ML20196J143

Proposed Tech Specs Adding EDG Kilowatt Indication to post- Accident Monitoring Instrumentation to Support CR-3 Restart Issue of EDG Load Mgt

ML20196J143
Person / Time
Site:	Crystal River
Issue date:	07/29/1997
From:	FLORIDA POWER CORP.
To:
Shared Package
ML20196J141	List: 3F0797-21, TS Change Request 209,Rev 1 for License DPR-72,adding EDG Kilowatt Indication to post-accident Monitoring Instrumentation to Support CR-3 Restart Issue of EDG Load Mgt ML20196J143
References
NUDOCS 9708040010
Download: ML20196J143 (50)

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ATTACHMENT D FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 i

DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 i

l TECHNICAL SPECIFICATION CHANGE REQUEST NOTICE 209, REVISION 1 i

l STRIKEOUT / SHADOW PAGES i

l Deletions are indicated by strikeout.

Additional and replacement text are indicated by shading.

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i ggSo2MRUTg82 P

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PAM Instrumentation 3.3.17 Table 3.3.17-1 (page 1 of 1)

Post Accident Monitoring Instrumentation CONDITIONS REFERENCED FROM FUNCTION REQUIRED CHANNELS REQUIRED ACTION D.1 1.

Wide Range Neutron Flux 2

E l

2.

RCS Hot Leg Temperature 2

E 3.

RCS Pressure (Wide Range) 2 E

l 4.

Reactor Coolant Inventory 2

F 5.

Borated Water Storage Tank Level 2

E 6.

High Pressure Injection Flow 2 per injection line E

j 7.

Containment Sump Water Level (Flood Level) 2 E

j 9.

Containment Pressure

( h Expected Postf 2

E Accident Range) 9.

Containment Pressure (Wide Range) 2 E

10.

Containment Isolation Valve Position 2 per penetration (a) (b)

E

]

11.

Containment Area Radiation (High Range) 2 F

12.

Containment Hydrogen Concentration 2

E 13.

Pressurizer Level 2

E 14.

Steam Generator Water Level (Start-up Range) 2 per OTSG E

15.

Steam Generator Water Level (Operating Range) 2 per OTSG E

16.

Steam Generator Pressure 2 per OTSG E

37.

Emergency Feedwater Tank Level 2

E 18.

Core Exit Temperature (Backup)

-.... - - f i E

3 per core; quadrant 19.

Emergency Feedwater Flow 2 per OTSG E

26

Low PressyelInjection. Flow 2

H 21(

[Degreesy f Subcooling F (d)

E 22.

Emergent / Diesel Generator kW Indication Fic)

E ta)

Only one position indication is required for penetrations with one Control Room indicator.

(b)

Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

(c): One indicator.per EDG (d)"

These two channels' of subcooling' margin are backed up by either'of two ' indications: of D

~

subcooling margin based on similar inputs.through the safety. Parameter Display-.

~

System (SPDS}, At least one-SPDS channel must be available to provide-this backupt Withlbothj SPDS. channels. INOPERABLE Condition C is-applicable.

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I Crystal River Unit 3 3.3-41 Amendment No.

PAM Instrumentation B

3.3.17 BASES LCO The following list is a discussion of the specified (continued) instrument Functions listed in Table 3.3.17-1.

1.

Wide Range Neutron Flux Two wide-range neutron flux monitors are provided for post-accident reactivity monitoring over the entire range of expected conditions.

Each monitor provides indication over the range of 10-8 to 100% log rated power covering the source, intermediate, and power ranges.

Each monitor utilizes a fission chamber neutron detector to provide redundant main control board indication.

A single channel provides recorded information in the control room.

The control' room indication of neutron flux is considered one of the primary indications used by the operator following an accident.

Following an event the neutron flux is monitored for reactivity control. The operator ensures that the reactor trips as necessary and that emergency boration is initiated if required. Since the operator relies upon this indication in order to take specified manual action, the variable is included in this LCo.

Therefore, the LCO deals specifically with this portion of the string.

2.

Reactor Coolant System (RCS) Hot Leo Temperature Two wide range resistance temperature detectors (RTD's), one per loop, provide indication of reactor coolant system hot leg temperature (Ts) over the range of 120* to 920 F.

Each T measurement provides an y

input to a control room indicator.

Channel B is also recorded in the control room.

Since the operator l

relies on the control room indication following an accident, the LCO deals specifically with this portion of the string.

Tu is a Twe A variable on which the operator bases manual actions required for event mitigation for which no automatic controls are provided.

Thic tcmperctur-e T.ccourement providcc input to the incdcquctc core eecling inctrumentction hich ic uccd to /crify the Crystal River Unit 3 B

3.3-126 Amendment No.

l 1

PAM Instrumentation B

3.3.17 i

BASES LCO 2.

Reactor Coolant System (RCS) Hot Leo Temnerature (continued) e*ictcnce of, or to take actienc to encurc the rcctoratien Of cubcccling margin Opccifically, a lecc cf adcquatc cubcccling margin during a cmall brcah LOCA requircc the opcratcr to trip the reacter ccclant pumpc

'"CP'c',

encurc high cr 10w prcccurc injcction, and raic; the Otcam generator levcic to the CCC level.

1 Once cubccoling margin ic rectored, the operatcr ic inctructed to rectart at 1cact onc "CP and thrcttic injection flow to maintain a opccified degrce of cu~cccling. Ancther manual action baccd on 'F, followc FollowingCa steam generator tube rupturer.

Tthe affected steam generator is to be isolated only after Tu f alls below the saturation temperature corresponding to the valves. pressure setpoint of the main steam safety For event monitoring once the RCP's are I

tripped, Tu is used along with the core exit temperatures and RCS cold leg temperature to measure i

the temperature rise across the core for verification i

of core cooling.

3.

RCS Pressure (Wide Rance)

RCS pressure is measured by pressure transmitters with a span of 0-3000 psig. Redundant monitoring ca j

is provided by two trains of instrumentation. pability Control i

room and remote shutdown panel indications are provided.

The cubccoling margin moniter can c1cc dicpla'i room indications are the primary indications reactor ccclant precourc u On dcmand.

The contro used by the operator during an accident.

Therefore, the LCO deals specifically with this portion of the i

instrument string.

RCS pressure is a Type A variable because the operator

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uses this indication to adjust parameters such as steam generator (OTSG) level or pressure in order to monitor and maintain a controlled cooldown of the RCS following a steam generator tube rupture or small break LOCA. In addition, HPI flow is throttled based Crystal River Unit 3 B

3.3-127 Amendment No.

I PAM Instrumentation B

3.3.17 BASES LCO 8,9.

Containment Pressure (!!a rrow Expected' Post-accident l

~

Range and Wide Range)

The containment pressure variable is monitored by two ranges of pressure indication.

!! arrow Expected posta accidentfrange (-10 to frG 70.'psig) and wide' range'- (0 ~ to 200 psig)' pressure indication each provide two channels of pressure indication. Channel A and B wide range containment pressure are recorded in the associated ' A' and ' B' EFIC Rooms. The low range is required in order to ensure instrumentation of the necessary accuracy is available to monitor conditions in the RB during DBAs.

The wide range instrument was required by Regulatory Guide 1.97 to be capable of monitoring pressures over the range of atmospheric to three times containment design pressure (approximately 165 psig). Thus, it was intended to monitor the RB in the event of an accident not bounded by the plant safety analysis (i.e.,

a Severe Accident).

These instruments are not assumed to provide information required by the operator to take a

mitigation action specified in the accident analysis.

As such, they are not Type A variables. However, the monitors are deemed risk significant (Category 1) and are included within the LCO based upon this consideration.

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l Crystal River Unit 3 B

3.3-131 Amendment No.

PAM Instrumentation B

3.3.17 BASES LCO

18. Core Exit Temperature (Backuo)

(continued) f ollowing a steam generator tube rupture or small break LOCA.

Operator actions to maintain a controlled cooldown, such as adjusting OTSG level or pressure, would be prompted by this indication. :n addi icn, the 1

core cnit thermoccuplco provide input tc the cubcccling margi:. monitor, hich ic a Typc A /ariabic.

j The cubccoling margin monitor takcc the averagc of the five highcot CCTc for cach of the ICC" trainc.

T. 0 channelc cncurc that a cingle f ailurc till not dicable the ability to dctcrmine the reprccentativc ccrc cnit tcmpcrcturc.

19.

Emergency Feedwater Flow EFW Flow instrumentation is provided to monitor operation of decay heat removal via the OTSGs. The EFW injection flow to each OTSG (2 channels per OTSG, one associated with each EFW injection line) is determined from a dif ferential pressure measurement calibrated to a span of 0 gpm to 1000 gpm.

Each differential pressure transmitter provides an input to a control room indicator and the plant computer.

EFW Flow is used by the operator to determine the need to throttle flow during accident or transient conditions to prevent the EFW pumps from operating in runout conditions or from causing excessive RCS cooldown rates when low decay heat levels are present.

EFW Flow is also used by the operator to verify that the EFW System is delivering the correct flow to each OTSG. However, the primary indication of this function is provided by OTSG level.

These instruments are not assumed to provide information required b,y the operator to take a

mitigation action specified in the safety analysis. As such, they are not TWe A variables. However, the monitors are deemed risk significant (Category 1) and are included within the LCO based upon this consideration.

Crystal River Unit 3 B

3.3-138 Amendment No.

PAM Instrumentation B

3.3.17 BASES LCO

20. " Low ~ Pressure"Iniection ' Flow Low pressure" injection' flow instrumentation ~ is provided to monitor 0 flow > to the : RCS"::following 1 a' large Ebreak LOCA.' iltfis also 'used to monitor LPI flow -during ' piggy back operation following)afsmall" break LOCA. 'The. low pressure ; injection flow to the reactor -(2. channelsk.one associated with each LPI injection.line) is determined from a dif ferential! pressure. measurement:, calibrated to a: span of'0;gpm. tov 50.00 gpm; The L LPI ? flow " indication 11s iused1 byy theloperatbri to throttle the Kflow; to s. 2000. gpm prior to switching the pump. suction 1: f rom -~ the BWST o to 4 the 2 RB 1 sump...

This assures adequate net positive suction head 1(NPSH) ::is maintained to the pump. The indication'is also used-to verify LPI flow to the reactorfas.a7 prerequisite-to termination. of: HPI flow.

