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Proposed Tech Specs Re one-time Extension of Performance Intervals for Certain TS Surveillance Requirements
	ML20081K319
Person / Time
Site:	Perry
Issue date:	03/24/1995
From:	; CENTERIOR ENERGY
To:	;
Shared Package
ML20081K315	List: ML20081K319; PY-CEI-NRR-1890, Application for Amend to License NPF-58 Re one-time Extension of Performance Intervals for Certain TS Surveillance Requirements;
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{{#Wiki_filter:3/4.3 INSTRUMENTATION

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3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION ! LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM i RESPONSE TIME as shown in Table 3.3.1-2. 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 Table 3.3.1-1 inoperable in one or more Functional Units;

1. Within one hour, verify sufficient char.nels remain OPERABLE or are in the tripped condition

to maintain trip capability in the Functional Unit, and

2. in one trip system Within and/or 6 hours, that place the trip system ** ininoperable the trippedchannel condit (s) ion,* and

3. restore the inoperable channels in the other trip Within12 system to an hours OP $RABLEstatusorplacetheminthetrippedcondition*.

Otherwise, take the ACTION required by Table 3.3.1-1 for the Functional Unit. SURVEILLANCE REGUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the pB TIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated auto 'c operation of all channels shall be performed at least once per 18 months. p

4.3.1.3 The REACTOR PROTECTION EM RESPONSE TIME of each reactor trip functional unit shown in Table shall be demonstrated to be within its limit st once per 18 mon ch test shall include at least one channel per tem such that all channels are tested at least once every N times 18 mon N is the total number of redundant channels in a specific reactor trip system. j 4.3.1.4 The provisions of Specification 4.0.4 are not applicable to the I CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION surveillances for the Intermediate Range Monitors for entry into their ap licable OPERATIONAL CONDITIONS Sg the surveil 1(as shown ances areinperformed Table 4.3.1.1-1) withinfrom 12 OPERA.IONAL hours after suchCONDITION entry. 1, provided I$$ WO O C)

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 oU required time, the ACTION required by Table 3.3.1-1 for the Functional Unit 88 shall be taken. o< y ** This ACTION applies to that trip system with the most inoperable channels; if or o ip systems have the same number of inoperable channels, the ACTION can y@o, be applied to either trip system. leutr A ) PERRY - UNIT 1 3/4 3-1 Amendment No. -44, -66, 67

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x y TABLE 3.3.1-2 9 Q REACTOR PROTECTION SYSTEM RESPONSE TIMES I C z RESPONSE TIME FUNCTIONAL UNIT (Seconds)

1. Intermediate Range Monitors:

a. Neutron Flux - High NA

b. Inoperative NA

2. Average Power Range Monitor :

a. Neutron Flux - High, Setdown NA

b. Flow Biased Simulated Thermal Power - High 5 0.09' #

c. Neutron Flux - High 50.09 54

d. Inoperative NA

3. Reactor Vessel Steam Dome Pressure - High $ 0.

w 4. Reactor Vessel Water Level - Low, Level 3 5 1. 1 5. Reactor Vessel Water Level - High, Level 8 s 1. w 6. Main Steam Line Isolation Valve - Closure 50, a 7. Deleted l

8. Drywell Pressure - High NA

9. Scram Discharge Volume Water level - High NA

10. Turbine Stop Valve - Closure s 0.06

11. Turbine Control Valve Fast Closure, Valve Trip System Oil Pressure - Low s 0.07#

3 12. Reactor Mode Switch Shutdown Position NA 2 13. Manual Scram NA h o

                                  =      ' Neutron detectors are exempt from response time testing.         Response time shall be measured from P        the detector output or from the input of the first electronic component in the channel.
                                        "Not including the simulated thermal power time constant specified in the COLR.
                                         # Measured from start of turbine control valve fast closure.

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A TABLE 4.3.1.1-1 E REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 7 E CHANNEL OPERATIONAL Z CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATIONfa) SURVEILLANCE REQUIRED

l. Intermediate Range Monitors:

a. Neutron Flux - High S/U,S,(b) W R 2 S W R 3,4,5

b. Inoperative NA W NA 2,3,4,5

2. Average Power Range Monitor:(f) .

a. Neutron flux - High, S/U,S,(b) W SA 2 Setdown S W SA 3, 5

b. Flow Biased Simulated M ta Themal Power - High 5,0(h) g g(d)(e) gg(m) y c. Neutron Flux - High S Q W(d), SA 1

d. In., perative NA Q NA 1, 2, 3, 5

3. Reactor Vessel Steam Dome Pressure - High S Q R Q(0) 1, 20 '

4. Reactor Vessel Water level - ge}

Low, level 3 S Q 1, 2 j 5. Reactor Vessel Water Level - High, Level 8

                                                                                 ')(n)(o) sI

g. S Q k 6. Main Steam Line Isolation g Valve - Closure NA Q R 1

_g 7. Deleted .u

                                                                          ~
                  @                                                                          TABLE 4.3.I'1-1 (Continued)

=

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 7 s Z (d) This calibration shall consist of the adjustment of the APRM channel to confom to the power values

                  -         calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 125% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of AATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reaching 25% of RATED THERMAL POWER. To functionally implement this protective function during entry into single loop operation, APRM channel gain adjustments may be made in lieu of adjusting the APRM Flow Blased Simulated Thermal Power-High Trip setpoint and Allowable Value equations for a period not to exceed 72 hours, provided the criteria in Note b to Table 2.2.1-1 are met. Any APRM channel gain' adjustments made i                            in compliance with Specifications 2.2.1 and 3.3.1 shall not be included in determining the' absolute difference.

w (e) This calibration shall consist of the adjustment of the APRM flow biased channel to confom to a calibrated 1 flow signal. (f) The LPRMs shall be calibrated at least once per 1000 MWD /T using the TIP system. (g) Calibrate trip unit setpoint at least once per 92 days. l (h) Verify measured core flow (total core flow) to be greater than or equal to established core flow at the existing loop flow (APRM % flow). (i) This calibration shall consist of verifying that the simulated thermal pomr time constant is within the limits specified in the COLR. k (j) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per E Specification 3.10.1. 2 f+ (k) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or y 3.9.10.2. a (1) This function is not required to be OPERABLE when Drywell Integrity is n>t required. r , ps 3 'O ps a ) The MCALIBRATION shall exclude the low reference transmitters, these transmitters sha11 D x ' P calibrated at least once per 18 months, uc.ur Ek.& au test me.y b<. u.hard to be

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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 t PERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1 4.3.2.2 LOGIC# SYSTEM FUNCTIONAL TESTS and simulated automa

eration 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 Table 3. . hall be demonstrated to be within its limit at least once per 18 month hh test shall include at least one channel per trip r, system such that all channels are tested at least once every N times 18 monthbs# where N is the total number of redundant channels in a specific isolation trip system. s C6xne) Calibub yn'od weq beedended a idenEfed bI nobe 'c' on ' 7aMe 4.3.1 - ( . e Lo3h .Sgsb Cmba t T<a priod q t<ex6,d ed as 3 id<eha 6 no te '4' on luk 4.3.1- t

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e ? TABLE 3.3.2-3. ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME

                - TRIP FUNCTION                                      RESPONSE TIME (Seconds)#               ,

1. , PRIMARY CONTAINMENT ISOLATION

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

b. Drywell Pressure - High .

NA :

c. Containment and Dryll Purge Exhaust Plenum Radiation - High* $ 10(* g

d. Reaqtor Vessel Water Level - Low, Level 1 NA .

e. Maridal Initiation NA

2. MAIN STEAM LINE ISOLATION q ,

Reactor Vessel Water Level - Low, Level I !

a. s 1.0*/s 10(*)

b. Main Steam Line Radiation - High NA I#3 l' t

c. Main Steam Line Pressure - Low < l .0*/s 10(*)*

d. Main Steam Line Flow - High 2 0.5*/s 10(*) d l e. Condenser Vacuum - Low NA I f. Main Steam Line Tunnel Temperature - High NA

g. Main Steam Line Tunnel A Temperature - High NA

h. Turbine Ballding Main Steam Line i Temperature - High NA

            . .         1. Manual Initiation                                 NA                  ,

3. SECONDARY CONTAINMENT ISOLATION .

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

b. Drywell Pressure - High NA _

c. Manual Initiation NA ,

4. REACTOR WATER CLEANUP SYSTEM ISOLATION

a. A Flow - High NA

b. A Flow Timer NA

c. Equipment Area Temperature - High NA

d. Equipment Area A Temperature -'High NA ,

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

f. Main Steam Line Tunnel Ambient Temperature - High .NA

g. Main Steam Line Tunnel A Temperature - High NA

h. SLCS Initiation NA i

1. Manual Initiation NA PERRY - UNIT 1 3/4 3-21 Amendment NoS8 ,

TABLE 3,3.2-3 (Continued) ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#

5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION

a. RCIC Steam Line Flow - High NA

b. -RCIC Steam Supply Pressure - Low NA

c. RCIC Turbine Exhaust Diaphragm Pressure - High NA

d. RCIC Equipment Room Ambient Temperature - High NA

e. Deleted

f. Main Steam Line Tunnel Ambient l- .

Temperature - High NA

g. Main Steam Line Tunnel A Temperature - High NA

h. Main Steam Line Tunnel Temperature Timer NA

1. RHR Equipment Room Ambient Temperature - High NA

j. RHR Equipment Room A Temperature - High NA l

k. RCIC Steam Line Flow High Timer NA ;

1. Drywell Pressure - High NA

m. Manual Initiation NA

6. RHR SYSJEM ISOLATION

a. RHR Equipment Area Ambient Temperature - High NA

b. RHR Equipment Area A Temperature - High NA

c. RHR/RCIC Steam Line Flow - High NA

d. Reactor Vessel Water Level - Low, Level 3 NA e.

Reactor Vessel (RHR Cut-in Permissive) Pressure - High NA

f. Drywell Pressure - High NA

g. Manual Initiation NA (a) Isolation system instrumentation response time specified includes the.

diesel generator starting and sequence loading delays. , (b) Radiation detectors are exempt from response time testing. Resp ee ti-^ ht irs elec%e measured from The trontrcomponent'in detector output or the input ,oJA thaard. {c) fe9pmWtesho my R o kdat te h eeMW (ofg$ ,,gA f o 7e, , e 4tnr sfries, in ru on r . No generator delays assumed. Isolation system instrumentation response time for associated valves except MSIVs.

      # Isolation sy:, tem instrumentation response time specified for the Trip Function actuating each containment isolation valve shall be added to the isolation time for each valve to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.

PERRY - UNIT 1 3/4 3-22 Amendment No. AA,59

g TABLE 4.3.2.1-1 f ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE RE0VIREMENTS E CHANNEL OPERATIONAL-Z CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REOUIRED

1. PRIMARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level -

Low, level 2 S Q R)C 1, 2,- 3 and #

b. Drywell Pressure - High ## S Q R(b) 1, 2, 3

c. Containment and Drywell Purge Exhaust Plenum Radiation -

High S Q 1, 2, 3 and * -1

d. Reactor Vessel Water Level -

Low, level 1 S R( (f) 1, 2, 3 and # l

e. Manual Initiation NA R NA 1, 2, 3 and

5 2. MAIN STEAM LINE ISOLATION w a. Reactor Vessel Water Level -

A, Low, Level 1 S Q Rb )E 1, 2, 3 l

b. Main Steam Line Radiation - gqg)

High S Q R *** l

c. Main Steam Line Pressure -

Low S Q Rcb> g

d. Main Steam Line Flow - High S Q R 1, 2, 3

e. Condenser Vacuum - Low S Q R(b) 1, 2**, 3**

f. Main Steam Line Tunnel
&" Temperature - High S Q R 1, 2, 3 l @ g. Main Steam Line Tunnel & A Temperature - High S Q R 1, 2, 3 [. @ h. Turbine Building Main Steam Line Temperature - High S Q R 1, 2, 3 l
1. Manual Initiation NA R NA 1, 2, 3 8

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m A E t 5 TABLE 4.3.2.I-I (Continued) Z m ISOLATION ACTUATION INSTRUNENTATION SURVEILLANCE REQUIREMENTS
~
i I CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL TRIP FUNCTION CONDITIONS IN WHICH
. CHECK TEST CALIBRATION SURVEILLANCE REQUIRED
3. SECONDARY CONTAINNENT ISOLATION

a. Reactor Vessel Water Level - Low, level 2 S Q RS )(')(d) 1, 2, 3 and #

b. Drywell Pressure - High ## S Q 1, 2, 3

c. Manual Initiation NA R NA 1, 2, 3 and *

4. REACTOR WATER CLEANUP SYSTEM ISOLATION w a. A Flow - High
'a b. A Flow Timer S Q R 1, 2, 3 w c. Equipment Area Temperature -
NA Q R 1, 2, 3
~ High
d. Equipment Area Ventilation S Q R 1, 2, 3 A Temperature - High S Q R 1, 2, 3

e. Reactor Vessel Water Level - Low, level 2 S Q RS ) I')C ) 1, 2, 3

f. Main Steam Line Tunnel Ambient Temperature - High S Q R 1, 2, 3

g. Main Steam Line Tunnel A Temperature - High S Q R 1, 2, 3 g h. SLCS Initiation NA Q(*) NA 1, 2, 3 g 1. Manual Initiation NA R NA 1, 2, 3 k
a i!F
p f v
] TABLE 4.3.2.1-1 (Continued) f ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS I E CHANNEL OPERATIONAL Z CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH - TRIP FUNCTION . CHECK _ TEST CALIBRATION . SURVEILLANCE REQUIRED
! 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION l a. RCIC Steam Line Flow - High S Q R*' I, 2, 3
b. RCIC Steam Supply Pressure -
. Low S Q R*) 1, 2, 3
c. RCIC Turbine Exhaust Diaphragm Pressure - High 5 R) S 1, 2, 3 Q

d. RCIC Equipment Room Ambient Temperature - High 5 Q R 1, 2, 3

e. Deleted

f. Main Steam Line Tunnel Ambient g Temperature - High S Q R 1, 2, 3 a g. Main Steam Line Tunnel y A Temperature - High S' Q -R 1, 2, 3 m h. Main Steam Line Tunnel Temperature Timer NA Q R 1, ' 2, 3

1. RHR Equipment Room Ambient Temperature - High 5 Q R 1, 2, 3

j. RHR Equipment Room a Temperature - High -S Q R 1, 2, ' 3

k. RCIC Steam Line Flow High Timer NA (t 1, 2, 3

1. Drywell Pressure - High S F R*)
R 1, 2, 3 y m. Manual Initiation MA N NA 1, 2, 3 8o. i.'
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A TABLE 4.3.2.I-1 (Continued) 53 7 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS
-)
E CHANNEL OPERATIONAL. Z CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED
^
6. RHR SYSTEM ISOLATION

a. RHR Equipment Area Ambient :
Temperature - High S Q R 1, 2, 3
b. RHR Equipment Area A Temperature - High S Q R 1, 2, 3

c. RHR/RCIC Steam Line Flow - High S Q R*) 1, 2, 3

d. Reactor Vessel Water Level -
R* Low, Level 3 ## 5 Q R )gg) 1, 2, 3 Y e. Reactor Vessel (RHR Cut-in g Permissive) Pressure - High S Q R 3 gg) 1, 2, 3
f. Drywell Pressure - High ## S Q R 'U) 1, 2, 3

g. Manual Initiation NA R NA 1, 2, 3
When handling irradiated fuel in the primary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
** When any turbine stop valve is greater than 90% open and/or the key locked bypass switch is k in the normal position.
R *** OPERATIONAL CONDITION 1 or 2 when the mechanical vacuum pump lines are.not isolated. 2 # During CORE ALTERATION and operations with a potential for draining the reactor vessel. 5 (a) Each train or logic channel shall be tested at least every other 92 days. E e Tr c ich a&c m .
.E (e) (MMtL.CAuEb\ TIDE % b Uftad"O b N "%
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l p INSTRUMENTAT10N 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 3.3.3-2 and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Table 3.3.3-APPLICABILITY: As shown in Table 3.3.3-1. ACTION:
a. With an ECCS actuation instrumentation channel trip setpoint less 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. ;

c. With either ADS trip system "A" or "B" inoperable, restore the inoperable trip system to OPERABLE status:

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

2. Within 72 hours, provided either the HPCS or the RCIC system is inoperable.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to less than or equal to 100 psig within the following 24 hours. SURVEILLANCE REQUIREMENTS l 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for ERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.3.1 . ration of 4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated autom . all channels shall be performed at least once per 18 month .
' 3-3 4.3.3.3 The ECCS RESPONSE TIME of each ECCS trip function shown in Tab ch !
shall be demonstrated to be within the limit at least once per 18 month test shall include at least one channel per tri m such that all channels are tested at least once every N times 18 month - > N is the total number of redundant channels in a specific ECCS trip system. I weW PERRY - llNil 1 3/4 3-27 i
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TABLE 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES ECCS RESPONSE TIME (Seconds) A. DIVISION 1 TRIP SYSTEM
1. RHR-A (LPCI MODE) AND LPCS SYSTEM

a. Reactor Vessel Water Level - Low, 13 N Level 1

b. Drywell Pressure - High $ 37

c. LPCS Pump Discharge Flow - Low (Bypass) NA

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

e. Reactor Vessel Pressure - Low (LPCI Injection NA Valve Permissive)

f. LPCI Pump A Start Time Delay Relay NA

g. LPCI Pump A Discharge Flow - Low (Bypass) NA

h. Manual Initiation NA

2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"

a. Reactor Vessel Water Level - Low, Level 1 NA .

b. Manual Inhibit NA NA

c. ADS Timer

d. Reactor Vessel Water Level - Low, NA Level 3 (Permissive)
LPCS Pump Discharge Pressure - High NA e. (Permissive) NA
f. LPCI Pump A Discharge Pressure - High (Permissive) NA

g. Manual Initiation B. DIVISION 2 TRIP SYSTEM

1. ,RHR B AND C (LPCI MODE)
(*
a. Reactor Vessel Water Level - Low, $3 Level I < 37

b. Drywell Pressure - High RA

c. Reactor Vessel Pressure - Low (LPCI Injection Valve Permissive)
LPCI Pump B Start Time Delay Relay NA d. LPCI Pump Discharge Flow - Low (Bypass) NA e. NA
f. Manual Initiation 1
PERRY - UNIT 1 3/4 3-35 l
i TABLE 3.3.3-3 (Continued) EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES 1 TRIP FUNCTION RESPONSE TIME (Seconds)
2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"

a. Reactor Vessel Water Level - Low, NA Level 1 >

b. Manual Inhibit NA

c. ADS Tiedr NA

d. Reactor Vessel Water Level - Low, NA Level 3 (Permissive)

e. LPCI Pump B and C Discharge NA Pressure - High (Permissive)

f. Manual Initiation NA C. DIVISION 3 TRIP SYSTEM i

1. HPCS SYSTEN

a. Reactor Vessel Water Level - Low, <2 Level 2

b. Drywell Pressure - High < 27 -

c. Reactor Vessel Water Level - High, NA Level 8

d. Condensate Storage Tank Level - Low NA

e. Suppression Pool Water Level - High NA

f. HPCS Pump Discharge Pressure - High NA

g. HPCS System Flow Rate - Low NA

h. Manual Initiation NA D. LOSS OF POWER

1. 4.16 kv Emergency Bus Undervoltage# NA (Loss of Voltage)

2. 4.16 kv Emergency Bus Undervoltage# NA (Degraded Voltage)
The Loss of Voltage and Degraded Voltage functions are common to Division 1, Mdh %) % Leededed & %e LesupleNon ch Or NS Vehelof olt$ P RY - UNIT 1 3/4 3-36 :
A TABLE 4.3.3.1-1 E EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 7 E CHANNEL OPERATIONAL Q CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH
- TRIP FUNCTION CHECK TEST CALIBRATION ,
SURVEILLANCE REQUIRED A. DIVISION 1 TRIP SYSTEM
1. RHR-A flPCI MODE) AND LPCS SYSTEM

a. Reactor Vessel Water Level - g, Low, Level 1 S Q R(*)(g))(c) 1, 2, 3, 4*, 5*

b. Drywell Pressure - High S Q 1, 2, 3

c. LPCS Pump Discharge Flow - Low (Bypass) S Q R(*) 1, 2, 3, 4*, 5*

d. Reactor Vessel Pressure - Low S Q R(*) 1, 2, 3, 4*, 5*
(LPCS Injection Valve Permissive) m e. Reactor Vessel Pressure - Low S Q R(') 1, 2, 3, 4*, 5* 2 (LPCI Injection Valve Permissive) m f. LPCI Pump A Start Time Delay
& Relay NA Q 1, 2, 3, 4*, 5*
u g. LPCI Pump A Flow - Low (Bypass) S Q(*) R 1, 2, 3, 4*, 5* Q
h. Manual Initiation NA R NA 1, 2, 3, 4*, 5*

2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#

a. Reactor Vessel Water Level -
Low, level 1 S Q R(*) I,2,3
b. Manual Inhibit NA Q 1, 2, 3 y c. ADS Timer NA Q Qr 1, 2, 3 g d. Reactor Vessel Water Level - (*) g )
p Low, level 3'(Permissive) S Q 1, 2, 3 ; a e. LPCS Pump Discharge 5 Pressure - High (Permissive) S Q R(*) 1, 2, 3
= f. LPCI Pump A Discharge ? Pressure - High (Permissive) S q R(*) 1, 2, 3 a g. Manual Initiation NA RM NA 1, 2, 3
T gj TABLE 4.3.3.1-1 (Continued)
o
~'
i i
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS ~
l 55 CHANNEL
" OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION . SURVEILLANCE REOUIRED

8. DIVISION 2 TRIP SYSTEM i 1. RHR B AND C (LPCI MODE)

a. Reactor Vessel Water Level -
Low, Level 1 S Q If*)(gge) 1, 2, 3, 4*, 5*
b. Drywell Pressure - High S Q i t) (') 1, 2, 3

c. Reactor Vessel Pressure - Low (LPCI Injection Valve Permissive) S Q RD') 1, 2, 3, 4*, 5*

d. LPCI Pump B Start Time Delay Relay NA Q Q 1, 2, 3, 4*, 5* -
us e. LPCS Pump Discharge Flow - Low 1 (Bypass) S Q R 1, 2, 3, 4*, 5*
f. Manual Initiation NA R NA 2, 2, 3, 4*, 5*

2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#

a. Reactor Vessel Water Level -
Low, level 1 S Q R)(b)0') 1, 2, 3
b. Manual Inhibit NA Q 1 , 12 , 3

c. ADS Timer NA Q 1, 2, 3

d. Reactor Vessel Water Level - /'Q y Low, Level 3 (Permissive) S Q R"') Ch0C') 1, 2, 3 g e. LPCI Pump B and C Discharge c Pressure - High (Permissive) S R"') 1, 2, 3 l'r f. Manual Initiation NA ') NA 1, 2, 3
- - - - , - - _ _ - _ _ _ _ - - , - . _ - . - _ _ - - _ - - - w -- -- -_v - -
at- ---
o TABLE 4.3.3.1-1 (Continued) 7 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E CHANNEL U OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION $URVEILLANCE REQUIRED C. DIVISION 3 TRIP SYSTEM

l. HPCS SYSTEM

a. Reactor Vessel Water level -
Low, Level 2 S Q R(*) D X 0 1, 2, 3, 4*, 5*
b. Drywell Pressure-High ## S Q R(') b) -1, 2, 3

c. Reactor Vessel Water Level -
High, Level 8 S Q (* N ') 1, 2, 3, 4*, 5*
d. Condensate Storage Tank Level -
Low
e. Suppression Pool Water S Q R(*) 1, 2, 3, 4*, 5*
w i Level - High S Q R(*) 1, 2, 3, 4*, 5* w .f . HPCS Pump Discharge Pressure - 4, High S R(*) 1, 2, 3, 4*, 5*
g. HPCS System Flow Rate - Low h.
S M R(*) 1, 2, 3, 4*, 5* Manual Initiation ## NA NA 1, 2, 3, 4*, 5* D. LOSS OF POWER
1. 4.16 kV Emergency Bus Under- NA NA R 1, 2, 3, 4**, 5**
voltage (Loss of Voltage) .
2. 4.16 kv Emergency Bus Under- S M R 1, 2, 3, 4**, 5**
voltage (Degraded Voltage) F g # Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig. p ## The injection function of Drywell Pressure - High and Manual Initiation are not required to be g OPERABLE with indicated reactor vessel water level on the wide range instrument greater than the Level 8 setpoint coincident with reactor pressure less than 450 sig. g
When the system is re ired to be OPERAB r-Specifica o .5.3.
** equired when ESF pment equ to be OPERA . ' w- s
O a) Calibrate tri t setpoint at i ast once per 92 days. G Acaet C.Ausrchoc MM bt_ dt~ M to W wPdou d M N b 4 l gyhp ,
,sss s s - + - ~ A~ ~ - ^
i l i' y l INSTRUMENTATION 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION LINITING CONDITION FOR OPERATION I i, 3.3.5 The reactor core isolation cooling (RCIC) system actuation instrumenta-tion channels shoini'in Table 3.3.5-1 shall be OPERA 8LE with their trip set-points set consistent with'the values shown in the Trip Setpoint column of Table 3.3.5-2. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3 with reactor steam done pressure greater than 150 psig. ACTION: ,
a. With a RCIC system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.5-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted corsistent with the Trip Setpoint value.

