Text

Forwards List of Tech Spec Changes to Accompany Full Power Ol,Representing Minor Editorial Corrections,Clarifications & Enhancements.Justifications & Proposed marked-up Tech Spec Pages Also Encl
	ML20199E755
Person / Time
Site:	Perry
Issue date:	06/18/1986
From:	Edelman M; CLEVELAND ELECTRIC ILLUMINATING CO.
To:	Harold Denton; Office of Nuclear Reactor Regulation
References
	PY-CEI-NRR-0469, PY-CEI-NRR-469, NUDOCS 8606240053
	Download: ML20199E755 (76)
	v • d • e

THE CLEVELAND ELECTRIC ILLUMIN ATING COMPANY P.o box 5000 - CLEVELAND, oHlo 44101 - TELEPHONE (216) 622-9800 - lLLUMINATING BLOG - 55 PUBLICSOUARE Serving The Best Location in the Nation MURRAY R. EDELMAN VICE PRESIDENT NucuAR June 18, 1986 PY-CEI/NRR-0469 L Mr. Harold R. Dentoa, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Perry Nuclear Power Plant Docket No. 50-440 Changes to Technical Specifications for Full Power Licensing

Dear Mr. Denton:

This letter provides a list of changes that the Cleveland Electric Illuminating Company request be included in the Technical Specifications which will accompany the full power operating license for the Perry Nuclear Power Plant-Unit 1. These items represent minor editorial corrections, clarifications, and enhancements to the Technical Specifications. The justifications and the proposed markup pages are attached.

None of these changes affect CEI's ability to safely operate the Perry Nuclear Power Plant-Unit I under its current license. Thus, no amendment of the present low power license is being requested. If you have any questions, please call me.

_gou s, Very A / truly,7

/s

/

Murray R. Edeiman Vice President Nuclear Group Attachments MRE:njc cc: Jay Silberg, Esq D. Eisenhut J. Keppler John Stefano (2) R. Bernero C. Norelius J. Grobe W. Butler C. Paperiello G. Lainas R. Knop 8606240053 86061 %

PDR ADOCK 05000440 '

p PDR I \

Attachment PY-CEI/NRR-0469 L Justification Technical Specifications Table 1.2 Figure 3.2.1-1 Figure 3.2.1-2 Table 3.3.2-1 4.6.1.2.c Table 3.6.4-1 4.6.6.2.b 3.8.4.1 Table 3.8.4.1-1 Table 3.12.1-1 The above Technical Specifications contained typographical / editorial errors. The proposed changes correct these errors and add clarification.

6 O

e

TABLE 1.2

(-

OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE

1. POWER OPERATION Run Any temperature

2. STARTUP Startup/ Hot Standby Any temperature

3. HOT SHUTDOWN Shutdown #'*** > 200 F

4. COLD SHUTDOWN Shutdown #'##'*** 1 200 F

5. REFUELING

Shutdown or Refuel **'# 1 140*F l

i

The reactor mode switch may be placed in the Run, Startup/ Hot Standby, or Refuel position to test the switch interlock functions and related instrumentation provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

1. The reactor mode switch may be placed in the Refuel position while a single ;

control rod drive Specification is being removed from the reactor pressure vessel per 3.9.10.1. !

Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

- Cee Special Test Exceptions 3.10.1 and 3.10.3.

- - The reactor mode switch may be placed in the Refuel position while a single control rod is being recoupled provided that the one rod-out interlock is 4

OPERABLE.

o r ulfhdrawn

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PERRY - UNIT 1 1-11

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AVERAGE PLANAR EXPOSURE (mwd /t)

MAXIMUM AVERAGE PLANAR LINEAR HEAT ll GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE -

I INITIAL CORE FUEL TYPES-P&-6 RIH 99- 8PB SR8 2l9

! Figure 3.2.1-1 i

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' GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE INITIAL CORE FUEL TYPES 90-6AlHM B FB s- R S / *7 (;

Figure 3.2.1-2 i

l i

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j TABLE 3.3.2-1 (Continued) h ISOLATION ACTUATION INSTRUNENTATION i A

. VALVE GROUPS MINIMUM APPLICABLE c OPERATED BY OPERA 8LE CHANNELS OPERATIONAL I

F, TRIP FUNCTION SIGNAL i -e PER TRIP SYSTEM (a) _ CONDITION ACTION !

N

5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION i

a.

b.

RCIC Steam Line Flow - High RCIC Steam Supply Pressure -

9 h 1, 2, 3 27

)

c.

Low RCIC Turbine Exhaust 9 @g 1,2,3 27

, Diaphragm Pressure - High 9 II) 2 1,2,3 27

d. RCIC Equipment Room Ambient i Temperature - High 9 1 1, 2. 3 27 I

e. RCIC Equipment Room A j 1 Temperature - High 9 1 1,2,3 27 1

i' f. Main Steam Line Tunnel

} '

C Ambient Temperature - High 9 1 1,2,3 27 I g. Main Steam Line Tunnel i & Temperature - High 9 1 1,2,3 27

h. Main Steam Line Tunnel i, Temperature Timer 9 1 1,2,3 27

1. RHR Equipment Room Ambient Temperature - High 9 1 1,2,3 27

j. RHR Equipment Room A j} Temperature - High 9 1 1,2,3 27

k. RCIC Steam Line Flow High ,

i Timer 9 1 1,2,3 27 l 1. Drywell Pressure - High 9(h) 1 1,2,3 . 27

m. Manuel Initiation S I} 1 1,2,3 26 t

\ . - . .

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

I 1

CONTAINMENT SYSTEMS C SURVEILLANCE REQUIREMENTS (Continued)

1. Confirms the accuracy of the tes; by verifying that the differ-ence beteen the supplemental data and the Type A test data is within 0.25 L,. The formula to used is:, b e

~

[L, + L ,- 0.25 L,] 1 Le i El *o 'm + 0.25 L,3,where Lc"

) '

supplemental test result; L, = superimposed leakage; L ,=

l measured Type A leakage.

2. Has duration sufficient to esta;11;h accurately the change in

1eakage rate between the Type A test and the supplemental test.

I 3. Requires the quantity of gas injected into the primary contain-ment or bled from the primary containment during the supple-1 mental test to be between 0.75 L, and 1.25 L,.

) d. Type 8 and C tests shall be conducted with gas at P , 11.31 psig*,

j at intervals no greater than 24 months except for tIsts involving:

1. Air locks, l .

2. Main steam line isolation valves,

3. Valves pressurized with fluid from a seal system, l 4. All containment isolation valves in hydrostatically tested lines

{ per Table 3.6.4-1 which penetrate the primary containment, and j 5. Purge supply and exhaust isolation valves with resilient

material seals.

e. Air locks shall be tested and demonstrated OPERABLE per surveillance i

Requirement 4.6.1.3.

I-

f. Main steam line isolation valves shall be leak tested at least once per la months.

g. Leakage from isolation valves that are sealed with fluid from a seal

! system may be excluded, subject to the provisions of Appendix J of 10

CFR 50 Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P maintain,12.44psig,andthesealsystemcapacityisadequateto system pressure for at least 30 days.

h. All containment isolation valves in hydrostatically tested lines per i Table 3.6.4-1 which penetrate the primary containment shall be leak !

tested at least once per 18 months. '

i. Purge supply and exhaust isolation valves with resilient material j seals shall be tested and demonstrated OPERA 8LE per surveillance !

Requirements 4.6.1.8.3. and 4.6.1.8.4. *

j. The provisions of Specification 4.0.2 are not applicable' to Specifications 4.6.1.2.a. 4.6.1.2.b, 4.6.1.2.c 4.6.1.2.d. and 4.6.1.2.e.

( "Unless a hydrostatic test is required per Table 3.6.4-1.

l PERRY - UNIT 1 , 3/4 6-5 l

m ,

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, 3. 4. 4 - l Table 4,6 4+ Containment and Drywell Isolation Valves

R g NOTES: a. Isolation valve for instrument ifne which penetrates the containment, conforms to the requirements j -< of Regulatory Gutde 1.11. The In-service Inspection (ISI) program w111 provide assurance of the operabillty and integrity of the 1solatton provisions. Type "C" testing will not be performed on i

E the instrument ifne isolation valves. The instrument ifnes will be within the boundaries of the 4

t Q Type "A" test, open to the media (containment atmosphere or suppression pool water) to which they w111 be exposed under postulated accident conditions. Three except1ons to the above are penetratiens P401, P318, and P425. Isolation valves for these three penetrations include the He analyzer and Post

] Accident Sampling System valves. These valves are norma 11y closed post-LOCA, opened only inter-i mittently, and will receive Type C tests.

1

+ . b. Hydrostatic leak test at 11.10Pa.

I c. See spectfication 3.3.2, Table 3.3.2-1, for isolation signal (s) associated with each valve

] groups 1-9. Valve groups 10-13, 16 and 17 are as fo11ews: .

li Valve Group 10 - MSIV Leakage Control System Valve Group 11 - Reacter Rec 1rculation System Valve Group 12 - Combustible Gas Control System i 1:* Valve Group 13 - Orywell Vacuum Relief System

Valve Group 16 - HPCS

]

j ? Valve Group 17 - LPCS

! O d. Test connection valve.

4 L. e. Remote manually controlled valve.

~

f. Check valve.

l

g. See section 3/4.4.7, " Main Steam Line Isolation Valves." =

h. . During Type C testing, valve stem and bonnet are checked for leaks as potential secondary contafament bypass leakage paths, i f. Net required to be OPERABLE in OPERATIONAL COW ITION **.

) J. Not required to be OPERABLE fn OPERATIONAL COMITIONS 1, 2 and 3.

1 -

}

Standard closure time, based upon nominal pfpe diameter, is approximately 12 inches / min for gate valves and approximately 4 inches / min for globe valves.

f **When handling frradiated fuel in the primary containment anet during CORE j, ALTERATIONS and operations with a potential for draining the reactor j vessel.

l l

l

f CONTAINMENT SYSTEMS k SURVEILLANCE REQUIREMENTS (Continued) i

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:

i 1. Verifyir.g that the subsystem satisfies the in place penetration j

testing acceptance criteria of less than 0.05% and uses the test 1

procedure guidance in Regulatory Positions C.5.a. C.S.c and

! C j & sy.S.d staaoflow Regulatory rate is 2000 Guide scfm 1.52, 2 105. Revision 2, March 1978, M tM f

N/<yu.My i 2. Verify i {ng within 31 days after removal that a laborat

' 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 J Position C.6.a of Regulatory Guide 1.52, Revision 2 March 1978, by showing a methyl iodide penetration of less than 0.1755 when tested at a tenperature of 30*C and at a relative humidity of 705 in accordance with ASTM D3803; and I;

3. Verifying a subsystem flow rate of 2000 scfm i 10% during system i operation when tested in accordance with ANSI M510-1980. The i

installed air flow monitor can be used to detenmine flow in i lieu of the pitot traverse.'

( c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying

] within 31 days after removal that a laboratory analysis of a repre-l i

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 i of 30*C and at a relative humidity of 705 in accordance with l ASTM D3803; j d. At least once per 18 months by: .

1. Performing a system functional test which includes simulated '

automatic actuation of the systes throughout its emergency j operating sequence for the LOCA.

l

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 ;

1 15.

3. Verifying that the filter train starts and isolation dampers

) open on each of the following test signals:

l a.

1 Manual initiation from the control room, and 1

b.

l Simulated automatic initation signal.

1

4. Verifying that the heaters dissipate 20 kw i 10% when tested

] ( in accordance with ANSI N510-1980.

