ML20077E403

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Proposed Tech Spec Change Request 86 Re Administrative Changes
ML20077E403
Person / Time
Site: Beaver Valley
Issue date: 07/14/1983
From:
DUQUESNE LIGHT CO.
To:
Shared Package
ML20077E391 List:
References
NUDOCS 8307280282
Download: ML20077E403 (19)
v • d • e


Text

._

ATTACHMENT A 1.

Remove pages 3/4 2-14, 3/4 2-15, 3/4 2-16, 3/4 3-2, 3/4 3-3, 3/4 3-4, 3/4 3-7, 3/4 3-8, 3/43-15,3/43-18,3/43-19,3/43-20, 3/4 3-21, 3/4 6-5, 3/4 6-10, 3/4 8-6, 6-2 2.

Insert pages 3/4 3-2, 3/4 3-3, 3/4 3-4, 3/4 3-7, 3/4 3-8, 3/4 3-15, 3/4 3-18, 3/4 3-19, 3/4 3-20, 3/4 3-21, 3/4 6-5, 3/4 6-Sa, 3/4 6-10, 3/4 8-6, 6-2 o

8307280282 830714 0500033]

DR ADOCK p

TABLE 3.3-1 D

REACTOR TRIP SYSTDI INSTRUMENTATION 9

MINIMUM b

TOTAL NO.

CHANNELS CHANNELS APELICABLE h

FUNCTIONAL UNIT OF CllANNELS TO TRIP OFLRABLE MODES ACTION 1.

Manual Reactor Trip 2

1 2

t, 2 and

  • 12

)

2.

Power Range, Neutron Flux 4

2 3

L, 2 2

l 3.

Power Range, Neutron Flux 4

2 3

L, 2 2

liigh Positive Rate 4.

Power Range, Neutron Flux, 4

2 3

t, 2 2

l liigh Negative Rate

(

$g 5.

Intermediate Range, Neutron Flux (2)2 1

2 1, 2 and

  • 3 6.

Source Range, Neutron Flux ( }

gy A.

Startup 2

I 2

2# and

  • 4 1

gw B.

Shutdown 2

0 L

3, 4 and 5 5

o 7.

Ove rtempe ra tu re a '"

j Three: Loop Operation 3

2 2

1, 2 2

Two Loop operation 3

1**

2 1, 2 9

8.

Ove rpower a T Three Loop operation 3

2 2

1, 2 2

Two Loop Operation 3

1**

2 1, 2 9

9.

Pressurizer Pressure-Low 3

2 2

1, 2 7

(Above P-7)

I l

  1. liigh Voltage to detector may be de-energized above P-6.

1 I

TABLE 3.3-1 (Continued) us REACTOR TRIP SYSTEM INSTRUMENTATION n

N MINIMUM yp TOTAL NO.

CHANNELS CHANNELS APPLICABLE P'

FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION Q

g 10.

Pressurizer Pressure--liigh 3

2 2

L, 2 7

U 11.

Pressurizer Water Level--liigh 3

2 2

1, 2 7

y (Above P-7)

I 12.

loss of Flow - Single Loop 3/ loop 2/ loop in 2/ loop in 1

7 l

(Above P-8) any oper-each oper-4 ating loop ating loop El Q

13.

Loss of Flow - Two Loops 3/ loop 2/ loop in 2/ loop t

7 i

(Above P-7 and below P-8) two oper-each oper-

]

y td ating loops ating loop a

z, I

C 14.

Steam Generator Water (4) 3/ loop 2/ loop in 2/ loop in 1, 2 7

l gy Level--Low-Low any oper-each oper-y ating loops ating loop 15.

Steam /Feedwater Flow 2/ loop-level 1/ loop-level 1/ loop-level t, 2 7

Mismatch and Low Steam and coincident and Generator Water Level 2/ loop-flow with 2/ loop-flow mismatch 1/ loop-flow mismatch or mismatch in 2/ loop-level same loop and 1/ loop-flow mismatch

)

16.

Undervoltage-Reactor Coolant 3-l/ bus 2

2 1

7 Pumps (Above P-7) l 17.

Underf requency-Reactor Coolant 3-1/ bus 2

2 1

7 Pumps (Above P-7) l f

i l

TABLC 3.3-1 (Continuai) m REACf0R THIP SYSTD1 ItBURUMEtTfATION MINItui TOfAL PD.