b. With one or more RCIC system actuation instrumentation channels i inoperable, take the ACTION required by Table 3.3.5-1.
SURVEILLANCE REQUIREMENTS : 4.3.5.1 Each RCIC system actuation instrumentation channel shal'; be demon-strated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL i TEST and NEL CALIBRATION operations at the frequencies shown in Table :
4. 3. 5.1-1 # l 4.3.5.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automa peration of all channels shall be performed at least once per 18 months # l i.
*Nvl Cahhdh pna q be w,a a iacsb by a %'n, . T a le 4 . ~6 . 5 .1 - t .
eloGic Syna fupaiopat7m psa q L, Aa,Jas iA,4$a by 4 'c% Tau < 03.s.1- i . l l PERRY - UNIT 1 3/4 3-50
A TABLE 4.3.5.1-1 E REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 7 E CHANNEL Z CHANNEL FUNCTIONAL CHANNEL
- FUNCTIONAL UNITS CHECK _ TEST CALIBRATION
a. Reactor Vessel Water Level -
low, level 2 S Q
N 's 'T
b. Reactor Vessel Water level - S Q R DXD High, level 8

c. Condensate Storage Tank level -
Low S Q R(
d. Suppression Pool Water Level -
High S Q R(*)
e. Manual Initiation NA R NA Y
i C3 g 1")pibratest@ibsetpoint ,at-1 eIst'oney-924ays s v7~~u f uuMn A uta ntowd.gfhey dt n or a a u (9 Ca w au tAteawnoa en d *-a =~g&k (c) tout systm sacamntr ~4 u chm +o k R , EPrk rJrodicg ooky , B
!!F l
INSTRUMENTATION l i l 3 /4. 3. 6 CONTROL ROD BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6. The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2. APPLICABILITY: As shown in Table 3.3.6-I. ACTION:
a. With a control rod block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, 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 the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, take the ACTION required by Table 3.3.6-1.
( SURVEILLANCE REQUIREMENTS 4.3.6.1 Each of the above required control rod block trip systems.and instrumentation channels shall be demonstrated OPERABLE by the performance the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operatio 8 for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.61. 4.3.6.2 The provisions of Specification 4.0.4 are not applicable to the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION surveillances for the Intermed-iate Range Monitors and Source Range Monitors for entry into their applicable OPERATIONAL CONDITIONS (as shown in Table 4.3.6-1) from OPERATIONAL CONDITION 1 provided the surveillances are performed within 12 hours after such entry. l
*The APRM flow biased instrumentation need not be declared inoperable upon entering single recirculation loop operation provided the setpoints are adjuste itpounqqc.4pecif4catign 3:4;l g M mNrtod r"ti k<<fh'Ad bi KON ' 0 '4 k O*W !
4 '"" '3 " g pm e wtTEsr pviod mrq Mtt h* " not t I o^ T , PER - NI 3/4 3-55 Amendment No. #1,61,66
.~
N y TABLE 4.3.6-1 El Q CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REOUIREMENTS h CHANNEL OPERATIONAL 5
CHANNEL FUNCTIONAL CHANNEL TRIP FUNCTION CONDITIONS IN WHICH CHECK TEST CALIBRATION"' SURVEILLANCE REQUIRED I. ROD PATTERN CONTROL SYSTEM
a. Low Power Setpoint

b. RWL - High Power Setpoint NA S/Utb ) Q SAf 1, 2 NA S/U(b',,Q SAf 1

2. APBd

a. Flow Biased Neutron Flux - Upscale Flow Biased S w

1) NA S/U ) SA( 1

2) High Flow Clamped SAR)

b. Inoperative NA NA S/U*S
S/U ', Q Q Q NA I 1' 2' 5
c. Downscale NA S/U*) SA l d.
t Neutron Flux - Upscale, Startup NA S/U"',' Q Q SA 2, 5 [ 3. SOURCE RANGE MONITORS h a. Detector not full in NA S S/U ',W NA 2*** 5
b. Upscale NA S/U*),W(d'

c. Inoperative R 2*** 5 NA S/US),W NA 2** 5

d. Downscale NA S/U(b',W(d' R N5

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in NA S/U*),W NA 2, 5

b. Upscale NA S/U*),W(d' R 2, 5 P c. Inoperative NA S/U*',W NA 2, 5
[ d. Downscale Nr. S/U b),ytd> R 2, 5
5. SCRAM DISCHARGE VOLUME

a. Water Level - High NA Q R# 1, 2, 5* l

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW (4 a. Upscale NA /U Q SA 1 l

7. REACTOR MODE SWITCH SHUTDOWN POSITION NA NA 3, 4
l 1 l i e TABLE 4.3.6-1 (Continued) . CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE RE0.UIREMENTS NOTES:
a. Neutron detectors may be excluded from CHANNEL CALIBRATION.

b. Within 7 d,ays prior to startup.

c. The CHANNEL CALIBRATION shall exclude the flow reference transmitt these transmitters shall-be4alibrated erJ 8-mont s ext (e & NY
% t<e ma.y w catue A u vu c~etm_at-least_once_[W e- of m s <^de utwg,e mg p-setp6fiitrard3 rtfTeDfIr g_ 1 * #- '
d ALT j e Aog h ubh4 to kk c=eithco o E Ek (4r
^
Mb eg och (< , v- - m .-
With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
** With IRMs on range 2 or below. # Calibrate trip unit setpoint at least once per 92 days. l ;
e s PERRY - UNIT 1 3/4 3-60 Amendment No.-M, 67
INSTRUMENTATION REMOTE SHUTDOWN ',YSTEM INSTRUMENTATION AND CONTROLS LIMITING CONDITION FOR OPERATION 3.3.7.4 The remote shutdown system instrumentation and controls shown in Table 3.3.7.4-1 shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS I and 2. ACTION:
a. With the number of OPERABLE remote shutdown system instrumentation channels less than required by Table 3.3.7.4-1, restore the inoperable channel (s) to OPERABLE status withic. 7 days or be in at least HOT SHUTDOWN within the next 12 hours.

b. With the number of OPERABLE remote shutdown system controls less than required in Table 3.3.7.4-1, restore the inoperable control (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within '
the next 12 hours. - e c. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.3.7.4.1 Each of the above required remote shutdown system instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.4-1.* 4.3.7.4.2 Each of the above remote shutdown controls shall be demonstrated OPERABLE by verifying its espability to perform its intended function (s) at least once per 18 months, e - e tw,a camu m ,ou,a t a wha y a.te 'a' a Tak 4 2.1.4-1. u PERRY - UNIT 1 3/4 3-73
r ; 4 TABLE 4.3.7.4-1 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBR&TJ,0
1. Reactor Vessel Pressure M

2. Reactor Vessel Water Level M

3. Safety / Relief Yalve Position M NA

4. Suppression Pool Water Level M R

5. Suppression Pool Water Temperature M R

6. Drywell Pressure M R

7. Drywell Temperature M R

8. RHR System Flow . M R

9. Emergency Service Water Flow to RHR M R '
Heat Exchanger
10. Emergency Service Water Flow to Emergency M R Closed Cooling Heat Exchanger

11. RCIC System Flow M R

12. RCIC Turbine Speed M R

13. Emergency Closed Cooling System Flow' M R

14. Inboard MSIV Position M NA i
tycu wet. canuATw up, osaa w p,,b,a sa,& Aptt, y,L,;;4,. t9o r axv<agy t, ama,a i, a 4 ,a a#3 u,,Ma r,p.,,3 am. PERRY - UNIT 1 3/4 3-76
INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.5 The accident monitoring instrumentation channels shown in Table 3.3.7.5-1shall,peOPERABLE. APPLICABILITY: As shown in Table 3.3.7.5-1. ACTION: with one or more accident monitoring instrumentation channels inoperable, take the AC/iGN required by Table 3.3.7.5-1. SURVEILIANCE REQUIREMENTS 4.3.7.5 Each of the above required accident monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL. CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.5-( e e tl Clie%el CaldraWew pededmq l e e>{eged a gjfggg 4 g. Tak 4.31.5 -l . PERRY - UNIT 1 3/4 3-77
A g TABLE 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE0UIREMENTS E APPLICABLE 5 CHANNEL CHANNEL OPERATIONAL m INSTRUMENT CHECK CALIBRATION CONDITIONS
1. Reactor Vessel Pressure M '~
i 1, 2, 3
2. Reactor Vessel Water Level

a. Fuel Zone M R @) 1, 2, 3

b. Wide Range M N 1, 2, 3

3. Suppression Pool Water Level M R- 1, 2, 3

4. Suppression Pool Water Temperature M R 1, 2, 3

5. Primary Containment Pressure M R 1,2,3

6. Primary Containment Air Temperature M R 1, 2, 3

7. Drywell Pressure M R 1,2,3

8. Drywell Air Temperature M R 1, 2, 3

9. Primary Containment and Drywell Hydrogen Concentration Analyzer and Monitor NA Q* 1, 2, 3 t' 10. Safety / Relief Valve Position Indicators M R_ 1, 2, 3

11. Primary Containment /Drywell Area Gross Y Gamma Radiation Monitors M *N 1, 2, 3 8 12. Offgas Ventilation Exhaust Monitor # M R 1, 2, 3

13. Turbine Building / Heater Bay Ventilation Exhaust Monitor # M R 1, 2, 3

14. Unit 1 Vent Monitor # M R 1, 2, 3

15. Unit 2 Vent Monitor # M i< 1, 2, 3

16. Neutron Flux
> a. Average Power Range M R 1,2,3 8 b. Intermediate Range M R 1, 2, 3 E. c. Source Range M R 1, 2, 3, s 17. Primary Containment Isolation Valve M R 1, 2, 3 E Position U
8 *Using sample gas containing:
a. One volume percent hydrogen, balance nitrogen,

b. Four volume percent hydrogen, balance nitrogen.
**The CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a one point calibration check of the detector be wJ0-4/hr with an installed'or portable gamma source. .
igh and iritermE81 Tie range D19 systhnobh,gs -niC
@ Chet true<*om My - h e *dde_ b ' cr4** #
40% h W f M*d_'__N N _ . - ~-
1
INSTRUMENTATION 3/4.3.9 PLANT SYSTEMS ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.9 The plant systems actuation instrumentation channels shown in : Table 3.3.9-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.9-2. APPLICABILITY: As shown in Table 3.3.9-1. ACTION:
a. With a plant system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.9-2, 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 plant systems actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.9-1.
SURVEILLANCE RE0VIREMENTS 4.3.9.1 Each plant system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the O TIONAL CONDITIONS and at the frequencies shown in Table 4.3,9.I 1.4 4.3.9.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated autom tric operation of all channels shall be performed at least once per 18 month .4 m
,/ ,/ x ~ ~~~ ~~ } ..
A(,w m (Au6(Chou Etr '*d M'N h"\'A g e d c, ' c, a '[ A 6 t t- 4.3 i . i - t h> L t e:. 4 & d' bi 6c 9 O n- (04tt bys h FootheaAt TL W Qusod
\
ge s( ' b o e T A 9 L. 4 3.f i, \ - \ - s u / s 3/4 3-98 Amendment No. -30, 67 PERRY - UNIT 1
4 s
,s TABLE 4.3.9.1-1 9
PLANT SYSTEMS ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS g CHANNEL OPERATIONAL
~
CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH [ TRIP FUNCTION CHECK-- TEST CALIBRATION SURVEILLANCE REQUIRED
1. CONTAINMENT SPRAY SYSTEM
, a. Drywell Pressure - High S Q R 1, 2, 3
b. Containment Pressure - High S Q R* 1,2,3

c. Reactor Vessel Water Level - '
d. Low, Level 1 Timers S Q *W . 1, 2, 3 (1) Subsystem A and B NA Q R 1, 2, 3 (2) Subsystem B NA Q R 1, 2, 3
e. Manual Initiation NA R NA 1, 2, 3-

2. FEEDWATER SYSTEM / MAIN TURBINE TRIP SYSTEM
[ a. Reactor Vessel Water Level - High,
1. Level 8 S Q R* g ( g 1 l

3. SUPPRESSION POOL MAKEUP SYSTEM

a. Drywell Pressure - High S Q 1, 2, 3

b. Reactor Vessel Water Level -
Low, level 1 S Q *(cS ( b) 1, 2, 3
c. Suppression Pool Water Level - Low S Q R* 1, 2, 3

d. Suppression Pool Makeup Timer NA Q Q 1, 2, 3 g e. SPMU Manual Initiation NA R NA 1, 2, 3 o
k
Calibrate trip unit setpoint at least once r 921
~ ,' l g) Che' CWord'* *" b ""E" ( QFk h" ' '
a @dqoA@ ,
g d p (-datocc At TM MW bt eMM k' F % esa rM g p. 7 '
MACTOR COOLANT SYSTEM 3/4.4.? SAFETY VALVES l SAFETY /RELIFF VALVES LIMITING CONDITION FOR OPERATION 1 3.4.2.1 Of the following safety / relief valves, the safety valve function of at least 7 valves and the relief valve function of at least 6 valves other than those satisfying the safety valve function requirement shall be OPERABLE with the specified lift settings: , Number of Valves Function Setooint* fosial i 8 Safety 1165 i 11.6 psi 6- Safety 1180 i 11.8 psi ; 5 Safety 1190 i 11.9 psi ; I Relief 1103 i 15 psi i 9 Relief 1113 15 psi ' 9 Relief ~ 1123 i 15 psi APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. . ACTION: i
a. With the safety and/or relief valve function of one or more of the above required safety / relief valves inoperable, be in at least HOT SHUTDOWN i within 12 hours and in COLD SHUTDOWN within the next 24 hours.

b. With one or more safety / relief valves stuck open, close the stuck open i safety / relief valve (s); with suppression pool average water temperature i 110*F or greater, place the reactor mode switch in the Shutdown position.

c. With one or more safety / relief valve tail-pipe pressure switches inoperable, !
restore the inoperable switch (es) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within : the following 24 hours. [
d. With either relief valve function pressure actuation trip system "A" or "B" !
inoperable, restore the inoperable trip system to OPERABLE status within i 7 days; otherwise, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.4.2.1.1 The tail-pipe pressure switch for each safety / relief valve shall be l demonstrated OPERABLE with the setpoint verified to be 30 -5 psig by performance ; of a: !
a. CHANNEL FUNCTIONAL TEST at least once per 92 days, and a l l

b. CHANNEL CALIBRATION at least once per 18 months.
]
4.4.2.1.2 The relief valve function pressure actuation instrumentation shall be l demonstrated OPERABLE # by performance of a: .. l
a. CHANNEL FUNCTIONAL TEST, including calibration of *the trip unit, at least once per 92 days. l CHANNEL CALIBRATION, LOGIC SYSTEM FUNCTIONAL TEST ** and simulated b.
automatic operation of the entire system at least once per 18 month
The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.
** SRV solenoid energization shall be used alternating between the "A" solenoid and L the "B" solenoid.
l
# When a channeLis placed in an inoperable status solely for performance of l re4ifired'Surveilling.cn ry into associated ACTIONS may begyed-for up tohility.
3 s4telief 6%hours provided kuhW hkp6<uddures the associated-TUnction W- 6 FA rdvdi initiation maintai o u tn%. cap / :
@T-MINI-T1Wf -7 6 '67
f REACTOR COOLANT SYSTEM SAFETY / RELIEF VALVES LOW-LOW SET FUNCTION LIMITING CONDITION FOR OPERATION 3.4.2.2 . The relief valve function and the low-low set function of the following reactor coolant system safety / relief valves shall be OPERABLE with the following settings: Low-Low Set Function Relief Function Setooint* (osia) i 15 osi Setooint* (osia) Valve No. - Qpin Close 9Agn Elgig IB21-F051D 1033 926 1103 15 psi 1003 20 psi IB21-F051C 1073 936 1113 15 psi 1013 20 psi , 1821-F051A 1113 946 1113 15 psi 1013 20 psi 1821-F051B 1113 946 1113 15 psi 1013 20 psi 1821-F047F 1113 946 1113 15 psi 1013 20 psi IB21-F051G 1113 946 1113 15 psi 1013 20 psi APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. ACTION:
a. With the relief valve function and/or the low-low set function of one of the above required reactor coolant system safety / relief valves inoperable, restore ,
the inoperable relief valve function and the low-low set function to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
b. With the relief valve function and/or the low-low set function of more than one of the above required reactor coolant system safety / relief valves inoperable, be in at least H0T SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.

c. With either relief valve / low-low set function pressure actuation trip system "A" or "B" inoperable, restore the inoperable trip system to OPERABLE status within 7 days; otherwise, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours. !
SURVEILLANCE RE0UIREMENTS 4.4.2.2.1 The relief valve function and the low-low set function pressure actuation I instrumentation shall be demonstrated OPERABLE # by performance of a: l
a. CHANNEL FUNCTIONAL TEST, including calibration of the trip unit, at least once per 92 days. l

b. CHANNEL CALIBRATION, LOGIC SYSTEM FUNCTIONAL TEST and simul automatic !
operation of the entire system at least once per 18 months. !
The lift setting pressure shall correspond to ambient cor :itions of the valves at ,
nominal operating temperatures and pressures.
# 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 hours ,yevidedAsoci aM Ta G urmatnwn- Low .Lqw_Se initiation capability.
M k e&sul<4 k L yshed 3b Wp@L nhe(p~ ty.
-A PERRY - UNIT 1 3/4 4-8 Amendment No. 67
REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.3.1 The fol]Qwing reactor coolant system leakage detection systems shall be OPERABLE:
a. The drywell floor drain sump and equipment drain sump flow monitoring system.

b. Any 2 of the following:

1. Drywell atmosphere particulate radioactivity monitoring system.

2. Drywell atmosphere gaseous radioactivity monitoring system.

3. Upper drywell air coolers condensate flow rate monitoring system.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. ACTION:
a. With the drywell floor drain sump or equipment drain sump flow monitor-ing system inoperable, operation may continue for up to 30 days provided ;
that the upper drywell coolers condensate flow rate monitoring system is one of the two systems OPERABLE per 3.4.3.1.b.
b. With only one of the systems required by 3.4.3.1.b OPERABLE operations may continue for up to 30 days provided:

1. The drywell floor drain sump and equipment drain sump flow monitoring system is OPERABLE.

2. Grab samples of the drywell atmosphere are obtained and analyzed at least once per 24 hours. 1

c. Otherwise be it. at least HOT SHtfTDOWN within the next 12 hours and in I COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated OPERABLE by: I
a. Drywell atmosphere particulate and gaseous monitoring systems-performance of a CHANNEL CHECK at least once per 12 hours, a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.

b. Drywell floor drain and equipment drain sump flow monitoring system-performance of a CHANNEL FUNCTIONAL TEST at least onc 31 days and a CHANNEL CALIBRATION at least once per 18 months.

c. Upper drywell air coolers condensate flow rate monitoring system-performance of a CHANNEL FUNCTIONAL TEST at least onc >r 31 days
-.CUANNEL-GL4BginN MDst onc er nths.
(tr (Au,b ofruj,j k be pniormed dunn4 % 4l(E!.ydai,,c, ou(W - J/u 44
a REACTOR C00LAtiT SYSTEM SURVEILLANCE REQUIREMENTS f 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:
a. Monitoring the drywell atmospheric particulate or gaseous radioactiv-ity at least once per 12 hours (not a means of quantifying leakage), '

b. Monitoring the drywell floor and equipment sump flow rate at least once per 12 hours,

c. Monitoring the drywell upper drywell air coolers condensate flow rate at least once per 12 hours, and

d. Monitoring the reactor vessel head flange leak detection system at least once per 24 hours.
4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5 and verifying the leakage of each valve to be within the , specified limit:
a. At least once per 18 months,**

b. Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate, and The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3.
i
**P.I.V. leak test extension to the first refueling outage is permissible for ~
each Reactor Coolant System P.I.V. listed in Table 3.4.3.2-1, which are identified in letter PY-CEI/NRR-0714L (dated September 11, 1987) as needing a plant outage to test. For this one time test interval, the provisions of Specification 4.0.2 are not applicable.
~
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~
r. '
' EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS i
l 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:
l
1. Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valv,e,is filled with water. 1

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

2. Performing a CHANNEL CALIBRATION of the ECCS discharge line
" keep filled" pressure alarm instrumentation.
d. For the HPCS system, at least once per 18 months, verifying that the suction is automatically transferred from the condensate storage tank to the suppression pool on a condensate storage tank low water level signal and on a suppression pool high water level signal.
l l
*Except that an automatic valve capable of automatic return to its ECCS position de of operation.
a#.m a#ps -% PERRY - UNIT 1 3/4 5-4
' EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
e. for the ADS by:

1. At least once per 31 days, performing a CHANNEL FUNCTIONAL TEST of the safety related instrument air system low pressure alarm system.

2. At least once per 18 months:
a) ' Performing a system functional test which includes simulated automatic actuation of the system throughout its emergen operating sequence, but excluding actual valve actuation b) Manually ** opening each ADS valve when the reactor steam dome pressure is greater than or equal to 100 psig* and observing that either:
1) The control valve or bypass valve position responds accordingly, or

2) There is a corresponding change in the measured steam flow, or

3) The safety relief valve discharge pressure switch indicates the valve is open.
c) Performing a CHANNEL CALIBRATION of the safety related j instrument air system low pressure alarm system and verifying an alarm setpoint of > 155 psig with an allowable value of > 151.9 psig on decreasing pressure.
?
uThe provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test. Q*A05 solenoid energization shall be used alternating between ADS Division 1 and ADS Division 2.
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v PERRY - UNIT 1 3/4 5-5 Amendment No. 6 l t
CONTAlf4 MENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.4.1 Each containment isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replace-ment work is perfonned on the valve or its associated actuator, control or power circuit.by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time. 4.6.4.2 Each automatic containmen lation valve shall be demonstrated l OPERABLE at least once per 18 mont verifying that on an isolation test signal each automatic isolation valve actuates to its isolation position. 4.6.4.3 The isolation time of each power operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5. L extenJ<d do Oc (0mp(chen 0- N O Ve' eI'n3 odhy . \ tt w' PERRY - UNIT 1 3/4 6-29 Amendment No. 44 (Next page is 3/4 6-40)
l
)
i
. CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the subsystem by:
1. '
Verifying that the subsystem satisfies the in place penetration
t'esting acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.S.a, C.S.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, while operating the system at a flow rate of 2000 scfm i 10%.
2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, by showing a methyl iodide penetration of less than 0.175% when tested at a temperature of 30*C and at a relative humidity of 70% in accordance with ASTM D3803; and

3. Verifying a subsystem flow rate of 2000 scfm i 10% during system operation when tested in accordance with ANSI N510-1980. .The installed air flow monitor can be used to determine flow in lieu of the pitot traverse.

c. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a :
of Regulatory Guide 1.52, Revision 2, March 1978, by showing a methyl iodide penetration of less than 0.175% when tested at a temperature of 30 C and at a relative humidity of 70% in accordance with , ASTM 03803;
d. At least once per 18 months by:

1. Performing a system functional test which includes simulated automatic actuation of the syst roughout its emergency operating sequence for the LOCA *

2. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.0 inches water gauge while operating the filter train at a flow rate of 2000 scfm '
i 10%.
3. Verifying that the filter train starts and isolation dampers open on each of the following test signals:

a. Manual initiation from the control r and i

b. Simulated automatic initation signal w

4. Verifying that the heaters dissipate 20 kw i 10% when tested NSIO-l h M se ac,,u 4 4k u.+u~ q s M 4c% ck .
I
> ERR -t 3/ -4 i
I 3/4.7 PLANT SYSTEM 5 3/4.7.1 COOLING WATER SYSTEMS EMERGENCY SERVICE WATER SYSTEM (LOOPS A, B, C) LIMITING CONDITION FOR OPERATION
\
3.7.1.1 The emergency service water (ESW) loop (s) shall be OPERABLE which is associated with systems or components which are required to be OPERABLE. Each OPERABLE ESW loop'shall be comprised of:
a. One OPERABLE ESW pump, and

b. An OPERABLE flow path capable of taking suction from Lake Erie and transferring water through the associated systems and components '
heat exchanger (s) that are required to be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5 and *. ACTION: With an emergency service water loop (s) inoperable which is as,sociated with system (s) or component (s) required to be OPERABLE, declare the associated system (s) or component (s) inoperable and take the ACTION required by the appifcable Specification (s). SURVEILLANCE REQUIREMENTS 4.7.1.1 The above required emergency service water system loops shall be demonstrated OPERABLE:
a. At least once per 31 days by verifying that each valve in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

b. At least once per 18 month uring shutdown by verifying that each automatic valve servicing safety related equipment actuates to the correct position on a LOCA test signal.
*When handling irradiated fuel in the Fuel Handling Building or primary containment.
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r, - 3/4.7 PLANT SYSTF.MS EMERGENCY CLOSED COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 The emergency closed cooling (ECC) loop (s) shall be OPERABLE which is associated with systems or components which are required to be OPERABLE. Each OPERABLE ECC loop shall be comprised of:
a. One OPERABLE ECC pump, and

b. An OPERABLE flow path capable of transferring water through the associated systems and components heat exchanger (s) that are required to be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5. ACTION: With an emergency closed cooling loop (s) inoperable which is associated with system (s) or component (s) required to be OPERABLE, declare the associated system (s) or component (s) inoperable and take the ACTION required by the applic'able Specification (s). SURVEILLANCE REQUIREMENTS 4.7.1.2 The above required emergency closed cooling loop (s) shall be demonstrated OPERABLE:
a. At least once per 31 days by verifying that each valve in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

b. At least once per 18 mont uring shutdown by verifying that each automatic valve servicing safety related equipment actuates to the correct position on a LOCA test signal.
d * *
PERRY - UNIT 1 3/4 7-2
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communi-cating with the subsystem by:

1. Verifyfng that the subsystem satisfies the in place penetration !
testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.S.a, C.S.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 30000 scfm i 10L
2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978 by showing a methyl iodide penetration of less than 1% when tested at a temperature of 30*C and at a relative humidity of 70% in accordance with ASTM D3803; and

3. Verifying a subsystem flow rate of 30000 scfm i 10% during sub-system operation when tested in accordance with ANSI N510-1980.
The installed air flow monitor can be used to determine flow in lieu of a pitot traverse,
d. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Posi-ton C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, by showing a methyl iodide penetration of less than 1% when tested at a temperature of 30 C and at a relative humidity of 70% in accordance with ASTM D3803.

e. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 4.9 inches water gauge while operating the subsystem at a flow rate of 30000 scfm i 10L

2. Verifying that on each of the below emergency recirculation mode actuation test signals, the subsystem automatically switches to the emergency recirculation mode of operation and the isolation dampers close within 10 se ds:
a) High Drywell Pressur M i b) Low Reactor Water Level-Level 1@(] { c) High radiation from control room ventilation duct j m - m -- m ._ T3 lcNow c[ Sc. h 6 c.hel% o eda 3e , PERRY - UNIT 1 M
f PLANT SYSTEMS 3/4.7.4 SNUBBERS LINITING CONDITION FOR OPERATION 3.7.4 All snubbers shall be OPERABLE. l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. OPERATIONAL CONDITIONS 4 ! and 5 for snubbers located on systems required OPERABLE in those OPERATIONAL I CONDITIONS. .. . L ACTION: With one or more snubbers inoperable, within 72 hours replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.4.g on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that_ system. SURVEILLANCE REOUIREMENTS , 4.7.4 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of , Specification 4.0.5.
a. Inspection Tvoes l As used in this specification, type of snubber shall mean snubbers of the same design and manufacturer, irrespective of capacity.

b. Visual Inspections A visual inspection of all snubbers shall be performed according to +
the schedule determined by Table 4.7.4.1. Snubbers are categorized as inaccessible or accessible during reactor o Each of these categories (inaccessible and accessible)peration. may be inspected independently. The visual inspection for each type of snubber shall be determined based on the criteria provided in Table 4.7.4-1 and the initial inspection interval utilizing this criteria shall be 18-months, beginning from the conclusion of the last visual inspection conducted during RF04.g. 3
? .. t
[ Y M ha g6L,4 bo W 6 4(thou d N P dh f- og '% - [
- l PERRY - UNIT 1 3/4 7-8 Amendment No. 68 i
PLANT SYSTEMS SURVEllLANCE REQUIREMENTS (Continued)
e. Functional Tests During the first refueli shutdown and at least once per 18 months thereafter during shutd wt& ajiepresentative sample of snubbers shall be tested using one of t f6110 wing sample plans for each type of snubber. The sample plan shall be selected prior to the test period and cannot be changed during the test period. The Nuclear. Regulatory Commission shall be notified in writing pursuant to 10 CFR 50.4 of the sample plan selected prior to the test period or the sample plan used
! in the prior test period shall be implemented: 1) At least 10% of the total of each type of snubber shall be functionally tested either in-place or in a bench test. For each snubber of a type that does not meet the functional test acceptance criteria of Specification 4.7.4.f., an additional 5% of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested; or
2) A representative sample of each type of snubber shall be functionally tested in accordance with Figure 4.7.4-I. "C" is the total number of snubbers of a type found not meeting the acceptance requirements of Specification 4.7.4.f. The cumulative number of snubbers of a type tested is denoted by "N". At the end of each day's testing, the new values of "N" and "C" (previous day's total plus current day's increments) shall be plotted on Figure 4.7.4-1. If at any time the point plotted falls on or above the " Reject" line all snubbers of that type shall be functionally tested. If at any time the point plotted falls on or below the
" Accept" line, testing of snubbers of that type may be terminated. .
When the point plotted lies in the " Continue Testing" region, additional snubbers of that type shall be tested until the point falls in the " Accept" region or the " Reject" region, or all the snubbers of that type have been tested._ Testing equipment failure during functional testing may invalidate that day's testing and allow that day's testing to resume anew at a later time, providing all snubbers tested with the failed equipment during the day of equipment failure are retested; or
3) An initial representative sample of 55 snubbers of each type shall be functionally tested. For each snubber type which does not meet the functional test acceptance criteria, another sample of at least one-half the size of the initial sample shall be tested until. the total number tested is equal to the initial sample size multiplied by the factor,1 + C/2, where "C" is the number of snubbers found which do not meet the functional test acceptance criteria. The results from this sample plan shall be plotted using an " Accept" line which follows the equation N - 55(1 + C/2). Each snubber .
point should be plotted as soon as the snubber is tested. If the point plotted falls on or below the " Accept" line, testing of that type of snubber may be terminated. If the point plotted fails above the " Accept" line, testing must continue until the point l f alls on orybe_ low-the~Accep_t"Jne77Wbtpnubbers of that l T
' type have been tested.
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- % mendment A No.66 ,n PERRYcjJRll-V^ -- - 3/4'7 %%
1 l ELECTRICAL POWER SYSTEMS r SURVEILLANCE REQUIREMENTS 4.8.1.1.'1 Each of the above required independent circuits between the offsite transmission network and the onsite Class IE distribution system shall be:
a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments and indicated power availabil , and Demonstrated OPERABLE at least once per 18 mont uring shutdown b.
by transferring unit power supply from the normal circuit to the alternate circuit. 4.8.1.1.2 Each of the above required diesel generators shall be demonstrated OPERABLE:
a. In accordance with the frequency specified in Table 4.8.1.1.2-1 on a STAGGERED TEST BASIS by:

1. Verifying the fuel level in the day tank.

2. Verifying the fuel level in the fuel storage tank.

3. Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank.

4. Verifying the diesel starts from ambient conditions and acceler-ates to at least 441 rpm for Div 1 and Div 2 and 882 rpm for Div 3 in less than or equal to 10 seconds *. The generator voltage and frequency shall be 4160 1 420 volts and 60 1 1.2 Hz within 10 seconds
after the start signal for Div 1 and Div 2 and 13 seconds

af ter the start signal for Div 3.
5. Verifying the diesel generator is synchronized, loaded to between 5600 and 5800 kw** for diesel generators Div 1 and Div 2 and loaded to greater than or equal to 2600 kw for diesel generator Div 3 in less than or equal to 60 seconds *, and operates with this load for at least 60 minutes.

6. Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.
*All diesel generator starts for the purpose of this Surveillance Requirement may be preceded by an engine prelube period.
The diesel generator start (10 sec)/ load (60 sec) from ambient conditions shallAll be performed other engine at least starts for once per 184 days in these surveillance tests. the purpose of this surveillance testing may be preceded by other warmup pro- 1 cedures recommended by the manufacturer so that the mechanical stress and wear on the diesel engine is minimized.
"This band is meant as guidance to avoid routine overloading of the engine.
Loads in excess of this band shall not invalidate the test; the loads, however, ot%e' TPs'rttww&600- kw w>
-,% g ow .: 0 63 h. edadeJ b Se ce >
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) _
7. Verifying the pressure in all air start receivers for each diesel generator to be greater than or equal to 210 psig.

b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to 1 hour by checking' for and removing accumulated water from the day tank.

c. At least,once per 92 days by checking for and removing accumulated water from thh fuel oil storage tanks.

d. By sampling new fuel oil in accordance with ASTM D4057-88 prior to the addition to the storage tank and:

l. By verifying prior to addition to the storage tanks that the sample has:
a) An API Gravity of within 0.3 degrees at 60 F or a specific gravity of within 0.0016 at 60/60 F, when compared to the supplier's certificate; or an absolute specific gravity at 60/60*F, of greater than or equal to 0.83 but less than or equal to 0.89; or an API gravity at 60*F of greater than or equal to 26 degrees but less than or equal to 39 degrees, when tested in accordance with ASTM D1298-85, l b) A kinematic viscosity at 40 C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes, when test-ing in accordance with the tests specified in ASTM D975-89, if
- gravity was not detennined by comparison with the supplier's certification, c) A flash point equal to or greater than 125"F, when tested in accordance with the tests specified in ASTM 0975-89, d) No visible freetwater or particulate contamination when tested in accordance with ASTM D4176-86.
2. By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM 0975-89 are met when tested in accordance with the tests specified in ASTM D975-89.

e. At least once every 31 days by obtaining a sample of fuel oil from the storage tanks in accordance with ASTM D2276-88, and verifying that total-particulate contamination is less than 10 mg/ liter when tested in accord-ance with ASTN 02276-88.

f. At least once per 18 months *,
ring shutdown, by:
1. Subjecting the diesel to an inspection in accordance with instructions prepared in conjunction with its manufacturer's recommendations for this class of standby service.

2. Verifying the diesel generator capability to reject a load of greater than or equal to 1400 kw (LPCS pump) for diesel generator Div 1, greater than or equal to 729 kw (RHR B pump or RHR C pump)
*For any start of a diesel, the diesel must be loaded in accordance with the manufacturer's reconmendations. *"xcept 4.8.1.1.2. f.1 to ce performed every refueling outage, for the Div i and l . :. ~ c
ELECTRICAL POWER SYSTEMS j f)) ] SURVEILLANCE REQUIREMENTS (Continued) 1 for diesel generator Div 2, and greater than or equal to 2400 kw l (HPCS pump) for diesel generator.Div 3 while maintaining speed less than nominal speed plus 75% of the difference between nominal speed and the overspeed trip setpoint or 15% above nominal, whichever is less. .
3. ' Verifying the diesel generator capability to reject a load of l 5800 kw for diesel generators Div 1 and Div 2 and 2600 kw for '
diesel generator Div 3 without tripping. The generator voltage shall not exceed 4784 volts for Div 1 and Div 2 and 5000 volts for.Div 3 during and following the load rejection.
4. Simulating a loss of offsite power by itself, and:
a) For divisions 1 and 2:
1) Verifying de energization of the emergency busses and load shedding from the emergency busses.

2) Verifying the diesel generator starts
on the auto-start signal, energizes the emergency busses with per-manently connected loads within 10 seconds, energizes the auto-connected loads through the load sequence .
(individual load timers) and operates for greater than or equal to 5 minutes wnile its generator is so loaded. After energization, the steady state voltage and frequency of the emergency busses shall be main-tained at 4160 1 420 volts and 6011.2 Hz during this test. b) For division 3: Verifying de-energization of the emergency bus. 1)
2) Verifying the-diesel generator starts
on the auto-start signal, energizes the emergency bus with the per-manently connected loads within 13 seconds and operates for greater than or equal to 5 minutes while its gen-erator is so loaded. After energization, the steady "All diesel generator starts for the purpose of this Surveillance Requirement may be preceded by an engine prelube period. The diesel generator start
. (10 sec)/ load (60 sec) from ambient conditions shall be perfomed at least ,
, once per 184 days in these surveillance tests. All other engine starts for the purpose of this surveillance testing may be preceded by other warmup pro-cedures recommended by the manufacturer so that the mechanical stress and wear on the diesel engine is minimized. PERRY - UNIT 1 3/4 8-6 Amendment No. 22
. ~ - . ., - - . . .- . . - --.- --- - , ~ _ -
t ELECTRICAL POWER SYSTEMS gg g i
-SURVEILLANCE REQUIREMENTS (Continued) state voltage and frequency of the emergency bus shall be maintained at 4160 1 420 volts and 60 i 1.2 Hz during this test.
5. Verifying that on an ECCS actuation test signal, without loss
of offsite power, the diesel generator ' starts
on the auto-start !
signal and operates on standby for greater than or equal to' 5 minutes. The generator voltage and frequency shall be 4160 , i 420 volts and 60 i 1.2 Hz within 10 seconds after the auto- -; start signal for Div 1 and Div 2 and within 13 seconds after the , auto-start signal for Div 3; the steady state generator voltage , and frequency shall be maintained within these limits during this , , test. L 6. Simulating a loss of offsite power in conjunction with an ECCS l actuation test signal, and: ' a) For divisions 1 and 2:
1) Verifying de-energization of the emergency busses and load shedding from the emergency busses.

2) Verifying the diesel generator starts
on the auto-start signal, energizes the emergency busses with permanently ;
connected loads.within 10 seconds, energizes the auto-connected emergency loads and operates for greater than or equal to 5 minutes while its generator is , loaded with the emergency loads. After energization, the steady state voltage and frequency of the emergency ' busses shall be maintained at 4160
420 volts and 60 i 1.2 Hz during this test.
b) For division 3:
1) Verifying de-energization of the emergency bus. ;

2) Verifying the diesel generator starts
on the auto- l start signal, energizes the emergency bus with its loads and the auto-connected emergency loads.within 13 seconds and operates for greater than.or equal to 5 minutes while its generator is loaded with the emer- i gency loads. After energization, the steady state j l
l
*All diesel generator starts for the purpose of this Surveill'ance Requirement may be preceded by an engine prelube period. The diesel generator start (10 sec)/ load (60 sec) from ambient conditior.s shall be performed at least once per 184 days in these surveillance tests. All other engine starts for ,
the purpose of this surveillance testing may be preceded by other warmup pro- i cedures recommended by the manufacturer so that the mechanical stress and 1 wear on the diesel engine is minimized.
/
PERRY - UNIT 1 3/4 8-7
. , .- _ .. _ _ _ _ _. ~ .
ELECTRICAL' POWER' SYSTEMS 8 - SURVEILLANCE REQUIREMENTS (' Continued) voltage and frequency of the emergency bus'shall be. maintained at 4160 1 420 volts and 60 1_1.2 Hz during this test.
7. Verifying that all automatic diesel generator tripe ere auto- }
matica11y bypassed with an ECCS actuation signai except: '
.,a) For divisions 1 and-2, engine overspeed and generator r differential current.
b) For division 3, engine overspeed and generator differential current. ,
8. Verifying the diesel generator operates for at least 24 hou.s.
, During this test, the diesel generator shall be loaded to between -
6800-7000 kw for the first two hours and between 5600-5800 kw** ' for the remaining 22 hours for diesel generator Div 1 and Div 2. - The Div 3 diesel generator shall be loaded to greater than or equal to 2860 kw for the first two hours of this test and 2600 kw for the remaining 22 hours of this test. The generator voltage and frequency shall be 4160 1 420 volts and 60 1 1.2 Hz within i 10 seconds after the start signal for Div 1 and Div 2 and within ' 13 seconds after the start signal for Div 3; the steady state ' . generator voltage and frequency shall.be maintained within these limits during this test. Within 5 minutes after completing this ; 24-hour test, perform Surveillance Requirement 4.8.1.1.2.f.4.a.2 and b.2* or perform Surveillance Requirement 4.8.1.1.2.f.6.a.2. , and b.2.* i
9. Verifying that the auto-connected loads to each diesel generator do not exceed 7000 kw for diesel generator Div 1.and Div 2 and t 2860 kw for diesel generator Div 3. ,

10. Verifying the diesel generator's capability toi :
a) Synchronize with the offsite power source while the generator f is loaded with its emergency loads upon a simulated ; restoration of offsite power, ! b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.
11. Verifying that w'ith the diesel generator' operating in a test mode I and connected to its bus, a simulated ECCS actuation signal over-rides the test mode by (1) returning the diesel generator to !
1
*If Surveillance Requirements 4.8.1.1.2.f.4.a.2 and b.2 or 4.8.1.1.2.f.6.a.2 l r and b.2 are not satisfactorily completed, it is not necessary to repeat the preceding 24 hour test. Instead, the diesel generator Div 1 or Div 2 may be <
operated at 5600-5800 kw or diesel generator Div 3 may be operated at 2600 kw - for one hour or until operating temperatures have stabilized. '
**This band is meant as guidance to avoid routine overloading of the engine.
Loads in excess of this band shall not invalidate the. test; the loads, however, shall not be less than 5600 kw nor greater tLan 7000 kw.
- PERRY - UNIT 1 3/4 8-8 Amendment No. 42 i
d' _ .__1 _ _ . _ _ _ _ _ _ _ _ _ _ . - . _ - - - _ , - - - _ --
ELECTRICAL POWER SYSTEMS .y SURVEILLANCE REQUIREMENTS (Continued) l to standby operation, and (2) automatically energizes the emer- l gency loads with offsite power. I
12. Verifying that each fuel transfer pump transfers fuel from the fuel storage tank to the day tank of each diesel.

13. Verifying that the automatic load sequence timers are OPEPABLE with the interval between each load block within + 10% of its
' design interval for diesel generators Div 1 and DIV 2.
14. Verifying ~that the following diesel generator lockout features prevent diesel generator starting only when required:

a. For diesel generators Div 1 and Div 2:

1) Control room switch in pull-to-lock (with local / remote switch in remote).

2) local / remote switch in local

3) Barring device engaged

4) Inop/ Normal switch in inop

b. For diesel generator Div 3:

1) Emergency run/stop switch in stop ;

2) Maintenance / auto / test switch in maintenance

3) Control room switch in pull-to-lock position 9 At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting all three diesel generators simultaneously, during shutdown, and verifying that all three diesel generators accelerate to at least 441 rpm for diesel generators Div 1 and Div 2 and 882 rpm for diesel generator Div 3 in less than or equal to 10 seconds,

h. At least once per 10 years by:

1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite'on equivalent solution, and

2. Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code in accordance with ASME Code Section 11 Article IWD-5000.
4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall be I reported to the Commission pursuant to Specification 6.9.2 within 30 days. j Reports of diesel generator failures shall include the information recommended ; in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. ! If the number of failures in the last 100 valid tests of any diesel generator ' is greater than or equal to seven, the report shall be supplemented to in-clude the additional infonnation recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. l 1 PERRY - UNIT 1 3/4 8-9 Amendment No. J2,M , 37 l I
Attcchnent 3 PY-CEI/NRR-1890L Page 1 of 4 SIGNIFICANT HAZARDS CONSIDERATION The standards used to arrive at a determination that a request for amendment i involves no significant hazards considerations are included in the Commission's l Regulations, 10 CFR 50.92, which state that the operation of the facility in I 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 previously evaluated, or (3) involve a significant reduction in a margin of ' safety. The proposed amendment 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 proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
1 The proposed TS change requests a one-time extension of the surveillance ) intervals related to: a) RPS Instrumentation calibration, LSFTs, and ; response time testing; b) Isolation Actuation System Instrumentation calibration, LSFTs, and response time testing; c) ECCS Actuation Instrumentation calibration, LSFTs, and response time testing; d) Control Rod Block Instrumentation calibration and LSFTs; e) Remote Shutdown Instrumentation and Controls calibration and operability testing; f) Accident Monitoring Instrumentation calibration; g) Plant Systems Instrumentation calibration and LSFTs; h) Primary Containment automatic valve actuation; i) Reactor Coolant System Pressure Isolation Valve (PIV) testing; j) system automatic initiation testing; and, k) Emergency Diesel Generator inspection and testing. Also proposed is the re-establishment of the baseline for the "N times 18 months" cumulative surveillance interval for response time testing. The discussion in the License Amendment Request demonstrates the following: i) Rosemount transmitter calibration period extension is acceptable based on Rosemount Report D8900126, Revision A which supported extension of the calibration interval from 18 months to 30 months based on the reduction in the drift allowance;
11) Extrapolation of plant specific calibration data is acceptable in supporting the extension of other calibration surveillance intervals to RFO-5; iii) LSFT interval extension is acceptable based on the NRC Safety Evaluation Report (Peach Bottom Atomic Power Plant, Units 2 and 3, dated August 2, 1993) which supported extension of the interval for LSFT from 18 to 24 months. This was based on the small probability of relay or contact failure relative to mechanical component failure probability and, therefore, the increase in LSFT interval represented no significant change in the overall safety system unavailability;
Attach;;nt 3 PY-CEI/NRR-1890L Page 2 of 4 iv) Response time testing interval extension for Isolation Actuation and ECCS Actuation instrumentation channels is acceptable based on the BVR Owners Group (BVROG) Licensing Topical Report NED0-32291 (January 1994) which provided the necessary justification for elimination of response time testing and, therefore, provides a suitable argument for extending the 4 interval for a short period of time. The NRC approved the use of NED0-32291 { as a basis for License Amendment Requests, with additional conditions l specified, in a letter to the BVROG in December 1994. v) Response time testing interval extension for RPS Instrumentation channels is acceptable because: i) there are redundant sensors that can initiate the scram function; ii) one-out-of-two redundancy exists in every individual instrument channel within each trip functions lii) several redundant and diverse instrument channels are provided which can detect and generate a scram signal; iv) the failure probability is e small fraction of the total control rod insertion (scram) failure probability; v) failure of instrumentation in the sluggish mode is a small fraction of its overall failure modes; and iv) NRC Safety Evaluation Report dated August 2, 1993 (Peach Bottom Atomic Power Station, Units 2 and 3 docket) has previously provided approval for extension of the RPS response time testing surveillance interval from 18 to 24 months. vi) Response time testing interval extension for the Main Steam Line isolation is acceptable because 1) redundancy and diversity exist in individual instrument channels within a trip function; ii) instrumentation response time is a small fraction of the overall response time of the actuating device; iii) instrumentation failure probability is a very small portion of the total MSIV failure probability; and, iv) failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure modes. vii) Containment Isolation Valve leakage determination and actuation interval extension is acceptable based on: i) redundancy provided in the design of the penetrations; ii) the pericJic testing of the valves during power operation; and, iii) the short period of time the interval is being extended. viii) Reactor Coolant System PIVs have exhibited lov as-found leak rates as measured during the last refueling ou: age; there is substantial margin available for the PIVs from the as-left leakage to the allowed TS leakage; the requested extension of the surveillance interval is small; and the conclusion of NUREG-1463, " Regulatory Analysis for the Resolution of Generic Safety Issue 105: Interfacing System Loss-of-Coolant Accident in Light i Water Reactors" (July 1993), and the confirmation of the PNPP Individual l Plant Examination that the ISL9CA (for which PIVs are provided to prevent) is not a risk concern to BVRs ot PNPP. ix) System initiation and actuation testing interval is acceptable based on l the periodic testing of components during power operation and the short period of time the interval is being extended. x) Emergency Diesel Generator testing interval extension is acceptable based on: i) the past testing results which support extension for the short period of time; 11) the testing that is done during power operation; and, iii) the short period of time the interval is being extended.
1 Attechnenti 3 ~ PY-CEI/NRR-1890L- j Page 3 of 4 i i xi) The re-establishment of the baseline for the "N times 18 months" i cumulative surveillance interval for_ response time testing is acceptable in ; that the extension of the cumulative interval would not be for more than the 'I individual extension requested and justified herein. Therefore, from the above it is shown that the proposed change will not significantly increase the probability of an accident previously evaluated. ,
^1
2. The proposed change would not create the possibility of a new or different kind of accident from any accident previously evaluated. 7 The proposed TS change requests a one-time extension of the surveillance l
. intervals for instrument calibration, instrument channel LSFT and response i time testing, containment isolation valve leakage determination and ' actuation, PIV leak rate determination, system actuation testing, and diesel generator inspection and testing. The proposed changes do not necessitate a ! physical alteration to the plant (no new or different type of equipment will be installed). The requested extension durations are small as compared to : the overall interval allowed by TS; drift data supports extension of the i' calibration intervals; NRC and industry evaluations support extension of LSFT; industry evaluations and redundancy in system design support extension of response time testing; past testing and periodic testing provides i confidence of no effect on equipment availability by extending the j surveillance interval. Therefore, the change does not create the possibility of a new or different kind of accident from any accident f previously evaluated. j In addition, the requested re-establishment-of the baseline at RFO-5 for the -l i
"N time 18 months" cumulative surveillar.ce interval for response time testing is acceptable in that the cumulative surveillance interval vill not i be extended by more than that which is proposed for individual response time 1 tests during RFO-5. The individual response time test surveillance interval !
extensions have been justified herein. The justification for individual response time test surveillance interval extensions applies to the cumulative surveillance interval extension which is requested and will be granted-by allowing the re-establishment of the baseline of the "N times 18 months" surveillance interval to the response time testing dates for those response time tests to be performed during RFO-5. The proposed changes do not necessitate a physical alteration to the plant (no new or different type , of equipment will be installed). Therefore, the change does not create the . possibility of a new or different kind of accident.
3. The proposed change will not involve a significant reduction in the margin ;
of safety. j The proposed TS change requests a one-time extension of the surveillance intervals for instrument calibration, instrument channel LSFT, and response ; time testing, containment isolation valve leakage determination and j actuation, PIV leak rate determination, system actuation testing, and diesel '! generator inspection and testing. The proposed changes do not necessitate a ; physical alteration to the plant (no new or different type of equipment will be installed). In that the requested extension durations are small as compared to the overall interval allowed by TS, drift data supports : extension of the calibration intervals, NRC and industry evaluations support extension of LSFT, industry l l
Attechment 3 PY-CEI/NRR-1890L Page 4 of 4 evaluations and redu..dsney in system design support extension of response time testing, past testing and periodic testing provides confidence of no effect on equipment availability by extending the surveillance interval, the change does not involve a significant reduction in the margin of safety. ; In addition, the requested re-establishment of the baseline at RFO-5 for the "N times 18 months" cumulative surveillance interval for response time testing is acceptable in that the cumulative surveillance interval vill not be extended by more than that which is proposed for individual response time tests during RFO-5. The individual response time test surveillance interval extensions have been justified herein. The justification for individual response time test surveillance interval extensions applies to the cumulative surveillance interval extension which is requested and will be granted by allowing the re-establishment of the baseline of the "N times 18 months" surveillance interval to the response time testing dates for those response time tests to be performed during RFO-5. The proposed changes do not necessitate a physical alteration to the plant (no new or different type of equipment vill be installed). Therefore, the change does not involve a significant reduction in the margin of safety.
, . . - - > - A L re~-. 4- o A 4e c + a a s a s +.
,' sf i L. i t: Enclosures PY-CEI/NRR-3890L E I I ! I I l I i 1
l ENCLOSURE 1 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l LOGIC SYSTEM FUNCTIONAL TESTING FOR ; TECHNICAL SPECIFICATION SR 4.3.1.2, TABLE 4.3.1.1-1, ITEM 13, SR 4.1.3.1.4.a.1 AND-SR 4.1.3.1.4.a.2 . REACTOR PROTECTION SYSTEM MANUAL SCRAM AND SCRAM DISCHARGE VOLUME VENT AND DRAIN VALVE OPERABILITY - REACTOR PROTECTION SYSTEM Technical Specification SF. 4.3.1.2 requires a LSFT and simulated automatic . actuation of all channels of the Reactor Protection System (RPS) at least once 3 per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Manual Scram functional unit (TS Table 4.3.1.1-1, ! Item 13) provides for manual initiation of the RPS logic including closure and opening (upon logic reset) of the Scram Discharge Volume (SDV) Vent and Drain Valves, as required by TS SRs 4.1.3.1.4.a.1 and 4.1.3.1.4.a.2. The Manual Scram functional unit of TS Table 4.3.1.1-1 and the SDV Vent and Drain Valves closure within 30 seconds and opening per TS SRs 4.1.3.1.4.a.1 and 4.1.3.1.4.a.2, i respectively, require a surveillance interval extension for a nominal period of 24 days to reach the most conservative projected start of RFO-5. However, in that the Manual Scram functional unit is required to be OPERABLE in Operational l' Conditions 1, 2, 3, 4 and 5, a total extension of the surveillance interval for this functional unit to the most conservative projected end of RFO-5 for a nominal period of 116 days.is required. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BVR Ovners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of ' the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant , change in the overall safety system unavailability." The SDV Vent and Drain Valves are required to be cycled at least once per 92 - days by TS SR 4.1.3.1.1.a thereby verifying that the valves are capable of closing and opening. The logic providing input to close the valves within 30 seconds and open the valves is the only remaining portion of the SRs which requires extension. In that this logic is the subject to the evaluation above, and the valves are cycled periodically during the operating cycle, the extension of the surveillance intervals for SRs 4.1.3.1.4.a.1 and 4.1.3.1.4.a.2 for a nominal period of 37 days to reach the most conservative projected start of RFO-5 is justified. The evaluation above is applicable to PNPP and the surveillance interval extension for the TS Table 4.3.1.1-1, Item 13 functional unit (Manual Scram) for a nominal period of 116 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. l 5
l 1 i ENCLOSURE 2 l JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l TECHNICAL SPECIFICATION SR 4.3.1.1, TABLE 4.3.1.1-1, ITEM 2.b(i) l APRM - FLOV BIASED SIMULATED THERMAL POVER - HIGH TIME CONSTANT ! CALIBRATION REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.1, Table 4.3.1.1-1, Item 2.b(i) requires the Reactor Protection System Average Power Range Monitor Flow Biased Simulated
. Thermal Power - High Time Constant to be demonstrated OPERABLE by performance of-a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The calibration of the Time Constant is ,
essentially a response time test. This testing requires an extension for a ; nominal period of 47 days to reach the most conservative projected start of l RFO-5. The extension would have no substantial effect on plant safety because:
a. There are several redundant APRMs that can initiate the scram operation.

