Text

Proposed Tech Specs Pages,Increasing Present 100% Authorized Rated Thermal Power Level of 3579 Megawatts Thermal to 3758 Megawatts Thermal
	ML20211P287
Person / Time
Site:	Perry
Issue date:	09/09/1999
From:	; CENTERIOR ENERGY
To:	;
Shared Package
ML20138D182	List: ML20211P287; PY-CEI-NRR-2420, Application for Amend to License NPF-58,increasing Present 100% Authorized Rated Thermal Power Level of 3579 Megawatts Thermal to 3758 Megawatts Thermal.Proprietary SAR Encl. Proprietary SAR Withheld,Per 10CFR2.790(a)(4);
References
	NUDOCS 9909130178
	Download: ML20211P287 (67)
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r Attrchm:nt 2 PY-CEl/NRR-2420L Page 1 of 24 l

PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS i

i i

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9909130178 990909 ADOCK 05000440 PDR PDR

'P

1 PY-CE1MRR-2420L Page 2 or24

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renewal. Such safe and leaseback transactions are subject to January 23,1987, as supplemented on Ma a the i'

e application of Directorof the Office of Nuclear Reactor Regulation dai consenting to such transactions. Specifically, a lessor and

1987, acquire an interest under these transactions are prohibited i

e directly or indirectly any control over the licenses of PNPP rcising subsequently amended, are fuity appli n

. Forpurposes as may be interest to that lessor as long as the license for PNPP U ccessorin these financial transactions shall have no effect o 1

a ns in effect Nuclear facility throughout the term of the license se forthe Perry ;

(b) Further, the licensees are also required to notify the NRC i of these transactions ii the PNPP Operatin property insurance co;v(e) rage for PNPP Unit 1; a as part othem that may have an adverse effect on the safe opera C. This license shall be deemed to contain and is subject to the acility.

Commission's regulations set forth in 10 CFR Chapter i an conditions specifiedin the provisions of the Act and to the rules, regulations, and orders o ect to allapplicable j

hereafter in effect; and is subject to the additional conditions s i

ommission nowor below:

ecified orincorporated (1)

Maximum Power Level 358

' FENOC is authorized to operate the facility at reactor excess conomo(r.egawatts thermal (100% power) in accordance with thec ed herein.

om (2)

Technical Soecrfications Protection Plan contained in Appendix B, a ronmental are hereby incorporated into the license. FENOC shall op mendment No. 81, accordance with the Technical Specifications and the Env Plan.

yn w

n (3)

Antitrust conditions

a. Cleveland Electric illuminating Company, Duquesne Lig Edison Company OES Nuclear, Inc., Pennsylvania Pow o

, and the Amendment No. 96

fy L Definitions RR.2420L Page 3 of 24

l 1.1 Definitions (continued)

HINIMUH CRITICAL POWER RATIO (HCPR)

The MCPR shall be the smallest critical power ratio (CPR) that exists in the core for ea of fuel.

The CPR is that power in the asse that is calculated by application of the' mbly tie assembly to experience boiling trans divided by the actual assembly operating power.

MODE A H0DE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature. and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE-OPERABILITY A system subsystem. division.

device shall be OPERABLE or have. or it is capable of :erforming its specified safetyILITY w function s) and ween all necessary attendant instrumen(tation electrical power. controls, normal or emergency

. cooling and seal water itbrication and other auxiliary equipmen.t that are red for the system. subsystem. division.

. or device to perform its specified sa function s are also capabl their related su(pp) ort function (s).e of performing RATED 'HERHAL POWER (RTP)

RTP shall be a total reactor

r. ate to the reactor coolant of heat trans r 37gg REACTOR PRuiti.i10N The RPS RESPONSE TIME shall be that time SYSTDI (RPS) RESPONSE from whr. the monitored paraatter exceeds its RPS TIME trip s ioint at the channel sensor until de-ene zation of the scram pilot valve solenoi The response time may be measured b i

means of any series of sequential measured.ps so that the entire res overlap total ste or ponse time is surveillance requirements. Exceptions are stated in th (continued)

PERRY - UNIT 1 1.0-5 Amendment No. 69. 77

E c

PY CEUNRR 2420L Page 4 of 24 e'.

2.0 SAFETY LIMITS (SLs) 2.0 2.1 SLs i

2.1.1-Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 g or core flow

< 10% rated core flow:

3.% N r

o THERMAL POWER shall be s RTP l

2.1.1.2 With the reactor steam dome pressure a 785 psig and core flow

> 10% rated core flow:

MCPR shall be 21.09 for two recirculation loop ' operation or a 1.11 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant Sv! item Pressure SL t)

Reactor steam dome pressure shall be s 1325 psig.

I 2.2 SL Violations With any SL violation. the following actions shall be completed:

2.2.1 Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , notify the NRC Operations Center, in accordance with 10 CFR 50.72.

2.2.2 Within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 Insert all insertable control rods.

2.2.3 Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , notify the plant manager and the corporate executive.

responsible for overall pl. ant nuclear safety.

(continued) l PERRY - UNIT 1 2.0-1 Amendment No.104 l '-

4 i

.a.

[

Centrol Rod OPERABILITY 2420L 3.1.3 Page 5 of 24 ACTIONS (continued)

CONDITION REQUIRED. ACTION COMPLETION TIME _

D.

NOTE---------

D.1 Restore compliance 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Not applicable when with BPWS.

2 0.2 Restore control rod 4 h'ours Two or more inoperable to OPERABLE status, control rods not in compliance with banked position withdrawal not separa(BPWS) and sequence ted by two or more OPERABLE control rods.

t E.

Required Action and E.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, C, or D not met.

E 1

Nine or more control rods inoperable.

4 d

PERRY - UNIT 1 3.1-9 Amendment No. 69

Central Rod Patt:rn L

-2420L 3.1.6 Pgc 6 of 24 3.1 REACTIVITY CONTROL SYSTEMS r-3.1.6 Control Rod Pattern LCO 3.1.6 OPERABLE control rods shall comply with the requirements of the banked position withdrawal sequence (BPWS).

I9.01o APPLICABILITY:

MODESIand2withTHERMALPOWERshTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more OPERABLE A.1

NOTE---------

control rods not in compliance with BPWS.

Affected control rods may be bypassed in Rod Action Control System (RACS in accordance w)th i

SR 3.3.2.1.9.

Move associated 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months control rod (s) to correct position.

A.2 Declare assoc' fated 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months control rod (s) inoperable.

(continued) i i

?

PERRY - UNIT 1 3.1-18 Amendment No. 69

r t

MLHGR py.C 1/M 2420L Page 7 of24 3.2 POWER DISTRIBUTION LIMITS

.r 3.2.1 AVERAGEPLANARLINEARHEATGENERATI0k" RATE (APLHGR)

LC0 3.2.1 All APLHGRs shall be less than or equal to the limits specified in the COLR.

THERMALPOWERhhRTP.

APPLICABILITY:

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1

A.

Any APLHGR not within A.1 Restore APLHGR(s) to" 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limits.

within limits.

B.

Required Action and 8.1 Reduc ERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Com i

Time not met. pletion to <

.g,g s

l SURVEILLANCE REQUIREMENTS SURVEILLANCE l

FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or e to the limits specified in the COLR. qual Once within l

I rs after 1

h TP amt 23Efo l

l 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter PERRY - UNIT 1 3.2-1 Amendment No. 69 l

i MCPR Y-CE RR-2420L Page 8 of 24 3.2 POWER DISTRIBUTION LIMITS e

3.2.2 MINIMUN CRITICAL POWER RATIO (MCPR)

LC0 3.2.2 All MC?Rs shall be greater than or equal to the MCPR operating limits specified in the COLR.

APPLICABILITY:

THERMAL POWER 2 (TP.

ACTIONS CONDITION l

REQUIRED ACTION COMPLETION TIME A.

Any MCPR not within A.1 Restore MCPR(s) to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limits.

within limits.

B.

Required Action and 8.1 Redu ERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Com Time not met. pletion to <

TP.

2331o s_

i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal l

to the limits specified in the COLR.

Once ylthin I

rs after.

It RTP 4

ale 8%

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter PERRY - UNIT 1 3.2-2 Amendment No. 69

F 1

l Attachnxnt 2 LHGR i

PYCELHRR-2420L 3,2,3

)

Page 9 of24 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.

2.3B 7o APPLICABILITY:

THERMAL POWER ;t RTP.

ACTIONS CONDITION REQUIRED ACTION CCMPLETION TIME A.

Any LHGR not within A.1 Restore LHGR(s) to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limits.

within limits.

l l

B.

Required Action and 8.1 Redu ERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion to <

Time not met.

pTP.

{

-234 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to Once within the limits specified in the COLR.

I rs after:

l 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter 1

l l

PERRY - UNIT 1 3.2-3 Amendment No. 69

r II N R-2420L RPS Instrumentation Pop: 10 0f 24 ACTIONS (continued)

CONDITION REQUIRED. ACTION COMPLETION TIME _

D.

Required Action and 0.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, l

B, or C not met.

Table 3.3.1.1-1 for the channel.

E.

As required by E.1 Reduc ERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Required Action D.1 to <

RTP.

i and referenced in Table 3.3.1.1-1.

%gjbq 3.>

F.

As required by F.1 Reduc ERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Required Action D.1 to <

TP.

and referenced in Table 3.3.1.1-1.

l St3 896 G.

As required by G.1 Be in H0DE 2.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Required Action D.1 and referenced in Table 3.3.1.1-1.

H.

As required by H.1 Be in H0DE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action D.1 and referenced in Table 3.3.1.1-1.

I.

As required by I.1 Initiate action to Immediately Required Action D.1 and referenced in fully insert all Table 3.3.1.1-1.

insertable control rods in core cells containing one or more fuel assemblies.

i PERRY - UNIT 1 3.3-2 Amendment No. 69

1 m2 RPS Instrumentation py.ct:l/NRR 2420L Page ll of 24 3*3*I.1 SURVEILLANCE REQUIREMENTS

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

1.

Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2.

When a channel is placed in an inoperable status solely for performance o required Surveillances, entry into associated Conditions and Required maintains RPS trip capability. Actions may be delayed for up to 6 h SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.1.2

NO T E-------------

Not required to be performe il 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER 2 RTP.

2.38 %

Verify the absolute difference between 7 days the average power range monitor (APRM) channels and the calculated p s 2% RTP while operating at h TP.

g,gg 4.

SR 3.3.1.1.3 Adjust the channel to conform to a calibrated flow signal.

7 days i

SR 3.3.1.1.4

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

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST.

7 days (continued)

PERRY - UNIT I 3.3-3 Amendment No. 69

et 2 RPS Instrumentation evawna-2420t Page il of 24 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.16 Verify Turbine Stop Valve Closure and Turbine Contr 011 Pressure ol Valve Fast Closure Trip 18 months Low Functions are bypassed when THERNAL POWER is RTP.

t 38 'h)

(

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SR 3.3.1.1.17 Calibrate flow reference transmitters.

18 months t

SR 3.3.1.1.18

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

1.

Neutron detectors are excluded.

2.

For Functions 3. 4 and 5 in Table 3.3.1.1-1, the channel sensors are excluded.

t 3.

For Function 6 "n" equals 4 channels for the xarxse.

STAGGERE) T5T BASof determining

.......................IS Frequency.the Verify the RPS RESPONSE TIME is within limits.

18 nonths on a STAGGERED TEST BASIS l

e PERRY - UNIT 1 3.3-6

^.. ::b%d. No. 59, 77

RPS Instrtimentatica whm nt 2 Py(LI/NRR.2420L 3.3.1.1 i

Page 13 of 24 Table 3.3.1.1*1 (P89* 1 of 33 Reactor Protection System instrumentation

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7 Co elTIONS APPLICA8LE REQUltED REFERENCED

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MODES 04 OINER CNANNELs FROM sPECIFIED PER TalP REeuttED sLAVEILLAME FUNCTION Coe l110Ns SYSTEM ACTION D.1 kt0UIREMNTS At s e.a.yj Vagag 1.

Intermediate Range Monitors a.

