ML20058E803

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Proposed Tech Specs,Relocating Reactor Trip & ESF Actuation Sys Response Time Limits to FSAR
ML20058E803
Person / Time
Site: Farley  Southern Nuclear icon.png
Issue date: 11/24/1993
From:
SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20058E796 List:
References
NUDOCS 9312070177
Download: ML20058E803 (52)
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Text

'

P Joseph M. Farley Nuclear Plant' - Units 1 and 2 Technical Specification Changes Associated with It,0 cation of Reactor Trip and Engineered Safety Feature Actuation System Response Time Limits Pace Chanae Instructions t

Unit 1

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Remove Paae Insert Pace 3/4 3-1 3/4 3-1 j

3/4 3-10 3/4 3-10 3/4 3-11 3/4 3-11 3/4 3-15 3/4 3-15 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-33 8 3/4 3-1 B-3/4 3-1 B 3/4 3-2 B 3/4 3-2 i

Unit 2 Remove Pace Insert Pace 3/4 3-1 3/4 3-1 3/4 3-10 3/4 3-10 3/4 3-11 3/4 3-11 3/4 3-15 3/4 3-15 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32 B 3/4 3-1 B 3/4-3-1 f

B 3/4 3-2 B 3/4 3-2 l

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3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION i

i 3.3.1 - As a minbnum, the reactor trip system instrumentation channels and -

interlocks of Table 3.3-1 shall be OPERABLE.

l APPLICABILIIY1 As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1 SURVEILLANCE RE0VIREMENTS r

i 4.3.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Taole 4.3-1.

4.3.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days. The total j

interlock function shall be demonstrated OPERABLE at least once per 18 months.

4.3.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months.*.

l Each test shall _ include at least one logic train such that both logic trains are tested et least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as

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shown in the " Total No. of Channels" column of Table 3.3-1.

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Neutron detectors are exempt from response time testing.

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. FARLEY-UNIT 1 3/4 3-11

. AMENDMENT NO.

4 INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shewn in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip setpoint column _ of Table 3.3-4.

l_

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a.

With an ESFAS instrumentation channel or interlock trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2.

4.3.2.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at least once per 18 months.

4.3.2.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each-ESFAS function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one logic train such that both logic trains are te'sted at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" column of Table 3.3-3.

FARLEY-UNIT 1 3/4 3-15 AMENDMENT NO.

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I FARLEY-UNIT 1 3/4 3-32 AMENDMENT NO.

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i 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip and Engineered Safety Feature Actuation System instrumentation and interlocks ensure that 1) the associated Engineered Safety Feature Actuation action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, 2) the specified coincidence logic'is maintained, 3) suf-ficient redundancy is maintained to permit a channel to be out of service'for testing or maintenance, and 4) sufficient system functional capability.is available for protective and ESF purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses. The surveillance requirements-specified for these systems ensure that the overall system functional capa-bility is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

The Engineered Safety Features Actuation System interlocks perform the functions indicated below on increasing the required parameter, consistent with the setpoints listed in Table 3.3-4:

P-11 Defeats the manual block of safety injection actuation on low pressurizer pressure.

P-12 Defeats the manual block of safety injection actuation on low steam line pressure.

P-14 Trip of all feedwater pumps, turbine trip, closure of feedwater isolation valves and inhibits feedwater control valve modulation.

On decreasing the required parameter the opposite function is performed at reset setpoints, with the exception' of P-12 as noted below:

P-12 Allows manual block of safety injection actuation on low steam line pressure. Causes steam line isolation on high steam flow. Affects steam dump blocks (i.e., prevents premature block of the noted function).

t The measurement of response time at the specified frequencies provides i

assurance that the reactor trip and ESF actuation associated with each channel is completed within the time limit assumed in the accident analyses, i

Response time limits for the Reactor Trip System and Engineered Safety Features Actuation System are maintained in Tables 7.2-5 and 7.3-16 of the Farley FSAR, respectively.

No credit was taken in the analyses for those channels j

with response times indicated as not applicable.

F FARLEY-UNIT 1 B 3/4 3-1

' AMENDMENT NO.

INSTRUMENTATION BASES 9

REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION-SYSTEM

[

INSTRUMENTATION (Continued)

Response time may be demonstrated by any series of sequential, overlapping j

or total channel test measurements provided that such tests demonstrate the

't total channel response time as defined.

Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or

2) utilizing replacement sensors with certified response times.

3/4.3.3 MONITORING INSTRUMENTATIOJ 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures _that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

t Alarm / trip setpoints for the containment purge have been established for a purge rate of 5,000 scfm in all MODES and for purge rates of 25,000 scfm and 50,000 scfm in MODES 4, 5, and 6. -_The containment purge setpoints are based on a release in which Xe-133 and Kr-85 are the predominant isotopes, on the 5.6 x 10', Appenjix B, Table 2, MPC values for these isotopes and on a X/Q of 10 CFR 20 sec/m at the site boundary.

The Alarm / trip setpoint for the fuel storage pool area has been established I

based on a flow rate of 13,000 scfm; a release in which Xe-133 and Kr-85 are the predominant isotopes, on the 10 CFR 20, pppend{x B, Table 2, MPC values for these isotopes and on a X/Q of 5.6 x 10~ sec/m at the site boundary.

3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating-each detector used and determining the acceptability of its voltage curve.

For the purpose of measuring F (Z), F [#g,WCAP-8El8, June 1976 and F a full incore flux map o

is used.

Quarter-core flux maps, as define in used in recalibration of the excore neutron flux detection system.

Full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.

FARLEY-UNIT 1 B 3/4 3-2 AMENDMENT N0.

l 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.

l APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1 SURVEILLANCE RE0VIREMENTS 4.3.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown.

in Table 4.3-1.

4.3.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days.

The total interlock function shall be demonstrated OPERABLE at least once per 18 months.

4.3.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months.

l-Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.

Neutron detectors are exempt from response time testing.

l FARLEY-UNIT 2 3/4 3-1 AMENDMENT NO.

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INSTRUMENTATION j

3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION j

3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation.

channels and interlocks shown in Table 3.3-3 shall be OPERABLE.with their trip setpoints set consistent with the values shown in the Trip Setpoint column of' Table 3.3-4.

l.