b. One-out-of-two redundancy exists in every individual instrument channel within each trip function.

c. The instrumentation failure probability is a very small fraction of the total control rod insertion (scram failure) probability.

d. There are several redundant and diverse instrument channels which-can detect and generate a scram signal (i.e., flux, pressure, etc.)

e. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure..
Extension of the Peach Bottom Atomic Power Station Units 2 and 3 surveillance intervals for RPS response time testing from 18 to 24 months was accepted by the " NRC in the Safety Evaluation Report dated August 2, 1993. The justification cited above is very similar to that provided in the NRC Safety Evaluation Report which states:
"The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS system is such that either subchannel can trip a trip system and that both trip systems must trip to cause a reactor trip. The logic is such that a single failure vill neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent redundancy in' the RPS system, the impact of extending the response time surveillance interval on system availability is small."
Based on the above, a one-time extension of the RPS Instrumentation, 15 Table 4.3.1.1-1, Item 2.b(i) calibration (response time testing) surveillance interval is justified.
ENCLOSURE 3 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.1.1, TABLE 4.3.1.1-1, ITEM 2.b(m) TECHNICAL SPECIFICATION SR 4.3.6.1, TABLE 4.3.6-1, ITEMS 2.a(c) AND 6.a(c) RPS/ CONTROL R0D BLOCK RECIRCULATION FLOW TRANSHITTER CALIBRATION' REACTOR PROTECTION SYSTEM AND CONTROL ROD BLOCK INSTRUMENTATION Technical Specification (TS) SR 4.3.1.1, Table 4.3.1.1-1, Item 2.b(m) and TS SR 4.3.6.1, Table 4.3.6, Items 2.a(c) and 6.a(c) requires the Recirculation Flow Reference Transmitters which provide input to the Average Power Range Monitors Flow Biased Thermal Power - High Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). The flov transmitters, Rosemount Model 1153 flov transmitters, vill require an extension of the SR intervals cited in TS Table 4.3.1.1-1, Item 2.b(m) and TS Table 4.3.6-1, Items 2.a(c) and 6.a(c) for a nominal period of 47 days to reach the most conservative projected start of RF0-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.) This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. It should also be noted that, the PNPP Setpoint Calculation for the Recirculation Loop Flov indicates that instrument loop shift is significantly less than the existing 3% of rated flow margin which is maintained between the Nominal Trip Setpoint (111%) and the Allovable Value (114%). Actual loop drift (calculated) is less than 1.5%. 1 The existing PNPP setpoint calculations for Rosemount transmitters and l associated trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. i l
\
1 1 l l l ,
ENCLOSURE 4 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR . TECHNICAL SPECIFICATION SR 4.3.1.2, TABLE 4.3.1.1-1, ITEM 2.b APRM FLOV BIASED SIMULATED THERMAL POWER - HIGH REACTOR PROTECTION SYSTEM Technical Specification SR 4.3.1.2 requires a LSFT and simulated automatic actuation of all channels of the Reactor Protection System at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Average Power Range Monitor: Flow Biased Simulated Thermal Power functional unit (TS Table 4.3.1.1-1, Item 2.b) requires a surveillance , interval extension for this functional unit's portion of the LSFT for a nominal period of 20 days to reach the most conservative projected start of RFO-5. . As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3,- surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs), '
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 20 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. i i r i t '
f 1 Y ENCLOSURE 5 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR , TECHNICAL SPECIFICATION SR 4.3.1.3, TABLE 3.3.1-2, ITEM 2.b APRM - FLOV BIASED SIMULATED POVER - HIGH REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.3 requires the Response Time of the Reactor Protection System (RPS) Instrumentation shown on Table 3.3.1-2 be demonstrated to be within the. limits at least once per 18 months. Each of the tests are to
-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 reactor trip system. Trip function Item 2.b, Average Power Range Monitor Flow Biased Simulated Thermal Power - High, of Table 3.3.1-2 vill become overdue prior to the beginning of RFO-5 conservatively scheduled to begin February 15, 1996. This RPS response time test requires an extension for a nominal period of 47 days to reach the scheduled start of RFO-5.
The extension would have no substantial measurable effect on plant safety because:
a. There are redundant sensors that can initiate the scram operation.

b. One-out-of-two redundancy exists in every individual channel within each trip function.

c. The instrumentation failure probability is a very small fraction of the total control rod insertion (scram failure) probability.

d. There are several redundant and diverse instrument channels which can detect and generate a scram signal (e.g., flux, pressure, etc.).

e. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure.
Extension of the Peach Bottom Atomic Power Station, Units 2 and 3 surveillance intervals for RPS response time testing from 18 to 24 months was accepted by the NRC in the Safety Evaluation Report dated August 2, 1993. The justification cited above is very similar to that provided in the NRC Safety Evaluation Report which states:
"The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS system is such that either subchannel can trip a trip system and that .
both trip systems must trip to cause a reactor trip. The logic is such l that a single failure vill neither cause nor prevent a required reactor ! scram. The licensee states that, based on the inherent redundancy in j the RPS system, the impact of extending the response time surveillance ! interval on system availability is small." ! Based on the above, a one-time extension of the RPS Instrumentation, TS Table 3.3.1-2, Item 2.b, response time testing surveillance interval is justified.
O- J ENCLOSURE 6 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL' SPECIFICATION -SR 4.3.1.3, TABLE 3.3.1-2, ITEM 2.c ! APRM - NEUTRON FLUX - HIGH REACTOR PROTECTION SYSTEM INSTRUMENTATION-Technical Specification SR 4.3.1.3 requires the Response Time of the Reactor Protection System (RPS) Instrumentation shown on Table 3.3.1-2~be demonstrated to be within the limits at least once per 18 months. Each of.the tests are to i 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 reactor trip system. Trip function Item 2.c, Average Power Range Monitor Neutron Flux - High, of Table 3.3.1-2 vill become overdue prior to the beginning of RFO-5 conservatively scheduled to begin February 15, 1996. This RPS response time test requires an extension for a nominal period of 47 days to reach the scheduled start of RFO-5. The extension would have no substantial measurable effect on plant safety because:
a. There are redundant sensors that can initiate the scram operation. '

b. One-out-of-tvo redundancy exists in every individual channel within each trip function.

c. The instrumentation failure probability is a very small fraction of the total control rod insertion (scram failure) probability.

d. There are several redundant and diverse instrument channels which can detect and generate a scram signal (e.g., flux, pressure, etc.).

e. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure.
Extension of the Peach Bottom Atomic Power Station, Units 2 and 3 curveillance i intervals for RPS response time testing from 18 to 24 months was accepted by the l NRC in the Safety Evaluation Report dated August 2, 1993. The justification i cited above is very similar to that provided in the NRC Safety Evaluation Report ! which states: i "The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS system is such that either subchannel can trip a trip system and that both trip systems must trip to cause a reactor trip. The logic is such that a single failure vill neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent redundancy in the RPS system, the impact of extending the response time surveillance interval on system availability is small." Based on the above, a one-time extension of the RPS Instrumentation, TS Table 3.3.1-2, Item 2.c, response time testing surveillance interval is justified. 1
I l ENCLOSURE 7 l JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.1.1, TABLE 4.3.1.1-1, ITEM 3 REACTOR VESSEL STEAM DOME PRESSURE - HIGH CALIBRATION REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.1. Table 4.3.1.1-1, Item 3 requires the Reactor Protection System Reactor Vessel Steam Dome Pressure - High Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once
~
per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). Two channels of the instrumentation-(channels A and C), containing Rosemount Model 1153 pressure transmitters, vill require an extension of the SR interval cited in TS Table 4.3.1.1-1, Item 3 for a nominal period of 40 days to reach the most conservative projected start of RFO-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drif t allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift < value of 0.13% SP (2 sigma). The existing PNPP setpoint calculations for Rosemount transmitters and associated trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. i
r ? ENCLOSURE 8 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.1.2, TABLE 4.3.1.1-1, ITEM 3 REACTOR VESSEL STEAM DOME PRESSURE - HIGH , REACTOR PROTECTION SYSTEM INSTRUMENTATION , Technical Specification SR 4.3.1.2 requires a LSFT and simulated automatic actuation of all channels of the Reactor Protection System (RPS) at least once , per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel Steam Dome Pressure - High functional unit (TS Table 4.3.1.1-1, Item 3) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 40 days to reach the most conservative projected start of RFO-5. + As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to ;
~
extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, ' surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the~BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps ' and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small ; relative'to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." p The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 40 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is ' justified. s e T
I I ENCLOSURE 9 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.1.3, TABLE 3.3.1-2, ITEM 3 REACTOR VESSEL STEAM DOME PRESSURE - HIGH REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.3 requires the Response Time of the Reactor Protection System (RPS) Instrumentation shown on Table 3.3.1-2 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to 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 reactor trip system. Trip function Item 3, Reactor Vessel Steam Dome Pressure - High, of Table 3.3.1-2 vill become overdue prior to the beginning of RF0-5 conservatively scheduled to begin February 15, 1996. This RPS response time test requires an extension for a nominal period of 26 days to reach the scheduled start of RFO-5. The extension would have no substantial measurable effect on plant safety because:
a. There are redundant sensors that can initiate the scram operation.

b. One-out-of-two redundancy exists in every individual channel within each trip function,

c. The instrumentation failure probability is a very small fraction of the total control rod insertion (scram failure) probability.

d. There are several redundant and diverse instrument channels which can detect and generate a scram signal (e.g., flux, pressure, etc.).

e. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure.
Extension of the Peach Bottom Atomic Power Station, Units 2 and 3 surveillance intervals for RPS response time testing from 18 to 24 months was accepted by the NRC in the Safety Evaluation Report dated August 2, 1993. The justification cited above is very similar to that provided in the NRC Safety Evaluation Report which states:
"The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS system is such that either subchannel can trip a trip system and that l both trip systems must trip to cause a reactor trip. The logic is such ;
that a single failure vill neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent redundancy in the RPS system, the impact of extending the response time surveillance interval on system availability is small." Based on the above, a one-time extension of the RPS Instrumentation, TS Table 3.3.1-2, Item 3, response time testing surveillance interval is justified. i
ENCLOSURE 10 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.1.1, TABLE 4.3.1.1-1, ITEM 4 REACTOR VESSEL VATER LEVEL - LOV, LEVEL 3 CALIBRATION REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.1, Table 4.3.1.1-1, Item 4 requires the Reactor Protection System Reactor Vesse't Vater Level- Lov, Level 3 Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). One , channel of the instrumentation (channel D), containing a Rosemount Model 1153 pressure transmitter, vill require an extension of the SR interval cited in TS Table 4.3.1.1-1, Item 4 for a aominal period of 8 days to reach the most conservative projected start c f RF0-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Pevision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report nupported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly 01: quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing PNPP setpoint calculations for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Thtrefore, the requested extension is justified.
ENCLOSURE 11 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.1.2, TABLE 4.3.1.1-1, ITEM 4 REACTOR VESSEL VATER LEVEL - LOV, LEVEL 3 REACTOR PROTECTION SYSTEM Technical Specification SR 4.3.1.2 requires a LSFT and simulated automatic actuation of all channels of the Reactor Protection System (RPS) at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor vessel Vater Level - Lov, Level 3 functional unit (TS Table 4.3.1.1-1, Item 4) requires a surveillance interval extension for this fonctional unit's portion of the LSFT for a nominal period of 8 days to-reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval l extension for a nominal period of 8 days is bounded by the. interval' accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. l t
1 ENCLOSURE 12 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.1.3, TABLE 3.3.1-2, ITEM 4 ; REACTOR VESSEL VATER LEVEL - LOV, LEVEL 3 REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.3 requires the Response Time of the Reactor Protection System (RPS) Instrumentation shown on Table 3.3.1-2 be demonstrated to be within the limits at least_once per 18 months. Each of the tests are to , 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 reactor trip system. Trip function Item 4, Reactor Vessel Water Level - Lov, Level 3, of Table 3.3.1-2 vill become overdue prior to the beginning of RFO-5 conservatively scheduled to begin February 15, 1996. ' This RPS response time test requires an extension for a nominal period of 7 days to reach the scheduled start of RFO-5. The extension vould have no substantial measurable effect on plant safety because:
a. There are redundant sensors that can initiate the scram operation. ,

b. One-out-of-two redundancy exists in every individual channel within each trip function.

c. The instrumentation failure probability is a very small fraction of the total control rod insertion (scram failure) probability.

d. There are several redundant and diverse instrument channels which can detect and generate a scram signal (e.g., flux, pressure, etc.).

e. The failure of instrumentation in the sluggish responding mode is a i small fraction of its overall failure.
Extension of the Peach Bottom Atomic Power Station, Units 2 and 3 surveillance intervals for RPS response time testing from 18 to 24 months was accepted by the NRC in the Safety Evaluation Report dated August 2, 1993. The justification cited above is very similar to that provided in the NRC Safety Evaluation Report which states:
"The RPS system consists of two independent trip systems with at least ,
two subchannels of a parameter per trip system. The logic of the RPS system is such that either subchannel can trip a trip system and that both trip systems must trip to cause a reactor trip. The logic is such ; that a single failure vill neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent redundancy in the RPS system, the impact of extending the response time surveillance interval on system availability is small." Based on the above, a one-time extension of the RPS Instrumentation, TS Table 3.3.1-2, Item 4, response time testing surveillance interval is justified. 1 l
7_ ENCLOSURE 13 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.1.1, TABLE 4.3.1.1-1, ITEM 5 REACTOR VESSEL VATER LEVEL - HIGH, LEVEL 8 CALIBRATION REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.1, Table 4.3.1.1-1, Item 5 requires the Reactor Protection System Reactor Vessel Vater Level- High, Level 8 Instrumentation to be demonstrated OPERABLE by performance of a channel calibratio3 at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). One channel of the instrumentation (channel D), containing a Rosemount Model 1153 pressure transmitter, vill require an extension of the SR interval cited in TS Table 4.3.1.1-1, Item 5 for a nominal period of 8 days to reach the most conservative projected start of RFO-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.) This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified.
f ENCLOSURE 14 ! I JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR i TECHNICAL SPECIFICATION SR 4.3.1.2, TABLE 4.3.1.1-1, ITEM 5 REACTOR VESSEL VATER LEVEL - HIGH, LEVEL 8 l REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.2 requires a LSFT and simulated automatic , actuation of all channels of the Reactor Protection System (RPS) at least'once per 18 months (with a maximum allovable surveillance interval extension of.4.5 months per TS 4.0.2). The Reactor Vessel Vater Level - High, Level 8 functional unit (TS Table 4.3.1.1-1, Item 5) requires a surveillance interval extension for , this functional unit's portion of the LSFT for a nominal period of 8 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of - the logic system, but by that of the mechanical components, (e.g., pumps ' and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small : relative to the probability of mechanical component failure, increasing ! the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 8 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is { justified.
ENCLOSURE 15 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.1.3, TABLE 3.3.1-2, ITEM 5 REACTOR VESSEL VATER LEVEL - HIGH, LEVEL 8 REACTOR PROTECTION SYSTEM INSTRUMENTATION i Technical Specification SR 4.3.1.3 requires the Response Time of the Reactor Protection System (RPS) Instrumentation shown on Table 3.3.1-2 be demonstrated to be within the limits at least once per 18 nonths. Each of the tests are to 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 reactor trip system. Trip function Item 5, Reactor Vessel Water Level - High, Level 8, of Table 3.3.1-2 vill become overdue prior to the beginning of RFO-5 conservatively scheduled to begin February 15, 1996. This RPS response time test requires an extension for a nominal period of 7 days to reach the scheduled start of RFO-5. The extension vould have no substantial measurable effect on plant safety because:
a. There are redundant sensors that can initiate the scram operation.

b. One-out-of-two redundancy exists in every individual channel within each trip function,

c. The instrumentation failure probability is a very small fraction of the total control rod insertion (scram failure) probability,

d. There are several redundant and diverse instrument channels which can detect and generate a scram signal (e.g., flux, pressure, etc.).

e. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure.
Extension of the Peach Bottom Atomic Power Station, Units 2 and 3 surveillance intervals for RPS response time testing from 18 to 24 months was accepted by the NRC in the Safety Evaluation Report dated August 2, 1993. The justification cited above is very similar to that provided in the NRC Safety Evaluation Report which states:
"The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS system is such that either subchannel can trip a trip system and that both trip systems must trip to cause a reactor trip. The logic is such that a single failure vill neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent redundancy in the RPS system, the impact of extending the response time surveillance interval on system availability is small."
Based on the above, a one-time extension of the RPS Instrumentation, TS Table 3.3.1-2, Item 5, response time testing surveillance interval is justified.
b ENCLOSURE 16
+
JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.1.3, TABLE 3.3.1-2, ITEM 6 MAIN STEAM LINE ISOLATION VALVE - CLOSURE REACTOR PROTECTION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.1.3 requires the Response Time of the Reactor Protection System (RPS) Instrumentation shown on Table 3.3.1-2 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to 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 l channels in a specific reactor trip system. Trip function Item 6, Main Steam
~
j Line Isolation Valve - Closure, of Table 3.3.1-2 vill become overdue prior to
~
the beginning of RF0-5 conservatively scheduled to begin February 15, 1996. This RPS response time test requires an extension for a nominal period of 51 days to reach the scheduled start of RFO-5. The extension vould have no substantial measurable effect on plant safety , because:
a. There are redundant sensors that can initiate the scram operation.

b. One-out-of-two redundancy exists in every individual channel within ;
each trip function.
c. The instrumentation failure probability is a very small fraction of the total control rod insertion (scram failure) probability.
t
d. There are several redundant and diverse instrument channels which can detect and generate a scram signal (e.g., flux, pressure, etc.).

e. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure.
Extension of the Peach Bottom Atomic Power Station, Units 2 and 3 surveillance j intervals for RPS response time testing from 18 to 24 months was accepted by the NRC in the Safety Evaluation Report dated August 2, 1993. The justification cited above is very similar to that provided in the NRC Safety Evaluation Report ; which states:
"The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS system is such that either subchannel can trip a trip system and that both trip systems must trip to cause a reactor trip. The logic is such that a single failure vill neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent redundancy in the RPS system, the impact of extending the response time surveillance interval on system availability is small."
Based on the above, a one-time extension of the RPS Instrumentation, TS Table 3.3.1-2, Item 6, response time testing surveillance interval is justified. j l l l
t
-1 ENCLOSURE 17 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTEM ION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.1.2, TABLE 4.3.1.1-1, ITEM 12100)
CHANNEL FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.6.1, TABLE 4.3.6-1, ITEM 7 REACTOR MODE SVITCH SHUTD0VN POSITION REACTOR PROTECTION SYSTEM AND CONTROL ROD BLOCK INSTRUMENTATION Technical Specification SR 4.3.1.2 requires a LSFT and simulated automatic actuation of all channels of the Reactor Protection System'at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). Technical Specification SR 4.3.6.1 requires a Channel Functional Test Control Rod Block trip systems and instrumentation at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Mode Switch Shutdown Position functional unit (TS Table 4.3.1.1-1, Item 12 and TS Table 4.3.6-1, Item 7) requires a surveillance interval extension for this functional unit's portion of the LSFT and Channel : Functional Test for a nominal period of 24 days to reach the most conservative projected start of RFO-5. However, in that this functional unit is required to be OPERABLE in Operational Conditions 1, 2, 3, 4 and 5, a total extension of the surveillance interval to the most conservative projected end. of RFO-5' for a nominal period of 116 days is required. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, l surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of ; the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent - basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing ; the logic system functional test interval represents no significant change in the overall safety system unavailability." i The Channel Functional Test for the Control Rod Block trip function of the Reactor Mode Switch Shutdown Position provides a verification of the logic (alarm and trip functions) within the channel. Therefore, the evaluation above is applicable to the Channel Functional Test for the Control Rod Block i instrumentation. The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 116 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
F' ENCLOSURE 18 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.2.1, TABLE 4.3.2.1-1, ITEM 1.a REACTOR VESSEL VATER LEVEL- LOV, LEVEL 2 CALIBRATION ISOLATION ACTUATION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.2.1, Table 4.3.2.1-1, Item 1.a requires the Isolation Actuation System Reactor Vessel Water Level - Lov, Level 2 Instrumentation to be demonstrated OPERAELE by performance of a channel calibration at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.2.1-1, Item 1.a for a nominal period of 47 days to reach the most conservative projected start of RFO-5. However, this instrumentation is required in Operational Conditions 4 and 5 during periods of core alterations or operations with a potential for draining the reactor vessel; therefore, extension of the surveillance interval for a nominal period of 139 days is required to reach the most conservative projected end of the refueling outage. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. l l l
s ENCLOSURE 19 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.2, TABLE 4.3.2.1-1, ITEMS 1.a, 3.a AND 4.e REACTOR VESSEL' VATER LEVEL - LOV, LEVEL 2
' PRIMARY CONTAINMENT ISOLATION, SECONDARY CONTAINMENT ISOLATION AND REACTOR VATER CLEANUP SYSTEM ISOLATION ISOLATION ACTUATION INSTRUMENTATION i
Technical Specification SR 4.3.2.2 requires a LSFT and simulated automatic actuation of all channels of the Isolation Actuation System at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months '. per TS 4.0.2). The Reactor Vessel Vater Level - Lov, Level 2 functional unit for Primary Containment, Secondary Containment and Reactor Vater Cleanup System isolation (TS Table 4.3.2.1-1, Items 1.a, 3.a and 4.e, respectively) requires a surveillance interval extension for these functional units' portion of the LSFT for a nominal period of 47 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant-change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 47 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. l l
i 1 ENCLOSURE 20 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.2, TABLE 4.3.2.1-1, ITEM 1.c CONTAINMENT AND DRYVELL PURGE EXHAUST PLENUM RADIATION - HIGH PRIMARY CONTAINMENT ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.2 requires a LSFT and simulated automatic actuation of all channels of the Isolation Actuation System at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Containment and Dryvell Purge Exhaust Plenum Radiation - High functional unit for Primary Containment isolation (TS Table 4.3.2.1-1, Item 1.c) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 50 days to reach the most conservative projected start of RF0-5. However, this instrumentation is required in Operational Conditions 4 and 5 during periods of core alterations or operations with a potential for draining the reactor vessel; therefore, extension of the surveillance interval for a nominal period of 142 days is required to reach the most conservative projected end of the refueling outage. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 142 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
e I s ENCLOSURE 21- l JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION ! RESPONSE TIME TESTING FOR' l
., TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 2.c i
MAIN STEAM LINE PRESSURE - LOV : I MAIN STEAM LINE ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation Instrumentation shown on Table 3.3.2-3 be demonstrated to.be within the limits at least once per 18 months. Each of the tests are to include at least one channel per trip system such that all channels are tested at .' east once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system. Trip function Item 2.c, Main Steam Line .; Isolation, Main Steam Line Pressure - Lov, of Table 3.3.2-3 vill become overdue prior to the beginning of RFO-5 conservatively scheduled to begin February 15, 1996. This RPS response time test requires an extension for a nominal period of 36 days to reach the most conservative projected start of RFO-5. The extension would have no substantial measurable effect on plant-safety because:
a. There are redundant sensors that can initiate the scram operation.

b. One-out-of-two redundancy exists in every individual channel within each trip function.

c. The instrumentation failure probability is a very small fraction of the total control rod insertion (scram failure) probability.

d. There are several redundant and diverse instrument channels which can detect and generate a scram signal (e.g., flux, pressure, etc.).

e. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure.
Extension of the Peach Bottom Atomic Power Station, Units 2 and 3 surveillance intervals for RPS response time testing from 18 to 24 months was accepted by the NRC in the Safety Evaluation Report dated August 2, 1993. The justification cited above is very similar to that provided in the NRC Safety Evaluation Report which states:
"The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS system is such that either subchannel can trip a trip system and that both trip systems must trip to cause a reactor trip. The logic is such that a single failure vill neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent redundancy in the RPS system, the impact of extending the response time surveillance interval on system availability is small."
Based on the above, a one-time extension of the RPS Instrumentation, TS Table 3.3.2-3, Item 2.c, response time testing surveillance interval is justified. l I l
ENCLOSURE 22 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.2.1, TABLE 4.3.2.1-1, ITEM 1.d REACTOR VESSEL VATER LEVEL- LOV, LEVEL 1 CALIBRATION ISOLATION ACTUATION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.2.1, Table 4.3.2.1-1, Item 1.d requires the Isolation Actuation System Reactor Vessel Vater Level - Lov, Level 1 l' Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.2.1-1, Item 1.d for a nominal period of 45 days to reach the most , conservative projected start of RFO-5. However, this instrumentation is l required in Operational Conditions 4 and 5 during periods of core alterations or operations with a potential for draining the reactor vessel; therefore, extension of the surveillance interval for a nominal period of 137 days is required to reach the most conservative projected end of the refueling outage. l In February 1^90, Rosemount published a report, "30 Month Stability Specificatio- ar Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount R rt D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated .ogust 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from C.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified.
ENCLOSURE 23 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.2, TABLE 4.3.2.1-1, ITEM 1.d REACTOR VESSEL VATER LEVEL - LOV, LEVEL 1 PRIMARY CONTAINMENT ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.2 requires a LSFT and simulated automatic actuation of all channels of the Isolation Actuation System at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel Vater Level - Lov, Level 1 functional unit for Primary Containment isolation (TS Table 4.3.2.1-1, Item 1.d) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 44 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilitiei of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 44 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
i l I l ENCLOSURE 24 ) JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.2, TABLE 4.3.2.1-1, ITEM 1.e MANUAL INITIATION PRIMARY CONTAINMENT ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.2 requires a LSFT and simulated automatic actuation of all channels of the Isolation Actuation System at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Manual Initiation functional unit for Primary Containment isolation (TS Table 4.3.2.1-1, Item 1.e) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 48 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 48 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. I l I 1
ENCLOSURE 25 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.2.1, TABLE 4.3.2.1-1, ITEM 2.a REACTOR VESSEL VATER LEVEL- LOV, LEVEL 1 CALIBRATION ISOLATION ACTUATION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.2.1, Table 4.3.2.1-1, Item 2.a requires the Isola ion Actuation System Reactor Vessel Vater Level - Lov, Level 1 Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.2.1-1, Item 2.a for a nominal period of 47 days to reach the most conservative projected start of RFO-5. However, this instrumentation is required in Operational Conditions 4 and 5 during periods of core alterations or operations with a potential for draining the reactor vessel; therefore, extension of the surveillance interval for a nominal period of 139 days is required to reach the most conservative projected end of the refueling outage. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance froc 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month , calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters ! and trip unit channels are bounding. There is adequate allovance in the I calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified.
i ENCLOSURE 26 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.2, TABLE 4.3.2.1-1, ITEM 2.a REACTOR VESSEL VATER LEVEL - LOV, LEVEL 1 MAIN STEAM LINE ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.2 requires a LSFT and simulated automatic ^ actuation of all channels of the Isolation Actuation System at least once per 18 months (with 3 maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel Water Level - Lov, Level 1 functional unit for Main Steam Line isolation (TS Table 4.3.2.1-1, Item 2.a) requires a surveillance interval extension for this functional unit's portion of the'LSFT for a nominal period of 47 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance intetival extension for a nominal period of 47 days is bounded by the interval accepted on the Peach Bottom cocket; therefore, the surveillance interval extension is justified.
1 l I ENCLOSURE 27 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 2.a REACTOR VESSEL VATER LEVEL - LOV, LEVEL 1 MAIN STEAM LINE ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation Instrumentation shown on Table 3.3.2-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to 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 reactor trip system. Trip function Item 2.a, Main Steam Line Isolation, Reactor Vessel Vater Level - Lov, Level 1, of Table 3.3.2-3 vill become overdue prior to the beginning of RFO-5 conservatively scheduled to begin February 15, 1996. This RPS response time test requires an extension for a nominal period of 39 days to reach the most conservative projected start of RFO-5. The extension vould have no substantial measurable effect on plant safety because
a. There are redundant sensors that can initiate the scram operation,

b. One-out-of-two redundancy exists in every individual channel within each trip function.

c. The instrumentation failure probability is a very small fraction of the total control rod insertion (scram failure) probability.

d. There are several redundant and diverse instrument channels which can detect and generate a scram signal (e.g., flux, pressure, etc.).

e. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure.
Extension of the Peach Bottom Atomic Power Station, Units 2 and 3 surveillance intervals for RPS response time testing from 18 to 24 months was accepted by the NRC in the Safety Evaluation Report dated August 2, 1993. The justification cited above is very similar to that provided in the NRC Safety Evaluation Report which states:
"The RPS system consists of two independent trip systems with at least ,
two subchannels of a parameter per trip system. The logic of the RPS ! system is such that either subchannel can trip a trip system and that i both trip systems must trip to cause a reactor trip. The logic is such that a single failure vill neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent redundancy in ! the RPS system, the impact of extending the response time surveillance : interval on system availability is small." Based on the above, a one-time extension of the RPS Instrumentation, TS Table I 3.3.2-3, Item 2.a response time testing surveillance interval is justified.
i ENCLOSURE 28 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.2.1, TABLE 4.3.2.1-1, ITEM 2.b HAIN STEAM LINE RADIATION - HIGH CALIBRATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.1, Table 4.3.2.1-1, Item 2.b requires the Isolation Actuation Instrumentation, Main Steam Line Radiation - High trip function of the Main Steam Line Radiation Monitors to be calibrated at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2) by performing a channel calibration. The Main Steam Line Radiation - High trip function, which trip the mechanical vacuum pump isolation valves, is required to be OPERABLE in Operational Conditions 1 and 2 when the mechanical l vacuum pump lines are not isolated and requires extension of the surveillance interval of for a nominal period of 32 days to reach the most. conservative projected start date for RFO-5. The mechanical vacuum pump is only utilized in Operational Conditiog 2 during l plant startup. In that this Operational Condition is only experienced for a very small period of time during the operating cycle and vill noc be experienced unless there is a plant shutdown, extension of the interval to an approximate period of 24 months vould not adversely affect safety. The calibration data from the past refueling outages has shown that the additional drift which may be expected by the monitors for this surveillance interval extension period should not exceed the setpoint by more than 9%. Calibration of the monitors vill be performed prior to the expiration of the interval if a plant shutdown is experienced during this cycle and, thus, there vould be no concern. Therefore, based on the expected small amount of drift above the setpoint, the very limited period of time that the trip provided by this monitor would be needed, and the fact that if a plant shutdown was to occur the monitors vould be calibrated, the extension of the surveillance interval for a nominal period of 32 days to reach the most conservative start date for RFO-5 is justified. 1
I i i ENCLOSURE 29 l JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR ; TECHNICAL SPECIFICATION SR 4.3.2.2, TABLE 4.3.2.1-1, ITEM 2.b HAIN STEAM LINE RADIATION - HIGH MAIN STEAM LINE ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.2 requires a LSFT and simulated automatic actuation of all channels of the Isolation Actuation System at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Main Steam Line Radiation - High functional unit for Main Steam Line Isolation (TS Table 4.3.2.1-1, Item 2.b) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 32 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 32 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. l
f ENCLOSURE 30 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 2.c MAIN STEAM LINE PRESSURE - LOW MAIN STEAM LINE ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation Instrumentation shown on Table 3.3.2-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to 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 reactor trip system. Trip function Item 2.c, Main Steam Line Isolation, Main Steam Line Pressure - Lov, of Table 3.3.2-3 vill become overdue prior to the beginning of RFO-5 conservatively scheduled to begin February 15, 1996. This RPS response time test requires an extension for a nominal period of 36 days to reach the most conservative projected start of RFO-5. The extension would have no substantial measurable effect on plant safety because:
a. There are redundant sensors that can initiate the scram operation.

b. One-out-of-two redundancy exists in every individual channel within each trip function.

c. The instrumentation failure probability is a very small fraction of the total control rod insertion (scram failure) probability.

d. There are several redundant and diverse instrument channels which can detect and generate a scram signal (e.g., flux, pressure, etc.).

e. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure.
Extension of the Peach Bottom Atomic Power Station, Units 2 and 3 surveillance intervals for RPS response time testing from lf to 24 months was accepted by the NRC in the Safety Evaluation Report dated August 2, 1993. The justification cited above is very similar to that provided in the NRC Safety Evaluation Report which states:
"The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS system is such that either subchannel can trip a trip system and that both trip systems must trip to cause a reactor trip. The logic is such that a single failure vill neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent redundancy in the RPS system, the impact of extending the response time surveillance interval on system availability is small."
Based on the above, a one-time extension of the RPS Instrumentation, TS Table 3.3.2-3, Item 2.c, response time testing survaillance interval is justified.
i 1 l ENCLOSURE 31 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.3, TABLE 3.3.2-3, ITEM 2.d MAIN STEAM LINE FLOV - HIGH MAIN STEAM LINE ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.3 requires the Response Time of the Isolation Actuation Instrumentation shown on Table 3.3.2-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to 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 reactor trip system. Trip function Item 2.d, Main Steam Line Isolation, Main Steam Line Flow - High, of Table 3.3.2-3 vill become overdue prior to the beginning of RFO-5, conservatively scheduled to begin February 15, 1996. This RPS response time test requires an extension for a nominal period of 36 days to reach the most conservative projected start of RFO-5. The extension would have no substantial measurable effect on plant safety because:
a. There are redundant sensors that can initiate the scram operation.

b. One-out-of-two redundancy exists in every individual channel within each trip function.

c. The instrumentation failure probability is a very small fraction of the total control rod insertion (scram failure) probability.

d. There are several redundant and diverse instrument channels which can detect and generate a scram signal (e.g., flux, pressure, etc.).