Neutron 2

3 N

st 3.3.1.1.1 Fim. Nigh SR 3.3.1.1.4 1 122/125 st 3.3.1.1.6 divialens of SR 3.3.1.1.7 fuit scote st 3.3.1.1.13 a 3.3.1.1.15 5(*)

i 3

I st 3.3.1.1.1 s 122/125 st 3.3.1.1.5 divistene of st 3.3.1.1.13 st 3.3.1.1.15 futi seete b.

Inap 2

3 N

st 3.3.1.1.4 NA st 3.3.1.1.15 5(e) j 3

I st 3.3.1.1.5 NA a 3.3.1.1.15 2.

Average Power Range Monitors l

a.

Neutron 2

Flun aNIgh, 3

N st 3.3.1.1.1 s 301 RTP setdoun SR 3.3.1.1.4 st 3.3.1.1.7 st 3.3.1.1.s o,62gW+63.8%

a 3.3.1.1.11 SR 3.3.1.1.15 b.

Fteu 9 faced 1

3 E

R 3.3.1.1.1 i

sledated Thenoot M 3.3.1.1.2 i

Power-Nigh and R 3.3.1.1.3 TT131 RTP"'

N 3.3.1.1.8 st 3J.1.1.9 st 3.3.1.1.11 st 3.3.1.1.14 st 3.3.1.1.1$

St 3.3.1.1.1T SR 3.3.1.1.18 (continued) l (a)

With enr centret red withd ami from a core teLL centelnine one or more fuel tteu assemblies.

(b)

A.t. sbte.Vetue is a ti.n w.....

TP town reset for single loop eparation per LCO 3.4.1

.e.

.675+ +3 57o

~

PERRY - UNIT 1 3.3-7 Amendment No. 69

\\

RPS Instrumentatirn PY<E1/NRR.2420L 3.3.1.1 Page 14 of 24 Table 3.3.1.1 1 (page 2 of 3)

Reactor Protection system Instrumentation

{

4 APPLICABLE CONDITIOWs MODES OR REQUIRED REFERENCED OTNER CHANNELS Ft0M SPECIFIED PER TRIP REeUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REGulREMEWis

, VALyg 2.

Averese Power tense Monitors (continued) c.

flued Neutron 1

3 G

st 3.3.1.1.1 1 120% RTP fIun - NIsh SR 3.3.1.1.2 St 3.3.1.1.8 st 3.3.1.1.9 st 3.3.1.1.11 st 3.3.1.1.15 st 3.3.1.1.18 4

j d.

Inap 1,2 3

N st 3.3.1.1.8 NA st 3.3.1.1.9 st 3.3.1.1.15 3.

Asoctor vessel steen 1,2 2

N st 3.3.1.1.1 s 1079.7 Dome Pressure-Nigh st 3.3.1.1.9 psig st 3.3.1.1.10 st 3.3.1.1.13 st 3.3.1.1.15 st 3.3.1.1.18 4

Reactor vessel Water 1,2 2

N st 3.3.1.1.1 1 177.1 Levet -Low, Level 3 st 3.3.1.1.9 inches st 3.3.1.1.10 3*b st 3.3.1.1.13 st 3.3.1.1.15 SR 3.3.1.1.18 5.

Reester vessel Water t 2SE TP 2

F st 3J.1.1.1 s 220.1 Level = Nigh, Level 8 at 3.3.1.1.9 inches s

at 3J.1.1.10 Et 3.3.1.1.13 st 3.3.1.1.15 st 3.3.1.1.18 6.

Mein steen Isolation 1

8 G

st 3.3.1.1.9 2 125 vaive -Closure SR 3.3.1.1.13 etosed st 3.3.1.1.15 st 3.3.1.1.18 7 Drywell Pressure-algh 1,2 2

N st 3.3.1.1.1 1 1.88 SR 3.3.1.1.9 pele at 3.3.1.1.10 st 3.3.1.1.13 st 3.3.1.1.15 (continued)

PERRY - UNIT 1 3.3-8 Amendment No. 69

RPS Instrumentation py.cEl/NRR.2420L 3.3.1.1 Page l$ of 24 Table 3.3.1.1 1 (psee 3 of 3)

Reactor Protection system Instrumentation APPLICA8LE CONDITIONS MODES OR REQUIRED REFERENCED OTNER CNANNELS FROM SPECIFIED PER TRIP REGuinED sutVEILLANCE ALLOWLSLE FUNCTION CONDITIONS SYSTEM ACTION 0.1 REeultEMENTS VALUE 8.

seren Discharge Volume Water Level-Nigh s.

Trenamitter/ Trip unit 1,2 2

N st 3.3.1.1.1 2 38.87 inches st 3.3.1.1.9 st 3.3.1.1.10 st 3.3.1.1.13 st 3.3.1.1.15 5(8) 2 i

st 3.3.1.1.1 s 38.87' Inches at 3.3.1.1.9 st 3.3.1.1.10 st 3J.1.1.13 st 3.3.1.1.15 b.

F1 eat switch 1,2 2

N st 3.3.1.1.9 5 626 ft st 3.3.1.1.13 11.5 inches st 3.3.1.1.15 elevation 5(*)

2 I

st 3.3.1.1.9 s 626 ft SR 3J.1.1.13 11.5 inches SR 3.3.1.1.15 elevation 9.

Turbine step Wat9e Ctesure 1

ATP 4

E a 3J.1.1.9 s FK closed at 3.3.1.1.13 39.T, st 3.3.1.1.15 at 3.3.i.i.a st 3.3.1.1.18

10. Turbine Centrol Valve t

TP 2

E st 3.3.1.1.9 t 465 psis Fast Cteeure, Trip OIL Pressure-Lou SR 3J.1.1.13 83 3J.1.1.15 i

4 M 3 J.1.1.16

'~

m 3.3.1.1.18 11 Reactor Mede 1,2 2

u st 3.3.1.1.12 NA sultch-shutdoun Position SR 3.3.1.1.15 5(a) 2 I

st 3.3.1.1.12 NA st 3.3.1.1.15

12. Manuel scram i

1,2 2

N st 3.3.1.1.5 NA a 3.3.1.1.15 5(*)

2 I

a 3.3.1.1.5 NA at 3J.1.1.15 ta)

With enr centrol red withdrawn from a core cett containing one or more fuel asseebtles PERRY - UNIT 1 3.3-9 Amendment No. 69

(

l

I l

Centrol Rod Black Instrumentatica Att.enent 2 py(El/NRR.2420L Page 16 of 24 3.3.2.1 SURVEILLANCE REQUIREMENTS r

NOTES 1.

Refer to Table 3.3.2.1-1 to determine which SRs aprily for e Block Function.

o od 2.

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

unction SURVEILLANCE FREQUENCY SR 3.3.2.1.1

NOTE-Not required to be perfoi after THERMAL POWER is >

until I hour

....____...____________.pf05 TP.

I l

Perform CHANNEL FUNCTIONAL TEST.

92 days SR 3.3.2.1.2

------- NOTE Not required to be perf af r THERMAL POWER is until I hour s

TP.

RTP an 3,3h 66,9 Perform CHANNEL FUNCTIONAL TEST.92 days SR 3.3.2.1.3

NO TE-Not required to be performed until I hour after any control rod is withdrawn in MODE 2.

Perform CHANNEL FUNCTIONAL TEST.92 days (continued)

PERRY - UNIT 1 3.3-16 Amendment No. 69 i

~

Centrol Rod Black Instrumentatica Attachrwn 2 PYCEl/NRR 2420L Page 17 of 24 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.1.4

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

Not required to be perfo until I hour after THERMAL POWER is s TP i a

MODE 1.

l9.0'fo l

Perform CHANNEL FUNCTIONAL TEST.

92 days 1

SR 3.3.2.1.5 Calibrate the low sower setpoint trip u

The All le Value shall be 92 days 19S7o

' "" I (*

33.37, g

SR 3.3.2.1.6 Verify the RWL high power Functio' s not 92 days bypassed when THEEi omrR is >

l RTP.

.2$

w SR 3.3.2.1.7 Perform CHANNEL CALIBRATION.

184 days SR 3.3.2.1.8

-- -------NOTE Not required to be performed until.I hour after reactor mode switch is in the shutdown position.

Perform CHANNEL FUNCTIONAL TEST.

18 months 1.

(continued) i l

PERRY - UNIT 1 3.3-17 Amendment No. 69

Control Rod Bl d Inst a ntation Py.cotm an.2420L Page 18 of 24 3.3.2.1 i

{

T, Table 3.3.2.1 1 (post-1 of 13 Control Red Block Instrsamentation APPLICA8LE MODES OR OfflER

~

FLAICTION SPECIFIED REQUIRED CONDITIONS CIUulNELS SLAVEll13 M

{

REQUIEEnuffS 1.

Rod Pettern control system

s. Rod withdrawal tialter (e) 2 SR 3J.2.1.1 SR 3.3.2.1.4 SR 3.3.2.1.9 (b) 2 SR 3J.2.1.2 e i SR 3J.2.1.3 SR 3.3.2.1.7

b. Rod pottern centrotter M

82 3J.2.1.9 1

,2 2

sa 3J.2.1.3 SR 3.3.2.1.4 SR 3.3.2.1.5 SR 3.3.2.1.7 2.

Reetter Mode Switch-Shutdown Position St 3J.2.1.9 (d) 2 SR 3.3.2.1.3 (s) TIIERMAL POER (b)

Tumuu. PouEn >@aTP.nd shtP.

(c) With TilERMAL POWR 8 gg TP.

(d)

Reactor mode switch in the shutdown position.

le PERRY - UNIT 1 3.3-13 Amendment No. 85

EOC-RPT Instrumentatien Ana.wm2 PY{T.UNRR-2420L 3.3.4.1 Pmge 190f 24 3.3.INSTRUNENTATION r-3.3.4.1 End of Cycle Recirculation Pump T' rip (EOC-RPT) Instrumentation LCO 3.3.4.1 Two channels per trip system for each E0C-RPT instrumentation Function listed below shall be OPERABLE Turbine stop Valve (TSV) Closure; and a.

b.

Turbine Control Valve (TCV) Fast Closure, Trip 011 Pressure--Low.

APPLICABILITY:

THERMAL POWER h 40%

speed.

TP with any recirculation pump in fast ACTIONS 25 ff 16'

NOTE-Separate Condition entry is allcwed for each channel.--------------------------

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more required A.1 Restore channel to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months channels inoperable.

OPERABLE status.

QR A.2 MTE-Not applicable if inoperable channel is the result of an inoperable breaker.

Place channel in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months trip.

(continued)

)

PERRY - UNIT 1 i

3.3-26 l

Amendment No. 69

EOC-RPT Instrumentatien PY-CI:I/NRR 2420L

*4+1 Page 20 of 24 ACTIONS (continued)

~#

CONDITION REQUIRE [D ACTION COMPLETION TIME l

B.

One or more Functions B.I with E0C-RPT trip Restore EOC-RPT trip 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months l

capability not capability.

maintained.

1 l

C.

Required Action and C.1 Remove the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion Time not met, recirculation pump fast speed breaker from service.

08 C.2 Redu ERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to <

{.

SURVEILLANCE REQUIREMENTS

NOTE---------

When a channel is placed in an inoperable status solely for performan may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided the as E0C-RPT trip capability.

SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CNANNEL FUNCTIONAL TEST.

92 days 1

(continued)

PERRY - UNIT 1 3.3-27 Amendment No. 69

hyTrI*[a[.2420L EOC-RPT Ir.strumentatica Page 21 of 24 3.3.4.1 SURVEILLANCE REOUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.4.1.2 Perform CHANNEL CALIBRATION.The Allowable Values shall be:

18 months a.

TSV Closure: 5 7% closed; and b.

TCV Fast closure, Trip 011 Pressure-Low: 1 465 psig.

SR 3.3.4.1.3 Perform LOGIC SYSTEM FUNCTIONAL TEST, including breaker actuation.

18 months SR 3.3.4.1.4 Verify TSV Closure and TCV Fast Closure, 18 months Trip 011 Pressure-Low Functions not bypassed when THERMAL POWER is 1 TP.

i ss.%

[

W SR 3.3.4.1.5

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

Breaker arc suppression time may be assumed from the most recent perfomance of SR 3.3.4.1.6.

.__ 4 -

Verify the EOC-RPT SYSTEM RESPONSE TIME is within limits.

