ML17347B628

Forwards Response to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47 'Safety Implications of Control Sys in LWR Nuclear Power Plants.'

ML17347B628
Person / Time
Site:	Turkey Point
Issue date:	03/19/1990
From:	Goldberg J FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-A-47, REF-GTECI-SY, TASK-A-47, TASK-OR GL-89-19, L-90-108, NUDOCS 9003260402
Download: ML17347B628 (13)

Text

.ACCELERATED DISTIGBUTION DEMONSTPD,TION SYSTEM

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REGULATORY INFORMATION DXSTRXBUTION SYSTEM (RIDS)

ACCESSION NBR 9003260402 DOC.DATE: 90/03/19 NOTARXZED:

FACIL:50-250 Turkey Point Plant, Unit 3, Florida Power and Light

~ +~ C DOCKET 05000250 50-251 Turkey Point Plant, Unit 4, Florida Power and Light C 05000251 AUTH. NAME AUTHOR AFFXLIATION GOLDBERG,J.H. Florida Power & Light Co.

RECIP.NAME RECIPIENT AFFXLIATION Document Control Branch (Document Control Desk)

SUBJECT:

Responds to Generic Ltr 89-19, "Request for Action re Resolution of USI A-47."

DXSTRIBUTION CODE: A001D COPIES RECEIVED:LTR t ENCL ~ SIZE:

TITLE: OR Submittal: General Distribution NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 LA 1 1 PD2-2 PD 1 1 EDISON,G 5 5 INTERNAL: NRR/DET/ECMB 9H 1 1 NRR/DOEA/OTS B 1 1 1 1 NRR/DST 8E2 1 1 NRR/DST/SELB 8D 1 1 NRR/DST/SICB 7E 1 1 NRR DST SRXB 8E 1 1 NUDOCS-ABSTRACT .1'1 1 t 1 0 OGC/HDS2 0 EG F~E 01 1 1 RES/DSIR/EIB 1 1 EXTERNAL: LPDR 1 1 NRC PDR NSIC 1 1 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP. US TO REDUCE WASTE! CONTACT THE. DOCUMENT CONTROL DESK, ROOM PI-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISIRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEEDl TOTAL NUMBER OF COPXES REQUIRED: LTTR 21 ENCL 19

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P.O. Box 14000, Juno Beach, FL 33408-0420 KERCH 'j 9 1990 L-90-108 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:

Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Generic Letter 89-19 Request for Action Related to Resolution of Unresolved Safety Issue A-47 "Safety Implications of Control Systems in LWR Nuclear Power Plants" Pursuant to 10 CFR 50.54 f As a result of the technical resolution of USI A-47, "Safety Implications of Control Systems in LWR Nuclear Power Plants," the NRC concluded that protection should be provided for certain control system failures, and, for certain plants, that selected emergency procedures should be modified to assure that plant transients resulting from control system failures do not compromise public sa'fety.

Generic Letter 89-19, issued September 20, 1989, provided recommendations for control system design and procedural modifications for resolution of USI A-47. Specifically, Generic Letter 89-19 recommended that. all Westinghouse plant designs provide automatic steam generator overfill protection to mitigate main feedwater (MFW) overfeed events, and that plant procedures and technical specifications include provisions to periodically verify the operability of the MFW overfill protection and ensure that the automatic overfill protection is operable during reactor power operation.

Florida Power & Light Company's response to the recommendations in Generic Letter 89-19 is attached.

Should there be any questions regarding the attached information, 0 please contact us.

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FiO Very truly yours, GO OIA RO 82~

GU J. H. Goldberg Acr OA Executive Vice President WcX N JHG/TCG/gp (9 I OQ OQ Attachment

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cc: Stewart D. Ebneter, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Plant an FPL Group company

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STATE OF FLOR7DA )

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COUNTY OF PALM BEACH )

Z. H. Goldber being first duly sworn, deposes and says:

That he is Executive Vice President, of Florida Power & Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.

J. H. Goldber Subscribed and sworn to before me this

~today of W64Lcl, 19 9+.

C NOTHER'I ->PtiBX,7C; in and for the County of Palm Bezel.," State of Florida PBotary fsub1i(, State of Horirfa June 1, 1993 Kty Commission Expires My Commission expires f] Inw c I<.

