Text

Proposed Tech Specs Re Nuclear Instrumentation Sys Power Range Daily Surveillance Requirement
	ML20155J456
Person / Time
Site:	Farley
Issue date:	11/06/1998
From:	; SOUTHERN NUCLEAR OPERATING CO.
To:	;
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ML20155J445	List: ML20155J439; ML20155J456;
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	NUDOCS 9811120164
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- w FNP Unit 1 Current Technical Specifications Reactor Trip System Instrumentation Channed Page Unit 1 Resision Page 3/4 3-14 Replace

,4 9811120164 981106 I" PDR ADOCK 05000348i P PDRif I

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. l r l TABLE 4.3-1 (Continued) ]

l 4 " TABLE NOTATION

,With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1) -

If not performed in previous 7 days.

MTQ (2) -

Heat balance only, above 15% of RATED THERMAL POh k Adjus h el ifA Q::lut; di"'------ 1: cr gt:r_9 --J n-- r ___

g g ,g e.deu\at dwereie.emanRE.%iN g WW m 4h*w + 2. litt to tJJ ~ T W e M e to excore axial flux di ference every 31 F .

Recalibrate if the absolute difference is greater than or equal to 3 percent.

(4) -

Manual ESF functional input check every 18 months. l (5) -

Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) -

Below the P-6 (Block of Source Range Reactor Trip) setpoint.

Upon reaching P-6 from MODE 2 the CHANNEL CHECK must be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

(8) - Logic only, if not performed in previous 92 days.

(9) - CHANNEL FUNCTIONAL TEST will consist of verifying that each channel indicates a turbine trip prior to latching the turbine and indicates no turbine trip prior to P-9.

(10) -

If not performed in the previous 31 days.

(11) -

Independently verify OPERABILITY of the undervoltage and shunt trip circuitry for the Manual Reactor Trip Function.

(12) -

Verify reactor trip breaker and reactor trip bypass breaker open upon actuation of each Main control Board handswitch.

(13) -

Local manual shunt trip prior to placing breaker in service.

Local manual undervoltage trip prior to placing breaker in service.

(14) -

Undervoltage trip via Reactor Protection System.

(15) -

Local manual shunt trip.

FARLEY-UNIT 1 3/4 3-14 AMENDMENT NO.138

,7 +=

TARTE d _ h1 I cnnt- 1 nu ed )

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TARf2 NOTATTON With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1) -

If not performed in previoum 7 days.

(2) -

Heat balance only, above 15% of RATED THERMAL POWER (RTP).. Adjust NIS channel if calorimetric calculated power exceeds NIS indicated power by more than +2% RTP.

(3) -

Compare incore to excore axial flux difference every 31 EFPD.

Recalibrate if the absolute difference is greater than or equal to 3 percent.

(4) -

Manual ESF functional input check every 18 months.

(5) -

Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(6) -

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) -

Below the P-6 (Block of Source Range Reactor Trip) setpoint. Upon reaching P-6 from MODE 2 the CHANNEL CHECK must be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

(8) -

Logic only, if not performed in previous 92 days..

(9) -

CHANNEL FUNCTIONAL TEST will consist of verifying that each channel indicates a turbine trip prior to latching the turbine and indicates no turbine trip prior to P-9.

(10) -

If not performed in the previous 31 days.

(11) -

Independently verify OPERABILITY of the undervoltaga and shunt trip circuitry for the Manual Reactor Trip Function.

(12) -

Verify reactor trip breaker and raactor trip bypass breaker open upon actuation of each Main Control Board handswitch.

(13) -

Local manual shunt trip prior to placing breaker in service. Local manual undervoltage trip prior to placing breaker in service.

(14) -

Undervoltage trip via Reactor Protection System.

(15) - Local manual shunt trip.

FARLEY-UNIT 1 3/4 3-14 EMENDMENT NO.

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FNP Unit 2 Current Technical Specifications Reactor Trip System Instrumentation Chanced Paa; i

L] nit I Revn. .sson 1

Page 3/4 3-14 Replace l l

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l TABLE 4.3-1 (Continued) l 1

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,, ,, TABLE NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

I (1) -

If not performed in previous 7 days. !

Nb l (2) -

Heat balance only, above 15% of RATED THERMAL PO Adjus MIS

-- ---- W hannelj fg2-_r bt: Ef f err:n ir crr:tir *M_O rrrO -

sneakele C.*\cuWheci ; matt exceesis Ar.S unbishack p=M b~y mm uL %evi+,2,% greg l D I - N pare incore t erence every 3 l EFPD. Recalibrate if the absolute difference is greater than or equal to 3 percent.

(4) -

Manual ESF functional input check every 18 months. l (5) -

Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

, (6) -

Neutron detectors may be excluded from CHANNEL CALIBRATION.

t f

(7) -

Below the P-6 (Block of Source Range Reactor Trip) setpoint.

Upon reaching P-6 from MODE 2 the CHANNEL CHECK must be l performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

(8) -

Logic only, if not performed in previous 92 days.

l l (9) -

CHANNEL FUNCTIONAL TEST will consist of verifying that each

! channel indicates a turbine trip prior to latching the turbine and indicates no turbine trip prior to P-9. j

? l l (10) -

If not performed in the previous 31 days. l (11) -

Independently verify OPERABILITY of the undervoltage and shunt trip cireuitry for the Manual Reactor Trip Function.

(12) -

Verify reactor trip breaker and reactor trip bypass breaker open upon actuation of each Main Control Board handswitch. I t

I (13) -

Local manual shunt trip prior to placing breaker in service.

Local manual undervoltage trip prior to placing breaker in service.

t i

l (14) - Undervoltage trip via Reactor Protection System.

(15) -

Local manual shunt trip.

t FARLEY-UNIT 2 3/4 3-14 AMENDMENT NO.130

l Tant.e A,3-1 t enne t nismA t TART.E MnTaTTON With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1) -

If not performed in previous 7 days.

(2). -

Heat balance only, above 15% of RATED THERMAL POWER (RTP). Adjust NIS channel if calorimetric calculated power exceeds NIS indicated power by more than +2% RTP.

(3) -

Compare incore to excore axial flux difference every 31 EFPD.

Recalibrate if the absolute difference is greater than or equal to 3 percent.

(4) -

Manual ESF functional input check every 18 months.

(5) -

Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

J (6) -

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) -

Below the P-6 (Block of Source Range Reactor Trip) setpoint. Upon reaching P-6 from MODE 2 the CHANNEL CHECK must be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

(8) Logic only, if not performed in previous 92 days.

(9) -

CHANNEL FUNCTIONAL TEST will consist of verifying that each channel indicates a turbine trip prior to latching the turbine and indicates no turbine trip prior to P-9.

(10) - If not performed in the previous 31 days.

(11) -

Independently verify OPERABILITY of the undervoltage and shunt trip circuitry for the Manual Reactor Trip Function.

' (12) -

Verify reactor trip breaker and reactor trip bypass breaker open upon actuation of each Main Control Board handswitch.

(13) -

Local manual shunt trip prior to placing breaker in service.

Local manual undervoltage trip prior to placing breaker in service.

(14) -

Undervoltage trip via Reactor Protection System.

(15) -

Local manual shunt trip.

FARLEY-UNIT 2 3/4 3-14 AMENDMENT NO.

1 7 em j ATTACllMENT 11 1

I FARLEY NUCLEAR PLANT l

l IMPROVED TECHNICAL SPECIFICATIONS CHANGE RI. QUEST NIS POWER RANGE CHANNEL DAILY SURVEILLANCE REQUIREMENT l

l l

FNP Unit 1/2 Technical Specifications Changed Page List j l

FNP Unit 1/2 Technical Specifications Marked-up Page )

i FNP Unit 1/2 Technical Specifications Typed Page FNP Unit 1/2 Bases Changed Pages List FNP Unit 1/2 Bas:s Marked-up Pages FNP Unit 1/2 Bases Typed Pages

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Farley Nuclear Plant

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improved Technical Specifications Reactor Trip System Instrumentation Changed Paac Unit 1/2 Revision

. Page 3.3.1-9 Replace e .

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l RTS Instrumentation l

3.3.1

! SURVEILLANCE REQUIREMENTS

N OTE- - -

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function. 1

. l l

SURVEILLANCE FREQUENCY !