18 months on a STAGGERED TEST BASIS SR 3.3.4.1.6 Determine RPT breaker are suppression 60 months time.

PERRY - UNIT 1 3.3-28 Amendment No. 69

Primary Containmsnt and Dryw ll 1:clation Instrume mhrnern 2 PY.CE 1/NRR.2420L Page 22 of 24 3.3.6.1 Table 3.3.6.1 1 (page 1 of 6)

Primary Contairveent erws Crywelt isolat en Instrumentation APPLICAsLE CONDITIONS M(BES OR REQUIRED REFERENCED OTNER CHAhMELS FR(M FUNCTION SPECIFIED PER TRIP REQUIRED 4W WEILLANCE l

COWITIONE SYSTEM ACTION C.)

RteulREfENTS ALLoutsLE 1.

Main Steam Uneisolation

., VALg s.

Reador VesselWater Level-Low Low Low.

113 2

D SR 3.3A1.1 a 14.3 inches Level 1 SR 3.3A1.2 SR 3.3A1.3 SR 3.3A1A SR.3.3A1.5 b.

Main Steam Line 8R 3.3A1A Pressure. Low 1

2 E

8 R 3 2 A 1.1 SR 3.3A.12 2 7952 psig SR 32A.13 SR 3.3A.1A 8R 316.1.6 Able c.

Main Steam Line SR 316.1A Flow. Higt) 1.2.3 2 perMSL D

SR 316.1.1 SR 316.1.2 l

s SR 3.3A1.3 1

SR 316.1A S R 3 1 6.1 E 8R 316.1A d.

Condenser Vacuum -

1,2(a) low 2

D SR 32A1.1 a 7AIndies 3(a)

SR 32A1.2 Hgveauum SR 3.3A1.3 SR 316.1A Main Steam Line Pipe SR 3AA1A e.

1A3 2

D S R 3 1 6.1.1 s166.9'F 1

TunnelTemperature-Higt!

S R 3.3 A 1.2 SR 33A1A SR 33A1A f.

Main Steam Line SR 33A1.7 Turbine Builde' g 1.2.3 2

D Temperature Hegh SR 316.1.1 s 133.9'F SR 3.3A12 SR 3.3A1A SR 316.1.5 g.

Manualinitiation 1,2,3 2

G SR 316.1.5 NA 2.

Prl.

Containmentand Drywell

a. KoectorVesselWater level-Low Low. Level 2 1,2,3 2M N

M

.6.1."

a.127.6 inches i

,8 j',Lj* {

l E

$* :i:N (a)

With any turbine stop valve not closed.

(continued)

(b)

Required to intttate the assectated drywett isolation function PERRY - l#HT 1 3.3 54 d eui. No. 93 L

Anachnmit 2 PY CEl/NRR-2420L rage 23 or24 Main Tud e D ass Dstem 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Main Turbine Bypass System l

LCO 3.7.6 The Main Turbine Bypass System shall be OPERABLE.

APPLICABILITY:

THERMAL POWER 1 RTP.

ACTIONS CONDITION l

REQUIRED ACTION COMPLETION TIME A.

Main Turbine Bypass A.I Restore Main Turbine 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months System inoperable.

Bypass System to OPERABLE status.

)

B.

Required Action and B.1 Reduc ERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion Time not met.

to <

RTP.

3 13 3b SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify one complete c turbine bypass valve.yc1,e of each main 31 days SR 3.7,.d.2 Perform a system functional test.

18 months SR.3.7.6.3 Verify the TURBINE BYPASS SYSTEM RESP 0615E TIME is within limits.

18 months PERRY - UNIT I 3.7-13 Amendment No. 69

Programs and Manuals 5.5 Attachnent 2 PY{El/NRR 2420L Stp24 Programs and Manuals 5.5.12 Primary Containment Leakaae Rate Testino Proaram (continued)

BN-TOP-1 methodology may be used for Type,A tests.

The corrections to NEI 94-01 which are identified on the Errata Sheet attached to the NEI letter. "A)pendix J Workshop Questions and Answers," dated Marc 119,1996 are considered an integral part of NEI 94-01..

The containment isolation check valves in the Feedwater p(Technical Specification 5.5.6).enetrations are tested per the Ins

\\6 G.40 psiO' The peak calculated primary. containment internal orehs ure esign basis loss of~ coolant accidentg.. n 7.00 c g%I W st,,f\\lc Qm2

) is u d t.b.J /The maximum allowable primary containment leakage rate L. shall be 0.20% _of primary containment air wei ht WHfdItaggppeak containment pressure (P.g). per day at the 3,

Leakage rate acceptance criteria are:

a.

Primary containment leakage rate acceptance criterion is s 1.0 L. However, during the first unit startup following testing, performed in accordance with this Program, the leakage rate acce for the T and Type C tests,ptance criteria are < 0.6 L,A tests; ype B and s 0.75 L, for the Type b.

Air lock testing acceptance criteria are:

1)

Overal.1 air lock' leakage rate is s 2.5 scfh when tested at 2 P.

2)

For each door, leakage rate is s 2.5 scfh when the gap between the door seals is pressurized to a P..

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 are ap?rogram.)licable to the Primary Containment Leakage Rate Testing (continued)

PERRY - LNIT 1 5.0-15a Aiu ded. No.105

Att:chment 3 PY-CEl/NRR-2420L Page 1 of 4 CLARIFICATION OF CHANGES NOT MADE TO TECHNICAL SPECIFICATIONS i

.i

l Att: chm:nt 3 i

PY-CEl/NRR-2420L Page 2 of 4

{

RESOLUTION OF TECHNICAL SPECIFICATION CHANGES NOT MADE

, Technical";+: :r: x

.m

. W Resolution,

mm-

,aca'-

4

1. Page 3.1-13, SR 3.1.4.1

1. This value is retained at 40% Rated Thermal Power (RTP)

Frequency -40% RTP f r p wer uprate. This is justified based on the following:

4 This value is not based on any safety analysis or analytic limit (e.g., it is not related to the Turbine Stop Valve / Turbine Control Valve (TSV/TCV) scram bypass setpoint changes).

The value for this parameter is based on engineering judgment and recognition of the operctional restrictions associated with control rod scram time testing below the Rod Pattern Controller (RPC) - Low Power Setpoint (LPS).

The intent is to provide a " reasonable window" above the RPC-LPS such that it precludes the need for out-of-sequence rod withdrawals; but also ensures that scram time testing is completed within a " reasonable time". The current value (40% RTP) is acceptable, considering the typically larger thermal margins at these lower power levels and considering the additional surveillances that ensure control rod OPERABILITY. e.g., frequent verification of adequate accumulator pressure (SR 3.1.5.1), and scram time testing after work on control rods or the control rod drive system that could affect scram time (SR 3.1.4.3).

Retaining this at 40% RTP, therefore, does not invalidate the safety design basis or the intent of this SR.

2. Page 3.1-13, SR 3.1.4.4

2. Same as item 1 above.

Frequency - 40% RTP

3. Page 3.3-7, Table 3.3.1.1-1,

3. This value is retained at 20% RTP for power uprate. This is Function 2a Allowable Value -

justified based on the following:

This value is not based on a specific safety analysis or 20% RTP analytic limit.

This Function indirectly ensures that reactor power does not exceed the reactor core safety limit at low pressure and low flow conditions (i.e.,25% RTP pre-uprate and 23.8% RTP post-uprate) before the mode switch is placed in the run position.

The Allowable Value for this parameter is based on preventing significant increases in power when THERMAL POWER is less than the reactor core safety limit at low pressure and low flow conditions (i.e.,25%

RTP pre-uprate and 23.8% RTP post-uprate).

Attachm:nt 3 PY-CEl/NRR-2420L Page 3 of 4 RESOLUTION OF TECHNICAL SPECIFICATION CHANGES NOT MADE MTechnicalSpecification.

, a : or im mResolution emu mW-

4. Page 3.3-7, Table 3.3.1.1-1,

4. This value is retained at 113% RTP for power uprate. This Function 2b Allowable Value-is justified based on the following:

113% RTP in terms of absolute power, the power uprate safety analyses are based on a 5% increase (i.e., proportional to the increase in RTP) in the analytic limit for this Function.

Therefore, the analytic limit and Allowable Value for this Function are unchanged when expressed in terms of relative power.

5. Page 3.3-8, Table 3.3.1.1-1,

5. Same as item 4 above.

Function 2c Allowable Value -

120% RTP

6. Page 3.4-1, Comp!etion Time

6. This value is retained at 5% RTP for power uprate. This is for Condition C -5% RTP justified based on the following:

The BWR Owners' Group (BWROG) guidelines for Stability I

interim Corrective Actions to better address startup and low power maneuvering conditions has been accepted by the NRC and supported by the Perry Nuclear Power Plant (PNPP). Because these guidelines are intended for use at PNPP until replaced by a stability long-term solution, modification of the PNPP Technical Specifications is not appropriate (Reference 9/9/94 PNPP letter to NRC -

" Response to Generic Letter 94-02 Regarding Long-Term

)

Solutions and Upgrade of Stability interimCorrective Actions"

- PY-CEl/NRR-1855L).

7. Page 3.4.5, Figure 3.4.1-1

7. Same as item 6 above.

8. Page 3.4-8, SR 3.4.3.1,

8. This value is retained at 25% RTP for power uprate. This is Note 2 - 25% RTP justified based on the following:

This value is not based on any safety analysis or analytic limit (e.g., it is not related to the low pressure / low flow safety limit which was changed from 25 to 23.8% RTP elsewhere in the markups).

The value for this parameter is based on engineering judgment and, as indicated in the Bases for this SR, takes in to consideration operational constraints and operating conditions necessary to obtain repeatable and meaningful data.

)

i

I Attachmrnt 3 PY-CEl/NRR-2420L Page 4 of 4 RESOLUTION OF TECHNICAL SPECIFICATION CHANGES NOT MADE

wTechnical Specification +,

Y'

.w.

> w;c Resolution : ;'e.*

s

9. Page 3.4-30, SR 3.4.11.8,

9. This value is retained at 30% RTP for power uprate. This is justified based on the following:

Note and SR 3.4.11.9 Note -

30% RTP The Bases for these SRs (Page B 3.4-62,2*' paragraph) currentiy states: Plant specific test data has determined that the bottom headis not subject to temperature stratification with natural circulation at powerlevels as low as 30% of RTP or with any single loop flow rate when the recirculation loopjet pump flowis > 50% of rated core flow. Thus, if Thermal Power is S30% RTP orif recirculation loop jet pump flow is s50% of rated core flow, THERMAL POWER and recirculation loop flow increases shall be suspended.

Therefore, SR 3.4.11.8 and SR 3.4.11.9 have boen modified by a Note that requires the Surveillance to be met only when THERMAL POWER or recirculation loop flow is being increased when the above conditions exist. Based on the logic of this paragraph, maintaining this value at 30% is conservative and does not invalidate the intent of the spec.

10. Page 3.6-36, 10.This value is retained at 1% RTP for power uprate. This is LCO 3.6.2.1 a., b., c, and justified based on the following:

Condition A - 1% RTP The Bases (bottom of page B 3.6-71) currently states:

...when the reactoris producing power essentially eauivalent to 1% RTP, heat input is annroximately eaual to normalsystem heatlosses.

The difference between 1% and (1% + 1.05) is considered negligible based on the following:

1% was selected based on engineering judgment (i.e., it is not based on a specific safety analysis or analytic limit).

it is not practicable to measure the difference.

it is expected, based on engineering judgment, that a 5%

increase in this value would not result in exceeding the design basis maximum allowable values for primary containment temperature or pressure if an accident were to occur at this low power.

11. Page 3.6-37, 11.Same as item 10 above.

Required Action B.1 and Condition C - 1% RTP 1

Att: chm:nt 4 PY-CEl/NRR-2420L Page 1 of 38 PROPOSED CHANGES TO TECHNICAL SPECIFICATION BASES "INFORMATION ONLY"

^ ((

Reactor Core SLs PY 2001.

