERIAL: NLS-85-300
Dear Mr. Denton:
Carolina Power R Light Company hereby submits additional information concerning fire protection at the Shearon Harris Nuclear Power Plant (SHNPP).
The attached information is submitted in response to Safety Evaluation Report (SER) Open Item No. 8 and questions from the Auxiliary Systems Branch reviewer concerning safe and alternate shutdown capability.
Responses to the following five items are attached:
l.
High/Low Pressure Interface (Q010.00) 2.
Double Fusing (QOI0.50) 3.
Breaker Coordination Testing (QOI0.53a) 0.
Auxiliary Control Panel Procedures 5.
Safe Shutdown Methodology If you have any further questions on the subject or require additional information, please contact me.
Yours very truly, 3HE/ccc (l9503DK)
Attachment S. R.
tmmerman Manager Nuclear Licensing Section CC:
Mr. B. C. Buckley (NRC)
Mr. O. Chopra (NRC-PSB)
Mr. G. F. Maxwell (NRC-SHNPP)
Dr. 3. Nelson Grace (NRC-RII)
Mr. N. Wagner (NRC-ASB)
Mr. Travis Payne (KUDZU)
Wake County Public Library 85~Oa>O2O4 S5OOOO40 5O926 l
Agocx 05o >DRI l, E
Mr. H. A. Cole Mr. Daniel F. Read (CHANGE/ELP)
Mr. Wells Eddleman Mr. 3ohn D. Runkle Dr. Richard D. Wilson Mr. G. O. Bright (ASLB)
Dr. 3. H. Carpenter (ASLB)
Mr. 3. L. Kelley (ASLB) 411 Fayetteville Street
~ P. O. Box 1551
~ Raleigh, N. C. 27602
Shearon Harris Nuclear Power Plant NRC uestion 010.00, Hi h/Low Pressure Interface Provide a discussion of the fire protection features which prevent the possibility of a fire induced LOCA through a high/low pressure interface.
Revised Res onse The following summary describes the systems being protected to exclude the possibility of a fire induced LOCA.
Reactor Coolant S stem Hi h/Low Pressure Interface Summar The following Reactor Coolant System high-low pressure interfaces rely on redundant electrically controlled devices to maintain primary system integrity.
l.
RHR Suction Isolation Valves (FSAR Figure 5.0.7-1)
IRH-V500 SB redundant to IRH-V501 SA IRH-V502 SB redundant to IRH-V503 SA 2.
Power Operated Relief System (FSAR Figure 5.1.2-2)
IRC-P527 SN in series with IRC-V526 SN IRC-P528 SN in series with IRC-V527 SN IRC-P529 SN in series with IRC-V528 SN The three sets of valves are in parallel 3.
Letdown Isolation System (FSAR Figure 9.3.0-1) 2CS-V512 SA, 2CS-V511 SA, 2CS-V513 SA in parallel ICS-L500 SN in series with ICS-L501 SN 0.
Primary Sampling System (FSAR Figure 9.3.2-1) 2SP-V21-SA-I 2SP-V22-SB-I 2SP-V I I I-SB-I redundant to 2SP-V23-SA-I 2SP-V12-SA-I redundant to 2SP-V I I-SB-I 2SP-Y2-SA-I redundant to 2SP-V I-SB-I 2SP-V116-SA-I OSP-V38 I-I OSP-V382-I 2SP-V113-SB-I 2SP-VII'-SB-I 2SP-V115-SB-I 5.
Reactor Coolant Vent System (FSAR Figure 5.1.2-1 and 5.1.2-2 to be revised to include this system).
See response to DSER Open Item No. 07 2RC-V280 SB redundant to 2RC-V281 SA 2RC-V282 SB redundant to 2RC-V283 SA The above valves may be aligned to provide flow to either 2RC-V280 SA 2RC-V285 SB (1950JDK/ccc
)
Spurious operation due to stray voltages between circuits within equipment (control panels, switchgear, auxiliary relay panels (ARP), etc.) is considered for the reactor coolant system high-low pressure interfaces.
1.
RHR Suction Isolation Valves (Fig. 1)
The cabling has been analyzed to ensure that the in-series isolation valves meet the criteria of 10 CFR 50, Appendix R, Section III.G. The SHNPP RHR suction isolation valve design meets the criteria of III.G.
2.