SR 3.3.1.1 NOTE Not required to be performed for source range instrumentation until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is < P-6.

l Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.2 NOTES ,

1. Adjust NIS annelif obsc!qt di"crence is I WA c. w:me. ec cateutated Qv exce7d) 3 tESI (cded power more. %n + 2.% rr , l

. o require to e orme un i ours I after THERMAL POWER is 2 15% RTP.

1 1

Compare results of calorimetric heat balance 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months l calculation to Nuclear Instrumentation System (NIS) i I

channel output.

SR 3.3.1.3 NOTES I

1. Recalibrate NIS channelif absolute difference l is 2 3%.

2. Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is 2 15% RTP.

Compare results of the incore detector 31 effective full measurements to NIS AFD. power days (EFPD) l Farley Units 1 and 2 3.3.1-9 Amendment No. (Unit 1)

Amendment No. (Unit 2)

RTS Instrumsntation 3.3.1 SURVEILLANCE REQUIREMENTS NOTE -

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 NOTE Not required to be performed for source range instrumentation until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is < P-6.

Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.2 -

NOTES

1. Adjust NIS channel if calorimetric calculated power exceeds NIS indicated power by more than + 2% RTP.

2. Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is 215% RTP.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Compare results of calorimetric heat balance calculation to Nuclear instrumentation System (NIS) channel output.

SR 3.3.1.3 NOTES

1. Recalibrate NIS channelif absolute difference is 2 3%.

2. Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is 215% RTP.

Compare results of the incore detector 31 effective full measurements to NIS AFD. Power days (EFPD)

Farley Units 1 and 2 3.3.19 Amendment No. (Unit 1)

Amendment No. (Unit 2)

-. .-. -. . - . . -. -. - .. .- . . . . = . . - - - . _ . . .

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Farley Nuclear Plant Improved Technical Specifications Bases Reactor Trip System Instrumentation Chanced Panes Unit 1/2 Revision Page B 3.3.1-50 Replace Page B 3.3.1-51 Replace Page B 3.3.1-52 Replace Page B 3.3.1-53 Replace Page B 3.3.1-54 Replace Ptge B 3.3.1-55 Replace Page B 3.3.1-56 Replace Page B 3.3.1-57 Replace Page B 3.3.1-58 Add Page B 3.3.1-59 Add l

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Improved Technical Specifications Bases Markups INSERT 1 If the calorimetric is performed at part power (< 50% RTP), adjusting the NIS channel indication in the increasing power direction will assure a re4or trip below the safety analysis limit (s 118% RTP).

Making no adjustment to the NIS channel indication in the decreasing power direction due to a part power calorimetric assures a reactor trip consistent with the safety analyses.

INSERT 2

'Ihis allowance does not preclude malang indicated power adjustments, if desired, when the calorimetric calculated power is less than the NIS channel iaAle=tM power. To provide close agreement between indicated power and calorimetric power and to preserve operating margin, the NIS channels are normally adjusted when operating at or near full power during steady-state conditions. However, discretion must be exercised if the NIS channel indicat~i power is adjusted in the decreasing power direction due to a part power calorimetric (< 50% RTP). ' Ibis action could introduce a non-conservative bias at higher power levels which could result in an NIS reactor trip abow the safety analysis limit (> 118% RTP). The cause of the non-conservative bias is the decreased accuracy of the calorimetric at reduced power conditions, as discussed in Westmghouse Technical Bulletin, ESBU-TB-92-14-R1, "Decalibration Effects Of Calorimetric Power Level Measurements On The NIS High Power Reactor Trip At Power Levels Less Than 70% RTP,"(Ref.13). To assure a reactor trip below the safety analysis limit, the Power Range Neutron Flux - High bistables are set s 85% RTP: 1) whenever the NIS channel indir=+~i power is adjusted in the decreasing power direction due to a part power calorimetric below 50% RTP; and 2) for a post refueling startup. Before the Power Range Neutron Flux - High bistables are re-set s 109% RTP, the NIS channel calibration must be confirmed based on a calorimetric performed ;t 50% RTP.

INSERT 3 A power level of 15% RTP is chosen based on plant stability, i.e., automatic rod control capability and turbine generator synchronized to the grid.

INSERT 4

13. WattagWse Technical Bulletin, ESBU-TB-92-14-RI, "Decalibration Effects Of Calorimetric Power Level Measurements On The NIS High Power Reactor Trip At Power Levels I.ess 'Ihan 70% RTP."

RTS Instrumentation

, .. B 3.3.1 BASES SURVEILLANCE SR 3.3.1.2 REQUIREMENTS (continued) tea e power SR 3.3.1.1:ornpares e calorimetric he balance cal - lon t M hanM c4lculded power NIS cha el output ery 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If ie calorimetri xceeds the NIS

[ rasset 1}. channe by 2% RTP, the NIS is not declared inoperable, but

{ re.Q must be adjusted. If the NIS channel output cannot be properly

'usted, the channel is declar ~

rable. = -

ca\edmetrie TeigM y.wer exc.eed ca\cu\akel PoweV o o es mo ify .1. first Note indicates that t e ch nel output shall be adjusted consistent with the calorimetri if thybevlme d.llerence beti.een the NIS channel output ..c and :pesults b3m.re. %n + ce:c.i..mm .5 > 2% RTP. The second Note clarifies that this Surveillance is required only if reactor power is 215% RTP and that 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed for performing the first Surveillance after reaching 15% RTP.J: lcv.cr pcyscr lcve:5, calcr:mctdc data are lnaccura:c.

{x.oswr g - -- -

$t agtperence, wbeen The Frequency of every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is adequate. It is based u operating experience, considering instrument reliability and operating history data for instrument drift. Together these factors demonstrat he chang in the absciute difference betvicca NlC and heat balance anA We.uts cknne\ bd.'64Na power;;jige:s .f swepe.N exce"DRfP

+ ts wFv.^% in a_ny_24 te.cl ) hour period.

n additioiGio3Foiroom operators peno ica y monitor redundant indications and alarms to detect deviations in channel outputs.

SR 3.3.1.3 SR 3.3.1.3 compares the incore system to the NIS channel output every

, 31 EFPD. If the absolute difference is 2 3%, the NIS channelis still OPERABLE, but must be recalibrated.

If the NIS channel cannot be properly recalibrated, the channelis declared inoperable. This Surveillance is performed to verify the f(ol) input to the overtemperature AT Function.

Two Notes modify SR 3.3.1.3. Note 1 indicates that the excore NIS channel shall be recalibrated if the absolute difference between the incore and excore AFD is 2 3%. Note 2 clarifies that the Surveillance is required only if reactor power is 215% RTP and that 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed for performing the first Surveillance after reaching 15% RTP.

I (continued)

Farley Units 1 and 2 B 3.3.1 50 Revision 0

RTS Instrumentation B 3.3.1

.BA$$S SURVEILLANCE SR 3.3.1.14 (continued)

REQUIREMENTS As appropriate, each channel's response must be verified every 18 months on a STAGGERED TEST BASIS. Each verification shall include at least one Logic train such that both Logic trains are verified at least once per 36 months. Testing of the final actuation devices is included in the testing. Response times cannot be determined during unit operation because equipment operation is required to measure i

response times. Experience has shown that these components usually l L

. pass this surveillance when performed at the 18 month Frequency. ;

Therefore, the Frequency was concluded to be acceptable from a ~l

! reliability standpoint. l SR 3.3.1.14 is modified by a Note stating that neutron detectors are l excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response.

l REFERENCES- 1. FSAR, Chapter 7. I

2. FSAR, Chapter 6.

3. FSAR, Chapter 15,

4. IEEE-279-1971

5. 10 CFR 50.49.

6. WCAP 13751, FNP RTS/ESFAS Setpoint Methodology Study.

7. WCAP-10271-P-A, Supplement 2, Rev.1, June 1990.

8. FSAR, Table 7.2.5.

9. RPS Functional System Description (FSD)- A- 181007.

10. WCAP 12925, Median Signal Selector (MSS).

11. WCAP 13807/13808, Elimination of Feedwater Flow trip via implementation of MSS.

12. Joseph M. Farley Nuclear Power Plant Unit 1 (2) Precautions, l-seart 4 Limitations and Setpoints - U - 266647 (U - 280912).

t.