APPLICABILITY: As shown in Table 3.3-3 ACTION:

a.

With an ESFAS instrumentation channel or interlock trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent l

with the Trip Setpoint value.

b.

With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.

i SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shewn'in Table 4.3-2.

4.3.2.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at.least once per 18 months.

4.3.2.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one logic train such that both logic _ trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in r

the " Total No. of Channels" column of Table 3.3-3.

4 FARLEY-UNIT 2 3/4 3-15 AMENDMENT N0.

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FARLEY-UNIT 2 3/4 3-32 AMENDMENT NO.

t 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip and Engineered Safety Feature Actuation System instrumentation and interlocks ensure that 1) the associated Engineered Safety Feature Actuation action ar.d/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, 2) the specified coincidence logic is maintained, 3) suf-ficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4)' sufficient system functional capability is available for protective and ESF purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses. The surveillance requirements

~

specified for these systems ensure that the overall system functional-capa-bility is maintained comparable to the original design standards.

The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

The Engineered Safety Features Actuation System interlocks perform the functions indicated below on increasing the required parameter, consistent with the setpoints listed in Table 3.3-4:

P-11 Defeats the manual block of safety injection actuation on low pressurizer pressure.

P-12 Defeats the manual block of safety injection actuation on low steam line pressure.

P-14 Trip of all feedwater pumps, turbine trip, closure of feedwater i

isolation valves and inhibits feedwater control valve modulation.

On decreasing the required parameter the opposite function is performed at reset setpoints, with the exception of P-12 as noted below:

P-12 Allows manual block of safety injection actuation on low steam line pressure.

Causes steam line isolation on high steam flow. Affects steam dump blocks (i.e., prevents premature block of the noted function).

The measurement of response time at the specified frequencies provides assurance that the reactor trip and ESF actuation associated with each channel is completed within the time limit assumed in the accident analyses.

Response time limits for the Reactor Trip System and Enginrereo rafety Features Actuation System are maintained in Tables 7.2-5 and 7.3-1b of the Farley FSAR, respectively. No credit was taken in the analyses for those channels with response times indicated as not applicable.

FARLEY-UNIT 2 B 3/4 3-1 AMENDMENT NO.

I

'I

INSTRUMENTATION l

t BASES i

REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (Continued) l Response time may be demonstrated by any series of sequential, overlapping.

I or total channel test measurements provided that such tests demonstrate the-total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or

2) utilizing replacement sensors with certified response times.

3/4.3.3-MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual.

channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

Alarm / trip setpoints for the containment purge have been established for 8

a purge rate of 5,000 scfm in all MODES and'for purge rates of 25,000 scfm and 50,000 scfm in MODES 4, 5, and 6.

The containment purge setpoints are based-on a release in which Xe-133 and Kr-85 are the predominant isotopes, on-the 5.6 x 10', Appenjix B, Table 2, MPC values for these isotopes and on a X/Q of 10 CFR 20 sec/m at the site boundary.

The Alarm / trip setpoint for the fuel storage pool area has been established based on a flow rate of 13,000 scfm; a release'in which Xe 133 and Kr-85 are for these isotopes and on a X/Q of 5.6 x 10'pppend{x B, Table 2, MPC values j

the predominant isotopes, on the 10 CFR'20, sec/m at the site boundary.

3/4.3.3.2 MOVABLE INCORE DETECTORS f

The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of i

this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating -

each detector used and determining the acceptability of its voltage curve.

[

For the purpose of measuring Fo; define [$,WCAP-8N8, 'une 1976, i

Z),F and F a full incore flux map is used.

Quarter-core flux maps, as in l

used in recalibration-of the excore neutron flux detection system.

Full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.

i 6

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

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Hand-Marked Pages Technical Specifications Changes Associated with Relocation of Reactor Trip and Engineered Safety Feature l

Actuation System Response Time limits 4

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l lj 3/4. 3 INSTRUMENTATION

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3/4.3.'1 REACTOR TRIP SYSTEM INSTRUMENTATION i

l LIMITING CONDITION FOR OPERATION f

As a minimum, the reactor trip system instrumentation channels and 3.3.1 interlocks of Table 3.3-1 shall be OPERABLE." "re"^'er ""re t

APPLICASILITY: As shown in Table 3.3-1.

ACTION:

I As shown in Table 3.3-1.

l f

i SURVEILLANCE REQUIREMENTS t

i Each reactor trip system instrumentation channel shall be demonstrated l

4.3.1.1 l

OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1.

The logic for the interlocks shall be demonstrated OPERABLE prior to The total l

4.3.1.2 each reactor startup unless performed during the preceding 92 days.

interlock function shall be demonstrated OPERABLE at least once per 18 months.

j The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 sonths.*

l

4. 3.1. 3 Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.

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FARLEY-UNIT 1

-3/43-1 AMENDMENT N0..

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TABt.E 1;]-2

%s REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES E!INCTIONAL UNIT RESPONSE TIME i

T l

1.

Man \\ Reactor Trip Not Applicable Power \\

N i

2.

Ran g Neutron Plux a.

High f 0.5 seconds

  • b.

Lov Not Applicable M

\\

N 3.

Power Range, Neutro N 1ux, h

High Positive Rate Not Applicable 4.

Power Range, Neutron Plax, M

High Negative Rate Not Applicable l

w 5.

Intermediate Range, Neutron F1 Not Applicable w

6.

Source Range, Neutron Flux Not Applicable p

b 7.

Overtemperature AT f 6.0 seconds

  • 8 o

8.

Overpower er Not Applicable y

\\

i 9.

Pressurizer Pressure--Lov f 2.0 seconds g

10. Pressurizer Pressure--High f 2.0 seconds
11. Pressurizer Vater Level--High

\\ Not Applicable D'

\\ \\

g

\\

i c3

\\

x h
  • Neutron detectors are exempt from response time testing. Response time of the neutro ux signal portjon i

of the channel shall be measured from detector output or input of first electronic compo in channel.

2 l

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TABLE 3.3-2 (Continued)

REACTOR TRiff y 3.iEM INSTRUMENTATION RESPONSE TIMES

\\

\\

RESPONSE TIME FUNCTIONA 9

12.

