Text

Proposed Improved Tech Specs
	ML20108A258
Person / Time
Site:	Perry
Issue date:	04/26/1996
From:	; CENTERIOR ENERGY
To:	;
Shared Package
ML20108A249	List: ML20108A258; PY-CEI-NRR-1999, Application for Amend to License NPF-58,revising Improved Tech Specs.Changes Either Administrative in Nature or Reflect Minor Technical Errors Discovered During Implementation Process;
References
	NUDOCS 9605020193
	Download: ML20108A258 (43)
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py.cruwan.i m t SLC System Atixhmen:2 Page I of 34 3.1.7 3.1 REACTIVITY CONTROL. SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LC0 3.1.7 Two SLC subsystems shall be OPERABLE.

,a.

~

APPLICABILITY:

MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION

. (,,,.! COMPLETION TIME A.

One SLC subsystem

-A.1 RestorezSLC 7 days inoperable.

subsystem to OPERABLE status.

s Two SLC subsystems B.1 Restore one SLC 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months i operable subsystem to OPERABLE status.

C.

Required Action and C.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met.

PERRY - UNIT 1 3.1-20 AmendmentNo.fy 9605020193 960426 PDR ADOCK 05000440 P

PDR

PY<E1/NRR 1999L Control Rod Block Instrumsntation Page 2 of 34 3.3.2.1

.s Table 3.3.2.1 1 (page 1 of 1)

Control Rod Block Instrumentation m.=me===

APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS REQUIREMENTS 1.

Rod Pattern control System a.

Rod withdrawat limiter (a) 2 SR 3.3.2.1.1 SR 3.3.2.1.6 SR 3.3.2.1.9 h

I 4

-Q SR 3.3.2.1.2 SR 3.3.2.1.5 SR 3.3.2.1.7

heg, SR 3.3.2.1.9 b.

Rod pattern controtter

^\\

2 2

SR 3.3.2.1.3

'tg SR 3.3.2.1.4 gy j

SR 3.3.2.1.5 4uh'Kr gl SR 3.3.2.1.7 SR 3.3.2.1.9 2.

Reactor Mode Switch -Shutdown Position (d) 2 SR 3.3.2.1.8 (a) THERMAL POWER > 70% RTP.

(b) THERMAL POWER > 35% RTP and 5 70% RTP.

(c) With THERMAL POWER s 20% RTP.

(d)

Reactor mode switch in the shutdown position.

1 l

PERRY - UNIT 1 3.3-19 AmendmentNo.//

l PY-CEI/NRR-1999L

.PAM Instrumentation Page J of 34 3.3.3.1 Table 3.3.3.1 1 (page 1 of 1)

Post Accident Monitoring Instrumentation m::

CONDITIONS REFERENCED FROM FUNCTIDN REQUIRED REQUIRLD CNANNELS ACTION D.1 1.

Reactor Steam Dome Pressure 2

E 2.

Reactor Vesset Water Levet-Wide Range 2

+

E 3.

Reactor Vessel Water Level-fue! Zone 2

E 4 Suppression Poot Water Level 2

5 E

S w ession Poot Sector Water Tenperature I

2.C E

6.

Drywell Pressure s,,,

2 7 Drywett Air Tenperature E

2 8.

E.

Primary Containment /Drywell Area Cross Gersna Radiation Monitors 2

F 9.

Penetration flow Path, PCIV Position

10. Primary containment a5 Drywell H T 2perpenet[gtg E

flow path Concentration Analyzer and Monitor]

4 b

2

11. Primary Containment Pressure E

2

12. Primary Containment Air Tenperature E

2 E

(a)

Not required for isolation valves whose associated penetration flo w path is isolated.

(b) control room indication channel.Only one position indication channet is required for penetration fl ow paths with only one instatted (c)

Monitoring each of eight sectors.

PERRY - UNIT 1 3.3-23 Nrandment No. g5

PY-CEl/NRR.1999L ECCS Instrumentation l

Pue 4 cf 34 3.3.5.1 i

Tabte 3.3.5.1 1 (page 3 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS i

i MODES OR REFERENCED l

DTHER REQUIRED FROM SPECIFIED CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE l

FUNCTION CONDITIONS FUNCTIDW ACTION A.1 REQUIREMENTS VALUE 9

2.

LPCI B and LPCI C SLAsystans (continued) e.

LPCI Ptsp B 1,2,3, 1 per pum E

SR 3.3.5.1.1 a 1450 spm i

and LPCI Ptmp C Discharge 4(a) 5(a)

SR 3.3.5.1.2 Itow-Low SR 3.3.5.1.3 (Bypass)

SR 3.3.5.1.5 SR 3.3.5.1.6

f. Manual Init(stian 1,2,3, 1

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

High Pressure Core Spray (HPCS) System a.

Reactor vesset 1,2,3, 4

B SR 3.3.5.1.1 a 127.6 inches l

Dater Level-Low Low, Level 2 4(a) 3(a)

SR 3.3.5.1.2 gg 3,3,$,g,y SR 3.3.5.1.5 h

SR 3.3.5.1.6 b.

Drywell 1,2,3 4@3 B

SR '3.3.5.1.1 s 1.88 psig l

Pressure - High SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 c.

Reactor vesset 1,2,3, 4

B SR 3.3.5.1.1 s 221.7 inches Water 4 '),5(a)

SR 3.3.5.1.2 f

Level - High, Level 8 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 d.

Condensate 1,2,3, 2

D

a 3.3.5.1.1

= 59,700 gettons Storage Tank Level-Low 4(C) 5(C)

En 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 e.

Sq pression Pool 1,2,3 2

D SR 3.3.5.1.1 5 18 ft 6 inches Water Level -mish SR 3.3.5.1.2 l

SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6 (continued) l (a)

When associated stbsystem(s) are required to be OPERABLE.

h5 tsa rewired to initiate the associated diesel generathand '6"h +-A ben HPCS is OPERABLE for conpliance with LCO 3.5.2, *ECCS -Shutdown,"

(c) storage tank while tank water levet is not within the timits of SR 3.5.2.2.

W 4

PERRY - UNIT 1 3.3-41 Amendment No. 59, %

\\

l l

PY-CEf tJRR-1999L Primary Containment and Drywell Isolation Instrumentat Attachn.. 2 Page $ of 34 3.3.6.1 1

Table 3.3.6.1 1 (page 1 of 6)

Primary Contaltvoent and Drywell Isolation Instrunientation

=

APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP, REGulRED SURVEILLANCE OJNCTION Colelfl0NS STSTEM, ACTION C.1 REQUIREMENTS ALLOWABLE -

VALUE 1.

Main Steam Line Isolation Reactor Yesset Water a.

1,2,3 2.-

D SR 3.3.6.1.1 a 14.3 inches Level-Low Low Low, Level 1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.6 b.

Main Steam Line 1

2 E

SR 3.3.6.1.1

= 795.0 psta Pressure - Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 c.

Main Steam Line SR'3.3.6.1.6 Flow-H1sh 1,2,3 2 per MSL D

SR ~5.3.6.1.1 s 191 pstd SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3 4 1.5 l

SR 3.3.6.1.6 d.

Condenser Vacuta-Low 1,2(a),

2 D

SR 3.3.6.1.1 a 7.6 inches 3(a)

SR 3.3.6.1.2 Hg vacuum SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 Main Steam Line Pipe 1,2,3 2

D SR 3.3.6.1.1 s 158.9'F e.

Tunnet Tenperature-nigh SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7 f.

Main Steam Line Turbine 1,2,3 2

0 SR 3.3.6.1.1 8 138.9'F BulIding Teaperature-Nigh SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 g.

Manual Initiatiot 1,2,3 2

G SR 3.3.6.1.5 NA Primary Contairnent and Drywell 2.