PERRY - UNIT 1 3/4 6-49 I ,

1 i

)

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ELECTRICAL POWER SYSTEMS CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES f

i LIMITING CONDITION FOR OPERATION '

f i

3.8.4.1 All containment penetration conductor overcurrent protective devices

shown in Table 3.8.4.1-1 shall be OPERABLE. -

! i i APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

[

ACTION:

a. With one or more of the containment penetration conductor overcurrent I

! protective devices shown in Table 3.8.4.1-1 inoperable, declare the affected I

system or component inoperable and apply the appropriate ACTION statement i for the affected system and.

i

1. For 13.8 kV circuit breakers, de-energize' the 13.8 kV circuit (s) l by tripping the associated redundant circuit breaker (s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and verify the redundant circuit breaker to be tripped at least once 3 '

per 7 days thereafter. .

,y b'. (

2. akers, remove W M 9 the inoperable circuit 1 breaker (sk..JICvoltcircuitFor 120 ,M ervice by n.,in; :;t the brea i and verify the inoperable breaker (s) to be ;d;d ;.0 at least once i per 7 days thereafter. fe,ged ;

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in l

] COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . !

b. The provisions of Specification 3.0.4 are not applicable to overcurrent i

devices in 13.8 kV ci uits which have their redundant circuit breakers ;

j tripped or to 120 c. Iff, Jolt circuits which have the inoperable circuit l

, breaker tsched-out. N ;

$'t'hfcd SURVEILLANCE REQUIREMENTS

! i 4.8.4.1 Each of the containment penetration conductor overcurrent protective l

] devices shown in Table 3.8.4.1-1 shall be demonstrated OPERABLE:

)

a. At least once per 18 months:

l

1. By verifying that the medium voltage 13.8 kV circuit breakers -

are OPERA 8LE by selecting, on a rotating basis, at least 10% of '

the circuit breakers and performing: ; )I

~

a) A CHANNEL CALIBRATION of the associated protective relays,

' b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and overcurrent con-l trol circuits function as designed, and i PERRY - UNIT 1 3/4 8-21 1

l 1

f TABLE 3.8.4.1-1 .

CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES 13.8 KV LOAD OVERCURRENT PROTECTION

~

Primary Secondary 1833-C001A L1106 1R22-S012 1833-C0018 L1205 1R22-S013 120VLOADbRCIRCUIT 1821-81X (Sp. Htr.) 1R25-5097-C81 NA*

1821-83X (Sp. Htr.) 1R25-5097-C82 NA*

1 1821-85X (Sp. Htr.) 1R25-5093-C81 NA*

i 1821-8760X8 1R25-5043-C820 -

NA*

1 1821-8756XS 1R25-5043-C818 NA*

1821-8758X8 1R25-5043-C819 NA*

1821-8754XS 1R25-5043-C817 NA*

1821-8752X8 1R25-5047-C811 NA*

, 1833-82'X (Sp. Htr. ) 1R25-5093-C87 NA*

ls33-spX (Sp. Htr.) 1R25-5093-C88 NA*

i I a43-spX (Sp. Htr.)

1R25-5097-C85 NA*

!' I 1833-si rX (Sp. Htr.)

[ ,

9 fj_1533-BAX (Sp. Htr.)

1R25-5097-C86 NA*

1R25-5093-C89 NA*

,- ,' fy 1533-5;EX (Sp. Htr.) 1R25-5093-C810

NA* ,

g ::233-s;4X 1R25-5097-C87 233-8;4X(Sp.

(Sp. Htr.)

NA* '

Htr.) 1R25-5097-C88 NA*

1833-817X (Sp. Htr.) 1R25-5093-C811 NA*

1833-819X (Sp. Htr.) 1R25-5097-C89 NA* j 1833-821X (Sp. Htr.) 1R25-5093-C84 NA* i

(; 1833-823X (Sp. Htr.) 1R25-5097-C810 NA*

i {

IC11-8250X (Sp. Htr.) 1R25-5093-C85 NA*

1C11-C1X NA* 1H13-P653-C81 j

IC41-89X8 (Sp. Htr) 1R25-5043-C821 NA*

1E51-83X8 1R25-5043-C824 NA*

1 1E51-81X8 1R25-5043-C823 NA*

' 3 1F42-8FX (Sp. Htr.) 1R25-5097-C83 NA*

, 1G33-81X (Sp. Htr.) 1R25-5095-C81 NA*

1G33-83X (Sp. Htr.)

1R25-5081-C82 NA*

j

' (3 1G33-85X (Sp. Htr.) 1R25-5079-C81 NA*

1G33-87X (Sp. Htr.) 1R25-5079-C82 NA*

1G33-89X (Sp. Htr.) 1R25-5081-C81 NA*

't

Protected by fuse. !

1 Pfl PERRY - UNIT 1 3/4 8-23

-. . .-.-__.w-- -

--- - - - - - - - -- - - - - - - - *- - ~ ~ ~ " ~ ~~ ~ ~

Attachment PY-CEI/NRR-0469 L i

a Justification i

Technical Specifications 3.3.5 i Table 3.3.7.5-1 Table 4.3.7.5-1 i 3.3.8 3.4.1.2 3.4.1.3 3.4.3.1 3.6.1.1.2 1 3.6.5.3 j 3.6.6.1 1

3.7.3 3.11.1.3 3.11.2.5 3.11.3 The above Technical Specifications contained certain deferrals to operability requirements from initial fuel load to prior to exceeding 5%

of rated thermal power. This change request would delete these deferrals.

= These deferrals should be deleted because they will not be applicable af ter exceeding 5% of rated thermal power, which will be authorized by the full power license.- At that time, the systems will have been declared operable.

l j

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..__m.__ _ _ . _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . . _ _ ._ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ . _ . _ _ _ _ _

\

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[ INSTRUMENTATION 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION l

LIMITING CONDITION FOR OPERATION i 3.3.5 The reactor core isolation cooling (RCIC) system actuation instrumenta-tion channels shown in Table 3.3.5-1 shall be OPERABLE with their trip set-points set consistent with the values shown in the Trip Setpoint column of i Table 3.3.5-2.

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2 and th 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 r

' channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

, b. With one or more RCIC system actuation instrumentation channels

(' inoperable, take the ACTION required by Table 3.3.5-1.

l SURVEILLANCE REQUIREMENTS i

4.3.5.1 Each RCIC system actuation instrumentation channel shall be demon-

! strated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL l TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 1 4.3.5.1-1.

l .,

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

i t

1 l

. \

I "Not required to be OPERABLE until after non-nuclear heatup following initial ,

( criticality.

PERRY - UNIT 1 -

3/4 3-50 i l

-l l '

l l

. . - - - ........__,2,. - . _ - - - _ ...--.._....-.-_...-:. . . - . . . - . . .-.

] -

1 < r e ., .

TA8tE 3.3.7.5-1 A

g ACCIDENT MONITORING INSTRUMENTATION r ,

E MININUM APPLICA8LE Q REQUIRED NUMBER CHANNELS OPERATIONAL

INSTRUMENT OF CHANNELS OPERA 8LE CONDITIONS ACTION

1. Reactor Vessel Pressure 2 2.

1 1,2,3 80 J Reactor Vessel Water Level 2 1 1,2,3 80 j . 3. Suppression Pool Water Level 2 1 1,2,3

4. 80

Suppression Pool Water Temperature 16, 2/ sector 8, 1/ sector 1,2,3 80 1 5. Primary Containment Pressure 2 1 1,2,3

6. 80 I,

Primary Containment Air Temperature 2 1 1,2,3 80

7. Drywe11 Pressure 2 1 1,2,3 80 l 8. Drywell Air Temperature 2 1 1,2,3 80 i 9. Primary Containment and Drywell Hydrogen i

Concentration Analyzer and Monitor 2 1 1,2,3 80

] !:' 10. Safety / Relief Valve Position Indicators ** 2/ valve 1/ valve 1,2,3

80

?

11. Primary Containeen rywell Area Gross Gamma .

Radiation Monito # 2* 1* 1,2,3 81 i

a 12. Offgas Ventilation Exhaust Monitor

' 1 1 1,2,3 81

13. Turbine Bu' iding/ Heater Bay Ventilatinn Exhaust i Monitor b 1 1 1,2,3 81 l 14. Unit 1 Vent Monitor, ,,, 1 , 1 1,2,3 81 l 15. Unit 2 Vent itor 1 1 1,2,3 81 l 16. Neutron F1 j a. Average Power Range 2 1,2,3 1 80

! b. Intermediate Range 2 1 1,2,3 80 i c. Source Range am 2 1 1,2,3 80

{ 17. Primary Containment Isolation Valve Position '

2/ valve 1/ valve 1,2,3 -82 i

i =

! Each for primary containment and drywell.

am j

one channel consists of a pressure switch on the SRV discharge pipe, the other channel corsists of a temperature sensor on the SRV discharge pipe.

i nan i One channel consists of the open limit switch, and the other channel consists of the closed limit switch

{ for each automatic containment isolation valve in Table 3.6.4-1,a. -

! High and intermediate range D19 system noble gas monftors.

! required to be OPERA 8LE prior to exceeding 5% of RATED THERMAL POWE (i

_ m p i

TA8tE 4.3.7.5-1 A

g ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4

g CHANNEL CHANNEL APPLICABLE OPERATIONAL q INSTRUMENT CHECK CALIBRATION CONDITIONS

1. Reactor Vessel Pressure M R 1, 2, 3

2. Reactor Vessel Water Level M R 1, 2, 3

. 3. Suppression Pool Water Level M 4.

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

5. Primary Containment Pressure M 6.

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

7. Drywell Pressure M R 1, 2, 3 i

8. Drywell Air Temperature M 9.

R 1,2,3 Primary Containment and Drywell Hydrogen Concentration Analyzer and Monitor NA Q* 1,2,3

10. Safety /Reifef Valve Position Indicators M R 1, 2, 3 w 11. Primary Containment /Drywell Area Gross Gamma Radiation Monito M R** 1,2,3

12. Offgas Ventilation Exhaust Monitor M R 1, 2, 3

13. Turbine Bufidi ater Bay Ventilation Exhaust Menitor

M R 1, 2, 3

14. Unit 1 Vent Monitor

M , R 1, 2, 3

15. Unit 2 Vent after

M R 1, 2, 3

16. Neutron Flu

a. Average Power Range M R 1, 2, 3'

b. Intermediate Range M R 1, 2, 3

c. Source Range M R 1, 2, 3

17. Primary Containment Isolation Valve Positi M . R 1, 2, 3 -

Usino 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 below 10 R/hr with an installed or portable gamma source.

High and intermediate ran y D19 system noble gas monitors.

i @Not required to be OPERA 8 E prior to exceeding 5% of RATED THERMAL POWE%

_ _ _ _ _ _ - __ _--_ __ _ + _ w __ _

l

)

INSTRUMENTATION 1

3/4.3.8 TUR8INE OVERSPEED PROTECTION SYSTEN LINITING CONDITION FOR OPERATION 3.3.8 At least one turbine overspeed protection system shall be OPERA 8LE.

j APPLICA8!LITY: OPERATIONAL CONDITIONS I and 1 & '

ACTION:

} l

a. With one turbine control valve or one turbine stop valve per high t

pressure turbine steam line inoperable, and/or with one turbine intercept or intermediate stop valve per low pressure turbine staae line inoperable, restore the inoperable valve (s) to OPERA 8LE status 1 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or close at least one valve in the affected steam 1

i line or isolate the turbine from the steals supply within the next '

l 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the above required turbine overspeed protection system otherwise j !

inoperable, within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months isolate the turbine from the steam supply. !

i SURVEILLANCE REQUIREMENTS 4.3.8.1 The provisions of Specification 4.0.4 are not applicable.