A.

Loss o ow - Single Loop

< 1.0 seconds a

(Above P-135

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\\

h 8.

Loss of Flow - % Loops 1 1.0 seconds g

(Above P-7 and be g P-8)

13. Steam Generator Water leve ow-Low 1 2.0 seconds g

M

14. Steam /Feedwater Flow Mismatch and Het Appitcable L

Low Steam Generator Water Level d

1 1.2 seconds AC.

15. Undervoltage-Reactor Coolant Pumps
16. Underfrequency-Reactor Coolant Pumps 1 0.6 seconds s*

t M

17. Turbine Trip Not Appilcable S

A.

Low Auto Stop 011 Pressure Not Appilcable h

8.

Turbine Throttle Valve Closure 18.

Safety injection Input from ESF Not Appilcable b Appilcable 19.

Reactor Coolant Pump Breaker Position Trip Not (cable

20. Reactor Trip System Interlocks g

i Not App 11

21. Reactor Trlp Breakers Not Applicable \\
22. Automatic Trip Logic y

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INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4,:nd with P.:0?ONS: TIM;; e; ;hean in Tet'e 0.0 5.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

With an ESFAS instrumentation channel or interlock trip setpoint a.

less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL C.'.JCK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2.

4.3.2.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at least once per 18 months.

4.3.2.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.

A

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FARLEY-UNIT 1 3/4 3-15 AMENDMEET NO. ibE e

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TABLE 3.3-5

/[

ENGINEERED SCEfY FEATURES RESPONSE TIMES RESPONSETIMEINSECOND[

i INITIATING SICNAL AND FUNCTION NotApplicable!

1.

Manual Snfety Injection (ECCS) a.

Feedwater Isolation Not Applica Reactor Trip (SI)

Not Appli le Containment Isolation-Phase "A"

' Not Ap '

able Containment Purge Isolatlo:

Not licable No plicable

(

Auxiliary Feedwater Pumps I

Service Water System j

Applicable MiotApplicable Containment Air Coolers

!NotApplicable b.

Containment Spray Not Applicable Containment Isolation-Phase "B" Containment Purge Isolation Not Applicable c.

Containment Isolation-Phase "A" Not Applicable l

Containment Purge Isolation Not Applicable d.

Steam Line Isolation Not-Applicable i

2.

Containment Pressure-High i 27.0(I) f a.

Safety Injection (E

< 2.0 b.

Reactor Trip (from

< 32.0(6) c.

Feedwater Isolati n d.

Containment Iso tion-Phase "A" i 17.0(4}/27.0(5) e.

Containment P ge Isolation 1 5.0 f.

Auxiliary dwater Pumps Not Applicable 177.0(N/87.0(5) g.

Service ter System h.

Conta ent Air Cooler Fan i 27.4 i

/

/

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/

/

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FARLEY-UNIT 1 3/4 3-29 AMENDMENT NO. 26

i 2

4

  • )Nis PNA IA/T6AJTIDA]4LLYLEFT BL&ll TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES

[

?

/

INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SEC0fipS 3.

Pressurizer Pressure-Lov

[

Safety Injection (ECCS) f27.0'*'/12.0/

a.

b.

Reactor Trip (from SI) f 2.0

[

c.

Feedvater Isolation f 32.0

I d.

Containme.

Isolation-Phase "A" f 17.0

e.

Containment Purge Isolation f 5.0 f.

Auxiliary Feedvater Pumps Not Ap cable g.

Service Va*.er System f.77

    • /87.0'*'

((12.0/22.0

4.

Differential Pressure Between Steam Lines-High a.

Safety Injection (ECCS)

I b.

Reactor Trip (from SI) 1 2.0 c.

Feedvater Isolation 1 32.0

d.

Containment Isolation-Phase "A" f 17.0'*8/27.0

i e.

Containment Purge Isolation Not Applicable f.

Auxiliary Feedvater Pumps Not Applicable g.

Service Vater System 1 77.0'*8/87.0

~

5.

Steam Flow in Two Steam Lines oincident with T,y --Low-Lov a.

Steam Line Isolation Not Applicable l

t 6.

Steam Line Pressure-Lov i

a.

Safety Injection CCS) f 12.0/22.0 ts:

b.

Reactor Trip (

m SI) f 2.0 FeedvaterIs[ation

< 32.0

c.

d.

Containmen Isolation-Phase "A" f 17.0/27.0:s>

l e.

Contains t Purge Isolation Not Applicable f.

Auxill y Feedvater Pumps Not Applicable j

g.

Serv e Vater System j 77.0/87.0

b.

St m Line Isolation f 7.0 j

i

/

/

i FARLEY - UNIT 1 3/4 3-30 AhENDMENT NO. 26, 89,

}

l 92 l

~ -

Tes Psss hlThr/77cKlALL LEfr E k/d TABLE 3.3-5 (Continued) i ENGINEERED SAFET FEATURES RESPONSE TIMES

/

RESPONSETIMEINSECOND5 INITIATING SIGNAL AND FUNCTION

)

/(

7.

& ;ainment Pressure--High-H g I

a.

Steam Line Isolation 5 7.0 l

/

8.

Containment Pressure--High-High-High a.

Containment Spray 1 45.0 b.

Containment Isolation-Phase "B" Not A icable

((.5 9.

Steam Generator Water Level--High-Hich a.

Turbine Trip

/[5 32.0(6) b.

Feedwater Isolation 10.

Steam Generator Water Level -- Low-Low I2)

,i 60.0 a.

Motor-driven Aux. Feedwater Pumps

)

b.

Turbine-driven Aux. Feedwater P 1 60.0

11. Undervoltage RCP a.

Turbine-driven Aux. Feedwa i 60.0 12.

S.I. Sicnal a.

Motor-driven Auxilia 1 60.0 Feedwater Pumps

13. Trip of Main Feedwater Motor-driven Aux. Feelater Pumps Not Applicable a.

14.

Loss of Power a.

4.16 kv Emer ncy Bus Undervoltage (7)

(Loss of V age) b.