Since~ 1'ow: pressure :inj ection" flow.Fis a1 Type 'A vari'able' i

on which the operator. bases, manual' actions required for event, mitigation foriwhich;no. automatic-controls"are providedrit has beenfincludedEin-thiscLCO;

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21' Decrees"of subcoolina Two channelslof subcooling. margin withrinputs from.RCS

~

hot leg temperature (T,,), core exit. temperature', and RCS pressure are provided. Multiple: core exit temperatures are auctioneered ; withi only theYhighestt temperature being:1. input:to-the monitor.'DA note >has.beenEadded to indicate that.the~ two channels oflsubcooling margin'are backed;:up'.by either of:two: indications 7of subcooling margin based : on-: similareinputs;:- through :1the H Safety Parameter? Display ~ System 0 (SPDS)'.. ': Atfleast4lone9 SPDS channel must be available to provide this backup? C

.<With both SPDS channels INOPERABLE,.xCondition eis applicables ;This-is consideredinecessary becauseEthe core exit thermocouple inputs to the: subcooling: margin monitors are not environmentally qualified.- :The T.

inputs to ; the subcooling : margin-monitors < and SPDS operate.over a-range.of 120x to 920'F. The : core - exit temperature - inputs operate 3 over aay range : of 9150 o to 2000*F and 150 to :E2500 Fi for7 the "subcooling1. margin monitors and SPDS, respectively.

RCS: pressure:Linputs operate;over a' range of-200!to;2500lpsig.

Crystal River Unit 3 B

3.3-138a Amendment No.

LCO

21. #Dectrees Tof "Subcoolind" (continded)

The subcoolingTmargin monitors arefused 7to. verify.:the existenceJ :of, 'or: to take : actionsf. - ton -: ensure : ' the restoration of subcooling margin. ! Specifically,Ja" loss oftadequate subcooling margin;duringta:LOCA; requires the operator to trip the. reactor coolant: pumps : (RCP's);:,

to ensure 'high or low pressure -injection,- and raise the steam :: generator levels to x the.' inadequate subcooling margin' level. Since-de variable 4 on::which1 the grees of subcooling. ista Type -A c

operator'- basesi manual? actions required : f or/: event: mitigation : for. which t noyautomatic control. s. _:o a.r.e. provided, i.t 0 h..a. s j b.ee. n?inc. lude. d sin th. is.

221 l:Emerdency Dissel" Generator $kW Indication The Eme rgency Die s el' Generatorsl( EDG) ! providesTs tahdby L{ emergency} L electrical poweriin.~ the' case:.'of!T LossN of Offsite!PowerH(LOOP). EDG kilowatt L(kW) rindicationfis providedLin the control room'to monitor the operational-statuscofethe-EDGs.

EDG.p6wer (kW) output !LinNeation: is T aETy 4

because EDG: kW-indication provides D the pe Al variable

.controleroom operator EDG load management:.. capabilities. ?EDGiload managementlenables the operator >to base manual actions of-load-start and stop)fortevent mitigation; l

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Crystal River Unit 3 B

3.3-138b Amendment No.

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i ATTACHMENT E I

i FLORIDA POWER CORPORATION i

CRYSTAL RIVER UNIT 3 i

DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 3

TECHNICAL SPECIFICATION CHANGE REQUEST NOTICE 209, REVISION 1 4

j REVISION BAR PAGES 1

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PAM Instrumentation 3.3.17 Table 3.3.17-1 (page 1 of 1)

Post Accident Monitoring Instrisnentation CONDITIONS REFERENCED FROM FUNCTION REQUIRED CHANNELS REQUIRED ACTION D.1 1.

Wide Range Neutron Flux 2

E 2.

RCS Hot leg Temperature 2

E 3.

RCS Pressure (Wide Range) 2 E

4 Reactor Coolant Inventory 2

F 5.

sorated Water Storage Tank Level 2

E 6.

High Pressure injection Flow 2 per injection line E

7.

Contairunent Sump Water Level (Flood Level) 2 E

d 8.

Containment Pressure (Expected Post-Accident 2

E l

Range) 9.

Containment Pressure (Wide Range) 2 E

10. Containment Isolation Valve Position 2 per penetration (a)(b)

E

11. Containment Area Radiation (High Range) 2 F

i

12. Containment Hydrogen Concentration 2

E

13. Pressurizer Level 2

E

14. Steam Generator Water Level (Start-up Range) 2 per OTSG E

15. Steam Generator Water Level (Operating Range) 2 per OTSG E

16. Steam Generator Pressure 2 per OTSG E

17. Emergency Feedwater Tank Level 2

E

18. Core Exit Temperature (Backup) 3 per core quadrant E

l

19. Emergency Feedwater Flow 2 per OTSG E

20. Low Pressure Injection Flow 2

E

21. Degrees of Subcooling 2(d)

E

22. Emergency Diesel Generator kW Indication 2(C)

E (a) only one position indication is required for penetrations with one Control Room indicator.

(b) Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secu W.

(c) One indicator per EDC.

l (d) The,e two channels of subcooling margin are backed up by either of two indications of subcooling margin based on similar inputs through the Safety Parameter Display System (SPDS). At least one SPDS channel must be available to provide this backup. With both SPDS channels INOPERABLE, Condition C is applicable.

Crystal River Unit 3 3.3-41 Amendment No.

\\

j PAM Instrumentation l

B 3.3.17 l

BASES 1

LC0 The following list is a discussion of the specified (continued) instrument Functions listed in Table 3.3.17-1.

i 1.

Wide Ranae Neutron Flux Two wide-range neutron flux monitors are provided for j

post-accident reactivity monitoring over the entire i

range of expected conditions.

Each monitor provides indication over the range of 10-8 to 100% log rated i

power covering the source, intermediate, and power ranges.

Each monitor utilizes a fission chamber neutron detector to provide redundant main control board indication. A single channel provides recorded l

information in the control room.

The control room t

indication of neutron flux is considered one of the primary indications used by the operator following an accident.

Following an event the neutron flux is monitored for reactivity control.

The operator

]

ensures that the reactor trips as necessary and that emergency boration is initiated if required.

Since l

the operator relies upon this indication in order to l

take specified manual action, the variable is included i

in this LCO. Therefore, the LC0 deals specifically l

with this portion of the string.

i 2.

Reactor Coolant System (RCS) Hot Leo Temoerature i

Two wide range resistance temperature detectors j

(RTD's), one per loop, provide indication of reactor coolant system hot leg temperature (T ) over the range

{

of 120* to 920*F.

Each T measuremenf provides an g

j input to a control room indicator.

Channel B is also recorded in the control room.

Since the operator 4

relies on the control room indication following an j

accident, the LCO deals specifically with this portion j

of the string.

T is a Type A variable on which the operator bases g

]

manual actions required for event mitigation for which no automatic controls are provided.

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Crystal River Unit 3 8 3.3-126 Amendment No.

l

PAM Instrumentation B 3.3.17 BASES LC0 2.

Reactor Coolant System (RCS) Hot Lea Temoerature (continued)

Following a steam generator tube rupture, the affected I steam generator is to be isolated only after T fall s g

below the saturation temperature corresponding to the pressure setpoint of the main steam safety valves.

For event monitoring once the RCP's are tripped, T is g

used along with the core exit temperatures and RCS cold leg temperature to measure the temperature rise across the core for verification of core cooling.

3.

RCS Pressure (Wide Ranae)

RCS pressure is measured by pressure transmitters with a span of 0-3000 psig.

Redundant monitoring capability is provided by two trains of instrumentation. Control room and remote shutdown panel indications are provided.

The control room I

indications are the primary indications used by the operator during an accident. Therefore, the LC0 deals specifically with this portion of the instrument string.

RCS pressure is a Type A variable because the operator uses this indication to adjust parameters such as steam generator (OTSG) level or pressure in order to monitor and maintain a controlled cooldown of the RCS i

following a steam generator tube rupture or small break LOCA.

In addition, HPI flow is throttled based (continued)

Crystal River Unit 3 B 3.3-127 Amendment No.

l

s PAM Instrumentation B 3.3.17 BASES LCO 8,9. Containment Pressure (Expected Post-Accident Ranae and

~

(continued)

Wide Ranae)

The containment pressure variable is monitored by two ranges of pressure indication.

Expected post-accident range (-10 to 70 psig) and wide range (0 to 200 psig) pressure indication each provide two channels of pressure indication.

Channel A and B wide range containment pressure are recorded in the associated

'A' and 'B' EFIC Rooms.

The low range is required in order to ensure instrumentation of the necessary accuracy is available to monitor conditions in the RB during DBAs.

The wide range instrument was required by Regulatory Guide 1.97 to be capable of monitoring pressures over the range of atmospheric to three times containment design pressure (approximately 165 psig).

Thus, it was intended to monitor the RB in the event

~

of an accident not bounded by the plant safety analysis (i.e., a Severe Accident).

These instruments are not assumed to provide information required by the operator to take a i

mitigation action specified in the accident analysis.

As such, they are not Type A variables.

However, the monitors are deemed risk significant (Category 1) and are included within the LC0 based upon this consideration.

s (continued)

Crystal River Unit 3 B 3.3-131 Amendment No.

s PAM Instrumentat. n B 3.3.17 BASES LC0 18.

Core Exit Temoerature (Backup)

(continued) following a steam generator tube rupture or small break LOCA. Operator actions to maintain a controlled cooldown, such as adjusting 0TSG level or pressure, would be prompted by this indication.

l 19.

Emeraency Feedwater Flow EFW Flow instrumentation is provided to monitor operation of decay heat removal via the OTSGs.

The EFW injection flow to each 0TSG (2 channels per OTSG, one associated with each EFW injection line) is determined from a differential pressure measurement calibrated to a span of 0 gpm to 1000 gpm.

Each differential pressure transmittr r provides an input to a control room indicator and.th olant computer.

EFW Flow is used by the operatt to determine the need to throttle flow during accid <

or transient conditions to prevent the EFW p mps from operating in runout conditions or from causing excessive RCS cooldown rates when low decay heat levels are present.

EFW Flow is also used by the operator to verify that the EFW System is delivering the correct flow to each 0TSG.

However, the primary indication of this function is provided by 0TSG level.

These instruments are not assumed to provide information required by the operator to take a mitigation action specified in the safety analysis.

As such, they are not Type A variables. However, the monitors are deemed risk significant (Category 1) and are included within the LC0 based upon this consideration.

i (continued)

Crystal River Unit 3 B S.3-138 Amendment No.

PAM Instrumentation B 3.3.17 BASES i

LCO 20.

