Text

Proposed Tech Specs Re Incorporation of Results of Work Performed by BWROG
	ML20087H891
Person / Time
Site:	Perry
Issue date:	05/01/1995
From:	; CENTERIOR ENERGY
To:	;
Shared Package
ML20087H888	List: ML20087H891; PY-CEI-NRR-1932, Application for Amend to License NPF-58.Amend Would Incorporate Results of Work Performed by BWROG Re Elimination of Selected Response Time Testing Requirements;
References
	NUDOCS 9505050104
	Download: ML20087H891 (40)
	v • d • e

.

Attachrant 2 PY-CEI/NRR-1932 L DEFINITIONS

~

DRYWELL INTEGRITY (continued) f.

The suppression pool is in compliance with the requirements of Specification 3.6.3.1.

g.

The sealing mechanism associated with each drywell penetration; e.g.,

welds, bellows or 0-rings, is OP,ERABLE.

w.

E-AVERAGE DISINTEGRATION ENERGY

'1.12 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTBi (ECCS) RESPONSE TIME 1.13 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their requireid positions, pump dis-charge pressures reach their required values, etc. Timo shall include diesel

~

generator starting and sequence loading delays where applicable.

The response t., '

time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured. 4-END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESP 3MSE TIME 1.14 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:

a.

Turbine stop valves, and b.

Turbine control valves.

3 The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

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sM Ge t ma

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PERRY - UNIT 1 1-3 Amendment No. 42 9505050104 950501 ADOCK05000J4O PDR P

[I Attachatnt 2 PY-CEI/NRR-1932 L Page 2 of 16 DEFINITIONS FREQUENCY NOTATION 1.17 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

FUEL HANDLING BUILDING INTEGRITY 1.18 FUEL HANDLING BUILDING (FHB)'INTIG'RITY shall exist when:

a.

The doors in each access to the 620 foot elevation of the FHB are closed, except for normal entry and exit.

b.

The FHB railroad track door is closed.

c.

The fuel handling area floor hatches are in place.

d.

The FHB ventilation system is in compliance with Specification.

3.7.7.1.

The shield blocks are installed adjacent to the Shield Building.

e.

GASE0US RADWASTE TREATMENT (0FFGAS) SYSTEM 1.19 The GASE0US RADWASTE TREATMENT (0FFGAS) SYSTEM is the system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgasses from the main condenser evacuation system and provid-ing for delay or holdup for the purpose of reducing the tctal radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.20 IDENTIFIED LEAKAGE shall be:

a.

Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or b.

Leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.

ISOLATION SYSTEM RESPONSE TIME 1.21 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions.

Times shall include diesel generator starting and sequence loading delays where applicable.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is cd s\\.kJ a % inMv,1) g, yt,lL o r e.,

Ogostutd 3 PERRY - UNIT 1 1-4 Amendment No. 6

~

PY-CEI/NRR-1932 L Page 3 of 16 DEFINITIONS f.

The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows or 0-rings, is OPERABLE.

PROCESS CONTROL PROGRAM (PCP) 1.34 The. PROCESS CONTR.0L PROGRAM shall.contajn the current'.. formulas; sampling, '..

analyses ' tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 61, 10 CFR Part 71 and Federal' and State regulations, burial ground requirements and other requirements governing the disposal of the radioactive waste.

PURGE - PURGING 1.35 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER i

1.36 RATED THERMAL POWER shall be a total reactor core heat transfer rate to i

the reactor coolant of 3579 MW.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.37 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measuredy g"LLt%.

s are s%ke..(

sn REPORTABLE EVENT

^ ~M *

0%v rwd 3 1.38 A REPORTABLE EVENT shall be any of those conditions specified in i

10 CFR 50.73.

l R00 DENSITY I

1.39 R00 DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted i

is equivalent to 100% ROD DENSITY.

SECONDARY CONTAINMENT INTEGRITY 1.40 SECONDARY CONTAINMENT INTEGRITY shall exist when:

e a.

All penetrations terminating in the annulus and required to be closed during accident conditions are either:

PERRY - UNIT 1 1-7 Amendment No.pp

I 3/4.3 INSTRUMENTATION Attschm:nt 2 PY-CEI/NRR-1932 L 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION Page 4 of 16 LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor arotection system innt m anuauion channels s iown in h >1e 3.3.1-1 shall be OP ERABlyith th: '"~r"^A FaGEi tun m ap E3?=

r :: :t: = ::: = :: 3. J. Q APPLICABILITY:

As shown in Table 3.3.1-1.

ACTION:

a.

With one channel required by Table 3.3.1-1 inoperable in one or more Functional Units place the inoperable channel and/or that trip system in thetrippedcondition*within12 hours, b.

With two or more channels required by Iable 3.3.1-1 inoperable in one or more Functional Units; 1.

Within one hour, verify sufficient channels remain OPERABLE or are in the tripped condition

to maintain trip capability in the Functional Unit, and 2.

Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months place the inoperable channel s in one trip system and/orthattripsystem**inthetrippedconiion,*and 3.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months restore the inoperable channels in the other trip i

systemtoanOPERABLEstatusorplacetheminthetrippedcondition*.

Otherwise, take the ACTION required by Table 3.3.1-1 for the Functional Unit.

SURVEILLANCE RE0VIREMENTS 4.3.1.1 Each reactor rotection system instrumentation channel shall be i

demonstrated OPERABLE the performance of the CHANNEL CHECK CHANNEL FUNCTIONAL TEST and C NELCALIBRATIONoperationsfortheOP$ RATIONAL i

CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

i 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

i 4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unitc.c. T. :

m: :.:.'- Dsha11 be demonstrated to be within its limit at least once per 18 mont per trip system such that al f1 c'I n Each test shall include at least one channeliannels a months where the total (number of redundant channels in a specific reactor Lt t s. u-a ha w.;W.d at od o r s o r e. cept % re s p nse. hs it. Nag. Fge

}

. system A

'a n 4s b.W 5.4. And sensorJ on cadwAc4 Ow ru pnse.

g.3.1.9 ine provisions of spectrication 4.v.4 are not applicapie to the % 6 CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION surveillances for the Intermediate Range Monitors for entry into their applicable OPERATIONAL bm%,sD'__ /

CONDITIONS (as shown in Table 4.3.1.1-11 from OPERATIONAL CONDITION 1, provided the survelliances are performed within h2 hours after such entry.

An inoperable channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur.

In these cases, if the inoperable channel is not restored to OPERABLE status within the required time, the ACTION required by Table 3.3.1-1 for the Functional Unit shall be taken.

- This ACTION applies to that trip system with the most inoperable channels; if both trip systems have the same number of inoperable channels, the ACTION can be applied to either trip system.

PERRY - UNIT 1 3/4 3-1 Amendment No. -41,46, 67 mwz--r-.

wa-w-

+ - -

w

. m eim

. +

---em+i u.----+-

T' 9

d TABLE 3.3.1-2 h

REACTORPROTECTIONSYSTEMRESPONSETIMEh

\\

O c

\\

.A '

5

]

FUNCTIONAL VNIT ESPONSE TIME V

(Seconds) 1.

Interme late Range Monitors:

a.

Ne ron Flux - High NA

'f b.

Ino rative NA p

2.

Average Po er Range Monitor.:

c a.

Neutro Flux - High, Setdown NA e

b.

Flow Bi sed Simulated Thermal Power - High s 0.09,,

7, c.

Neutron lux - High 5 0.09 e

d.

Inoperati e NA 7

3.

Reactor Vessel team Dome Pressure - High s 0.35 6'

4.