4.16 kv ergency Bus Undervoltage (7)

(Degra Voltage)

/

/

\\

/

/

\\

/

/

/

FARLEY-UNIT 1 3/4 3-31 AMENDMENT NO. 26

.a

}

~

\\

_ _ _.. -.. - ~ ~ ~

TWG Pbb /dT&dnodalL LWT &sVK TABLE 3.3-5 (Continued)

[!

~

TABLE NOTATION

., _. _ ~

3 (1) Diesel generator starting and sequence loading delays included.

Respon time limit includes opening of valves to establish SI path and attai of discharge pressure for centrifugal charging pumps, and RHR pumps.

(2) One 2/3 any Steam Generator (3) On 2/3 in 2/3 Steam Generators

[

(4) Diesel generator starting and sequence loading delay not inc ed.

Offsite power available. Respone time limit includes openi of valves to establish SI path and attainment of discharge pressure fo entrifugal charging pumps.

/

(5) Diesel generator starting and sequence loading delays neluded. Response.

time Ifmit includes opening of valves to establish S path and attainment of discharge pressure for centrifugal charging p (6) Verification shall include testing of all instru ntation, the isolation valves (MOV-3232A, 3232B, 3232C) and the control valves (FCV-478, 479, 488, 489, 498, 499). The isolation valves function within 30 seconds and the control valves within 5 seconds.

1 l

(7) The respone time shall include the ti -

associated with the

)

undervoltage relays as determined in

.3-4 plus an additional second associated with interposing nd circuit operation.

/

,/

/

/

\\

/

)

/

/

/

/

/

/

/

FARLEY-UNIT 1

-32 AMENDMENT NO. 26

)

+,

l

( SfbT&M M/D EAbsLl&&R&D SAf6N 03RDAkh TIM & LIMIT 3~ Fat IML-PfAC7bt ik/P INS M HENTATION

) ACTL/s77ad SWTEM M6Mst4//3rd&D Id TABU-S 7.2 5 sta ~l3-/G Oc 0%F' et.WF.tA2,35/WDk.K }9 j

BASES s

i i

REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM TRSIRUMENTAT3DN (Continued)

The measurement of response time at the specified frequencies providr3 assurance that the reactor trip and ESF actuation associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor responsc time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or

2) utilizing replacement sensors with certified response times.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level taip setpoint is exceeded.

Alarm / trip setpoints for the containment purge have been established for a purge rate of 5,000 scfm in all MODES and for purge rates of 26,.000 scfm and 50,000 scfm in MODES 4, 5, and 6.

The containment purge setpoints are based on a release in which Xe-133 and Kr-85 are the predominent isotopas, on the 10 CFR 20 AppengixB. Table 2,MPCvaluesfortheseisotopesandonaX/Qof 6

5.6 x 10 sec/m at the site boundry.

The Alarm / trip setpoint for the fuel storage pool area has been established t e, d on a flow rate of 13,000 scfm; a release in which Xe-133 and Kr-85 are thepredominentisotopes,onthe10CFR20,,gppendigB, Table 2,MPCvalues for these isotopes and on a X/Q of 5.6 x 10 sec/m at the site boundry.

3/a.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained fro.n use of this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.

N For the purpose of measuring F (Z), F

, and F,y a full incoCe flux map q

is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the excore neutron flux detection system. Full incore flux maps or symmetric incere thimbles may be used for. monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.

i FARLEY-UNIT 1 B 3/4 3-2 AMENDMENT NO. M

+

I 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING COND5 TION FOR OPERATION As a minimum, the reactor trip system instrumentation channels and 3.3.1 th:- '-

interlocks of Table 3.3-1 shall be OPERABLE.ef t' ?.E!?ONSE TIMEE ::

7:b1: 2.2-2.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REOUIREMENTS

4. 3~.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATIOk and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequenc1tc shown in Table 4.3-1.

I The logic for the interlocks shall be demonstrated OPERABLE prior to 4.3.1.2 The total each reactor startup unless performed during the preceding 92 days.

interlock function shall be demonstrated OPERABLE at least once per 18 months.

j The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip funeggen 4.3.1.3 shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the l

total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.

i 1

l i

W de Dent.1bes, och 6#664p7 FEcu RespodMTIME Teimk>.

l FARLEY-UNIT 2

. 3/4 b l AfMUDMEMT Ab.

4

(.

\\

TABLE 3.3-2 RFA ' T. TRIP SYSTEN INSTRUMENTATION RESPONSE TIMES c

s % 0NAL UNIT F

RESPONSE TIME 1.

Nanual g or Trip Not Applicable i

w 2.

Power Range, ron Flux g

.l a.

High

< 0.5 seconds

  • b.

Lov

\\

ot Applicable 3.

Power Range, Neutron g

i High Positive Rate Not Applicable s

4.

Power Range, Neutron Flux, C

High Negative Rate Not Applicable M

l

{

5.

Intermediate Rance, Neutron Flux Not Applicable 1

6.

Source Range, Neutron Flux Not Applicable y

7.

Overtemperature AT f 6.0 seconds *

,l l

E 1

j 8.

Overpover AT Not Applicable g

i i,

\\

g\\

]

f 2.0 seconds 9.

Pressurizer Pressure-Low f2 I

\\

.0 seconds k

10.

Pressurizer Pressure-High

11. - Pressurizer Vater Level-High

% Applicable M

\\g i

a E

N 8

  • Neutron detectors are exempt-from response time testing..ofthechannelshallbemeasuredfromdetectoroutputorinputoffirs Response time of the neutron f x t

hannel.

a N

4 e.a=

I l

TABLE 3.3-2 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES

  1. '*\\

Qe N

FUNCTI % L UNIT RESPONSE TIME

-e 12.

A.

f Flow - Single Loop

~

(Ab

-8)

$ 1.0 seconds B.

Loss of A - Two Loops

[g (AboveP-7a%belowP-8)

$ 1.0 seconds 13.

Steam Generator Wate q1--Low-Low i 2.0 seconds Steam /FeedwaterFlowMissab.and 14.

i Low Steam Generator Water Level Hot Applicable

_ 1.2 seconds th 15.

Undervoltage-Reactor Coolant P 16.