Low Pressure Iniection Flow (continued)

Low pressure injection flow instrumentation is provided to monitor flow to the RCS following a large break LOCA.

It is also used to monitor LPI flow i

during piggy back operation following a small break LOCA. The low pressure injection flow to the reactor (2 channels, one associated with each LPI injection 4

i line) is determined from a differential pressure measurement calibrated to a span of 0 gpm to 5000 gpm.

The LPI flow indication is used by the operator to throttle the flow to 12000 gpm prior to switching the pump suction from the BWST to the RB sump. This assures adequate net positive suction head (NPSH) is maintained to the pump. The indication is also used to verify LPI flow to the reactor as a prerequisite to termination of HPI flow.

Since low pressure injection flow is a Type A variable on which the operator bases manual actions required for event mitigation for which no automatic controls are provided, it has been included in this LCO.

21. Dearees of Subcoolina Two channels of subcooling margin with inputs from RCS hot leg temperature (T ), core exit temperature, and RCSpressureareproviUed.

Multiple core exit temperatures are auctioneered with only the highest temperature being input to the monitor. A note has been added to indicate that the two channels of subcooling margin are backed up by either of two indications of subcooling margin based on similar inputs through the Safety Parameter Display System (SPDS). At least one SPDS channel must be available to provide this backup. With both SPDS channels IN0PERABLE, Condition C is applicable. This is considered necessary because the core exit thermocouple inputs to the subcooling margin monitors are not environmentally qualified.

The T inpds u g

the subcooling margin monitors and SPDS operate over a range of 120 to 920*F.

The core exit temperature inputs operate over a range of 150 to 2000*F and 150 (continued)

Crystal River Unit 3 B 3.3-138A Amendment No.

-. -.. - - -. - ~- -. - -.

PAM Instrumentation B 3.3.17 BASES 4

LC0_

21.

Dearees of Subcoolino (continued) to 2500*F for the subcooling margin monitors and SPDS, respectively.

RCS pressure inputs operate over a range of 200 to'2500 psig.

1 The subcooling margin monitors are used to verify the existence of, or to take actions to ensure the i

[

restoration of subcooling margin.

Specifically, a loss of adequate subcooling margin during a LOCA requires the operator to trip the reactor coolant pumps- (RCP's), to ensure high or low pressure injection, and raise the steam generator levels to the inadequate subcooling margin level.

Since degrees of subcooling is a Type A variable on which the operator bases manual actions required for event mitigation for t

i which no automatic controls are provided, it has been j

included in this LCO.

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22.

Emeraency Diesel Generator. kW Indication i

The Emergency Diesel Generator (EDG) provides standby (emergency) electrical power in the case of Loss of 2

Offsite Power (LOOP).

EDG kW indication is provided in the control room to monitor the operational status of the EDG.

EDG Power (kW) output indication is a type A variable 1

because EDG kW indication provides the control room j

operator EDG load management capabilities.

EDG load management enables the operator to base manual actions of load start and stop for event mitigation.

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(continued)

Crystal River Unit 3 B 3.3-138B Amendment No.

U.S. Nuclear Regulatory Commission Attachment F 3F0797-21 Page 1 of 14 i

ATTACHMENT F EVALUATION OF SAFETY PARAMETER DISPLAY l

SYSTEM AGAINST DESIGN CRITERIA OF REGULATORY GUIDE 1.97 FOR SUBCOOLING MARGIN MONITOR 4

1 Introduction J

Florida Power Corporation (FPC) is conducting an in-depth review of the Emergency Operating Procedure (EOP's). As part of this review, it was determined that the use of

" degrees of subcooling" in the EOP's was consistent with a Reg. Guide 1.97 Type A l

variable. FPC had previously specified " degrees of subcooling" as a Reg. Guide 1.97 Type B variable. Due to the way the variable is used in the EOP's, it should be listed j

2 as a Type A variable. Reg. Guide 1.97 specifies that instrumentation that monitors Type 1

A variables should be designed to Category 1 criteria. The design guidance in Reg.

Guide 1.97 for Category 1 devices is equivalent to safety related devices. Many of the i

components associated with the current subcooling indication are non-safety related.

An evaluation of the monitors' design was performed and the impact of the non-safety related devices was assessed. From this evaluation, it was concluded that the major impact on plant safety was from the core exit thermocouples which are input to the monitors. These thermocouples are not environmentally qualified.

As a compensatory measure, a modification was made to the plant Safety Parameter j

Display System (SPDS). Core exit thermocouple inputs were added to the pressures and temperatures already part of the system and programming was added to calculate i

subcooling margin. Alarms were included in the programming to alert operators when the subcooling margin does not meet minimum specified values. The SPDS displays are used to display the subcooling margin in several different formats including full screen when an alarm occurs. The control room operators have been trained on the use of the new system. All of the inputs used by SPDS to calculate subcooling margin are environmentally qualified. These interim enhancements were completed prior to startup from the Spring 1996 Refuel 10 outage and are described in our July 8,1996 letter to the NRC (3F0796-03).

As a follow-up to the interim SPDS modification, FPC has evaluated alternatives for a permanent modification to upgrade the subcooling margin monitors. Alternatives from "do nothing" to installation of a fully qualified, safety related system were evaluated and discussed. The emphasis of the decision making process was on making changes that produced the greatest safety enhancement for the resources expended. Input from the Operations Department weighed heavily in the decision-making process. The operators have been very favorably impressed with the man-machine interface provided by the SPDS display. As a result, for the permanent resolution to this issue, FPC has decided to modify the SPDS a second time to enhance its reliability and gain further compliance with the design recommendations of Regulatory Guide 1.97 for Category 1 instruments.

This report describes the current configuration of the SPDS relative to the Reg. Guide 1.97 design recommendations and also the planned enhancements. The enhancements will be instalbd during Refuel 11, currently scheduled for the fall of 1999.

U.S. Nuclear Regulatory Commission Attachment F 3F0797-21 Page 2 of 14 General The subcooling margin monitoring instrumentation at Crystal River 3 (CR-3)is comprised of two train, of instruments, each with inputs as described below. The instrument string configurations are depicted graphically on the three figures included at the end of this attachment.

f Each train of subcooling margin monitoring instrumentation has the following inputs:

Two Reactor Coolant System (RCS) hot leg temperature signals Two RCS pressure wide range signals Two RCS pressure narrow range signals Eight incore thermocouple signals l

Two of the hot leg temperature signals, one for each monitor, originate at RC-4A-TE1 l

(Train A) and RC-4B-TE4 (Train B) and go to separate remote shutdown auxiliary cabinets in the respective 4160 volt engineered safeguards switchgear room. These i

signals are fully qualified up to this point. From there they go into the non-nuclear 4

instrumentation (NNI) cabinets in the main control room and to the SPDS where the j

subcooling margin calculation is performed. This information is displayed on the SPDS displays mounted above the main control board.

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The other two hot leg temperature signals, one for each monitor, originate at RC-4B-TE1 (Train A) and RC-4A-TE4 (Train B) and go directly to separate NNI cabinets in the main control room and from there to the SPDS. The components associated with these l

signals, which are located in a harsh environment. are qualified to the requirements of

)

l 10 CFR 50.49.

Two wide range RCS pressure signals feed both trains of the subcooling margin monitor.

They originate at RC-3A-PT3 and RC-3B-PT3 and go to separate engineered safeguards l

l cabinets in the main control room. These signals are fully qualified up to this point.

From there they go to the SPDS where they are used in subcooling margin calculations when RCS pressure is greater than 600 psig.

l I

The "A" side narrow range RCS pressure signal feeds both trains of the subcooling margin monitor. It originates at RC-147-PT and goes to the "A" remote shutdown auxiliary cabinet in the "A" 4160 volt engineered safeguards switchgear room. This signal is fully qualified up to this point. From there it goes to the SPDS where it is used l

in subcooling margin calculations when RCS pressure is below 600 psig. As part of the enhancement to the subcooling margin instrumentation feeding the SPDS, a redundant, "B" side narrow range pressure signal will be provided to both trains of the subcooling margin monitor. This narrow range RCS pressure signal will originate at RC-148-PT, and go to the "B" remote shutdown auxiliary cabinet in the "B" 4160 volt engineered i

safeguards switchgear room. The signal will be fully qualified up to this point. From there it will go to the SPDS where it will be used in subcooling margin calculations when 4

i RCS pressure is below 600 psig.

l l

U.S. Nuclea: Regulatory Commission Attachment F 3F0797-21 Page 3 of 14 The sixteen incore thermocouple signals go into the Reactor Coolant Inventory and Tracking System (RCITS) cabinet in the control complex. At this point, eight of the sixteen signals go to one subcooling margin monitor and the other eight go to the other monitor. These signals are fully qualified up to this point. The signals go from the RCITS cabinets to the SPDS, eight to each train, where they are auctioneered in software so that the highest reading incore temperature is used in the subcooling margin calculation.

l

SUMMARY

OF PLANNED ENHANCEMENTS FOR SAFETY PARAMETER DISPLAY SYSTEM l

The following enhancements are planned for SPDS as delineated in Attachment 8 of FPC's letter dated September 27,1996 (3F0996-05, Reference 1):

j a.

The addition of a redundant "B" side narrow ranae Reactor Coolant System (RCS) pressure sianal.

l Currently, there is a question regarding the environmental qualification of the containment penetration associated with the "B" side narrow range RCS pressure signal.

The enhancements referred to in item 1, Environmental Qualification, " Degree of Compliance", will evaluate the penetration connector and upgrade it, if necessary, to meet the requirements of 10 CFR 50.49. The instrument string will then be used as a redundant RCS narrow range pressure input to SPDS. This signal will originate at RC-148-PT, and go to the "B" remote shutdown auxiliary cabinet in the "B" 4160 volt engineered safeguards switchgear room. The signal will be fully qualified up to this point. From there, it will go to the SPDS where it will be used in subcooling margin i

calculations when RCS pressure is below 600 psig.

l b.

Enhancement of the seismic desian of the existina SPDS system components, Currently, the SPDS computer equipment in the main control room is installed in open-air metal frame racks.

The enhancement discussed in item 1, Seismic Qualification,

" Degree of Compliance", refers to the modifications that will be made to the SPDS computer equipment racks to enhance their anchorage. All major components required i

for the operation of the system will be evaluated cnd physically restrained to prevent l

excessive movement and minimize the probability of damage during seismic events.

c.