Isolation

( f u. a c.

di

[ WyttRf a.

Reactor Vessel Water 1,2,3 Level-Low Low, Level 2 H

SR 3.3.6.1.1 a 127.6 inches

{

SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 (continued)

(a)

With any turbine stop valve not closed.

R' equired to initiate the associated drywett isolation function (b)

PERRY - UNIT 1 3.3-54 knendment No. 69,

Prinary Ccntainment and Drywell Isslaticn Instrumentat.icn pyggg,gg99t mchinen 2 Page 6of 34 3.3 6.1 Table 3.3.6.1 1 (page 2 of 6)

Primary Containment and Drywelt isolation Instrumentation APPLICA8LE CONDITIONS MODES OR REQUIRED REFERENCED' OTHER CHANNELS FROM

$PECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ActIDN C.1 REQUIREMENTS VALUE 2.

Primary Containment and Drywell isolation

(

Reactor vessel Water s.

hj (c)

LeveMow Low, Level 2 L

SR 3.3.6.1.1 t 127.6 inches (continued)

ER 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR '3.3.6.1.5 b.

Drywell Pressure-High 1,2,3 4-H SR,3,3(6.1.1 5 1.88 psig l

SR 3.3.6.1.2

' SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 c.

Reactor Vessel Water 1,2,3 Level - Low Low Low, F

SR 3.3.6.1.1 1 14.3 inches Level 1 (ECCS SR 3.3.6.1.2 Divisions 1 and 2)

SR 3.3.6.1.3 j

SR 3.3.6.1.4 SR 3.3.6.1.5 (c) 3 SR 3.3.6.1.1 a 14.3 Inches h e.fca n b,A bu h]

~

SR 3,3.6.1.2 byv at[,\\ g SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 d.

Drywell Pressure - High 1,2,3 2

F SR 3.3.6.1.1 5 1.88 psig (ECCS Divisions 1 and 2)

SR 3.3.6.1.2 SR 3.3.6.1.3

(

SR 3.3.6.1.4 SR 3.3.6.1.5 e.

Reactor Vessel Water 1,2,3 4

F SR 3.3.6.1.1 t 127.6 inches Level -Low Low, Level 2 (HPCS)

SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 (c) 4 L

SR 3.3.6.1.1 2127.6 inches SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 f.

Drywell Pressure - High 1,2,3 4

F SR 3.3.6.1.1 5 1.88 psig (HPCS)

SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 p.

Containment and Drywell 1,2,1 b(

Purge Exhaust Plentsn F

SR 3.3.6.1.1 5 4.0 nR/hr Radiat ion - High SR 3.3.6.1.2 above SR 3.3.6.1.4 background SR 3.3.6.1.5 (continued)

(b)

Required to initiate the drywell isolation function.

(c)

During CORE ALTERATIONS, and operations with a potential for draining th e reactor vessel.

PERRY - UNIT 1 l

3.3-55 Amendment No. #

1 PY-CEIMRR 1999L Primary Containment and Drywell Isolation Instrumentation Anachment 2 Page 7of 34 3.3.6.1 t

Table 3.3.6.1 1 (page 3 o' 6)

Primary Contelnaent and Drywell Isolatitn Instrtamentation APPL 1 CABLE CONDITIONS l

MODES OR REQUIRED REFERENCED OTHER CNANNELS FROM SPECIFIED PER TRIP REQUIRED EURVEILLANCE ALLOWABLE FUNCTION CCWITIONS SYSTEM ACT10N'C.1 REQUIREIENTS l

VALUE 2.

Primary contalisment and Drywell Isolation

s. Containannt and (d) 2 K

SR 3.3.6.1.1 8 4.0 mR/hr above Drywatt Purge Exhaust Ptersse Radiation-High SR 3.3.6.1.2 backgrotmd (continued)

SR 3.3.6.1.4 LN' p* a'W !.! f.'.3 T

Mt 6 sagagg. pk h.

Manuel Initiation 1,2,3 G

BR 3.3 4,1.5 NA l

I (d) 2 K

SR 3.3.6.1.5 NA 3.

seactor core Isolation cooling (acIn system Isolation a.

RCIC Steam Line 1,2,3 1

F SR 3.3.6.1.1 8 298.5 inches Flow-High SR 3.3.6.1.2 water SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.',.1.5 b.

RCIC Steun Line ? low 1,2,3 1

F SR 3.3.6.1.2 k 3 seconds and I

flee Detoy SR 3.3.6.1.4 s 13 seconds SR '3.3.6.1.5 c.

RCIC Steam Stpply Line 1,2,3 1

F SR 3.3.6.1.1 k 55 psig Pressure-Low SR 3.3.6.1.2 I

SR 3.3 6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 d.

RCIC Turbine Exhaus,t 1,2,3 2

F SR 3.3.6.1.1 s 20 psig Diaphrase Pressure-Nigh SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 e.

RCIC Ecysipment Area 1,2,3 1

F SR 3.3.6.1.1 s 145.9'F Anblent 7enperature - High SR 3.3.6.1.4 SR 3.3.6.1.5 i

SR 3.3.6.1.7 f.

Main Steam Line Pipe 1,2,3 1

F SR 3.3.6.1.1 s 158.9'F Tunnet Tenperature - hlgh SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7 (continued)

(b)

Required to initiate the drywell isolation function.

(d)

, irradiated fuel assenblies in primary containment.During CORE ALTERATIONS, op vessel, and movement of 4

PERRY - UNIT 1 3.3-56 Anendment No. 59 D

~

Primary Containment and Drywell Isolation Inst PY<Emm Anachment 2 Page 8of 34 3.3.6.1 Table 3.3.6.1 1 (pege 6 of 6)

Primary Contelnment and Drywell Isolation Instrumentation APPLICABLE

{

MODES OR CONDITIONS J

REFERENCED OTHER REQUIRED FROM FUNCTION SPECIFIED CHANNELS PER REQUIRED CONDITIONS TRIP SYSTEM ACTION C.1

, REQUIEBENTS ALLOWABLE SURVEILLANCE VALUE 5.

Rint System Isolation s.

RNR Egsipment Ares Anblent 1 per, area F

SR 3.34.1.1 8 159.9'F h

Temperature-sigh SR 3.SA1.4 r

SR 3.3 A 1.5 1,O,6*'

b.

Reactor Vessel Water SR 3.34.1.7 Levet-Lew, Level 3 2

F SR 3.3 A 1.1 a 177.1 l

SR 3JA1.2 inches C4Jt.c4 S c.

O d S *te.

st 3.3A1.3 Sfc/$L/ha

$ !*3 c

k b,

,4,5 M

,5 J

SR 3.3A1.1 a 177.1 l

SR 3.34.1.2 inches SR ' 3.3A T.3' SR 3.3A1 4 c.

Reactor Vesset Steam SR 3.3 A 1'.5 Dome Pressure-High 1,2,3 2

F SR 3.3 A 1.1 5 150 psig SR 3.311.2 SR 3.611.3 SR 3.6.11.4 1

d.

DrywetL Pressure-High 1,2,3 2

SR 3.6.1.1'.5 i

7 SR '3.3 A 1.1 s 1.88 psig SR 3.3A1.2 SR 3.3 A T.3 SR 3.3A1.4 l

e.

Manual initiation SR 3.3A1.5 I

1,2,3 2

C SR 3.3 A 1.5 NA (e) With reactor vesset steam dome pressure less than the RNR cut i

=

n permissive pressure.