Pasc 2 0f 38 8 2,1,1 BASES APPLICABLE 2.1.1.1~

Fuel Claddina Inteority (continued)

SAFETY ANALYSES this flow is approximatindicate that the fuel assem xwer at M.fo 7o desicn osakimfac'. ors.ely 3.35 Mwt. Witi the this corres:xmds to a mtt@ AL i El 5@RTP.

Thus, a TIER limit R

or reactor pressure < MAL POWER 33g is wnse ve.

~

785 psig 2.1.1.2 5;P.E 1he fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the lim is not violated.

Since the parameters that result in fuel damage are not directly observable during reactor op the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the be11nning of the region in which fuel damage could oc Alb. hough it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods. the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit.

the procedures used to calculate the critic in an uncertainty in the value of the critical power.

Therefore, the fuel cladding integrity SL is defined as the more than 99.9% of the fuel rods in the c avoid boiling transition, considering the power distribution within the core and all uncertainties.

combines all the uncertainties in operating the procedures used to calculate critical power. meters and probability of the occurrence of boiling transition is The determined using the approved General Electric critical power correlations Details of the fuel cladding integrity SL calculation are.given in Reference 2.

includes a tabulation of the uncertainties used in theR the parameters used in the MCPR SL stati h iinued)

PERRY - UNIT 1 B 2.0-3 Revision No. 0

PY-CEIMRR 2420L Control Rod OPERABILITY Page 3 0f3:

B 3.1.3 BASES y

ACTIONS C.1 and C.2 (continued)

The allowed Completion Times are reasonable, cons{

small number of allowed ino)erable control rods and pr time to insert and disarm t1e control rods in an, orderly manner and without challenging plant systems.

D.1 and D.2 19.0 70 Out of sequence control rods may increase the potential

~1vity worth of a drop)ed control rod during a CRDA.

TP. the generic banced position withdrawal sequence At

(

) analysis (Ref. 7) requires inserted control rods not-l in com)11ance with BPWS to be separated b OPERAB.E control rods in all directions, y at least two including the diagonal.

Therefore, if two or more inoperable control rods are not in compliance with BPWS and not separated by at least two OPERABLE control rods, action must be taken to restore compliance with BPWS or restore the control rods to OPERABLE status. A Note has been added to the.Cond clarify that the Condition is not applicable when >

to since the BPWS is not required to be followed under i

conditions, as described in the Bases for LCO 3.1.6.

e allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is acceptable. The considering the low probability of a CRDA occurring. (gg j

g If any Required Action and associated Completion Time of Condition A. C. or 0 are not met or nine or more inoperable.

control rods exist which the LCO does the plant must be brou.1ht to a MODE in not apaly.

plant must be brought to ODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .To achieve ".his This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active 19*0%

function (i.e.. scram) of the control rods.

i rol rods The ntaber of RTP (i.e.permittad to be inoperable when operatin) above value spec. no CRDA considerations) could be m ino)erable control rods could be indicative of a generic problem,should be undertaken. pro)1em and investigation The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to and without challenging plant systems. reach MODE 3 (continuedi PERRY - UNIT 1 B 3.1-1B Re. vision No.1

[y$",n.242" Control Rod Pattern B 3.1.6 BASES APPLICABLE banked position withdrawal sequence (BPWS h0$

SAFETY ANALYSES (continued)

Reference 8.

The BPWS is applicab frm the condition f all control rods fully inserted t 01 iP (Ret. 11 7or the BPWS. the control rods are requ ed to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions (e.g.,

between notches OB and 12). The banked positions are defined to minimize the maximum incremental control worths without being overly restrictive during normal plant operation. The generic BPWS analysis (Ref. 8) also evaluated the effect of fully inserted inoperable control number (i.e., eight) and distribution of fully ins inoperable control rods.

Rod pattern control satisfies the requirements of Criterion 3 of the NRC Policy Statement.

LC0 Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the BPWS.

This LC0 only applies to OPERABLE control rods.

For inoperable control rods required to be inserted, separate irements are specified in LCO 3.1.3 " Control Rod r

OP ILITY." consistent with the allowances for im2Je control rods,in the BPWS.

p, APPLICABILITY In MODES 1 and 2. when THERHAL POWER is s RTP.

CRDA is a Design Basis Accident (DBA) and, thereTore, coupliance, with the assumptions of When THERMAL POWER is safety analysis is required.

19.0$

com,roi roo udivuratt RTP. there is no credible t results in a control rod worth that could exceed the 280 cal /gm fuel damage limit during a CRDA (Ref. 1).

In H0 DES 3, 4. and 5. since the reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, te SDH ensures that the consequences of a CRDA are a

able. since the reactor will remain subcritical with a ac

~

sing e control rod withdrawn.

(continued)

PERRY - UNIT 1 B 3.1-35 Revision No. 1

^"*g.,

Control Rod Pattern Page5of38 8 3,1,g BASES (continued) c ACTIONS A.1 and A.2 With one or more OPERABLE control rods no with the prescribed control rod sequence,t in compliance taken to either correct the control rod pattern or decaction ina the associated control rods inoperable within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .lare Noncompliance with the prescribed sequence may be th

___' '3 of " double notching." drift 10$ng scram valv 19.0 k cocling water m 1ent. lea recuction to d control rod pai RTP before establishing the correctpower' not in conpliance with the prescribed s. The number of OPE eight to prevent the operator from a is limited to ing to correct a control rod pattern that significantly iates from the prescribed sequence. When the control r compliance with the prescribed sequence od pattern is not in all control rod movement should be stopped except for moves needed to correct the control rod pattern, or scram if warranted.

Required Action A.1 is modified by a Note which allows control rods to be bypassed in Rod Action C control rods to be returned to their correct position.

ensures that the control rods will be moved to the corre This -

position.

A control rod not in compliance with the prescribed sequence is not considered ino>erable except as ired by Required Action A.2.

OPERABI.ITY of ceritrol is determined by com

" Control Rod Scram Times"pliance with LCO 3.1.3: LCD 3.1.4

and LCO 3.1.5. " Control Rod Scram Accumulators." The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, considering the restrictions on the number of i

allowed out of sequence control rods and the low probability of a CRDA oc of sequence. curring during the time the control rods are out

\\

B.1 and B.2 1

the control rod pattaen significantly deviates fr prescribed sequence. Control rod withdrawal should be suspended insediately to prevent the po deviation from the prescribed sequence.tential for further Control rod insertion to correct control rods withdrawn be allowed position is allowed since. in general. yond their insertion of control rods has less 1spact on control rod worth than h dinued)

PERRY - UNIT 1 8 3.1-36 Revision No.1

Control Rod Pattern

-2420L Page 6 of 38 B 3.1.6 BASES e

~

ACTIONS B.1 and B.2 (continued) withdrawals have.

that allows the affected control rods to be bypa in accordance with SR 3.3.2.1.9 to allow insertion only.

i with BPWS. the reactor mode switch must shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . With the reactor mode does not meet the applicability requirements The allowed Completion Time of I hour is reasonable to allow insertion of control rods to restore appropriate relative to the low probabil 11ance, and is of a occurring with the control rods out of sequence. CRDA SURVEILLANCE SR 3.1.6.1 REQUIREMENTS The control rod pattern is verified to be in compliance with the BPWS at a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency, ensuri the asstaptions of the CRDA analyses are met. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months requency of this Surveillance was devel consideri that the primary check of the control pattern c ance with the BPWS is performed by the RPC (LCO 3.3.2.1).

The RPC provides control rod blocks to enforce the required control rod WF A9.ohsenuence and is ran" ired to be OP a

" Modifications to the Requirements for Control Rod l

REFERENCES 1.

July 1987. Drop Accident Hitigating Systems." BWR Owners 2.

USAR. Section 15.4.9.

i 3.

NUREG-0979. "NRC Safety Evaluation Report Related to the Final Design Approval of the GESSAR II BWt/6 Nuclear Island Design. Docket No. 50-447."

Section 4.2.1.3.2. April 1983.

4.

NUREG-0800. " Standard Review Plan." Section 15.4.9.

" Radiological Consequences of Control Rod Drop Accident (BWR)." Revision 2. July 1981.

(s iinued)

PERRY - UNIT 1 B 3.1-37 Revision No. 1

Y CEl/N -2420L Page 7 of 38 g 3.2.1 BASES r

APPLICABLE core flow increases) over a range of power SAFETY ANALYSES (continued) conditions.

>ower deaendent multipliers. MAPFAC,ie generated.

to initial core flow levels at power levels below th. response n-which turbine stop valve closure and tur ose at 23.7%

core flow MAPEAC._ limits are provided for operation at p levels RTP and the power level.

viously mentioned bypass' re reduced by MAPFAC a t APLHGR limits are at various operating conditions to ens,re that all fuel design criteria are met u

for normal operation and A00s.

analysis code is provided in Reference 6.A complete discussion of determined APLHGR limits are adequate maximum oxidation limits of 10 CFR 50.46.

with the requirements of 10 CFR 50. Appendix The PCT following a postulated LOCA is a average heat generation rate of all the rods of a fuel i

assembly at any axial location and is not stro assembly.

LHGR of the highest powered fuel rod assumed analysis divided its local peaking factor. A conservative multi lier is applied to the LHGR assumed in the LOCA analysis o account for the uncertainty associated with the measurement of the APLHGR.

For single recirculation loop operation. the MAPFAC multiplier is limited to a maximum value which is specified in the COLR.

This multiplier is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one recirculation loop available resulting in a more severe cladding heatup during a LOCA.

Statement.The APLHGR satisfies Criterion 2 of the NRC Policy (continued)

PERRY - UNIT 1 B 3.2-2 Revision No. O

I fYC HGR 2420L Page 8 of 38 BASES (continued)

LC0 The APLHGR limits specified in the COLR are the re fuel design. DBA. and transient analyses. For two recirculation loops operating, the limit is detennined multiplying the smaller of the MAPFACf and HAPFAC times the exposure dependent APLHGR limits.

recirculation loop in operation. in conformance with theWit requirements of LCO 3.4.1. " Recirculation Loops Ope i

the limit is determined by multiplying the exmsure dependent APLHGR limit by the smallest of MAP?ACr. H and the limiting value smeified for single recirculation,,

loop operation in the CO.R. which has been determin specific single recirculation loop analysis (Ref. 2).

APPLICABILITY The APLHGR limits are primarily derived from fuel des evaluations and LOCA and transient analyses that are a to occur at high power levels.

Design calculations and tie margin to the required APLHGR limitsoxrating e trend continues down to the po mases This when entry into NODE 2 occurs.wer rance o I t3 1" RTP Whenin rapid scram initiation during any significa DN thereby effectively removing any APLHGR limit compliance

^

o in H00E 2.

RTP. the reactor oTherefore, at THERHAL POWER levels 6143 LHGR limits: thus,perates with substantial margin to l

th this LCO is not required.

(

ACTIONS M

thrJ8J,tofy,g If any APLHGR exceeds the required limit an assumption regarding an initial condition of the DBA=and. transient analyses may not be met.

to restore the APLHGR(s) to within the required limit such that the plant will be operating within analyzed The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Coupletion Time is sufficient to re APLHGR(s) to within its limit and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR out of specification.

k ed.inued)

PERRY - UNIT 1 8 3.2-3 Revision No. O

[.

M y

-2420L Page 9 of 38

*l BASES e

ACTIONS B.l (continued)

If the APLHGR cannot be restored to within its required be brought to a MODE or other specifie g

tha 10 does not apply. Io achieve this status. T POWEl must be reouced to 925BRTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowed Completion Time is 11Fasonable, b The 7

I order 7 man. to rWe THERMAL POWER to <g on operating.

ernar' ence 3.8

~- RTP in an ner and witnour, cnaiienging p' alif systems.

SURVEILLANCE SR 3.2.1.1 REQUIREMENTS 23.g$

i APLHGRs are required to be initia '

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is = "

iculated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter. They are c RTP and then every '

to the specified within the assumptions of the safety analysis Frequency is based on both engineering judgment andThe' 24 h recognition of the slowness of changes in power distribution under normal c The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after ons.

THERNAL POWER RTP is achieved, is acceptable given the -

large inhe in to operating limits at low power levels.

3,g w

REFERENCES 1.

NEDE-24011-P-A " General Electric Standard Application -

for Re' actor Fuel. GESTAR-II" (latest approved revision).

2.