Underfrequency-Reactor Coolant Pumps 1 0.6 seconds sD M

h 17.

Turbine Trip y

A.

Low Auto'Stop 011 Pressure Not Applicable d

B.

Turbine Throttle Valve Closure Not Applicable 1

18.

Safety Injection Input from ESF

\\

Not Appilcable 19.

Reactor Coolant Pump' Breaker Position. rip ot Applicable k

h 20.

Ratctor Trip System Interlocks No licable

(

ri. Re.ctor Tri, reaxers Not App bie

22. Automatic Trip Logic Not Applica N

%N N N 4

w..--

..L.-.--- =-..--.

-e n.

ma.

m

,e.m,,ea.

s.

n t..c.--

,,,,v.-

,--e e

w s,

n.

_g j

I INSTRUMENTATION

_(.

'3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION i

3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.end-with Pi;PCM E T!MES :: ;h;wn in T;bi; 3.0-5, APPLICABILITY: As shown in Table 3.3-3.

ACTION:

With an ESFAS instrumentation channel or interlock trip setpoint a.

less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the-applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

i b.

With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS a

I 4.3.2.1 Each ESFAS instrumentation channel shall be demonstrate'd OPERABLE by.-

the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies _shown in Table 4.3-2, 4.3.2.2 The logic for the interlocks shal'1 be demonstrated OPERABLE during the automatic actuation logic test.

The total interlock function shall be--

demonstrated OPERABLE at least once per 18 months.

4.3.2.31 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one logic train such that both. logic trains are tested at least once per 36 months and one channel per function such that all' channels are tested at-least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in-the " Total No. of Channels" Column of Table 3.3-3.

~

e J

FARLEY-UNIT 2_

3/4 3-15

- Aur.uomeur uo.

5/5 fh66 /U1GA/770Alt>LL Vuf 7 &bt/K TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECO V

1.

Manual

/

a.

Safety Injection (ECCS)

NotApplicab/

Feedwater Isolation NotApplic[e Reactor Trip (SI)

NotAppl/able Containment Isolation-Phase "A" Not A icable Containment Purge Isolation Not plicable Auxiliary Feedwater Pumps N/ Applicable Service Water System

/otApplicable Containment Air Coolers

/NotApplicable b.

Containment Spray

/

Not Applicable Containment Isolation-Phase "B" Not Applicable Containment Purge Isolation Not Applicable c.

Containment Isolation-Phase "A' Not Applicable Conthinment Purge Isolation Not Applicable d.

Steam Line Isolation Not Applicable 2.

Containment Pressure-High a.

Safety Injection (EC 1 27.0(I) b.

Reactor Trip (from SI

< 2.0 FeedwaterIsolatig I32.0(6) c.

d.

Containment Is tion-Phase "A" 1 17.0(4)/27.0(5) e.

Containment ge Isolation 15.0 AuxiliarypdwaterPumps Not Applicable f.

ServicepterSystem 1 77.0(4)/87.0(5) g.

Contapent Air Cooler Fan i 27.4 h.

/

\\

/

/

~

/

/

/

FARLEY-UNIT 2 3/4 3-29 i.

J

t..

I THS P&s& lAframckJMLVUFr St.a/K

  • TABLE 3.3-5 (Continued)

/[

I

{.

ENGINEERED SAFETY FEATURES RESPONSE TIMES 7.NITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SE@NDS

[

3.

Pressurizer Pressure-Lov a.

Safety Injection (ECCS)

$27.0'*'/12.0[

b.

Reactor Trip (from SI) f 2.0

/

c.

Feedvater Isolation f 32.0

d.

Containment Isolation-Phase "A" f 17.0

e.

Containment Purge Isolation f 5.0 f.

Auxiliary Feedvater Pumps Not plicable g.

Service Vater System 3

.0/87.0'*'

4.

Differential Pressure Between Steam Lines-High a.

Safety Injection (ECCS) f f 12.0/22.0

b.

Reactor Trip (from SI) f 2.0 c.

feedvater Isolation f 32.0

d.

Containment Isolation-Phase "A" f 17.0/27.0

e.

Containment Purge Isolation Not Applicable f.

Auxiliary Feedvater Pumps Not Applicable 3

g.

Service Vater System

< 77.0/87.0

5.

Steam Flow in Two Steam Li Coincident with T,y--Low-Lov a.

Steam Line Isolati Not Applicable l

6.

Steam Line Pressure-

)

a.

Safety Inject (ECCS)

$ 12.0/22.0'58 b.

Reactor Tri from SI) f 2.0 I

c.

Feedvater olation f 32.0

d.

Contain t Isolation-Phase "A"

$ 17.0/27.0

e.

Conta:

ent Purge Isolation Not Applicable f.

Aux

..ary Feedvater Pumps Not Applicable g.

S ice Vater System f 77.0/87.0

q h./

team Line Isolation f 7.0

/

/

/

FARLEY - UNIT 2 3/4 3-30 AMENDMENT NO.

85

)

h l

.l-i

.==

Tus Rau titrextridsit ? (LFT duk TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SE S-7.

Containment Pressure--High-High a.

Steam Line Isolation 57.0

/

8.

Containment Pressure--High-High-High a.

Containment Spray 5 45 b.

Containment Isolation-Phase "B" No pplicable 9.

Steam Generator Water Level--High-High

[

!(($2.5 a.

Turbine Trip

$ 32.0(6) b.

Feedwater Isolation 10.

Steam Generator Water Level -- Low-Low a.

Motor-driven Aux. Feedwater Pum

)

< 60.0 b.

Turbine-driven Aux. Feedwater

)

5 60.0 11.

Undervoltage RCP a.

sine-driven Aux. Fee ump 5 60.0 12.

S.I. Signal a.

Motor-driven Auxili

< 60.0 Feedwater Pumps

~

13.

Trip of Main Feedwater Pu s Motor-driven Au/ Feedwater Pumps Not Applicable a.

14.

Loss of Power a.

4.16 kv Em ency Bus Undervoltage (7)

(Lossoffitage) b.