Power Operated Relief System (Fig. 2)
The cabling for each series of PORV's and block valves is routed to meet the requirements of 10 CFR 50, Appendix R, Section III.G in order to prevent a fire-induced LOCA from occurring through an isolation path.
Spurious operation due to the application of stray voltages between cables within a common raceway resulting from fire-induced damage has been considered in the case of the power operated relief system.
However, spur ious operation due to three-phase stray voltages between cables is not considered a credible event.
The SHNPP Power Operated Relief System design meets the criteria of 10 CFR 50, Appendix R, Section III.G.
3.
Letdown Isolation System (Fig. 3)
The cabling has been analyzed per 10 CFR 50, Appendix R, Section III.G so that either 1CS-L500-SN-I or 2CS-V511, V512, and V513 SA-1 are available to isolate the letdown system.
The SHNPP letdown isolation system design meets the criteria of III.G.
0.
Primary Sampling System The Primary Sampling System consists of 3/8" tubing with a flow restrictor in the primary loop outlet.
The loss of fluid in this system due to any malfunction caused by a fire would not cause a LOCA condition since the discharge from the charging pumps willadequately make-up for any loss in the system.
Therefore, no analysis of the cabling to meet 10 CFR 50, Appendix R, Section III.G is deemed necessary.
5.
Reactor Coolant Vent System (Fig. 0)
The cabling has been analyzed per 10 CFR 50, Appendix R, Section III.G so that either 2RC-V285SB-I and 2RC-V280SA-I or 2RC-V280SB, V281SA, V282SB, V283SA-I are available to isolate the reactor coolant vent system.
The SHNPP reactor coolant vent system design meets the criteria of III.G.
(1950JDK/ccc
)
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Shearon Harris Nuclear Power Plant NRC uestion No. 010.50, Double Fusin An electrical isolation deficiency has been identified in which fuses in transfer switches might have had to be replaced in order to maintain a hot shutdown condition.
This deficiency is discussed in the attachment (Appendix A). Provide notification as to whether a similar problem exists in the transfer switches in the Shearon Harris plant. If such a deficiency exists in the Shearon Harris plant, you should further advise us as to how this deficiency willbe corrected as discussed in Appendix A.
Revised Res onse An electrical isolation deficiency such as that described does not exist at the Shearon Harris plant.
120VAC essential control circuits cables have been isolated with transfer switches and provided with redundant fuses in the motor control centers or transfer panels.
Essential 125VDC control circuits which have both positive and negative leads within the same cable (potential for wire to wire faults) willalso be provided with redundant fuses in the switchgear/power center or transfer panels to further ensure their operability after transfer.
As a general rule in the Harris plant design, only the positive leg of the essential circuits are switched at the Main Control Board.
The negative legs in the Main Control Board are for the indicating light circuits of air operated and solenoid operated valves as shown on the attached figure. The grounding of a positive leg of an essential 125VDC circuit coincident with the grounding of one of the negative legs in another cable having the same DC common is considered unlikely in view of the following design features of the Shearon Harris Main Control Board:
1.
2.
30 low combustible loading, use of fire retardant materials, use of enclosed switch modules, use of individual flex conduit enclosed panel wiring.for each switch module.
Al
~
r Based on the above, CPdcL feels it has demonstrated that non-grounded cable-to-cable DC shorts are improbable.
(1950 JOK/ccc )
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TYPICAL PLUCs DETAIL
Shearon Harris Nuclear Power Plant NRC uestion 010.53a, Associated Circuits Your response to Question 010.28 regarding associated circuits appears to be ambiguous, in part.
In order to clarify your response, please provide the following information:
a.
With regard to associated circuits having a common power source, you stated that power feeder(s) from buses, power centers and motor control centers are provided with breakers or fuses to isolate fire-induced electrical shorts so as to prevent
, tripping of or damage to the power source.
Show that you have considered suitability of the inter rupting devices in accordance with the criteria of Section Il.b.2 of the letter forwarded to you on October 18, 1983 (also attached herein as Appendix B pages 0 and 5).
Revised Res onse Section 8.3.1.1.2.11 of the Shearon Harris FSAR provides a description of the electric circuit protection system provided. Allelectrical equipment is designed and manufactured to applicable ANSI, NEMA, IEEE and other industry standards.