Power Source l

Another important enhancement to the SPDS system is discussed in Item 3, Power Source, " Degree of Compliance" Currently, the power for the SPDS computer system is being supplied from a non-safety related inverter power source. Although this is a reliable source of power, the proposed changes will separate SPDS into a train "A" and train "B" powered, redundant system, from safety related inverters which are backed up by the station standby power sources (emergency diesel generators).

_.~.

)

U.S. Nuclear Regulatory Commission Attachment F 3F0797-21 Page 4 of 14 l

SPECIFIC CRITERIA FROM REGULATORY GUIDE 1.97 l

1.

Equipment Qualification Environmental Qualification RG Recommendation l

The instrumentation should be qualified in accordance with Regulatory Guide 1.89,

" Qualification of Class 1E Equipment for Nuclear Power Plants," and the methodology described in NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment."

l Instrumentation whose ranges are required to extend beyond those ranges calculated l

in the most severe design basis accident event for a given variable should be qualified using the guidance provided in paragraph 6.3.6 of ANS-4.5.

Qualification applies to the complete instrumentation channel from sersor to display where the display is a direct-indicating meter or recording device. If the instrumentation i

channel signal is to be used in a computer-based display, recording, or diagnostic program, qualification applies from the sensor up to and including the channel isolation j

device.

1 l

Dearee of Compliance - Full (Subsecuent to Modifications')

All equipment used for subcooling margin monitoring which is located in a harsh environment has been reviewed. The sensing devices for core exit temperature, wide l

range RCS pressure, RCS hot leg temperature and associated cables, connections, and building penetrations are all qualified to the requirements of 10 CFR 50.49. Presently the "A" side narrow range RCS pressure sensing device, cables, connections and penetrations are qualified to the requirements of 10 CFR 50.49.

Following enhancements to the SPDS subcooling margin instrumentation, the "B" side narrow range RCS pressure signal will be qualified to these requirements as well.

l 1

Seismic Qualification RG Recommendation The seismic portion of qualification should be in accordance with Regulatory Guide 1.100, " Seismic Qualification of Electric Equipment for Nuclear Power Plants."

l Instrumentation should continue to read within the required accuracy following, but not j

necessarily during, a safe shutdown earthquake.

Dearee of Comoliance - Partial l

The safety related portions of the system are all s?ismically designed in accordance with the CR-3 seismic licensing basis. This includes the RCITS cabinets, which are equally

U.S. Nuclear Regulatory Commission Attachment F 3F0797-21 Page 5 of 14 l

qualified.

The remote shutdown auxiliary cabinets and NNI cabinets have been l

evaluated as part of the resolution of USl A-46. The remote shutdown auxiliary cabinets were determined to be seismically adequate. Two of the four NNI cabinets were determined to be seismically adequate. The remaining two cabinets are not bolted together and were therefore classified as outliers. This deficiency will be corrected in l

accordance with FPC's USI A-46 outlier resolution program.

The "A" and "B" SPDS multiplexers and multiplexer servers are not seismically qualified but are housed in seismically qualified cabinets. The SPDS computers and displays are not seismically qualified but are either supported seismically, seismically restrained from becoming missiles, or housed in seismically qualified cabinets. All major components necessary for the operation of the system will be physically restrained to prevent motion and minimize the probability of damage during a seismic event.

Justification for Deviation The plant is located in a very low seismic risk region. The probability that a seismic event will occur and damage this equipment is very low. In the unlikely event damage does occur as a result of a seismic event, corrective actions will be initiated to restore the equipment in accordance with the actions required by the technical specifications.

2. Redundancy RG Recommendation No single failure within either the accident monitoring instrumentation, its auxiliary supporting features, or its power sources concurrent with the failures that are a condition or result of a specific accident should prevent the operators from being presented the information necessary for them to determine the safety status of the plant and to bring the plant to and maintain it in a safe condition following that accident. Where failure of one accident-monitoring channel results in information ambiguity (that is, the redundant displays disagree) that could lead operators to defeat or fail to accomplish a required safety function, additional information should be provided to allow the operators to deduce the actual conditions in the plant. This may be accomplished by providing additional independent channels of information of the same variable (addition of an identical channel) or by providing an independent channel to monitor a different variable that bears a known relationship to the multiple channels (addition of a diverse channel).

Redundant or diverse channels should be electrically independent and physically i

separated from each other and from equipment not classified important to safety in j

accordance with Regulatory Guide 1.75, " Physical Independence of Electric Systems,"

l up to and including any isolation device. Within each redundant division of a safety system, redundant monitoring channels are not needed except for steam generator level instrumentation in two-loop plants.

Dearee of Compliance - Full (Subseouent to Modification) in the final configuration, two independent channels of subcooling margin information are provided.

The two channels will be electrically independent.

No single active

. - ~

U.S. Nuclear Regulatory Commission Attachment F 3F0797-21 Page 6 of 14 l

l l

l component failure can result in the loss of both channels. In the event of failure of one of the channels resulting in ambiguous information, diverse subcooling margin l

information is available from the use of steam tables using fully qualified and redundant pressure and temperature indications.

i

3. Power Source RG Recommendation i

The instrumentation should be energized from station standby power sources as i

provided in Regulatory Guide 1.32, " Criteria for Safety-Related Electrical Power Systems for Nuclear Power Plants," and should be backed up by batteries where momentary interruption is not tolerable.

i Dearee of Compliance - Full (Subseauent to Modification) in the final configuration, the AC power for the redundant channels will be supplied from l

independent, safety related inverters backed up by the station standby power sources (emergency diesel generators). The DC for the inverters is from independent, safety t

l related station batteries. The AC power supply for the NNI cabinets is from these same i

sources and is also auctioneered with redundant supplies (VBDP-1 and VBDP-7) fed from non-safety related regulated instrument buses. One of these supplies (VBDP-7) is powered by an inverter with AC and DC power from non-safety related supplies.

l

4. Channel Availability RG Recommendation The instrumentation channel should be available prior to an accident except as provided l

in paragraph 4.11, " Exception," as defined in IEEE Std. 279-1971, " Criteria for Protection l

Systems for Nuclear Power Generating Stations," or as specified in the technical specifications.

Dearee of Compliance - Full l

FPC is proposing a technical specification change to add the subcooling margin monitors l

to Technical Specification LCO 3.3.17, Post-Accident Monitoring Instrumentation. The Required Actions for one or two channels inoperable are delineated in the technical specifications.

I

5. Quality Assurance RG Recommendation l

The recommendations of the following regulatory guides pertaining to quality assurance should be followed:

U.S. Nuclear Regulatory Commission Attachment F l

3F0797-21 Page 7 of 14 Regulatory Guide 1.28

" Quality Assurance Program Requirements (Design and Construction)"

Regulatory Guide 1.30

" Quality Assurance Requirements for the Installation, (Safety Guide 30) Inspection, and Testing of Instrumentation and Electric Equipment" Regulatory Guide 1.38

" Quality Assurance Requirements for Packing, Shipping, l

Receiving, Storage, and Handling of items for Water-Cooled l

Nuclear Power Plants" l

Regulatory Guide 1.58

" Qualification of Nuclear Power Plant inspection, t

l Examination, and Testing Personnel" l

l Regulatory Guide 1.64

" Quality Assurance Requirements for the Design of Nuclear Power Plants" l

Regulatory Guide 1.74

" Quality Assurance Terms and Definitions" l

Regulatory Guide 1.88

" Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records" Regulatory Guide 1.123

" Quality Assurance Requirements for Control of Procurement of items and Services for Nuclear Power Plants"

-Regulatory Guide 1.144

" Auditing of Quality Assurance Programs for Nuclear Power l

Plants" Regulatory Guide 1.146

" Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants" Reference to the above regulatory guides (except Regulatory Guides 1.30 and 1.38) is being made pending issuance of a revision to Regulatory Guide 1.28 that is under development (Task RS 002-5) and that will endorse ANSI /ASME NOA-1-1979, " Quality i

j Assurance Program Requirements for Nuclear Power Plants."

Dearee of Comraliance - Partial The safety related portions of the system have been qualified in accordance with the FPC Quality Assurance Program which complies with the requirements of 10 CFR 50, Appendix B and has been approved by the NRC. FPC has committed to all of the Regulatory Guides listed above except for Regulatory Guide 1.28. FPC's commitment to this guidance is documented and clarified in Table 1-3 of the CR-3 Final Safety Analysis Report (FSA8). Commitments related to Regulatory Guide 1.28 are contained in FPC's commitment to Regulatory Guide 1.33 and through FPC's compliance with 10 4

l CFR 50, Appendix B. The recommendation is fully met for these components.

The two RCS pressure channels (narrow range and wide range) are safety related up through an isolation device in the engineered safeguards cabinets. One of the RCS hot

l l

l U.S. Nuclear Regulatory Commission Attachment F 3F0797-21 Page 8 of 14 leg temperature channels is safety related up through an isolation device in a remote shutdown auxiliary cabinet. All 15 incore temperature channels (8 per SPDS train) are safety related through the Reactor Coolant Inventory & Tracking (RCITS) cabinets.

The remainder of the components are non-safety related, but satisfy the quality.

assurance recommendations of Reg. Guide 1.97 for Category 2 variables.

Justification for Deviation The human factors advantages of this personal computer-based system more than outweigh the loss of quality resulting from a lack of 10 CFR 50, Appendix B qualification for all equipment. The equipment is continually in service and monitored for proper operation by control room operators. Any malfunction will be promptly detected. There will be a high degree of confidence that the system will function when called upon to do so. Since the SPDS functions continuously failures will be readily evident and corrective actions can be initiated expeditiously.

6. D.'splav s. ' Recording RG Flecommendation Continuous real-time display should be provided. The indication may be on a dial, digital display, CRT, or stripchart recorder. Recording of instrumentation readout information should be provided for at least one redundant channel.

If direct and immediate trend or transient information is essential for operator information or action, the recording should be continuously available on redundant dedicated recorders. Otherwise, it may be continuously updated, stored in computer memory, and displayed on demand.

Intermittent displays such as data loggers and scanning recorders may be used if no significant transient response information is likely to be lost by such devices.

Dearee of Compliance - Full (Subseauent to Modification)

Continuous real-time display is provided on two redundant SPDS displays. Recording of subcooling margin is done at one second intervals and stored in computer memory.

In the final configuration, this information will have the capability to be displayed in trend format.

j

7. Range RG Recommendation if two or more instruments are needed to cover a particular range, overlapping of instrument span should be provided. If the required range of monitoring instrumentation results in a loss of instrumentation sensitivity in the normal operating range separate instruments should be used.