Reactor Vessel W ter Level - Low, Level 3 w

s 1.05 l

2 5.

Reactor Vessel Wa er Level - High, Level 8 s 1.05 6.

Main Steam Line Is lation Valve - Closure wa 7.

Deleted s 0.06 g

l 8.

Drywell Pressure -

gh NA 9.

Scram Discharge Volu Water Level - High NA

(,

fj

10. Turbine Stop Valve -

osure s 0.06 11.

Turbine Control Valve st Closure, Valve Trip System y

i Oil Pressure - Low

\\

r s 0.07#

12. Reactor Mode Switch Shut wn Position NA k

j 13.

Manual Scram NA E

f i

2o

[

~ Neutron' detectors are exempt from response time testing Response t me shall be measured from '

? 5 %

P the detector output or from the i ut of the first electronic compon t in the channel.

%Ag

..Not including the simulated thrmal ower time constant specified in t COLR.

uM R-

{hE

Measured from start of turbine contro valve fast closure.

g,L y O

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INSTRUMENTATION Attach: ant 2 PY-CEI/NRR-1932 L 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION Page 6 of 16 LIMITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trio Setooint column of Table 3.3.2-g.: =. 50a!""' EYS?5". ".55==P CIS n ;tn.. 6 T:th 3.3.tp APPLICABillfY:

As shown in Table 3.3.2-1.

ACTION:

a.

With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent i

with the Trip Setpoint value.

b.

With one channel required by Table 3.3.2-1 inoperable in one or more Trip Functions, place the inoperable channel and/or that trip system in the tripped condition

within:

1.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for Trip Functions common to RPS instrumentation,.and 2.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Trip Functions not common to RPS instrumentation.

c.

With two or more channels required by Table 3.3.2-1 inoperable in one or more Trip Functions; 1.

Within one hour, verify for automatic Trip Functions that sufficient channels remain OPERABLE or are in the tripped condition

to maintain isolation capability for the Trip Function, and l

2.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for Trip Functions common to RPS instrumentation, and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Trip Functions not common to RPS instrumentation, place the inoperable channel (s) in the tripped condition *.

Otherwise, take the ACTION required by Table 3.3.2-1 for the Trip Function.

An inoperable channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur.

In these cases, j

if the inoperable channel is not restored to OPERABLE status within the required time, the ACTION required by Table 3.3.2-1 for the Trip Function shall be taken.

I PERRY - UNIT 1 3/4 3-9 Amendment No. 4HL, 67

Attachm nt 2 PY-CEI/NRR-1932 L i

Page 7 of 16 INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.

4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function (shown in T&t,le 3.3." ' shall be demonstrated to be within its limit at least once per 18 months.,,Each test shall include at least one channel per trip system 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 isolation trip system. _

(

n.

bt.h b Q f S ch ((

R.,% ( V

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(gm A&h tu p ue M e teshn3, l

PERRY - UNIT 1 3/4 3-10

Attechment 2 i

, h E 3.3.2-3 p

[bOLATIONSYhTEMINSTRUMENTATIONRSPONSET I

RESPONSE TIME (Seconds)#

1.

RIMARY CONTATNMENT ISOLATI a.

Reactor Vessel Water Leve - Low, Level 2 N

b.

Drywell Pressure - High NA ContainmentandDrpellPur Exhaust Plenum c.

adiation - High*

s 10(

d.

actor Vessel Water Level -

w, Level 1 NA e.

Ma ual Initiation NA 2.

MAIN STE LINE ISOLATION a.

Reacto Vessel Water Level - Low, L el I s 1.0*/s 10(

b.

Main St am Line Radiation - High NA

['

c.

Main Ste Line Pressure - Low

$ 1.0*/s 10(*)*

d.

Main Stea Line Flow - High s 0.5 /s }0(*)"

e.

Condenser acuum - Low NA s

f.

Main Steam ine Tunnel Temperature - High NA g.

Main Steam L e Tunnel A Temperature - High NA h.

Turbine Build g Main Steam Line Temperature -

gh NA 1.

Manual Initiatto NA 3.

SECONDARY CONTAINMENT OLATION a.

Reactor Vessel Water Level - Low, Level 2 NA b.

Drywell Pressure - Hi h NA c.

Manual Initiation NA

4. REACTOR WATER CLEANUP SYSTEM ISO TION a.

A Flow - High NA b.

A Flow Timer NA c.

Equipment Area Temperature - igh NA f

d.

Equipment Area A Temperature - High NA e.

Reactor Vessel Water Level - Lo Level 2 NA f.

Main Steam Line Tunnel Ambient Temperature - High NA g.

Main Steam Line Tunnel A Temperatu

- High NA h.

SLCS Initiation NA 1.

Manual Initiation NA pap dh nj 3 ( (-).

o i

PERRY - UNIT 1 3/4 3-21 Amendment NoS8 i

Attechment 2 TABLE 3.3.2-3 ((onti P -CEI/

932 L

]ISOLATIONSYSTEMINSTRUMENTATIONRESPONSETIM

[TRIPFUNCTION RESPONSE TIME (Secon M 5.

REACTOR CORE ISOLATION 00 LING SYSTEM ISOLA ON RCIC Steam Line Flo - High NA RCIC Steam Supply Pre sure - Low NA c.

RCIC Turbine Exhaust D'aphragm Pressure - Hi h NA d.

RCIC Equipment Room Amb ent Temperature - Hig NA

~

l e.

Deleted f.

ain Steam Line Tunnel Am ient Temperature - High A

g.

M in Steam Line Tunnel A Te erature - High N

h.

Ma Steam Line Tunnel Tempe ture Timer NA i.

RHR quipment Room Ambient Te erature - High NA j.

RHR uipment Room A Temperatu

- High NA k.

RCIC eam Line Flow High Timer NA 1.

Drywel Pressure - High NA m.

Manual itiation NA 6.

RHR SYSTEM ISOL TION a.

RHR Equipmen Area Ambient Temperature - High NA

\\

b.

RHR Equipment Area A Temperature - High NA c.

RHR/RCIC Steam Line flow - High NA d.

Reactor Vessel ater Level - Low, Level 3 NA e.

Reactor Vessel (

R Cut-in Permissive)

Pressure - High NA f.

Drywell Pressure - High NA g.

Manual Initiation NA (a)

Isolation system instrumenta ion response time specifie includes the diesel generator starting and sequence loading delays.

(b) Radiation detectors are exempt om response time testing.

esponse time shall be measured from detector tput or the input of the fi st electronic component in the chann

Isolation system instrumentation res onse time for MSIVs only.

diesel

{

generator delays assumed.

Isolation system instrumentation respon e time for associated valves except MSIVs.

Isolation system instrumentation response me specified for the Trip Function actuating each containment isolatio valve shall be added to th isolation time for each valve to obtain ISOLA ION SYSTEM RESPONSE TIME for each valve.

k* 7 PERRY - UNIT 1 3/4 3-22 Amendment No. 44,59

Attechment 2 PY-CEI/NRR-1932 L-Page 10 of 16 INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3g lg GTd with DI-RGENCi CGRE CGGLING SYSTEP RESPONSE TITC o>Wud in :m 11_ y APPLICABILITY:

As shown in Table 3.3.3-1.

ACTION:

With an ECCS actuation instrumentation channel trip setpoint less a.

conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value, b.

With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1.

With either ADS trip system "A" or "B" inoperable, restore the inoperable c.

trip system to OPERABLE status:

1.

Within 7 days, provided that the HPCS and RCIC systems are OPERABLE, or, 2.