4.3.8.2 The above required turbine overspeed protection system shall be demonstrated OPERABLE:

j a. At least once per 7 des by:

1. Cycling each of the following valves through at least one <

complete cycle from the running position:

a) For the overspeed protection control system;

1) Six low pressure turbine intercept valves, and t

\

2) Four high pressure tuttine control valves.

i

! b) For the electrical overspeed trip systes and the mechanical l

' 1 overspeed trip system; i s

1) Four high pressure turbine stop valves, and l

2) Six low pressure turbine intermediate stop valves, and i 3) Four high pressure turbine control valves.

i I ( [NotrequiredtobeOPERA8LEpriortoexceeding15ofRATEDTHERMALPOWERfor (the first tim fe -

PERRY - UNIT 1 3/4 3-96 i

l REACTOR COOLANT SYSTEM .'

JET PUMPS LIMITING CONDITION FOR OPERATION i 3. 4.1. 2 All jet pumps shall be OPERABLE. t APPLICABILITY: OPERATIONALCONDITIONSIand2h ACTION:

With one or more jet pumps inoperable, be in at least NOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCEREpUIREMENTS l

4.4.1.2 Each of the above required jet pumps shall be demonstrated OPERA 8LE prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by determining recirculation loop flow, total core flow and diffuser-to-lower plenum differantial pressure for each jet pump and veri,fying that no two of the following conditions occur when the recirculation loops are

( operating at the same flow control valve position.

a. The indicated recirculation loop flow differs by more than 105 from the estaclished flow control valve position-loop flow characteristics.

i b. The indicated total core flow differs by more than 10K from the established total core flow value derived from recirculation loop flow measurements.

c. The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from established patterns by more than 10%.

, t required to be OPERA 8LE prior to nuclear heatup 1

[

e

. PERRY - UNIT 1 3/4 4-4

- . , - . - - , - , - - - - - - , , - -- -r - . , . , , , . - - - . _ , , , . . - - - - , , - , , - - ,_-,.~-,,,e?,,_.--,_-.,,--.---w...n.., -

REACTOR COOLANT SYSTEN RECIRCULATION LOOP FLOW LIMITING CONDITION FOR OPERATION 3.4.1.3 Recirculation loop ficw mismatch shall be maintained within:

a.

5% of rated recirculation flow with core flow greater than or equal to 70% of rated core flow.

b.

10% of rated recirculation flow with core flow less than 70% of rated core flow.

! APPLICABILITY: OPERATIONAL. CONDITIONS 18 and 2

^

I ACTION:

i With either: recirculation loop flows different by more than the specified limits, a.

Restore the recirculation loop flows to within the specified Ifait within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

( b.

Declare the recirculation loop with the lower flow not in operation and take the ACTION required by Specification 3.4.1.1.

O SURVEILLANCE REQUIREMENTS 4.4.1.3 Recirculation loop flow mismatch shall be verified to be within the limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i N

" cial Test Exception 3.10._4.

(#Notrequiredto.DeUrtRABLEpriortonuclearheatup.f1 -

i PERRY - UNIT 1 -

3/4 4-5 i

- - . - ,- -- - . , - - - , . - . - ~ r-, -_, - - , , - - . , , , - - - . - - --- --- - --,

4 3

l t

j. !

REACTOR C0OLANT SYSTEN

\ \

3/4.4.3 REACTOR C00LANT SYSTEN LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.3.1 The following reactor coolant system leakage detection systems shall be OPERABLE- I

a. The drywell atmosphere particulate or gaseous radioactivity monitoring

system, !

i b. The drywell floor drain sump and equipment drain sump flow monitoring

] system, and

! c. The upper drywell air coolers condensate flow rate monitoring system.

1 I l APPLICA81LITY: OPERATIONAL CONDITIONS 1, 8 nd h

\

ACTION:

1 1 1

With only two of the required leakage detection systems 0PERA8LE, operation may continue for up to:

a. 30 days when the required gaseen.s and particulate radioactive i monitoring system is inoperable provided grab samples of the drywell :

j atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or l I

l b. 30 days when the drywell floor drain sump or equipment drain simp a flow monitoring system is inoperable, or l j c. 30 days when the upper drywell air coolers condensata flow rate monitoring system is inoperable.

Otherwise, be in at least NOT SHUTDOW witMn the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD i 5HUTDOW within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

j $URVEILLANCE REQUIREMNT$

j 4.4.3.1 The reactor coolant systes leakage detection systems shall be demon-

strated 0PERA8LE by: l

) a. Drywell atmosphere particulate or gaseous monitoring systems-performance of a CHAfstEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a

{' CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per la months.

j b. Drywell floor drain and equipment drain simp flow monitoring systen-i perforsmance of a CHANNEL FilNCTIONAL TEST at least once per 31 days

] and a CHANNEL CALISRATION at least once per la months.

c. Upper drywell air coolers condensate flow rate monitoring systee-performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months. ,

i aihe crywell floor drains sump and equipment drain sump flow monitoring system I , and the upper drywell air coolers condensate flow rate monitoring system are j not required to be OPERA 8LE prior to nuclear heatup. j -

PERRY - UNIT 1 ,

3/4 4-9

}

t.____.-.___ _ __ __-.._ _ _ _ ___..._____m-

~

l l

I j i 3/4.6.1 PRIMARY CONTAINNENT .

PRIMARY CONTAINNENT INTEGRITY - SHUTOOWN LIMITING CONDITION FOR OPERATION 3.6.1.1.2 PRIMARY CONTAINMENT INTEGRITY

shall be maintained.

APPLICABILITY:

4

' i When irradiated is being handled in the primary containment, and during

{ CORE ALTERATION , and operations with a potential for draining the reactor vessel. Under these conditions, the requirements of PRIMARY CONTAINMENT k-j INTEGRITY do not apply to normal operation of the inclined fuel transfer i systas.

ACTION:

Without PRIMARY CONTAIMiENT INTEGRITY, suspend handling of irradiated fuel in i

the primary containment, CORE ALTERATIONS, and operations with a potential for draining the reactor vessel.

SURVEILLANCE REQUIREMENTS 1

i (

, 4.6.1.1.2 PRIMARY CONTAIMiENT INTEGRITY shall be demonstrated:

ie 1 a. At least once per 31 days by verifying that all primary containment i penetrations not capable of being closed by 0PERABLE primary contain-j ment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deacti-4 wated automatic valves secured in position, except as provided in j

f Table 3.6.4-1 of Specification 3.6.4.

b. By verifying each p' rimary containment air lock is in compliance with 1 the requirements of Specification 3.6.1.3.

i l

l,

The primary containment leakage rates in accordance with Specification 3.6.1.2 art not applicable.

.[F'he primary coni.ainment equipment nascn 1s nos required to oe ciosea ang L sealed during initial fuel load. f

{ PERRY - UNIT 1 -

3/4 6-2 .

, CONTAlfetENT SYSTEMS ORWELL VACUUM BREAKERS LINITING CONDITION FOR OPERATION i

3.6.5.3 All drywell vacuum breakers shall be OPERABLE and closed. s APPLICABILITY: OPERATIONAL CONDITIONS 1, nd 3N -

j ACTION:

a. With one drywell vacuum breaker inoperable for opening but known to be j closed, restore the inoperable vacuum breaker to OPERA 8LE status within i

i 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i b. With one drywell vacuum breaker open, restore the open vacuum breaker to the closed position within I hour or be in at least NOT SHUT-DOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ,

1 i~

c. With the position indicator of an OPERABLE dryw11 vacuum breaker inoper-able, verify the vacuum breaker to be closed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by local indication. Otherwise, declare the vacuus breaker inoperable. ;

i' SURVEILLANCE REQUIREMENTS I 1

! ~

I 4.6.5.3 Each drywell vacuum breaker shall be:

a. Verified closed at least once per 7 dvs.

! b. Demonstrated OPERA 8LE:

1. At least once per 31 days by a). Cycling'the vacuum breaker and associated isolation valve j -

through at least one complete cycle of full travel.

i b) Verifying the position indicators 0PERA8tE by observing l

4 expected valve movement during the cycling test.

! 2. At least once per 18 months by:

i i

a) Verifying the pressure differential required to open the vacuum breaker, from the closed position, to be less than or equal to 0.5 psid (containment to drywell), and b) Verifying the position indicators 0PERA8LE by performance l of a CHANNEL CALIBRATION.

3. By verifying the OPERA 8ILITY of the vacuum breaker isolation I valve differential pressure actuation instrumentation with the l

opening setpoint 5 -0.810 inch water gauge dp by performance of a: .

a) CHANNEL FUNCTIONAL TEST at least once per 31 days, and '

j b) CHANNEL CALIBRATION at least once per 18 months.

t j

' *Not required to be OPERA 8LE until af ter non-nuclear heatup following initia criticality. _

1 l PEARY - UNIT 1 . 3/4 6-44 . .

i

i 4

( CONTAINMENT SYSTEMS 3/4.6.6 SECONDARY CONTAINMENT SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.6.1 SECONDARY CONTAINMENT INTEGRITY shall be maintained.

l

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and i

ACTION:

Without SECONDARY CONTAIMfENT INTEGRITY:

s. In OPERATIONAL CONDITION 1, 2 or 3, restore SECONDARY CONTAINMENT

!- INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN withiff the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. In Operational Condition spend handling of irradiated fuel in the primary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable. ,

SURVEILLANCE REQUIREMENTS i

(

e 4.6.6.1 SECONDARY CONTAIMENT INTEGRITY shall be demonstrated by:

1

a. Verifying at least once pier 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that the pressure within the secondary containment is less than or equal to 0.40 inches of vacum water gauge.

1 i b. Verifying at least once per 31 days that:

1. The primary containment equipment hatch is closed and sealed and the shield blocks are installed adjacent to the shield -

building.

2. The door in each access to the secondary containment is closed, except for routine entry and exit.

! 3. All penetrations terminating in the annulus not capable of being i closed by OPERA 8LE automatic isolation valves and required to be .

i closed during accident conditions are closed by valves, blind l flanges, or deactivated automatic valves secured in position. '

4 "When irradiated fuel is being handled in the primary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

[#The primary containmen; equipment hatch is not required to be closed and .

T sealed and the shield blocks are not required to be installed adjacent to the

, shield building during initial fuel load. Surveillance Requirements 4.6.6.1.a and 4.6.6.1.b.1 are not required to demonstrate secondary containment integrity ,

( during initial fuel load. f PERRY - AINIT 1 ,

3/4 6-47

PLANT SYSTEMS 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE flow path capable of automatically taking suction from the sup- -

pression pool and transferring the water to the reactor pressure vessel.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3 pressure greater than 150 psig. h reactor steam dome ACTION:

With the RCIC system inoperable, operation may continue provided the HPCS system is OPERABLE; restore the RCIC system to OPERA 8LE status within 14 days or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reducethe within reactor steam following 24donc pressure to less than or equal to 150 psig hours.

SURVEILLANCE REQUIREMENTS 4.7.3 The RCIC systes shall be demonstrated OPERABLE: -

( a. At least once per 31 days by:

o 1.

Verifying by venting at the high point vents that the systes piping from the pump discharge valve to the system isolation -

valve is filled with water. <

. 2. .

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

3. Verifying that the pump flow controller is in the correct position, b.

When tested pursuant to Specification 4.0.5 by verifying that the RCIC pump develops a flow of greater than or equal to 700 gpa in the test flow path with a system head corresponding to reactor vessel operating. pressure when steam is being sup 1020 + 25 - 100 psig (steam dose pressure)* plied to the turbine at

The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test. I l

'#Not required to be OPERABLE until after non nuclear heatup following initial criticality.

s l

PERRY - UNIT 1 3M 7-6 u .