OIANNFIS OIANNEIS APPLICABIE Q

FUNCfIONAL IRIIT OF OIANNEIS TO TRIP OPERABLE F0DFS KTIOT N

18. Turbine Trip ( '

A.

Auto Stcp Oil Pressure 3

2 2

1 7

g B.

Turbine Stcp Valve Closure 4

4 4

1 8

19.

Safety Injection Input (6) l frun ESF 2

1 2

1, 2 1

20.

Reactor Coolant Punp Breaker Position Trip y

8 (Above P-7) 1/ breaker 2

1/ breaker 1

11 yR Per q)er-4-

ating locp Y

$ d' 21.

Reactor Trip Breakers 2

1 2

1, 2*

1 5g 22.

Autanatic Trip Lol c 2

1 2

1, 2*

1 i

c)

IIILow setpoint block permittal by I'10 (2)f4anual bypass pennittel by P-10 or rack mountol bypass switches during testing (3)tianual bypass ocrmitted above P-6, blockal above P-10 (4)Autcmatically blockal by closing locp-stcp valves (5)Dlocko3 below P-9 (6) Low Pressure Pressurizer Safety Injection block pemitted by I'll q

Iow Steamline Pressure Safety Injection block permittal by P-11 Manual Block permittoi after Safety Injection Syston Reset arx1 P-4

(

i

TABLE 3.3-1 (continued)

With a channel associated with an operating loop inoperable, ACTION 9 restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in HOT STA!!DBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1.

ACTION 10 - Not applicable.

ACTION 11 - With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Charnels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the reactor trip breakers.

REACTOR TRIP SYSTEM INTERLOCKS DESICNATION CONDITION AND SETP0 INT FUNCTION P-6 1/2 Neutron flux (intemediate range)

Allows manual block of above setpoint source range reactor trip 2/2 Neutron flux (intermediate range)

Defeats the block of below setpoint source range reactor trip P-7 2/4 Neutron flux (power range) below Blocks reactor trip on:

setpoint (from P-10)

Low flow or reactor or coolant pump breakers 1/2 Turbine impulse chamber pressure open in more than two below setpoint (from P-13) loops, undervoltage, underfrequency, pressur-izer low pressure, and pressurizer high level P-8 2/4 Neutron flux (power range) below Blocks reactor trip on setpoint low flow P-9 2/4 Neutron flux (power range) below Blocks reactor trip on setpoint turbine trip F-10 3/4 Neutron flux (power range) above Allows manual block of setpoint power range (low setpoint) reactor trip BEAVER VALLEY - UNIT 1 3/4 3-7 l

PROPOSED WORDING

TABLE 3.3-1 (continued)

DESIGNATION CONDITION AND SETPOINT FUNCTION Allows manual block of intermediate range reactor trip and intermediate range rod stops (C-1)

Blocks source range reactor trip (back-up for P-6) 3/4 Neutron flux (power range)

Defeats the block of below setpoint power range (low setpoint) reactor trip Defeats the block of intermediate range reactor trip and intermediate range rod stops (C-1)

Input to P-7 P-13 1/2 Turbine impulse chamber pressure Input to P-7 above setpoint P-14 2/3 Steam generator water level above Closes all feedwater setpoint on any steam generator control valves Trips all main feedwater pumps which closes the pump discharge valves Actuates turbine trip BEAVER VALLEY - UNIT 1 3/4 3-8 PROPOSED WORDING

TAPLI' 3.3-3 ENGINEEPED SAFEfY FEAfDRE ACfUATICtl SYSTE21 INSTRUFEfffATIGI tn 9

MINItt1M N

TOPAL 10.

OIAPUEIS OfANFEIS APPLICAfY.E FUNCfIONAL UNIT O' OIANNELS

'IO TRIP

_OPERAIE.E 110DFS M'PIM1 C

M 1.

SAFEfY I!UECfION AND i

EEEIMMPER ISOIATION g

a.

Manual Initiation 2

1 2

1, 2, 3, 4 18 II b.

Autanatic Actuaticn 2

1 2

1,2,3,4 13 Logic c.

Containnent 3

2 2

1,2,3 14

,o Pressure-Iligh 8}t' d.

Pressurizer 3

2 2

1, 2, 3#

14 m

Pressure - Low DY c.