e. The failure of instrumentation in the sluggish responding mode is a small fraction of its overall failure.
Extension of the Peach Bottom Atomic Power Station, Units 2 and 3 surveillance intervals for RPS response t!me testing from 18 to 24 months was accepted by the < I NRC in the Safety Evaluation Report dated August 2, 1993. The justification cited above is very similar to that provided in the NRC Safety Evaluation Report which states:
"The RPS system consists of two independent trip systems with at least two subchannels of a parameter per trip system. The logic of the RPS ;
system is such that either subchannel can trip a trip system and that ; both trip systems must trip to cause a reactor trip. The logic is such that a single failure vill neither cause nor prevent a required reactor scram. The licensee states that, based on the inherent redundancy in the RPS system, the impact of extending the response time surveillance 1 interval on system availability is small." l Based on the above, a one-time extension of the RPS Instrumentation, TS Table l 3.3.2-3, Item 2.d, response time testing surveillance interval is justified. 1
ENCLOSURE 32 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.2.1, TABLE 4.3.2.1-1, ITEM 3.a REACTOR VESSEL VATER LEVEL- LOV, LEVEL 2 CALIBRATION ISOLATION ACTUATION SYSTEM INSTRUMENTATION 4 Technical Specification SR 4.3.2.1, Table 4.3.2.1-1, Item 3.a requires the Isolation Actuation System Reactor Vessel Water Level - Lov, Level 2 Instrunentation to be demonstrated OPERABLE by performance of a channel , calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table , 4.3.2.1-1, Item 3.a for a nominal period of 47 days to reach the most conservative projected start of RFO-5. However, this instrumentation is ' required in Operational Conditions 4 and 5 during periods of core alterations or operations with a potential for draining the reactor vessel; therefore, extension of the surveillance interval for a nominal period of 139 days is required to reach the most conservative projected end of the refueling outage. In February 1990, Rosemount published a report, "30 Month Stability ' Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the , requested extension is justified. 1 l 1 w n
l ENCLOSURE 33~ i JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.2.1, TABLE 4.3.2.1-1, ITEM 4.e. REACTOR VESSEL VATER LEVEL- LOV, LEVEL 2 CALIBRATION ISOLATION ACTUATION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.2.1, Table 4.3.2.1-1, Item 4.e requires the Isolation Actuation System Reactor Vessel Water Level - Lov, Level 2 Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). The level-transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.2.1-1, Item 4.e for a nominal period of 47 days to reach the most-conservative projected start of RFO-5. However, this instrumentation is required in Operational Conditions 4 and 5 during periods of core alterations or operations with a potential for draining the reactor vessel; therefore, extension of the surveillance interval for a nominal period of 139 days is required to reach the most conservative projected end of the refueling outage. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the i drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. t 1 1 1 _I
1 ENCLOSURE 34 l JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR ! TECHNICAL SPECIFICATION SR 4.3.2.2, TABLE 4.3.2.1-1, ITEM 5.m MANUAL INITIATION REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.2 requires a LSFT and simulated automatic actuation of all channels of the Isolation Actuation System at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Manual Initiation functional unit for Reactor Core Isolation Cooling System Isolation (TS Table 4.3.2.1-1, Item 5.m) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 38 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 38 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
l 1
-l f
i ENCLOSURE 35 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION 1
-TECHNICAL SPECIFICATION SR 4.3.2.1, TABLE 4.3.2.1--1, ITEM 6.d REACTOR VESSEL VATER LEVEL - LOV, LEVEL 3 CALIBRATION ISOLATION ACTUATION SYSTEM INSTRUMENTATION Technical Specification SR 4.3.2.1, Table 4.3.1.2-1, Item 6.d requires the i Isolation Actuation System Reactor Vessel Vater Level- Lov, Level 3 Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). One channel of the instrumentation (channel D),
containing a Rosemount Model 1153 pressure transmitter, vill require an extension of the SR interval cited in TS Table 4.3.2.1-1, Item 6.d for a nominal period of 8 days to reach the most conservative projected start of RFO-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3' docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the , calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. l
l ENCLOSURE 36 l JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.2, TABLE 4.3.2.1-1, ITEM 6.d REACTOR VESSEL VATER LEVEL - LOV, LEVEL 3 RHR SYSTEM ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.2 requires a LSFT and simulated automatic actuation of all channels of the Isolation Actuation System at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel Water Level - Lov, Level 3 functional unit for RHR System Isolation (TS Table 4.3.2.1-1, Item 6.d) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 8 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 8 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
ENCLOSURE 37 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.2.1, TABLE 4.3.2.1-1, ITEM 6.e ' REACTOR VESSEL (RHR CUT-IN PERMISSIVE) PRESSURE - HIGH CALIBRATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.1, Table 4.3.2.1-1, Item 6.e requires the Isolation Actuation System Reactor Vessel (RHR Cut-In Permissive) Pressure - High Instrumentation to be demonstrated OPERABLE by-performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). Two channels of the instrumentation (channels A and C),'containing Rosemount Model 1153 pressure transmitters, vill require an extension of the SR interval cited in TS Table 4.3.2.1-1, Item 6.e for a nominal period of 40 days to reach the most conservative projected start of RFO-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The existing PNPP setpoint calculations for the trip units for Rosemount transmitters and associated trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). ' Therefore, the requested extension is justified. l
ENCLOSURE 38 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.2, TABLE 4.3.2.1-1, ITEM 6.e REACTOR VESSEL (RHR CUT-IN) PRESSURE - HIGH RHR SYSTEM ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.2 requires a LSFT and simulated automatic actuation of all channels of the Isolation Actuation System at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel (RHR Cut-In) Pressure - High functional unit (TS Table 4.3.2.1-1, Item 6.e) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 40 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 40 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
ENCLOSURE 39' JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.2.2, TABLE 4.3.2.1-1, ITEM 6.f- l DRYWELL PRESSURE - HIGH RHR SYSTEM ISOLATION ISOLATION ACTUATION INSTRUMENTATION , Technical Specification SR 4.3.2.2 requires a LSFT and simulated automatic actuation of all channels of the Isolation Actuation System at least once per 18 months (with a maximum allovable surveillance interval extension of-4.5 months per TS 4.0.2). The Dryvell Pressure - High functional unit for RHR System Isolation (TS Table 4.3.2.1-1, Item 6.f) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of ' 5 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BWRs), i prepared by the BVR Ovners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps '
and valves), which are consequently tested on a more frequent basis...Since the probability of-a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant , change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 5 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. . l l l
\
ENCLOSURE 40 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR
-TECHNICAL SPECIFICATION SR 4.3.2.2, TABLE 4.3.2.1-1, ITEM 6.g MANUAL INITIATION RHR SYSTEM ISOLATION ISOLATION ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.2 requires a LSFT and simulated automatic actuation of all channels of the Isolation Actuation System at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months
.per TS 4.0.2). The Manual Initiation functional unit for RHR System Isolation (TS Table 4.3.2.1-1, Item 6.g) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 5 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BUR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components,_(e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 5 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
l i ENCLOSURE 41 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEM A.I.a REACTOR VESSEL VATER LEVEL- LOV, LEVEL 1 CALIBRATION DIVISION 1 TRIP SYSTEM - RHR-A (LPCI MODE) AND LPCS SYSTEM EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.1, Table.4.3.3.1-1, Item A.1.a requires the Emergency Core Cooling System Reactor Vessel Vater Level - Lov, Level 1 Instrumentation for the RHR-A (LPCI Mode) and LPCS System, Division 1 Trip System to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.3.1-1, Item A.1.a for a nominal period of 44 days to reach the most conservative projected start date for RFO-5. However, this instrumentation is required to be OPERABLE in Operational Conditions 4 and 5 as required by TS 3.5.2; therefore, extension of the surveillance interval for a nominal period of 159 days is required to reach the most conservative projected end of RFO-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. 1 1
! ENCLOSURE 42 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.2, TABLE 4.3.3.1-1, ITEM A.1.a REACTOR VESSEL VATER LEVEL - LOW, LEVEL 1 i ) DIVISION 1 TRIP SYSTEM - RHR-A (LPCI MODE) AND LPCS SYSTEM EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.2 requires a LSFT and simulated automatic actuation of all channels of the Emergency Core Cooling System at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel Vater Level - Lov, Level 1 functional unit for RHR-A (LPCI Mode) and LPCS System System, Division 1 Trip System (TS Table 4.3.3.1-1, Item A.1.a) requires a surveillance interval extension for this ; functional unit's portion of the LSFT for a nominal period of 28 days to reach j the most conservative projected start of RFO-5. However, this instrumentation I is required to be OPERABLE in Operational Conditions 4 and 5 as required by TS 3.5.2; therefore, extension of the surveillance interval for a nominal period of 1 120 days is required to reach the most conservative projected end of RFO-5. ] As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months: i
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Ovners Group (NEDC-30936P) show that the overall . safety systems' reliabilities are not dominated by the reliabilities of I the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of meschanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 120 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
l ENCLOSURE 43 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.3, TABLE 3.3.3-3, ITEM A.1.a REACTOR VESSEL VATER LEVEL - LOV, LEVEL 1 DIVISION 1 TRIP SYSTEM - RHR-A'(LPCI H0DE) AND LPCS SYSTEM EMERGENCY CORE COOLING SYSTEM INSTRUMENTATION Technical Specification SR 4.3.3.3 requires the Response Time of the Emergency Core Cooling System Instrumentation shown on Table 3.3.3-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to 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 reactor trip system. Trip function Item A.1.a, Division 1 Trip System - RHR-A (LPCI Mode) and LPCS System, Reactor Vessel Vater Level - Lov, Level 1, of Table 3.3.3-3 vill become overdue prior to the beginning of RFO-5 conservatively scheduled to begin February 15, 1996. This RPS response time test requires an extension for a nominal period of 50 days to reach the most conservative projected start of RF0-5. However, this instrumentation is required to be OPERABLE in Operational Conditions 4 and 5 for Emergency Core Cooling System per TS 3.5.2 and 3.5.3; therefore, extension of the surveillance interval for a nominal period of 142 days is required to reach the most conservative projected end of the refueling outage. Regulatory Guide 1.118 (Revision 2) states:
" Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is verified by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstrated that the changes in response time beyond acceptable limits are accompanied by changes in performance characteristics which are detectable during routine tests."
On January 14, 1994, the BVR Owner's Group submitted a Licensing Topical Report (LTR) prepared by the General Electric Company, NED0-32291, " System Analyses For Elimination Of Selected Response Time Testing Requirements", January 1994, for NRC review. The information contained therein justifies a one-time extension of the surveillance requirement interval of TS Table 3.2.3-3, Item A.1.a. The LTR provided justification for the elimination of selected BWR Response Time Testing or Tests (RTT), as defined in the Instrument Society of America (ISA) Standard S67.06, from the plant TS SRs. The analyses included the affected instrumentation loops which could potentially impact the instrument loop response time. In addition, plant operating experiences were reviewed to identify response time failures and how they were detected. The failure modes identified vere then evaluated to determine if the effect on response time would be detected by other testing requirements contained in the TS. Based on the analyses presented in the LTR, it was concluded that there were no failure modes which will affect the response time of the instrumentation loop which vould not be detected by other surveillances such as channel calibration (at least once per 31 days for trip setpoint calibration and at least once per 18 months for sensor calibration), channel functional tests (at least once per 31 days), channel checks (at least once per 12 hours), or other techniques. [It should be
.. - . . .~ _ . . .- ._- _. . . .
i 1 6 noted that the "other techniques" phrase applies to Rosemount transmitter slow , oil loss determination.] j In addition, individual instrument channel response time delays for specific trip functions (on the order of a fraction of second) have very~little safety significance. Redundancy and diversity exits in most instrumentation trip , functions (e.g., flux, water level, pressure). For the Emergency Core Cooling System, the instrumentation response times are a small fraction of the overall , response times of the actuating devices. . Based cn1 the above, a one-time extension of the Emergency Core Cooling System Actuation Instrumentation TS SR 4.3.3.3 Table 3.3.3-3, Item A.l.a response time testing surveillance interval is justified. I l l l _,_l
ENCLOSURE 44 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.2, TABLE 4.3.3.1-1, ITEM A.1.b DRYVELL PRESSURE - HIGH DIVISION 1 TRIP SYSTEM - RHR-A (LPCI MODE) AND LPCS SYSTEM EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.2 requires a LSFT and simulated automatic actuation of all channels of the Emergency Core Cooling System at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Dryvell Pressure - High functional unit for RHR-A (LPCI Mode) and LPCS System System, Division 1 Trip System (TS Table 4.3.3.1-1, Item A.1.b) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 28 days to reach the most conservative projected start of RF0-5. However, this instrumentation is required to be OPERABLE in Operational Conditions 4 and 5 as required by TS 3.5.2; therefore, extension of the surveillance interval for a nominal period of 120 days is required to reach the most conservative projected end of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 120 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
,. . . - - - - .. - . . - . - - . - . ,~ . . . . ,
r ENCLOSURE'45' JUSTIFICATION FOR SURVEILLANCE INTERVAL' EXTENSION- l TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEM A.2.a REACTOR VESSEL VATER LEVEL- LOV, LEVEL 1 l CALIBRATION- ;
-DIVISION 1 AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM 'A' ,
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.1,' Table 4.3.3.1-1,; Item A.2.A requires the 'f
' Emergency Core Cooling Systen Reactor Vessel' Vater Level - Lov, Level 1-Instrumentation for the Division 1 Automatic'Depressurization System. Trip System 'A' to be demonstrated OPERABLE by performance of a channel calibration at least j
once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). .The level transmitters, Rosemount Model 1153 level transmitters, vill ! i require an extension of the.SR interval cited.in TS TableL4.3.3.1-1, Item'A.2.a for a nominal period of 45 days to reach the most conservative projected start' of RFO-5. ; In February 1990, Rosemount published a report, "30 Month Stability ; Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900125, Revision A) [ accepted by NRC Safety Evaluation - ! Report dated August 2, 1993 on the Peach Bottom Atomic Power Station,. Units 2 ' and 3 docket.) This report supported the extension of.the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the: ' drift allowance from 0.29% URL (2 sigma) for 18' months to 0.20% URLi(2 sigma) for 30 months. In addition, applicable setpoint. calculations assumed'18; month " calibration of the trip interval for trip units. However, the trip units are-calibrated either monthly or quarterly, depending on the TS requirement for-
channel functional testing. The setpoint calculations for'the trip units ,
utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift ; value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the ; calculations'for 30 month transmitter. drift and no potentia 1' impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore,.the , requested extension is justified. t
.i f
i e f I 1
. . . . - . -- - . ~ . ~ . - . . . - , . . - . . .( +
1 q i ENCLOSURE 46 : JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION ~ LOGIC SYSTEM FUNCTIONAL TESTING FOR :!
- TECHNICAL SPECIFICATION SR 4.3.3.2, TABLE 4.3.3.1-1, ITEM A.2.a . -,
REACTOR VESSELLVATER LEVEL - LOV, LEVEL 1 "j DIVISION 1 AUTOMATIC DEPRESSURIZATION SYSTEM TRIP-SYSTEM 'A' !
'i EMERGENCY CORE COOLING SYSTEM ACTUATION' INSTRUMENTATION 7 '
Technical Specification SR 4.3.3.2 requires a LSFT andLsimulated automatic actuation of all channels of the Emergency Core Cooling System at least once per s i
.18 months (with a maximum allowable surveillance interval extension of 4.5 i months'per TS 4.0.2). The Reactor Vessel Water Level - Lov, Level.1 functional unit for the Division 1 Automatic Depressurization System Trip System 'A(TS' l Table 4.3.3.1-1, Item A.2.a) requires a surveillance interval extension-for this !
functional unit's portion of the LSFT for a nominal period;of 40 days to reach the most conservative projected start of RFO-5. l As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to ' extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, t surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling' vater reactors (BVRs), ' ~!
prepared by the BVR'0wners Group (NEDC-30936P) show that the overall ! safety systems' reliabilities are not dominated by the reliabilities of the logic system .but by that of the mechanical components,-(e.g., pumps ; and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact. failure is'small' .) relative to the probability of mechanical component failure, increasing : the logic system functional test interval represents no significant ' change.in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval ~ f
. extension for a nominal period of 40 days is bounded by the interval accepted on f the Peach Bottom docket; therefore,'the surveillance interval-extension is ;
justified. i h t
')
l t
- .r - - - - .,m.. . , - , _ _ _ _ _ _ _ _ _ _
ENCLOSURE 47 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEM A.2.d REACTOR VESSEL VATER LEVEL - LOV, LEVEL 3 (PERMISSIVE) CALIBRATION DIVISION 1 AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM 'A' EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.1, Table 4.3.3.1-1 Item A.2.d requires the Emergency Core Cooling System Reactor Vessel Vater Level - Lov, Level 3 (Permissive) Instrumentation for the Division 1 Automatic Depressurization System Trip System 'A' to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.3.1-1, Item A.2.d for a nominal period of 32 days to reach the most conservative projected start of RFO-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculatione for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changer are required). Therefore, the requested extension is justified. 1 1
7, : l ENCLOSURE'48 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.2, TABLE 4.3.3.1-1, ITEM A.2.d .! REACTOR VESSEL VATER LEVEL - LOV, LEVEL'3 (PERMISSIVE) DIVISION 1 AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM 'A'- .i EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION i Technical Specification SR 4.3.3.2 requires a LSFT and simulated' automatic ; actuation of all channels of the Emergency Core Cooling System at least once-per-18 months (with a maximum allovable. surveillance interval extension of 4.5 months per TS 4.0.2).; The Reactor Vessel Water Level - Lov, Level 3 (Fermissive) functional unit for the Division 1 Automatic Depressurization System Trip System 'A' (TS Table 4.3.3.1-1,. Item A.2.d) requires.a surveillance interval extension for this: functional. unit's portion of the LSFT for a nominali period of 30 days to reach the most conservative projected start of RFO-5. , As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to , extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance' intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors:(BVRs),
prepared by the BWR Owners Group (NEDC-30936P)'show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently. tested on a more frequent basis...Since the probability of a relay or contact ~ failure is small- ' relative to the_ probability of mechanical component failure, increasing ' the logic system functional test interval represents no significant-change in the overall safety system unavailability."
~
t The evaluation above is applicable to PNPP and the surveillance' interval, extension _ for a nominal period of 30 days 'is bounded by the interval accepted Ani the Peach Bottom docket; therefore, the surveillance interval extension is justified. I 9 4 a
\
l 1 l
r ENCLOSURE 49 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION l LOGIC SYSTEM FUNCTIONAL TESTING FOR j TECHNICAL SPECIFICATION SR 4.3.3.2, TABLE 4.3.3.1-1, ITEMS A.2.g AND B.2.f l AUTOMATIC DEPRESSURIZATION SYSTEM 'A' AND 'B' TRIP SYSTEMS l MANUAL INITIATION j EMERGENCY CORE COOLING SYSTEM INSTRUMENTATION l l Technical Specification SR 4.3.3.2 requires a LSFT and simulated automatic ! actuation of all channels of the Emergency Core Cooling System at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Automatic Depressurization System 'A' and 'B' Trip j Systems manual initiation functional units (TS Table 4.3.3.1-1, Items A.2.g and i ! B.2.f) require surveillance interval extensions for these functional units' f portion of the LSFT for a nominal period of 22 days to reach the most ( conservative projected start of RFO-5. l As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, l surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Ovners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of rechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 22 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
l ENCLOSURE 50. JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEM B.1.a REACTOR VESSEL VATER LEVEL- LOV, LEVEL 1 CALIBRATION DIVISION 2 RHR-B AND C (LPCI MODE) EMERGENCY CORE COOLING SYSTEM ACTUJ. TION INSTRUMENTATION Technical Specification SR 4.3.2.1, Table 4.3.3.1-1, Item B.1.a requires the Emergency Core Cooling System Reactor Vessel Vater Level - Lov, Level 1 Instrumentation for the Division 2 RHR-B and C (LPCI Mode) to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allovable entension of 4.5 months per TS 4.0.2). The level - transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.3.1-1, Item B.1.a for a nominal period - of 44 days to reach the most conservative projected start date for RFO-5. i However, this instrumentation is required to be OPERABLE in Operational Conditions 4 and 5 as required by TS 3.5.2; therefore, extension of the ; surveillance interval for a nominal period of 159 days is required to reach the i most conservative projected end of RFO-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" ! (Rosemount Report D8900126, Revision A) (accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval ; for the transmitters from 18 months to 30 months based on a reduction in the drif t allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. l
ENCLOSURE 51 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.2, TABLE 4.3.3.1-1, ITEM B.1.a REACTOR VESSEL VATER LEVEL - LOV, LEVEL 1 DIVISION 2 RHR B AND C (LPCI MODE) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.2 requires a LSFT and simulated automatic actuation of all channels of the Emergency Core Cooling System Actuation Instrumentation at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel Vater Level - Lov, Level 1 functional unit for the Division 2 RHR B and C (LPCI Mode) (TS Table 4.3.3.1-1, Item B.1.a) requires a surveillance interval extension for a nominal period of 44 days for these functional units' portion of the LSFT to reach the projected start of RFO-5. However, these functional units are required to be OPERABLE in Operational Conditions 4 and 5 as required by TS 3.5.2; therefore, surveillance interval extensions are required for a nominal period of 159 days to reach the most conservative projected end of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 159 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
ENCLOSURE 53 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.3, TABLE 3.3.3-3, ITEM B.1.a REACTOR VESSEL VATER LEVEL - LOV, LEVEL 1 DIVISION 2 TRIP SYSTEM - RHR-B AND C (LPCI MODE) AND LPCS SYSTEM EMERGENCY CORE COOLING SYSTEM INSTRUMENTATION Technical Specification SR 4.3.3.3 requires the Response Time of the Emergency Core Cooling System Instrumentation shown on Table 3.3.3-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to 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 reactor trip system. Trip function Item B.1.a, Division 2 Trip System - RHR-B and C (LPCI Mode), Dryvell Pressure - High, of TS Table 3.3.3-3 vill become overdue prior to the beginning of RFO-5 conservatively scheduled to begin February 15, 1996. This RPS response time test does not require an extension to reach the projected start of RFO-5. However, this instrumentation is required to be OPERABLE in Operational Conditions 4 and 5 for Emergency Core Cooling System per TS 3.5.2 and 3.5.3; therefore, extension of the surveillance interval for a nominal period of 21 days is required to reach the most conservative projected end of the refueling outage. Regulatory Guide 1.118 (Revision 2) states:
" Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is verified by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstrated that the changes in response time beyond acceptable limits are accompanied by changes in performance characteristics which are detectable during routine tests."
On January 14, 1994, the BVR Owner's Group submitted a Licensing Topical Report (LTR) prepared by the General Electric Company, NED0-32291, " System Analyses For Elimination Of Selected Response Time Testing Requirements", January 1994, for NRC review. The information contained therein justifies a one-time extension of the surveillance requirement interval of TS Table 3.2.3-3, Item B.1.a. The LTR provided justification for the elimination of selected BVR Response Time Testing or Tests (RTT), as defined in the Instrument Society of America (ISA) Standard 567.06, from the plant TS SRs. The analyses included the affected instrumentation loops which could potentially impact the instrument loop response time. In addition, plant operating experiences were reviewed to identify response time failures and how they were detected. The failure moder identified vere then evaluated to determine if the effect on response time vould be detected by other testing requirements contained in the TS. Based on the analyses presented in the LTR, it was concluded that there were no failure modes , which vill affect the response time of the instrumentation loop which would not l be detected by other surveillances such as channel calibration (at least once per 31 days for trip setpoint calibration and at least once per 18 months for , sensor calibration), channel functional tests (at least once per 31 days), f channel checks (at least once per 12 hours), or other techniques. [It should be ! noted that the "other techniques" phrase applies to Rosemount transmitter slow
]
i oil loss determination.] i t In addition, individual instrument channel _ response time delays.for specific- t trip functions (on the order of a fraction of second) have very.little safety i
" significance. Redundancy and diversity exits in most instrumentation trip functions (e.g., flux, water level, pressure). -For the Emergency Core Cooling System,,the instrumentation response times are a small fraction of the overall '; -response times of the actuating devices.
Based on the above, a one-time extension of the Emergency Core Cooling System
~
Actuation Instrumentation TS SR 4.3.3.3, Table 3.3.3-3, Item B.1.a response time testing surveillance interval is justified. [ [ i
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t P i l i I t 1 1 I i
[ ENCLOSURE 53 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC-SYSTEM ~ FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.2, TABLE 4.3.3.1-1, ITEM B.l.b l DRYVELL PRESSURE - HIGH DIVISION 2 RRR B AND C (LPCI MODE) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.2 requires a LSFT and simulated automatic actuation of all channels of the Emergency Core Cooling System Actuation 1 Instrumentation at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Dryvell Pressure - High functional unit for the Division 2 RHR B and C (LPCI Mode) (TS Table 4.3.3.1-1, Item B.l.b) requires a surveillance interval extension for a nominal period of 25 days for these functional units' portion of the LSFT to reach the projected start of RFO-5. However, these functional units are required to be OPERABLE in Operational Conditions 4.and 5 as required by TS 3.5.2; therefore, surveillance interval extensions are required for a nominal period of 117 days to reach the most conservative projected end of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 117 days is bounded by the interval accepted , on the Peach Bottom docket; therefore, the surveillance interval extension is justified. ! 1
)
l l
ENCLOSURE 54 l JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEM B.2.a REACTOR VESSEL VATER LEVEL- LOV, LEVEL 1 CALIBRATION DIVISION 2 AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM 'B' l EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.2.1, Table 4.3.3.1-1, Item B.2.a requires the Emergency Core Cooling System Reactor Vessel Vater Level - Lov, Level 1 Instrumentation for the Division 2 Automatic Depressurization System Trip System 'B' to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.3.1-1, Item B.2.a for a nominal period of 44 days to reach the most conservative projected start date for RF0-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. i 1
ENCLOSURE 55 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.2, TABLE 4.3.3.1-1, ITEM B.2.a REACTOR VESSEL VATER LEVEL - LOV, LEVEL 1 DIVISION 2 AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Speci:ication SR 4.3.3.2 requires a LSFT and simulated automatic actuation of all Sannels of the Emergency Core Cooling System at least once per 18 months (with a : <imum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel Vater Level - Lov, Level 1 functional unit for the Division 2 Automatic Depressurization System Trip System (TS Table 4.3.3.1-1, Item B.2.a) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 40 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 40 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified, i I l l i
1 ENCLOSURE 56 JUSTIFICATION FOR SURVEILLANCE. INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEM B.2.d REACTOR VESSEL VATER LEVEL - LOV, LEVEL 3 (PERMISSIVE) CALIBRATION j DIVISION 2 AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM 'B' EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.1, Table 4.3.3.1-1, Item B.2.d requires the. Division 2 Automatic Depressurization System Trip System 'B' Reactor Vessel- - Vater Level - Lov, Level 3 (Permissive) Instrumentation to be demonstrated OPERABLE by performance of.a channel calibration at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.3.1-1, Item B.2.d for a nominal period of 32 days to reach the most conservative projected start of RFO-5. In February 1990, Rosemount published a report, "30 Month Stability ' Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.) This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL'(2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly'or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels aru bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required)'. Therefore, the requested extension is justified. l l l
ENCLOSURE 57 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION
-LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.2,. TABLE 4.3.3.1-1, ITEM B.2.d-REACTOR VESSEL VATER LEVEL - LOV, LEVEL 3 (PERMISSIVE)
DIVISION 2 AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM 'B' EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.2 requires a LSFT and simulated automatic actuation of all channels of-the Emergency Core Cooling System at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel Vater Level - Lov, Level 3' (Permissive) functional unit for the Division 2 Automatic Depressurization ' System Trip System 'B' (TS Table 4.3.3.1-1, Item B.2.d) requires a surveillance interval extension for this functional unit's portion'of the LSFT for a nominal period of 31 days to reach the most conservative projected start of RFO-5. i As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs), '
prepared by the BVR Ovners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not' dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." , The evaluation above is applicable to PNPP and the surveillance interval ' extension for a nominal period of 31 days is bounded by the interval accepted on-the Peach Bottom dockets therefore, the surveillance interval extension is justified. i l 1