I l

l

.U.S. Nuclear Regulatory Commission Attachment F l

3F0797-21 Page 9 of 14 l

l i

Dearee of Compliance - Full I

Reg. Guide 1.97, Table 3, recommends a range of 200 F subcooling to 35 F superheat. The SPDS range exceeds this recommendation.

8. Equipment identification RG Recommendation Types A, B, and C inttrum.,ts designated as Categories 1 and 2 should be specifically identified with a comn!on detignation on the control panels so that the operator can easily discern that they cre lr. tended for use under accident conditions.

l Dearee of Compliance - Full (Subseauent to Modificatic.is)

)

The SPDS displays will be labeled as Reg. Guide 1.97 instruments on the main control board. This will be done during Refuel 11 outage.

j

9. Interfaces RG Recommendation I

The transmist.!on of signals for other use should be through isolation devices that are

]

designed as part of the monitoring instrumentation and that meet the provisions of this document.

l Dearee of Cc1noliance - None j

Where signals are fed from safety related systems for use by the subcooling margin monitors, isolation devices are provided to protect the safety related systems from faults in the SPD.c No isolation devices are provicad to protect signals to the SPDS from faults in otner safety or non-safety related components.

Justification for Dev:ation l

S;nce this system 4. in operation and displays continuously, any malfunction ;aused by components conr cted to the system will be promptly detected by the plant operators.

Prompt correctiv actions can be taken to restore the equipment.

Failure of l

interconnected ce nponents during an event has a very low probability. Also, since the i

system is redur.J.nt, failure of a single channel will not cause a loss of function.

l

1 U.S. Nuclear Regulatory Commission Attachment F j

3F0797-21 Page 10 of 14 i

10. Servicing, Testing, and Calibration RG Recommendation Servicing, testing, and calibration programs should be specified to maintain the capability of the monitoring instrumentation. If the required interval between testing is less than i

the normal time intervai between plant shutdowns, a capability for testing during power operation shouH be provided.

j Whenever means for removing char.nels from service are included in the design, the j

design should facilitate administrative control of the access to such removal means.

The design should facilitate administrative control of access to all setpoint adjustments, module calibration adjustments, and test points.

Periodic checking, testing, calibration, and calibration verification should be in accordance with the applicable portions of Regulatory Guide 1.118, " Periodic Testing of E:ectric Power and Protection Systems," pertaining to testing of instrument channels.

(Note: Response time testing not usually needed.)

The location of the isolation device should be such that it would be accessible for maintenance during accident conditions.

Dearee of Comoliance - Full The subcooling margin monitors' design and FPC's program for calibrating and maintaining them complies with all of the above recommendations. In addition, the post-accident monitoring instrumentation technical specification Surveillance Requirements require a CHANNEL CHECK every 31 days and a CHANNEL CALIBRATION every 24 4

months.

11. Human Factors RG Recommendation The instrumentation should be designed to facilitate the recognition, location, replacement, repair, or adjustment of malfunctioning components or modules.

ihe monitoring instrumentation design should minimize the development of conditions that would cause meters, annunciators, recorders, alarms, etc., to give anomalous indications potentially confusing to the operator. Human factors analysis should be used in determining type and location of displays. To the extent practicable, the same instruments should be used for accident monitoring as are used for the normal i

operations of the plant to enable the operators to use, during accident situations, instruments with which they are most familiar.

U.S. Nuclear Regulatory Commission Attachment F 3F0797-21 Page 11 of 14 Dearee of Comoliance - Full Subcooling margin is displayed on the SPDS displays mounted above the main control board. Real time indication of subcooling margin is the most desirable method of monitoring the variable from a human factors perspective. Operators will be trained to use the SPDS displays as the primary means of determining subcooling margin.

Providing subcooling margin data on the SPDS displays provides unique flexibility in the presentation of information to the operator:

o Subcooling margin is automatically calculated from incore temperatures when SPDS determines no reactor coolant pumps are running, o

The SPDS display indicates if subcooling margin is being calculated from hot leg temperature or incore temperature.

o in the event of an alarm, subcooling margin information is enlarged to cover the entire display. The display colors change to white characters on a red background with the characters being as large as possible and still display the margin.

o A timer starts, in the event of an alarm, to aid the operators in determining how long adequate subcooling margin has been lost. This is useful in determining the need to trip all reactor coolant pumps within two minutes of the loss of adequate subcooling margin.

12. Direct Measurement RG Recommendation To the extent practicable, monitoring instrumentation inputs should be from sensors that directly measure the desired variables. An indirect measurement should be made only when it can be shown by analysis to provide unambiguous information.

Dearee of Compliance - Full Subcooling margin is not directly measurable. However the temperature and pressum that determine the degree of subcooling are measured directly.

SPDS TSAT Train "A"

I:e c toe !u.lg Cun t i o! Co m plo lois it El Contal Complex 121 ft El Control Cotuplex 145 f t.

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T) -

e O@

. _, _ }.;

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l NNI Cab 4 Q NNI Cab 4 [i NNI Cah 5 Q g,

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+ 24 vde Bus f

^ M

-l U.S. Nuclear Regulatory Commission Attachment F 3F0797-21 Page 14 of 14 1

4

}

g4A Incores

1. 4B Incores A" 1E Vital Bus "B" 1E Vital Bus Control l

Room

_L

1. 2A Primary

1. 3A Backup

1. 2B Primary

1. 3B Backup l

" A* 1E Vital Bus "A" 1E Vital Bus "B" 1E Vital Bus "B" 1E Vital Bus I

Network "A" Ethernet Hub connection "B" Ethernet Hub

" A" 1E Vital Bus "B" 1E Vital Bus i

" A" Mux Server "B" Mux Server

" A" 1E Vital Bus "B" 1E Vital Bus Room "A" Intelligent hoYneclon "B" Intelligent 3

Ethernet Hub Ethernet Hub

" A" 1E Vital Bus "B" 1E Vital Bus Main

" A" SPDS Display "B" SPDS Display Control

" A" 1E Vital Bus "B" 1E Vital Bus Board SPDS TS AT Configuration d

f

U.S. Nuclear Regulatory Commission Attachment G 3F0797-21 Page 1 of 7 ATTACHMENT G EVALUATION OF LOW PRESSURE INJECTION FLOW INSlRUMENTATION AGAINST DESIGN CRITERIA OF REGULATORY GUlOE 1.97 INTRODUCTION Florida Power Corporation (FPC) recently determined that the use in emergency operating procedures of low pressure injection flow instrumentation was consistent with 1

a Reg. Guide 1.97 Type A variable. A Type A variable is one of "those variables that provide primary information needed to permit the control room operating personnel to take the specified manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis events." The low pressure injection flow instrumentation is used to verify flow to the core prior to securing high pressure injection as part of the actions to assure long term post-accident core cooling.

The instrumentation being used for this purpose meets the design criteria for Reg. Guide 1.97 Category 1 instruments with the exception of the recording function. A plant change will be implemented during the current outage to provide the recording function.

SPECIFIC CRITERIA FROM REGULATORY GUIDE 1.97

1. Equipment Qualification Environmental Qualification RG Recommendation The instrumentation should be qualified in accordance with Regulatory Guide 1.89,

" Qualification of Class 1E Equipment for Nuclear Power Plants," and the methodology described in NUREG-0588, " Interim Staff Position on Eavironmental Qualification of Safety-Related Electrical Equipment."

instrumentation whose ranges are required to extend beyond those ranges calculated in the most severe design basis accident event for a given variable should be qualified using the guidance provided in paragraph 6.3.6 of ANS-4.5.

Qualification applies to the complete instrumentation channel from sensor to display where the display is a direct-indicating meter or recording device. If the instrumentation channel signal is to be used in a computer-based display, recording, or diagnostic p ogram, qualification applies from the sensor up to and including the channel isolation device.

U.S. Nuclear Regulatory Commission Attachment G 3F0797-21 Page 2 of 7

_Dearoe of Comoliance - Full All of the components used for low pressure injection flow indication which are credited as Category 1 and which are located in a harsh environment are qualified to the requirements of 10 CFR 50.49.

Seismic Qualification RG Recommendation The seismic portion of qualification should be in accordance with Regulatory Guide 1.100, " Seismic Qualification of Electric Equipment for Nuclear Power Plants."

Instrumentation should continue to read within the required accuracy following, but not necessarily during, a safe shutdown earthquake.

Dearee of Compliance - Full Allinstrumentation to be used for low pressure injection flow indication for Post Accident Monitoring is Category 1 and qualified to the requirements of IEEE 344-1975 and Regulatory Guide 1.100, Revision 1.

2. Redundancy RG Recommendation No single failure within either the accident monitoring instrumentation, its auxiliary supporting features, or its power sources concurrent with the failures that are a condition or result of a specific accident should prevent the operators from being presented the information necessary for them to determine the safety status of the plant and to bring the plant to and maintain it in a safe condition following that accident. Where failure of one accident-monitoring channel results in information ambiguity (that is, the redundant displays disagree) that could lead operators to defeat or fail to accomplish a required safety function, additional information should be provided to allow the operators to deduce the actual conditions in the plant. This may be accomplished by providing additional independent channels of information of the same variable (addition of an identical channel) or by providing an independent channel to monitor a different variable that bears a known relationship to the multiple channels (addition of a diverse channel).

Redundant or diverse channels should be electrically independent and physically separated from each other and from equipment not classified important to safety in accordance with Regulatory Guide 1.75,

• Physical Independence of Electric Systems,"

up to and including any isolation device. Within each redundant division of a safety system, redundant monitoring channels are not needed except for steam generator level instrumentation in two-loop plants.

Dearee of Compliance - Full 1

The low pressure injection system is comprised of redundant divisions each with a single qualified channel of flow instrumentation.

Per Regulatory Guide 1.97, Criterion 2,

U.S. Nuclear Regulatory Commission Attachment G 3F0797-21 Page 3 of 7 redundant monitoring channels for low pressure injection flow are not required (the last sentence in the Redundancy criterion for Category 1 in the Regulatory Guide 1.97 is taken as justification that redundant monitoring channels are not needed). Non-safety related circuits are isolated in accordance with Section 4.7 of IEEE 279, " Criteria for Nuclear Power Plant Protection Systems," dated August,1968. This is the design basis standard for the protection systems at CR-3. The CR-3 electrical system design is not required to meet the recommendations of Regulatory Guide 1.75.