Low Steanline Pressure D

C Three Locps 3/locp 2/ loop 2/locp 1, 2, 35 14 i

U operating any locp any locp j

Two locps 3/locp 2/locp any 2/any 1, 2, 3#

15 operating cperatinq cperating locp locp l

IIIManual b1cxi pennitted after Safety Injection Systen Reset arri P-4 l

TABLE 3.3-3 (Continucrl)

EtCINEEPED SAFETY F5'tfUPE ACTUATION SYSPIN IFBPPUfDffATICN

~

ni

o g

MINIMUM TOfAL 10.

OINitEIS OIMRYTS APPLICAPIE j

FUNCfICtRL UNIT OF OIMMIS TO TRIP OPERABLE MODES ACTICU 4.

SPEAM LINE ISOIATION N

l

)

~

a.

Manual 2/stean line 1/ steam line 2/cperating 1, 2, 3, 4 18

{

stean line b.

Autanatic 2

1 2

1, 2, 3, 4 13 2

Actuation Logic y

c.

Containment Pressure--

3 2

3 1, 2, 3 14 yw Intermediate-liigh-liigh M2

@w d.

Iow Steanline Pressure g&

Three Locps (Locp Stcp 3/locp 2/locp 2/locp 1, 2, 3#

14 g'

Operating Valves Open)

Any log)

Any itxp ti l1 Two Locps 3/locp 2/locp any 2/any 1, 2, 3#

15 Operating cperating cperating locp loop 4

e.

High Stean Pressure 3/locp 2/locp 2/cperating 3##, 4 18 l

Rate any locp locp 1

b I

I i

I r

TABLE 3.3-3 (Continued)

E101NEEhED SAFETY iTA'IURE JCIVATION SYSTEM INSTRUNEhTATION i

to

$f MINIhl51 10fAL NO.

OlA! EELS CliANNELS APPLICABLE f

FUFCTIONAL LNIT OF OfAhNELS 101 RIP OPERABLE FODES JCIION s

N 5.

'IURBIliE TRIP & FLE0 WATER g

ISOIATION i'4" a.

Steam Generator 3/ loop 2/ log; in 2/ loop in 1, 2, 3 14 Water IcVel-any qcr-each g ar-liigh-11igh ating loop ating loop 6.

LOSS OF POWER a.

4.16kv Bus 1/4.16kv Bus 1/4.16kV Bus 1/4kv Bus 1, 2, 3, 4 33

$[

1.

Ioss of Voltage g4 (trip fealer) ge 2.

Ioss of Voltage c

(start diesel) 1/4.16kv Bus 1/4.16kv Bus 1/4kv Bus 1, 2, 3, 4 33 ti b.

Grid Degraded Voltage 2/4.16kV Bus 2/ Bus 2/ Bus 1, 2, 3, 4 34 (4.16kv Bus) c.

Grid Degraded Voltage 2/480v Bus 2/ Bus 2/ bus 1, 2, 3, 4 34 (480v Bus)

Administrative block permitted for two-log) q;eration 3

J I

Table 3.3-3 (Continued)

TABLE NOTATION

  1. Trip function may be bypassed in this MODE below P-11.
    1. Trip function automatically bypassed above P-11, and is bypasced below P-11 when Safety Injection on low steam pressure is not manually bypassed.
      1. The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped mode.

ACTION STATEMENTS With the number of OPERABLE Channels one less than the ACTION 13 Total Number of Channels, be in HOT STANDBY within six hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one chanael may be bypassed for up to two hours for surveillance testing per Specification 4.3.2.1.1.,

provided the other channel is operable.

With the number of OPERABLE Channels one less than the ACTION 14 Total Number of Channels:

a.

Belev P-ll or P-12, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter exceeding P-11 or P-12; otherwise be in at least HOT STAND 5Y within the' following six hours.

b.

Above P-il and P-12, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.

With a channel associated with an operating loop inoperable, ACTION 15 restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.2.1.1.

ACTION 16 - With the number of OPERABLE Channels one less than the Total Number of Channels:

a.

Below P-11 or F-12, place the inoperable channel in the bvpass condition; restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding P-ll or P-12; otherwise be in at least HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

BEAVER VAILEY - UNIT 1 3/4 3-20 PROPOSED WORDING

TAB _LE 3.3-3 (Continued) b.

Above P-11 or P-12, demonstrate that the flinimum Channels OPERABLE requirement is met witin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; operation may continue with the inoperable channel bypassed and one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for testing per Specification 4.3.2.1.

ACTICN 17 -

With less than the Minimum Channels OPERABLE, operation may continue provided the containment purge and exhaust valves are maintained closed.