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ATTACHMENT Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Generic Letter 89-19 Request for Action Related to Resolution of Unresolved Safety Issue A-47 "Safety Implication of Control Systems,'in, LWR Nuclear Power Plants" Pursuant to "10 CFR '50.54 f At Turkey Point Nuclear Plant Units 3 and 4, Steam Generator Level Protection Channels I, II, III, and IV are designed to combine redundant sensors, independent"channel circuitry, coincident trip logic and different parameter measurements so that a safe and reliable system is provided that is single failure proof.

Channels I and is used for level IIcontrol.

are used for level protection while channel IV Channel III is used for both protection and control. The steam generator overfill protection at Turkey Point is initiated on a S/G Hi-Hi Level signal, based on a 2 out of 3 initiating logic which is safety related.

The steam generator protection and portions of the control systems utilize shared power sources. However, sufficient power and logic diversity exists to ensure steam generator overfill protection.

As detailed below, a review of the Turkey Point Nuclear Plant Units 3 and 4 S/G Level Control and Protection system and applicable procedures indicates that it falls within the acceptable NRC Category (Group I) gpr Westinghouse PWRs, and no further actions are required.

S G Protection Sufficient redundancy and separation exists between the channels and applicable power supplies. The protection channels (I, II and III) are located in separate analog racks, providing adequate physical separation. The major instruments are arranged to physically separate the protection equipment from the control equipment. Power cable routing was not specifically reviewed, but based on the vintage of the Turkey Point units, there is not reasonable assurance that present day separation criteria were used. However based on the total system design we believe that adequate protection exists such that the effects of environmental factors including fire, electrical transients, and physical accidents are reduced. Channel III electrical separation between protection and control is achieved by employing isolation amplifiers. Redundant HVAC systems are provided for equipment reliability. Redundant trains "A" and "B", ensure that the feedwater (FW) pumps are tripped and that the main and by-pass feedwater control valves are closed in a Hi-Hi S/G Level condition (804 on 2 out of 3 level transmitters). (See Table 1.)

The feedwater control valves are also provided with redundant components and redundant FW isolation signals to ensure closure.

In addition, these valves are designed to fail closed if the air supply to them is lost (See Table 2).

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TABLE 1 Table 1 is for S/G Loop "A"; S/G loops "B" and "C" are similar.

Steam Gen. Associated Prot. Channels Inst. Si nal In uts LT-474 At 80% Hi-Hi Level

~ Trip Main Turbine LT-475 ~

Trip FW pumps

~ Close FW Control Valves LT-476 TABLE 2 COMPONENTS Main FW Control Valves S/G A FCV-478 S/G B FCV-488 S/G C FCV-498 By-Pass FW Control S/G 'A FCV-479 Valves S/G B FCV-489 S/G C FCV-499 S G Level Control In order to maintain a programmed water level in the secondary side of the steam generators during normal plant operation, the steam generators are equipped with a three-element feedwater control system. The three-element feedwater control system in each steam generator receives input from the following four parameters 3): (see'able

1. Feedwater Flow

2. Steam Flow

3. Steam Generator Water Level 4 ~ Turbine First Stage Pressure The controllers continuously compare feedwater flow with steam flow, and a steam generator water level signal with a water level setpoint to regulate the main feedwater control valve position.

Manual control of the main feedwater control valve is provided by the auto-manual station on the control room console. A bypass feedwater control valve is provided around each main feedwater flow control valve. The bypass valve is used to control feedwater flow at loads below approximately 15 percent.

TABLE 3 This table is for S/G Loop "A"; S/G Loops "B" and "C" are similar.

Steam Gen Level Associated Cont. Channels Inst. Si nal In uts Channel III LT-476 FT-474 S/G Level Steam Flow FT-477 Feedwater Flow PT-446 Turbine First Stage Pressure Channel IV FT-475 Steam Flow FT-476 Feedwater Flow PT-447 Turbine First Stage Pressure Procedures Technical S ecifications The Turkey Point plant procedures (Table 4) and Technical Specifications, Section 4.1, Table 4.1-1 (Section 3/4.3.1, Table 4.3-1 of the Revised Technical Specifications submitted on June 5, 1989) include requirements to periodically verify the safety function of Lo-Lo level trip operability of the S/G Level Control Channels for the S/G Level Control And Protection System. The procedures which perform the function of verifying operability of the Lo-Lo level reactor trip also check the Hi Steam Generator Level (804) trip. Other procedures, not required by Technical Specifications, provide for monthly analog channel operational testing, periodic (CHANNEL) calibration and functional testing of instrumentation and operability tests of the feedwater control valves. Feedwater isolation testing, which includes tripping of the feedwater pumps and closure of the feedwater control valves, is performed with the unit in cold shutdown in accordance with the applicable engineering safeguards integrated test procedures.