A breaker coordination study is being performed for the Shearon Hanis Nuclear plant and periodic testing shall be performed to demonstrate that the overall scheme remains within the design criteria for 6.9KV and 080 volt safety-related circuit breakers.
The periodic testing willconsist of functionally testing at least 1096 of the 6.9KV safety related breakers and l096 of the 080V safety related breakers every 18 months.
For each circuit breaker found inoperable an additional 1096 of all circuit breaker of the inoperable type willbe functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.
Molded case circuit breakers that supply penetrations willbe tested as required by Technical Specifications.
(l950JDK/ccc
)
Shearon Harris Nuclear Power Plant NRC uestion, Auxiliar Control Panel Procedures In an informal request for information the NRC reviewer requested the following information of the procedure used to shutdown the plant from the AuxiliaryControl Panel:
l.
2.
3.
number and title of the procedure brief description of procedures contents location of controlled copy of the procedure to be used in the event of a control room fire.
~Res ense The procedure used to shutdown the plant from the AuxiliaryControl Panel is Abnormal Operating Procedure (AOP) -0, "Alternate Safe Shutdown in Case of Fire or Control Room Inaccessibility."
The procedure consists of those steps required to transfer control from the Main Control Room to the AuxiliaryControl Panel and AuxiliaryTransfer Panels and safely shutdown and cooldown the reactor to cold shutdown conditions.
The procedure provides instruction on the operation of those systems which can be relied upon in event of a fire which has affected all or part of the equipment and controls which are located in the Main Control Room fire area.
A controlled copy of AOP-0 will be maintained in the ACP room, (l950JOK/ccc
)
Shearon Harris Nuclear Power Plant NRC uestion, Safe Shutdown Methodolo Provide a summary of the SSNPP Fire Protection Safe Shutdown Analysis Methodology.
~Res onse Step I Each fire area in which essential safe shutdown components are located was analyzed separately.
Deviations were taken to analysis by fire area for Fire Areas 1-A-BALand 12-A-BALwhere the analysis was done as described in Table 9.5B-3 transmitted February 20, 1980.
These deviations are detailed in Table 9.5B-3 by safe shutdown analysis area.
Step II A single safe shutdown system was analyzed in the fire area of interest.
1.
By using "Equipment List:
10096 Power to Hot Standby" and/or "Equipment List: Hot Standby to Cold Shutdown" in conjunction with the applicable Safe Shutdown Analysis drawing(s), all essential safe shutdown equipment was investigated as follows:
ao b.
Cs Allessential safe shutdown equipment was located.
The functionally redundant counterparts for the equipment located in Step Il-l.a were located.
It was determined whether or not each piece of essential safe shutdown equipment is separated from its redundant counterpart as required by Section III.G of Appendix R to 10 CFR 50.
2.
Allcables connected to all the essential safe shutdown equipment identified in Steps Il.l.a and Il.l.b were investigated as follows:
as b.
C.
d.
The cable numbers for those cables which are connected to essential safe shutdown equipment were determined.
Each cable's function and whether it is essential for safe shutdown was determined.
Also, it was established which cables are functionally redundant.
The locat'ion, in conduit and/or tray, of each essential cable was determined.
Electrical Cable Tray and Conduit Drawings were used to locate plan points for cable located in cable trays.
It was determined whether or not each essential cable is separated from its functionally redundant counterpart(s) (both equipment and cabling) as required by Section III.G of Appendix R to 10 CFR 50.
3.
Those essential safe shutdown components (both equipment and cabling) that have their functionally redundant counterpart(s) located in the'same safe shutdown area were investigated as follows:
a.
The horizontal spacing between redundant components was determined.
b.
The presence or absence of intervening combustibles was verified by using the SHNPP Fire Hazard Analysis and the Electrical Cable Tray and Conduit Drawings.
(1950JDK/ccc
)
c.
For those components not meeting the separation criteria stated in Section III.G of Appendix R to 10 CFR 50, one or a combination of the following was done:
i.
The plant design was modified to conform with Section III.G.
ii.
A deviation request, with technical justification, identifying acceptable alternatives to the requirements of Section III.G was prepared.
0.
The analysis was performed as described in Step II for all remaining safe shutdown systems located in the fire area of, interest.
Step III When the analysis was completed for one fire area, the analysis was repeated, Steps I through III, for all remaining fire areas of the plant which contain essential safe shutdown equipment.
(1950 JDK/ccc )