(f)

Dnty one trip system required in MODES 4 and 5 with RHR Shutd With reactor steam cleme pressure greater than or equat to the RHRown Cooling (g) cut in permlssive tusd

)

? ""'

PERRY - UNIT 1 3.3-59 AmendmentNo.69M

i l

PY-CEl/NRR-1999L Recirculation Loops Operating Page 9 of 34 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 M itherd Two recirculation loops'shall be in operation with:

v 1.

Matched flows; and 2.

Total core flow and THERMAL POWER within limits.

gg b.

One recirculation loop shall be in operation with:

1.

Thermal power f 2500 MWt; j

2.

Total core flow and THERMAL POWER within limits; 3.

Required limits modified for single recirculation loop operation as specified in the COLR; and 4.

LCO 3.3.1.1, " Reactor Protection System (RPS)

Monitors Flow Biased Simulated Thermal

-Allowable Value of Table 3.3.1.1-1 reset for single loop operation.

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

)

Required limit and setpoint modifications for single 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after transition from two recirculation operation to single recir

...........................culation loop operation.

APPLICABILITY:

MODES 1 and 2.

_ ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Recirculation loop jet A.1 pump flow mismatch not Shut down one of the 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months within lir.iits.

recirculation loops.

(continued)

PERRY - UNIT 1.

3.4-1 AmendmentNo.((

1

{j[g*L RHR Shutdown Cooling System-Cold Sh m

Page 10of 34 i

I ACTIONS (continued)

CONDITION 1

REQUIRED ACTION COMPLETION TIME

~

%\\bt $ n B.

No RHR shutdown B.1 Verify reactor L 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from cooling subsystem in operation.

coolant q: !:r q[-

discovery of no i

by an alternate reactor coolant method.

\\

O circulation 1

l No recirculation pump M

l in operation.

Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months MQ

,thereafter B.2 Monitor reactor Once per hour I

coolant temperature and pressure.

l i

SURVEILLANCE REQUIREMENTS i

i SURVEILLANCE l

FREQUENCY SR 3.4.10.1 l

Verify one RHR shutdown cooling subsystem 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or recirculation pump is operating.

1

?

l l

i

't l

PERRY - UNIT 1 3.4-25 AmendmentNo.//

PY CE!/NRR-1999L RCS P/T Liaits Anahmem 2 Page 11 of 34 3.4.11 SURVEILLANCE REQUIREMENTS (continued)

. SURVEILLANCE FREQUENCY SR 3.4.11.8 4

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

Only required to be met in single loop operation during increases in THERMAL POWER or recirculation loop flow with the operating recirculation loop jet pump flow s 50% of rated core floy or THERMAL POWER s 30% of RTP, and with reactor vessel steam dome pressure it 25 psig.

Verify the difference between the bottom.,. (Once within head coolant temperature and the RPV

15 minutes temperature is s 100*F.

prior to an

{c.%\\

L increase in THERMAL POWER or an increase in loop flow SR 3.4.11.9

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

Only required to be met in single loop operation during increases in THERMAL POWER or recirculation loop flow with the operating recirculation loop jet pump flow s 50% of rated core flow, or THERMAL POWER s 30% of RTP, and the idle recirculation loop not isolated from the RPV.

Once within 15 minutes prior to an Verify the difference between the reactor increase in coolant temperature in the recirculation THERMAL POWER loop not in operation and the RPV coolant or an increase temperature is s 50*F.

in loop flow SR 3.4.11.10 The reactor vessel material surveillance In accordance specimens shall be removed and examined to with the determine changes in reactor pressure schedule vessel material properties.

required by 10 CFR 50, Appendix H PERRY - UNIT 1 3.4-30 AmendmentNo.I[

PY-CEl/NRR 19991.

Primary Containment Air Locks Page 12of 34 3.6.1.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.

(continued)-

C.3 Restore air lock to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

D.

Required Action and 0.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A.

M B. or C not met in MODE 1. 2. or 3.

D.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E.

Required Action and E.1 Suspend movement of Immediately associated Com letion Time of Condit on A.

recently irradiated fuel assemblies in B. or C not met during the primary movement of recently containment.

1 irradiated fuel i

assemblies in the M

primary containment.

or OPDRVs.

E.2 Initiate action to Immediately suspend OPDRVs.

l l

l 4

PERRY - UNIT 1 l

3.6-6 Amendment No. g, (6

i l

(

PYCEUNRR l%9L PCIVs Amhment 2 3.6.1.3 Pqe 13cf 34 I

t SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY l

SR 3.6.1.3.11

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

Only required to be met in MODES 1, 2,

)

and 3.

l Verify combined leakage rate of I gpm

NOTE-----

times the total number of PCIVs through SR 3.0.2 is not hydrostatically tested lines that applicable penetrate the primary containment is not l

exceeded when these isolation valves are tested at 2 1.1 P.

In accordance with 10 CFR 50, Appendix J, as modified by approved exemptions I

l 1

SR 3.6.1.3.12

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

Only required to be met in MODES 1, l

2, and 3.

= =

==-===--_

1 i

Verify each outboard 42 inch primary 18 months l

containment purge valve is blocked to restrict the valve from opening > 50*.

l SR 3.6.1.3.13

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

l Not required to be met when the Backup Hydrogen Purge System isolation valves i

are open for pressure control, ALARA or air quality considerations for personnel entry, or Surveillances or special testing of the Backup Hydrogen Purge System that require the valves to be open.

l Verity each 2 inch Backup Hydrogen Purge 31 days System isolation valve is closed.

4

[

PERRY - UNIT 1 3.6-19 Amendment No. 43, l

l

PY CEl/NRR 1999L Containm2nt Humidity Control

^= tant 2 -

P4114 of 34 3.6.1.12 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Required Action and 8.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A E

not met or in MODE 1, 2, or 3.

B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

\\

C.

Required Action and C.1 Suspend movement of Immediately associated Completion irradiated fuel Time of Condition A assemblies in the not met during primary containment.

movement of irradiated fuel assemblies in the M

primary containment, CORE ALTERATIONS, @

C.2 Suspend CORE Immediately l

OPDRVs.

f ALTERATIONS.

C.3 Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENT SURVEILLANCE FREQUENCY SR 3.6.1.12.1 Verify containment average temperature-24 hours to-relative humidity to be within limits.

( _-

PERRY - UNIT 1 3.6-35 Amendment No. Ml,

["1*$"

Primary Containment and Drywell Hydrogen Igniters Page 15cf 34 3.6.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

\\

SR 3.6.3.2.4 Verify each required igniter in 18 months

@ccessible areas develops a surface temperature of 21700*F.

[

i t

f i

i i

1 1

l

~

PERRY - UNIT 1 3.6-48 Amendment No. AS,

Co Mw 3

@cogMA-

@h o

PY-CE!/NRR 1999L bCIYS g,

h

( Ahb dt.

n 3.6.4.2 Page 16of34 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and D.I Suspend movement of Immediately associated Completion irradiated fuel Time of Condition A assemblies in the or B not met during primary containment, movement of irradiated fuel assemblies in the ANQ i

primary containment, during CORE 0.2 Suspend CORE Immediately ALTERATIONS, or during ALTERATIONS.

~

OPDRVs.

AN.Q D.3 Initiate action to Immediately.

suspend OPDRVs.

l SURVEILLANCE REQUIREMENTS l

SURVEILLANCE FREQUENCY SR 3.6.4.2.1

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

I.

Valves and blind flanges in high radiation areas may be verified by i

use of administrative means.

2.

Not required to be met for SCIVs that are open under administrative controls.

Verify each secondary containment 31 days isolation manual valve and blind flange that is required to be closed during accident condition's is closed.

l l

4 1

V j

PERRY - UNIT I 3.6-55 Amendment No. 69 l

p g,

Drywell Isolation Valves Amdmwnt 2 3*6*b'3 Pqe 17of 34 3.6 CONTAINMENT SYSTEMS 3.6.5.3 Drywell Isolation Valves i

LCO 3.6.5.3 Each drywell isolation valve, except for Drywell Vacuum Relief System valves, shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS

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

1.

