Text

Proposed Tech Specs Incorporating Miscellaneous Administrative Changes,Including Changes to Administrative Control Section
	ML20064A831
Person / Time
Site:	Perry
Issue date:	09/13/1990
From:	; CENTERIOR ENERGY
To:	;
Shared Package
ML20064A830	List: ML20064A829; ML20064A831;
References
	NUDOCS 9009280145
	Download: ML20064A831 (26)
	v • d • e

' *

(i .,

PY-CE1/NRR-1214 L Attachm2nt 3 Page 1 of 26 ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY............................................... 6-1 6.2 0RGANIZ'ATION................................................. 6-1 6.2.1 Corporate............................................... 6-1

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6.2.2 U ni t 5 ta f f . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .~. . . . . . . . . 6-1 i s ywi . w . .> . un . -1. sw w. yw.vuoww wii. ee e..e........... w ,

Table 6.2.2-1 Ninimum Shift Crew Composition......................... 6-6 6.2.3 INDEPENDENT SAFETY ENGIEERING GROUP Function .............................................. 6-7 Composition............................................ 6-7

( Responsibilities....................................... 6-7 Records................................................ 6-7 6.2.4 SHIFT TEC$#IICAL ADVI$0R. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7 6.3 UNIT STAFF QUALI FI CATIONS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7 4.4 TRAINING.................................................... 6-8 6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW ColetITTEE (PORC)

Function .............................................. 6-8 Composition ........................................... 6-8 Alternates............................................. 6-8

. Meeting Frequency ..................................... 6-8 Quorus................................................. 6-9 Respons i b i l i ti es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-0 t-

@V Records................................................ 6-10 PERRY - UNIT 1 xxv 9009280145 900913 PDR ADOCK 050004401 P PNU tm'

PY-CEI/NRR-1214 L Attechn3nt 3 Page 2 of 26 i TABLE 3.3.2-1-(Continued) i ISOLATION ACTUATION INSTRUMENTATION ACTION rat ACTION 20

h~

ACTION 21 - Close the affected systen isolation valve (s) within one hour or:

a. In 0PEftATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. In Operational Condition *, suspend CORE ALTERATIONS, handling of irradiated fuel in the primary containment and operations with a potential for draining the reactor vessel.

ACTION 22 -

Restore the manual initiation function to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or:

a. In OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. In OPERATIONAL CONDITION *, suspend CORE ALTERATIONS, 1 operations with a potential for draining the reactor vessel, and handling of irradiated fuel in the primary containment.

ACTION 23 -

Be in at least STARTUP with the associated isola 1.irn valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTl)0WI within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ind in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l ACTION 24 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 25 - Verify SECONDARY CONTAINMENT INTEGRITY with the annulus exhaust gas treatment system operating within one hour. N ACTION 26 -

Restore the manual initiation function to 0PERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or close the affected system isolation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affacted system inoperable.N ACTION 27 - Close the affected system isolation valves within one hour and declare the affected system inoperable.ff ACTION 28 - Within one hour lock the affected system isolation valves closed, or verify, by remote indication, that the valve (s) is closed and electrically disarmed, or isolate the penetration (s) and declare the affected system inoperable.

@TES

When handling. irradiated fuel in the primary containrent and during CORE ALTERATIONS and operations with a potential for draiqing the reactor vessel.

- When any turbine stop valve is greater than 90% open and/or the key locked Condenser Low Vacuum Bypass Switch is in the normal position.

Ouring CORE ALTERATIONS and operations with a poten:ial for draining the reactor vessel.

1. The provisions of Specification 3.0.4 are not applicable.

PERRY - UNIT 1 3/4 3-15

D- ,

PY-CEI/NKR-1214 L Attachment 3 ;l Page 3 of 26 Insert 1 Action 20 - In OPERATIONAL CONDITION 1, 2 or 3, be in at least 80T SHUTD0VN vithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD S5trIDOVN vithin the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

In OPERATIONAL CONDITION t, suspend CORE ALTERATIONS and operations with a potential for draining the rea'etor vessel.

NJC/ CODED /3761 I

i b i l PY-CE1/NRR-1214 L Attcchment 3 Page 4 of 26 l I li4STRUi4Ei(TAT 10N l TRAVER$f NG IN. CORE PROBE SYSTEM l i

(,lMITING CONDITION FOR OPERATION 1

3.3.7.7 The traversing in-core probe system shall be OPERABLE with either:

a. Five movable detectors, drives and readout equipment to map the core, and indexing equipment to allow all five detectors to be calibrated .

r in a common location. -

j l

OR i

b. With one or more i!P measurement locations inoperable, data may be I replaced by data obtained from that location's symetric counterpart if the. Substitute TIP data was obtained from an OPERABLE measurement 'I location; provided the reactor core is operating in a type A control 1 rod pattern and the total core TIP uncertainty for the present cycle ;

has been detemined to be less than L.7 percent (standard deviation). ';

Symetric counterpart data may be substituted for a maximum of ten i TIP measurement locations.