Those procedures required to be performed by Technical Specifications and those which are not are delineated on Table 4.

The NRC staff is currently reviewing the Nuclear Steam Supply System vendor specific revised standard technical specifications.

Recommendations for specific Limiting Conditions for Operation pertaining to steam generator overfill protection should be addressed as part of that review. Specific changes to Turkey Point Technical Specifications will be addressed upon completion of that review.

TABLE 4 Procedures Descri tion ~Fre uenc 3-SMI-071.1* Steam Generator Protection Set I (QR-3) Every 31 days Analog Channel Test when in Operating Modes 1, 2, or 3 and within 31 days 4-SMI-071.1* Steam Generator Protection Set I (QR-3) prior to entering Analog Channel Test Mode 3 from Mode 4

3-SMI-071.2* Steam Generator Protection Set II (QR-13)

Analog Channel Test 4-SM!-071.2* Steam Generator Protection Set II (QR-17)

Analog Channel Test 3-SMI-071.3* Steam Generator Protection Set III (QR-17)

Analog Channel Test 3-SMI-071.4* Steam Generator Protection Set III (QR-16)

Analog Channel Test 4-SMI-071.4* Steam Generator Protection Set III (QR-16)

Analog Channel Test 3-SMI-071.5* Steam Generator Protection Set I (QR-18)

Analog Channel Test 4-SMI-071.5* Steam Generator Protection Set III (QR-18)

Analog Channel Test 3-SMI-071.6* Steam Generator Protection Set IV (QR-25)

Analog Channel Test 4-SMI-071.6* Steam Generator Protection Set IV (QR-25)

Analog Channel Test 3-SMI-071.7* Steam Generator Protection Set IV (QR-24)

Analog Channel Test 4-SMI-071.7* Steam Generator Protection Set IV (QR-24)

Analog Channel Test

• Technical Specification required.

~TL P d ~0 ~Fre uenc 3-PMI-071.2* Steam Generator Level (Narrow Range) L-3-477, Refu cling L-3-487, L-3-497 Channel Calibration 4-PMI-071.2* Steam Generator Level (Narrow Range)

Protection Instrumentation Set 1 L-474,

, L-484, L-494 Channel Calibration 3-PMI-071.3* Steam Generator Level (Narrow Range)

Protection Instrumentation Set 1 L-474, L-484, L-494 Channel Calibration 4-PMI-071.3* Steam Generator Level (Narrow Range)

Protection Instrumentation Set II L-475, L-485, L-495 Channel Calibration 3-PMI-071.4* Steam Generator Level (Narrow Range)

Protection Instrumentation Set III L-476, L-486, L-496 Channel Calibration 4-PMI-071.4* Steam Generator Level (Narrow Range)

Protection Instrumentation Set III L-476, L-486, L-496 Channel Calibration 3-PMI-071.5 Steam Generator Level (Alternate Control)

L-478, L-488, L-498 Channel Calibration 4-PMI-071.5 Steam Generator Level (Alternate Control)

L-478, L-488, L-498 Channel Calibration 3-PMI-071.6 Steam Generator "3A" Steam/FW Flow Mismatch Protection Instrumentation Set III F-3-474 and F-3%77 Channel Calibration 4-PMI-071.6 Steam Generator "4A" Steam/FW Flow Mismatch Protection Instrumentation Set III F-4-474 and F-4-477 Channel Calibration.

3/4-OSP-074.2 S/G FW Control Valves Operability Test 18 Months 3-OSP-203* Engineered Safeguards Integrated Test 18 Months 4-OSP-203.1* Train A Engineered Safeguards Integrated Test 18 Months 4-OSP-203.2* Train B Engineered Safeguards Integrated Test 18 Months

• Technical Specification required.

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