Penetration flow paths, except for the 24 inch and 36 inch purge supp and exhaust valve penetration flow path, may be unisolated intermit under administrative controls.

2.

Separate Condition entry is allowed for each penetration flow path 3.

Enter applicable Conditions and Required Actions for systems made inoperable by drywell isolation valves.

4.

Enter applicable Conditions and Required Actions of LC0 3.6.5.1, "Drywell," when drywell isolation valve leakage results in. exceeding overall drywell by

_______________________ pass leakage rate acceptance criteria.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more A.1 penetration flow paths' Isolate the affected 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months with one drywell penetration flow path b use of at least isolation val.ve' inoperable, g\\,pJ one (E5!gFand de-activated automatic

(

valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

MQ (continued)

Q, PERRY - UNIT 1 3.6-65 Amendment No. 6&'

PY-CEUNRR 1999L DryWe11 Is01ation Valyes o

^*hment 2 3.6.5.3 Page la of 34 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.2

NOTE---------

Isolation devices in high radiation areas may be verified by use of administrative means.

Verify the affected Prior to penetration flow path entering MODE 2 is isolated.

or 3 from MODE 4, if not performed within the previous 92 days B.

One or more B.1 Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months penetration flow paths penetration flow path with two drywel'1 by use of at least isolation valves one closed and de-inoperable.

activated automatic hgadmawd valve? blind f1ange, m

i or GTE5affvalve with g(>

flow through the

3*Ettsecured.

C.

Required Action and C.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met.

AND C.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months PERRY - UNIT 1 3.6-66 Amendment No. Ji@3

4-1 1

o PY-CEI/NRR 1999L 1

Drywell Isolation Valves Pqe 19of 34 i

3.6.5.3 i

SURVEILLANCE REQUIREMENTS' SURVEILLANCE FREQUENCY i

i SR 3.6.5.3.1 Verify each 24 inch and 36 inch drywell 31 days.

purge supply and exhaust isolation valve 2

is sealed closed.

t-f-~-trequire---------------NOTE------------------h SR 3.6.5.3.2 j

No o be met w the Backup 134bMd*

Hydroge urge System ' solation valves 1

are n for press e control, ALA or gif quality co derations for rsonnel i

entry, or S veillances or cial i

i testing the Backup H ogen Purge Syst that require valves to be

'i j

o i

(((lI(??f" Verify h 2 inch Ba p Hydrog urge System isolation val e is clos

,)

g h

i i

SR 3.6.5.3.3

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

i 1.

Valves and blind flanges in high radiation areas may be verified by

)

use of administrative means.

2.

Not required to be met for drywell isolation valves that are open under administrative controls.

Verify each drywell isolation manual Prior to valve and blind flange that is required entering MODE 2 to be closed during accident conditions or 3 from is closed.

MODE 4, if not performed in the previous 92 days (continued)

PERRY - UNIT 1 3.6-67 Amendment No. 19f

PY CEl/NRR-1999L

~

Pqe 20of 34 3.7 PLANT SYSTEM 3.7.3 Control Room Emergency Recirculation. (CRER) System LCO 3.7.3 Two CRER subsystems shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the primary containment or fuel handling building, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One CRER subeystem A.1 Restore CRER 7 days inoperable.

subsystem to OPERABLE

status, B.

quired Action and

.B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

sociated Completion ime of Condition A ANQ I

not met in MODE 1, 2, or 3.

B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued) i q,4

/

PERRY - UNIT 1 3.7-4 Amendment No. If

PY{IUNRR-lM9L Control Room HVAC System 3.7.4 Pqe 2l of 34 3.7 PLANT SYSTEMS 3.7.4 Control Room Heating. Ventilating, and Air Conditioning (HVAC) System i

LCO 3.7.4 Two control room HVAC subsystems shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the primary containment or fuel handling building, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs).

ACTIONS CONDITION i

REQUIRED ACTION COMPLETION TIME i

A.

One control room HVAC A.1 Restore control room 30 days subsystem inoperable.

HVAC subsystem to OPERABLE status.

B.

Two control room HVAC B.1 Verify control room Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months l

subsystems inoperable.

air temperature is 5 90*F.~

AND l

B.2 Restore one control 7 days room HVAC subsystem to OPERABLE status.

C.

quired Ac.+. ion and C.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sociated Completion me of Condition A or 6NQ l

B not met in MODE 1, 2, or 3.

J4 2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months i

l (continued) l l

i PERRY - UNIT 1 3.7-8 Amendment No. gf,f

{,1 ] N Fuel Handling Building Ventilation Exhaust System Page 22 of 34 3.7.9

_ SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Operate each FHB ventilation exhaust subsystem for 210 continuous hours with 31 days heaters operating.

i SR 3.7.9.2 Perform FHB ventilation exhaust filter testing in accordance w In accordance Filter Testing { ogram (ith the Ventilation VFTP).

with the VFTP I

L SR 3.7.9.3 Perform a system functional test.

~

18 months 1

SR 3.7.9.4 Perform a CHANNEL FUNCTIONAL TEST of the FHB ventilation exhaust radiation monitor 92 days (noble gas) i 1

i l

i I

PERRY - UNIT 1 3.7-18 Amendment No.)WV

=-

c PY-CEI/NRR 1999L Amstmwnt 2 AC Sources-Operating Pqe 23 of 34 1

3"8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.3

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

1.

DG loadings may include gradual loading as recommended by the manufacturer.

2.

Momentary transients outside the load range do not invalidate this test.

3.

This Surveillance shall be conducted on only one DG at a time.

4.

This SR shall be preceded by, and immediately follow, without shutdown, a successful performance of SR 3.8.1.2 or SR 3.8.1.7.

Verify each DG operates for 2 60 minutes at a load 2 5600 kW and g 7000 kW for As specified in Division 1 and 2 DGs, and Table 3.8.1-1 2 2600 kW for Division 3 DG.

SR 3.8.1.4 Verify each day tank contains 2 al of 31 days fuel oil for Divisions 1 and 2 and i

[

~

2({jjfgal for Division 3.

b

[

SR 3.8.1.5 Check for and remove accumulated water from each day tank.

31 days SR 3.8.1.6 Verify the fuel oil transfer system operates to automatically transfer fuel oil 31 days from the storage tank to the day tank.

(continued)

PERRY - UNIT 1 3.8-6 Amendment No. /d[

PY<EUNRR-1999L AC Sources-Operating e

Page 24 or 34 3.8.1 i

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.9

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

1.

This Surveillance shall not be performed in MODE 1 or 2.

However, credit may be taken for unplanned

{

events that satisfy this SR.

2.

If performed with DG synchronized with offsite power, it shall be performed at a power factor s 0.9.

' ao ck),

\\

or 3

Verify each DG rejects greater than or 18 months equal to its associated single largest

{

t i

post-accident load.

Following load rejertion, engine speed is maintained less than nominal plus 75% of the difference between nominal speed and the overspeed trip setpoint, or 15% above nominal, whichever is less.

SR 3.8.1.10

NOTE [------------------

l d.

Mom tary tran nts out the 1 d

I d powe ctor r o no inval ate this t st.

1 is Surveillance shall not be performed in MODE 1 or 2.

However,g 4-credit may be taken for unplanned events' that satisfy this SR.

Verify each DG operating at a power factor 18 months s 0.9 does not trip and voltage is maintained s 4784 V for Division 1 and 2 DGs and s 5000 V for Division 3 DG during and following a load rejection of a load 2 5600 kW for Division 1 and 2 DGs and 2 2600 kW for Division 3 DG.