ACTION 18 -

With the number of OPERABLE Chanr.els one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

ACTION 33 -

With the number of OPERABLE Channels one less than the Total Number of Channels, the Emergency Diesel Generator associated with the 4kv Bus shcIl be declared inoperable and the ACTION Statements for Specifications 3.8.1.1 or 3.8.1.2, as appropriate shall apply.

ACTION 34 -

With the number of OPERABLE Channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until the perfonnance of the next required Channel Functional Test provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ENGINEERED SAFETY FEATURES INTEPLOCKS DESIGNATION CONDIT0N AND SETPOINT FUNCTION P-4 Reactor Trip Actuates turbine trip.

Closes main feedwater valves on Tavg below setpoint.

Prevents opening of main feedwater valves which were closed by safety injection or high steam generator water level.

Allows manual block of the automatic reactuation of safety injection.

Reactor not tripped Defeats the block of the automatic reactuation of safety injection.

P-ll 2/3 Pressurizer pressure below Allows manual block of setpoint safety injection actuation on low pressurizer pressure signal.

Blocks automatic opening of the power relief valves.

2/3 Pressurizer pressure above Defeats the manual block of setpoint safety injection actuation.

P-12 2/3 Tavg channels below setpoint Blocks steam dump below setpoint. Allows manual bypass of the steam dump j

block to cooldown condenser dump valves.

BEAVER VALLEY - UNIT 1 3/4 3-21 PROPOSED WORDING

.-,-,n

ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEftS A.C. LISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 4

3.8.2.1 The following A.C. electrical busses shall be OPERABLE and energized from sources of power other than the diesel generators with tie breakers open between redundant busses:

4160 volt Emergency Bus #IAE and 480V Emergency Bus 8N 4160 volt Energency Bus #IDF and 480V Emergency R1s 9P 120 volt A.C. Vital Bus #1 120 volt A.C. Vital Bus #II 120 volt A.C. Vital Bus #III 120 volt A.C. Vital Bus #IV APPLICABILITY:

MODES 1, 2, 3 and 4 ACTION:

With less than the above complement of A.C. busses OPERABLE, restore the inoperable bus to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOW!! within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.8.2.1 The specified A.C. busses shall be determined OPERABLE and energized from A.C. sources other than the diesel generators at least once per 7 days by verifying correct breaker alignment and indicated power availability.

i l

PEAVER VALLEY - UNIT 1 3/4 8-6 PROPOSED WORDING l

CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTECRITY LIMITING CONDITIONS FOR OPERATION i

i 3.6.1.6 The structural !.ntegrity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.1.

l APPLICABLITY: FODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of the containment not conforming to the above reguirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F.

SURVEILLANCE REQUIREMENTS 4.6.1.6.1 Liner Plate and Concrete The structural integrity of the containment liner plate and concrete shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by:

a.

a visual inspection of the accessible surf aces and verifying no apparent changes in appearance or other abnormal degradation.

b.

a visual inspection of accessible containment liner test channels prior to each Type A containment leakage rate test. Any contain-ment liner test channel which is found to be damaged to the extent that channel integrity is impaired or which is discovered with a vent plug removed, shall be removed and a protective coating shall be applied to the liner in that area.

c.

a visual inspectica of the dome area prior to each Type A contain-ment leakage rate test to insure the integrity of the protective coating. If a loss of integrity of the protective coating is observed, any vent plug to a test channel which may be in the area where the protective coating has failed shall be seal welded and l

then the protective coating shall be repaired.

4.6.1.6.2 Reports An initial report of any abnormal degradation of the containment structure detected during the above required tests and inspections shall be made within 10 days af ter completion of the surveil-lance requirements of this specification, and the detailed report shall l

be submitted pursuant to Specification 6.9.1 within 90 days af ter l

comple tion. This report shall include a description of the condition of the liner plate and concrete, the inspection procedure, the tolerances on cracking and the corrective actions taken.

i SEAVER VALLEY - UNIT 1 3/4 6-10 PROPOSED WORDlHC l

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Nuclear DiviSioi t

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VICE PRESIDENT E r-e

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ENGINEERING AND E

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VICE PRESIDENT CONSTRUCTION NUCLEAR DIVISION DIVISION

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HAHAGER g

NANAGER SUPERlWIENDEN2*l'j U

NANALGER MANAGER

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PEVLOND H

NUCLt*,R NUCLEAR

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ADMa',ETRA710sl SUPPORT SERVICES ENGINEERINC 1

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GER. ENGINEERING ORC 1

DIRECT 0k NUCLEAR STATICN STARTUP NUCLEAR STATION SUPERit TENDENT FIRE AND NUCLEAR ENGINEERING i

EUPERINTEhi'ENT.