USAR Chapter 15. Appendix 158..

3.

USAR. Chapter 15. Appendix 15F.

4.

USAR. Chapter 15. Appendix 15E.

5.

NED0-30130-P-A. "Stea(y State Nuclear Methods." April 1985.

(cedinued) l PERRY - UNIT 1 8 3.2-4 Revision No. O

PY CIW -2420L D

Page lo of38

*2 BASES APPLICABLE The MCPR operating limits derived from the transient SAFETY ANALYSES analysis are dependent on the operating core flow and (continued) state (MCPRe and MCPR,. resactively) to ensure adher fuel design limits during the worst transient thaf occurs with moderate frequency (Refs. 4. 5. and 6).

state tiermal hydraulic methods using t BWR simulator code (Ref. 7). MCPR, curves are provided Hanual and Non Loop Hanual operation. based o The result of a operation is the runout of only one 1 single failure recirculation loops are under i because both Manual operational modes allow s control.

1taneous runout ofNon Loop loops because a single controller regulates core flow.both Power dependent MCPR limits (MCPR transient code (Ref. 8).three dimensional BWR sim are determined by the o

Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, high and low flow

' operati limits are provided for operating RTP a mentioned bypass power level.

the previously-g The MCPR satisfies Criterion 2 of the NRC Policy Stataman LCO The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis.

The MCPR operatin larger of the MCPRr and MCPR,g limits are determined by the limits.

i APPLICABILITY The MCPR operating limits are primarily derived from trannient a t are assumed to occur at high power 2%$

leve s. Beluw

. the reactor is operating a'; a slow recirculation speed and the moderator va atio is small. Survei ance of thermal limits below unnecessary due to the large inherent margin &

R RTP is that the MCPR SL is not exceeded even if a limiti ensures.-

transient occurs.

(cur 1nued)

~

PERRY - UNIT 1 B 3.2-7 Revision No. O

242a B3 BASES r

APPLICABILITY Studies of the variation of limiting transient behavior ha (continued) been performed over the range of power and flow condi These studies encompass the range of key actual plant The results of these studies demonstrate tha 2185 expected between performance and the MCPR r

_~

~

\\ that marQins increase as Dower is regd

_ ts, and

%7% to E2k trend is expected to continue to tno' tA154; ed tn IP. This when entry into MODE 2 occurs.

When in~ t1ULP*l. the intermediate range monitor (IRM) provides rapid scram thereby effectively removing any MCPR comp H00E 2.

Therefore at THERMAL POWER levels <oncern in limits and t11s LCO is not required. reactor is o)eratin

- MCPR TP. the g

ACTIONS Al If any MCPR exceeds the required limit, an assumption analyses may not be met.regarding an initial condition of the to restore the MCPR(s) to within the required limit that the plant will be operating within analyzed conditions.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Com)letion Time is sufficient to restore the HCPR(s) to wit 11n its limit and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.

IL1 If the MCPR cannot be restored to within the required limit within the associated Completion Time. the plant must be

=

73,gg brought to a MODE or other specified condition in which the LCO does not apply. iTo achieve this status. THERMAL POWER must ce reauced to Completion Time is reasonable, based onRTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The experience to reduce THERMAL POWER to <

orderly man,ner and without challenging p in an tystems.

(continued) 23 8 PERRY - UNIT 1 8 3.2-8 Revision No. O

PY CEl/NRR-2420L Page 12 of 38 MCPR B 3.2.2 BASES (continued)

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS

~

2.3, o

MCPRs are required to be initiall 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 'RIERNAL POWER is =

iculated within z.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter. They are c RTP and then every within the assumptions of the safety analy to the specified i

Frequency is based on both engineering judgment andThe recognition of the slowness of changes in power distrib during normal ion.

THERMAL POWER The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after la e inhe RTP is achieved is acceptable given the lev ls.

n to operating limits at low power

,gg i

REFERENCES 1.

Nucleate Boiling or Dryout." June 1979.NUR 2.

NEDE-24011-P-A. " General Electric Standard Applica

)

for Reactor Fuel, GESTAR-II" (latest approved revision).

3.

Power Plant Unit 1. Reload 3 Cycle 4. Supple 4.

USAR. Chapter 15. Appendix 158.

5.

USAR Chapter 15. Appendix 15C.

6.

l USAR Chapter 15. Appendix 150.

7.

NEDE-30130-P-A. " Steady State Nuclear Methods." Ap 1985.

8.

Transient Modellfor Boiling Water Reacto 1978.

e i

\\

PERRY - UNIT 1 i

B 3.2-9 Revision No. 1 l

-C 1/N -2420L M

Page 13 of 38 BASES APPLICABLE operating limit specified in the COLR. The analysis also SAFETY ANALYSES the operating limit to account for A00 (continued) for densification power spiking.

The LHGR satisfies Criterion 2 of the NRC Policy S LC0 The LHGR is a basic assumption in the fuel design The fuel has been designed to operate at rated core with sufficient design ma in to the LHGR calculated to cause a 1% fuel cladding

~

astic strain.

limit to acconplish this The operating jective is specified in the COLR.

APPLICABILITY The LHGR limits are derived from fuel design analysis is limiting at power level conditions. At THERMAL POWER levels <

substantial ma RTP. the reactor is operating with required.

he LHGR limits and this LCO is not g,g g ACTIONS M

If any LHGR exceeds the required limit, an assungtion regarding an initial. condition of the fuel design analysis is not met.

Therefore, prompt action is taken to restore will be ope) rating within analyzed condi

-design limits of the fuel rods. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Coupletion Ti is sufficient to restore the LHGR(s) to within it me is acceptable based on the low probability of a transient o Design Basis Accident occurring simultaneously with the out of specification.

L1 If the LHGR cannot be restored to within its required limit within the associated Completion Time, the plant must be brought to a H0DE or other specified condition in which the LCO does not apply. A achieve this status must be reduced to < g RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months THERMAL PO The allowed h iinued_)

~ --

PERRY - UNIT 1 B 3.2-11 Revision No. O

9

.C IM -2420L Page 14 of 38 g3 BASES

,7 ACTIONS ILI. (continued) b.6%

Completion Time is reasonable, based on ating.

experience orderly man. to reduce THERMAL POWER to <RTP in an 1

ner and without challenging p systems.

SURVEILLANCE SR

3. 2. 3.1_

REQUIREMENTS 23,8 T, The LHGRs are recuired to be init 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after ThFJtHAL POWER isy alculated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter. They are TP and then every,

ed to the specified within the assumptions of the safety analysis\\

Frequency is based on both engineering judgment andThe 24 h under normal corecognition of the slowness of changes in p THERMAL POWER =

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after ons.

la e inherent is achieved. is acceptable given the lev ls.

o ope limits at lower power 3,gg REFERENCES 1.

m NUREG-0800, " Standard Review Plan," Section 4.2 II.A.2(g). Revision 2. July 1981.

i s

PERRY - UNIT 1 B 3.2-12 Revision No. O

$Y RPS Instrumentation E

2 COL nse is or38 8 3.3.1.1 BASES APPLICABLE 2.a.

Averaoe Power Rance Monitor Neutron Flux-Hich.

SAFETY ANALYSES. Setdown_

(continued)

LCO. and APPLICABILIlY With the IRMs at Range 9 or 10. it is possible that the Average Power Range Monitor Neutron Flux-High. Setdown Function will provide the primary trip signal for a corewi increase in power.

l No specific safety analyses take direct credit for the Average Power Range Monitor Neutron Flux-High. Setdown Function. However, this Function indirectl ensures that. A before the reactor mode switch is pla position ihn run when oper reactor power does not ating at low reactor pressure

/ (SL 2.1.1.

Therefore, it indirectly prevents fuel damage du si ficant reactivity increases with THERMAL POWER 23.6'h The APRM System is divided into two groups of channels wit four APRM channels inputting to each trip system.

to be bypassed. system is designed to allow one channel in ea The cause the associated trip system to trip.Any one APRM Average Power Range Monitor Neutron Flux-HighSix channels of with three channels in each trip system are req. Setdown, OPERABLE to ensure that no single failure will preclude auired t 1

scram from this Function on a valid signal.

l e of the entire core,In addition to provide adequate cove at least 14 LPRM in)uts are for each APRM channel, with at least two.PRM inputs rom each of the four axial levels at which the LPRMs are located.

The Allowable Value is based on )reventing increases in power when THERMAL 30WER is <

ificant 1

)

The Average Power Range Monitor Neutron Flux-Hi 23 88h Function must be OPERABLE during MODE 2 when con!1h.

"rol rods may be withdrawn since the potential for criticality exists.

In MODE 1. the Average Power Rang Monitor Neutron transients and the Rdl and RPC protect agains arovides pro ection against reactivity withdrawal error wents.

(Und.inued)

PERRY - UNIT 1 B 3.3-7 Revision No. 0

RPS Instrumentation C

2420L Page 16 of 38 B 3.3.1.1 BASES r

APPLICABLE 4

SAFETY ANALYSES.

Reactor vessel Water Level-Low. Level 3 (continued)

LCO. and energy exists in the RCS resulting in the APPLICABILITY transients and accidents.

Vessel Water Level-Low Low. Level 2 an Level 1 provide sufficient protection for level transients in all other MODES.

5.

Reactor Vessel Water Level-Hich. Level 8 High RPV water level indicates a potential problem with t reactivity associated with the introductio amount of relatively cold feedwater.

Therefore. a scram is initiated at Level 8 to ensure that MCPR is ma the MCPR SL.

The Reactor Vessel Water Level-High. Level 8 Function is one of the many Functions assumed to be OP and capable of providing a reactor scram during transients analyzed in Reference 3.

It is directly assumed in the analysis of feedwater controller failure, maximum demand (Ref. 4).

Reactor Vessel Water Level-High. Level 8 signals are initiated from four level transmitters that sense the water (reference leg) and the pressure due t water level'(variable leg) in the vessel.

The Reactor Vessel Water Level-Hi

. Level 8 Allowable Value is tie assumed transient.smcified to ensure th t the MCPR SL i Level 8 FunctionFour channels of the Reactor Vessel Water arranged in a one, with two channels in each tri system out-of-two required to be OPERABLE when TH

. are availab re POWER is rmir to ensure that no single instrument failure will scram f this Function on a valid signal.

th THERMAL a

POWER <

TP. thi unction is not required since MCPR is not a c e

bel TP.

234fe (cedinued)

PERRY - UNIT 1 B 3.3-13 Revision No. O

Atchment 4 RPS Instrumentation PY<EIMRR-2420L Page 17 of 38 B 3.3.1.1 BASES

=-

APPLICABLE 9.

Turbine stoo Valve closure SAFETY ANALYSES.

(continued)

LCO and Turbine Stop Valve Closure si APPLICABILITY switches at each stop valve. gnals are initiated by limit are associated with each sto) valve.Two independent limit switche to R)S trip system A: the otherOne o switches provides in RPS trip system B.

input-from four Turbine Stoi) Valve Closure cha an consisting of one limit swech.

Stop Valve Closure Function is sucThe logic for the Turbine must be closed to produce a scram.h that three or more TSVs to be high enough to detect inninent T reducing the severity of the subsequent pressure transie Eight channels of Turbine Sto channels in each trip system.p Valve Closure, with four scram from this Function if any three 15V This Function is required, consistent wi assumptions, whenever THERMAL POWER is

is Function is not required when THERHAL is since the Reactor vessel Steam Dome Pressure s <

RTP I

i Average Power Range Monitor Fixed Neutron Flux-HighHigh Functions are adequate to maintain the necessary sa margins. Enabling of this Function is normall automatica11y b 1shed 3gjo sta e )ressure:y 3ressure transmitters sensi turbi first t

t1erefore, to consider this

_E. the turbine bypass valves must remain shut at ion i

RTP.

result of the subcooling changes that affect th as turbine first stage pressure / reactor power rel i

For RTP operation with feedwater tem)erature a Set)oint of s 212 psig and an A110wa21e.Value l

tur)ine first statie pressure are provided for the bypass 18 ps function.

The Se: points and Allowable Values are reduced to the ng values based on the difference in temperatute.