7 .

l 1 . . ._ ~

l

_RADI0 ACTIVE EFFLUENTS

_ LIQUID RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.11.1.3 The LIQUID RADWASTE TREATMENT SYSTEM shall be OPERABLE and appropriate i portions of the system shall be used to reduce the release of radioactivity when ,

- the projected doses due to the liquid effluent from each reactor unit to i UNRESTRICTED AREAS (see Figure 5.1.1-1) would exceed 0.06 mrem to the total I body or 0.2 area to any organ, in a 31-day period.

APPLICABILITY: At all times h l

ACTION:

l

a. With radioactive liquid waste being discharged without treatment and in excess of the above Ifmits, and any portion of the liquid l j

radwaste treatment system'not in operation, prepare and submit to the i Commission, within 30 days pursuant to Specification 6.9.2, a Special Report which includes the following information: i j

1. Explanation of why liquid radwaste was being discharged without ;

treatment, identification of any inoperable equipment or sub- '

systems, and the reason for the inoperability, and I

2. Action (s) taken to restore the inoperable equipment to OPERABLE a

status, and l

3. Summary description of action (s) taken to prevent a recurrence. ,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. -

l SURVE1LLANCE REQUIRENENTS 4.11.1.3.1 Doses due to liquid releases from each reactor unit to UNRESTRICTED AREAS shall be projected at least once per 31 days, in accordance with method-ology and parameters in the ODCM.

4.11.1.3.2 The installed LIQUID RADWASTE TREATMENT SYSTEM shall be demonstrated OPERA 8LE by meeting Specifications 3.11.1.1 and 3.11.1.2.

l hNot required to be OPERABLE prior to initial criticality.

l l

PERRY - UNIT 1 -

3/4 11-6 l l

,, _ , _ _ , , , . _ --v *"-- - -"~" --' "

l 1

RADI0 ACTIVE EFFLUENTS i

VENTILATION EXHAUST TREATMENT SYSTEMS '

I I

LIMITING CONDITION FOR OPERATION i

3.11.2.5 The VENTILATION EXHAUST TREATMENT SYSTEMS shall be OPERABLE and appropriate portions ref the system shall be used to reduce releases of radio-activity when the projected dose due to gaseous effluent releases from each reactor unit to areas at and beyond the SITE BOUNDARY (see Figure 5.1.1-1) in a 31 day period would exceed 0.3 ares to any organ of a MEMBER OF THE FUBLIC.

APPLICABILITY: At all time b '

t ACTION:

a. With radicactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commis-

< sion within 30 days, pursuant to Specification 6.9.2, a Special Report which in:1udes the following information:

1. Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems which resulted in gase.'us radwaste being discharged without treatment, and the reason for the

( inoperability,

2. Action (s) ,taken to restore the inoperable equipment to OPERA.BLE status, and ,

3. Summary description of action (s) taken to prevent a recurrence. '

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

4. n.2.5.1 Doses due to gaseous releases from each reactor unit to areas at and beyond the SITE SOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the 00CM.

4. n.2.5.2 The installed VENTILATION EXHAUST TREATMENT SYSTEMS shall be demonstrated OPERABLE by meeting Specifications 3. n.2.1 and 3. u.2.3.

't l

l 1

C"Not required to be OPERABLE prior to initial criticality. T j I PERRY - UNIT 1 3/4 u-15

h. -, -- ?..-. , _, , - , - - . - ,-,. ,,, , ,.,, , - - - , - - . , - , , _ - ,

l t

s, _RADI0 ACTIVE EFFLUENTS 3/4.11.3 SOLID RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION '

3.11.3 Radioactive wastes shall be SOLIDIFIED or dewatered in accordance with the PROCESS CONTROL PROGRAM to meet shipping and transportation requirements during transit, and disposal site requirements when received at the disposal site.

APPLICABILITY: At all time i

ACTION:

a. With SOLIDIFICATION or dewatering not meeting disposal site and shipping and transportation requirements,. suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures and/or the solid waste system as necessary to prevent recurrence.

b.

With the SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, (1) test the Taproperly processed waste in each container to ensure that it meets burial ground and shipp,ing requirements and (2) take appropriate administrative action to prevent

( recurrence.

c.

The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIRENENTS

4. H.3.1 If the SOLIDIFICATION method is used, the PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g. , filter sludges, spent resins, evaporator bottoms, and sodium sulfate solutions).

a.

If ator test specimen fails to verify SOLIDIFICATION, the SOLID 1 FICA-TION of the batch under test shall be suspended until such time as additonal test specimens can be obtained, alternative SOLI 0IFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.

t

Not required to be OPERABLE prior to initial c

. l PERRY - UNIT 1 -

3/4 n-18

..r .

..s ,.

$. -._ , _ , - , , -. w. . w.,_ .- . ,,.,,rw. m,-_m.,,- .,,,,-L,s.--,,..ey,.-y n m. _- m- ,,,,-a.,.

Justification Technical Specification Table 4.3.6-1

( Control Rod Block Instrumentation Surveillance Requirements Technical Specification Table 4.3.6-1 presently lists the surveillance frequencies of the Rod Pattern Control System (RPCS) channel functional tests as startup, daily, and monthly, with respective footnotes. This change request would clarify the required surveillance frequencies of the RPCS channel functional tests to startup and monthly.

The intent of these surveillance requirements is to demonstrate operability of the Low Power Setpoint (LPSP) and the High Power Setpoint (HPSP). The purpose of the LPSP is to minimize individual rod worths to ensure the peak enthalpy of 280 cal /g will not be exceeded in the event of a control rod drop accident. As proven by analysis, this is not a concern at reactor power greater than 20% rated thermal power, and therefore the RPCS does not place any pattern restrictions on control rod movement above the LPSP. The reliability of the systems is such that testing monthly and

( prior to startup is adequate assurance of proper function.

The purpose of the HPSP is to prevent fuel damage in the event of errorteous rod withdrat from locations of high power density during higher power operation. Notes e and d of Table 4.3.6-1 presently require the channel functional test to be performed daily as power is increased above the HPSP, decreased below the HPSP, and at least once per 31 days while operation continues above the HPSP. The requirement to perform the surveillance when decreasing power below the HPSP serves no purpose.

Since most startups last less than 7 days, the startup surveillance frequency with footnote b will adequately demonstrate operability when increasing above the HPSP. The monthly frequency will remain at least once per 31 days, regardless of power level.

These proposed changes provide operational flexibility, clarify and meet the intent of the present wording, and have been approved at another recently licensed BWR. l C i l

= -, * * * , e

a $ g__

i

. TABLE 4.3.6-1 (Continued)

CONTROL' ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTES:

a. Neutron detectors may be excluded from CHANNEL CALIBRATION.

b. Within 2 t$ prior to startup, if .:t ;:-#: ;;d .;it.'.'- tM f r 4~a fc. ith one our for cont I rod move nt, ess rfo withD th previ s 24 ours and powe is i reas abov the S low L p er s oin and e RPC high ower etpo' t and as p is d eased &

n elow e RP hig power setpo t fo the Jrst me dur ng any 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months peri durf g p r inc ase dec ase.

l l

A least once r 31 ays ileoeratinconpnues th p r abo the

( Q CS 1 6 9 p se nt.f "

C f. The CHANNEL CALIBRATION shall exclude the flow reference transmitters, 1

l these transmitters shall be calibrated at least once per 18 months.

( *With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

l

Calibrate trip unit setpoint at least once per 31 days.

99 9

e PERRY - UNIT 1 3/4 3-60

_,-,._,,,~. ,, .-_,_,-.-,,-n,,e,__ , - , _ . -.,w., ,,.w, _ ,

Justification Technical Specification Table 3.3.7.1-1 Radiation Monitoring Instrumentation Technical Specification Table 3.3.7.1-1, Action 72, presently requires a portable continuous noble gas monitor to be operable in the Control Room when the Control Room Ventilation Radiation Monitor (Noble Gas) is inoperable. This change request would allow the Control Room Area Radiation Monitor (CRARM) to be used to meet the Action requirements.

The CRARM may be used to meet Action 72 for many reasons. Permanent instrumentation may always be used in lieu of portable instrumentation.

Action 72 requires a continuous monitor, which the CRARM is. Neither the CRARM nor any portable continuous noble gas monitor will auto-initiate any equipment, and there is no requirement to do so. Also, the setpoints of both the CRARM and the Control Room Ventilation Radiation Monitor (Noble Gas) are calculated using the same criteria; ensuring the dose to the Control Room operator remains less than SR in a 30 day period following a LOCA.

The CRARM meets all the requirements for a portable continuous noble gas monitor, and full credit should be taken for it in order to comply with Action 72.

1

)

,..s- q .} .a- -, - - .. ,.

TABLE 3.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION l ACTION ACTION 70 - With the required monitor inoperable, obtain and analyze at least one grab sample of the monitored parameter at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In addition, with the Unit 1 Vent noble gas monitor inoperable, restore the inoperable noble gas monitor to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or place the inoperable noble gas monitor in the tripper condition.

ACTION 71 - With the required monitor inoperable, release via this pathway may continue provided grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . .

f or N Cen t rol Roo m Rra R J a % Naitor ACTION 72 -

fWith the required monitor inoperable, assure a portable con-t tinuous noble gas monitor;is OPERABLE in the control room within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the inoperable monitor to OPERABLE status within 7 days, otherwise, initiate and maintain operation of the control room emergency filtration system in the isolation mode of operation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 73 -

With the number of channels OPERABLE less than required by Minimum Channels OPERABLE requirement, release via this pathway may continue for up to 30 days provided:

a. The offgas system is not bypassed, and

b. The offgas post-treatment monitor is OPERA 8LE, and

c. Grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 74 - With the required monitor inoperable, assure a portable area radiation monitor with the same alarm setpoint is OPERABLE in the vicinity of the installed monitor during any fuel movement.

If no fuel movement is being made, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 75 - With the required monitor inoperable, perform area surveys of the monitored area with portable monitoring instrumentation 'at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

PERRY - UNIT 1 3/4 3-64

-- --l

Justification Technical Specification 3.3.7.5 Technical Specification Bases 3/4.3.7.5 Accident Monitoring Instrumentation Technical Specification 3.3.7.5-1, Action 82, presently requires the plant to be placed in Cold Shutdown after some period of time with the number of operable accident monitoring instrumentation channels less than the ,

1 minimum required. This change request would allow the plant to remain at l power provided the valve position could be verified by use of alternate indication and that the instrumentation is returned to operable status the next time the valve is required to be demonstrated operable pursuant to Specification 4.0.5.

Action 82, as presently worded, is excessively punitive. It allows the use of alternate means of verifying valve position, but the time associated with the Action is not consistent with the testing requirements of the valves in accordance with Specification 4.0.5. The primary concern of containment isolation is that the valves isolate.

The control room indication of isolation valve position provides an easy means of determining valve position but it is certainly not the only means. System flow rate and/or pressure may also be used to verify a valve's position.

The surveillance frequency of containment isolation valves are set in accordance with ASME requirements. This frequency should be the basis on which instrumentation operability is required. Requiring a valve to change position in order to perform this position verification is unwarranted and may cause the unit to be shut down.

This proposed change will provide operational flexibility, and still provide adequate assurance of containment isolation valve position.

l

l. . ..

Table 3.3.7.5-1 (Continued)

ACCIDENT MONITORING INSTRUMENTATIONS ACTION STATEMENTS ACTION 80 -

a.