COORDINATOR SUPERINTENDENT TECHNICAL SAFETY ENGINEEll N

UNIT fl LV-2 SHIPPINGPURT SERVICES I

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DIRECTOR

,t RADict0GICAL SUPERINTENDENT DIRECTOR DIRECTOR COORDINATOR SUPIRINTENDENT ELECTRICAL OPERATIONS LICENSING ENVIR0tetENTAL OPERATIONS OFFSITE REVIEW ANALYTICAL ENGINEERING

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COORDINATOR AND COHPLIANCE AND RADIOLOGICAL QUALITY CONNITTEE SERVICES i

SAFETY PROGRAF!S CDNTROL

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RADIATION IDIRECTOR t(DIRECTORi

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OffSite Or9Tr.ization (Partial) i

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s CONTAINMEhT SYSTDiS CONTAINMENT AIR LOCKS s

LIMITING CONDITION EDR OPERATION x

N 3.6.1,3.

IMch containment air lock shall be OPERABLE with:

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

[

b.-

An overall air lock leakage rate of less than or equal to 0.05 L, at P,(38.3 psig).

_\\

APPLICABILITY: MODES 1, 2, 3 and 4.

. s ACTION:

3. ; m a.

With one containment air lock door inoperable:

w.,~

1.

Maintain the associated OPERABLE air lock door closed and either restore the associated inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or lock the associated OPERABLE sir lock door closed.

2.

Operation may then continue until performance of the next required overall air lock leakage test provided that the associated OPERABLE air lock door is verified to be locked closed at least once per 31 days.

f l

3.

Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> i

and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

4.

The provisions of Specification 3.0.4 are not applicable.

b.

With a containment air lock inoperable, except as the result

[

of an inoperable air lock door, maintain at least one air lock j

door closed; restore the inoperable air lock to 0PERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l LEAVER VALLEY - UNIT 1 3/46-5 PROPOSED WORDING l

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a.

Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following each containment entry, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, by verifying no detectable seal leakage when the gap between the docr seals is pressurized to greater quantifying the tota (38.3 psig) for at least 2 minutes, or by than or equal to P l air lock leakage to insure the require-ments of 3.6.1.3.b are met, b.

By conducting overall air lock leakage tests at not less than P (38.3 psig), and verifying the overall air lock leakage rate il within its limit:

1.

At least once per 6 months, # and 2.

Upon completion of maintenance which has been performed l

on the air lock that could affect the air lock sealing capability

  • c.

At least once per 18 months durina shutdown by verifying that only one door in each air lock can be opened at a time.

tThe provisions of Specification 4.0.2 are not applicable.

BEAVER VALLEY - UNIT 1 3/4 6-Sa PROPOSED WORDING

ATTACHMEllT B Safety Evaluation Proposed Change Request tio. 86 amends the Beaver Valley Power Station, Unit fio.1 Technical Specifications, Appenaix A concerning various administrative changes.

Descriotion and Purpose of Change 1

Section 3.2.6 - Axial Power Distribution, is being deleted in accordance with the NRC Safety Evaluation enclosed with Tech-nical Specification Amendment No. 9.

2.

Tables 3.3-1 and 3.3-3 have been revised to identify conditions under which operating bypasses will block the Reactor Trip System Instrumentation and Engineered Safety Feature actuation channe?s.

The pennissive descriptions have also been revised to conform to the Updated Final Safety Analysis Report (UFSAR) format.

3.

Section 3.8.2.1 - (insite Power Distribution Systems, has been revised tg require the 4160 volt energency bus and applicable 480 volt emergency bus to be operable.

4.

Section 3.6.1.1 - Containment Systems, has been revised to cor-rect a typographical error.

5.

The offsite organization, Figure 6.2-1, has been revised to update the title from Director to Superintendent, Personnel Administration and to incorporate the organization under the Manager, fluclear Engineering.

6.

Section 3.6.1.3 - Containment Air Locks, has been amended by revising the Action and Surveillance Requirements. The change te the Action statement consists of a provision to allow continued plant operation with one air lock door inoperable, provided that the operable air lock door is locked closed and verified to be locked closed at least once per 31 days. The locking of an air lock -ill be achieved by: disconnecting the power source of the D.C. motor driven pump unit which supplies the hydraulically operated latches on the personnel air lock doors, and by physically locking the emergency egress air lock doors (with lock and key to the outside door and a physical constraint to the inside door.)