(continued)

PERRY - UNIT 1 3.8-8 Amendment No. $

i o

PYGI/NRR 1999L Attachtnent 2 AC Sources-Operating Page 25of 34 3,8,1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.12

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

1.

All DG starts may be preceded by an engine prelube period.

2.

This Surveillance shall not be L

performed in MODE 1 or 2.

However, i

credit may be taken for unplanned events that satisfy this SR.

Verify on an actual or simulated Emergency 18 months Core Cooling System (ECCS) initiation signal each DG auto-starts from standby condition and:

In s 10 seconds for Divisions 1 and 2, a.

and s 13 seconds for Division 3 after auto-start and during tests, achieves voltage 2 3900 V and s 4400 V; b.

In s 10 seconds for Divisions 1 and 2, and s 13 seconds for Division 3 after auto-start and during tests, achiev frequency.2 58.8 Hz and s 61.2 Hz; a l

Operates for 2 5 minutes [e c.

g SR 3.8.1.13

.------------NOTE--------------------

This Surveillance shall not be performed in MODE,1, 2, or 3.

However, credit may be taken'for unplanned events that satisfy this SR.

Verify each DG's automatic trips are bypassed on an actual"or simulated ECCS 18 months initiation signal except:

a.

Engine overspeed; and b.

Generator differential current.

Q.,

(continued)

PERRY - UNIT 1 3.8-10 AmendmentNo.//

i PY<E!/NRR 1999L Amdmwnt 2 Pqe 26 of 34 AC Sources--Operating 3,8,1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.20

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

All DG starts may be preceded by an engine prelube period.

Verify, when started simultaneously from standby condition, the Division 1 and 2 DGs 10 years achieve GPVoltage 2 3900 V and s 4400 V and frequency 2 58.8 Hz and 5 61.2 Hz in s 10 I

seconds, and the Division 3 DG achieves a

-frequency 2 58.8 Hz in s 10 seconds, and a

}

voltage 2 3900 V and s 4400 V and frequency 2 58.8 Hz and 5 61.2 Hz in s 13 seconds.

1

,( _. -

PERRY - UNIT 1 3.8-15 Amendment No./s/

PYCEUNRR-IPML DC Sources--Shutdown Anwhment 2 3.8.5 Pqe 27 of 34 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources--Shutdown (bhe following DC electrical power subsystems shall be LC0 3.8.5 OPERABLE:

l One Class IE DC electrical power subsystem capable of a.

supplying one division of the Division 1 or 2 onsite Class IE electrical power distribution subsystem (s) required by LC0 3.8.8, " Distribution Systems -

Shutdown";

b.

One Class IE battery or battery charger, other than the DC electrical power subsystem in LCO 3.8.5.a, capable of supplying the remaining Division 1 or Division 2 onsite Class IE DC electrical power distribution subsystem when required by LCO 3.8.8; and The Division 3 DC electrical power subsystem capable of

($2uec supplying the Division 3 onsite Class IE DC electrical distribution subsyste p hen the Division 3 onsite Class IE DC electrica Fpower distribution subsystem is I

required by LC0 3.8.8.

APPLICABILITY:

MODES 4 and 5, During movement of irradiated fuel assemblies in the primary containment or fuel handling building.

...r l

A s PERRY - UNIT 1 3.8-28 Amendment No. A f

py.camR 1999L DC Sources-Shutdown 3.8.5 Page 28 of 34 t

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.2.2 Suspend movement of Immediately irradiated fuel assemblies in the primary containment and fuel handling building.

AND

\\

A.2.2 Suspend movement of Immediately.

irradiated fuel

(

assemblies in the I

primary containment and fuel handling building.

QD A.2.3 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel.

AND A.2.4 Initiate action to Immediately restore required DC electrical power subsystems to OPERABLE status.

l PERRY - UNIT 1 3.8-30 AmendmentNo.46

PY CEI/NRR 1999L Design Features Page 29of 34 4.O 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

k 'c# s 0.95 if fully flooded with unborated water, which a.

in ludes an allowance for uncertainties as described in Section 9.1.2 of the USAR; b.

A nominal fuel assembly center to center storage spacing of 7 inches within rows and 12 inches between rows in the storage racks in the upper containment pool; and c.

A nominal fuel assembl center to center storage spacing 4

of 6.625 incheii,' -,u.iiy a neutron poison material between storage spaces, in the high density storage racks in the j

fuel handling building.

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

k 'c# 5 0.95 if fully flooded with unborated water, which a.

in ludes an allowance for uncertainties as described in Section'9.1.1 of the USAR;j l

b.

A nominal 7 inch-center to center distance between fuel assemblies placed in storage racks.

4.3.2 Drainaae The sp.egt fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 594 ft 6 inches.

4.3.3 Capacity 4.3.3.1 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 4020 fuel assemblies.

4.3.3.2 No more than 190 fuel assemblies may be stored in the upper containment pool.

4 -

PERRY - UNIT 1 4.0-2 Amendment No. 8

+

py.cewgg.i999t Atrachment 2 Responsibility Page 30of 34 5.1 A

5.0 V

ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 and shall delegate in writing the succession to thi responsibility during his absence.

The plant manager, or his designeo, shall ap systems or equipment that affect hi(clear safety, and all

{

administrative procedures.

5.1.2 The shift supervisor (SS) shall be responsible for the control room command function.

During an9 absence of the SS from the i

with an active Senior Reactor Operatorcontrol room while designated to assume the control room co(mman)d function. license SRO any absence of the SS from the control room while the unit is in During MODE 4 or 5, an individual with an active SRO license or Reacto Operator license shall be designated to assume the control room command function.

D 4

PERRY - UNIT 1 5.0-1 Amendment No. A9

PY CEUNRR-1999L Attachtnent2 Unit Staff Qualifications Page 31of 34 53 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.)-1971 for comparable positions as modified i;y Specification 5.c.2.f, except for the radiatiori protection minager, who shall meet or exceed the qualifications of Hegulatory GulTE 1.8, September 197h7 Operators and Senior Reactor Operators,/he licensed Reactor

. ele shalpiscmers, exceea V 1

tr.;

= ; r ga,i.c;u;m. J 1..

20ppi; men ml requir=cr,ts specified4r, Sectione A.nd C J "r.cle:ure 1 of the $8erch 28,1

~

NRC lettei iv oli l iceriasca.f Camg % u d k a p d * <~k e

16 ( FA. s r.

PERRY - UNIT 1 5.0-4 Amendment No. fff

l PY CEl/NRR-1999L Programs and Manuals Auckunt 2 Pqe 32 cf 34 5.5 l

j 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Procram (contir.ued) f.

Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to red 0ce releases of radioactivity when the projected doses in a 1

period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I; 1

l g.

Limitations on the dose rate resulting from radioactive I

material released in gaseous effluents to areas beyond the site boundary as follows:

1.

for noble gases: s 500 mrem /yr to the total body and s 3000 mrem /yr to the skin, and 2.

for iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives

> 8 days: s 1500 mrem /yr to any organ; h.

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from the unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; j

i.

'.imitatio.ns on the annual and quarterly doses to a member of che public from iodine-131, iodine-133, tritium, and all radionuclidas in particulate form with half lives > 8 days

)

in gaseous effluents released from the unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and j.

Limitations on the annual dose or dose commitment to any member of the public due to releases of rcdioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

5.5.5 Component Cyclic or Transient limit This program provides controls to track the USAR, Section 3.9.1.1, cyclic and transient occurrences to ensure that reactor vessel is maintained within the design limits.

l (continued)

' (..