With.the number of OPERABLE accident monitoring instrumentation I channels less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channel (s) to OPERA 8LE status i within 7 days or be in at least HOT SHUTDOWN within the p2xt 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERA 8LE requirements of Table 3.3.7.5-1, restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least, HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ar.d in COLD SHUTDOWN with the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 81 - With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable Channel (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or:

a. Initiate the preplanned alternate method of monitoring the -

appropriate parameter (s), and

b. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the l l

system to OPERA 8LE status. '

ACTION 82 - !

a. th numb o 0 ac d t inst i h n s 4s t n e e i ni ri n tjen' umb C els j 1 3. 7 5 , ry a e( i o by c/a r-a i i ti n t , s re er 1 e n 1(

s u w i 3 s i a 1 st SH

{ i in u i C0 w n Qoll ing 24 urs. f - --

With the number of OPERA 8LE accident monitoring instrumentation !

channels less than the Minimum Channels OPERABLE requirements of Table 3.3.7.5-1, verify the valve (s) position by use of alter-nate indication methods; restore the inoperable channel (s) to OPERA

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INSTRUMENTATION

' l BASES MONITORING INSTRUMENTATION (Continued) 3/4.3.7.4 REMOTE SHUTDOWN INSTRUMENTATION AND CONTROLS l

The OPERASILITY of the remote shutdown monitoring instrumentation'and controls ensures that sufficient capability is available to permit shutdown and maintenance room. of HOT SHUTDOWN of the unit from locations outside of the control This capability is required in the event control room habitability is lost and is consistant with General Design Criteria 19 of 10 CFR 50.

3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0737, " Clarification of TMI Action Plan Requirements." November 1980.p 3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with information of the

(~ status of the neutron level in the core at very low power levels during startup and shutdown. At these power levels, reactivity additions shall not be made without this flux level infomation available to the operator. When the inter-mediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted.

' ; The SRMs are required OPERABLE in OPERATIONAL CONDITION 2 to provide for rod block capability, and are required OPERABLE in OPERABLE CONDITIONS 3 and 4 to provide monitoring capability which provides diversity cf protection to the mode switch interlocks. !

i 3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM J

The OPERABILITY of the traversing in-core probe system with the specified minimum complement of equipment ensures'that the measurements obtained from

{ use of this equipment accurately represent the spatial gama flux distribution i of the reactor core.

l The TIP system OPERABILITY is demonstrated by normalizing all probes '

(i.e., detectors) prior to performing an LPRM calibration function. Monitoring

' core thermal limits may involve utilizing individual detectors to monitor selected areas of the reactor core, thus all detectors may not be _ required to CPERABLE.

The OPERABILITY of individual detectors to be used for monitoring is demonstrated by comparing the detector (s) output with data obtained during the previous LPRM calibrations. "

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Justification Technical Specification Table 3.3.7.10-1 Radioactive Gaseous Effluent Monitoring Instrumentation Table 3.3.7.10-1, Radioactive Gaseous Effluent Monitoring Instrumentation, presently requires the Unit 1 Vent Radiation Monitor Noble Gas Activity Monitor to be operable at all times. This change request would modify what actions need to be taken when the Unit 1 Vent Radiation Monitor Noble Gas Activity Monitor is inoperable in Operational Conditions 1, 2 and 3 1

and Operational Conditions 4 and 5.

In Operational Conditions 1, 2 and 3 it is reasonable to immediately suspend containment drywell purge and vent (M14) in the event of an inoperable noble gas monitor. This is due to the concern of the contribution of the noble gas concentration by the M14 system to the offsite dose. However, in Operational Conditions 4 and 5, this concern does not exist as evidenced by the fact that the M14 system is not required to be operable in Operational Condition 4 and 5 and there is no limit on the hours that containment /drywell purge may be in operation. In addition, there are no requirements for primary containment integrity in operational Condition 4 and 5 unless handling irradiated fuel. It is not, therefore, appropriate to immediately suspend containment /drywell purge and vent (M14) in the event of an inoperable noble gas monitor. It is more appropriate to take grab samples at least one per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the Unit 1 Vent noble gas monitor is inoperable in Operational Conditions 4 and 5.

i

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TABLE 3.3.7.10-1 A

g RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION E MINIMUM CHANNELS q INSTRUMENT OPERABLE APPLICABILITY ACTION

1. OFFGAS VENT RADIATION MONITOR

a. Noble Gas Activity Monitor 1

121

b. Iodine Sampler 1

122 I' c. Particulate Sampler . 1

122

, d. Effluent System Flow Rate Monitor 1

123

$ e. Sampler Flow Rate Monitor 1

123

2. UNIT 1 VENT RADIATION MONITOR o a. Noble Gas Activity Monitor 1 # 62,3 125 sr sat

b. Iodine Sampler 1

122

c. Particulate Sampler 1

122

d. . Effluent System Flow Rate Monitor .1

123

e. Sampler Flow Rate Monitor 1

123 i

3 O N l ,

Justification Technical Specific-tion Table 4.3.7.10-1 Radioactive Gaseous Effluent Monitoring Instrumentation Technical Specification Table 4.3.7.10-1 presently contains a footnote (4)

~~#

on the weekly channel check requirements for the iodine and particulate samplers which monitor the gaseous effluent release paths. This change request would modify the footnote (4) to state that the performance of the 1

weekly channel check, which actually is the changing of the iodine cartridge and the particulate filter, does not render the system inoperable and the applicable Action statements need not be entered.

The effluent radiation monitors are composed of iodine and particualte samplers, and a noble gas activity monitor. The weekly channel check, to change the iodine cartridge and the particulate filter, requires the sample pump to be turned off for a brief period of time. During this j time, the noble gas activity monitor is not analyzing any sample, and therefore cannot peform its intended function. As a result, the Action statement for the noble gas activity monitor must be entered.

Since the brief inoperablity of the noble gas activity monitor is due to required surveillance tests, it seems excessive to be required to enter l the respective Action statements. Technical Specification surveillance tests were never intended to cause unnecessary hardships on plant ;

operations, and this change request provides the relief necessary to continue operation without entering the Actions. )

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. I TABLE 4.3.7.10-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT NONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATION At all times.

During main condenser offgas treatment system operation.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annuciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. Instrument indicates a downscale failure.

3. Instrument controls not set in operate mo,de.

(2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that par-ticipate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended energy and measure-ment range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(3) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

1. One volume percent hydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

(4) The iodine cartridges and particulate filters will be changed at least once per 7 days. fe,q,ma c c. of tUc Ct4 ANN E L- CH ECK SCb Pf f der %c cyc fcus in of cM) e , sa( fh C LLff GbbC h~

, A CM ^l r tntesustr n e ed n o t h e n fer ed, e

PERRY - UNIT 1 3/4 3-95

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Justification Technical Specification 3/4.4.3.2 Operational Leakage Surveillance Requirement 4.4.3.2.2 presently requires the testing of each reactor coolant pressure isolation valve in accordance with Specification 4.0.5, and provides relief from Specification 4.0.4 for entry into Operational Condition 3. This change request would provide relief from Specification 4.0.4 for entry into Operational Condition 2 or 3.

s The relief from Specification 4.0.4 to allow entry into Operational Condition 3 was intended to allow the plant to reach rated pressure and temperature for the purpose of testing the pressure isolation valves. The normal means of reaching rated pressure and temperature at Perry is to enter Operational Condition 2, Startup, and commence a reactor startup.

Heatup is achieved from nuclear heat, not from pump heat or decay heat alone. This change would provide the intended relief during the Operational Condition which Perry would normally use for reactor startup. i The relief for Operational Condition 3 is still required for specific test y applications not resulting in a reactor startup, such as non-nuclear heatup.

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REACTOR COOLANT SYSTEM 4

SURVEILLANCE REQUIREMENTS i

4.4.3.2.1 The reactor coolant systes 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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (not a means of quantifying leakage),

b. Monitoring the drywell floor and equipment sump flow rate at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

c. Monitoring the drywell upper drywell air coolers condensate flow rate at least once per 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months , and

d. Monitoring the reactor vessel head flange , leak detection system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.3.2.2 Each reactor coolant systes pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERA 8LE 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 intoOPERATIONALCONDITIONf.

2. o c 3 s

PERRY - UNIT 1 3/4 4-11

Justification Technical Specification 3.4.9.2 Cold Shutdown Technical Specification 3.4.9.2 is presently applicable in Operational Condition 4. This change request would require Specifcations 3.4.9.2 to be applicable during Operational Condition 4, when heat losses to the ambient are not sufficient to maintain Operational Condition 4.

The bases for this specification is to provide sufficient heat removal capability for the removal of core decay heat, and mixing to assure accurate temperature indication. During those times when core decay heat is low enough that heat losses to the ambient are capable of preventing an increase in reactor coolant temperature, there is no need for the RHR system and coolant recirculation. This change has been approved on several recently licensed BWR's.

O e

REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.9.2 Two# shutdown cooling mode loops of the residual neat removal (RHR) system shall be OPERABLE and, unless at least one recirculation pump is in operation, at least one shutdown cooling mode loop shall be in operation ** N with each loop consisting of at least: -

a. One OPERABLE RHR pump, and

b. T W OPERABLE RHR heat exchangers.

APPLICABILITY: OPERATIONAL CONDITION 4r w/im hed Mrtu d M( dmINnf

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ACTION: 'E Y No Mast /dih S#fM f/ #N A L i

a.

With less than the above required RHR shutdown cooling mode loops OPERABLE, within one hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop.

b.

With no RHR shutdown cooling mode loop or recirculation pump in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pra.nure at least once per hour.

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.9.2 At least one shutdown cooling mode loop of the residual heat removal system, recirculation pump or alternate method shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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m a y h ara 4t #0ne RHR shutdown cool g mode loop ma9,be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other loop is OPERABLE and in operation.

"The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided the other loop is OPERABLE.

,{(MTheshutdowncoolingmodeloopmayberemovedfromoperationd hydrostatic testing.

PERRY - UNIT 1 3/4 4-27

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Justification Technical Specification 3/4.6.1.2 Primary Containment Leakage Surveillance Requirement 4.6.1.2.j presently states the provisions of Specification 4.0.2 are not applicable to various specifications including Specification 4.6.1.2.e. This change requst would delete Specification 4.6.1.2.e from the list under Specification 4.6.1.2.J.

Surveillance Requirement 4.6.1.2.j states the provisions of Specification 4.0.2 do not apply to Specification 4.6.1.2.e, among others, in its entirety. This is not the case. Surveillance Requirement 4.6.1.2.e requires the containment airlocks to be tested and demonstrated operable per Surveillance Requirement 4.6.1.3. Surveillance Requirement 4.6.1.3 contains a footnote which says the provisions of Specification 4.0.2 do not apply to certain portions of the testing requirements (i.e. those required per 10CFR50 Appendix J). The listing of Specification 4.6.1.2.e under Surveillance Requirement 4.6.1.2.j is excessive and is in contradiction with Surveillance Requirement 4.6.1.3.

l w.,- y-w--, , - -

1 CONTAINMENT SYSTEMS

, SURVEILLANCE REQUIREMENTS (Continued)

1. Confirms the accuracy of the test by verifying that the differ-ence between the supplemental data and the Type A test data is within 0.25 L,. The formula to used is:

[L, + L ,- 0.25 L,3 i L cI El *o'm + 0.25 L,] where Le =

supplemental test result; L, = superimposed leakage; L , = ,

measured Type A leakage.

2. Has duration sufficient to establish 1 curately the change in leakage rate between the Type A test and the supplemental test.

3. Requires the quantity of gas injected ir.to the primary contain- l ment or bled from the primary containment during the supple-mental test to be between 0.75 L, and 1.25 L,. l

d. Type 8 and C tests shall be conducted with gas at P , 11.31 psig*, '

at intervals no greater than 24 months except for tlsts involving: ,

1. Air locks,

2. Main steam line isolation valves,

3. Valves pressurized with fluid from a seal system,

4. All containment isolation valves in hydrostatically tested lines

{

per Table 3.6.4-1 which penetrate the primary containment, and

5. Purge supply and exhaust isolation valves with resilient -

material seals.