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , provided either the HPCS or the RCIC system is inoperable.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to less than or equal to 100 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and l

CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the l

f requencies shown in Table 4.3.3.1-1.

4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

1 I

[4.3.3.

The ECCS RESPONSE ME of each ECCS trip f' ction shown in ble 3.3.3-3 shall b demonstrated to be w~ thin the limit at least nce per 18 mont s.

Each test shal include at least one channel per trip syste such that all c nnels are tested least once every N imes 18 months where N 's the total num er of Q dundant cha nels in a specific E S trip system (

PERRY - UNIT 1 3/4 3-27

--n-..

Attcch2:nt 2 PY-CEI/NRR-1932 L Page 11 of 16 TABLE 3.3.3-3 (EMERGENCYCORECOOLINGSYSTEMRESPONSETIMES RESPONSETIME(SecondsP j

ECOS

\\

A.

DJVISION 1 TRIP SYSTEM l

1.

R-A (LPCI MODE) AND LPCS SYSTEM a.

Reactor Vessel Water Level - Low, 1 37 evel 1 b.

D ell Pressure - High

< 37 c.

LP Pump Discharge Flow - Low (Bypass)

NA d.

Rea or Vessel Pressure - Low (LPCS Injectio NA Valv Permissive) e.

React Vessel Pressure - Low (LPCI Injection NA Valve reissive) f.

LPCI P A Start Time Delay Relay g,

LPCI P A Discharge Flow - Low (Bypass)

NA h.

Manual Ini lation NA 2.

AUTOMATIC DEPRE URIZATION SYSTEM TRIP SYSTEP. "A" i

a.

Reactor Vessel ater Level - Low, Level 1 NA b.

Manual Inhibit NA c.

ADS Timer NA d.

Reactor Vessel Wa er Level - Low, NA

~~

Level 3 (Permissiv )

\\

LPCS Pump Discharge Pressure - High NA e.

(Permissive) f.

LPCI Pump A Discharge ressure - High NA (Permissive) g.

Manual Initiation NA B.

DIVISION 2 TRIP SYSTEM 1.

RHR B AND C (LPCI MODE) a.

Reactor Vessel Water Level - L

< 37

~

Level 1 b.

Drywell Pressure - High 1 37 c.

Reactor Vessel Pressure - Low (L I NA Injection Valve Permissive) d.

LPCI Pump B Start Time Delay Relay NA e.

LPCI Pump Discharge Flow - Low (Byp s)

NA f.

Manual Initiation NA j

~

b I %Oa g

PERRY - UNIT 1 3/4 3-35

Attechment 2 PY-CEI/NRR-1932 L Page 12 of 16

]5MERGENCYCORECOOLINGSYSTEMRESPONSETIMES TRIP FU$CTION RESPONSE TIME (Seconds) 2.

AUT TIC DEPRESSURIZATION SYSTEM TRIP STEM "B" a.

R actor Vessel Water Level - Low, MA Le el 1 b.

Ma al Inhibit NA r

c.

ADS iner NA d.

Reac r Vessel Water Level - Low, NA Level (Permissive) e.

LPCI P B and C Discharge NA Pressur - High (Permissive) f.

Manual I itiation NA fC.

DIVISION 3 TRIP YSTEM

1. "HPCS SYSTEM a.

Reactor Vess 1 Water Level - Low,

<2 Level 2 b.

Drywell Press e - High

< 27 c.

Reactor Vessel ater Level - High, HA Level 8 d.

Condensate Stora Tank Level - Low NA e.

Suppression Pool ter Level - High NA f.

HPCS Pump Discharg Pressure - High NA g.

HPCS System Flow Ra

- Low NA h.

Manual Initiation NA D.

LOSS OF POWER 1.

4.16 kv Emergency Bus Underv 1tage#

NA (Loss of Voltage) 2.

4.16 kv Emergency Bus Undervolt ge#

NA (Degraded Voltage) l

~

N T\\G oy J~> A\\q u a,

i f Voltage and Degrad\\eVoltage functions are n to Division 1,

The Loss Division and Division 3.

\\

~

PERRY - UNIT 1 3/4 3-36

~

.-m

PY-CEI/NRR-1932 L Page 13 of 16 INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pump trip (EOC-RPT) system instrumentation channels shown in Table 3.3.4.2-1 shall be OPERABLE with their l

trip setpoints set consistent with the values shown in the Trin setnoint column of Table 3.3.4.2-1.gnd-with tha EurtnF-CYCLE RECIRCtitATf0NDP TRifb y

415]EM RESPONSE-TIMt n & in Table 3.3.A.2-3j j

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 40% of RATED THERMAL POWER.

ACTION:

a.

With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.

I b.

With the number of OPERABLE channels less than required by the i

Minimum OPERABLE Channels per Trip System requirement:

1.

Verify that a sufficient number of channels remain OPERABLE or are in the tripped condition to maintain EOC-RPT trip capability for both the turbine stop valve closure and turbine control valve fast closure Trip Functions within two hours, and 4

2.

Place the inoperable channel (s) in the tripped condition within 1

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Otherwise, either remove the associated recirculation pump fast speed breaker from service or reduce THERMAL POWER to less than 40%

of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

t i

PERRY - UNIT 1 3/4 3-44 Amendment No. 67

Attacharnt 2 PY-CEI/NRR-1932 L Page 14 of 16 INSTRUMENTATION i

SURVEILLANCE REQUIREMENTS 4.3.4.2.1 Each end-of-cycle recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.4.2.1-1.

4.3.4.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

4.3.4f2.3 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME of each trip function (iba ; i., Tab'e 3.3.0.2-3Dsha11 be demonstrated to be within its limit at least once per 18 months. Each test shall include at least the logic of one type of channel input, turbine control valve fast closure or turbine stop valve closure, such that both types of channel inputs are' tested at least once per 36 months. The measured time shall be added to the most recent breaker arc suppression time and the resulting END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be verified to be within its limits.

4.3.4.2.4 The time interval necessary for breaker arc suppression from energi-zation of the recirculation pump circuit breaker trip coil shall be measured at least once per 60 months.

PERRY - UNIT 1 3/4 3-45

t e

ABLE 3.3.4.2-3 1

-OF-CYCLE RECIRCULAMON PUMP TRIP SYSTEM RESP 0ihiE TIME g.

TRIP FUNCTION RESPONSENIME (Milliseconds)

E q

1.

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Attachant 2 1

PY-CEI/NRR-1932 L Page 16 of 16 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.1 ECCS division 1, 2 and 3 shall be demonstrated OPERABLE by:

a.

At least once per 31 days for the LPCS, LPCI and HPCS systems:

1.

Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.

2.

Verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct

position.

b.

Verifying that, when tested pursuant to Specification 4.0.5, each:

1.

LPCS pump develops a flow of at least 6110 gpm at a differential pressure greater than or equal to 128 psid for the system.

2.

LPCI pump develops a flow of at least 7100 gpa at a differential pressure greater than or equal to 24 psid for the system.

3.

HPCS pump develops a flow of at least 6110 gpm at a differential pressure greater than or equal to 200 psid for the system.

c.

For the LPCS, LPCI and HPCS systems, at least once per 18 months:

1.

Performing a system functional test which includes simulated l

automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injec-i tion of coolant into the reactor vessel may be excluded from this test.

1 2.

Performing a CHANNEL CALIBRATION of the ECCS discharge line T_

" keep filled" pressure alarm instrumentation.

/

d.

For the HPCS system, at least once per 18 months, verifying that the I

suction is automatically transferred from the r~ 1ensate storage tank to the suppression pool on a condensate storage tank low water level signal and on a suppression pool high water level signal.