1Farley Units 1 and 2 B 3.3.157 Revision 0

. - . - - . . -. . .- . - . - - - - . - . - . . - - - - - . - - _ _ - =.. - -

BASES

~

SNRVEILLANCE SR 3.3.1.2 REQUIREMENTS (continued) SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS channel output every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the calorimetric calculated power exceeds the NIS channel indicated power by more than + 2% RTP, the NIS channel is not declared inoperable, but must be adjusted. If the NIS channel output cannot be property adjusted, the channel is declared inoperable.

If the calorimetric is performed at part power (< 50% RTP),

adjusting the NIS channel indication in the increasing power direction will assure a reactor trip below the safety analysis limit (s 118% RTP). Making no adjustment to the NIS channelindication in the decreasing power direction due to a part power calorimetric assures a reactor trip consistent with the safety analyses.

This allowance does not preclude making indicated power adjustments, if desired, when the calorimetric calculated power is ;

less than the NIS channel indicated power. To provide close j agreement between indicated power and calorimetric power and to preserve operating margin, the NIS channels are normally adjusted i when operating at or near full power during steady-state conditions. !

However, discretion must be exercised if the NIS channelindicated power is adjusted in the decreasing power direction due to a part power calorimetric (< 50% RTP). This action could introduce a non-conservative bias at higher power levels which could result in an NIS reactor trip above the safety analysis limit (> 118% RTP). The cause of the non-consentative bias is the decreased accuracy of the calorimetric at reduced power conditions, as discussed in l Westinghouse Technical Bulletin, ESBU-TB-92-14 R1, "Decalibration Effects Of Calorimetric Power Level Measurements On The NIS High Power Reactor Trip At Power Levels Less Than 70% RTP," (Ref.13). To assure a reactor trip below the safety analysis limit, the Power Range Neutron Flux - High bistables are set s 85% RTP: 1) whenever the NIS channel indicated power is adjusted in the decreasing power direction due to a part power calorimetric below 50% RTP; and 2) for a post refueling startup.

Before the Power Range Neutron Flux - High bistables are re-set s 109% RTP, the NIS channel calibration must be confirmed based on a calorimetric performed 2 50% RTP.

Two Notes modify SR 3.3.1.2. The first Note indicates that the NIS channel output shall be adjusted consistent with the calorimetric calculated power if the calorimetric calculated power exceeds the (continued)

Farley Units 1 and 2 B 3.3.1-50 Revision 0

- ~ .. _ _ _.. _ .. _ . _ . .___ _ ... _ .. _ ._... _ . _ _ _ _ ._ ,

f h

fe BASES SURVEILLANCE SR 3.3.1.2 (continued)

- REQUIREMENTS NIS channel output by more than + 2% RTP The second Note

! clarifies that this Surveillance is required only if reactor power is 215% RTP and that 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed for performing the first Surveillance after reaching 15% RTP. A power level of 15% RTP is I-chosen based on plant stability, i.e., automatic rod control capability and turbine generator synchronized to the grid. !

The Frequency of every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is adequate. It is based on unit l- operating experience, considering instrument reliability and l l operating history data for instrument drift. Together these factors l

demonstrate that a difference between the heat balance calculated l power and the NIS channel indication of more than + 2% RTP is not i expected in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. :

In addition, control room operators periodically monitor redundant j indications and alarms to detect deviations in channel outputs. l l

l SR 3.3.1.3 SR 3.3.1.3 compares the incore system to the NIS channel output i every 31 EFPD If the absolute difference is 2 3%, the NIS channel ;

j is still OPERABLE, but must be recalibrated. l l 1

If the NIS channel cannot be properly recalibrated, the channelis l L

declared inoperable. This Surveillance is performed to verify the f(AI) input to the overtemperature AT Function.

, Two Notes modify SR 3.3.1.3. Note 1 indicates that the excore NIS i, channel shall be recalibrated if the absolute difference between the incore and excore AFD is 2 3%. Note 2 clarifies that the -

L Surveillance is required only if reactor power is 215% RTP and that L 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed for performing the first Surveillance after reaching 15% RTP.

l l- The Frequency of every 31 EFPD is adequate. It is based on unit l operating experience, considering instrument reliability and operating history data for instrument dnft. Also, the slow changes in neutron flux during the fuel cycle can be detected during this interval. 1 I

(continued) t q ' Farley Units 1 and 2 8 3.3.1-51 Revision 0

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W BASES SdRV5LI.ANCE SR 3.3.1.4 REQUIREMENTS (continued) SR 3.3.1.4 is the performance of a TADOT every 31 days on a STAGGERED TEST BASIS. This test shall verify OPERABILITY by actuation of the end devices.

The RTB test shall include separate verification of the undervoltage trip via the Reactor Protection System and the local manual shunt Np mechanism. The bypass breaker test shall include a local manual shunt trip and local manual undervoltage trip. A Note has been added to indicate that this test must be performed on a bypass breaker prior to placing it in service. The independent test of undervoltage and shunt trip circuitry for the bypass breakers for the manual reactor trip function is included in SR 3.3.1.12. No capability is provided for performing such a test at power.

The Frequency of every 31 days on a STAGGERED TEST BASIS is adequate. It is based on industry operating experience, considering lastrument reliability and operating history data.

'I SR 3.3.1.5 SR 3.3.1.5 is the performance of an ACTUATION LOGIC TEST.

The SSPS is tested every 31 days on a STAGGERED TEST BASIS, using the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation. Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function. The Frequency of every 31 days on a STAGGERED TEST BASIS is adequate. It is based on industry opersting experience, considering instrument reliability and operating histosy data.

SR 3.3.1.8 SR 3.3.1.6 is the performance of a TADOT and is performed every 92 days, as justified in Reference 7.

The SR is modified by a Note that excludes verification of setpoints from the TADOT. Since this SR applies to RCP undervoltage and underfrequency relays, setpoint verification requires elaborate bench calibration and is accomplished during the CHANNEL CAllBRATION.

(continued)

Farley Units 1 and 2 B 3.3.1-52 Revision 0

, - - e en -.--n-v,-,.n- ---.- w -- , , . - - .,--.-,n--

BASES j SURVEILLANCE SR 3.3.1.7

, REQUIREMENTS (continued) SR 3.3.1.7 is the performance of a COT every 92 days.

~

. A COT is performed on each required channel to ensure the rack

. components will perform the intended Function.

, Setpoints must be within the Allowable Values specified in Table 3.3.1-1.

1 The "as found" values are evaluated to ensure consistency with I

. (i.e., bounded by) the drift allowance used in the setpoint j methodology. The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.

i SR 3.3.1.7 is modified by a Note that provides a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months delay in the '

requirement to perform this Survfilance for source range instrumentation when entering MODE 3 from MODE 2. This Note allows a normal shutdown to proceed without a delay for testing in l MODE 2 and for a short time in MODE 3 until the RTBs are open I and SR 3.3.1.7 is no longer required to be performed. If the unit is to be in MODE 3 with the RTBs closed for > 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months this Surveillance must be performed prior to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entry into MODE 3.

l The Frequency of 92 days is justified in Reference 7.

l SR 3.3.1.8 SR 3.3.1.8 is the performance of a COT as described in SR 3.3.1.7, except it is modified by a Note that this test shall ,

include verification that the P-6 and P-10 interlocks are in their required state for the existing unit condition. The Frequency is modified by a Note that allows this surveillance to be satisfied if it has been performed within 92 days of the Frequencies prior to reactor startup and four hours after reducing power below P.10 and P-6. The Frequency of " prior to startup" ensures this surveillance is performed prior to critical operations and applies to the source, intermediate and power range low instrument channels. The Frequency of "4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reducing power below P-10" (applicable to the power range low channels) and "4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reducing power below P-6" (applicable to source range channels) allows a normal shutdown to be completed and the unit removed (continued)

Farley Units 1 and 2 B 3.3.1-53 Revision 0

BASES ,

SURVEILLANCE S R 3 3.1.8 (continued)

REQUIREMENTS from the MODE of Applicability for this surveillance without a delay to perform the testing required by this sunteillance. This surveillance is not required for the intermediate range instrumentation during power descension. The Frequency of every 92 days thereafter applies if the plant remains in the MODE of Applicability after the initial performances of prior to reactor startup and four hours after reducing power below P 10 or P-6. The MODE of Applicability for this surveillance is < P-10 for the power range low channels and < P-6 for the source range channels. Once the unit is in MODE 3, this surveillance is no longer required. If power is to be maintained < P-10 or < P-6 for more than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , then the testing required by this surveillance must be performed prior to the expiration of the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limit. Four hours is a reasonable time to complete the required testing or place the unit in a MODE where this surveillance is no longer required. This test ensures that the NIS source, intermediate, and power range low channels are OPERABLE prior to taking the reactor critical and after reducing power into the applicable MODE (< P-10 or

< P-6) for the power and source range instrumentation for periods >

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SR 3.3.1.9 SR 3.3.1.9 is a calibration of the excore channels to the incore channels. If the measurements do not agree, the excore channels are not declared inoperable but must be calibrated to agree with the incore detector measurements. If the excore channels cannot be adjusted, the channels are declared inoperable. This Surveillance is I performed to verify the f(AI) input to the overtemperature AT Function.