ENCLOSURE 58 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEM C.1.a ' REACTOR VESSEL VATER LEVEL - LOV, LEVEL 2 CALIBRATION DIVISION 2 AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM 'B' EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.1, Table 4.3.3.1-1, Item C.1.a requires the Division 2 Automatic Depressurization System Trip System 'B' Reactor Vessel Vater Level - Lov, Level 2 Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum-allovable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.3.1-1, Item C.1.a for a nominal period of 18 days to reach the most conservative projected start of RFO-5. However, this instrumentation is required in Operational Conditions 4 and 5 during periods of core alterations or operations with a potential for draining the reactor vessel; therefore, extension of the surveillance interval for a nominal period of 110 days is required to reach the most conservative projected end of the refueling outage. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.) This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending.on the TS requirement for ' channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified.
ENCLOSURE 59 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.2, TABLE 4.3.3.1-1, ITEMS C.1.a AND C.1.c REACTOR VESSEL VATER LEVEL - LOV, LEVEL 2, AND HIGH, LEVEL 8 DIVISION 3 TRIP SYSTEM - HPCS SYSTEM EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.2 requires a LSFT and simulated automatic actuation of all channels of the Emergency Core Cooling System Actuation Instrumentation at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel Water Level - Lov, Level 2 and Reactor Vessel Vater Level - High, Level 8 functional units for the Division 3 Trip System - HPCS System (TS Table 4.3.3.1-1, Items C.1.a and C.1.c) require surveillance interval extensions for a nominal period of 18 days for these functional units' portion of the LSFT to i reach the projected start of RFO-5. However, these functional units are required to be OPERABLE in Operational Conditions 4 and 5 as required by TS 3.5.2; therefore, surveillance interval extensions are required for a nominal ; period of 110 days to reach the most conservative projected end of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extensions for a nominal period of 110 days are bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified.
o q r ENCLOSURE 60 1 l JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING.FOR l TECHNICAL SPECIFICATION SR 4.3.3.3, TABLE 3.3.3-3, ITEM C.l.a ! REACTOR VESSEL VATER LEVEL - LOV, LEVEL 2 I DIVISION 3 TRIP SYSTEM - HPCS SYSTEM ,
-EMERGENCY CORE COOLING SYSTEM INSTRUMENTATION Technical Specification SR 4.3.3.3 requires the Response Time of the Emergency Core Cooling System Instrumentation shown on Table 3.3.3-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to 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 reactor trip system. Trip function Item C.1.a Division 3 Trip System - HPCS System, Reactor Vessel Vater Level - Lov, Level 2, of Table 3.3.3-3 vill become overdue prior to the beginning of RFO-5 conservatively scheduled to begin February 15, 1996. This RPS response time test requires an extensioTi for a nominal period of 47 days to reach the most conservative
projected start of RFO-5. Howevet, this instrumentation is required to be
OPERABLE in Operational Conditions 4 and 5 for Emergency Core Cooling System per TS 3.5.2 and 3.5.3; therefore, extension of the surveillance interval for a nominal period of 139 days is required to reach the most conservative projected end of the refueling outage.
Regulatory Guide 1.118 (Revision 2) states:
" Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is verified by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstrated-that the changes in response time beyond acceptable limits are accompanied by changes in performance characteristics which are detectable during routine tests."
On January 14, 1994, the BVR Owner's Group submitted a Licensing Topical Report (LTR) prepared by the General Electric Company, NED0-32291, " System Analyses For Elimination Of Selected Response Time Testing Requirements", January 1994, for NRC review. The information contained therein justifies a one-time extension of the surveillance requirement interval of TS Table 3.2.3-3, Item C.1.a. The LTR provided justification for the elimination of selected BVR Response Time Testing or Tests (RTT), as defined in the Instrument Society of America (ISA) Standard S67.06, from the plant TS SRs. The analyses included the affected instrumentation loops which could potentially impact the instrument loop response time. In addition, plant operating experiences were reviewed to identify response time failures and how they were detected. The failure modes identified were then evaluated to determine if the effect on response time would be detected by other testing requirements contained in the TS. Based on the analyses presented in the LTR, it was concluded that there vere no failure modes which vill affect the response time of the instrumentation loop which would not be detected by other surveillances such as channel calibration (at least once per 31 days for trip setpoint calibration and at least once per 18 months for sensor calibration), channel functional tests (at least once per 31 days), i channel checks (at least once per 12 hours), or other techniques. [It should be
noted that the "other techniques" phrase applies to Rosemount transmitter slov oil loss determination.] In addition, individual instrument channel response time delays for specific trip functions (on the order of a fraction of second) have very little safety significance. Redundancy and diversity exits in most instrumentation trip functions (e.g., flux, water level, pressure). For the Emergency Core Cooling System, the instrumentation response times are a small fraction of the overall response times of the actuating devices. Based on the above, a one-time extension of the Emergency Core Cooling System Actuation Instrumentation TS SR 4.3.3.3, Table 3.3.3-3, Item C.l.a response time testing surveillance interval is justified. , I 1 f t l 1
l 1 l ENCLOSURE 61 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION j LOGIC SYSTEM FUNCTIONAL TESTING FOR i TECHNICAL SPECIFICATION SR 4.3.3.2, TABLE 4.3.3.1-1, ITEM C l.b i l DRYVELL PRESSURE - HIGH DIVISION 3 TRIP SYSTEM - HPCS SYSTEM EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.2 requires a LSFT and simulated automatic actuation of all channels of the Emergency Core Cooling System Actuation Instrumentation at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Dryvell Pressure - High functional unit for the Division 3 Trip System - HPCS System (TS Table 4.3.3.1-1, Item C l.b) does not require a surveillance interval extension for these functional units' portion of the LSFT to reach the projected start of RFO-5. However, these functional units are required to be OPERABLE in Operational Conditions 4 and 5 as required by TS 3.5.2; therefore, a surveillance interval extension is required for a nominal period of 88 days to reach the most conservative projected end of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BVR Ovners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 88 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. l
ENCLOSURE 62 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION RESPONSE TIME TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.3, TABLE 3.3.3-3, ITEM C.1.b DRYWELL PRESSURE - HIGH DIVISION 3 TRIP SYSTEM - HPCS SYSTEM EMERGENCY CORE COOLING SYSTEM INSTRUMENTATION Technical Specification SR 4.3.3.3 requires the Response Time of the Emergency Core Cooling System Instrumentation shown on Table 3.3.3-3 be demonstrated to be within the limits at least once per 18 months. Each of the tests are to 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 reactor trip system. Trip function Item C.1.b, Division 3 Trip System - HPCS System, Dryvell Pressure - High, of Table 3.3.3-3 vill become overdue prior to the beginning of RFO-5 conservatively scheduled to begin February 15, 1996. This RPS response time test requires an extension for a nominal period of 2 days to reach the most conservative projected start of RFO-5. Regulatory Guide 1.118 (Revision 2) states:
" Response time testing of all safety related equipment, per se, is not required if, in lieu of response time testing, the response time of the safety equipment is verified by functional testing, calibration checks or other tests, or both. This is acceptable if it can be demonstrated that the changes in response time beyond acceptable limits are accompanied by changes in performance characteristics which are detectable during routine tests."
On January 14, 1994, the BVR Owner's Group submitted a Licensing Topical Report (LTR) prepared by the General Electric Company, NED0-32291, " System Analyses For Elimination Of Selected Response Time Testing Requirements", January 1994, for NRC review. The information contained therein justifies a one-time extension of the surveillance requirement interval of TS Table 3.3.3-3, Item C.1.b. The LTR provided justification for the elimination of selected BVR Response Time Testing or Tests (RTT), as defined in the Instrument Society of America (ISA) Standard 567.06, f rom the plant TS SRs. The analyses included the affected instrumentation loops which could potentially impact the instrument loop response time. In addition, plant operating experiences were reviewed to identify response time failures and how they were detected. The failure modes identified vere then evaluated to determine if the effect on response time vould be detected by other testing requirements contained in the TS. Based on the analyses presented in the LTR, it was concluded that there were no failure modes which will affect the response time of the instrumentation loop which would not be detected by other surveillances such as channel calibration (at least once per 31 days for trip setpoint calibration and at least once per 18 months for sensor calibration), channel functional tests (at least once per 31 days), channel checks (at least once per 12 hours), or other techniques. [It should be noted that the "other techniques" phrase applies to Rosemount transmitter slow oil loss determination.] In addition, individual instrument channel response time delays for specific
1 4 4 trip functions (on the order of a fraction of second) have very little safety i significance. Redundancy and diversity exits in most' instrumentation trip functions (e.g., flux, water level, pressure). For the Emergency Core Cooling ' System, the instrumentation response times are a small fraction of the overall response times of the actuating devices. 1 Based on the above, a one-time extension of the Emergency Core Cooling System
- Actuation Instrumentation TS SR 4.3.3.3, Table 3.3.3-3, Item C.1.b response time testing surveillance interval is justified. +
e f i i i I s 9 a l f
F
]
ENCLOSURE 63 JUSTIFICATION FOR SURVEILLANCE IRIERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.3.1, TABLE 4.3.3.1-1, ITEM C.1.c REACTOR VESSEL VATER LEVEL - LOV, LEVEL 8 CALIBRATION DIVISION 3 TRIP SYSTEM - HPCS SYSTEM EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.1, Table 4.3.3.1-1, Item C.1.C requires the Division 3 Trip System - HPCS System Reactor Vessel Vater Level - Lov, Level 8 Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.3.1-1, Item C.1.c for a nominal period of 18 days to reach the most conservative projected start of RF0-5. However, this instrumentation is required in Operational Conditions 4 and 5 during periods of core alterations or l operations with a potential for draining the reactor vessel; therefore, ' extension of the surveillance interval for a nominal period of 110 days is required to reach the most conservative projected end of the refueling outage. In February 1990, Rosemount published a report, " 30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. l l __m --_ _ _ _ . _ _ _ _ _ _ _ _ _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ - _ _ _ _ . _ _ _ _ _ . . . _ _ _ _ - _
1 ENCLOSURE 64 ! JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION I LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.3.2, TABLE 4.3.3.1-1, ITEM C.1.h MANUAL INITIATION DIVISION 3 TRIP SYSTEM - HPCS SYSTEM EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.3.2 requires a LSFT and simulated automatic actuation of all channels of the Emergency Core Cooling System Actuation Instrumentation at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Manual Initiation functional unit for the Division 3 Trip System - HPCS System (TS Table 4.3.3.1-1, Item C.1.h) does not requires a surveillance interval extension for this functional unit's portion of the LSFT to reach the projected start of RFO-5. However, this functional unit is required to be OPERABLE in Operational Conditions 4 and 5 as required by TS 3.5.2; therefore, a total surveillance interval extension is required for a nominal period of 88 days to reach the most conservative projected end of RF0-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 88 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. l
I ENCLOSURE 65 i JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.5.1, TABLE 4.3.5.1-1, ITEM a l REACTOR VESSEL VAidR LEVEL- LOV, LEVEL 2-CALIBRATION l REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.5.1, Table 4.3.5.1-1. Item a requires the Reactor Core Isolation Cooling System Beactor Vessel Vater Level - Lov, Level 2 Instrumentation to-be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.5.1-1, Item a for a nominal period of 45 days to reach the most conservative 1 projected start of RF0-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" . (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the , drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month . calibration of the trip Jnterval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement-for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant ; safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. T
l ENCLOSURE 66 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.5.2, TABLE 4.3.5.1-1, ITEM a REACTOR VESSEL VATER LEVEL - LOV, LEVEL 2 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.5.2 requires a LSFT and simulated automatic l actuation of all channels of the Reactor Core Isolation Cooling System Actuation Instrumentation at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel f . Vater Level - Low Level 2 functional unit (TS Table 4.3.5.1-1, Item a) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 44 days to reach the most conservative projected start i of RFO-5. i As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to J extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months: i I ) "T.ndustry reliability studies for boiling water reactors (BVRs), l prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." l The evaluation above is applicable to PNPP and the surveillance interval l extension for a nominal period of 44 days is bounded by the interval accepted on l the Peach Bottom docket; therefore, the surveillance interval extension is l justified. I
ENCLOSURE 67 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.5.1, TABLE 4.3.5.1-1, ITEM b ; REACTOR VESSEL VATER LEVEL - HIGH, LEVEL 8 CALIBRATION l REACTOR CORE ISOLATION COOLING SYSTEM INSTRUMENTATION j Technical Specification SR 4.3.5.1, Table 4.3.5.1-1,-Item b requires the Reactor , Core Isolation Cooling System Reactor Vessel Vater Level - High, Level 8-Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allovable extension of ' 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table ' 4.3.5.1-1, Item b for a nominal period of 32 days to-reach the most conservative projected start of RFO-5. In February 1990, Rosemount published a report, "30 Month Stability - Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2
and 3 docket.) This report supported the extension of the calibration interval ,
for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip. units are calibrated either monthly or quarterly, depending on the~TS requirement for channel functional testing. The setpoint calculations for the trip units
~
utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the ' calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified.
ENCLOSURE 68 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.5.2, TABLE 4.3.5.1-1, ITEM b REACTOR VESSEL VATER LEVEL - HIGH, LEVEL 8 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION Technical Specification SR 4.3.5.2 requires a LSFT and simulated automatic actuation of all channels of the Reactor Core Isolation Cooling System Actuation Instrumentation at least once per 18 months (with a maximum allowable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel Water Level - High, Level 8 functional unit (TS Table 4.3.5.1-1, Item b) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 31 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, , surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BWR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 31 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. l
c 1 I I I ENCLOSURE 69-JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION 4.3.7.4, TABLE 4.3.7.4-1, ITEM 1 , REACTOR VESSEL PRESSURE CALIBRATION REMOTE SHUTDOWN SYSTEM INSTRUMENTATION Technical Specification SR 4.3.7.4, Table 4.3.7.4-1, Item 1 requires the Remote Shutdown System Reactor Vessel Pressure Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). Two channels of the instrumentation (channels A and C), containing Rosemount Model 1153 pressure transmitters, vill require an extension of the SR interval cited in TS Table 4.3.7.4-1, Item 1 for a nominal period of 52 days to reach the most conservative projected start of RFO-5. This functional unit does not provide any trip or alarm function, but, rather, is for indication only. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Hcdct 1152, 11 3 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters'from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. The existing PNPP setpoint calculations for the Rosemount transmitters which provide this indication function are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified.
ENCLOSURE 70 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.7.4.1, TABLE 4.3.7.4-1, ITEM 2 REACTOR VESSEL VATER LEVEL REMOTE SHUTD0VN SYSTEM INSTRUMENTATION Technical Specification SR 4.3.7.4.1, Table 4.3.7.4-1, Item 2 requires the Remote Shutdown System Reactor Vessel Vater Level Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.7.4-1, Item 2 for a nominal period of 44 days to reach the most conservative projected start of RFO-5. This functional unit does not provide any trip or alarm function, but, rather, is for indication only. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottem Atomic Power Station, Units 2 and 3 docket.) This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. The existing PNPP setpoint calculations for the Rosemount transmitters which provide this indication function are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified.
1 I I { ENCLOSURE 71' j JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.7.5, TABLE 4.3.7.5-1, ITEM 2 l I REACTOR VESSEL VATER LEVEL CALIBRATION ACCIDENT MONITORING INSTRUMENTATION Technical Specification SR 4.3.7.5, Table 4.3.7.5-1, Item 2 requires the Accident Monitoring System Reactor Vessel Vater Level Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per. , 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.7.5-1, Item 2 for a nominal period of 43 days to reach the most conservative projected start of RFO-5. This functional unit does not provide any trip or alarm function, but, rather, is for indication only. In February 1990, Rosemount published a report, "30 Month stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 mor,ths to 0.20% URL (2 sigma) for 30 months. The existing setpoint calculations for the Rosemount transmitters providing this incation function are bounding. There is adequate allowance in the calculations ' for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. l
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, ENCLOSURE'72 -JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL' SPECIFICATION.SR 4.3.7.5, TABLE 4.3.7.5-1, ITEM 11 .
PRIMARY CONTAINMENT /DRYWELL AREA GROSS GAMMA RADIATION MONITORS CALIBRATION ACCIDENT'HONITORING SYSTEM INSTRUMENTATION j Technical Specification SR 4.3.7.5, Table 4'3.7.5-1, Item,ll requires the. Accident Monitoring System Instrumentation, Primary Containment /Dryvell Area Gross Gamma Radiation Monitors to be demonstrated OPERABLE at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2) by performing a channel calibration. The Primary Containment / Dryvell Area Gross' . Gamma Radiation Monitors, which are' digital monitors and provide an alarm a function, are required to be OPERABLE in Operational Conditions 1,'2, and 3~and require extension of the. surveillance interval of for a nominal period of749-days to reach the most conservative projected start date for RFO-5. , The calibration data from the past two refueling outages for the Primary Containment /Dryvell Area Gross Gamma Radiation Monitors, which utilize ASI. KD-1000 Gross Gamma Radiation Monitors, has shown that these monitors have ,
-reached a maximum of 9% of the Leave-As-Is-Zone during a 22 month period. The 9% deviation is most probably attributable to inaccuracies in~the calibration i setup and M&TE equipment at the time performed rather than the monitors themselves. An intrinsic quality of the digital E-prom as used for this monitor l is that drift is nonexistent over any period of time. Therefore, extension of ~
the interval to approximately 25 months should not cause any increase in drift.- However, if drift could be attributed to the E-prom, the drift to be expected I
. vould not be appreciably greater than that seen in a 22-month period and'the- -{
setpoint would not be affected. Therefore, the extension of the surveillance a interval for a nominal period of 49 days is justified. l t e i e t [ + r
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'l' s ' ENCLOSURE 73 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION j 1
LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL' SPECIFICATION SR 4.3.9.2, TABLE 4.3.9.1-1,-ITEMS 1.a AND 3.a DRYWELL PRESSURE - HIGH CONTAINMENT SPRAY SYSTEM AND SUPPRESSION POOL MAKEUP SYSTEM ' PLANT SYSTEMS ACTUATION INSTRUMENTATION Techr.ical Specification SR 4.3.9.2 requires a LSFT and simulated.autometic actuation of all channels of the Plant Systems Actuation Instrumentation at least once per 18 months-(vith a maximum allovable surveillance interval.
' extension of 4.5 months per TS 4.0.2). The Dryvell Pressure - High. functional units for the Containment Spray System and'the Suppression Pool Makeup System .
(TS Table'4.3.9.1-1, Items.l'.a and 3.a, respectively) require surveillance. interval extensions for these functional' units' portion of the LSFT.for a nominal period of 25 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety. Evaluation Report (dated August 2, 1993) related.to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and'3, surveillance intervals from 18 to 24 months: ;
" Industry reliability studies for boiling water reactors (:BWRs), ,
prepared'by the BWR Owners Group (NEDC-30936P) show that the overall-safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components,-(e.g., pumps: and valves), which are consequently tested on.a more frequent
. basis...Since the probability of a relay or contact failure is small relative'to the probability of mechanical component failure, increasing.
the logic: system functional test interval represents no significant' change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 25 days is bounded by the interval accepted on
.the Peach Bottom docket; therefore, the surveillance-interval ~ extension is justified.
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i i l ENCLOSURE 74 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.9.1, TABLE 4.3.9.1-1, ITEM 1.c , REACTOR VESSEL VATER LEVEL - LOV, LEVEL 1 CALIBRATION PLANT SYSTEMS ACTUATION INSTRUMENTATION Technical Specification SR 4.3.9.1, Table 4.3.9.1-1, Item 1.c requires the Plant Systems Reactor Vessel Vater Level - Lov, Level 1 Instrumentation to be demonstrated OPERABLE by nerformance of a channel calibration at least once per 18 months (with a maximum allova'.la extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.9.1-1, Item 1.c for a nominal period of 45 days to reach the most conservative projected start of RFO-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. 1
i i
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ENCLOSURE 75 . i JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION j
+ ' LOGIC SYSTEM FUNCTIONAL TESTING FOR- !
TECHNICAL SPECIFICATION SR 4.3.9.2, TABLE 4.3.9.1-1, ITEMS 1.c AND 3.b REACTOR VESSEL VATER LEVEL - LOV,~ LEVEL 1 f CONTAINMENT SPRAY SYSTEM ANL SUPPRESSION POOL MAKEUP SYSTEM- j PLANT SYSTEMS ACTUA! ION INSTRUMENTATION 1 Technical Specification SR 4.3.9.2 requires a LSFT and simulated automatic , actuation of all channels of the Plant Systems Actuation Instrumentation at least once per 18 months (with a maximum allowable surveillance-interval . [' extension of 4.5 months per TS 4.0.2). The Reactor Vessel Vater Level - Lov, Level 1 functional ~ units for the cortainment Spray System and the Suppression ~ !' Pool Makeup System (TS Table 4.3.9.; 1, Items 1.c and 3.b, respectively)' require '
' surveillance interval _ extensions for these functional units' portion of the LSFT for a nominal period of 44 days to reach the most conservative projected start . ;
i of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to , extension of the Peach Bottom' Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to'24' months: :
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show-that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of.the mechanical components, (e.g., pumps. , and valves), which are consequently tested on a more frequent ; basis...Since the probability of a relay or contact failure is small ! relative to the probability of mechanical component failure, increasing ! the logic system functional test inttrval represents no significant , change in the overall safety system unavailability." i The evaluation above is applicable to PNPP and the surveillance interval ; extension for a nominal period of 44 days is bounded by the interval accepted on j the Peach Bottom docket; therefore, the surveillance interval extension is j [ justified. : f r l 9
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ENCLOSURE 76 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.3.9.1, TABLE 4.3.9.1-1, ITEM 2.a FEEDWATER SYSTEM / MAIN TURBINE TRIP SYSTEM REACTOR VESSEL VATER LEVEL - HIGH, LEVEL 8 CALIBRATION PLANT SYSTEMS ACTUATION INSTRUMENTATION T Technical Specification SR 4.3.9.1, Table 4.3.9.1-1, Item 2.a requires the Reactor Vessel Water Level - High, Level 8 trip function of the Feedvater System / Main Turbine Trip System portion of the Plant Systems Actuation Instrumentation to be demonstrated OPERABLE at least once per 18 months-(vith a maximum allovable extension of 4.5 months per TS 4.0.2) by performing a channel calibration. The Reactor Vessel Level - High, Level 8 trip function is required to be OPERABLE in Operational Condition 3 and require extension of the surveillance interval of for a nominal period of 21 days to reach the most conservative projected start date for RFO-5. As reflected in the Perry Nuclear Power Plant Reload Analysis, a new analytical limit was established for the Reactor Vessel Level - High, Level 8 trip function. The resultant margin gained from this action translates into an additional allocation for drift to accommodate an additional 18 months of operation. Additionally, the trip function is only required in Operational Condition 3. This Operational Condition is only experienced for a very small period of time during the operating cycle and vill not be experienced unless there is a plant shutdown. Therefore, extension of the interval to an approximate period of 23.5 months, which vill only include a very small period of time for which this trip function is required and is well within the drift allowance for the trip function, is justified. .P k - n - , . . - r
l l ENCLOSURE 77 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.3.9.2, TABLE 4.3.9.1-1, ITEMS 2.a REACTOR VESSEL VATER LEVEL - HIGH, LEVEL 8 FEEDVATER SYSTEM / MAIN TURBINE TRIP SYSTEM PLANT SYSTEMS ACTUATION INSTRUMENTATION Technical Specification SR 4.3.9.2 requires a LSFT and simulated automatic actuation of all channels of the Plant Systems Actuation Instrumentation at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The Reactor Vessel Vater Level - High, Level 8 functional unit for the Feedvater System / Main Turbine Trip System (TS Table 4.3.9.1-1, Item 2.a) requires a surveillance interval extension for this functional unit's portion of the LSFT for a nominal period of 22 days to reach the most conservative projected start of RF0-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic sJstem, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 22 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. i i I I l
I i i I ENCLOSURE 78 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION , TECHNICAL SPECIFICATION SR 4.3.9.1, TABLE 4.3.9.1-1, ITEM 3.b i REACTOR VESSEL VATER LEVEL - LOV, LEVEL 1 CALIBRATION PLANT SYSTEMS ACTUATION INSTRUMENTATION Technical Specification SR 4.3.9.1, Table 4.3.9.1-1, Item 3.b requires the Plant Systems Reactor Vessel Vater Level - Lov. Level 1 Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). The level transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.3.9.1-1, Item 3.b for a nominal period of 45 days to reach the most conservative projected start of RFO-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.) This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. 1 i i 1 1
1 I ENCLOSURE 79 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.4.2.1.2.b l RELIEF VALVE FUNCTION PRESSURE ACTUATION INSTRUMENTATION CALIBRATION Technical Specification SR 4.4.2.1.2.b requires the relief valve function pressure actuation instrumentation of the Reactor Coolant System Safety / Relief Valves to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allovable extension of 4.5 months per ' TS 4.0.2). The pressure transmitters, Rosemount Model 1153 pressure transmitters, vill require an extension of the SR interval cited in TS SR 4.4.2.1.2.b for a nominal period of 28 days to reach the most conservative ; projected start of RFO-5. . In February 1990, Rosemount published a report, "30 Honth Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are ' calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The setpoint calculations for the trip units-utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip uni channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. i i t i 1 l l
ENCLOSURE 80 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.4.2.1.2.b RELIEF VALVE FUNCTION PRESSURE ACTUATION INSTRUMENTATION Technical Specification SR 4.4.2.1.2.b requires a LSFT and simulated automatic actuation of the Relief Valve Function Pressure Actuation Instrumentation of required Reactor Coolant System Safety / Relief Valves at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). This LSFT requires a surveillance interval extension for a nominal period of 33 days to reach the most conservative projected start of RF0-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, - surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact' failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 33 days is bounded by the interval-accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. I l
ENCLOSURE 81 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.4.2.2.1.b RELIEF VALVE FUNCTION PRESSURE ACTUATION INSTRUMENTATION CALIBRATION Technical Specification SR 4.4.2.1.2.b requires the relief valve low-lov set function pressure actuation instrumentation of the Reactor Coolant System Safety / Relief Valves to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). The pressure transmitters, Rosemount Model 1153 pressure transmitters, vill require an extension of the SR interval cited in TS SR 4.4.2.2.1.b for a nominal period of 28 days to reach the most conservative projected start of RFO-5. In February 1990, Rosemount published a report, "30 Honth Stability Specification For Rosemount Hodel 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval I for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are l calibrated either monthly or quarterly, depending on the TS requirement for j channel functional testing. The setpoint calculations for the trip units ; utilized a drift value of 0.23% SP (2 sigma) which bounds the required drift value of 0.13% SP (2 sigma). The existing setpoint calculations for the trip units for Rosemount transmitters and trip unit channels are bounding. There is adequate allowance in the calculations for 30 month transmitter drift and no potential impact on plant safety analyses (i.e., no analytic limit changes are required). Therefore, the requested extension is justified. i t
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d ENCLOSURE 82 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION LOGIC SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.4.2.2.1.b RELIEF VALVE FUNCTION PRESSURE ACTUATION INSTRUMENTATION Technical Specification SR 4.4.2.2.1.b requires a LSFT and simulated automatic actuation of the Relief Valve Function and the Low-Low Set Function Pressure Actuation Instrumentation of the required Reactor Coolant System Safety / Relief Valves at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). This LSFT requires a surveillance interval extension for a nominal period of 33 days to reach the most conservative projected start of RFO-5. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 33 days is bounded by the interval accepted on the Peach Bottom docket; therefore, the surveillance interval extension is justified. l l l
ENCLOSURE 83 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.4.3.1.b DRYVELL FLOOR DRAIN AND EQUIPMENT DRAIN SUMP FLOV MONITORING SYSTEM CALIBRATION REACTOR COOLANT SYSTEM LEAKAGE DETECTION SYSTEM Technical Specification SR 4.4.3.1.b requires the Dryvell Floor Drain and Equipment Drain Sump flow monitoring system portion of the Reactor Coolant System Leakr.ge Detection System to be demonstrated OPERABLE at least once per 18 months (v11h a maximum allowable extension of 4.5 months per TS 4.0.2) by performirg a channel calibration. The Dryvell Floor Drain and Equipment Drain Sump flow monitoring system, which provides an alarm function, is required to be OPERABLE in Operational Conditions 1, 2, and 3 and requires extension of the surveillance interval of for a nominal period of 20 days to reach the most conservative projected start date for RFO-5. The calibration data from the past two refueling outages for the Dryvell Floor Drain and Equipment Drain Sump flow monitoring system, which utilize a Gould-Model PD-3218 Differential Pressure Transmitter and Barton 368 Level Transmitter, has shown that these transmitters have reached a maximum of 66% of the alarm setpoint calculation drift allocation during a 22 month period. In that the extension for the surveillance calibration interval for these transmitters vould provide for only an approximate 24 month interval, the drift to be expected would not exceed the drift allowance and the alarm function vould not be affected. Therefore, the extension of the surveillance interval for a nominal period of 20 days is justified.