ECCS RE 590W SE T\\rnE bU Re 3.

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Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation.

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PERRY - UNIT 1 3/4 5-4 s

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PY-CEI/NRR-1932 L Page 1 of 3 SIGNIFICANT HAZARDS CONSIDERATION The etandards used to arrive at a determination that a request for amendment involves no significant hazards considerations are included in the Coamission's Regulations, 10CFR50.92, which state that the operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any occident previously evaluated, or (3) involve a significant reduction in the strgin of safety.

The proposed change has been reviewed with respect to these three factors and it has been determined that the proposed change does not involve a significant hazard because:

1.

The changes da not involve a significant increase in the probability or consequences of an accident previously evaluated.

For those proposed changes dealing with the elimination of selected response time test requirements, the purpose of the proposed Technical Specification change is to eliminate response tt'.a testing requirements for selected components in the Reactor Protection System, Isolation System, and Emergency Core Cooling System. The BWR Owners' Group has completed an evaluation which demonstrates that the response time testing is redundant to other Technical Specification required testing. Inese other tests, in conjunction with actions taken in response to NRC Bulletin 90-01, " Loss of Fill-Oil in Transmitters Manufactured by Rosemount," and Supplement 1, are sufficient to identify failure modes or degradations in instrument response time and ensure operation of the assoClAted systems within acceptable limits. There are no known failure modes that can be detected by response time testing that cannot also be detected by the other required Technical Specification testing. This evaluation was documented in NED0-32291, "Gystem Analyses for Elimination of Selected Response Time Testing Requirements," January 1994, and the letter from T. Green to P. Loeser dated April 15, 1994 which were approved by an NRC Safety Evaluation dated December 28, 1994. The applicability of this evaluation to the Perry Nuclear Power Plant (PNFP) has been confirmed, In addition, PNPP will complete the additional actions identified in the NRC staff's Safety Evaluation of NED0-32291.

Because of the continued application of other existing Technical Specifiention required tests such as channel calibrations, channel checks, channel fvactional tests, and logic system functional tests, the response times of these systems will be maintainoi within the acceptance limits assumed in plant safety analysis and required for successful mitigation of an initiating event. The proposed Technical Specification changes do not affect the capability of the associated systems to perform their intended function within their rem 41 red response time, nor do the proposed changes themselves affect the operation of any equipment. As a result the proposed changes dealing with alimination of selected response time tests do not involve a significant increase in the probability or the consequences of an l

accident previously evaluated.

1 i

PY-CEI/NRR-1932 L Page 2 of 3 For those changes dealing with moving the surveillance requirement for ECCS RESPONSE TIME testing from the instrumentation section to the system section of the Technical Specifications, no change in testing requirements (other than the elimination of the instrument loops implemented as part of the NEDO-32291 changes) has been introduced. The relaxation in Applicability does not increase the probability or the consequences of an accident previously evaluated, since there are no design basis events during OPERATIONAL CONDITION 4 and 5 where ECCS systems art relied upon.

For those changes dealing with relocation of the response time limits from Technical Specification Tables and into the Updated Safety Analysis Report (USAR), the proposed changes are administrative in nature in that the test requirements and time limits are still requirements, but the placement of the limits have been relocated from the Technical Specifications and into the USAR. Therefore these changes do not involve a significant increase in the probability or the consequences of an accident previously evaluated.

2.

The changes do not treate the possibility of a new or different kind of accident from any previously evaluated.

None of the proposed Technical Specification changes affect the capability of the associated systems to perform their intended function within the acceptance limits assumed in plant safety analyser and required for successful mitigation of an initiating event. The proposed changes also do not change the manner in which any plant equipment is operated.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3.

The changes do not involve a significant reduction in the margin of safety.

The current Technical Specification response times are based on the maximum allovable value assumed in the plant safety analyses. These analyses conservatively establish the margin of safety. As described above, the proposed Technical SpecfU ation changes do not affect the capability of the associated systems to p; form their intended function within the alleved response time used as the basis for the plant safety analyses.

Plant and system response to an initiating event vill remain in compliance within the assumptions of the safety analyses, and therefore the margin of safety is not affected.

Although not explicitly evaluated, the proposed Technical Specification changes dealing with response time testing elimination vill provide an improvement to plant safety and operation by reducing the time safety i

systems are unavailable, reducing safety system actuations, reducing plant shutdown risk, limiting radiation exposure to plant personnel, and eliminating the diversion of key personnel te conduct unnecessary testing.

Therefore, the proposed changes do not result in a significant reduction in a margin of safety, and may result in an overall increase in the margin of safety.

l I

l PY-CEI/NRR-1932 L Page 3 of 3 l

1 The changes dealing with relocation of the time response limits from the Technical Specifications to the USAR is an administrative change that does not affect either the requirements to perform response time testing or the limits associated with the response time tests.

Future changes to the limits vill be controlled by 10CPR50.59. Therefore, this portion of the l

change does not result in a significant decrease in a margin of safety.

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Attechment 4 3/4.3 INSTRUMENTATION PY-CEI/NRR-1932 L Page 1 of 4 BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a.

Preserve the integrity of the fuel cladding.

b.

Preserve the integrity of the reactor coolant system.

c.

Minimize the energy which must be absorbed following a loss-of-coolant accident, and d.

Prevent inadvertent criticality.

This specification provides the limiting condition: for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maini.enance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems. The bases for the trip settings of the RPS are discussed in the bases for Specificatior. 2.2.1.

Specified surveillance intervals and surveillance and raintenance outage times have been determined in accordance with NEDC-30851P, "Tec5nical Specification Improvement Analysis for BWR Reactor Protection System," as approved by the NRC and documented in the NRC Safety Evaluation Report (SER) letter to T. A. Pickens from A. Thadani dated July 15, 1987.

Action b.1 is intended to ensure that appropriate actions are taken if a loss-of-function situation occurs during repairs of multiple inogerable, untripped instrument channels.

In regards to ACTION b.1 RPS"tripcapability is considered to be mair.tained when each " Functional Unit" identified in Table 3.3.3-1 has sufficient channels OPERABLE or in the tripped condition such that both trig systems will enerate a trip signal upon receipt of a valid signal from that Functional Unit" without the need to consider a further single failure event).

The Functional Units identified in Table 3.3.1-1 are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances entry into associated ACTIONS may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided the assoclated Functional Unit maintains trip capability.

Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable ACTIONS taken. This Note is based on the RPS reliability analysis assumption that 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months i

allowance does not significantly reduce th e probability that the RPS will trip when necessary.

The measurement of response time at the specified frequencies provides assurance that the protective functions associated wittteach channel are completed within the t'me limit assumed in the safety analyses. 6 mm;mter**thacecm= f 4mm isimia n.mt 2;;licBl@' Response tlme may be demonstrated by G

ts.m p =

1 any series of sequential, overlapling, or total channel test measurement, provided such tests demonstrate the total channel response time as defined. Sensor response i

l time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

i PERRY - UNIT 1 B 3/4 3-1 Amendment No. 67

Attachmsnt 4 INSTRUMENTATION PY-CEI/NRR-1932 L Page 2 of 4 BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints QM resper.52 -iawfor isolation of the reactor systems.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P, Supplement 2, " Technical Specification Imgrovement Analysis for BWR Instrumentation Common to RPS and ECCS Instrumentation, as approved by the NRC and documented in the NRC Safety Evaluation Report (SER) letter to D.N. Grace from C.E. Rossi dated January 6, 1989, and NEDC-31677P, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation" as approved by the NRC and documented in the NRC SER letter to S.D.