Two Notes modify SR 3.3.1.9. Note 1 states that neutron detectors may be excluded from the calibration. Note 2 specifies that this Surveillance is required only if reactor power is > 50% RTP and that 7 days are allowed for completing the surveillance after reaching 50% RTP.

The Frequency of 18 months is based on plant operating experience and has proven sufficient to maintain the calibration of the excore detectors.

(continued)

Farley Units 1 and 2 8 3.3.1-54 Revision 0

BASES SURVEll' LANCE SR 3.3.1.10 REQUIREMENTS (continued) A CHANNEL CALIBRATION is performed every 18 months, or approximately at every refueling. CHANNEL CAllBRATION is a ,

complete check of the instrument loop, including the sensor. The

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test verifies that the channel responds to a measured parameter within the necessary range and accuracy. .

CHANNEL CAllBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology. The "as found" values are evaluated to ensure consistency with I.e., bounded -

by the drift allowance used in the setpoint methodology.

The Frequency of 18 months is based on the assumption of an 18 month calibration intervalin the determination of the magnitude of equipment drift in the setpoint methodology and the need to perform this surveillance under the conditions that apply during a plant outage. Operating experience has shown these components usually -

pass the surveillance when performed on the 18 month Frequency.

This SR is modified by two Notes. Note 1 states that neutron detectors are excluded from the CHANNEL CAllBRATION. The ---

CHANNEL CAllBRATION for the power range neutron detectors consists of a normalization of the detectors based on a power calorimetric and flux map performed above 15% RTP. The CHANNEL CALIBRATION for the source range and intermediate range neutron detectors consists of obtaining the detector plateau or preamp discriminator curves and evaluating those curves. This Surveillance is not required for the NIS power range detectors for entry into MODE 2 or 1, and is not required for the NIS intermediate range detectors for entry into MODE 2, because the unit must be in at least MODE 2 to perform the test for the intermediate range detectors and MODE 1 for the power range detectors. Note 2 states that this test shall include verification that the time constants are adjusted to the prescribed values where applicable. The OTAT, OPAT, and the power range neutron flux rate functions contain required time constants.

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(continued) l Farley Units 1 and 2 8 3.3.1-55 Revision 0

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BASES SURVEILLANCE SR 3.3.1.11 REQUIREMENTS (continued) SR 3.3.1.11 is the performance of a COT of RTS interlocks every 18 months. This COT is intended to verify the interlock Logic only.

The Frequency is based on the known reliability of the interlocks and the multichannel redundancy available, and has been shown to be acceptable through operating experience.

SR 3.3.1.12 SR 3.3.1.12 is the performance of a TADOT of the Manual Reactor Trip, RCP Breaker Position, and the St Input from ESFAS. This i TADOT is performed every 18 months. The test shallindependently verify the OPERABILITY of the undervoltage and shunt trip mechanisms for the Manual Reactor Trip Function for the Reactor Trip Breakers and Reactor Trip Bypass Breakers. The Reactor Trip Bypass Breaker test shallinclude testing of the automatic undervoltage trip.

The Frequency is based on the known reliability of the Functions and the multichanne: redundancy available, and has been shown to be acceptable through operating experience.

The SR is modified by a Note that excludes verification of setpoints J from the TADOT. The Functions affected have no setpoints )

associated with them. 1 SR 3.3.1.13 i

SR 3.3.1.13 is the performance of a TADOT of Turbine Trip )

Functions. This TADOT will consist of verifying that each channel '

indicates a Turbine trip prior to Latching the turbine and indicates no turbine trip prior to P-9. A Note states that this Surveillance is not required if it has been performed within the previous 31 days.

Verification of the Trip Setpoint does not have to be performed for this Surveillance.

(continued)

Farley Units 1 and 2 B 3.3.1-56 Revision 0

i BASES SURVEILLANCE SR 3.3.1.14 !

REQUIREMENTS (continued) SR 3.3.1.14 verifies that the individual channel / train actuation l response times are less than or equal to the maximum values l assumed in the accident analysis. Response time testing acceptance criteria are included in FSAR, Table 7.2.5 (Ref. 8). ;

Individual component response times are not modeled in the analyses. !

The analyses model the overall or total elapsed time, from the point I at which the parameter exceeds the trip setpoint value at the sensor I to the point at which the equipment reaches the required functional state (i.e., control and shutdown rods fully inserted in the reactor core).

For channels that include dynamic transfer Functions (e.g., lag, lead / lag, rate / lag, etc.), the response time test may be performed with the transfer Function set to one, with the resulting measured response time compared to the appropriate FSAR response time.

Alternately, the response time test can be performed with the time constan's set to their nominal value, provided the required response time is analytically calculated assuming the time constants are set at

' their nominal values.

Response time may be verified by actual tests in any series of l sequential, overlapping or total channel measurements, or by l summation of allocated sensor response times with actual test on the remainder of the channelin any series of sequential or overlapping '

measurements. Allocations for specific pressure and differential pressure sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) in place, onsite, or offsite (e.g. vendor) test measurements, or (3) utilizing vendor engineering specifications.

WCAP - 13632, Revision 2, " Elimination of Pressure Sensor Response Time Testing Requirements," provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. The allocations for these sensor response times must be verified prior to placing the sensor in operational service and re-verified following maintenance that may adversely affect response time. In general, electric repair work does not impact (continued)

Farley Units 1 and 2 B 3.3.1-57 Revision 0

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BASES

'SbRVEILLANCE SR 3.3.1.14 (continued)

REQUIREMENTS response time provided the parts used for repair are of the same type and value. One example where time response could be affected is replacing the sensing assembly of a transmitter.

Response time verification for other sensor types must be demonstrated by test.

As appropriate, each channel's response must be verified every 18 months on a STAGGERED TEST BASIS. Each verification shall include at leact one Logic train such that both Logic trains are verified at least once per 36 months. Testing of the final actuation devices is included in the testing. Response times cannot be determined during unit operation because equipment operation is required to measure response times. Experience has shown that these components usually pass this surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.3.1.14 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluduq the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response.

REFERENCES 1. FSAR, Chapter 7.

2. FSAR, Chapter 6.

3. FSAR, Chapter 15.

4. IEEE-279-1971.

5. 10 CFR 50.49.

6. WCAP 13751, FNP RTS/ESFAS Setpoint Methodology Study.

7. WCAP-10271-P-A, Supplement 2, Rev.1, June 1990.

(continued)

Farley Units 1 and 2 B 3.3.1-58 Revision 0

BASES REFERENCES 8. FSAR, Table 7.2.5.

(continued)

9. RPS Functional System Description (FSD) - A - 181007,

10. WCAP 12925, Median Signal Selector (MSS).

l' 11. WCAP 13807/13808, Elimination of Feedwater Flow trip via Implementation of MSS.

12. Joseph M. Farley Nuclear Power Plant Unit 1 (2) Precautions, Limitations and Setpoints - U - 266647 (U - 280912).

13. Westinghouse Technical Bulletin, ESBU-TB-92-14-R1, "Decalibration Effects Of Calorimetric Power Level Measurements On The NIS High Power Reactor Trip At Power Levels Less Than 70% RTP."