4.16kfmergencyBusUndervoltage (7)

(Degr ved Voltage)

/

/

p/

/

/

FARLEY-UNIT 2 3/4 3-31

Th95 SJ'66 /kl7bA/7MAll/J Vl.&f 7-Br.AA/E

~

TABLE 3.3-5 (Continued)

/

/[

TABLE NOTATION (I)

Diesel generator starting and sequence loading delays included.

esponse time limit includes opening of valves to establish SI path and #Ltainment of discharge pressure for centrifugal charging pumps, and RH umps.

(2)

On 2/3 any Steam Generator

/

(3)

On 2/3 in 2/3 Steam Generators (4)

Diesel generator starting and sequence loading de' not included. Off-site power available.

Response time limit inclu (

opiiiing of valves to establish SI path and attainment of discharge ssure for centrifigal charging pumps.

/

(5)

Diesel generator starting and sequence loa ng delays included. Response time limit includes opening of valves o establish SI path and attainment of discharge pressure for centrifu rging pumps, g

(6)

Verification shall include test 1 instrumentation, the isolation valves (MOV-3232A, 3232B, 3232 he control valves (FCV-478, 479, i

488,489,498,499). The iso valves must function within 30 seconds and the control valves wit conds.

(7) The response time shall the time delay associated with the undervoltage relays as ed in Table 3.3-4 plus an additional second associated with posing relay and circuit operation.

/

/

/

/

/

\\

/

/

/

/

/l

/

/

i FARLEY-UNIT 2 3/4 3-32

'g

f OS/hR$b llM& l/M//.T RW 7M-84f1DO T2/P f Wf7'6M AA/D GAlst&ceJD MWRadTtfths

_INSTRUHENTATION AC12/47/'QL/.$T$7t:ht A$GF hf4/A/7*AldfD 18/ TAA/J-f y,g 5 g,pg 7,3,3 g yggpyg7_,.ypfg,g3,y gfgg I

BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (Continued)

The measurement of response time at the specified frequencies provides assurance that the reactor trip and ESF actuation associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or

2) utilizing replacement sensors with certified response times.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels av e continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

Alarm / trip setpoints for the containment purge have been established for a purge rate of 5.,000 scfm in all MODES and for purge rates of 25,000 scfm and 50,000 scfm in MODES 4, 5, and 6.

The containment purge setpoints are based on a release in which Xe-133 and Kr-85 are the predominent isotopes, on the 10 CFR 29 Appengix B, Table 2, MPC values for these isotopes and on a X/Q of 6

5.6 x 10 sec/m at the site boundry.

The Alarm / trip setpoint for the fuel storage pool area has been established based on a flow rate of 13,000 scfm; a release in which Xe-133 and Kr-85 are thepredominentisotopes,onthe10CFR20,,gppendig8, Table 2,MPCvalues for these isotopes and on a X/Q of 5.6 x 10 sec/m at the site boundry.

3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the.

reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.

ForthepurposeofmeasuringF(Z),Fh,andF a full incore flux map q

is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be

~

used in recalibration of the excore neutron flur. detection system.

Full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.

1 FARLEY-UNIT 2 B 3/4 3-2 A&uMOuulT Mo.

i,

l 4

f ENCLOSURE 5 i

I l

Proposed FSAR Mark-ups Associated with Relocation of Reactor Trip and Engineered Safety Feature Actuation System Response Time Tables f

f l

t I

t D

a

?

.. i 6

-E r.f e

I

FNP-FSAR-7

- PAGE 1 0F 11 -

~

)'

7.2.3.1 Inservice Tests and Inspections Periodic surveillance of the reactor trip system is performed to ensure proper protective action.

This surveillance consists of checks, calibrations,*wv4 channel funct.ional, testing which are summarized as follows:

M ET%

A.

Checks A check consiets of a qualitative determination of acceptability by observation of channel behavior I

during operation.

It includes comparison of the channel with other independent channels measuring the same variable.

Failures such as blown instrument fuses, defective indicators, or faulted amplifiers are noticeable by simple observation of the functioning of Furthermore, in many cases the instrument or system..

such failures are revealed by alarm or annunciator action, and a check supplements this type of surveillance.

B.

Calibration A channel calibration consists of adjustment of

~

channel output such that it responds, within acceptable range and accuracy, to known values of the parameter which the channel measures.

Calibration encompasses the entire channel including alarm and/or trip; it also includes the channel functional testing discussed below.

ihus, the calibration ensures the acquisition and presentation of accurate information.

C.

Channel Functional Testing A channel functional test consists of injecting a simulated signal into the signal conditioning portion of the channel to verify its operability, including alarm and/or trip initiating action.

ME D. llkl%rtT il

)

2ne m2nimum frequency for checks, calibration,Vand Met >Ge-TMe testing are defined in the plant technical specifications.

7.2.3.2 Feriodic Testing' of the Nuclear Instrumentation

)

System The following periodic tests of the nuclear instrumentation system are performed:

)

7.2-40

INSERT 1 1

- PAGE 2 0F 11 -

Response time testing consists of any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined in the Technical Specifications. Sensor response time verification may be demonstrated by either 1) in place, onsite or.offsite test j

measurements or 2) utilizing replacement sensors with certified response times.

i The measurement of response time at the specified frequencies provides assurance that the reactor trip associated with each channel is completed within the time limit assumed in the accident analyses. The response time limits for the reactor trip system are provided in Table 7.2-5.

j l

i i

f

TABLE.b4_-4 7. 2-6 [.5Nd47-g oc 2)

.s REACTOR TRIF SYSTEM INSTRUMENTATION RESPONSE TIMES t

"i RESPONSE TIME FUNCTIONAL UNIT E

Not Applicable 1.

Nanual Reactor Trip 2.

Fover Range, Neutron Flux 5 0.5 seconds

  • a.

High Not Applicable b.

Lov 3.

Fover Range, Neutron Flux, High Fositive Rate Not Applicable 4.

Fover Range, Neutron Flux, Not Appilcable l)

High Negative Rate 5.

Inter.ediate Range, Neutron Flux Not Appilcable

{

y Not Applicable 6.

Source Range, Neutron Flux y

N

~1 f 6.0 seconds

  • l 7.

Overtemperature at Not Applicable 8.

Overpower ST f 2.0 seconds 9.

Pressurizer Pressure--Lov 10.