ENCLOSURE 84 ; l JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.4.3.1.c i DRYVELL AIR COOLERS CONDENSATE FLOV RATE MONITORING SYSTEM CALIBRATION REACTOR COOLANT SYSTEM LEAKAGE DETECTION SYSTEM Technical Specification SR 4.4.3.1.c requires the Dryvell Air Coolers Condensate flow rate monitoring system portion of the Reactor Coolant System Leakage Detection Systems to be demonstrated OPERABLE at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2) by performing a channel calibration. The Dryvell Air Coolers Condensate flow rate monitoring system, which provides an alarm function, is required to be OFERABLE in Operational Conditions 1, 2, and 3 and requires extension of the surveillance interval of for a nominal period of 19 days to reach the most conservative projected start date for RF0-5. The calibration data from the past two refueling outages for the Dryvell Air Coolers Condensate flow rate monitoring system, which utilizes a Rosemount 8712 flow transmitter, has shown that these transmitters have reached a maximum of 30% of the allowed drift between the Setpoint and the Allowable Value during a 22 month period. In that the extension for the surveillance calibration interval for these transmitters vould provide for only an approximate 24 month interval, the drift to be expected would not exceed the allowed drift and the required alarm function vould not be affected. Therefore, the extension of the surveillance interval for a nominal period of 19 days is justified.
l ENCLOSURE 85 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION FOR TECHNICAL SPECIFICATION 4.4.3.2.2 REACTOR COOLANT PRESSURE ISOLATION VALVE LEAKAGE Technical Specification SR 4.4.3.2.2 requires that each reactor coolant system pressure isolation valve specified in TS Table 3.4.3.2-1 be demonstrated OPERABLE by leak testing pursuant to TS 4.0.5 and verifying leakage to be limited to less than 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at least once per 18 months (with a maximum allovable extension of 4.5 months per TS 4.0.2). This SR requires extension of the surveillance interval for valves 1E21F005, 1E21F006, 1E51F065, and 1E51F066 for nominal period of 50 days to reach the most conservative start date for RF0-5. Valves 1E21F005 and lE21F006 which isolate the Low Pressure Core Spray, were tested during RFO-4 and each exhibited as-found leakage of 0.0 gpm. Valves 1E51F065 and 1E51F066, RCIC/RHR 'A' Head Spray isolation, were tested during RFO-4 and exhibited as-found leakage of 0.0 gpm and 0.017 gpm, respectively. The allovable leakage per TS 3.4.3.2.d is 5.0 gpm each for valves 1E21F005 and 1E21F006, and 3.0 gpm each for valves 1E51F065 and 1E51F066. As can be seen, the observed leakage is well within the allovable leakage. It should be further noted that NUREG-1463, " Regulatory Analysis for the Resolution of Generic Safety Issue 105: Interfacing System Loss-of-Coolant Accident in Light Vater Reactors" (July 1993), Section 2.5, "BVR Results" states:
"An ISLOCA analysis was performed as part of the GI-05 research program is documented in NUREG/CR-5928. The work consisted of screening analyses and bounding calculations on the systems identified as potentially susceptible to an ISLOCA. A BVR/4 was examined, and the following systems were explicitly addressed: reactor core isolation cooling, high pressure coolant injection, core spray, RER, reactor water cleanup, and control rod drive. The study concluded that ISLOCA is not a concern."
i The above is applicable to PNPP in that the design and accident response of a BVR/6 are very similar to a BVR/4 for this issue. l In addition, the PNPP Individual Plant Examination (IPE) (transmitted to'NRC by PY-CEI/NRR-1517L dated July 15, 1992), Section 3.1.3.6.1 states:
"An evaluation was performed on the potential paths of an interface LOCA. The paths noted in section 4.4.25 of NUREG/CR-4550 were identified as the dominant paths contributing to interface LOCA for the Perry plant. The Grand Gulf and Perry design are identical for these paths. Using the data in NUREG/CR-4550 the core damage frequency for an interfacing LOCA is less than 10(-8)/yr."
Based on the lov as-found leakage rate of the subject valves, the margin available from the TS allowed leakage, and the conclusion of NUREG-1463 and the PNPP IPE that the ISLOCA (which is the rationale for the PIVs in the system
design) is not a risk concern for BVRs, including PNPP, the extension of the surveillance interval for 50 days is justified. ,
ENCLOSURE 86 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.5.1.c.1 LPCI, LPCS, AND HPCS SYSTEMS EMERGENCY CORE COOLING SYSTEMS Technical Specification SR 4.5.1.c.1 requires a system functional test, including simulated automatic actuation, of LPCI, LPCS and HPCS systems the throughout their emergency operating sequence, excluding the actual injection of coolant into the reactor vessel, at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). These system functional tests vill require surveillance interval extensions for a nominal period of 25 days to reach the most conservative projected start of RFO-5. However, the ECCS is required to be OPERABLE in Operational Conditions 4 and 5 per the requirements of TS 3.5.2; therefore, extension of the surveillance intervals for a nominal period of 117 days is required to reach the most conservative projected end of RF0-5. In that the system functional test is essentially a Logic System Functional Test, extension is justified based on the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months, which states:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The simulated automatic actuation test is supplemented during the operating cycle by tests performed on instrumentation and mechanical components of the system. These include CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS, Inservice Testing and other component verification surveillances as specified by the TS. In that the simulated automatic actuation test is required to assure system operability during the operating cycle, and periodic testing is performed during the operating cycle, the requested extension to the surveillance interval is justified. The LSFT evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 117 days is bounded by the interval accepted on the Peach Bottom docket. The simulated automatic actuation test is supplemented during the cycle to assure operability. Based on these considerations, the surveillance interval extension is justified. l 1 I
l l l ENCLOSURE 87 l JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION SYSTEM FUNCTIONAL TESTING FOR TECHNICAL SPECIFICATION SR 4.5.1.e.2.a AUTOMATIC DEPRESSURIZATION SYSTEM CHANNELS A AND B EMERGENCY CORE COOLING SYSTEMS Technical Specification SR 4.5.1.e.2.a requires a system functional test, including simulated automatic actuation of the Automatic Depressurization System throughout its emergency operating sequence, excluding the actual valve actuation, at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). This system functional test vill require a surveillance interval extension for a nominal period of 22 days to reach the most conservative projected start of RFO-5. In that the system functional test is essentially a Logic System Functional Test, extension is justified based on the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months, which states:
" Industry reliability studies for boiling water reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The simulated automatic actuation test is supplemented during the operating cycle by tests performed on instrumentation and mechanical components of the system. These include CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS, Inservice Testing and other component verification surveillances as specified by the TS. In that the simulated automatic actuation test is required to assure system operability during the operating cycle, and periodic testing is performed during the operating cycle, the requested extension to the surveillance interval is justified. The LSFT evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 22 days is bounded by the interval accepte' on the Peach Bottom docket. The simulated automatic actuation test is supplemer.ted during the cycle to assure operability. Based on these considerations, the surveillance interval extension is justified.
ENCLOSURE 88 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION FUNCTIONAL TEST FOR TECHNICAL SPECIFICATION SR 4.6.1.4.C.1 MSIV LEAKAGE CONTROL SYSTEM Technical Specification SR 4.6.1.4.c.1 requires that a functional test, including simulated actuation throughout the operating sequence, be performed on the HSIV Leakage Control subsystems at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2). The functional test also includes verification that each automatic valve in the flow path actuates to its correct position and the blower starts. This SR requires extension of the surveillance interval for a nominal period of 21 days to reach the most conservative project start of RFO-5. The functional test provides a testing of the logic and functional components of the MSIV Leakage Control System. The functional components, consisting of the valves and blowerc, are tested periodically during the operating cycle in accordance with TS SRs 4.6.1.4.a and 4.6.1.4.b. Therefore, the functional test provides for testing of the logic in conjunction with the functional components. As stated in the NRC Safety Evaluation Report (dated August 2, 1993) related to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:
" Industry reliability studies for boiling vater reactors (BVRs),
prepared by the BVR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." The simulated actuation test is supplemented during the operating cycle by tests performed on instrumentation and mechanical components of the system. These include CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS, Inservice Testing and other component verification surveillances as specified by the TS. In that the simulated actuation test is required to assure system operability during the operating cycle, and periodic testing is performed during the operating cycle, 7 l the requested extension to the surveillance interval is justified. The LSFT evaluation above is applicable to PNPP and the surveillance interval extension for a nominal period of 21 days is bounded by the interval accepted on the Peach Bottom docket. The simulated actuation test is supplemented during the cycle to assure operability. Based on these considerations, the surveillance interval extension is justified. i i l l L _ _ _ _ l
ENCLOSURE 89 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SR 4.6.1.4.d.2 REACTOR VESSEL PRESSURE CALIBRATION MSIV LEAKAGE CONTROL SYSTEM INSTRUMENTATION Tachnical Specification SR 4.6.1.4.d.2 requires the MSIV Leakage Control System Reactor Vessel Pressure Instrumentation to be demonstrated OPERABLE by performance of a channel calibration at least once per 18 months (with a maximum allowable extension of 4.5 months per TS 4.0.2). The pressure transmitters, Rosemount Model 1153 level transmitters, vill require an extension of the SR interval cited in TS Table 4.6.1.4.d.2 for a nominal period of 52 days to reach the most conservative projected start of RFO-5. In February 1990, Rosemount published a report, "30 Month Stability Specification For Rosemount Model 1152, 1153, 1154 Pressure Transmitters" (Rosemount Report D8900126, Revision A) [ accepted by NRC Safety Evaluation Report dated August 2, 1993 on the Peach Bottom Atomic Power Station, Units 2 and 3 docket.] This report supported the extension of the calibration interval for the transmitters from 18 months to 30 months based on a reduction in the drift allowance from 0.29% URL (2 sigma) for 18 months to 0.20% URL (2 sigma) for 30 months. In addition, applicable setpoint calculations assumed 18 month calibration of the trip interval for trip units. However, the trip units are calibrated either monthly or quarterly, depending on the TS requirement for channel functional testing. The MSIV Leakage Control System is a manually inititated system. The Reactor Vessel Pressure transmitters and trip units provide a permissive and trip function for operation of the Leakage Control System blowers and valves. The setpoint for transmitters and trip units was established based on the time that the Leakage Control System should be shutdown, if operating. The allowable value for this setpoint was established as 1.5 times the Leave-As-Is-Zone to provide a reasonable operability limit for the transmitters and trip units. Sufficient margin exists between the Nominal Trip Setpoint and the site-established Allovable Value to justify the surveillance interval extension. The margin is significantly greater than the 30 month transmitter drift and trip unit drift uncertainty values applicable to the Rosemount instruments. Therefore, the requested extension is justified.
ENCLOSURE 90 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTEh TON FOR TECHNICAL SPECIFICATION SR 4.6.4.2 CONTAINMENT ISOLATION VALVES AUTOMATIC ACTUATION CONTAINHENT SYSTEMS Technical Specification SR 4.6.4.2 requires that each automatic containment isolation valve be demonstrated OPERABLE at least once per 18 months (with a maximum allovable surveillance interval extension of 4.5 months per TS 4.0.2) by verifying that, on an isolation actuation test signal, the automatic isolation valves actuate to their isolation position. Several valves in the RHR Isolation System, RCIC Isolation System, High Pressure Core Spray System and Nuclear Closed Cooling System require extension of the surveillance interval for a maximum nominal period of 53 days to reach the most conservative projected start of RFO-5. However, in that the containment isolation valves are required to be OPERABLE in Operational Conditions 4 and 5 when handling irradiated fuel in the primary containment, during CORE ALTERATIONS, and during operations with a potential for draining the reactor vessel, a total extension of the surveillance interval for a nominal period of 145 days is required. The valves which require extension of the surveillance interval, with the required extension, are: RHR ISOLATION (158 days extension) 1E12-F008, Shutdown Cooling Outboard Suction Isolation 1E12-F009, Shutdown Cooling Inboard Suction Isolation 1E12-F023, RHR A Head Spray Isolation 1E12-F037A, RHR A Upper Pool Cooling Isolation 1E12-F037B, RHR B Upper Pool Cooling Isolation RCIC ISOLATION (153 days extension) 1E51-F031, RCIC Pump Suppression Pool Suction Isolation 1E51-F063, RHR and RCIC Steam Supply Inboard Isolation 1E51-F064, RHR and RCIC Steam Supply Outboard Isolation 1E51-F068, RCIC Turbine Exhaust Shutoff 1E51-F076, RHR and RCIC Steam Supply Varm-up Isolation 1E51-F077, RCIC Exhaust Vacuum Breaker Secondary Isolation HIGH PRESSURE CORE SPRAY SYSTEM (101 days extension) 1E22-F004, HPCS Injection 1E22-F023, HPCS Test Valve to Suppression Pool NCC SYSTEM (82 days extension) 1P43-F140, NCC Containment Return Outboard Isolation IP43-F055, NCC Containment Supply Outboard Isolation IP43-F215, NCC Containment Return Inboard Isolation The penetrations providing isolation of primary containment have redundancy so that an active failure of any single valve or component does not prevent
. t containment isolation. In addition, periodic testing of the containment isolation system is performed during power operation, including Inservice Testing of valves. Based on the redundancy provided, testing during power operation, and the short, time duration for which the surveillance interval extensi on is requested, the extension forla nominal period of 145 days'is justified. I N b k h 1
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i
k ENCLOSURE 91 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION FOR TECHNICAL SPECIFICATION SRs 4.6.6.2.d.1 AND 4.6.6.2.d.3.b ANNULUS EXHAUST GAS TREATMENT SYSTEM SYSTEM FUNCTIONAL TEST Technical Specification SRs 4.6.6.2.d.1 and 4.6.6.2.d.3.8 require that the ! i Annulus Exhaust Gas Treatment System (AEGTS) be demonstrated OPERABLE at least once per 18 months (with a maximum allowable extension of the surveillance interval for 4.5 months per TS 4.0.2) by performance.of a system functional test, including simulated automatic actuation, and verification that the filter train starts and the isolation dampers open on a simulated automatic actuation signal. These SRs require extension of the surveillance intervals-for a nominal f period of 15 days to reach the most conservative projected start of RFO-5. : However, in that the AEGTS is required to be OPERABLE during Operational . Conditions 4 and 5 when irradiated fuel is being handled in primary containment and during CORE ALTERATIONS and operations with the potential for draining the reactor vessel, extension is required for a nominal period of 107 days to reach the most conservative projected end of RFO-5. , The AEGTS is partially tested in conjunction with the LPCI B and C LSFT and the
LPCS.and LPCI A LSFT. The logic used to initiate the AEGTS has been previously justified to allow its testing to be extended based on reliability studies presented in the BVR Owners Group topical report NEDC-30936P and has been j accepted by the NRC(reference NRC Safety Evaluation Report, dated August 2, 1993, for the Peach Bottom Atomic Power Station, Units 2 and 3).
E The automatic actuation surveillance' interval is juutified to be extended based r on the redundancy of the system involved, the Inservice Testing performed during the operating. cycle, and the short time duration for which the extension is i requested. Based on the above, the one-time extension of the surveillance interval for a nominal period of 107 days is justified. i 5 s
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i ENCLOSURE 92 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION FOR TECHNICAL SPECIFICATION SR 4.7.1.1.b EMERGENCY SERVICE VATER SYSTEM (LOOPS A, B, C) SYSTEM VALVE ACTUATION Technical Specification 4.7.1.1.b requires that the Emergency Service Water System loops be demonstrated OPERABLE at least once per 18 months (with a maximum allowable extension of the surveillance interval of 4.5 months per TS 4.0.2) by verifying that each automatic valve servicing safety.related equipment actuates to the correct position on a LOCA test signal. The valve actuation, which vill be tested in conjunction with the LPCI B and C LSFT, the LPCS and LPCI A LSFT, and the HPCS LSFT, vill require extension for maximum nominal period of 28 days to reach the most conservative projected start of RFO-5. However, these functional units are required to be OPERABLE in Operational Conditions 4 and 5 as required by TS 3.5.2; therefore, a total surveillance interval extension is required for a maximum nominal period of 120 days.to reach the most conservative projected end of RF0-5. The logie used to initiate the Emergency Service Vater System has been previously justified to allow its testing to be extended based on reliability studies presented in the BVR Owners Group topical report NEDC-30936P and has been accepted by the NRC (reference NRC Safety Evaluation Report, dated August 2, 1993, for the Peach Bottom Atomic Power Station, Units 2 and 3). , The automatic valve actuation surveillance interval is justified to be extended based on the redundancy of the system involved, the Inservice Testing performed during the operating cycle, and the short time duration for which the extension is requested. Based on the above the one-time extension of the surveillance interval for a maximum nominal period of 120 days is justified. l
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ENCLOSURE 93 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION FOR TECHNICAL SPECIFICATION SR 4.7.1.2.b AUTOMATIC VALVE ACTUATION EMERGENCY CLOSED COOLING VATER SYSTEM Technical Specification 4.7.1.2.b requires that the Emergency Closed Cooling Vater System loops be demonstrated OPERABLE at least once per 18 months (with a maximum allovable extension of the surveillance interval of 4.5 months per TS 4.0.2) by verifying that each automatic valve servicing safety related equipment actuates to the correct position on a LOCA test signal. The valve actuation, which will be tested in conjunction with the LPCI B and C LSFT, the LPCS and LPCI A LSFT, and the HPCS LSFT, vill require extension for maximum nominal period of 28 days to reach the most conservative projected start of RFO-5. However, these functional units are required to be OPERABLE in Operational Conditions 4 and 5 as required by TS 3.5.2; therefore, a total surveillance interval extension is required for a maximum nominal period of 120 days to reach the most conservative projected end of RF0-5. The logic used to initiate the Emergency Closed Cooling Vater System has been previously justified to allow its testing to be extended based on reliability studies presented in the BVR Owners Group topical report NEDC-30936P and has been accepted by the NRC(reference NRC Safety Evaluation Report, dated August 2, 1993, for the Peach Bottom Atomic Power Station, Units 2 and 3). The automatic valve actuation surveillance interval is justified to be extended based on the redundancy of the system involved, the Inservice Testing performed during the operating cycle, and the short time duration for which the extension is requested. Based on the above the one-time extension of the surveillance interval for a maximum nominal period of 120 days is justified. 1 l l 1 - e
5 ENCLOSURE 94 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION , FOR TECHNICAL SPECIFICATION SRs 4.7.2.e.2.a AND 4.7.2.e.2.b CONTROL ROOM EMERGENCY RECIRCULATION SYSTEM AUTOMATIC TRANSFER DURING ON HIGH DRYVELL PRESSURE AND REACTOR VESSEL VATER LEVEL - LOV, LEVEL'1 Technical Specifications 4.7.2.e.2.a and 4.7.2.e.2.b require that the Control Room Emergency Recirculation System be demonstrated OPERABLE at least once per , 18 months (with a maximum allovable extension of the surveillance interval of 4.5 months per TS 4.0.2) by verifying that the system transfer-to the emergency recirculation mode and the isolation dampers close upon initiation by High Dryvell Pressure and Reactor Vessel Water Level - Lov, Level 1 actuation tests 1 signals, respectively. The transfer to the emergency recirculation mode on initiation by the cited tests signals, which vill be tested in conjunction with the LPCI B and C LSFT, the LPCS and LPCI A LSFT, and the HPCS LSFT, will require extension for maximum nominal period of 28 days to reach the most conservative projected start of RFO-5. However, these functional units are required to be OPERABLE in Operational Conditions 4 and 5 as required by TS 3.5.2; therefore, a total surveillance interval extension is required for a maximum nominal period of 120 days to reach the most conservative projected end of RFO-5. The logic used to initiate the Control Room Emergency Recirculation System has been previously justified to allow its testing to be extended based on > reliability studies presented in the BVR Owners Group topical report NEDC-30936P and has been accepted by the NRC(reference VRC Safety Evaluation Report, dated August 2, 1993, for the Peach Bottom Atomic Power Station, Units 2 and 3). , The automatic actuation surveillance interval is justified to be extended based on the redundancy of the system involved, the Inservice Testing performed during the operating cycle, and the short time duration for which the extension is requested. Based on the above the one-time extension of the surveillance interval for a maximum nominal period of 120 days is justified. l l l l 1 l I
ENCLOSURE 95 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION FOR TECHNICAL SPECIFICATION SR 4.7.4.b SNUBBER VISUAL INSPECTION Technical Specification 4.7.4.b (as submitted by letter PY-CEI/NRR-1811L dated September 19, 1994) requires that a visual inspection of all snubbers be performed according to the schedule of TS Table 4.7.4-1. For the next inspection period the interval vill be 18 months (with a maximum allowable extension of the surveillance interval of 4.5 months per TS 4.0.2). This surveillance interval vill require an extension for a nominal period of 53 days to reach the most conservative start date for RF0-5. However, in that systems required to be OPERABLE in Operational Conditions 4 and 5 vill be affected by the snubber inspection, an extension of the interval to reach the most conservative end date for RFO-5 of 158 days is required. The =nubber visual inspection surveillance interval is based on the amount of service tius axperienced by a snubber at operating (environmental and dynamic) conditions. The snubber installed in a non-operating system is akin to a snubber stored on-the-shelf and experiences no appreciable degradation in this environment. Since startup of PNPP from RFO-4 in August 1994 to the most conservative end date of RFO-5 vill not exceed 22.5 months, the service time of the snubbers vill not exceed surveillance interval. In that the extension vill not provide any additional time than that assumed in the TS for snubber service, no effect on the OPERABILITY of the snubbers can reasonably be expected to be introduced by this extension. Therefore, the extension of the surveillance interval to the most conservative end of RFO-5 is justified, i i ________________________________I
ENCLOSURE 96 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION FOR TECHNICAL SPECIFICATION SR 4.7.4.e SNUBBER FUNCTIONAL TESTING Technical Specification 4.7.4.e requires that a representative sample of snubbers be functionally tested at least once per 18 months (with a maximum allowable extension of the surveillance interval of 4.5 months per TS 4.0.2). This surveillance interval vill require an extension for a nominal period of 51 days to reach the most conservative start date for RF0-5. However, in that systems required to be OPERABLE in Operational Conditions 4 and 5 vill be affected by the snubber functional testing, an extension of the interval to reach the most conservative end date for RFO-5 for a nominal period of 156 days is required. The snubber functional testing surveillance interval is based on the amount of service time experienced by a snubber at operating (environmental and dynamic) conditions. The snubber installed in a non-operating system is akin to a snubber stored on-the-shelf and experiences no appreciable degradation in this environment. Since startup of PNPP from RF0-4 in August 1994 to the most conservative end date of RFO-5 vill not exceed 22.5 months, the service time of the snubbers vill not exceed surveillance interval. In that the extension vill not provide any additional time than that assumed in the TS for snubber service, no effect on the OPERABILITY of the snubbers can reasonably be expected to be introduced by this extension. Therefore, the extension of the surveillance interval to the most conservative end of RF0-5 is justified. l l
ENCLOSURE 97 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION FOR TECHNICAL SPECIFICATION SR 4.8.1.1.1.b 0FFSITE TRANSHISSION CIRCUIT /0NSITE CLASS 1E CIRCUIT TRANSFER DIVISIONS 1, 2, AND 3 Technical Specification 4.B.l.1.1.b requires that the independent circuits between the offsite transmission network and the onsite Class 1E distribution system be demonstrated OPERABLE at least once per 18 months (with a maximum allovable extension of the surveillance interval of 4.5 months per TS 4.0.2) during shutdown by transferring unit power supply from the normal circuit to the alternate circuit. The SR for each Division vill not require an extension of the surveillance interval to reach the most conservative projected start of RF0-5. However, in that the Division 1, 2, and 3 electrical power systems are required to be OPERABLE during Operational Conditions 4 and 5, during movement of fuel in primary containment or the fuel handling building, during CORE ALTERATIONS, and during operations with the potential for draining the reactor vessel, extension of the surveillance intervals are required for Division 1 for a period of 114 days, Division 2 for a period of 35 days, and Division 3 for a period of 92 days to reach the most conservative end date of RF0-5. Four offsite power sources are available to the switchyard to provide an offsite source of power to the 4.16 kV emergency busses. The failure of any one of the offsite power sources supplying power to the busses does not result in a total loss of offsite power to the bus. The design of the offsite power to the busses provides a decreased likelihood that a total loss of offsite power vill occur. However, if a total loss of offsite power occurred and operation of the diesel generators was required, the requested extension vould have minimal impact on the circuit transfer failure probability. The extension of the surveillance interval for the diesel generator circuit transfer logic testing has in itself the same rationale for extension as LSFTs on other systems / components. Since the failure probability of the logic (relays, contacts, etc.) is reasoned, as documented in NEDC-30936P, to be less than the failure probability for the mechanical equipment (pumps, valves. etc.), the extension of the surveillance interval for the circuit transfer logic has minimal impact on the failure to function. And, since the mechanical components (diesel generators) are tested on a more frequent basis (i.e., monthly and quarterly), the probability of failure to function is further minimized. Therefore, the limited extensions of the surveillance intervals to reach the most conservative end date of RF0-5 has minimal impact on the failure probability and is justified. i j
ENCLOSURE 98 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION FOR TECHNICAL SPECIFICATION SR 4.8.1.1.2.f.1 DIVISION 3 DIESEL GENERATOR INSPECTION Technical Specification SR 4.8.1.1.2.f.1 requires that the Division 3 diesel generator be subjected to an inspection in accordance with the manufacturer's recommendations. This surveillance interval vill not require an extension of the surveillance interval to reach the most conservative start date for RFO-5. However, since Division 3 is. required to be OPERABLE in Modes 4 and 5 when HPCS is required to be OPERABLE, an extension of the surveillance interval to the Division 3 outage system vindow is required. This extension vill be no longer than the period to the most conservative end date of RFO-5 for a nominal period of 92 days. During the operating cycle, the diesel generators are subjected to testing every 31 days. This testing provides confidence of diesel generator integrity and capability to perform its intended function. This testing during the past cycles has not indicated any degradation of the diesel which would negate the extension of the surveillance as requested. In addition, the past internal inspections conducted in accordance with TS SR 4.8.1.1.2.f.1 have not revealed any degradation which would necessitate replacement of internal components, although, as a matter of course, several components have been replaced as preventive measures. The GM-EMD Owners Group has issued a program (similar to PNPP's) which has been endorsed by GM-EMD. This program defines the refueling cycle as "18-24 months". Hence, it is recognized by the industry at large that the operating cycles vary and the 18 month surveillance interval required by TS may be arbitrary. Based on the above, the requested extension of the surveillance interval for a nominal period of 92 days is justified. i
ENCLOSURE 99 JUSTIFICATION FOR SURVEILLANCE INTERVAL EXTENSION TECHNICAL SPECIFICATION SRs 4.8.1.1.2.f.2, 4.8.1.1.2.f.3, 4.8.1.1.2.f.4.a. 4.8.1.1.2.f.4.b, 4.8.1.1.2.f.5, 4.8.1.1.2.f.6.a, 4.8.1.1.2.f.6.b, 4.8.1.1.2.7.a, 4,8.1.1.2.f.7.b, 4.8.1.1.2.f.8, 4.8.1.1.2.f.9, 4.8.1.1.2.f.10, 4.8.1.1.2.f.11, 4.8.1.1.2.f.12, 4.8.1.1.2.f.14.a, 4.8.1.1.2.f.14.b EMERGENCY DIESEL GENERATOR 18-HONTH LOADING AND TESTING DIVISIONS 1, 2, AND 3 Technical Specification SRs cited above provide for verifying: i) load rejection capability; 11) de-energization and load shedding, and starting, re-energization, and loading on a simulated loss of offsite power, an ECCS actuation test signal, and a simulated loss of offsite power in conjunction with an ECCS actuation test signal; iii) bypass of automatic diesel generator trips on an ECCS actuation test signal; iv) continuous operation at rated load, voltage and frequency; v) auto-connected loads do not exceed maximum loading the diesel generator (s); vi) ability to synchronize with offsite power source while loaded, to transfer loads to the offsite power source, and to be restored to standby status; vii) override of the test mode by a simulated ECCS actuation signal causing return of the diesel generator to standby status and automatic energization of the emergency loads with offsite power; viii) load sequence timer operation; and, ix) lockout features prevent diesel generator starting only as required. These verifications are to be performed at least once per 18 months (with a maximum allovable extension of the surveillance interval of 4.5 months per TS 4.0.2) and require an extension of the surveillance intervals to reach the most conservative start date for RFO-5 for a nominal period of 13 days for Division 1. However, in that, per TS 3.8.1.2, a combination of one or more diesels may be required in Operational Conditions 4 and 5, and during periods of fuel handling in the primary containment and the fuel handling building, during CORE ALTERATIONS, and during operations with the potential for draining the reactor vessel, each of the diesel generators require extension of the surveillance interval to the most conservative end date for RFO-5 for total nominal period of 118 days for Division 1, for a total nominal period of 69 days for Division 2, and for a total nominal period of 52 days for Division 3. Four offsite power sources are available to the switchyard to provide an offsite source of power to the 4.16 kV emergency busses. The failure of any one of the offsite power sources supplying power to the busses does not result in a total loss of offsite power to the bus. The design of the offsite power to the busses provides a decreased likelihood that a total loss of offsite power vill occur. However, if a total loss of offsite power were to occur and operation of the diesel generators was required, the requested extension vould have minimal impact on the system failure probability. The extension of the surveillance interval for the diesel generator logic testing has in itself the same rationale for extension as LSFTs on other systems / components. Since the failure probability of the logic (relays, contacts, etc.) is reasoned, as documented in NEDC-30936P, to be less than the failure probability for the mechanical equipment (pumps, valves, etc.), the extension of the surveillance interval for the logic has minimal impact on the failure to function. And, since the mechanical components (diesel generators) are tested on a more frequent basis (i.e., monthly and quarterly), the probability of failure to function is further minimized. Therefore, the limited extensions of the surveillance intervals to reach the most conservative end date of RFO-5 has minimal impact on the failure probability and is justified.}}

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