Farley Units 1 and 2 B 3.3.1-59 Revision 0 l

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- ATTACHMENT III SAFETY ANALYSIS JOSEPH M. FARLEY NUCLEAR PLANT

. NIS POWER RANGE CHANNEL DAILY SURVEILLANCE REQUIREMENT TECHNICAL SPECIFICATIONS CHANGE i

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SAFETY ANALYSIS JOSEPH M. FARLEY NUCLEAR PLANT NIS POWER RANGE CHANNEL DAILY SURVEILLANCE REQUIREMENT TECHNICAL SPECIFICATIONS CHANGE INTRODUCTION When operating above 15% Rated Hermal Power (RTP), the current Farley Technical Specifications Nuclear Instrumentation System (NIS) Power Range daily Surveillance Requirement requires the adjustment of the Power Range channel (s) when the absolute difference between Power Range indicated power and wn=hry side calorimetric power is greater than 2% RTP. Compliance with this Technical Specifications requirement may result in a non-conservative channel calibration during reduced power operations. He proposed Technical Specifications change will resolve this undesirable condition by requiring adjustment of the NIS Power Range channel (s) only when the calorimetric calculated power is greater than the Power Range imiW~I power by + 2%.

In the interim, to ensure compliance with the Technical Specifications and conformance with the safety analyses, Farley has implemented admmistrative controls. However, these interim controls have resulted in negative operational and equipment impacts. He proposed surveillance change will also reduce the impact of the interim administrative controls.

He non-conservative NIS calibration issue and the proposed Technical Specifications change are applicable to other Westinghouse plants. Farley is the lead plant for the Westmghouse Owner's Group (WOG).

BACKGROUND Westinghouse Technical Bulletin ESBU-TB-92-14 R1, "Decalibration Effects Of Calorimetric Power Measurements On The NIS High Power Reactor Trip At Power Levels Less Than 70% RTP," dated February 6,1996, identified potential effects of decalibrating the NIS Power Range channels at part power operation. The decalibration can occur due to the increased uncenainty of the mn=4 y side power calorimetric when performed at part power (less than approximately 70% RTP). When NIS channel indication is reduced to match calculated power, the decalibration results in a non-conservative bias. He proposed change to the Technical Specifications removes the requirement to adjust the NIS Power Range channels when the indicated power is greater than the calorimetric calculated power by an absolute difference of > 2% RTP.

Westinghouse Technical Bulletin 92-14, " Instrumentation Calibration At Reduced Power," dated January 18,1993, was revised as a result of Westinghouse's review of ABB-CE Infobulletin 94-01, " Potential Nonconservative Treatment Of Power Measurement Uncertainty," dated June 21,1994. Both bulletins ;

addressed the potential decalibration effects on NIS Power Range indications and reactor trip setpoints due j to increased uncertainties associated with enmbry side power calorimetric measurements performed at (

low power levels. After review of the ABB-CE bulletin, Westinghouse determined that further information and clarification would be advisable and issued ESBU-TB-92-14-RI.

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'Jhc primary error contributor to the instrument uncertainty for a waaA=ry side power calorimetric measurement is the feedwater flow measurement, which is typically a AP measurement across a feedwater venturi. While the measurement uncertainty remams constant in AP as power decreases, when translated into flow, the uncertainty increases as a square term. 'Ihus a 1% flow error at 100% power can approach a 10% flow error at 30% RTP cven though the AP error has not changed. ESBU-TB-92-14-R1 depicted how the potential effects of this error incicase at lower power levels. In the example presented, for a 10% error in maaA= y side power calorimetnc, the NIS power range could be sufficiently biased in the non-censervative direction to preclude a reactor trip within the assumptions of the safety analy ses. For Farley, this event is the Rod Withdrawal From 10% RTP.

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'Ihere are six recommenA=' ions in the revised bulletin. Rxo.ws.dation Nos.1 - 5 are in concert with Farley practices and procedures. However, Recomn=htion No. 6 suggests that if the NIS Power Range indicates a higher power than the reada y side power calorimetric measurement at power levels below approximately 70%, the Power Range channel (s) should not be adjusted. 'Ihis recom==A= tion is in conflict with the Farley Technical Speci6 cations Power Range daily Surveillance Requirement, which requires channel adjustment whenever the absolute difference is > 2% above 15% RTP.

In response to ESBU-TB-92-14-RI, Farley determined that the i 2% RTP calorimetric power measurement uncertainty is valid for power levels 2 50% RTP based on the Farley-specific calorimetric measurement procedure. Farley also detennined that resetting the NIS Power Range (PR) High Neutron Flux High Setpoint reactor trip to s 85% RTP is an acceptable adminierative adjunct to continued performance of the calorimetric and adjustment of the NIS PR channels to reflect the calorimetric power below 50% RTP. 'Ihis second action is an interim solution that replaces Fwar3=* ion No. 6 of ESBU-TB-92-14-R1 and, thereby, obviates the conflict with the Technical Specifications. For long-term resolution, the WOG initiated a program (MUHP-3034) to obtain NRC approval to relax the present Technical Specifications requirements to always adjust NIS channels when indic=*~i power differs from calorimetric power by more than 2%. In LER 97-001-00, " Nuclear Instrumentation System Inaccuracies Below 50% Power," dated February 12,1997, Farley committed to evaluate the need for a Technical Specifications change based on the rcsuits of the WOG program Subsequently, Farley determined the proposed relaxation is a desirable resolution. Farley is the lead plant for this generic WOG program At Farley, the interim controls result in the following negative ima=c%

1. For calorimetric power determmations < 50% RTP, the NIS PR High Neutron Flux High Setpoint reactor trip bistables must be set at a nominal 85% RTP (or lower) if the NIS indir=*~i power is reduced by adjustment to reficct calorimetric power.

2. Subsequently, when operating power levels are increased above 50% RTP, power ascension delays can result while the NIS PR High Neutron Flux High Setpoint reactor trip bistables are reset to the TS nominal setpoint ofs 109% RTP.

3. Failure to reset the PR High Neutron Flux High Setpoint reactor trip bistables prior to increasing power above the adnunistrative setpoint would result in an inadvertent reactor trip.

4. The additional NIS PR bistable adjustments increase wear on the instrumentation.

The interim controls will not be elimia=*~i However, since the proposed surveillance change will preclude l una~~= y adjustments of the NIS Power Range channels, the above operational challenges will be reduced.

nispt:5.mse A - III - 2 10/29/98

PROPOSED TECHNICAL SPECIFICATIONS POWER RANGE SURVEILLANCE CHANGE Current Te'chnical Soccificatiom Surveillance Requirement 4.3.1.1 of the Current Technical Specifications (CTS) requires each Reactor Trip System instrumentation channel to be demonstrated operable by performance of the channel checks, calibrations and functional tests specified in Table 4.3-1. De NIS Power Range daily calibration is only applicable to Reactor Trip System Functional Units 2.A, PR High Neutron Flux High, and 2.B, PR High Neutron Flux Low. The current daily surveillance requirement is found in Table Notation (2), which states, " Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference is greater than 2 percent."

To prevent an undesired Power Range decalibration at part power operating conditions, the proposed change will not require a channel adjustment when indicated power is greater than calorimetric power.

Nevertheless, to ensure that the NIS PR channels are adjusted when indicated power is less than calorimetric power and to preserve the safety analyses assumptions, the proposed CTS change will state,

" Heat balance only, above 15% of RATED THERMAL POWER (RTP). Adjust NIS channel if calorimetric calculated power exceeds NIS indicated power by more than + 2% RTP." His proposed change is identical to the correspondmg Improved Technical Specifications revision, which follows.

%e CTS Bases are not impacted by this change.

In that Farley desires to implement the proposed changes as soon as practical, marked-up and typed CTS pages for Farley Units 1 and 2 are provided in Attachment I of this amendment request.

Imoroved Technical Soecifications By letter dated March 12,1998, Farley submitted a request to convert to the Improved Technical Specifications (ITS). %e proposed Power Range surveillance change impacts Surveillance Requirement (SR) 3.3.1.2 of the ITS submittal, which states, " Compare results of calorimetric heat balance calculation to Nuclear Instrumentation System (NIS) channel output." SR 3.3.1.2 Note No. I states, " Adjust NIS !

channel if absolute difference is > 2%." SR 3.3.1.2 Note No. 2 states, "Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is h 15% RTP." His ITS surveillance is only applicable to RTS Function No. 2.a, Power Range Neutron Flux High (see Table 3.3.1-1).