~

from

+26.5 F 1

AT(*F)

Setooint Allowable Value 0 < AT s 50 s 190 g

s 196 g

50 < AT s 100 s 168 g

s 174 g

100 < AT s 170 s 146 g

s 152 g

(continued)

PERRY - UNIT 1 B 3.3-17 Revision No.1

i fYCUp RPS Instrumentation 2420L Page 18 of38 B 3.3.1.1 BASES r

APPLICABLE

10. Turbine Control Valve Fast Closure. Trio Oil ~

SAFETY ANALYSES.Pressure-tow LCO. and

~

APPLICABILITY Fast closure of the TCVs results in the loss of (continued) that produces reactor pressure. neutron flux, and heat flu transients that must be limited.

transients that would result from the clo valves.

Pressure The Turbine Control Valve Fast Closure. Trip 011 Low Function is the primary scram signal for the i

generator load rejection event analyzed in Reference 4.

l required to be absorbed and, along with th For EOC-RPT System, ensures that the MCPR SL is not exceed e

Turbine Control Valve Fast Closure. Trip 011 Pressure-Low signals are initiated by the EHC fluid pressure at each control valve.

each control valve the signal from each transmit assigned to a separ, ate RPS logic ei This Functio gg%

must be enabled at THERMAL POWER =

l KtP. This is normally accomplished automaticall sensing turbine first stage pressure:

3ressure switches this Function OPFRARLE. the turbine bypass valves mustt1e gg remain shut at i T RTP. The basis for the setpoint of l

this automatic byp@ ass is identical to that d Turbine Stop Valve Closure Function.

The Turbine Control Valve Fast Closure. Trip 011 Pressure-Cow Allowable Value is s detect imminent TCV fast closure. elected high enough to Four channels of Turbine Control Valve Fast Closure. Trip 011 Pressure-Low Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure wil Jg7, areclude a scram from this Function on a valid signal.

unction is required, consistent with the This bsis assumptions, whenever THERMAL POWER is =

7.

Function is not required when THERMAL POWER is <d bis

$TP since the Reactor Vessel Steam Dome Pressure-Hig.1 a 15'1 Average Pcwer Range Monitor Fixed Neutrois Flux-High

~

Functions are adequate to maintain the necessary safety I

margins.

(continued)

PERRY - UNIT 1 B 3.3-18 Revision No. O

Attchment 4 py cri/sna-2420t a ion Page 19 of 38 B 3.3.1.1 L

BASES SURVEILLANCE SR 3.3.1.1.1_

(continued)

REQUIREMENTS Agreement criteria are determined by the plant staff based on a combination o the channel instrument uncertaint including indication and readability.

If a channel is outside the criteria it may be an indication that the instrument has drifted outside its limit.

The Frequency is based upon operating exmrience that demonstrates channel failure is rare.

channels during normal operational use of i

associated with the channels required by the LCO. plays SR 3.3.1.1.2 To ensure that the core average power APRMs are accurately indicating the true the APRMs are calibrated to tfie reactor power calculated from a heat balance.

The Frequency of once per 7 days is based on minor changes in LPRM sensitivity which could affect the APRM reading between perfonnances, SR 3.3.1.1.8.

A restriction to satisfying this SR when <

>ad RTP 3rovided that requires the SR to be met on t =J

)ecause it is difficult to accurately maintain 1 KIP 23.6 indication of THERNAL POWER consistent with a heat balanEe wnen RTP.

At low power levels, a h1 of accuracy is essary because of the large i ph deg 23 87 margin to thermal limits (MCPR and APLHGR).

^

At a TP. ~

l the Surveillance is required to have been satisfa performed within the last 7 days in r:cordance with ly SR 3.0.2.

A Note is ided which allows an increase in M RMAL POWER abov __if the 7 day Fwy is not met

?3.8%

per sR 3.u.z.

in Tn vent, the SR stat formed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reaching or exceeejng Twelve hours is based on operating exper; consideration of providing a reasonable tre in which to n

complete the SR.

23"8 SR 3.3.1.1.3 i

l The Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function uses the recirculation loop drive flows to vary the trip setpoint. This SR ensures that the total loop drive flow signals from the flow unit used to ksdinued)

PERRY - UNIT 1 B 3.3-25 Revision No. O

RPS Instrumentation

-2420L Pqe 20 of 38 B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.14 REQUIREMENTS

~

(continued)

The Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function uses an electronic filter circuit to generate a signal proportional'to the core THERMAL POWER from the APRM neutron flux signal.

dynamics that produce the relationship b This flux and the core THERMAL POWER.

The filter time constant the channel is accurately reflecting the de The Frequency of 18 months is based on engineering jud and reliability of the components.

SR 3.3.1.1.15 The LOGIC SYSTEM FUNCTIONAL TEST demonstra OPERABILITY of the required trip logic either actual or simulated automatic tri the receist of functional testin ignals. T1e Rod OPERABILITY."g of control rods, in L 3.1.3 " Control and SDV vent and drain valves,.in LCO 3.1.8 verlaps this Surveillance (to* Scram Discharge V Valves." o.

testing of the assumed safety function. provide complete Surveillance under the conditions tThe 18 month Fre outage and;the potential for an un) hat apply during a plant lanned transient if Surveillance were performed with t1e reactor at power. the Operating experience has shown that these commnents usually pass the S Frequency.urveillance when performed at the 13 month SR 3.3.1.1.16 8%

This SR ensures thati scrams initiated f Valve Closure and Turbine Control Valvast Closure.e Stop Turbin 011 Pressure-Low Functions will Trip bypassed when THERMAL POWER is =

inadvertently RTP. This involves calibration of the bypass channel.

the actual sthe instrument setpoint methodology are iiiceipo int. Because main turbine b.voass flow can affect this s int nonconservatively ine first stage pressur(THERKAL POWER is derived from t e). the mai rbine bypass valves must remain closed at THERMAL POWER =

to ensure that the calibration remains valid.

RTP ~

38 %

(cerd.inued)

PERRY - UNIT 1 B 3.3-30 Revision No. 1 i

RPS Instrumentation

-24201-Page 21 of 38 B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.1.16 (continued) 3hfo If any bypass channel setpoi Functions are bypassed at a:

s nonconservative (i.e turbine bypass valve (s) or o er rea either due to open., the RTP Turbine main Closure.Stop Valve Closure and Turbine Control Va Trip 011 Pressure-Low Functions are considered inoperable Alternatively. the by) ass ch in the cons.ervative condition (non)ypass)annel can be placed considered OPERABLE.nonbypass condition, this SR is If placed in the The Frequency of 18 months is based on engineeri and reliability of the conponents.

SR 3.3.1.1.18 This SR ensures that the individual channel response are less than or e accident analysis. qual to the maximum values assumed in the The RPS RESPONSE TIME acceptance criteria are included in Reference 10.

As noted, neutron detectors are excluded from RPS RES TIME testing because the principles of detector o virtually ensure an instantaneous res)onse time. peration addition, for Functions 3 In Functions, quired to be res. 4 and 5, t1e associated s are not re ponse time tested. For these t

1 components is required. response time testing for the remaining ch Reference 11.

This allowance is suppor:ed by STAGGERED TEST BASIS.RPS RESPONSE Frequency to be determined based on 4 channe system. in lieu of the 8 channels specified in Table 3.3.1.1-1 for the HSIV-Closure Function.

channels required to produce an RPS scram s Therefore, staggered testing results in res)gnal.

verification of these devices every 18 montis. This onse time cycle and is based upon plant operating ex shows that random failures of instrumentation compo,nents causing serious time degradation, but not channel failure.

are infrequent.

(continued)

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

Control Rod Block Instrumentation 1/h 2420L Page 22 of 38 B 3.3.2.1 t

BASES e

BACKGROUND during startup are such that only specified (continued) sequences and relative positions are allowed over 19.0 %

omrating range from all control rods inserted t T1e sequences effectively limit the potential a TP.

rate of reactivity increase during a CRDA.

The RPC. in and conjunction with the RCIS. will.nitiate control rod withdrawal and insert blocks when the actual sequence deviates b ond allowances from the specified sequence.

rod block 1 The ic circuitry is the same as that described above. The C also uses the turbine first stage pressure to determine when reactor power is above the power at which the RPC is automatically bypassed (Ref.1).

i With the reactor mode switch in the shutdown position. a to ansure that the shutdown condition is maintain function prevents criticality resulting from inadvertent This when the reactor mode switch is recuired to be shutdown channels, position. The reactor moce switch has two block circuit.with each providing inputs into a separate rod control rod block to all control rods.A rod block in either circuit l

APPLICABLE 1.a.

Rod Withdrawal Limiter SAFETY ANALYSES.

LCO. and 1he RWL is. designed to prevent violation of the MCPR SL APPLICABILITY and the cladding 15 plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event. The analytical methods and assumptions used in l

evaluating the RWE event are summarized in Reference 2.

A statistical analysis of RWE events was performed to determine the MCPR response as a function of withdrawal distance and initial operating conditions.

From these responses, the fuel. thermal performance was determined as a function of RWL allowable control rod withdrawal distance and power level.

I The RWL satisfies Criterion 3 of the NRC Policy Statement.

Two channels of the RWL are available and are required to be OPERABLE to ensure that no single instrument failure can preclude a rod block from this Function.

(continued)

PERRY UNIT 1 B 3.3-43 Revision No. O

{

l

Control Rod Block Instrumentatio t

24201, Page 23 of 38 B 3.3.2.1 BASES APPLICABLE 1 a.

SAFETY ANALYSES.

Rod Withdrawal Limiter (continued)

LCO and Nominal trip set points are specified in the setpoint APPLICABILITY calculations.

that the setpoints do not exceed the Allowable between successive CHANNEL CALIBRATIONS.

trip setpoint less conservative than the nominal tripOpera setpoint, but within its Allowable Value, is acceptable.

Trip setpoints are those predetermined values of outpu which an action should take place.

The setpoi compared to the actual process parameter (e.g.nts are power) and when the measured out)ut value of the process

, reactor parameter exceeds the setpoint. t1e associated device (e trip unit) changes state.

The analytic limits are derived from the limiting values of the process parameters obtain from the safety analysis.

from the analytic limits, corrected for caliThe Allowable Val process, and some of the instrument errors. bration.

i The trip setpoints are then determined accounting for the remain 1

instrtmient errors in this manner prov(e.g., drift).

ide adequate protection becauseThe trip s instrumenta tolerances. tion uncertainties, process effects, calibratiori (for channels that must function in harsh envi!

defined by 10 CFR 50.49) are accounted for.

33'%

The RWL is assumed t ate the consequences of an RWE event when operat RTP.

consequences of an Below this power level, the therefore the RWL is not required to be OPERABLE 1.b.

Rod Pattern Controller The RPC enforces the banked position withdrawal sequence (BPWS) to ensure that the initial conditions of the CR analysis are not violated.

The analytical methods and References 4 and 5. assumptions used in evaluating the CRDA a moved in groups, with all control rods assigned to aTh specific g positions.

required to be within specified banked compliance wi irements that the control rod se Rod Pattern."

BPWS are specified in LCO 3.1.6.quence is in

" Control (centinued)

PERRY UNIT 1 B 3.3-44 Revision No. O

l' Control Rod Block Instrumentation I

t PY CEIMRR.2420L Page 24 of 38 BASES APPLICABLE 1.b.

SAFETY ANALYSES.

Rod Pattern Controller (continued)

LCO. and The Rod Pattern Controller Function satisfies Crit APPLICABILITY the NRC Policy Statement.

o)erator control of control rod sSince the RPC is a back0p to clannel would be required to be OP uences. only a single LE to satisfy Criterion 3 (Ref. 5). However, the RPC is designed as a dual channel system and will not function without two OPERABLE channels.

Required Actions of LCO 3.1.3. " Control Rod OPERABILITY." and LCO 3.1.6 may necessitate bypassin l

individual control rods in the Rod Action Control System (RACS) to allow continued operation with inoperable control rods or to allow correction of a control rod pattern not in com)liance with the BPWS.

be -)ypassed as required by the conditions. and the R not considered inoperable provided SR 3.3.2.1.9 is met.

N.O Compliance with the BPWS. and therefo s re ILITY of the RTP. quired in MODES 1 and 2 ERMAL POWER When THERHAL POWER is >

le control rod configuration RTP. there is no results in a control worth that could exceed the 280 cal /gm fuel damage limit during a CRDA.