PERRY - UNIT I 5.0-9 Amendment No.,69"

PY CEIMRR 1999L Anachment 2 Progrcms and Manuals Page 33of 34 5.5

(

5.5 Programs and Manuals i

5.5.7 Ventilation Filter Testino Proaram (VFTP) i (continued) d.

Demonstrate for each of the ESF systems that the pressure s.

drop across the combined.HEPA filters and the charcoal adsorbers is less than the value specified below when tested a

in accordance with Regulatory Guide 1.52, Revision 2, and i

ANSI N510-1980 at the systen. flowrate specified below 10%:

i ESF Ventilation System Delta P _ Flowrate a)

Control Room Emergency Recirculation 4.9" H 0 30,000

~

he{m b)

Fuel Handling Building 4.9" H 0 15,000 c) Annulus Exhaust Gas Treatment 6.0" H 0 2,0004 Demonstrate that the heaters for each of the ESF systems e.

dissipate the value specified below i 10% when corrected to nomine.1 input voltage when tested in accordance with ANSI i

s N510-1980:

I, ESF Ventilation System Wattaae 1

4 a)

Control Room Emergency Recirculation b)

Fuel Handling Building 100 kW c) Annulus Exhaust Gas Treatment.

50 kW 20 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTPi test frequencies.

5.5.8 Exolosive Gas and Storace Tank Radioactivity Monitorina Procram This program provides controls for potentially explosive gas mixtures contained in the main condenser offga liquid 'to' rage tanks.

s

(

The program shall include:

The limits for concentrations of hydrogen in the main a.

condenser offgas treatr:ent system and a surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria (i.e., whether or not the s explosion); ys' tem is designed to withstand a hydrogen and (continued) m PERRY - UNIT 1 5.0-12 Amendment No. Af

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

PY-CEl.HRR.1999L tiigh Radiation Area Page 34 of 34 5.7 5.7 High Radiation Area 5.7.2 (continued)

Individuals qualified in radiation protection (e.g., hr.alth physics technicians) or personne. procedures l continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates s 3000 mrem /hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

5.7.3 In addition to the requirements of Specification 5.7.1, for individual high radiation areas accessible to personnel with radiation levels such that a major portion of the body could receive in I hour a dose 21000 mrem that are located within large areas such as reactor containment, where no enclosure exists for N4. u 7 purposes of locking, or that W r m t F# continuously guarded, and where no enclosure can be reasonably constructed around the l

individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.

5.7.4 In addition to the requirements and exemptions of Specifications that a major portion of the body could receive in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months a

> 3000 mrem, entry shall require an approved RWP which will specify dose rate levels in the immediate work area and the maximum allowable

.ay time for individuals in that area.

In lieu of the stay time specification of the RWP, continuous surveillance, direct or remote such as use of closed circuit TV cameras, may be made by personn,el qualified in radiation protection procedures to provide positive exposure control over activities within the areas.

PERRY - UNIT 1 5.0-20 Amendmenttio.//f

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

PY-CEl/NRR.1999L Page I of 2 l

l SIGNIFICANT HAZARDS CONSIDERATION I

The standards used to arrive at a determination that a request for amendment involves no significant hazards consideration are included in the Commission's regulations, I

10CFR50.92. This regulation states that a proposed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a i

significant reduction in a margin of safety.

j The proposed changes have been reviewed with respect to these three factors and it has been determined that the proposed changes do not involve a significant hazard because:

1. The proposed changes do not involve a significant increase in the probability or l

consequences of an accident previously evaluated.

l l

Eight of the proposed changes are administrative in nature and either correct errors or incorporate into the improved Technical Specifications a change which was approved l

by the NRC under Amendment 70 for the current Technical Specifications. Changing l

the classification of the Backup Hydrogen Purge System isolation valves from i

drywell isolation valves to primary containment isolation valves results in the same actions being taken in the event one of these valves is declared inoperable. However, i

the Completion Times are more restrictive for inoperable primary containment isolation valves than for inoperable drywell isolation valves. The proposed changes to the diesel generator fuel oil day tank minimum volumes provide more stringent i

requirements for operation of the facility to increase the reliability of the diesel generator fuel oil transfer pump operation. The more stringent requirements continue to ensure that the safety analysis and licensing basis are maintained. The proposed change to Specification 5.7.3 clarifies continuously guarding a high radiation area is an option, not a requirement. The proposed changes have been reviewed and determined to have no affect on accident conditions or assumptions.

Based on the above, the proposed changes do not significantly increase the probability or consequences of any accident previously evaluated.

1 l

1 I

PY CEI/NRR 1999L Page 2 or 2

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

As stated above eight of the proposed changes are administrative in nature and do not increase the possibility of any new or different kind of accident. Changing the classification of the Backup Hydrogen Purge System isolation valves from drywell isolation valves to primary containment isolation valves results in the same actions being taken in the event one of these valves is declared inoperable. However, the Completion Times are more restrictive for inoperable primary containment isolation valves than for inoperable drywell isolation valves. The proposed changes to the diesel generator fuel oil day tank minimum volumes do not involve installation of new or different equipment nor do they change the methods governing normal plant operations. These changes are also consistent with assumptions made in the safety analysis and licensing basis. Clarifying the controls ofhigh radiation areas will not impact existing or introduce any new accident precursors. The proposed changes do not cceate the possibility of a new or different kind of accident since they do not affect the reacter coolaet pressure boundary or reactivity controls. Consequently, no new failure mode: are introduced as a result of the proposed changes.

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

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

l The margin of safety is unchanged because the proposed administrative changes do not affect any design basis or accident assumptions. Changing the classification of the Backup Hydrogen Purge System isolation valves from drywell isolation valves to primary containment isolation valves results in the same actions being taken in the event one of these valves is declared inoperable. However, the Completion Times are more restrictive for inoperable primary containr.1ent isolation valves than for inoperable drywell isolation valves. The imposition of more restrictive requirements for the diesel generator fuel oil day tank minimum volumes results from the i

implementation of the Bases for the Technical Specification Surveillance Requirement. Clarifying the controls of high radiation areas is consister t with ALARA practices.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

{,...

~__.

l'Y-CEUNRR.1999L TABLE OF CONTENTS

$llI57/

4' 1.0 USE AND APPLICATION 1.1 Definitions 1.0-1 1.2 Logical Connectors.....

1.0-8 1.3 Completion Times.....

1.0-11 1.4 Frequency 1.0-24 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.0-1 2.2 SL Violations 2.0-1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs....................

B 2.0-1 B 2.1.2 Reactor Coolant System (RCS) Pressure SL........

B 2.0-6 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY..

3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.0-4 8 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY....

B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY B 3.0-10 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) 3.1-1 3.1.2 Reactivity Anomalies..

3.1-5 3.1.3 Control Rod OPERABILITY 3.1-7 3.1.4 Control Rod Scram Times 3.1-12 3.1.5 Control Rod Scram Accumulators.............

3.1-15 3.1.6 Control Rod Pattern 3.1-18 3.1.7 Standby Liquid Control (SLC) System 3.1-20 3.1.8 Scram Dischar ge Volume (SDV) Vent and Drain Valves...

3.1-24 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)

B 3.1-1 B 3.1.2 Reactivity Anomalies...

B 3.1-8 B 3.1.3 Control Rod OPERABILITY B 3.1-13 8 3.1.4 Control' Rod Scram Times B 3.1-22 8 3.1.5 Control Rod Scram Accumulators.

B 3.1-28 B 3.1.6 Control Rod Pattern B 3.1.

B 3.1.7 Standby Liquid Control (SLC) System B 3.1-38 B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves...

B 3.1-44 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) 3.2-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) 3.2-2 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) 3.2-3 8 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

B 3.2-1 B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

B 3.2-6 B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR).

B 3.2-10 (continued)

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Page got 7 TABLE OF CONTENTS 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation 3.3-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation.......