He proposed ITS change will revise SR 3.3.1.2 Note No. I to state, " Adjust NIS channel if calorimetric calculated power exceeds NIS indicated power by more than + 2% RTP."

An ITS Bases change is also required. The Bases change provides a summaryjustification for the surveillance change and clarifies when channel adjustments must be made. Specifically, the first paragraph of Bases SR 3.3.1.2 will be revised as follows.

"SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS channel output every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If the calorimetric calculated power exceeds the NIS channel indicatad power by more than + 2% RTP, the NIS channel is not declared inoperable, but must be adjusted. If the NIS channel output cannot be properly adjusted, the channel is declared inoperable."

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i The following two paragraphs will be inserted between the first paragraph and the second paragraph of Eases SR 3.3.1,2.

"If the calorimetric is performed at part power (< 50% RTP), adjusting the NIS channel indication in the increasing power direction will assure a reactor trip below the safety analysis limit (s 118% RTP).

Making no adjustment to the NIS channel in the decreasing power direction due to a part power

, calorimetric assures a reactor trip consistent with the safety analyses.

This allowance does not preclude making indicated power adjustments, if desired, when the calorimetric calculated power is less than the NIS channel indicit~i power. To provide close agreement between indicated power and to preserve operating margin, the NIS channels are nornally adjusted when operating at or near full power during steady-state conditions. However, discretion must oc exercised if the NIS channel indicated power is adjusted in the decreasing power direction due to a part power calorimetric (<

50% RTP). This action could introduce a non-conservative bias at higher power levels which could result in an NIS reactor trip above the safety analysis limit ( > 118% RTP). The cause of the non-conservative bias is the decreased accuracy of the calorimetric at reduced power conditions, as discussed in Westmghouse Technical Bulletin, ESBU-TB-92-14-RI, 'Decalibration Effects Of Calorimetric Power l Level Measurements On The NIS High Power Reactor Trip At Power levels Less Than 70% RTP' (Ref. l 13). To assure a reactor trip below the safety analysis limit, the Power Range Neutron Flux - High l bistables are set s 85% RTP: 1) whenever the NIS channel indic=*~4 power is adjusted in the decreasmg l power direction due to a part power calorimetric below 50% RTP; and 2) for a post refueling startup. l Before the Power Range Neutron Flux - High bistables are re-set s 109% RTP, the NIS channel calibration must be confirmed based on a calorimetric performed 2 50% RTP."

1he second paragraph of Bases SR 3.3.1.2 will be revised for consistency with the surveillance wording changes proposed in the first paragraph. In addition, the Bases information pertaining to the basis for not requiring performance of a ==aA=y power calorimetric measurement until reaching 15% RTP is being changed to reflect the correct licensing basis for Wdinghse PWR's. That is,15% RTP was chosen as the minimum power level for the NIS Power Range daily surveillance based on the Westinghouse NSSS design basis capability requirement of being able to achieve stable control system operation in the automatic control mode. The revision is as follows.

'Two Notes modify SR 3.3.1.2. The first Note indicates that the NIS channel output shall be adjusted consistent with the calorimetric calculated power if the calorimetric calculated power exceeds the NIS channel output by more than + 2% RTP. The second Note clarifies that this Surveillance is required only if reactor power is 215% RTP and that 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed for performing the first Surveillance after reaching 15% RTP. A power level of 15% RTP is chosen based on plant stability, i.e., automatic rod control capability and turbine generator synchronized to the grid."

1 The third paragraph of Bases SR 3.3.1.2 will also be revised for consistency with the surveillance wording changes proposed in the first paragraph as follows.

"The Frequency of every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is adequate. It is based on unit operatmg experience, considering instrument reliability and operatmg history data for instrument drift. Together these factors demonstrate that a difference between the heat balance calculated power and the NIS channel indication of more than

+ 2% RTP is not expected in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period."

The fourth paragraph of Bases SR 3.3.1.2 will not be revised nisprt:5.mse A - III - 4 10/29/98

,He subject Westinghouse Technical Bulletin will be added to the ITS Bases References as Reference No.

13. %c in'scrt is, "13. Westmghouse Technical Bulletin, ESBU-TB-92-14-RI, 'Decalibration Effects Of Calorimetric Power Level Measurements On ne NIS High Power Reactor Trip At Power Levels Less Than 70% RTP.'"

He ITS marked-up and typed pages are included in At:aciunent II of this submittal. %csc pages are based on the " clean typed copy" page 3.3.1-9 (ITS Volume 11) and " clean typed copy" pages B 3.3.1-50 and B 3.3.1-57 (ITS Volume 12), which were submitted by SNC letter dated April 24,1998. Due to text additions, pages B 3.3.1-50 through B 3.3.1-57 will be replaced and pages B 3.3.2-58 and B 3.3.1-59 will be added Review and approval of this licensing amendment change request is applicable to the ITS version as well, which will be incorporated into the ITS submittal when approved. In that Farley is the lead plant for WOG, the proposed ITS Technical Specifications arxi Bases changes are also applicable to NUREG-1431.

Ooerational and Safety Analyses Consideratigis When gain adjustments are perfonned on a power escalation, the NIS PR daily surveillance results in the NIS channel reflectmg the calorimetric calculated power with increasing accuracy up to approximately 100% RTP, When gain adjustments are performed at steady-state 100% RTP conditions, the NIS PR daily surveillance will adjust the PR channel for variations in iMiaW power due to changes in core power distributions with increasing burnup.

Normally, adjustment of the NIS channel indicated power in the decreasing power direction will be performed for operational reasons, such as, when operating at 100% RTP to restore operational margin to trip. Another example is when decreasing power and approaching Permissive P-10 reset (which automatically reinstates the PR High Neutron Flux Im Setpoint reactor trip) and there is a mismatch between NIS Power Range and NIS Intermediate Range iMimtM power levels. Adjustment ofindicated power in the decreasing power direction to more closely match the calonmetric calculated power may result in a closer agreement between the NIS Power Range and Intermediate Range channels, thus decreasing the possibility of an adverse interaction.

To ensure that the Power Range High Neutron Flux High Setpoint reactor trip signal will be generated prior to the safety analysis limit of 118% RTP, should operatmg conditions require that indicated power be decreased to match calculated calorimetric power based on data obtained below 50% RTP, Farley operatmg procedures will continue to specify that t5e PR High Neutron Flux High reactor trip setpoint be reduced to s; 85% RTP on all channels. He proposed ITS Bases change includes this administrative control requirement.

ANALYSIS He purpose of this analysis is to assess the impact of the proposed NIS Power Range surveillance change on the licensing basis and demonstrate that the. change will not adversely affect the subsequent safe operation of the plant, nisprts5.mse A - III - 5 10/29/98