{ e. Air locks shall be tested and demonstrated OPERABLE per Surveillance i j Requirement 4.6.1.3.

f. Main steam line isolation valves shall be leak tested at least once per 18 months.

g. Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J of 10 CFR 50 Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P maintain,12.44psig,andthesealsystemcapacityisadequateto system pressure for at least 30 days.

I h. All containment isolation valves in hydrostatically tested lines per l Table 3.6.4-1 which penetrate the primary containment shall be leak tested at least once per 18 months.

i. Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE per Surveillance Requirements 4.6.1.8.3. and 4.6.1.8.4.

i

j. The provisions of Specification 4.0.2 are not applicable to Specifications 4.6.1.2.a. 4.6.1.2.b, 4.6.1.2.c, 4.6.1.2.dx""

G. 6.1. z. e .

I "Unless a hydrostatic test is required per Table 3.6.4-1.

PERRY - UNIT 1 , 3/4 6-5

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Justification Technical Specification 3.6.1.3 Primary Containment Air Locks Technical Specification 3.6.1.3 presently provides an Action to be taken with one primary containment air lock door inoperable. This change request is to provide an Action to be taken with one primary containment air lock door in one or both air locks inoperable.

The Perry design provides two air locks into the primary containment each with two doors. Only one door in each air lock is required to be operable f'r o that air lock to perform its intended function and be consistent with analysis. The Action, as presently written, does not permit one air lock j door in one or both air locks to be inoperable, but rather only addresses I the Action to be taken with one primary containment air lock door

inoperable. Therefore, if an air lock door in both air locks was ,

inoperable, there would be no Action to cover this situation.

Specification 3.0.3 states that "when a Limiting Condition for Operation is not met, except as provided in the associated Action requirements, action shall be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place the unit in an Operational Condition in which the Specification does not apply by placing it, as applicable, in:....". This would require placing the unit in cold shutdown. Therefore, to preclude an unnecessary shutdown when an acceptable condition of one air lock door in one or both air locks inoperable exists, the Action requires revision to include "... in one or

) both air locks...". -

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CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION

3. 6.1. 3 Each primary containment air lock shall be OPERABLE with:' l l

a. Both doors closed except when the air lock is being used for' normal l transit entry and exit through the containment, then at least one air

, lock door shall be closed, and

b. An overall air lock leakage rate of less than or equal to 2.5 scf per hour at P,, 11.31 psig.

3 APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, 3, and f.

ACTION:

Jn o,,e e,c. /oon cde Iock5

a. With one primary containment air lock door41noperable: ,

1. Maintain at least the 0FERABLE air lock door closed

and either 1

restore the inoperable air lock door to CPERA8LE status within 24 l hours or lock the OPERABLE air lock door closed. !

i

2. Operation may then continue until performance of the next required i

overall air lock leakage test provided that the OPERA 8LE air lock i door is verified to be locked closed at least once per 31 days.

, 3. Othemise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and l in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4. Otherwise, in OPERATIONAL CONDITION f, suspend all operation involving handling of irradiated fuel in the primary containment, CORE ALTERATIONS, and operations with a potential for draining the reactor vessel.

i 5. The provisions of Specification 3.0.4 are not applicable.

cc i

b. With.the primary containment air lock inoperable in OPERATIONAL CONDI- I TIONS 1, 2, or 3, except as a result of an inoperable air lock door, main-tain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. With dprimary containment air lock inoperable, in OPERATIONAL CONDI- l 1

TION f, except as a result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPER-A8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or suspend all operations involving handling of irradiated fuel in the primary containment, CORE ALTERATIONS, and ope-rations with a potential for draining the reactor vessel. ,

j

When handling irradiated fuel in the primary containment, during CORE l.

ALTERATIONS, and operations with a potential for draining the reactor vessel.

( *Except during entry to repair an inoperable inner door, for a cumulative time not to exceed I hour per year. -

i i PERRY - UNIT 1 . 3/4 6-6 I l

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4 4

Justification

! Technical specification Table 3.6.4-1 l Containment and Drywell Isolation Valves i

! Technical Specification Table 3.6.4-1 presently requires the containment .

f and drywell isolation valves to be operable during specified operational 1

conditions. This change request would provide relief for two containment l isolation valves, and not require them to be operable while handling

! ' irradiated fuel in primary containment, during core alterations, and i operations with a potential for draining the reactor vessel (Operational Condition **).

4 t

The operability of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in 3

$ the event of a release of radioactive material to the containment

atmosphere or pressurization of the containment. In Operation Condition i ** , pressurization of the containment is not a concern, due to the initial conditions of the plant. The two valves (1P54-F726 and 1P54-F727) will i

i still be capable of isolating the containment atmosphere from the outside l environment, even while remaining opened.

l These valves provide the containment isolation for the fire protection l water lines. During normal system operation, the pipeline upstream and downstream of these valves is pressurized with water at approximately 75

! psig. Since this is considerably greater than the peak calculated ;

accident pressure for containment (P,, 11.31 psig), an increase in containment leakage will not occur with these valves open. a The only time the plant will be in Operational Condition ** is when it is

. in Cold Shutdown or Refueling. At this time, many personnel will be f inside containment performing surveillances, repairs, and maintenance activities. Should one of these activities result in a fire inside i

j containment, plant instructions (ONI-P54, Fire) direct the operator to I

l open these valves to provide the water source to the hose reels.

With the approval of this change request, the hose reels and standpipes inside the containment could be in a state of standby readiness in the event of a fire while in Operational Condition **.

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b. CONTAll#4ENT MANUAL ISOLATION VALVES (Continued)

M E Valve Penetration Valve gc) Maximum Secondary Test

-< Number Number Group Isolation Time Containment Pressure (Seconds) Bypass Path (Psig) '

g (Yes/No)

~

, [ IN27-F751 P106, P107, NA NA Yes- 11.31 -

P115, P429 f P305 NA 3 IPS3-F030 IP53-F035I ' *)) P305 NA 3 No No 11.31 11.31 IPS3-F040I ')

~

. P312 NA 3 No 11.31 IPS3-F045 I ') P312 NA 3

No 11.31 IP53-F536 /

F570 P305 NA NA Yes 11.31 1PS3-F541 /

F571 P312 NA NA Yes 11.31 R IP54-F726 P406 NA NA Yes 11.31

IP54-F727 @ P406 NA NA Yes 11.31 IP57-F015A P304 NA 15* No 11.31 IP57-F0158 P116 NA 15* No 11.31

! : IP87-F037I ') P401 NA 3 Yes (b)

IP87-F065 P318 NA 3 , Yes (a)

IP87-F071 P318 NA 3 Yes (a)

IP87-F074f ') P318 NA 3 Yes (a)

IP87-F077I ') P318 NA 3 Yes (a) f P413 NA 3 Yes 11.31 1P87-F049 1P87-F055 I)

') P413 3 I NA Yes 11.31 1P87-F046 ') P413 NA 3 Yes 11.31 IP87-F052I ')

P413 NA 3 Yes 11.31

, IP87-F083I ') P106, P107 NA 3 Yes 11.31 P115, P429 IP87-F264I ') P106, P107 NA 3 Yes 11.31 P115, P429 ,

i Justification Technical Specification 3.6.5.2 I Containment Humidity Control 1

Technical Specification 3.6.5.2, Containment Humidity Control, presently is applicable whenever primary containment integrity is required for Specification 3.6.1.1.1. This change request would require that Specification 3.6.5.2 be applicable whenever primary containment integrity is required for Specification 3.6.1.1.1 and 3.6.1.1.2.

) Technical Specification 3.6.5.1, Containment Vacuum Breakers, is required j to be operable whenever containment integrity is required for Specifications 3.6.1.1.1 and 3.6.1.1.2. The design basis for containment l 1

vacuum breakers includes assumptions with respect to containment humidity l l during those times when the containment vacuum breakers are required to be j operable.

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CONTAINMENT HUMIDITY CONTROL i .

LIMITING CONDITION FOR OPERATION 3.6.5.2 Containment average temperature and relative humidity shall be main-tained above the curve shown in Figure 3.6.5.2-1. ,

, APPLICABILITY: Whenever PRIMARY CONTAINMENT INTEGRITY is required for 4

Specification,3.6.1.1.1A R ne( 3. 6. l.1, 2 ACTION:

i i

With the containment average temperature / relative humidity not within the limits for acceptable operation as shown in Figure 3.6.5.2-1:

a. In OPERATIONAL CONDITION 1, 2 or 3, restore the average temperature /

relative humidity to within the limits for acceptable operation as shown in Figure 3.6.5.2-1 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within.the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. At all other times, either:

1. Maintain an unobstructed opening (s) in the containment that equals or exceeds the flow area provided by two open vacuum breakers, or

, 2. Deactivate the containment spray by closing at least one valve in each containment spray supply header and deenergizing the power supply to its motor operator.

SURVEILLANCE REQUIREMENTS 4.6.5.2 Containment average temperature / relative humidity shall be verified to l

be within the limits for acceptable operation curve shown in Figure 3.6.5.2-1:

a. At least once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

, b. By verifying the temperature instrumentation OPERABLE by performance j of a:

1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , i

2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and

3. CHANNEL CALIBRATION at least once per 18 months.

h 1

4 l PERRY - UNIT 1 3/4 6-44

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k Justification Technical Specification 4.6.5.2 Containment Humidity Control L

b j Technical Specification 4.6.5.2 presently requires surveillance of the containment temperature instrumentation at various intervals. This change request would delete these testing requirements from this technical specification.

I The intent of Technical Specification 4.6.5.2a and b is to ensure the !

temperature is within the limits of temperature / relative humidity for acceptable operation. The requirements for the channel check are performed by Technical Specification 4.6.1.7, Primary Containment Average Air Temperature. Channel functional tests on instrumentation which has no j alarm and/or trip functions serves no purpose as it is not consistent with the technical specification definition of Channel Functional Test. The channel celibration requirement is covered by the Perry Plant Maintenance Program. In addition, Technical Specification 4.6.1.7 requires a minimum number of operable air temperature detectors. Therefore, this change

! would delete redundant testing requirements and provide for consistency of 1

j testing of the same instrumentation serving multiple functions.

4 I

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1 l

CONTAINMENT HUMIDITY CONTROL

. LIMITING CONDITION FOR OPERATION 3.6.5.2 Containment average temperature and relative humidity shall be main-tained above the curve shown in Figure 3.6.5.2-1.

APPLICA8ILITY: Whenever PRIMARY CONTAINMENT INTEGRITY is required for Specification 3.6.1.1.1.

ACTION:

With the containment average temperature / relative humidity not within the limits for acceptable operation as shown in Figure 3.6.5.2-1:

a. In OPERATIONAL CONDITION 1, 2 or 3, restore the average temperature /

relative humidity to within the limits for acceptable operation as shown in Figure 3.6.5.2-1 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. At.all other times, either:

1. Maintain an unobstructed opening (s) in the containment that equals or exceeds the flow area provided by two open vacuum -

breakers, or

2. Deactivate the containment spray by closing at least one valve in each containment spray supply header and deenergizing the power supply to its motor operator.

SURVEILLANCE REQUIREMENTS 4.6.5.2 Containment average temperature / relative humidity shall be verified to be within the limits for acceptable operation curve shown in Figure 3.6.5.2-y>C-])

[ t least once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. By verlfying th'e temperature instrumentation OPERABLE by performance of a:

1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and

3. CHANNEL CALIBRATION at least once per 18 months.

)

s PERRY - UNIT 1 3/4 6-44 .

m

Justification Technical Specification 4.7.4.e.1 Snubbers Technical Specification 4.7.4.e.1 presently requires functional testing of snubbers. The initial sample size requirement is 10% of the total number of subject snubbers in the plant. This proposed change would reduce the required number of additional snubbers to be tested from 10% to 5% for every f ailed snubber discovered during functional testing.