Pressurizer Pressure--High f 2.0 seconds

11. 'Fressurizer Vater Level--Bigh Not Applicable

}In[I w

i e Neutron detectors are exempt from response time testing. Response time of the neutron fluz signal ggrtion 1

[i zk of the channel shall be measured from detector output or input of first electronic component in channel.

i

=

e j

i

n!'

l

\\

~

l Y

G i

l

e TABLE 2,2-2 'C:M=Q T.2-S (Zh667 20C 2)

>>E REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES l Q l

I RESPONSE TIME FUNCTIONAL UNIT q

12.

A.

Loss of Flow - Single Loop i 1.0 seconds

[V (Above P-8)

B.

Loss of Flow - Two Loops 1 1.0 seconds (Above P-7 and below P-8) 1 2.0 seconds

13. Steam Generator Water Level--Low-Low
14. Stene/Feedwater Flow Missatch and Not Applicable Low Steam Generator Water Level i 1.2 seconds
15. Undervoltage-Reactor Coolant Pumps 1 0.6 seconds

'1 Underfrequency-Reactor Coolant Pumps 16.

17. Turbine Trip Hot Applicable g

b) e A.

Low Auto stop 011 Pressure Not Applicable N

B.

Turbine Throttle Valve Closure Hot Appitcable

18. Safety injection Input from ESF Not Applicable Reactor Coolant Pump Breaker Position Trip 19.

Not Applicable

20. Reactor Trlp System Interlocks g

Not Applicable E

21. Reactor Trip Breakers Y

Mot Appilcable g

4)}

22. Automatic Trip Logic a

g

'51

}

S 6

.a-

-c.

u--.-

FNP-FSAR-7

- PAGE 5 0F 11 -

plant condition changes, the test frequency is accelerated to accommodate the situation until the marginal performance is resolved.

D.

The test interval discussed in' paragraph 5.2 of IEEE 338-1971 is developed primarily on past operating experience and modified, if necessary, to ensure that system and subsystem protection is reliably provided.

Analytical methods for determining reliability are not

-f used to determine test interval.

7.3.2.6 Evaluation of Compliance with IEEE 344-1971 Theseismictesting,assetforthinparagraph7.2.1.10, IEEE 33 8-1971,m IEEE 344-1971,0* WCAP-7706,* and WCAP-7705,* conforms to the guidelines set forth in IEEE 344-1971.00

7. 3. 2. 7 ResPcuJs& Time Temdb

/didBy 2 7.3.2.J"8 Further Considerations In addition to the considerations given above, a loss of instrument air or loss of component cooling water to vital equipment has been considered.

Assuming no other accident conditions, neither cause safety limits, as given in the technical specifications, to be exceeded.

Likewise, loss of l

either one of the two will not adversely affect the core or the reactor coolant system, nor will it prevent an orderly shutdown if this is necessary.

Furthermore, all pneumatically operatsd valves and controls will assume a preferred operating position upon loss of instrument air.

It is also noted that, for conservatism during the accident analyses (chapter 15), credit is not taken for the instrument air systems nor any control system benefit.

7.3.2.K9 Summary The effectiveness of the ESFAS is evaluated in chapter 15, based upon the ability of the system to contain the effects of Condition III and IV faults, including LOCAs and steam break accidents.

The ESFAS parameters are based upon the component

(

performance specifications which are given by the manufacturer or verified by test for each component.

Appropriate factors to account for uncertainties in the data are factored into the constants characterizing the system.

The ESFAS must detect Condition III and IV faults and generate signals which actuate the engineered safety features.

The t

system must sense the accident condition and generate the 7.3-19 REV 10 6/92

- PAGE 6 0F 11 -

INSERT 2 Response time testing consists of any series of sequential. overlapping or total' channel test measurements provided that s"-h tests demonstrate the total channel response time as defined in the Technica' Specifications.

Sensor response time verification may be demonstrated by eitt >r 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.

The measurement of response time at the specified frequencies provides assurance that the reactor trip associated with each channel is completed within the time-limit assumed in the accident analyses. The response time limits for the Engineered Safety Feature Actuation System are provided in Table 7.3-16.

1 9

l

\\

~

  • e W

dr T7 t' W ? 41'W92

rap -fsM-7 TABLE-s-a4 7.3-/6 (:%+ris4)

- PAGE 7 0F 11 -

ENGINEERED SAFETY FEATURES RESPONSE TIMES

~

RESPONSE TIME IN SECONDS INITIATING SIGNAL AND FUNCTION 1.

Manual _.

Not Applicable Safety Injection (ECCS) a.

Not Applicable Feedwater Isolation Not Applicable Reactor Trip (SI)

Not Applicable Containment Isolation-Phase "A" Not Applicable Containment Purge Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Service Water System Not Applicable Containment Air Coolers Not Appitcable b.

Containment Spray Not Applicable Containment Isolatior-Phase "B" i

Not Applicable Containment Purge Isolation Not Applicable Centainment Isolation-Phase "A" c.

Not Applicabla Containment Purge Isolation Not Applicable d.

Steam Line Isolation 2.

Containment Pressure-Hich i 27.0(I)

Safety Injection (ECCS) a.

b.

Reactor Trip (from SI) i 2.0 II)

< 32.0 Feedwater Isolation I4)/27.0(5) j c.

Containment Isolation-Phase "A" 117.0 d.

Containment Purge Isolation 5 S.0 e.

Not Applicable f.

Auxiliary Feedwater Pumps I4)/87.0(5)

$ 77.0 Service Water System

)

g.

h.

Containment Air Cooler Fan 5 27.4 s

8 e

'ENDMENT NO. 26 3/4 3-2 FAR_ LEY Q j

=

.r.

fifP Fst.4-7 TABLE 2 ? 5 S-5=O 7.3-l(o (5%T4RELO 0F 11 -

ENGINEERED SATETT FEATURES RESPDNSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME TN SEC0 HTS 3.

Pressurizer Pressure-Lov a.

Safety Injection (ECCS) 3 27.0'*'/12.0

b.

Reactor Trip (from SI) f 2.0 c.

Feedvater Isolation f 32.0

d.

Containment Isolation-Phase "A"

$ 17.0

e.