Floyd from C.E. Rossi dated June 18, 1990.

Action c.1 is intended to ensure that appropriate actions are taken if a loss-of-function situation occurs auring repairs of multiple, inoperable, untripped instrument channels.

In regards to ACTION c.1, " isolation capability" is considered to be maintained when sufficient channels are OPERABLE or in the tripped condition such that each " Trip Function" identified in Table 3.3.2-1 is capable of isolating the associated piping flow paths upon receipt of a valid signal from that " Trip Function" (without the need to consider a further single failure event). ACTION c.1 is not applicable to the Manual Initiation Trip Functions since they are not assumed in any accident or transient analysis. Thus, a total loss of manual initiation capability for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (as allowed by ACTION c.2) is permitted.

The Trip Functions identified in Table 3.3.2-1 are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated ACTIONS may be delayed as follows:for up ns other than 5.m provided the associated Trip Function maintains isolation capability.

Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable ACTIONS taken.

This Note is based on the reliability analysis assumpticn that 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance does not significantly reduce the probability that the isolation will occur when necessary.

Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety.

The setpoints of other instrumentation, where only the high or low end of +he setting have a direct bearing on safety, are established at a level away from the N. mal operating range to prevent inadvertent actuation of the systems involved.

Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connectu.

For D.C. opere.ted valves, a 3 second delay is assum 1 before the valve starts to move. For A.C. operated valves, it is assumed that the A.C. power supply is lost and is restored by startup of the emergency diesel generators.

In this event, a time of 13 seconds is assumed.efore the valve starts to move.

In addition to the pipe break, the failure of the D.C. operated valve is aumed; thus the signal delay (sensor response) is concurrent with the 13-second diesel :tartup. The safety analysis considers an allowable inventory loss in each case hich in turn determines the valve speed in conjunction with the 13-second delay.

It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isola ion functions.

PERRY - UNIT 1 B 3/4 3-2 Amendment No. 67

Attachmant 4 PY-CEI/NRR-1932 L Page 3 of 4 INSTRUMENTATION BASES 3.4.3.2 ISOLATION ACTUATION INSTRUMENTATION (Continued)

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation Instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements,+ trip setpointsCanTFw=: =9that will ensure effectiveness of the systems to provide the design rotection. Although the instruments are listed by system, in some cases the same nstrument may be used to send the actuation signal to more than one system at the same time.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30936P, Part 2, " Technical Specification Imgrovement Methodology (with Demonstration for BWR ECCS Actuation l

Instrumentation) as approved by the NRC and documented in the NRC Safety Evaluation Report (SER) letter to D.N. Grace from C.E. Rossi dated December 9, 1988 (Part 2).

ACTIONS 23, 31, 34 35 and 39 contain provisions to ensure that appropriate actionsaretakenifaloss-of-functionsituationoccursduringrepairsofmultiple, inoperable, untripped instrument channels.

In regard to ACTIONS 30, 31, 34 and 39,

" automatic actuation capability" is considered to be maintained when sufficient channels are OPERABLE (or are in the tripped condition for ACTIONS 30 and 34) such that each " Trip Function" identified in Table 3.3.3-1 is capable of initiating an ECCS function upon receipt of a valid signal from that " Trip Function" (without the need to consider a further single failure event.

For ECCS Divisions 1 and 2, each Trip Function should be able to initiate either) Division 1 or Div for ADS Trip Systems A and B, each ADS Trip Function should be able to initiate e System A or Trip System B; and for HPCS, the logic should be able to initiate HPCS.

The Trip Functions identified in Table 3.3.3-1 (except for those in Section D of the Table) are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated ACTIONS may be delayed as follows:

b) for up)to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Trip Functions (a for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Trip functions C.l.f. C.1.g, and C.I.h; and otherthanC.I.f,C.I.g,andC.I.hprov(dedtheassociatedTripFunctionorthe i

in the other Division maintains ECCS initiation redundant Trip Function (ion of the Surveillanc)e, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months capability. Upon complet allowance, the channel must be returned to OPERABLE status or the applicable ACTIONS taken. This Note is based on the reliability analysis assumption that 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance does not significantly reduce the probability that the ECCS will initiate when necessary.

i Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the oasis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.

PERRY - UNIl 1 B 3/4 3-2a Amendment No. 67 i

Attachm:nt 4 INSTRUMENTATION PY-CEI/NRR-1932 L Page 4 of 4 BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INd1RUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349, dated March 1971 and NED0-24222, dated December 1979, and Section 15.8 of the FSAR.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with GENE-770-06-01, " Bases for Changes to Surveillance Test Intervals and Allowed Out-0f-Service Times for Selected Instrumentation Technical Specifications" as approved by the NRC and documented in the NRC Safety Evaluation Report (SER) letter to R.D. Binz from C.E. Rossi dated July 21, 1992.

The end-of-cycle recirculation pump trip (E0C-RPT) system is an essential safety supplement to the Reactor Protection System. The purpose of the EOC-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle.

The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity.

Each E0C-RPT system trips both recirculation pumps, reducing coolant flow in order to reduce the void ccllapse in the core during two of the most limiting pressurization events.

The two events for which the E0C-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.

A fast closure sensor from each of two turbine control valves provides in. nut to the EOC-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second E0C-RPT system.

Similarly, a position switch for each of two turbine stop valves provides input to one E0C-RPT system; a pcsition switch from each of the other two stop valves provides input to the other EOC-RPT system.

For each E0C-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The operation of either logic will actuate the E0C-RPT system and trip both recirculation pumps.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with GENE-770-06-01, " Bases for Changes to Surveillance Test Intervals and Allowed Out-0f-Service Times for Selected Instrumentation Technical Specifications" as approved by the NRC and documented in the NRC SER letter to R. D. Binz from C. E. Rossi dated July 21, 1992.

Each E0C-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled.

The manual bypasses and the automatic Operating Bypass at less than 40% of RATED THERMAL POWER are annunciated in the control room.

The EOC-RPT s3 stem response time is the time assumed in the analysis between i

at, ion of valve motion and complete suppression of the electric arcQ@

ncluded in this time are:

the time from initial valve movement to reac ing the trip setpoint, the response time of the sensor, the response time of the system logic, and the time allotted for breaker arc suppression.

PERRY - UMT 1 B 3/4 3-3 Amendment No. 67

Definitions Attcchscnt 5 1.1 PY-CEI/NRR-1932 L Page 1 of 17 1.1 Definitions (continued) a EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM (ECCS) RESPONSE from when the monitored 3arameter exceeds its ECCS TIME initiation setpoint at t1e channel sensor until the ECCS equipment is ca)able of performing its safety function (i.e., t1e valves travel to their required positions, pump discharge pressures reach their required values. etc.).

Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. g __._

M END OF' CYCLE The EOC-RPT SYSTEM RESPONSE TIME shall be that RECIRCULATION PUMP TRIP time interval from initial movement of the -

(EOC-RPT) SYSTEM RESPONSE associated turbine stop valve or the turbine TIME control valve to complete suppression of the electric arc between the fully open contacts of the recirculation Jump circuit breaker. The response time may )e measured by means of any series of sequential. overlapping, or total steps so that the entire response time is measured.

ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that RESPONSE TIME time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. 4---

L, The maximum allowable primary containment leakage rate. L,, shall be 0.20% of primary containment air weight per day at the calculated peak containment pressure (P ).