The surveillance requirements have been revised to require essentially what is stated in the " Periodic Retest Schedule" of 10 CFR 50 Appendix J, Section III.D.2.

Also included is a change to require conducting overall air lock leakage tests when maintenance has been performed on that air lock.

Attachment B Safety Evaluation Page 2 Basis 1.

Is the probability of an occurrence or the consequence of an accident or malfunction of equipment important to safety as previously evaluated i: the UFSAR increased? M Reason:

The deletion of Section 3.2.6-Axisl Power Distribution is an administrative change and is consistent with the NRC Safety Evaluation enclosed with Techr.ical Speci-fication Amendment No. 9.

This change is not e safety concern, since it was only applica:;1e to Cycle 1 oper-ation.

The alteration of Table 3.3-1, Reactor Trip System In-strumentation and Table 3.3-3, Engineered Safety Feature Actuation System Instrumentation is administrative in nature and does not physically change plant safety related systems, components or structures. This change is being made in order to identify conditions under which operating bypasses will block the reactor trip or eng-ineered safety feature actuation channels and will not affect the function of any equipment or systems important to safety as addressed in the UFSAR Section 7.2, Reactor Trip System or Sectior 7.3, Engineered Safety Featurcs System.

The change to Section 3.8.2.1 - Onsite Power Dis-tribution Systems is administrative and will not affect any equipment or system as discussed in UFSAR Section 8.4, Station Service Systems or Section 8.5, En:ergency Power System. The change to Section 3.6.1.6 - Containment Structural Integrity, is being made to correct a typo-graphical error. The present specification references a non-existent section 4.6.1.7.

This change is admin-istrative and will not affact any equipment or system discussed in UFSAR Section 5.2, Containment Structure.

Figure 6.2-1, Offsite Organization, hss been revised to reflect reorganization changes. This change is not a safety concern and does not affect the UFSAR.

The prcposed changes made to Section 3.S.1.3 - Containment Air Locks, are administrative changes and do not involve any physical changes to the air locks or other p. ant equip-ment, nor do they degrade the leak tightness of the air lccks. The change to the Surveillance Requfrements remains essentially the same as 10 CFR 50 Appendix J Section III.D.2, periodic retest schedule of Type B test requirements.

For these reasons, the probability of an occurrence or the consequence of an accident or malfunction of equipment important to safety as previously evaluated in the UFSAR Sections 5.2.2.5.5, 5.2.4.8 and 5.6 is not increased.

~.

Attacheent B Safety Evaluation Page 3 2.

Is the probability for an accident or malfunction of a different type than previously evaluated in the Final Safety Analysis Report created? fpq Reason:

The proposed changes are administrative in nature and do not physically change the plant safety related systems, components or structures, therefore, the changes will not create the possibility for a new type of accident or malfunction of a different type than any previously evaluated in the UFSAR sections addressed above or the accident analysis of Section 14, 3.

Is the r.argin of safety as defined in the basis for any Technical Specification reduced? !!o.

Reason:

The Technical Specification BASES for the sections addressed above will not be affected by the proposed changes, as none of the systems or components will be physically changed or their function eitered in any way. Therefore, the margin of safety inherent in the applicable bases will not be reduced.

4.

Based on the above, is an unreviewed safety question involved? !!q Conclusion The proposed changes are administrative in nature end do not involve physical change to plant 3afety related systems, components or structures, will not increase the likelihood of a malfunction of safety related equipment, increase the consequences of an accident previously analyzed, nor create the possibility of a malfunction different than previously evaluated in the UFSAR. Therefore, it is concluded that, l

since the change does not involve an unreviewed safety question in accor-dance with 10 CFR 50.59, the proposed changes do not constitute a sig-i nificant hazards consideration. We have determined that this change will l

not authorize a significant change in the types or a significant increase l

in the amounts of effluents or in the authorized power level and will not result in any significant environmental impact. Therefore, pursuant to 10 'CFR 50.5(d)(4), no environmental impact statement, negative declar-ation or envircnmental impact appraisal is required.

The OSC and ORC have reviewed the proposed changes, and based on the above safety evaluation, it is concluded there is reasonable assur-ance that the public health and safety will not be endangered by opera-tions in the proposed manner.

t 1

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