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NIS Power Ranne Indication and RTS Functions Wiien operhting above 15% RTP, each Power Range channel is nomubed (i.e., calibrated) daily to match the thermal power calculation results based on the cernadary heat balance (i.e., calorimetric). The calibration is accomplished by adjusting the gain of each channel sumnung amplifier, such that the indi"*M power matches the calorimetric pos - The amplifier output (0% to 120% RTP) provides the input signals to the associated channel reactor trip, permissive and control interlock bastables, and the associated power indicators. Therefore, the proposed change to the NIS Power Range daily surveillance j potentially impacts the PR indications, RTS fLactions, control system functions, and miscellaneous alarm functions. These functions include: High Flux High Setpoint, High Flux Low Setpoint, High Positive Rate and High Negative Rate Reactor Trips; Permissives P-8, P-9 and F-10; Control Interlock C-2 (i.e., PR High Flux Rod Stop); automatic Reactor Contro! System nuclear power input; and PR Channel Desiation, Quadrant Power Tilt Ratio, and N-16 Leakage Detection System alarms Reactor power is monitored by the plant operators to ensure that the unit is operated within the limits of the Facility Operating License and safety analyses. The revision to the criteria for implementation of the daily surveillance will have a conservative effect on the PR channel indication (i.e., indicated power will be greater than actual power). With regard to the core safety limits, reactor power is one of four operating parameters with uncertainties explicitly used in the Revised Thermal Design Procedure (RTDP). The RTDP and safety analyses assume a reactor power uncertainty ofi 2% RTP. Farley-specific calculations presented in WCAP-12771, Revision 1, "Wes+inghn- Revised Thermal Design Procedure Instrument Uncertainty Methodology for Alabama Power Farley Nuclear Plant Units 1 and 2 (Upratmg to 2785 Mwt NSSS Power)," demonstrate that the secondary side power calorimetric measurement uncertainty at full power conditions is less than the RTDP assumption. Since the Farley-specific uncertainty calculation is not invalidated by the proposed PR surveillance method change, the PTDP and safety analyses reactor power uncertainty assumption ofi 2% RTP continues to be a hanading allowance for the core safety limits and safety analyses. Therefore, the NIS Power Range indications are not adversely impacted by the Pmposed change Farley-speciSc calculations have been perfornied for the followmg Power Range RTS functions: High Neutron Flux High Setpoint and High Neutron Flux Low Setpoint Reactor Trips; and Permissives P-8, P-9 and P-10. h calculation assumpticas account for the, daily PR calibration specified by the Technical Specifications. The setpoint uncertainty calculations demna trate conservative margm between the associated Technical Speci6 cations nommal trip setpoints and, when applicable, the correspondmg safety analysis limits. Since the daily calibration will continue to be performed and the maximum non-conservative error (i.e., when indiaW power is less than calorimetric power) will be 1; 2% PTD, the PR setpoint calculations, setpoints, and applicable safety analysis limits are not affected by the sur eillance change. With resper:t to the PR High Positive Rate and High Negative Reactor Trips, these trip functions are generated by time-delay rtlative-comparison circuits. As such, the NIS PR rate trips are not affected by the proposed change One potential non-conservative impact on the NIS RTS functions is evaluated herein. If the channel indication is gicater than the calorimetne power during a unit shutdown, the proposed change could delay the reset of Permissive P-10. Reset of P-10 (= 8% RTP) is required to enable the PR High Neutron Flux Iow Setpomt and IR High Neutron Flux reactor trips, which afford reactor protection for uncontrolled reactivity excursions from subcritical and low power (i.e., < 10% RTP). It is unlikely that a subcritical condition would be achieved before P-10 would reset. Nevertheless, ifindia'M power is greater than calorimetric power by a sufficient magnit"da (resulting in subcriticality without P-10 reset), the time duration natil P-10 reset would be very short. During this brief time interval, the PR High Neutron Flux High Setpoint reactor trip would provide core protection, as demonstrated by event specific analyses. Diverse protection is also afforded by the PR High Positive Rate, OTAT and OPAT reactor

( trips. Therefor:, the Power Range RTS functions are not adversely affected by the proposed change.

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! He Power Range input functions to the Reactor Control System are: Control Interlock C-2 (i.e., PR High Flyx Rod Stop), which blocks automatic and manual control rod withdrawal; and the nuclear power input signal (PR*44) to the power mi===*eh circuits associated with automatic reactor coolant system i temperature control. Rese are control system functions that are not required for safety (FSAR Chapter 7.7). Nevertheless, the proposed PR surveillance change continues to limit the maxunum allowed non-

conservative calibration error; therefore, the change will not adversely impact the NIS Power Range control system fi=ctinaa.

, Miscellaneous alarm functions also use input signals from the NIS Power Range channel (s). De functions j are: PR Channel Deviation, Quadrant Power Tilt Ratio (QPTR); and N-16 Ieakage Detection System.

. The Channel Deviation and QPTR alarms are generated by companson of the PR channel output signals. j In that these are relative compansons between channels, these alarm functions are not adversely affected by l j the ' proposed daily calibration change.

3 The N-16 Leakage Detection System associated with steam line radiation monitors R-70A, B, C may be

! - impacted by the proposed change since the proposed calibration change allows indicated power to be greater than calonmetric power. When greater than 20% power, the N-16 leakage Detection System i provides a continuous trend of the estimated " power-corrected" primary-tesaenad=ry leak rate, and it generates control room alarms if the leak rate increases above three threshold levels (alert, high, high-high).

De nuclear power signal is provided from NIS channel PR-43. A potential non-conservative impact on the leakage detection system is acceptable based on the followmg

1. %c N-16 May Detection System is a non-safety-related indication system that is considered to be an operational aid.

2. Other radiation monitors, such as the air ejector and steam generator blowdown monitors, provide diverse continuous primary-to-secondary leakage indication.

3. Reactor Coolant System leakage is periodically monitored by performance of the surveillance tests required by the Technical Speci6 cations

4. Actual primary-to-sec( ndary leak rates are determined by radiaehami*y analysis in accordance with plant procedures.

5. Normally, when operating at or near full power, PR-43 will be adjusted on a daily basis to match indicated power with calorimetric power. This plant practice results in the optimum channel calibration LOCA and LOCA-Fal='ad Aaalvses The following LOCA und LOCA relatM analyses are not adversely affected by the proposed modification of NIS Power Range daily survedinace: large and small break LOCA; reactor vessel and loop LOCA blowdown forces; post-LOCA long term core coohng subcriticality; post-LOCA long term core cooling mmamum flow; and hot leg switchover to prevent boron precipitation ne proposed madine=' ion does not effect the normal plant operatmg parameters, the safeguards systems actuation or accident mitigation capabilities important to LOCA mitiption, or the - sa-a': nan used in the LOCA-related accidents. The !

surveillance change does not create conditions more limiting than those ====ad in these analyses. In addition, the proposed moddication does not affect the Steam Generator Tube Rupture (SGTR) ana; :s methodology or assumptions, and it does not alter the SGTR event analysic results.

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Non-LOCA Related Analyses 4

l h non-LOCA safety analyses presented in Chapter 15 of the FSAR are not adversely affected by the i proposed NIS Power Range surveillance moddication %is modification does not affect normal plant

operatmg parameten, accident mitigation cyabilities, the assumatiana used in the non-LOCA transients, j or create conditions more limiting than those enveloped by the current non-LOCA analyses, brefore, the i conclusions presented in the FSAR remain valid.

Mechanical Components and Svetem_=

h surveillance modification as described does not affect the reactor coolant system coinponcet integrity or tre ability of the system to perform its intended safety function. W modtfication u described does not affed the integrity of a plant auxthary fluid system or the ability of the auxiliary systems to perform their !

design fhac*iaan.

l I&C Protection and Control Sv=*~a=

With the specific exception of the NIS Power Range reactor trip and indication functions, the proposed NIS Power Range daily surveillance change does not directly or indirectly involve additional electrical systems, m=; ==ts, or instrumentation considerations. Direct effects as well as indirect effects on equipment important to safety have been considered. Indirect effects include conditions or activities which involve non-safety-related electrical equipment which may affect Class IE, PAMS, or plant control electrical equipment. Consideraten has been given to seismic and envirnamaa*=8 qa=Hhtion, design and performance criteria per IEEE standards, Frae*ia==1 requirements, and plant Technical Specifications.

l h proposed change does not affect the plant normal operating design transients, margin to trip analysis, or low i .winure overpressure protection system l

An evaluatson herein determined that the proposed survedlance modification will ensure the performance of !

the NIS Power High Neutron Flux High Lepaint reactor trip function consistent with the safety analysis l assumptions. Deletion of the requirement to adjust the NIS Power Range channel (s) when iadic=*~I power is greater than calorimetric calculated power allows the channel (s) to not be adjusted in the non-conservative direction at part power. His allowance prevents the introduction of an error that has not been accounted for in the setpoint uncertamty calculations and the safety analyses associated with the NIS Power Range High Neutron Flux High Setpoint reactor trip fhac*iaa. Ifindicated poweris decreased to match a part power calonmetric poifvin.ed below 50% RTP, plant administrative controls ensure tae PR High Neutron Flux High Setpoint is reduced to s 85% RTP. Thus, the proposed madiheiaa does not have a pn*=*ial for identifie=*ian of an unreviewed safety question as it would relate to the safety-related function ofIAC systems.