Containment Purge Isolation f 5.0 f.

Auxiliary Feedvater Pumps Not Applicable g.

Service Vater System 3 77.0/87.0

4.

Differential Pressure Between Steam Lines-Eigh a.

Safety Injection (ECCS)

$ 12.0/22.0

b.

Reactor Trip (from SI) 3 2.0 c.

Feedvater Isolation f 32.0

d.

Containment Isolation-Phase 'A' f 17.0/27.0

e.

Containment Purge Isolation Not Applicable f.

Auxiliary Feedvater Pumps Not Applicable g.

Service Vater System 377.0/87.0

5.

Steam Flov in Two Steam Lines-Bigh Coincident with T,

--Low-Lov a.

Steam Line Isolation Not Applicable l

6.

Steam Line Pressure-Lov a.

Safety Injection (ECCS) f 12.0/22.0

b.

Reactor Trip (from SI) 3 2.0 c.

Feedvater Isolation f 32.0

d.

Containment Isolation-Phase "A" f 17.0/27.0

e.

Containment Purge Isolation Not Applicable f.

Auxiliary Feedvater Pumps Not Applicable

~

g.

Service Water System 3 77.0/87.0

h.

Steam Line Isolation f 7.0

~'

AMENDKENT NO.

,H,

)

TARLET - UNIT

i FdP FS M.- 7 i

TABLE :.24 'cutend) 7.3-b M'[ p d(

ENGINEERED SAFETY FEATURES RESPONSE TIMES RESPONSE TIME IN SECONOS i

INITIATING SIGNAL AND FUNCTION 7.

Containment Pressure--Mich-Hich s.

Steam Line Isolation 5 7.0 8.

Containment Pressure--Hich-Hich-Hich 5 45.0 a.

Containment Spray b.

. Containment Isolation-Phase "B" Not Applicable Steam Generator Vater Level--High-Hich 9.

i 2.5 a.

Turbine Trip IO) b.

Feedwater Isolation i 32.D

10. Steam Generator Vater Level -- Low-Low Motor-driven Aux. Feedwater Pumps (2) 5 60.0 a.

I3) b, Turbine-driven Aux. Feedwater Pump 1 60.0 i

11, Undervoltace'RCP

< 60.0 Turbine-driven Aux. Feedwater Pump a.

12.

S.I. Sicnal Motor-driven Auxiliary i 60.0 a.

Feedwater Pumps i

i

13. Trip of Main Feed ater Pumps Not Applicable Motor-driven Aux. Feedwater Pumps

)

a.

14.

Loss of Power f

4.16 kv Emergency Bus Undervoltage (7) j a.

(Loss of Voltage) 4.16 kv Emergency Bus Undervoltage (7) b.

(Degraded Voltage)

~

e i

$==-

Y 3/4 3-31 Q EY-UNIT 1 e

4

v

,a fs/P-BAR-7 TAstt v c " "-n

7. 3-w (.sw./ oc 4')

- PAGE 10 0F 11 -

TABLE NOTATION

/

(1) Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of dischargy pressure for centrifugal charging pumps, and RHR pumps.

(2) One 2/3 any' Steam Generator (3) On 2/3 in 2/3 Steam Generators (4) Diesel generator starting and sequence loading delay not included.

Offsite power available. Respone time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.

(5) Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.

(6) Verification shall include testing of all instrumentation, the isolation valves (MOV-3232A, 3232B, 3232C) and the control valves (FCV-478, 479, 488, 489, 498, 499). The isolation valves must function within 30 seconds and the control valves within 5 seconds.

(7) The respone time shall include the time delay associated with the undervoltage relays as determined in Table 3.3-plus an additional second associated with interposing relay and ci uit operation.

OC TM! HCHA/IC6L SM4tfICADOUT l-E e

F e

5 O

l

.o

{

FNP-FSAR-15

- PAGE 11 0F 11 -

A reactor trip signal acts to open two trip breakers connected in series which feed power to the control rod drive mechanisms

[

(CRDMs).

The loss of power to the mechanism coils causes the i

(

mechanisms to release the rod cluster control assemblies (RCCAs) which then fall by gravity into the core.

There are f

various instrumentation delays associated with each trip function, including delays in signal actuation, in opening the i

trip breakers, and in the release of the rods by the i

mechanisms.

The total delay to trip is defined as the time delay [

j f

(

from when the monitored parameter reaches the trip setpoint until l the rods are free and begin to fall.

The setpoint study is performed in the course of finalizing the design of the plant; however, many of the accident analyses in this chapter I

conservatively do not take credit for the control systems.

Table 15.1-3 refers to the overtemperature and overpower AT trip shown in figure 15.1-1A.

+

These trip setpoints bound the transition cores and a full core of VANTAGE 5 fuel.

The associated OTAT f(AI) penalty is shown in figure 15.1-1B.

For all the reactor trips, the difference between the trip setpoints assumed in the analysis and the nominal trip setpoints account for instrumentation channel error and setpoint error.

The plant technical specifications specify the nominal trip setpoints.

The calibration of protection system channels and the periodic d ermination of instrument response times are in accordance with the lant technical s eci{ications.

h3A

& tim 6 LIMITT FOf MAC722 TEtP387UU M MA/A03/iku /Al 738/k 72-S.

15.1.4 INSTRUMENTATION DRIFT AND CALORIMETRIC ERRORS l

The VANTAGE 5 fuel design features, the modified safety analysis assu=ptions, and the application of new methodologies (i.e.,

RTDP, WRB-1, and WRB-2) as discussed in section 4.4 (with respect

/

to the changes associated with the instrument uncertainties for I

the NSSS control parameters of power, pressure, temperature, and flow) are covered in reference 2.

i 15.1.5 ROD CLUSTER CONTROL ASSEMBLY INSERTION CHARACTERISTIC l

~

(

The negative reactivity insertion following a reactor trip is a function of the acceleration of the RCCAs and the variation in rod worth as a function of rod position.

With respect to accident analyses, the critical parameter is the time from the l

start of insertion up to the dashpot entry or approximately 85

[

percent of the rod cluster travel.

For accident analyses, it is

(

conservatively assumed that the insertion time to dashpot entry is 1

15.1-9 RIV 11 6/93

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