In the absence of a suitable snubber failure data base, it was required that for every failed snubber, an additonal 10% of that snubber type was to be tested. Subsequently, the ASME OM4 group developed a sampling plan which determined that only 50% of the initial sample size (10% in this case) need be tested for each failed snubber. This change has been approved on several recently licensed BWRs. 1 i

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PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

e. Functional Tests During the first refueling shutdown and at least once per 18 months '

thereafter during shutdown, a representative sample of snubbers shall be tested using one of the following sample plans for each type of i

snubber.

The sample plan shall be selected prior to the test period and cannot be changed during the test period. The NRC Regional Admin-istrator shall be notified in writing of the sample plan selected prior to the test period or the sample plan used in the prior test j period shall be implemented: ,

i l)' ' 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 of snubber shall be functionally tested until no m) ore failures are0rof th found ororuntil all snubbers of that type have been functionally tested; 1

2) A representative sample of each type of snubber shall be functionally tested in accordance with Figure 4.7.4-1. "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 I day's total Figure plus current day's increments) shall be plotted on 4.7.4-1.

If at any time the point plotted falls on or above tested. If at any all the " Reject" line snubbers of that type shall be functionally time the point plotted falls on or below the i " Accept" line, testing of snubbers of that type may be te ainated, 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.

during functional testing may invalidate that day s testing andTestin

, 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 i 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 i l

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 a: 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 falls e above the " Accept" line, testing must continue until the point falls on or below the " Accept" line or all the snubbers of that type have been tested.

PERRY - UNIT 1 3/4 7-10

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Justification Technical Specification 3/4.7.6 Main Turbine Bypass System Surveillance Requirement 4.7.6.a, presently requires each turbine bypass valve be cycled through at least one complete cycle at least once per 7 days. This change request would extend the surveillance test frequency to at least once per 31 days.

General Electric Company, the manufacturer of the main turbine at Perry, has issued a Technical Information Letter (TIL-969, May 22, 1984) on surveillance test intervals for main turbine bypass valves. This TIL recommended the surveillance frequency of turbine bypass valves be extended from 7 days to 31 days. GE has extensive experience in the design, operation and maintenance of large steam turbines, so this change would be in secordance with manufacturer's recommendations. In addition, the change has been approved on all recently licensed BWR's.

+

PLANT SYSTEMS 3/4.7.6 MAIN TUR8INE BYPASS SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6 The main turbine bypass system shall be OPERA 8LE.

APPLICABILITY: OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With the main turbine bypass system inoperable, restore the system to 6EllD@LE status within I hour or reduce THERMAL POWER to less than 25 THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REOUIREMENTS 4.7.6 The main turbine bypass system shall be demonstrated OPERA 8LE at least once per:

( a. / days by cycling each turbine bypass valve through at least one complete cycle of full travel, and '

b. 18 months by:

1. Performing a system functional test which includes simulated automatic actuation and verifying that each automatic valve actuates to its correct' position.

2. Demonstrating' TURBINE BYPASS SYSTEM RESPONSE TIME meets the following requirements when measured from the initial movement of the main turbine stop or control valve:

a) 80% of turbine bypass system capacity shall be established in less than or equal to 0.3 seconds, b) Bypass valve opening shall start in less than or equal to 0.1 seconds.

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O PERRY - UNIT 1 3/4 7-16 '

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i Justification Technical Specification Table 3.8.4.1-1 Containment Penetration Conductor Overcurrent Protective Devices Technical Specification Table 3.8.4.1-1 lists the containment penetration conductor overcurrent protective devices for Perry Unit 1. This change request would modify this list to only include those overcurrent protective devices connected to active devices inside the containment.

Technical Specification Table 3.8.4.1-1 presently lists all circuit breakers protecting containment electrical penetrations. Included in this list ace the circuits which feed the space heaters (Sp. Htr.) which were used to maintain the environmental qualifications of equipment during plant construction. Some space heaters are no longer used, and the supply circuit breakers have been administratively locked open. Since these breakers see no power, the testing requirements should be deleted as no potential exists for the penetration to burn out and become non-isolable.

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TABLE 3.8.4.1-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES 13.8 KV LOAD OVERCURRENT PROTECTION

~

Primary Secondary 1833-C001A L1106 1R22-5012 IB33-C0018 L1205 1R22-5013 120V LOAD O'R' CIRCUIT 1021 01X (Op. Ntr.) 1R25 0007 001 NA*

1 21 a;X (3r. at..) 1;23-;037-ca2 NA*

1321-23X (3p. at. . ) 1;23-3033-co; NA-1921-8760XS 1R25-5043-CB20 -

NA*

IB21-B756XB 1R25-5043-CB18 NA*

IB21-B758XS 1R25-SO43-CB19 NA*

1821-8754XB 1R25-5043-CB17 NA*

IB21-B752XB 1R25-5047-CB11 NA*

1933-B2X (Sp. Htr.) 1R25-5093-CB7 NA*

IB33-84X (Sp. Htr.) 1R25-5093-CB8 NA" 1833-86X (Sp. Htr.) 1R25-5097-CBS NA*

IS33-B8X (Sp. Htr.) 1R25-5097-CB6 NA*

IB33-810X (Sp. Htr.) 1R25-5093-CB9 NA*

1833-B12X (Sp. Htr.) 1R25-SO93-CB10 NA*

1833-814X (Sp. Htr.) 1R25-5097-CB7 NA*

IB33-816X (Sp. Htr.) 1R25-5097-CB8 NA*

1000 017X (;p. at. . ) 1:2;-3033-c;11 NA*

IB33-B13X (3p. Hir. ) 1R23-5037-CB3 nA-1B33-B21X (Sp. Hir. ) 1R20-3033-C34 NA*

1000 02;X (Sp. Hi. . ) 1R25-3037-CG1C NA-1c11 02:0X p. at..; 1R2;-Sc33-ca; aA*

1Cll-CIX NA* 1H13-P653-CB1 1C41-89XB (Sp. Htr) 1R25-5043-CB21 NA*

1E51-83XB 1R25-5043-CB24 NA*

1E51-BlXB 1R25-5043-CB23 NA*

1F42-85X (Sp. Htr.) 1R25-5097-CB3 NA*

1G33-B1X (3p. Hir. ) 1R23-SC35-CS NA*

1630 03X (Sp. Hi. . ) 1R2; S081-CS2 NA*

1G33 30X (Sp. HL. . ) 1R20 3073 CE NA*

1G33-87X (Sg. H L. . ) 1R2; SG73 002 NA*

10 3-23X (sg. H L. . ) 1R2;-3051-Cal HA-

" Protected by fuse.

PERRY - UNIT 1 3/4 8-23

l 1

1 TABLE 3.8.4.1-1 (Continued) 120V LOAD OR CIRCUIT OVERCURRENT PROTECTION Primary Secondary 1000 011X (Sp. Utc.) 1R25 5070 000 NA*

1G00 010X (0p. Hic.) 1R25 0070 004 NA* ,

1G33-315X (;g. H L. . ) 1R2; 5070 005 NA*

IC33 317X (Sg. Hi. . ) 1R2; 5070 000 NA' 1G41-81X -

1R25-5077-CB1 NA*

1R25-8516X '0R25-5054-CB7 NA*

1R25-B517X OR25-5054-CB13 NA*

1R25-8245X 1R25-5057-CB12 NA" 1P56-B1060X 1R25-5053-CB34

NA*

IP57-83XB 1R25-5043-CB15 NA*

1R25-8522X 1R25-S153-CB13 NA*

1R25-BS15X 1R25-5053-CB25 NA*

IM16-B7XB 1R25-5047-CB1 NA*

Id16-89XB 1R25-5047-CB3 NA*

IM16-B17XB 1R25-5047-CBS NA" 1M16-B19XB 1R25-5047-CB6 NA*

IE12-83XB 1R25-5047-CB7 NA*

IE12-87XB 1R25-5047-CB8 NA*

1E12-811XB 1R25-5047-CB9 NA*

1E12-B15XB 1R25-5047-CB10 NA*

" Protected by fuse.

PERRY - UNIT 1 3/4 8-24

i l

Justification Technical Specification 3.9.11.2 Low Water Level i

Technical Specification 3.9.11.2 is presently applicable in Operational Condition 4, when irradiated fuel is in the reactor vessel and water level is less than 22 feet 10 inches above the top of the reactor pressure vessel flange. This change request would require Specification 3.9.11.2

! to also be applicable when heat losses to the ambient are not sufficient i to maintain Operational Condition 5.

l

)

The bases for this specification is to provide sufficient heat removal capability for the removal of core decay heat, and mixing to assure accurate temperature indication. During those times when core decay heat j is low enough that heat losses to the ambient are capable of preventing an-I increase in reactor coolant temperature, there is no need for the RHR system and coolant recirculation. This change has been approved on several recently licensed BWRs.

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REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11.2 Two shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE and at least one loop shall be in operation," with each loop consisting of at least:

a. One OPERABLE RHR pump, and 2

b. Two OPERABLE RHR heat exchangers.

APPLICABILITY: OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is less than 22 feet 10 inches above the top of the reactor pressure vessel flange x and h e4.f losses to t/re 4ex4/enf are no f ACTION: ' ' ' ONO#N "N# MOW 5

a. With less than the above required shutdown cooling mode loops of the RHR system OPERABLE, within one hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop.

b. With no RHR shutdown cooling mode loop in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour.

SURVEILLANCE REQUIREMENTS - 4.9.11.2 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be verified to be in operation and circulating :

reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . '

l l

l "The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period.

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PERRY - UNIT 1 . 3/4 9-17

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Justification Technical Specification 3.10.1 Primary Containment Integrity /Drywell Integrity l Specification 3.10.1 allows certain provisions of containment integrity and drywell integrity to be suspended to permit the reactor mode switch to be placed in the startup position with the reactor vessel head and the drywell head removed. The applicability of this specification is Operational Condition 2, during low power Physics Tests. This change-request would allow suspension of these same requirements during the conduct of shutdown margin demonstrations as well.

Specification 3.10.1 presently places restrictions on reactor coolant temperature and rated thermal power, with actions required if these limits are exceeded. Shutdown margin demonstrations can easily be performed I within the bounds of the present restrictions. Shutdown margin is described as a physics test in Chapter 14 of the FSAR, but it is conducted in Operational Condition 5 in accordance with Specification 3.10.3. This change would revise the applicability of Specification 3.10.1 to allow its provisions to apply to the shutdown margin demonstrations. This change makes Specification 3.10.1 consistent with Specification 3.10.3 1

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l 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY /DRYWELL INTEGRITY LIMITING CONDITION FOR OPERATION o.Mk.1,% $ '

3.10.1 The provisions of Specifica ons 3. 6.1.1.1, 3. 6.1. 2, 3. 6.1. 3, 3. 6. 2.1, 3.6.2.3, 3.6.5.1, 3.6.5.2,aad 3.9.1 and Table 1.2 may be suspended to permit the reacter pressure vessel closure head and the drywell head to be removed and the drywell air lock door to be open when the reactor mode switch is in the Startup position during low power PHYSICS TESTS with THERMAL POWER less than 1% ;

of RATED THERMAL POWER and reactor coolant temperature less than 200*F.

1 APPLICABILITY: OPERATIONAL 40NSEf9N+; during icw power PHYSICS TESTS 3ec '

i . ACTION: 2 ** # '"

c h an teattens, With THERMAL POWER greater than or equal to 1% of RATED THERMAL-POWER or '

with the reactor coolant temperature greater than or equal to 200*F, -

1 immediately place the reactor mode switch in the Shutdown position.