(continued)

%qb o re Si d td b h [r clM d vd beotM m CLpr t.d f

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l PERRY - UNIT 1 1.0-3 Amendment No. C l

Attcchment 5 Definitions PY-cEIMRR-1932 L 11 Page 2 of 17 1.1 Definitions (continued)

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core for each class of fuel.

The CPR is that power in the assembly that is calculated by application of the a)propriate correlation (s) to cause some point in t1e assably to experience boiling transition.

divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE -0PERABILITY A system, subsystem, division, com)onent or device shall be OPERABLE or have 03ERABILITY when it is capable of aerforming its specified safety function (s) and w1en all necessary attendant instrumentation, controls, normal or emergency electrical power. cooling and seal water.

lubrication, and other auxiliary equipment that are required for the system, subsystem division, component, or device to perform its specified safety function (s) are a' Iso capable of performing their related support function (s).

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 3579 MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until A

de-energization of the scram pilot valve solenoids.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.c 3

(continued) b ( 5-tb cu t s, U < ) k % sn f, fsj % \\ s e v e,\\\\ m ftsprt.m(n\\3.}

PERRY - UNIT 1 1.0-5 Amendment No. C

Attcchment 5 RPS Instirumentation PY-CEI/NRR-1932 L 3.3.1.1-

-Page 3 of 17 SURVEILLN1CE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.16 i'erify Turbine Stop Valve Closure and 18 months Turbine Control Valve Fast Closure Trip 011 Pressure-Low Functions are not bypassed when THERMAL POWER is a 40% RTP.

SR 3.3.1.1.17 Calibrate flow reference transmitters.

18 months SR 3.3.1.1.18

NOTES------------------

1.

Neutron detectors are excluded.

A.Ecr L d.;,s1 3X For Function 6. "n" equals 4 channels for the aurpose of determining the N ^^J V is STAGGERE) TEST BASIS Frequency.

LW 3 33,i_,

A c6nnd Verify the RPS RESPONSE TIME is within 18 months on a Sons r s m limits.

STAGGERED TEST cxdv4 BASIS PERRY - UNIT 1 3.3-6 Amendment No. C

Attcchment 5 ECCS Instrumentation PY-CEI/NRR-1932 L 3.3.5.1 Page 4 of 17 SURVEILLANCE REA iMENTS

.._....______________..._______.-----NOTES------------------------------------

1.

Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.

2.

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows:

(a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 3.c. 3.f. 3.g. and 3.h: and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions other than 3.c. 3.f. 3.g. and 3.h, provided the associated Function or the redundant Function maintains ECCS initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.5.1.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST.

92 days SR 3.3.5.1.3 Calibrate the tr., unit.

92 days SR 3.3.5.1.4 Perform CHANNEL CALIBRATION.

92 days SR 3.3.5.1.5 Perform CHANNEL CALIBRATION.

18 months SR 3.3.5.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST.

18 months l

SR 3.

.1.7 Verify the CS RESPONSE TIM 5 e within 18 nths e a limits.

STAGG ED TE BASIS PERRY - UNIT 1 3.3-38 Amendment No. C l

l

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Attechment 5 ECCS Instrumentation PY-CEI/NRR4932 L 3.3.5.1 Page 5 of 17 Table 3.3.5.1-1 (page 1 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER REQUIRED FRJM SPECIFIED CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE 1.

Low Pressure Coolan' Injection-A (LPCI) and Low Pressure Core Spray (LPCS)

Stbsystems a.

Reactor vessel Water 1,2,3, 2(b)

B SR 3.3.5.1.1 a 14.3 inches Level - Low Low Low, SR 3.3.5.1.2 Level 1 4(*),5(a)

SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 Q^; _:.:..

-)%-

b.

Drywell Pressure - High 1,2,3 2(b)

B SR 3.3.5.1.1 s 1.88 psig SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 QL L L L *.?p c.

LPCI Pump A 1,2,3, 1

C SR 3.3.5.1.2 s 5.25 Start - Time Delay SR 3.3.5.1.4 seconds Retsy 4(a) 5(a)

SR 3.3.5.1.6 d.

Reactor Vessel 1,2,3 1

C SR 3.3.5.1.1 a 482.7 psig Pressure - Low (LPCS SR 3.3.5.1.2 and Injection valve SR 3.3.5.1.3 s 607.7 psig Permissive)

SR 3.3.5.1.5 SR 3.3.5.1.6 4(a) 5(*)

1 B

SR 3.3.5.1.1 a 482.7 psig SR 3.3.5.1.2 and SR 3.3.5.1.3 s 607.7 psig SR 3.3.5.1.5 SR 3.3.5.1.5 e.

Reactor vessel 1,2,3 1

C SR 3.3.5.1.1 a 462.5 psig Pressure-Low (LPCI SR 3.3.5.1.2 and Injection valve SR 3.3.5.1.3 s 512.5 psis Permissive}

SR 3.3.5.1.5 SR 3.3.5.1.6 4(a), $(a) 1 B

SR 3.3.5.1.1 a 462.5 psig SR 3.3.5.1.2 and SR 3.3.5.1.3 s 512.5 psig SR 3.3.5.1.5 SR 3.3.5.1.6 F.

LPCS Pung Discharge 1,2,3, 1

E SR 3.3.5.1.1 a 1200 gpm Flow - Low (Bypass)

SR 3.3.5.1.2 4(a) 5(a)

SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 (continued)

(a) When associated subsystem (s) are required to be OPERABLE.

(b) Also required to init' ate the associated diesel generator and AEGT subsystem.

PERRY - UNIT 1 3.3-39 Amendment No. C

n s.

+

..+..a x

'. Attechment 5 ECCS Instrumentation PY-CEI/NRR-1932 L 3.3.5.1 Page 6 of 17 Table 3.3.5.1-1 (page 2 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER REQUIRED FROM i

-SPECIFIED CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQU?REMENTS VALUE 1.

Low Pressure Coolant injectlon-A (LPCI) and Low Pressure Core Spray (LPCS)

Subsystems (contfrued) g.

LPCI Pump A Discharge 1,2,3, 1

E SR 3.3.5.1.1 a 1450 gpm Flow - Low (Bypass)

SR 3.3.5.1.2 4(*),5(a)

SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 h.

Manual Initiation 1,2,3, 1

C SR 3.3.5.1.6 NA 4(a) $(a) 2.

LPCI B and LPCI C Subsyste's a,

Reactor vesset water 1,2,3, 2(b)

B SR 3.3.5.1.1

= 14.3 inches Level - Low Low Low, SR 3.3.5.1.2 Levet 1 4(a) 5(a)

SR 3.3.5.1.3 SR 3.3.5.1.5 sn 3.3.5.1.6 i

.22.'$

i b.

Drywell Pressure-Nigh 1,2,3 2(b) 8 SR 3.3.5.1.1 s 1.88 psis SR 3.3.5.1.2 i

SR 3.3.5.1.3 SR 3.3.5.1.5 i

tn 3.3.5.1.6 5

?_3** @

c.

LPCI Pump e 1,2,3, 1

C SR 3.3.5.1.2 s 5.25 start - Time Delay SR 3.3.5.1.4 seconds Relay 4(*),5(*)

SR 3.3.5.1.6 d.

Ree: tor Vessel 1,2,3 1 per C

SR 3.3.5.1.1 a 468.0 psig Prossure-Low (LPCI subsystem SR 3.3.5.1.2 and i

Injection Valve SR 3.3.5.1.3 s 518.0 psis Permissive)

SR 3.3.5.1.5 for LPCI B; SR 3.3.5.1.6 and a 466.6 psig and s 516.6 psig for LPCI C 4(a) 5(a) 1 per 8

SR 3.3.5.1.1 a 468.0 psig subsystem SR 3.3.5.1.2 and SR 3.3.5.1.3 s 518.0 psig SR 3.3.5.1.5 for LPCI S; j

SR 3.3.5.1.6 and a 466.6 psig and 5 516.6 psig for LPCI C (continued) i (a) When associated stbsystem(s) are required to be OPERA 8LE.

j (b) Also required to initiate the associated diesel generator and AEGT subsystem.