3.3-10 3.3.2.1 Control Rod Block Instrumentation 3.3-15 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation...

3.3-20 3.3.3.2 Remote Shutdown System.................

3.3-24 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation 3.3-26 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation.....

3.3-29 3.3.5.1 Emergency Core Cooling System (ECCS)

Instrumentation 3.3-32 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation 3.3-44 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation 3.3-48 3.3.6.2 Residual Heat Removal (RHR) Containment Spray System Instrumentation....

3.3-60 3.3.6.3 Suppression Pool Makeup (SPMU) System Instrumentation 3.3-64 3.3.6.4 Relief and Low-Low Set (LLS) Instrumentation.

3.3-68 3.3.7.1 Control Room Emergency Recirculation (CRER) System Instrumentation 3.3-70 3.3.8.1 Loss of Power (LOP) Instrumentation 3.3-74 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring 3.3-77 8 3.3 INSTRUMENTATION B 3.3.1.1 Reactor Protection System (RPS) Instrumentation B 3.3-1 B 3.3.1.2 Source Range Monitor (SRM) Instrumentation.

B 3.3-33 B 3.3.2.1 Control Rod Block Instrumentation B 3.3-42 B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation.

B 3.3-51 B 3.3.3.2 Remote Shutdown System.

B 3.3-63 8 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation B 3.3-68 B 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation B 3.3-79 B 3.3.5.1 Emergency Core Cooling System (ECCS)

Instrumentation B 3.3-88 B 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation B 3.3-124 3 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation B 3.3-136 B 3.3.6.2 Residual Heat Removal (RHR) Containment Spray System Instrumentation...

B 3.3-173 B 3.3.6.3 Suppression Pool Makeup (SPMU) System l

Instrumentation B 3.3-184 (continued) l PERRY UNIT 1 ii Revision No. O

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TABLE OF CONTENTS Page 3 or 7 B 3.3 INSTRUMENTATION (continued)

B 3.3.6.4 Relief and Low-Low Set (LLS) Instrumentation......

B 3.3-195 B 3.3.7.1 Control Room Emergency Recirculation (CRER) System Instrumentation B 3.3-201 B 3.3.8.1 Loss of Power (LOP) Instrumentation B 3.3-211 B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring..

B 3.3-218 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating 3.4-1 3.4.2 Flow Control Valves (FCVs).

3.4-6 3.4.3 Jet Pum]s 3.4-8 3.4.4 Safety /lelief Valves (S/RVs).

3.4-10 3.4.5 RCS Operational LEAKAGE.

3.4-12 3.4.6 RCS Pressure Isolation Valve (PIV) Leakage....

3.4-14 3.4.7 RCS Leakage Detection Instrumentation 3.4-16 3.4.8 RCS Specific Activity 3.4-19 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown.............

3.4-21 3.4.10 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown 3.4-24 3.4.11 RCS Pressure and Temperature (P/T) Limits 3.4-26 3.4.12 Reactor Steam Dome Pressure 3.4-32 l

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 Recirculation Loops Operating B 3.4-1 8 3.4.2 Flow Control Valves (FCVs).

B 3.4-9 B 3.4.3 Jet Pumps B 3.4-13 8 3.4.4 Safety / Relief Valves (S/RVs)

B 3.4-17 B 3.4.5 RCS Operational LEAKAGE B 3.4-22 B 3.4.6 RCS Pressure Isolation Valve (PIV) Leakage...

B 3.4-27 B 3.4.7 RCS Leakage Detection Instrumentation B 3.4-32 B 3.4.8 RCS Specific Activity B 3.4-39 8 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown.....

B 3.4-43 B 3.4.10 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown B 3.4-48 B 3.4.11 RCS Pressure and Temperature (P/T) Limits B 3.4-53 B 3.4.12 Reactor Stean. Dome Pressure B 3.4-63 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - 0)erating 3.5-1 3.5.2 ECCS -S lutdown 3.5-6 3.5.3 RCIC System 3.5-10 (continued)

PERRY - UNIT 1 iii Revision No. O

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Page 4 or 7 TABLE OF CONTENTS B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS -0)erating B 3.5-1 B 3.5.2 ECCS-Slutdown..

B 3.5-15 B 3.5.3 RCIC System B 3.5-20 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary Containment--Operating 3.6-1 3.6.1.2 Primary Containment Air Locks 3.6-3 3.6.1.3 Primary Containment Isolation Valves (PCIVs)....

3.6-9 3.6.1.4 Primary Containment Pressure...

3.6-20 3.6.1.5 Primary Containment Air Tem 3.6-21 Low-Low Set (LLS) Valves. perature 3.6.1.6 3.6-22 i

3.6.1.7 Residual Heat Removal (RHR) Containment Spray System..................

3.6-24 3.6.1.8 Feedwater Leakage Control System (FWLCS).

3.6-26 3.6.1.9 Main Steam Isolation Valve (MSIV) Leakage Control System (LCS) 3.6-27 3.6.1.10 Primary Containment-Shutdown.......

3.6-29 3.6.1.11 Containment Vacuum Breakers 3.6-31 3.6.1.12 Containment Humidity Control 3.6-34 3.6.2.1 Suppression Pool Average Temperature 3.6-36 3.6.2.2 Suppression Pool Water Level...

3.6-39 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling 3.6-40 3.6.2.4 Suppression Pool Makeup (SPMU) System 3.6-42 3.6.3.1 Primary Containment Hydrogen Recombiners........

3.6-44 3.6.3.2 Primary Containment and Drywell Hydrogen Igniters...

3.6-46 3.6.3.3 Combustible Gas Mixing System 3.6-49 3.6.4.1 Secondary Containment 3.6-51 3.6.4.2 Secondary Containment Isolation Valves (SCIVs).

3.6-53 3.6.4.3 Annulus Exhaust Gas Treatment (AEGT) System 3.6-56 3.6.5.1 Drywell 3.6-59 3.6.5.2 Drywell Air Lock....

3.6-61 3.6.5.3 Drywell Isolation Valves.

3.6-65 3.6.5.4 Drywell Pressure 3.6-69 3.6.5.5 Drywell Air Temperature 3.6-70 3.6.5.6 Drywell Vacuum Relief System 3.6-71 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment-0perating B 3.6-1 B 3.6.1.2 Primary Containment Air Locks B 3.6-5 B 3.6.1.3 Primary Containment Isolation Valves (PCIVs)

B 3.6-15 B 3.6.1.4 Primary Containment Pressure B 3.6-31 B 3.6.1.5 Primary Containment Air Temperature B 3.6-34 8 3.6.1.6 Low-Low Set (LLS) Valves B 3.6-37 (continued)

PERRY - UNIT 1 iv Revision No. O

UN" TABLE OF CONTENTS Page 5 of 7 B 3.6 CONTAINMENT SYSTEMS (continued)

B 3.6.1.7 Residual Heat Removal (RHR) Containment Spray System....................

B 3.6-41

)

B 3.6.1.8 Feedwater Leakage Control System (FWLCS)........

B 3.6-46 i

B 3.6.1.9 Main Steam Isola +. ion Valve (MSIV) Leakage Control System (LCS)........

B 3.6-49 i

B 3.6.1.10 Primary Containrr.ent-Shutdown.........

B 3.6-53 B 3.6.1.11 Containment Vacuum Breakers B 3.6-57 B 3.6.1.12 -Containment Humidity Control....

B 3.6-63 B 3.6.2.1 Suppression Pool Average Temperature.......

B 3.6-68 B 3.6.2.2 Suppression Pool Water Level..............