3. Power Source RG Recommendation The instrumentation should be energized from station standby power sources as provided in Regulatory Guide 1.32," Criteria for Safety-Related Electrical Power Systems for Nuclear Power Plants," and should be backed up by batteries where momentary interruption is not tolerable.

Dearee of Compliance - Full All of the components used for low pressure injection flow indication which are credited as Category 1 are powered from vital buses which are backed up by the station batteries and the emergency diesel generators. The design of the electrical systems at CR-3 satisfies the IEEE 308 proposed criteria for Class 1E electrical systems, dated June, 1969. In addition, the design is in compliance with the intent of 10 CFR 50, Appendix A, Criterion 17.

The CR-3 electrical system design is not required to meet the recommendations of Regulatory Guide 1.32.

4. Channel Availability RG Recommendation The instrumentation channel should be available prior to an accident except as provided in paragraph 4.11, " Exception," as defined in IEEE Std. 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations," or as specified in the technical specifications.

Dearee of Compliance - Full FPC is proposing a Technical Specification change to add the low pressure injection flow instrumentation to Technical Specification LCO 3.3.17, Post-Accident Monitoring Instrumentation.

The Required Actions for one or two channels inoperable are delineated in the Technical Specifications.

U.S. Nuclear Regulatory Commission Attachment G

)

3F0797-21 Page 4 of 7

5. Quality Assurance RG Recommendation The recommendations of the following regulatory guides pertaining to quality assurance should be followed:

Regulatory Guide 1.28 "O.uality Assurance Program Requirements (Design and Construction)"

Regulatory Guide 1.30

" Quality Assurance Requirements for the Installation, (Safety Guide 30) Inspection, and Testing of Instrumentation and Electric Equipment" Regulatory Guide 1.38

" Quality Assurance Requirements for Packing, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants" Regulatory Guide 1.58

" Qualification of Nuclear Power Plant inspection, Examination, and Testing Personnel" Regulatory Guide 1.64

" Quality Assurance Requirements for the Design of Nuclear Power Plants" Regulatory Guide 1.74

" Quality Assurance Terms and Definitions" Regulatory Guide 1.88

" Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records" Regulatory Guide 1.123

" Quality Assurance Requirements for Control of Procurement of items and Services for Nuclear Power Plants" Regulatory Guide 1.144

" Auditing of Quality Assurance Programs for Nuclear Power Plants" Regulatory Guide 1.146

" Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants" Reference to the above regulatory guides (except Regulatory Guides 1.30 and 1.38) is being made pending issuance of a revision to Regulatory Guide 1.28 that is under development (Task RS 002-5) and that will endorse ANSI /ASME NOA-1-1979, " Quality Assurance Program Requirements for Nuclear Power Plants."

t Dearee of Compliance - Full All of the components used for low pressure injection flow indication which are credited as Category 1 have been qualified in accordance with the FPC Quality Assurance 1

I Program which complies with the requirements of 10 CFR 50, Appendix B and has been approved by the NRC. FPC has committed to all of the Regulatory Guides listed above

L l

U.S. Nuclear Regulatory Commission Attachment G 3F0797-21 Page 5 of 7 except for Regulatory Guide 1.28. FPC's commitment to this guidance is documented and clarified in Table 1-3 of the CR-3 Final Safety Analysis Report (FSAR).

Commitments related to Regulatory Guide 1.28 are contained in FPC's commitment to Regulatory Guide 1.33 and through FPC's compliance with 10 CFR 50, Appendix B.

6. Display and Recording RG Recommendation Continuous real-time display should be provided. The indication may be on a dial, digital display, CRT, or stripchart recorder. Recording of instrumentation readout information should be provided for at least one redundant channel.

If direct and immediate trend or transient information is essential for operator information or action, the recording should be continuously available on redundant dedicated recorders. Otherwise, it may be continuously updated, stored in computer memory, and displayed on demand.

Intermittent displays such as data loggers and scanning recorders may be used if no significant transient response information is likely to be lost by such devices.

Dearee of Compliance - Full (Subseauent to Modification)

Continuous real-time display is provided. FPC willinstall a modification to add inputs to the plant computer to provide the recording function. This modification will be installed during the current outage. It should be noted that the plant computer is not powered by the emergency diesel generator-backed vital busses. It is, however, powered from a highly reliable non-safety related battery-backed power source.

7. Range RG Recommendation if two or more instruments are needed to cover a particular range, overlapping of instrument span should be provided. If the required range of monitoring instrumentation results in a loss of instrumentation sensitivity in the normal operating range separate instruments should be used.

Dearee of Compliance - Full The range of the low pressure injection flow instrumentation (0 - 5000 gpm) has been evaluated against its required post-accident function and determined to be adequate.

I G

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l U.S. Nuclear Regulatory Commission Attachment G 3F0797-21 Page 6 of 7 l

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8. Equipment identification i

i l

RG Recommendation Types A, B, and C instruments designated as Categories 1 and 2 should be specifically identified with a common designation on the control panels so that the operator can easily discern that they are intended for use under accident conditions.

Dearee of Comoliance - Full l

All of the indicators used for low pressure injection flow indication which are credited as Category 1 have been labeled on the main control board so the operator can easily discern that they are Reg. Guide 1.97 qualified instruments.

i

9. Interfaces l

RG Recommendation The transmission of signals for other use should be through isolation devices that are designed as part of the monitoring instrumentation and that meet the provisions of this document.

Dearee of Compliance - Full i

As stated previously, non-safety related circuits are isolated in accordance with Section 4.7 of IEEE 279, " Criteria for Nuclear Power Plant Protection Systems," dated August, 1968. This is the design basis standard for the safety related protection systems at CR-3.

10. Servicing, Testing, and Calibration RG Recommendation Servicing, testing, and calibration programs should be specified to maintain the capability of the monitoring instrumentation. If the required interval between testing is less than the normal time interval between plant shutdowns, a capability for testing during power operation should be provided. Whenever means for removing channels from service are included in the design, the design should facilitate administrative control of the access l

to such removal means.

The design should facilitate administrathe control of access to all setpoint adjustments, I

module calibration adjustments, and test points.

l l

Periodic checking, testing, calibration, and calibration verification should be in accordance with the applicable portions of Regulatory Guide 1.118," Periodic Testing of Electric Power and Protection Systems," pertaining to testing of instrument channels.

(Note: Response time testing not usually needed.)

U.S. Nuclear Regulatory Commission Attachment G 3F0797-21 Page 7 of 7 The location of the isolation device should be such that it would be accessible for maintenance during accident conditions.

Dearee of Compliance - Full The design of the components used for low pressure injection flow indication which are credited as Category 1 and FPC's program for calibrating and maintaining them complies with all of the above recommendations.

In addition, the post-accident monitoring instrumentation technical specification Surveillance Requirements require a CHANNEL CHECK every 31 days and a CHANNEL CALIBRATION every 24 months.

11. Human Factors RG Recommendation The instrumentation should be designed to facilitate the recognition, location, replacement, repair, or adjustment of malfunctioning components or modules.

The monitoring instrumentation design should minimize the development of conditions that would cause meters, annunciators, recorders, alarms, etc., to give anornalous indications poten+ially confusing to the operator. Human factors analysis should be used in determining type and location of displays. To the extent practicable, the same instruments should be used for accident monitoring as are used for the normal operations of the plant to enable the operators to use, during accident situations, instruments with which they are most familiar.

Dearee of Compliance - Partial The meters on the main control board used for Regulatory Guide 1.97 compliance are not the normal use meters. A proposed project to upgrade the normal use meters is being reviewed for implementation during Refuel 11. Implementation of that project would bring us into full compliance.

12. Direct Measurement RG Recommendation To the extent practicable, monitoring instrumentation inputs should be from sensors that directly measure the desired variables. An indirect measurement should be made only when it can be shown by analysis to provide unambiguous information.

Dearee of Compliance - Full Low pressure injection flow is measured directly by a ficw element and differential pressure transmitter which provides a signal to operate the indicator.

d U.S. Nuclear Regulatory Commission Attachment H 3F0797-21 Page 1 of 8 i

ATTACHMENT H EVALUATION OF EMERGENCY DIESEL GENERATOR KILOWATT METERS AGAINST DESIGN CRITERIA OF REGULATORY GUIDE 1.97 INTRODUCTION 4

Florida Power Corporation (FPC) is implementing plant modifications, procedure revisions, and technical specification revisions at Crystal River Unit 3 as part of a strategy to ensure that the emergency diesel generators are able to handle post-accident loads. This strategy involves adding loads to or removing loads from the EDGs as 4

needed to ensure that actions required by safety systems post-accident are l

accomplished without exceeding the capacity & the EDGs. The operators take these actions of EDG load management based on monitoring the instrumentation of the kilowatt (kW) load on the EDGs (kW meters). This use of the kW meters is consistent j

with a Regulatory Guide 1.97 Type A variable. A type A variable is one of "those variables that provide primary information needed to permit the control room personnel i

to take the specified manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design ba5l5 events".

)

The instrumentation being used for this purpose meets the design criteria for Reg. Guide i

1.97 Category 1 instruments with the exception of the recording function. A plant change will be implemented prior to restart from this outage that will add the required recording i

function.

SPECIFIC CRITERIA FROM REGULATORY GUIDE 1.97 4

i i

1. Equipment Qualification Environmental Qualification RG Recommendation The instrumentation should be qualified in accordance with Regulatory Guide 1.89,

" Qualification of Class 1E Equipment for Nuclear Power Plants," and the methodology described in NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment."

Instrumentation whose ranges are required to extend beyond those ranges calculated in the most severe design basis accident event for a given variable should be qualified using the guidance provided in paragraph 6.3.6 of ANS-4.5.

Qualification applies to the complete instrumentation channel from sensor to display where the display is a direct-indicating meter or recording device. If the instrumentation channel signal is to be used in a computer-based display, recording, or diagnostic program, qualification applies from the sensor up to and including the channel isolation device.

U.S. Nuclear Regulatory Commission Attachment H 3F0797-21 Page 2 of 8 Dearee of Compliance - Full None of the components used for EDG kW load indications are located in a harsh post-accident environment. The new kW meters are located in the Main Control Room, rated as a mild environment (EQ).

Seismic Qualification RG Recommendation The seismic prtion of qualification should be in accordance with Regulatory Guide 1.100, " Seismic Qualification of Electric Equipment for Nuclear Power Plants."

Instrumentation should continue to read within the required occuracy following, but not necessarily during, a safe shutdown earthquake.