~ -

N SURVEILLANCE REQUIREMENTS -

j 4.10.1 The THERMAL POWER and reactor coolant temperature shall be verified to

, be within the limits at least once per hour during low power PHYSICS TEST 5 4 or

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J Justification Technical Specification Bases 3/4.10.3 Shutdo#n Margin Demonstrations Specification 3.10.3 allows certain provisions of the technical specifications to be suspended in order to perform shutdown margin demonstrations. The bases indicates shutdown margin demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur. The bases for this special test j exception is not accurate. A description of shutdown margin demonstration is included in Chaper 14 of the Perry FSAR and the GE Startup Test Specification. Both documents indicate that once shutdown margin has been demonstrated, rods are continued to be withdrawn until criticality is achieved. This is consistent with industry practice. The change to Bases 3/4.10.3 would accurately describe that criticality is properly monitored and controlled, not prevented.

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BASES 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY /DRYWELL INTEGRITY The requirements for PRIMARY CONTAINMENT INTEGRITY and DRYWEll. INTEGRITY are not applicable during the period when open vessel tests are being performed during the low power PHYSICS TESTS.

! 1 3/4.10.2 ROD PATTERN CONTROL SYSTEM l

In order to perform the tests required in the technical specifications it ;

is necessary to bypass the sequence restraints on control rod movement. The additional surveillance requirments ensure that the specifications on heat generation rates and shutdown margin requirements are not exceeded during the period when these tests are being performed and that individual rod worths do '

not exceed the values assumed in the safety analysis.

3/4.10.3 SHtfTDOWN MARGIN DEMONSTRATIONS _ . _

) Performance of shutdown margin demonstrations with the vessel head removed j requires additional restrictions in order to ensure that criticality tr: xt j j crr. These additional restrictions are specified in this LCO. /f 3/4.10.4 RECIRCULATION LOOPS 48ANbM med marm/M' This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.

3/4.-10.5 TRAINING STARTUPS -

This special test exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling RCS temperature with one RHR subsystem aligned in the shutdown cooling mode in order to minimize contaminated water discharge to the radioactive waste disposal system.

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PERRY - UNIT 1 8 3/4 10-1 I y--.-- - .e-,m.e--,..-.$),.-.--my------w,.. - - . , ,-__,y-,m-,.--.. - . _ . , _ m--_--- - - . - - - . . , , . --.__,.,-m , -

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Justification Technical Specification Table 3.12.1-1, Radiological Environmental Monitoring Program i Technical Specification Table 3.12.1-1, Radiological Environmental

! Monitoring Program, Item 4.a, presently requires milk samples to be collected in certain locations within specific distances of the plant

site. This change request would increase the distance in which milk samples may be collected to fulfill the requirements of the sampling program.

To meet the requirements of the sampling program, CEI purchases milk samples f rom private individuals near the plant site. Since public ,

participation in this program is voluntary some individuals no longer wish to participate, and there are no longer sufficient participants to meet I the requirements of the technical specification. To be capable of obtaining milk samples now, and to avoid possible changes in the future, we are increasing the distances in which we are allowed to collect samples. The control location distance will remain unchanged.

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m TABLE 3.12.1-1 (Continued)

! g RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM i e g Number of Samples y Exposure Pathway and Sampling and y and/or Sample Sample Locations gg) Type and Frequency Collection Frequency .of Analysis

! 3. Waterborne (Contfoued) -

b. Drinking One sample of each of one to Composite sample I-131 analysis on each three of the nearest water over 2-week period (5) composite when the dose s

supplies that could be when I-131 analysis affected by its discharge. calculated from the consump-is performed; monthly tion of the water is greater

. composite otherwise.

l One' sample from a control than 1 arem per year.(6) Con-location. posite for gross beta and

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gamma isotopic analysesI4) a, monthly. Composite for tritfun analysis quarterly.

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c. Sediment One sample from area wIth Semiannually. Gamma isotopfc analysisI4) from existing or potentfal sentannually.

shoreline recreat1onal value.

. 4. Ingestion

a. Milk Samples from sliking animals Seminenthly when Gamma isotopicI4) and I-131 In three locations within animals are ori analysis semimonthly when

/0 A km distance having the pasture; monthly at animals are on pasture;-

highest dose potential. If other times. monthly at other times.

there are'none, then one

, sample from allking animals h of three areas between

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i th:2 1 : ;. ;:r yr. One jgj,,,j ,f,,,c po c,,pj /, g sample from allking animals at a control location .

} 15 to 30 km distant and in the j least prevalent wind direction.

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1 Justification j Technical' Specification Figure 6.2.1-1, Figure 6.2.2-1 i

Technical Specification Figure 6.2.1-1 and Figure 6.2.2-1 presently describe the corporate and unit staff organizations at Perry. The revised

]

l figures reflect a realignment of the Perry organization while maintaining i

j the existing functions (ref. letter M. R. Edelman to R. M. Bernero, l PY-CE1/NRR-0471L, May 30, 1986).

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4 Justification Technical Specification 6.5.1 Plant Operations Review Committee (PORC)

~

Technical Specification 6.5.1 presently describes the requirements of the PORC at Perry. This change request would revise the composition due to a recent organizational realignment (ref. letter M. R. Edelman to R. M. Bernero, PY-CEI/NRR-0471 L, May 30,1986) i 1

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l ADMINISTRATIVE CONTROLS

( 6. 4 TRAINING i

6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Perry Training Section General Super-visor, and shall meet or exceed the requirements and recommendations of Sec-tion 5.5 of ANSI N18.1-1971 and Appendix A of 10 CFR Part 55 and the supplemen-

tal requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all Itcensees, and shall include familiarization with

. relevant industry operational experience.

6. 5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC)

I FUNCTION 6.5.1.1 The PORC shall function to advise the Managers, Perry Plant Departments, on all matters related to nuclear safety. .

nypnggygny' V!s - C/mik<nl Man,4 *r htEr k YM " '" h *" S 6.5.1.2 The PORC shall be composed of the: 5#9for eer Chairman: Manager, Perry Plant Operations Department Vice-Chairman / Member: Manager, Perry Plant Technical Department Vice-Chairman / Member: Technical Superintendent, Perry Plant Technical Department

( Member: General Supervisor, Operations Section Member: General Supervising Engineer, Technical Section j Member: General Supervisor, Maintenance Section Member: Reactor Engineer Member: General Supervising Engineer, Radiation Protection Section Member: Plant Health Physicist Member: General Supervising Engineer, Instrumentation

%w and Control Section ,

ALTERNATE 5 S*"W fY '"'")

C.v h uce NA Engih ut, Atc w nny &

6. 5.1. 3 All alternate members shall be appointed in writing by the PORC Chaiman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PORC activities at any one time.

MEETING FREQUENCY j

6.5.1.4 The PORC shall meet at least once per calendar month and as convened by the PORC Chairman or his designated alternate. l l l QUORUM i !

6.5.1.5 The quorum of the PORC necessary for the performance of the PORC i

responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least four

members including alternates.

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PERRY - UNIT 1 6-8 tiewk te :

i

(,merd f ywNos) defideer, Oah)e lhMn vif .

$'es+Io n . 1

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4 Justification Technical Specification 6.9.1.8 Monthly Operating Reports 1

I l Technical Specification 6.9.1.8 requires routine reports of operating

! statistics and shutdown experience to be submitted on a monthly basis to i

the Director, Of fice of Management and Program Analysis, U.S. Nuclear f j

{ Regulatory Commission, Washington, D.C. 20555, with a copy to the i

i Regional Administrator of the Regional Office no later than the 15th of

! each month following the calendar month covered by the report.

The attached change would require the reports to be sent to the Director, Division of Automated Information Services, U.S. Nuclese Regulatory Commission, Washington, D.C. 20555.

i The Office of Management and Program Analysis has been disbanded. The 1

j change assures the reports reach their intended location in the NRC.

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ADMINISTRATIVE CONTROLS SENIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continue made during the reporting period to the PROCESS CO OFFSITE DOSE CALCULATION MANUAL (ODCM), pursuant to Specification respectively, as well as any to major change to Liquid, Gaseous, or Solid Radwaste 4

Treatment Systems pursuant Specification 6.15.

tified by the Land Use Census pursuant to Specification 3.

The Semiannual Radioactive Effluent Release Reports i, hall also include the following:

an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected w leading to liquid holdup tanks exceeding the limits of Specification...31114 .

MONTHLY OPERATING REPORTS f Do've% % oh /VA'um S 6.9.1.8 Routine reports of operating statist l n&# sw.w be submitted on a monthly basis to the Director,~ff L m'nd shutdown experience shall Prew 1. . ~. .g ;, r,d

^='pis, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, 4

with a copy to the Regional Administrator of the Regional Office no later than the 15th of each month following the calendar month covered by the report .

( _SPECIAL REPORTS 6.9.2 Regional Office within the time period specified for each repo 6.9.3 Safety / relief valve failures will be reported to the Regional Administrator system within 30 of the Regional Office of the NRC via the Licensee Event Report days.

6.10 RECORD RETENTION 6.10.1 i In addition to the applicable record retention requirements of Title at least theofminimum 10, Code Federalperiod Regulations, the following records shall be retained for indicated.

6.10.2 The following records shall be retained for at least 5 years:

a.

. Records power and logs of unit operation covering time interval at each level.

1 b.

Records and logs of principal maintenance activities, inspections, nuclearand repair, replacement of principal items of equipment related to safety,

c. All REPORTA8LE EVENTS.

1 d.

Records of surveillance activities, inspections, and calibrations required by these Technical Specifications.

PERRY - UNIT 1 6-21 ~

P f

k Justification i

j Technical Specification 6.9.4 Special Reports i

License Condition 2.C.6 requires that CEICO comply with the requirements l of the fire protection program as specified. This proposed change would specify what reporting is required if the requirements of the fire f

I protection program are not met.

i The addition of Specification 6.9.4 will clarify the reporting required , ;

j should the requirements of the fire protection program not be met. !

]

l 6.9.4 Violations of the requirements of the fire protection program 3

j described in the Final Safety Analysis Report which would have

) adversely affected the ability to achieve and maintain safe shutdown in the event of a fire shall be reported to the Regional i

j Administrator of the Regional Of fice of NRC via the Licensee Event ,

! Report system within 30 days.

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ADMINISTRATIVE CONTROLS k

SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued made during the reporting period to the PROCESS CON OFFSITE DOSE CALCULATION MANUAL (00CM), pursuant to Specifications 6.

respectively, as well as anytomajor change6.15.

to Liquid, Gaseous, or Solid Radwaste Treatment Systems pursuant Specification tified by the Land Use Census pursuant to Specification 3.1 The Semiannual Radioactive Effluent Release Reports shall also include the following:

an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected w leading to liquid holdup tanks exceeding the limits of Specification 3.11.14 ..

MONTHLY OPERATING REPORTS

6. 9.1. 8 be submitted on a monthly basis to the Director, Office of Ma Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C

. 20555, with a copy to the Regional Administrator of the Regional Office no later than the 15th of each month following the calendar month covered by the report .

i SPECIAL REPORTS 6.9.2 Regional Office within the time period specified for each repor 6.9.3 Safety / relief valve failures will be reported to the Regional Administrator ofdays. the Regional Office of the NRC via the Licensee Event Report system5%

6,9f within

/h 3p';fkek m a -f 6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title at least theofminimum 10, Code Federal Regulations, the following records shall be retained for

period indicated. ,

6.10.2 The following records shall be retained for at least 5 years:

a.

Records and logs of unit operation covering time interval at each power level, b.

Records and logs of principal maintenance activities, inspections, repair, and nuclear replacement of principal items of equipment related to safety.

c. All REPORTABLE EVENTS.

d.

Records of surveillance activities, inspections, and calibrations required by these Technical Specifications.

PERRY - UNIT 1 6-21 *

ML20199E755

Text

Dear Mr. Denton:

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