I PERRY - UNIT 1 3.3-40 Amendment No. C

i l

i Atttchment 5 ECCS Instrumentation PY-CEI/NRR-1932 L 3.3.5.1 Page 7 of 17 Table 3.3.5.1 1 (page 3 of 5)

Emergency Core Cooling System Instrunentation 1

APPLICABLE CONDITIONS MODES OR REFERENCED OTHER REQUIRED FROM SPECIFIED CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE 2.

LPCI B and LPCI C Subsystems (continued) e.

LPCI Fu, B 1,2,3, 1 per punp E

SR 3.3.5.1.1 a 1450 gpm and LPCI Punp C SR 3.3.5.1.2 Discharge 4(a) 5(a)

SR 3.3.5.1.3 F t ow - Low SR 3.3.5.1.5 (Bypass)

SR 3.3.5.1.6 f.

Manual Initiation 1,2,3, 1

C SR 3.3.5.1.6 NA 4(a) 5(a) 3.

High Pressure Core Spray (HFCS) System a.

Reactor Vessel 1,2,3, 4(b)

B SR 3.3.5.1.1 a 127.6 inches Water Level - Low SR 3.3.5.1.2 Low, Level 2 4(a) 5(a)

SR 3.3.5.1.3 rJt 3.3.5.1.5 SR 3 3_S.

k 3.3.3.i.?

b.

Drywell 1,2,3 4(b)

B SR 3.3.5.1.1 s 1.88 psig Pressure - High SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 ii;: :.:.M ',DL c.

Reactor Vessel 1,2,3, 4

B SR 3.3.5.1.1 s 221.7 inches Water SR 3.3.5.1.2 L evel - High, de)$(a)

SR 3.3.5.1.3 Level 8 SR 3.3.5.1.5 SR 3.3.5.1.6 d.

Condensate 1,2,3, 2

D SR 3.3.5.1.1 a 59,700 gettons Storage Tank SR 3.3.5.1.2 Level - Low 4(C) 5, CCI SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 e.

Suppression Pool 1,2,3 2

D SR 3.3.5.1.1 s 18 ft 6 inches Water Level - High SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 (continued)

(a) When associated subsystem (s) are required to be OPERABLE.

(b) Also required to initiate the associated diesel generator.

(c) When HPCS is OPERABLE for conpliance with LCO 3.5.2, "ECCS - Shutdown," and aligned to the condensate storoge tank while tank water level is not within the limit of SR 3.5.2.2.

PERRY - UNIT 1 3.3-41 Amendment No. C i

Attech=nt 5 Primary Containment and Drywell Isolation Instrumentation PY-CEI/NRR-1932 L 3.3.6.1 Page 8 of 17 SURVEILLANCE REQUIREMENTS l

..______._____________.______.....---NOTES------------------------------------

1.

Refer to Table 3.3.6.1-1 to determir.e which SRs apply for each Function.

2.

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided the associated Function maintains isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST.

92 days SR 3.3.6.1.3 Calibrate the trip unit.

92 days SR 3.3.6.1.4 Perform CHANNEL CALIBRATION.

18 months SR 3.3.6.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST.

18 months duti.We SR 3.3.6.1.6 erify the ISOLATION SYSTEM RESPONSE TIME 18 months on a for the main steam isolation valves is STAGGERED TEST within limits.

BASIS

- - - - hb \\ t _ - - - - _ _

d S e n.h t s art tw tLJ t)

PERRY - UNIT 1 3.3-53 Amendment No. C

Attcchment 5 ECCS -Operating PY-CEI/NRR-1932 L 3.5.1 Page 9 of 17 SURVEILLANCE REOUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.5

NOTE--------------------

Vessel injection / spray may be excluded.

Verify each ECCS injection / spray subsystem 18 months actuates on an actual or simulated automatic initiation signal.

SR 3.5.1.6

NOTE--------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or 18 months simulated automatic initiation signal.

SR 3.5.1.7

NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify each ADS valve opens when manually 18 months on a actuated.

STAGGERED TEST BASIS for each valve solenoid r %..- _ _ _ _

S R. 1 s. l.9, Ecesa u g*g, w y; h 6 cA v d e d Wid, N ECU REif oW sE 11 mom 3

/

h + F_ Eg.eatb F_C. U sE ttM- / spay sA4 4ht - ir i

3 J,% L?t 3 f

D-PERRY - UNIT 1 3.5-5 Amendment No. C

~Attechnent 5-RPS Instrumentation PY-CEI/NRR-1932 L B 3.3.1.1 i

Page 10 of 17 BASES SURVEILLANCE SR 3.3.1.1.16 (continued)

REQUIREMENTS If any bypass channel setpoint is nonconservative (i.e.

the Functions are bypassed at a 40% RTP either due to open main i

turbine bypass valve (s) or other reasons), then the affected Turbine Stop Valve Closure and Turbine Control Valve Fast Closure. Trip Oil Pressure-Low Functions are considered i

inoperable. Alternatively, the by) ass channel can be placed in the conservative condition (non)ypass).

If placed in the nonbypass condition, this SR is met and the channel is considered OPERABLE.

The Frequency of 18 months is based on engineering judgment and reliability of the components.

SR 3.3.1.1.18 This SR ensures that the individual channel response times t

are less than or equal to the maximum values assumed in the

]?*djbyhe accident analysis. The RPS RESPONSE TIME acceptance criteria are included in Reference 10.

bnMmQ d As noted, neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation E N 43 A E\\ c.J virtually ensure an instantaneous response time. p k RPS RESPONSE TIME tests are conducted on an 18 month

" * " ^

t

% + r tA b b STAGGERED TEST BASIS.

Note 2 requires STAGGERED TEST BASIS Frequency to be determined based on 4 channels er trip

{

D"*'k*'

system, in lieu of the 8 channels specified in able it.A<J. %," % 3t 3.3.1.1-1 for the MSIV Closure Function. This Frequency is-based on the logic interrelationships of the various p*

D " *-

channels required to produce an RPS scram signal.

t

.k. b s k es W n g (,

Therefore, staggered testing results in res)onse time This

erification of these devices every 18 montis.

v N Com ^ 3 bd requency is consistent with the typical industry refueling tot.nd s g3

ycle and is based upon plant operating experience, which phows that random failures of instrumentation components l

W "g

VJua r u n.ttdg g causing serious time degradation, but not channel failure.

oNwa m (. UN are infrequent.

J b 3 REFERENCES 1.

USAR. Figure 7.2-1.

2.

USAR. Section 5.2.2.

3.

USAR. Section 6.3.3.

(continued)

PERRY - UNIT 1 B 3.3-31 Revision No. C

Attcchment 5 RPS Instrumentation PY-CEI/NRR-1932 L B 3.3.1.1 Page 11 of 17 BASES l

REFERENCES 4.

USAR. Chapter 15.

(continued) 5.

USAR. Section 15.4.1.

l 6.

NED0-23842. " Continuous Control Rod Withdrawal in the Startup Range." April 18, 1978.

7.

USAR. Section 15.4.9.

8.

Letter. P. Check (NRC) to G. Lainas (NRC). "BWR Scram Discharge System Safety Evaluation." December 1.1980, as attached to NRC Generic Letter dated December 9, 1980.