Dearee of Comoliance - Full Allinstrumentation to be used for EDG kW load indication for Post Accident Monitonng is Category 1 and qualified to IEEE 344-1975 and Reg. Guide 1.100, Revision 1.

2. Redundancy RG Recommendation No single failure within either the accident monitoring instrumentation, its auxiliary supportil.g features, or its power sources concurrent with the failures that are a condition or result of a specific accident should prevent the operators from being presented the information necessary for them to determine the safety status of the plant and to bring the plant to and maintain it in a safe condition following that accident. Where failure of one accident-monitoring channel results in information ambiguity (that is, the redundant displays disagree) that could lead operators to defeat or fail to accomplish a required safety function, additional information should be provided to allow the operators to deduce the actual conditions in the plant. This may be accomplished by providing additional independent channels of information of the same variable (addition of an identical channel) or by providing an int.spendent channel to monitor a different variable that bears a known relationship to the multiple channels (addition of a diverse channel). Redundant or diverse channels should be electrically independent and physically separated from each other and from equipment not classified important to safety in accordance with Regulatory Guide 1.75, " Physical independence l

of Electric Systems," up to and including any isolation device. Within each redundant division of a safety system, redundant monitoring channels are not needed except for steam generator level instrumentation in two-loop plants.

Decree of Compliance - Full Separation distances for affected equipment will be in accordance with the CR-3 Electrical Design Criteria Electrical Circuit Physical Separation requirements. EDG load

U.S. Nuclear Regulatory Commission Attachment H 3F0797-21 Page 3 of 8 instrumentation is comprised of one kW meter per train. Per Regulatory Guide 1.97, Criterion 2, last sentence, redundant monitoring channels for EDG kW indication are not required. Non-safety related circuits are isolated in accordance with Section 4.7 of IEEE 279, " Criteria for Nuclear Power Plant Protection Systems," dated August,1968. This is the design basis standard for the protection systems at CR-3.

3. Power Source RG Recommendation The instrumentation should be energized from station standby power sources as provided in Regulatory Guide 1.32, " Criteria for Safety-Related Electrical Power Systems for Nuclear Power Plants," and should be backed up by batteries where momentary interruption is not tolerable.

Dearee of Compliance - Full The combined kW/kVAR transducers and kW indicators will be externally powered using dedicated circuits. The kW/kVAR transducers and kW indicators for the EDG-1 A will be powered from VBDP-5, which receives onsite backup power from EDG-1A.

The kW/kVAR transducers and kWindicators for the EDG-1B will be powered from VBDP-6, which receives onsite backup power from EDG-1B. The VBDPs are considered to be i

highly reliable sources of power. The design of the electrical systems at CR-3 satisfies the IEEE 308 proposed criteria for Class 1E electrical systems, dated June,1969. In addition, the design is in compliance with the intent of 10 CFR 50, Appendix A, Criterion l

17.

I

4. Channel Availability l

RG Recommendation The instrumentation channel should be available prior to an accident except as provided in paragraph 4.11, " Exception," as defined in IEEE Std. 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations," or as specified in the technical specifications.

Dearee of Compliance - Full FPC is proposing a Technical Specification change to add the EDG load indication to Technical Specification LCO 3.3.17, Post-Accident Monitoring Instrumentation. The Required Actions for one or two channels inoperable are delineated in the Technical j

Specifications.

U.S. Nuclear Regulatory Commission Attachment H 3F0797-21 Page 4 of 8

5. Quality Assurance RG Recommendation The recommendations of the following regulatory guides pertaining to quality assurance should be followed:

Regulatory Guide 1.28

" Quality Assurance Program Requirements (Design and Construction)"

Regulatory Guide 1.30

" Quality Assurance Requirements for the Installation, (Safety Guide 30)lnspection, and Testing of Instrumentation and Electric Equipment" l

Regulatory Guide 1.38

" Quality Assurance Requirements for Packing, Shipping, Receiving, Storage, and Handling of items for Water-Cooled Nuclear Power Plants" Regulatory Guide 1.58

" Qualification of Nuclear Power Plant inspection, Examination, and Testing Personnel" Regulatory Guide 1.64

" Quality Assurance Requirements for the Design of Nuclear Power Plants" Regulatory Guide 1.74

" Quality Assurance Terms and Definitions" Regulatory Guide 1.88

" Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records" Regulatory Guide 1.123

" Quality Assurance Requirements for Control of Procurement of items and Services for Nuclear Power Plants" Regulatory Guide 1.144

" Auditing of Quality Assurance Programs for Nuclear Power Plants" Regulatory Guide 1,146

" Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants" Reference to the abova regulatory guides (except Regulatory Guides 1.30 and 1.38) is being made pending issuance of a revision to Regulatory Guide 1.28 that is under development (Task P.S 002-5) and that will endorse ANSI /ASME NOA-1-1979, " Quality Assurance Program Requirements for Nuclear Power Plants."

Dearee of Compliance - Full All of the components used for EDG load indication which are credited as Category 1 have been qualified in accordance with the FPC Quality Assurance Program which complies with the requirements of 10 CFR 50, Appendix B and has been approved by the NRC. FPC has committed to all of the Regulatory Guides listed above except for

... - __- ~

i U.S. Nuclear Regulatory Commission Attachment H 3F0797-21 Page 5 of 8 Regulatory Guide 1.28. FPC's commitment to this guidance is documented and clarified in Tabic 1-3 of the CR-3 Final Safety Analysis Report (FSAR). Commitments related to Regulatory Guide 1.28 are contained in FPC's commitment to Regulatory Guide 1.33 and through FPC's compliance with 10 CFR 50, Appendix B.

i

6. Display and Recording RG Recommendation Continuous real-time display should be provided. The indication may be on a dial, digital display, CRT, or stripchart recorder. Recording of instrumentation readout information should be provided for at least one redundant channel.

If direct and immediate trend or transient information is essential for operator information or action, the recording should be continuously available on redundant dedicated l

recorders. Otherwise, it may be continuously updated, stored in computer memory, and i

displayed on demand.

Intermittent displays such as data loggers and scanning l

recorders may be used if no significant transient response information is likely to be lost by such devices.

l Dearee of Compliance - Full (Subseauent to Modification)

Continuous real-time display is provided. The new kW indicators are 0-4,000 kW digital display. FPC will install a modification to add inputs to the Recall /SPDS computer to provide the recording function. This modification will be installed before restart from the current outage.

7. Range RG Recommendation if two or more instruments are needed to cover a particular range, overlapping of instrument span should be provided, if the required range of monitoring instrumentation results in a loss of instrumentation sensitivity in the normal operating range separate instruments should be used.

Dearee of Compliance - Full The new EDG kWindicators have a range of 0-4,000 kW. This range adequately covers l

the range of EDG loads expected for post-accident conditions.

I

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l U.S. Nuclear Regulatory Commission Attachment H 3F0797-21 Page 6 of 8 l

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8. Equipment identification j

RG Recommendation Types A, B, and C instruments designated as Categories 1 and 2 should be specifically identified with a common designation on the control panels so that the operator can easily discern that they are intended for use under accident conditions.

Dearee of Compliance - Full (Subseauent to Modifications)

Both of the indicators used for EDG load indication which are credited as Category 1 will be labeled on the main control board so the operator can easily discern that they are Reg. Guide 1.97 qualified instruments. This will be done prior to restart from the current outage.

9. Interfaces 1

RG Recommendation The transmission of signals for other use should be through isolation devices that are designed as part of the monitoring instrumentation and that meet the provisions of this document.

Regree of Compliance - Full i

Non-safety related circuits are isolated in accordance with Section 4.7 of IEEE 279,

" Criteria for Nuclear Power Plant Protection Systems," dated August,1968. This is the design basis standard for the safety related protection systems at CR-3.

10. Servicing, Testing, and Calibration RG Recommendation Servicing, testing, and calibration programs should be specified to maintain the capability of the monitoring instrumentation. If the required interval between testing is less than the normal time interval between plant shutdowns, a capability for testing during power operation should be provided. Whenever means for removing channels from service are included in the design, the design should facilitate administrative control of the a' :ess to such removal means.

The design should facilitate administrative control of access to all setpoint adjustments, module calibration adjustments, and test points.

Periodic checking, testing, calibration, and calibration verification should be in i

accordance with the applicable portions of Regulatory Guide 1.118, " Periodic Testing of Electric Power and Protection Systems," pertaining to testing of instrument channels.

(Note: Response time testing not usually needed.)

l

U.S. Nuclear Regulatory Commission Attachment H 3F0797-21 Page 7 of 8 l

The location of the isolation device should be such that it would be accessible for maintenance during accident conditions.

Dearee of Compliance - Full The EDG kW transducers and indicators are fully accessible for periodic inspections and calibration.

The EDG kW indicators have a removable plug on the rear and the indicators can be easily removed from the front of the MCB. The indicators were ordered with a programming and calibration kit for easy setup. The design of the components used for EDG load indication which are credited as Category 1 and FPC's program for calibrating and maintaining them comp!ies with all of the above recommendations, in addition, the post-accident monitoring instrumentation technical specification Surveillance Requirements require a CHANNEL CHECK every 31 days and a CHANNEL CAllBRATION every 24 months.

11. Human Factors RG Recommendation The instrumentation should be designed to facilitate the recognition, location, replacement, repair, or adjustment of malfunctioning components or modules.

The monitoring instrumentation design should minimize the development of conditions that would cause meters, annunciators, recorders, alarms, etc., to give anomalous indications potentially confusing to the operator. Human factors analysis should be used in determining type and location of displays. To the extent practicable, the same instruments should be used for accident monitoring as are used for the normal operations of the plant to enable the operators to use, during accident situations, instruments with which they are most familiar.

Dearee of Compliance - Full The EDG kW indicators meet the human factors guidelines for indicators as addressed in SP-5145, a study performed for FPC by Gilbert / Commonwealth entitled " Human Factors Design Conventions for the Contrt Room Specification and Criteria" The report provided specific human factors criteria for panel layout, labels and location aids, and annunciators. Important considerations in the report included ensuring the operator can read a display while operating related controls, and labeling is clear and consistent.

Operations personnel choose the colors to be used for the indicators (red for EGDG-1A and green for EGDG-1B). The availability of power to the indicators will be visible on the indicators. Each VBDP breaker has circuit status indication lights to inform the operator that the circuits are energized. EDG load indications instrumentation design and installation, including equipment and displays, meet RG 1.97 human factor guidelines. The EDG kW indicators have a digital display of 0 to 4,000 kW. In addition to the digital display the EDG kW indicators have a bar graph display for improved human factors.