In MODES 3 and 4. all control rods are required to be inserted in the core.

In MODE 5 a single control rod can be withdrawn from a cor. since only e cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable. since the reactor will be suberitical.

2.

Reactor Mode Switch-Shutdown Position During MODES 3 and 4. and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical:

reactivity insertion. events are analyzed,therefore, no positive lhe Reactor Mode Switch-Shutdown Position control rod withdrawal block l

ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.

i The Reactor Mode Switch-Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement.

h d.inued)

PERRY UNIT 1 B 3.3-45 Revision No. 0 1

l E0CAPT Instrumentatiok-

^"$$$a#

B 3.3.4.1 y

2420L e or3s

=-

BACKGROUND 1

(continued) per recirculation pump. One trip system trips one of the two E0C-RPT breakers for each recirculation pum) and the second trip system trips the other EOC-RPT breacer for each recirculation pump.

APPLICABLE The TSV Closure and the TCV Fast Closure. Trip Oil SAFETY ANALYSES.

LCO. and Pressure-Low Functions are designed to trip the recirculation pumps from fast speed operation in the event APPLICABILIlY of a turbine trip or generator load rejection to mitigate the neutron flux. heat flux, and 3ressure transients. and to increase the margin to the MCPR S..

and assumptions used in evaluating the turbine trip andTh that assume E0C-RPT. are sunnarized in Refere and 3.

To mitigate pressurization transient effects, the EOC-RPT must trip the recirculation pumps from fast speed o)eration after initiation of initial closure movement of eitler the TSVs or the TCVs.

The combined effects of this trip and a scram reduce fuel bundle power more rapidly than does a scram alone. resulting in an increased margin to the MCPR SL. lhe E0C-RPT function is aut cally dis when turbine first stage pressure is <

RTP.

i EOC-RPT inst'rumentation satisfies Criterion 3 oNRC Policy Statement.

The OPERAB LITY of the EOC-RPT is dependent on the OPERABILITY of the individual instrumentation channel Functions.

Each Function must have a required number of

~

OPERABLE channels in each trip system, with their set within the specified Allowable Value of SR 3.3.4.1.2. points The actual setpoint is calibrated consistent with ap)licable setpoint methodology assumptions.

includes the associated EOC-RPT breakers. Channel O Each channel (including the associated EOC-RPT breakers) must also respond within its assumed response time.

Allowable Values are specified for each EOC-RPT Function specified in the LCO.

in the setpoint calculations. Nominal tri) setpoints are s)ecified A c1annel is inopera)le if kuit.inued)

PERRY - UNIT 1 B 3.3-69 Revision No. 0 1

fy EOC-RPT Instrumentation L

2420L Page 26 of 38 B 3.3.4.1 BASES APPLICABLE SAFETY #MLYSES. Turbine Ston Valve closure (continued)

LCO. and Closure of the TSVs is determined by a limit switch on APPLICABILITY stop valve.

There is one limit switch associated wi stop valve and the signal from each limit switch 1'th each assigned to a separate trip channel.

387, s

Closure is such that two or more TSys must be closed t produce an EOC-l This Function must be enabled at THERMAL POWER RTP.

This is normally accomplished autGuaucaily DT tierefore, to consider this Functionssure t stage 3ressure:

OPERAB.E. the t e by 3gg THERMAL POWER RTP. pass valves must remain shut at Four channels of TSV Closure. withl two woin. a in trip system, are required to be i

OPERABLE to ensure that no single instrument failure will f

i preclude an E0C-RPT from this Function on a valid signal.

The TSV Closure Allowable Value is selected high enough t detect inminent TSV closure.

gg7, This protection is required, consistent with the s M

analysis assumptions, whenever THERMAL POWER is a with any recirculation ptap in fast speed. Belowht. RTP or TP with the recirculation pump in slow speed, the Reatt5r Vessel Steam Dame Pressure-High and the Average Pow Monitor (APRM) Fixed Neutron Flux-High Functions of the i

Reactor Protection System (RPS) are adequate to maintain th necessary safety margins.

The automatic enable setpoint is feedwater temperature l

dependent a's a result of the subcooling changes that affect

- 1 i '-

the turbine first stage pressure / reactor For operation with feedwater temperature =

ationshi) of s 212 psig and an Allowable Value of 5 2 a setpotit I.-

1he Setpoints and Allowable Values are reduc ig turbine 8

+25.5*F M irwing values based on the difference in temperature fr om ATMF)

Setooint-Allowa31e Value O < A" s 50 r190 sig s 1% sig

,. c".

50 < AT s 100 s 168 sig s 174 sig N

100 < AT s 170 s 146 sig s 152 sig (continued)

PERRY - UNIT 1 i

B 3.3-71 Revision No. 1

r i

g E0C RPT Instrumentation Pv.crwaa-2420L Page 270f38 B 3.3.4.1 BASES e

APPLICABLE SAFETY ANALYSES.TCV Fast Closure. Trio 011 Pressure-Low LCO. and results in the loss of a heat sink that produce APPLICABILITY (continued) aressure, neutron flux, and heat flux transients that must

>e limited.

Closure. Tri) Oil Pressure-Low in anticipation of t transients t1at would result from the closure'of these valves.

The E0C-RPT decreases reactor power and aids the during the worst case transient. reactor scram in ensuring Fast closure of the TCVs is determined by measuring the EHC fluid pressure at each control valve.

switch associated with each control valve, and the signalT from each switch is assigned to a separate trip channe The logic for the TCV Fast Closure. Trip 011 Pressure l.

Function is such that two or more TCVs must be closed L

(pressure switch trips) to produce an EOC-RP 1 2his Function must be enabled at THERMAL POWER k is normally accomplished automatically by pr91Ehre PORKIP. Th transmitters sensing turbine first stage pressure:

therefore, to consider this Function OPERABLE. th bypass valves must remain shut at bine Four channels of TCV Fast Closure. THERMA 3TP.

Tri with two channels in each tri

.ow, are required to be OPERABLE to ensure that no si e instrument failure.will preclude an E0C-RPT from this ion on a valid signal.

The TCV Fast Closure Trip 011 Pressure-Low Allowab e Value 1

is selected high enough to detect inminen closure.

whenever the INERMAL POWER is = *This protection alysis. M l

inv recirculating pump in fast speed.

elow (TP or with recirculation Dome Pressure p s in slow s)eed, the r Vessel Steam Functions of the hiand the AMlM F1xed Neutton Flux-High safety margins. The turbine first stage pressure /reactora

)

power relationshio for the setpoint of the automatic enable is identical to that described for TSV closure.

(continued)

PERRY - UNIT 1 B 3.3-72 Revision No. O

7a/d u20t.

E0C-RPT Instrumentation Page 23 of 38 B 3.3.4.1 BASES ACTIONS U (continued) where placin an E0C-RPT).g the inoperable channel in trip would result in or if the inoperable channel is the result of an inoperable breaker. Condition C must be entered and its Required Actions taken.

M actions are taken if multiple. inoperable. u maintaining E0C-RPT trip cchannels within the same unction not r

bility. A Function is considered to be maintaini sufficient channels are OP EOC-RPT trip capability when t

EOC-RPT System will generate a trip signal from th Function on a valid signal and both recirculation pumps ca be tripped from fast speed operation.

OPERABLE or in tri). channels of the Function in thei breakers to be OPERABLE or in trip.and the associated EO The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is sufficient' for the operator to take corrective action, and takes into account the instrumentation during this period. likelihood of an It is also consistent with the 2. hour Completion Time provided purpose is to preclude a MCPR violation.

C.1 and C.2 SF%

With any Required Action and associated met. THERMAL POWER must be reduced to <letion Time not RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Alternately the associated rec rculation pump fast speed breaker may, be removed from service since this I

performs the intended function of the instrumentation.

allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable, b The operating experience to reduce THERMAL POWER to <

on fran full power conditions in an orderly manner and RTP challenging plant systems.

thout (continued) 38%

PERRY - UNIT 1 8 3.3-74 Revision No. O

1 E0C-RPT Instrumentation

-C

-2420L Page 29 of 38 BASES SURVEILLANCE SR 3.3.4.1.3 REQUIREMENTS

~

(continued)

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

included as a part of this test overlapping the L SYSTEM FUNCTIONAL TEST. to associated safety function. provide complete testing of the Therefore. if a breaker is would also be inoperable. incapable of operating, the as The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unalanned transient if Surveillance were performed with t1e reactor at power. the 0)erating experience has shown these components usually p t1e Surveil Frequency. lance test when performed at the 18 month SR 3.3.4.1.4 3ffg This SR ensures that an E0C-RPT initiated from t Closure and TCV Fast Closure. Trip 011 Pressure he TSV Functions Low POWER is not be inadvertently bypassed when THERHAL This involves caTibration of the bypass duumb.

te margins for the instrtment setpoint methodologips are incorporated into the actual setpoint.

8ecause main turbine by3 ass flow can affect this setpoint l

g nonconservatively (THERiAL POWER is derived from first stage sure). t>e maji ine bypass valves must remain closed at THLMMAL P:WER R RTP to ensure that the calibration remains valid.

If ass channel's setpoint is servative (i.e..

Functions are b ssed at 387p RTP either due to open main turbine ss valves or reasons). the:affected TSV Closure a TCV Fast Closure. Trip 011 Pressure-Low Functions are considered inoperable.: Alternatively, the by) ass ch in the conservative condition (non)ypass)annel can be placed If placed in the nonbypass condition, this SR is met and the channel considered OPERABLE.

The Frequency of 18 months has shown that channel bypass l

failures between successive tests are rare.

t (continued)

PERRY - UNIT 1 B 3.3-76 Revision No. O L

Primary Containment and Drywell Isolation Instru fy

-2420L Page 30of 38 BASES APPLICABLE 1.b.

Main Steam Line Pressure-Low SAFETY ANALYSES.

LCO. and Low MSL pressure ind the turbine pressure'icates that there may be a problem with APPLICABILITY reactor vessel water level condition and the R (continued) down more than 100*F/ hour if the pressure loss is allow continue.

The Main Steam Line Pressure-Low Function directly assumed iri the analysis of the pressure regulator failure (Ref. 2).

For this event, the closure of the HSIV ensures that the RPV temperature change limit (100*F/ hour) not reached.

In addition, this Function supports actions is to ensure that Safety Limit 2.1.1.1 is not exceeded.

Function closes the MSIVs prior to pressure decreasing b (This 785 psig, which results in reducing reactor power to <

ram due to MSIV closure, thus TP.)

g The HSL low pressure signals are initiat transmitters that are connected to the MSL header rom four separated from each othetransmitters are arranged suc The j

detect low MSL pressure.r, each transmitter is able to Four channels of Main Steam Line that no single instrument failure can prec function.

The Allowable Value was selected to be high enough to prevent excessive RPV depressurization.

j The Main Steam Line Pressure-Low Function is only require to be OPERABLE in MODE 1 since this is when the a transient can occur (Ref. 2).

This Function isolates the Group 6 valves.

1.c.

Main Steam Line Flow-Hiah Main Steam Line Flow-High is provided to detect a break of the MSL and to initiate closure of the MSIVs.

were allowed to continue flowing out of the break. theIf the steam reactor would depressurize and the core could uncover 1

the RPV water level decreases too far, fuel damage cou.

If ld to prevent or minimize core damage.Therefore. the iso occur.

Flow-High Function is directly assumed in the analThe Main Ste the main steam line break (MSLB) accident (Ref.1).ysis of isolation action, along with the scram function of the RPS.

The (continued)

PERRY - UNIT 1 B 3.3-142 Revision No. O

n r y a-2420t.

Primary Containmentjpg:ng BASES '

e-BACKGROUND e.

The containment leakage rates are in compliance with

-(continued) the requirements of Specification 3.'.1.1 and Specification 3.6.1.3:

f.

The suppression pool is OPERABLE: and g.

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

This Soecification ensures that the performance of the primary containment, in the event of a DBA, meets the assumptions used in the safety analyses of References 1 and 2.

SR 3.6.1.1.1 leakage rate requirements are in conformance with 10 CFR 50, Appendix J, Option B (Ref. 3),

as modified by approved exemptions.

APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures and t eratures of the limiting DBA without exceeding the desi n leakage rate.

The DBA that postulates the maximum release of radioactive

'^

u.)

material within primary containment is a LOCA.

In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.

l Analytical methods and assumptions involving the primary containment are presented in References 1 and 2.

The safety analyses assume a mechanistic fission product release following a DBA, based on NUREG 1465, which forms the basis for determination of offsite doses. The fission product i

i release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded.

[T maximum a lowable leaka ratefortheYMmary 4 co ainment (L. is 0.20% by ight of the inment and dr 11 air per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months at th design basis maximum peak ntainment ressure (P,) o 7.80.psig (Re 4).

Primary containment satisfies Criterion 3 of the NRC Policy Statement.

(continued)

PERRY - UNIT' 1 B 3.6-2 Revision No. 2 ry-caman.2420t.

Primary Containment--Operating Page 32 cf38 B 3.6.1.1 BASES.

~

l SURVEILLANCE REQUIREMENTS SR 3.6.1.1 (continued)

~

Sections III.A.1(d). III.A.5(b)(2). III.B.3 and c.

outboard MSIVs (including the volume up t

' outboard MSIV before seat drain line valves) ar required to be vented and drained for Type A testing and the main steam line isolation valve leak rate exempted from inclusion in the overall integrated primary containment leak rate and the combined local leak' rate (Reference 8),

d.

Section III.D.1(a) - The third Type A test for each when the plant is shutdown for the 10-yea inservice inspection (Reference 8).

L Section III.D.3 - Type C local leak rate testing may e.

be performed at other convenient intervals in addition to shutdown during refueling, but at intervals no greater than 2 years (Reference 8).

As left leakage prior to the first startup after performing i

a required leakage test is required to be < 0.6 L for combined Type B and Type C leakage, and s 0.75 L,for overall Type A leakage. At all other times betwe,en required leakage rate. tests, r.he acceptance criteria is based on an overall Type A leakage limit of < 1.0 L.

At < 1 offsite dose consequences are bounded by the assu.0 L, the the safety analysis.

mptions of REFERENCES 1,.

USAR. Section 6.2.

^

2.

USAR. Section 15.6.S.

l 3.

DELETED 10 CFR 50. Appendix J. Option B.

n w w wW 4.

PY CCIME-1510w A-- 2' laa2.)

w w

5.

Letter fr Edelman),om NRC (B.J. Yournblood) to CEi (M.R.

" Performance of nhe Preoperational Containment Integrated Leak Rate Test - Peri l. Nuclear Power Plant. Unit 1." dated June 10, 1933.

kent.inued) l PERRY UNIT 1 B 3.6-5 Revision No. 1

y ry.comna-2420t.

Suppression Pool Water Level 1

r.s. n ons B 3.6.2.2 B 3.6. CONTAINHENT SYSTEMS B 3.6.2.2 Suppression Pool Water Level BASES BACKGROUND 1he suppression pool is a concentric open container of with a stainless steel liner, which is located at the bottom of the primary containment.

The suppression pool is during a reactor blowdown from safety /

discharges or from a loss of coolant accident (LOCA).

Core Isolation Coolingsuppression pool must also conden The provides the main emerge (ncy w)ater supply sour reactor vessel.

The high water level limit and the low water level limit (indicated level of 18 ft 6 inches and 17 ft 9.5 incn respectively). are nominal values assiming a zero differential pressure across the drywell wall.

include the water volume of the containment portion of theT pool, the horizontal vents. and the weir annulus (includi encroachments).

containment LOCA response analyses wasTh g

shMew which with tie maximum negative drywell-to-containmen L or is ft 6 inches differential pressure (-0.5 psid) and primary containment to secondary containment differential pressure (1.0 psid).

This volume was used to maximize the negative effect of the suppression pool wate temperature response r volume on the drywell pressur and The suppression poolivolume used in pg 2ana +are containment LOCA response analyses was(fy.7Mft. which 9.5 8

contairumnt nmL and corresponds to 'an indi level of 17 flEB3nches with the maxis to-containmentTIfferential pressure (sa positive drywell-2.0 psid was used to maximize the containment pressure a)n. This voltme d

gpature response results of the long ters analyses.

The

' ger h4was sey' con BRXM.bj5(([jki$fer$5t1N[daNy,ses for

+

rawr.aun w Ene suppression pool.

(ccidinued)

PERRY - UNIT 1 B 3.6-75 Revision No. 1

r l

PY<El/NRR-2420L Suppression Pool Water Level Page 34 of 38 8 3.6.2.2 y

BASES r

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mm -

- er w w. 4 A -- AA Tcontinued]

.i PERRY - UNIT 1 B 3.6-75a Re'/ision No.1

py.CE1/NRR.2420L r gessoras SPMU System B 3.6.2.4 BASES N

/-

APPLICABLE volume of [d23 t' in the upper containment pool an SAFETY ANALYSES suppressio (continued)

)

The SPMU System satisfies Criterion 3 of the NRC P Statement LCO maintain peak suppression pool water design limits (Ref. 1).

are met, two SPMU subsystems must be OPERABLE.T in the event of an accident, at least one subsystem isTher OPERABLE. assuming the worst case single active failur The SPMU System is OPERABLE when the upper co water temperature is s 110'F. the piping is intact, and the system valves are OPERABLE water levels of the upwr c. Additionally, the combined suppression pool must a within limits.ontainment pool and the When the suppression mol level is maintained 2.2 inches greater than required by..C0 3.6.2.2. " Suppression Pool Water Level" allowed upper containment pool water level limit is redu to 22 ft 5 inches.

APPLICABILITY In MODES 1. 2. and 3. a DBA could cause heatup and pressurization of the primary containment.

i reduced due. to the pressure and temper In MODES 4 i

these MODES.' Therefore, maintaining the SPitt System OPERABLE is not required in HDDE 4 ~or 5.

ACTIONS L.1 When the combined water level of the upper containmen and suppression mol11s not within limits it is inadequate to ensure that tie suppression pool heat sink capability matches the safety analysis assumptions.

A sufficient quantity of water is necessary to ensure long term energy sink capabilities of the suparession pool and maintain water coverage over the uppennost. iorizontal vents.

volume has a relatively large impact on heat sinkLoss of water capability.

Therefore, the combined water level of the to within limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . upper contairstent pool and s is sufficient to provide makeup water to either theThe 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months suppression pool or the upper containment pool to restore (cedinued)

PERRY - UNIT 1 B 3.6-85 Revision No. 1

.,y

i N N 2420L

. Main Turbine Bypass System Page 36 0f 38 B 3.7 PLANT SYSTEMS 8 3,7,6 r

B 3.7.6 Main Turbine Bypass System BASES BACKGROUND The Main Turbine Bypass System is designed to control s pressure when reactor steam generation exceeds turbine requirements during unit startup. sudden load reduction. and cooldown.

It allows excess steam flow from the reactor to 2 g,g'fo the condenser without goit ca cit of the system 1s' rough the turbine.

The bypass t

ly stem rated steam (nominal) of the Nuclear Steam N

i S

M~E-inbecapacityof

. Sudden load reductions wi without reactor scram. the steam bypass can be accommodated The Main Turbine Bypass System consists of two valve chests connected to the main st lines between the main steam isolation valves and t turbine s valves.

operated hydraulic cylinders.Each of these valves is sequentially The bypass valves are controll Bypass and Pressure Regulating System, as d USAR. Section 7.7.1.5 (Ref. 1 normally closed, and the press)u. The bypass valves are turbine control valves, directing all steam flow to there regu turbine.

restricts steam flow to the turbineIf the speed control unit control.s the system pressure by open the 3ressure regulator When the ing tie bypass valves.

chest, th ss valves open, the steam flows from the bypass--

connecting piping, to the pressure b'reakdown assemblies..

reduce the re.a series of orifices are used to further condenser. steam pressure before the steam enters the APPLICABLE the design basis feedwater controller failure. m SAFETY ANALYSES demand event, described in the USAR. Section 15.1.2 (Ref. 2).

Opening the bypass valves during the pressurization event mitigates the increase in reactor vessel pressure, which affects the MCPR during the event.

i The Main Turbine Bypass System satisfies Criterion 3 of the NRC Policy Statement.

(continued)

PERRY - UNIT 1 B 3.7-25 Revision No. 1 L

r PY-CELHRR 2420L Page 37 of 38 Main Turbine Bypass System BASES (continued)

B 3.7.6 LCO The Main Turbine Bypass System is required to be limit peak pressure in the main steam lines and mainta cause raoid pressurization, such that th is not ex'ceeded.

An OPERABLE Main Turbine Bypass System require valves to open in response to increasing main steam lin pressure.

applicable analysis (Ref. 2).This response is within the ass APPLICABILITY RTP to ensure that the fuel claddin

=

violated durinand the cladding 1% plastic strain limit ar

,Ee

m i

e demand event. g the feedwater controller failure,e not i

As discussed in the Bases for LCO 3.2.1 maximum

~

" AVERAGE PLANAR LINEAR HEAT GENER LCO 3.2.2. " MINIMUM CRITICAL and margin to these limits exists <

RATIO (MCPR)." sufficient Therefore. these l

requirements are only necessary this power -level.

operatino t or above I

g3,g ACTIONS Q

If the Main Turbine Bypass bypass valves inoperable), S em is inoperable <(one or more l

basis transient analysis may not be met.asstaptions of.the ~ design Under such Main Turbine Bypass System to OPERABLE s l

the Required Action and the low probabilit The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months occurring durin Bypass System. g this period requiring the Main Turbine El 3.7 If the Main Turbine B ns System ca OPERABLE status withi the associat

_ Mletion Time.

ree,ored to 1HERMAL POWER must be reduced to <

theApplicabilitysection,operatio@rrat<

RT' i

scussed in sufficient margin to the required limits, results in Turbine Bypass System is not required to p the Main demand event.

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable.

(continuedi PERRY - UNIT 1 B 3.7-26 Revision No.1 L

r-PY CEIM -2420L ras.3s v3s Control Rod Testing-Operating B 3.10.7 r'

BASES APPLICABLE As described in LC0 3.0.7. compliance with Special SAFETY ANALYSES 0)erations LCOs is optional. and therefore, no criteria of (continued) t1e NRC Policy Statement apply.

provide flexibility to perform certain operations bySpe appropriately modifying requirements of other LCOs.

provided in their respective Bases. discussion of th A

LC0 As described in LCO 3.0.7. compliance with this Special Operations LCO is optional. Control rod testing may be

..C0 3.1.6. and during these tests 3erfonned in comp no ex performed with a seque.1.6 are nece,ssary.ce)tions to requirements of LCO 3

or testing the requirements of LCO 3.1.6 may be suspended, p additional administrative controls are placed on the test to ensure that the asstanptions of the s for the test sequence remain valid pecial safety analysis When deviating from the must be bypassed in the Rod Action Contro Assurance that the test sequence is followed can be

> rov)ided by a second licensed operator or other qualified mem er of the technical staff verifying conformance to the appr)oved test sequenc normally ap)e. These controls are consistent with those lied to operation in the startup range as defined inMR 3.3.2.1.9. when it is necessary to deviate from the prescribed sequence (e.

rod that must be fully inserted)g.. an inoperable control

_f0a Control rod testing

_;0 APPLICABILITY RATED THERMAL POWER, with THERMAL POWER greater tha existing LCOs on pow. is adequately controlled by the block instrumentation. Control rod movement during be)erformed within the constraints of LCO 3.2

" PLA(AR LINEAR HEAT GENERATION RATE (APL (9.07, <

"MINIMlM CRITICAL POWER RATIO (MCPR)." LC s

HEAT GENERATION RATE JLHGRLi and LCO 3.3.2.1,

~

With THERM 8L

~ PUWER less Inan or equal I provisions of this Special RATED THERMAL POWER the z

perform special tests that are not in conformance with the prescribed control rod sequences of LCO 3.1.6.

While in h dinued)

PERRY - UNIT 1 B 3.10-30 Revision No.1

F ATTACHMENT 1 SAFETY ANALYSIS REPORT i

FOR PERRY NUCLEAR POWER PLANT 5% THERMAL POWER UPRATE

ML20211P287

Text

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