9.

NED0-30851-P-A " Technical Specification Improvernent Analyses for BWR Reactor Protection System."

March 1988.

10.

GE DSDS 22A3771AJ.

1 l

S dem And su ll. NF O o -31L9 ;;.

r 3

3 3

EliaA Am 8 SA&J L.spsse Ge T e. A 'n g k<. g v s rce (d.1 ds ov ar 5 \\% 9 3

i 5

PERRY - UNIT 1 B 3.3-32 Revision No. C

Attcch2:nt 5 ECCS Instrumentation PY-CEI/NRR-1932 L B 3.3.5.1 Page 12 of 17 BASES (continued)

SURVEILLANC SR 3.3.5.1.7 A

REQUIREMENTS (continued)

This SR ensures that t individual channel ponse times re less than or equal t the maximum values as med in the l

a ident analysis.

Respon time testing accepta e cri eria are included in Re ence 5.

ECCS R PONSE TIME tests are co ucted on an 18 month STAGGER TEST BASIS. This Frequ cy is consistent with ge typical i stry refueling cycle a is based upon plant N

operating e erience, which shows th random failures of instrumentati components causing ser us response time

's degradation, but et channel failure, ar infrequent.

s, s

REFERENCES 1.

USAR. Section 5.2.

2.

USAR. Section 6.3.

3.

USAR, Chapter 15.

4.

NEDC-30936-P-A. "BWR Daners' Group Technical Specification Improvement Analyses for ECCS Actuation Instrumentation. Part 2." December 1988.

h ARJ Sect h 6.3. Tab k 6.3 h M i

I 1

PERRY UNIT 1 B 3.3-123 Revision No. C

Attachmont 5 Primary Containment Isolation Instrumentation PY-CEI/NRR-1932 L B 3.3.6.1 Page 13 of 17 BASES SURVEILLANCE SR 3 3.6.1.4 REQUIREMENTS (continued)

CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.

The Frequency of SR 3.3.6.1.4 is based on the assumption of the magnitude of equipment drift in the setpoint analysis.

SR 3.3.6.1.5

~

The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel.

The system functional testing performed on PCIVs in LCO 3.6.1.3 and on drywell isolation valves in LCO

! g' g' 3'.

7 3.6.5.3 overlaps this Surveillance to provide complete testing of the assumed safety funcMon. The 18 month

'3 SQ 3,.3.b. l. (,'

Frequency is based on the need to s erform this Surveillance under the conditions that apply during a plant outage and g(" gd the potential for an unplanned transient if the Surveillance ckm%\\s % 3 were performed with the reactor at power.

Operating experience has shown these components usually pass the o re_ ex d v h / Fr w Surveillance when performed at the 18 month Frequency.

O. 3 p e. E h Ta.Ah rputh, SR 3.3.6.1.6 Ro q snse. b h.3 %

This SR ensures that the individual channel response times g

D are less than or equal to the maximum values assumed in the YO 9 % Ne ng accident analysis. Testing is performed only on channels Cg*

g ^P"'4#

fwheretheassumedresponsetimedoesnotcorrespondtothe i

For channels assumed to

{dieselgenerator(DG)starttime.

b t qv,'re). % 3

. respond within the DG start time, sufficient margin exists in the 10 second start time when compared to the typical j.s3so(bA tl h channel response time (milliseconds) so as to assure adequate response without a specific measurement test. The UU mCA 9.

[ instrument response times must be added to the PCIV closure l

J times to obtain the ISOLATION SYSTEM RESPONSE TIME.

ISOLATION SYSTEM RESPONSE TIME acceptance criteria are I

included in References 7 and 8.

i>

(continued)

PERRY UNIT 1 B 3.3-173 Revision No. C

Attechnent 5-Primary Containment Isolation Instrumentation PY-CEI/NRR-1932 L B 3.3.6.1

- Page 14 of 17 BASES SURVEILLANCE SR 3.3.6.1.6 (continued)

REQUIREMENTS ISOLATION SYSTEM RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. This test Frequency is '

consistent with the typical industry. refueling cycle and is based upon plant operating experience that shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent.

REFERENCES 1.

USAR. Section 6.3.

2.

USAR, Chapter 15.

3.

NED0-31466. " Technical Specification Screening Criteria Application and Risk Assessment."

November 1987.

4.

USAR. Section 9.3.5.

5.

NEDC-31677-P-A. " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation."

June 1989.

6.

NEDC-30851-P-A Supplement 2. " Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation." March 1989.

7.

USAR. Section 15.1.3.

8.

USAR Section 15.6.

D 1

0 - h19 1, S

hm bnh sc3 Ee r y

D 4 L e s & A l L y,-3c T L b h n 3 e_ g [ rc + e_ n\\ J [ k n v a r $

9 M 6%.

W PERRY UNIT 1 B 3.3-174 Revision No. C 1

___,e,-

.=.

I ECCS-Operating l

PY-CEI/NRR-1932 L B 3.5.1 Page 15 of 17 BASES SURVEILLANCE SR 3.5.1.7 (continued) i REQUIREMENTS performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified. per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

SR 3.5.1.6 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.

The Frequency of 18 months on a STAGGERED TEST BASIS ensures that both solenoids for each ADS valve are alternately tested. The Frequency is based on the need to perform this Surveillance under the conditions that apply just prior to or during a startup from a plant outage and the )otential for un)lanned transients.

Operating experience las shown that t1ese components usually pass the SR when performed at the 18 month Frequency. which is based on the refueling TA SEQ _T cycle. Therefore, the Frequency was concluded to be g

acceptable from a reliability standpoint.

REFERENCES 1.

USAR. Section 6.3.2.2.3.

2.

USAR. Section 6.3.2.2.4.

3.

USAR. Section 6.3.2.2.1.

4.

USAR. Section 6.3.2.2.2.

5.

USAR. Section 15.6.6.

6.

USAR. Section 15.6.4.

7.

USAR. Section 15.6.5.

8.

10 CFR 50. Appendix K.

9.

USAR. Section 6.3.3.

10.

10 CFR 50.46.

11. USAR. Section 6.3.3.3.

(continued)

PERRY - UNIT 1 B 3.5-13 Revision No. C

^

i

Attcchment 5 PY-CEI/NRR-1932 L Page 16 of 17 INSERT B13 SR 3.5.1.8 This SR ensures that the ECCS RESPONSE TIMES are within limits for each of the ECCS injection and spray subsystems. This SR is modified by a note which identifies that the associated ECCS actuation instrumentation is not required to be response time tested. Response time testing of the remaining subsystem components is required. This is supported by Reference 14.

Response time testing acceptance criteria are included in reference 15.

ECCS RESPONSE TIME tests are conducted every 18 months. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor.

at power. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Attcchsent 5 ECCS~-Operating PY-CEI/NRR-1932 L B 3.5.1 Page 17 of 17 BASES REFERENCES 12.

Memorandum from R.L. Baer (NRC) to V. Stello. Jr.

(continued)

(NRC). " Recommended Interim Revisions to LCO's for ECCS Components." December 1. 1975.

13.

USAR. Section 5.2.2.4.1.

kr N. NEOO - 3M S t, D "Syhm knalyu b E l E, nd [r oS $ d u h t.l k u e s n n % b Te sh ng EtgviOt me d i, Senvcr3 R 1%.

\\$. V ShR,Sttk1E fo.3, Tcd t (o.3 - l.

L l

l PERRY - UNIT 1 B 3.5-14 Revision No. C

ML20087H891

Text

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