RTS and ESFAS henaia*=

With the specific exception of the NIS Power Range indication and reactor trip functions, the proposed modification to the Power Range daily surveillance, does not affect the Reactor Trip System (RTS.) or the F=pW Safety Feature Arenatian System (ESFAS) =a*pa:a*= nis proposed modification does not change the current trip setposts or instrument operability requirements identified in the Technical mispt:5.mse A - III - 8 10/29/98

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Specifications. .h modi 6 cation should ensure the operability of the NIS Power Range reactor trip at part power conditions after normahntion at 100% RTP conditions consistent with the safety analysis assumptions. brefore, the proposed mochfication has no effect on the RTS and ESFAS safety functions.

Other Safetv-Related Areas and Analyses

& following safety-related areas and analyses are not affected by the proposed surveillance modification:

Contamment Integrity Analyses (Short Term /Long Term LOCA Release); Main Steamhne Break (MSLB)

Mass and Energy Release; Radiological Analyses; Probabilistic Risk Assessment; and Emergency Response Procedures.

SUMMARY

/ CONCLUSION

'Ihe proposed Technical Specifications change modi 6es the Power Range daily Surveillance Requirement by only requiring a calibration adjustment when PR indicated power is less than the calculated ma ry celorimetric power by > 2% RTP. The detailed analysis presented herein assessed the potential impact of ik proposed daily surveillance change on applicable Farley safety analyses and NIS Power Range

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indications, RTS functions, and control system functions. W assessments h=trated that the change will not adversely affect the Farley design basis safety analyses, Power Range functions, or the subsequent safe operation of the plant.

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ATTACHMENT IV 1

i 10 CFR 50.92 SIGNIFICANT HAZARDS EVALUATION !

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i JOSEPH M. FARLEY NUCLEAR PLANT l l

NIS POWER RANGE CHANNEL DAILY SURVEILLANCE .

I REQUIREMENT TECHNICAL SPECIFICATIONS CHANGE i

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10 CFR 50.92 SIGNIFICANT HAZARDS EVALUATION

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JOSEPH M. FARLEY NUCLEAR PLANT NIS POWER RANGE CHANNEL DAILY SURVEILLANCE REQUIREMENT TECHNICAL SPECIFICATIONS CHANGE INTRODUCTION When operating above 15% Rated 'Ibermal Power (RTP), the current Farley Technical Specifications Nuclear Instrumentation System (NIS) Power Range daily Surveillance Requirement requires the adjustment of the Power Range channel (s) when the absolute difference between Power Range india'ad power and whry side calorimetric power is greater than 2%

RTP. Compliance with this Technical Specifications requirement may result in a non-conservative channel calibratiou during reduced power operations. & proposed Technical Specifications change will address the potential non-conservatism by requiring adjustment of the NIS Power Range channel (s) only when the calorimetric calculat;d power is greater than the Power Range india'ad power by + 2%. In the interim, to ensure compliance with the Technical Specifications and conformance with the safety analyses, Farley has implemented administrative controls.

& potential non-conservative NIS calibration issue and the proposed Technical Specifications change are applicable to other Westmghouse plants. Farley is the lead plant for the Westmghouse Owners Group (WOG).

PROPOSED CHANGE Surveillance Requirement 4.3.1.1 of the Current Technical SpeciSc.ations (CTS) requires each Reactor Trip System (RTS) instrumentation channel to be demonstrated operable by performance of the channel checks, calibrations and fharti-1t ests specified in Table 4.3 L M NIS Power Range (PR) daily calibration (i.e., surveillance requirement) is found in Table Notation (2). To prevent nn undesired Power Range decalibration at part power operatmg conditions, the proposed chang .ill not hequire a channel adjustment when indiatad power is greater than calorimetric power. Nevertheless, to ensure that the NIS PR channels are adjusted when indiatad power is less than calorimetric power by greater than 2% RTP and to preserve the safety analyses assumptions, the proposed CTS Table 4.3-1 Note (2) revision will state, " Heat balance only, above 15% of RATED THERMAL POWER (RTP). Adjust NIS channel if calorimetric calculated power exceeds NIS i Ucated pewer by more than + 2% RTP."

h proposed change to the criterion for implementation of the Power Range daily surveillance is also applicable to Surveillance Requirement (SR) 3.3.1.2 Note No. I end the associated ITS Bases of the Farley Improved Technical Specifications (ITS) submittal (reference SNC letters dated March 12,1998 and Apr3 24,1998). Review and approval of this licensing amendment change request is applicable to the ITS version as well, which will be incorporated into the ITS submittal when approved.

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. 10,CFR 50.92 EVALUATION When operatmg above 15% RTP, each Power Range channel is normalized (i.e., calibrated) daily to match the thermal power calculation results based on the =eendary heat balance (i.e.,

calorimetric). The calibration is accomplished by adjusting the gain of each channel summing amplifier, such that the indic=W power matches the calorimetric power. The amplifier output (0%

to 120% RTP) provides the input signals to the associated channel reactor trip, permissive and control interlock bistables and the associated power level indications. Therefore, the proposed change to the NIS Power Range daily surveillance potentially impacts the PR indications, RTS functions, control system functions, and miscellaneous alarm functions. These functions include:

"% Flux High Setpoint, High Flux Low Setpoint, High Positive Rate, and High Negative Rate Reactor Trips; Permissives P-8, P-9 and P-10; Control Interlock C-2 (i.e., PR High Flux Rod Stop); automatic Reactor Control System nuclear power input; and PR Channel Deviation, Quadrant Power Tilt Ratio, and N-16 Leakage Detection System alarms

& detailed analysis presented in Att=chawat III nece==ed the potential impact of the proposed daily surveillance change on applicable Farley safety analyses and NIS Power Range indications, RTS functions, and control system functions. b analysis also assessed the potential impact on the PR High Neutron Flux High Setpoint reactor trip functon and the associated safety analysis limit when channel adjustments are made during specific operatmg conditions, & assessments demaa=trated that the proposed CTS and ITS changes will not adversely affect the Farley design ,

basis safety analyses, NIS Power Range safety functions, or the subsequent safe operation of the l plant.

As required by 10 CFR 50.91 (a)(1), an analysis has been provided to dcincestrate that the proposed license amendment revising Technical Specifications NIS Power Range daily Surveillance Requirement does not involve a significant hazards consideration. The analysis suppcits the following conclusions with respect to 10 CFR 50.92.

1. Does the proposed surveillance change involve a significant increase in the probability or consequences of an accident previously evaluated?

'Ibe proposed surveillance change does not signiScently increase the probability or consequences of an accident previously evaluated in the FSAR. This modification does l not directly initiate an accident. The consequences of accidents previously evaluated in the l FSAR are not adversely affected by this proposed change because the change to the NIS Power Range ch:.mel adjustment requirement ensures the conservative response of the channel even at part power levels.

2. Does the proposed surveillance change create the possibility of a new or different kind of accident from any accident previously evaluated?

h proposed surveillance change does not create the possibility of a new or different kind of accident than any accident already evaluated in the FSAR. No new accident scenarios, failure mechamsms, or limiting single failures are introduced as a result of the proposed change The proposed Technical Spacincitions change does not challenge the performance or integrity of any safety-related systems hrefore, the possibility of a new or different Had of accident is not creatad nasprt:5.mse A - IV - 2 08/27/98

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. . 3 , ,, Does the proposed surveillance change involve a significant reduction in a margin of l safety? l h proposed surveillance change does not involve a significant reduction in a margin of safety. De proposed change does require a revision to the criterion for implementation of Power Range channel adjustment based on secondary power calorimetric calculation; however, the change does not eliminare any RTS surveillances or alter the frequency of surveillances required by the Technical Speci6 cations. W revision to the criterion for implementation of tlw daily surveillance will have a conservative effect on the performance of the NIS Power Range channel, particularly at part power after normahzation at 100%

RTP conditions. The nominal trip setpoints specified by the Technical Specifications and the safety analysis limits assumed in the transient and accident analysis are unchanged.

He margin of safety associated with the acceptance criteria for any accident is unchanged.

hrefore, the proposed change will not significantly reduce the margin of safety as defined in the Technical Specifications.

CONCLUSION Based on the precedmg information, it has been determined that the proposed change to the NIS Power Range daily surveillance does not involve a signi6 cant hazards consideration as defined in 10 CFR 50.92 (c).

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Text

SUMMARY

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