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U.S. Nuclear Regulatory Commission Attachment H 3

l 3F0797-21 Page 8 of 8 i

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\\

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12. Direct Measurement 1

)

RG Recommendation 1

i To the extent practicable, monitoring instrumentation inputs should be from sensors that directly measure the desired variables. An indirect measurement should be made only when it can be shown by analysis to provide unambiguous information.

Dearee of Compliance - Full EDG load indication is measured by the direct correlation from CT and PT ratios. The EDG cts and pts are attached directly to the EDG power bus.

l 1

l U.S. Nuclear Regulatory Commission Attachment I 3F0797-21 Page 1 of S l

ATTACHMENT I RESPONSE TO NRC REQUEST FOR INFORMATION FPC - NRC TELEPHONE CONFERENCE OF DECEMBER 10,1996 Item 1:

EVALUATION OF LOW PRESSURE INJECTION FLOWINSTRUM ENTATION AGAINST DESIGN CRITERIA OF REGULATORY GUIDE 1.97 FOR USE IN POST-ACCIDENT MONITORING NRC Reauest/ Concern Provide a description of how the LPI flow instrumentation met the design criteria for Category 1 instruments per Regulatory Guide 1.97. A similar description was provided for the SPDS in Attachment 8 of the September 27,1996 letter._

FPC Respon_se See the evaluations of the subcooling margin monitors, LPI flow instruments, and kW indication provided in Attachments F, G, and H, respectively.

Item 2:

AVAILABILITY OF WlDE RANGE RCS PRESSURE SIGNALS NRC Ouestion/ Concern Clarify the signal availability for RCS wide range pressure to the redundant SPDS channels. (The 9/27/96 letter, Attachment 8, page 2, paragraph 5, states "Two wide range RCS pressure signals feed both trains of the subcooling margin monitor." FPC was to clarify that this really means "Two wide range RCS pressure signals each feed both trains of the subcooling margin monitor."

FPC Response Each of the two wide range Reactor Coolant System (RCS) pressure signals (RC-3A-PT3 & RC-3B-PT3) feed both trains of the subcooling margin monitor. Both pressure signals are available to both trains of SPDS. In the normal configuration, the "A" train of SPDS uses the "A" train RCS pressure signal, RC-3A-PT3, and the "B" train of SPDS uses the "B" train RCS pressure signal,RC-3B-PT3. The redundant signal is available, although not used in the calculation.

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U.S. Nuclear Regulatory Commission 3F0797-21 Page 2 of 5 Item 3:

REDUNDANCY OF CORE EXIT TEMPERATURE SENSORS NRC Question / Concern Provide more detail regarding the level of redundancy of the core exit thermocouples and justify only requiring 3 out of the 4 thermocouples per quadrant be OPERABLE to meet the LCO.

FPC Response The CR-3 reactor has fifty-two (52) core exit thermocouples. All of these thermocouples are input to the plant computer where they are available for indication and trending. The following discussion summarizes how the thermocouples are currently used in the plant:

Sixteen (16) of these thermocouples were upgraded to safety related in accordance with commitments under NUREG 0737 and Reg. Guide 1.97. These 16 thermocouples are input to: 1) the plant computer for recording and trending,2) the SPDS for the Saturation Margin Monitor Calculation (T-SAT) which currently provides backup to the existing T-SAT displays that are discussed in the following paragraph. These 16 thermocouples will be physically and electrically separated to meet the requirements of Reg. Guide 1.97 Type A, Category 1. Note that the requested change to TS Table 3.3.17-1 for changing Core Exit Temperature (Backup) from "2 sets of 5" to "3 per core quadrant" is j

independent of the Refuel 11 design improvement, and 3) three recorders located in the i

control room for use by the operators as a backup method for calculating T-SAT j

manually using steam tables. Should this be necessary, the highest indicated j

temperature would be used in the calculation regardless of the location in the core.

Thus, only one of the 16 available signals would be used by the operator if inis post i

accident backup method is needed.

Twelve (12) non-safety related thermocouples are used as: 1) input to the existing digital T-SAT displays located on the control board, and 2) input to the SPDS where the average of the highest five thermocouples is output on the SPDS displays as core exit temperature.

The remaining twenty-four (24) thermocouples are non-safety related and are input to the plant computer for recording and trending.

Requiring 12 of these signals to be OPERABLE assures a representative distribution across the core.

It is noted that the current Technical Specification for the Core Exit Temperature requires two sets of 5 to be OPERABLE. Increasing the minimum subset from 2 sets of 5 (10 total) to 3 per quadrant (12 total) enhances the minimum compliment of instrumentation avaihble to the operator in two ways: (1) it increases the minimum number of available signals from 10 to 12, and (2) requires a minimum core quadrant coverage requirement of 3 out of the available 4 thermocouples. Hence, this requested Technical Specification would result in improved core thermocouple numbers and core coverage which will result in an improvement in overall safety.

U.S. Nuclear Regulatory Commission Attachment I l

3F0797-21 Page 3 of 5 l

i l

Item 4:

SIGNAL PROCESSING FOR REACTOR COOLANT SYSTEM PRESSURE i

SIGNALS NRC Question / Concern Clarify the signal path for RCS pressure (Attachment 8, page 7 of 9/27/96 letter).

1 FPC Response 1

The two wide range RCS pressure signals originate at RC-3A-PT3 and RC-3B-PT3 and go to separate Engineered Safeguards cabinets in the main control room.

The i

instrumentation is safety related up through an isolation device in those cabinets. The signals then go from the isolation device directly to the SPDS.

The two low range RCS pressure signals originate at RC-147-PT and RC-148-PT and go to separate remote shutdown cabinets in the 4160 volt engineered safeguards switchgear rooms. The low range instrumentation is safety related up through an isolation device in the those cabinets. The signals then go from the isolation device directly to the SPDS.

Item 5:

JUSTIFICATION FOR ONE LPI FLOW INSTRUMENTATION CHANNEL NRC Question / Concern Provide justification for only requiring one channel of LPI flow instrumentation per train.

FPC Response The low pressure injection system is comprised of redundant divisions, each with a single qualified channel of flow instrumentation.

The criterion for Redundancv of Category 1 instrumentation states that redundant monitoring channels are not needed except for steam generator level instrumentation in two-loop plants. FPC interprets this statement as an indication that redundant monitoring channels for low pressure injection flow are not required.

Item 6:

POWER SUPPLY CONFIGURATION FOR ENHANCED SAFETY PARAMETER DISPLAY SYSTEM AND COMPARISON AGAINST REGULATORY GUIDE 1.75 NRC Question / Concern Describe the power supply isolation design as compared to Regulatory Guide 1.75.

Document that the Reg. Guide is not applicable to CR-3.

i U.S. Nuclear Regulatory Commission 3F0797-21 Page 4 of 5 FPC Response The design for the T-Sat Upgrade Project is on-going. An important aspect of this project is reconfiguring of the power source to enhance the system's reliability. Plans j

currently call for having the "A" train of SPDS fed from an "A" train safety-related standby power source and likewise the "B" train of SPDS be from a "B" train safety related l

standby power source.

In order to accomplish this objective, electrical devices will be utilized to isolate the Non-l Class 1E SPDS system from the Class 1E power system. In a phone conversation with the NRC, a question was raised regarding the applicability of Regulatory Guide 1.75,

" Physical Independence of Electrical Systems," and how this Reg. Guide is being applied to this design.

Although this Reg. Guide contains pertinent information regarding isolation devices, it is not part of CR-3's licensing Basis; therefore, it is not referred to in FPC's transmittals. In lieu of Reg. Guide 1.75, FPC's Electrical Design Criteria for Electrical Circuit Physical Separation and Cable Tray Loading will be used. This design criteria identifies acceptable isolation devices that can be used for electrically isolating l

Class 1E to Non-Class 1E circuits. The design criteria states that the isolation devices l

must demonstrate, by a product class test, that the maximum credible voltage or current l

transient applied to the device's Non-Class 1E side will not degrade the operation of the l

circuit connected to the device's Class 1E or associated side below an acceptable level, and shorts, grounds or open circuits occurring in the Non-Class 1E side will not degrade the circuit connected to the Class 1E or associated side below an acceptable level. The j

following devices, when properly applied and qualified can be used for isolation:

l l

a.

Amplifiers b.

Control Switches c.

Fiber Optic Couplers d.

Photo-optical Couplers l

e.

Relays l

f.

Transducers g.

Power Packs h.

Current Transformers i.

Circuit Breakers A fuse may be used as an isolation device to isolate a Class 1E circuit from non-Class 1E circuits provided that the following requirements are met:

(

a.

Each fuse type shall be tested (i.e. resistance measurement to verify overcurrent i

protection as designed).

1 b.

Fuses shall provide the design overcurrent protection capability for the life of the fuse.

c.

The fuse time-overcurrent trip characteristics for all current faults shall cause the fuse to open prior to the initiation of opening of any upstream interrupting device.

i l

U.S. Nuclear Regulatory Commission Attachment I 3F0797-21 Page 5 of 5 l

d.

The power source shall be capable of supplying the necessary current under fault conditions to ensure the proper coordination without loss of function of Class 1E loads.

e.

The fuse size shall be less than the continuous rating of the conductor taking into l

account any derating that might apply.

t The above criteria will be considered and applied to this part of the design. The design intent is to feed the new SPDS power circuits from a safety related vital bus distribution panel (VBDP). Electrical isolation will be accomplished in the VBDP panel via a breaker l

l and fuse arrangement. An electrical calculation will be generated to ensure that there

)

l is proper coordination between the fuse and the breaker.

l item 7:

AVAILABILITY OF BACKUP PRESSURE AND TEMPERATURE SIGNALS l

FOR SUBCOOLING MARGIN DETERMINATION NRC Question / Concern l

State that qualified RCS pressure and temperature signals are available to calculate subcooling margin using steam tables. Site the most recent Regulatory Guide 1.97 compliance submittal.

FPC Response i

f RCS pressure and temperature (both T-hot and incore) are available from indicators on the main control board that are fully qualified to the recommendations of Reg. Guide l-1.97. This instrumentation is described in FPC's most recent submittal on Reg. Guide 1.97 instrumentation dated December 5,1990 (3F1290-01) on pages 8,11, and 18. This l

instrumentation can be used, in conjunction with steam tables, to calculate subcooling l

margin.

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