B 3.6-73 B 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling B 3.6-77 B 3.6.2.4 Suppression Pool Makeup (SPMU) System B 3.6-81 i

B 3.6.3.1 Primary Containment Hydrogen Recombiners........

B 3.6-88 B 3.6.3.2 Primary Containment and Drywell Hydrogen Igniters....

B 3.6-93 8 3.6.3.3 Combustible Gas Mixing System B 3.6-99 B 3.6.4.1 Secondary Containment B 3.6-104 B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)..

B 3.6-109 B 3.6.4.3 Annulus Exhaust Gas Treatment (AEGT) System B 3.6-115 B 3.6.5.1

'Drywell B 3.6-120 B 3.6.5.2 Drywell Air Lock...

B 3.6-124 8 3.6.5.3 Drywell Isolation Valves.

B 3.6-132 B 3.6.5.4 Drywell Pressure.....

B 3.6-140 B 3.6.5.5 Drywell Air Temaerature B 3.6-143 B 3.6.5.6 Drywell Vacuum Relief System..............

B 3.6-146 3.7 PLANT SYSTEMS 3.7.1 Emergency Service Water (ESW) System-Divisions 1 and 2 3.7-1 3.7.2 Emergency Service Water (ESW) System-Division 3 3.7-3 3.7.3 Control Room Emergency Recirculation (CRER) System...

3.7-4 3.7.4 Control Room Heating. Ventilation. and Air Conditioning (HVAC) System 3.7-8 3.7.5 Main Condenser Offgas 3.7-11 3.7.6 Main Turbine Bypass System..

3.7-13 3.7.7 Fuel Pool Water Level 3.7-14 3.7.8 Fuel Handling Building....

3.7-15 3.7.9 Fuel Handling Building Ventilation Exhaust System 3.7-16 3.7.10 Emergency Closed Cooling Water (ECCW) System...

3.7-19 8 3.7 PLANT SYSTEMS B 3.7.1 Emergency Service Water (ESW) System-Divisions 1 and 2 B 3.7-1 B 3.7.2 Emergency Service Water (ESW) System-Division 3 B 3.7-7 B 3.7.3 Control Room Emergency Recirculation (CRER) System B 3.7-10 (continued)

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.o TABLE OF CONTENTS l

B 3.7 PLANT SYSTEMS (continued) j B 3.7.4 Control Room Heating. Ventilation and Air Conditioning (HVAC) System...

B 3.7-16 B 3.7.5 Main Condenser Offgas B 3.7-21 B 3.7.6 Main Turbine Bypass System...............

B 3.7-24 i

B 3.7.7 Fuel Pool Water Level B 3.7-28 I

B 3.7.8 Fuel Handling Building (FHB)..

B 3.7-31 l

B 3.7.9 Fuel Handling Building (FHB) Ventilation Exhaust System...................

B 3.7-34 8 3.7.10 Emergency Closed Cooling Water (ECCW) System......

B 3.7-40 i

l 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources-0]erating 3.8-1 3.8.2 AC Sources -Slutdown........

3.8-17 3.8.3 Diesel Fuel 011. Lube Oil, and Starting Air 3.8-21 3.8.4 DC Sources-0)erating 3.8-24 3.8.5 DC Sources -Slutdown........

3.8-28 3.8.6 Battery Cell Parameters 3.8-32 3.8.7 Distribution Systems-0)erating 3.8-36 3.8.8 Distribution Systems-Slutdown...

3.8-38 B 3.8 ELECTRICAL POWER SYSTEMS l

B 3.8.1 AC Sources-0)erating B 3.8-1 B 3.8.2 AC Sources - Slutdown..................

B 3.8-35 l

B 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8-42 i

B 3.8.4 DC Sources-0)erating B 3.8-52 8 3.8.5 DC Sources-Slutdown B 3.8-61 i

B 3.8.6 Battery Cell Parameters B 3.8-65 l

B 3.8.7 Distribution Systems-0)erating B 3.8-72 l

B 3.8.8 Distribution Systems-Slutdown.

B 3.8-82 3.9 REFUELING OPERATIONS l

3.9.1 Refueling Equipment Interlocks 3.9-1 3.9.2 Refuel Position One-Rod-Out Interlock 3.9-2 3.9.3 Control Rod Position.......

3.9-4 3.9.4 Control Rod Position Indication 3.9-5 3.9.5 Control Rod OPERABILITY-Refueling 3.9-7 3.9.6 Reactor Pressure Vessel (RPV) Water Level-Irradiated Fuel 3.9-8 3.9.7 Reactor Pressure Vessel (RPV) Water Level-New Fuel or Control Rods 3.9-9 3.9.8 Residual Heat Removal (RHR)-High Water Level 3.9-10 3.9.9 Residual Heat Removal (RHR)-Low Water Level 3.9-13 8 3.9 REFUELING OPERATIONS B 3.9.1 Refueling Equipment Interlocks.

B 3.9-1 (continued) t PERRY - UNIT 1 vi Revision No. 0 l

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[fm*0" TABLE OF CONTENTS Page 7 of 7 B 3.9 REFUELING OPERATIONS (continued)

B 3.9.2 Refuel Position One-Rod-Out Interlock B 3.9-5 B 3.9.3 Control Rod Position..

B 3.9-9 B 3.9.4 Control Rod Position Indication B 3.9-12 B 3.9.5 Control Rod OPERABILITY-Refueling.....

B 3.9-16 B 3.9.6 Reactor Pressure Vessel (RPV) Water Level-Irradiated Fuel B 3.9-19 8 3.9.7 Reactor Pressure Vessel (RPV) Water Level-New Fuel or Control Rods B 3.9-22 B 3.9.8 Residual Heat Removal (RHR)-High Water Level B 3.9-25 B 3.9.9 Residual Heat Removal (RHR)-Low Water Level B 3.9-30 3.10 SPECIAL OPERATIONS 3.10.1 Inservice Leak and Hydrostatic Testing Operation 3.10-1 3.10.2 Reactor Mode Switch Interlock Testing 3.10-4 3.10.3 Single Control Rod Withdrawal-Hot Shutdown 3.10-6 3.10.4 Single Control Rod Withdrawal-Cold Shutdown 3.10-9 3.10.5 Single Control Rod Drive (CRD) Removal-Refueling 3.10-13 3.10.6 Multiple Control Rod Withdrawal-Refueling 3.10-16 3.10.7 Control Rod Testing-Operating 3.10-18 3.10.8 SHUTDOWN MARGIN (SDM) Test-Refueling 3.10-19 8 3.10 SPECIAL OPERATIONS B 3.10.1 Inservice Leak and Hydrostatic Testing Operation.

B 3.10-1 B 3.10.2 Reactor Mode Switch Interlock Testing B 3.10-6 B 3.10.3 Single Control Rod Withdrawal-Hot Shutdown B 3.10-11 B 3.10.4 Single Control Rod Withdrawal-Cold Shutdown......

B 3.10-16 B 3.10.5 Single Control Rod Drive (CRD) Removal-Refueling B 3.10-21 B 3.10.6 Multiple Control Rod Withdrawal-Refueling B 3.10-26 B 3.10.7 Control Rod Testing-Operating..

B 3.10-29 B 3.10.8 SHUTDOWN MARGIN (SDM) Test-Refueling B 3.10-33 4.0 DESIGN FEATURES 4.1 Site Location 4.0-1 4.2 Reactor Core.

4.0-1 4.3 Fuel Storage..

4.0-2 j

5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.0-1 5.2 Organization...

5.0-2 5.3 Unit Staff Qualifications 5.0-4 5.4 Procedures.

5.0-5 5.5 Programs and Manuals 5.0-6 5.6 Re orting Requirements 5.0-16 5.7 Hi h Radiation Area 5.0-19 PERRY - UNIT 1 vii Revision No. O

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