, APPLICABILITY: When the traversing in-core probe is used for:

L: a.* Recalibration of the LPRM detectors, and

i b.* Monitoring the APLHGR, LHER. MCPR. ;,, ^.~C(. )_ i ACT!*.!j,: '

^

With the traversing in-core probe system inoperable, do not use the system for the above applicable monitoring or talibration functions. The provisions of a Specifications 3.0.3 and 3.0.4 are not applicable.

l SURVEILLANCE REQUIREMENTS i 4.3.7.7 The traversing in-core probe system shall be demonstrated OPERABLE by I normalizing each of the above recuirg detector outputs _within 72_ hours rior' ,

to use when required for the['f*M:'En'J: ' :^/-- cacwe qppb 4.,

l LPRii ca M M A M .q 94. Man

%s l.

0nly the detector (s) in the location (s) of interest are required to be OPERABLE.

PE"tRY - UNIT 1 3/4 3-82 Amendment No. 25 t.

j PY-CEI/NRR-1214 L !

Attachment 3 .l

, Page $ of 26 j h

DEFINITIONS-

@', I DRWELL INTEGRITY (continued) '

f. The suppression pool is in compliance with the requirements of 1

Specification 3.6.3.1. ;

\

g. The sealing mechanise associated with each drywell penetration; e.g., :

welds, bellows or 0 rings, is OPERA 8LE. i i

I-AVERAGE DISINTEGRATION ENEltGY i

I 1.12 I shall be the average, weighted in proportion to the concentration of

each radionuclide in the reactor coolant at the time of sampling, of the sum of i the average _ beta and gamma energies per disintegration, in MeV, for isotopes, ,

with half lives greater than 15 minutes, making up at least 955 of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEN (ECCS) RESPONSE TIME l 1.13 The EMERGENCY C0RE COOLING SYSTEM (ECCS) RESPONSE TIE shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e. , the valves travel to their required positions, pump dis-charge pressures reach their required values, etc. Times shall include diesel

. generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured. 1 l

l EW-0F-CYCLE RECIRCULATION Ptw TRIP SYSTEN RESPONSE TIME i

1.14 The END-0F-CYCLE RECIRCULATION PLMP TRIP SYSTEM RESPONSE TIE shall be I that time interval to complete suppression of the electric are between the fully open contacts of the recirculation pump circuit breaker from initial ,

movement of the associated:

)

a. Turbine stop valves, and

b. Turbine control valves.

1 The response time may be measured by any series of sequential, overlapping or )

DSLR4. total steps such that the entire response time is measured.

A NON OF LIMITING POWER DEltSITY 1.15 The FRA MITING POWER DENSITY (FLPD) s LNGR existing at a given locatio by the speci limit for that bundle I type. !

l FRACTION OF RATED R j

i S. 1. TION OF RATED THERMAL POWER (FRTP) shall be the Bl4L I k"T POWER divided by the RATED TNERNAL POWER.

l l l

l PERRY - UNIT 1 1-3 1

1

[- .__

i e PY-CE!/NRR- 1214 L Attechtent 3 i

Page 6 of 26 i

u.o f

DEFINITIONS ,

3.gN, 1.0 DEFINITIONS M 1.1 ACT10N....................................................... 1-1

1. 2 AVERAGE PLANAR EXP050RE...................................... 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................... 1-1 1.4 CHANNEL CALIBRATION...............................'........... 1-1 1.5 CHANNEL CHECK..........'...................................... 1-1 1.6 CHANNEL FUNCTIONAL TEST...................................... 1-1 1.7 CORE ALTERATION.............................................. 1-2 1.8 CORE MAXIRM FRACTION 0F LIMITING POWER DENSITY. . . . . . . . . . . . . . 1-2 1.9 CRITICAL POWER RAT 10......................................... 1-2 1.1V DOSE EQUIVALENT I-131........................................

1-2

{,

' 1.11 DRYWELL INTEGRITY.....,...................................... 1-2 1.12 E- AVERAGE DISINTEGRATION ENEMY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.13 EERGENCY CORE COOLING SYSTS (ECCS) ESPONSE TIM........... 1-3 1.14 De-0F-CYCLE RECIRCULATION MBF TRIP SYSTM RESPONSE TME.... 1 ......_..-............-__n

...........n. . . . . .

1.17 FREQUENCY N0TATION...........................................

1-4

. .- 1.18 FUEL HANDLING BUI LDING INTEGRITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1-4 3 1.19 GASE0US RADWASTE TREATENT (OFFGAS) SYSTS. . . . . . . . . . . . . . . . . . .

1.20 IDENTIFIED LEAKAGE........................................... 1-4

1. 21 I CALATION SYSTEN RESPONSE TIE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.22 LIN2 TING CONTROL R00 PATTERN................................. 1-5 1.23 LINEAR HEAT GENERATION RATE.................................. 1-5 PERRY - UNIT 1 1

PY-CEI/NRR-12141.

Attackm2nt 3 Page 7 of 26 l-INSTRUMENTATION -

SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 18 months by perfomance of a CHANNEL CALIBRATION of the turbine overspeed protection instrumentation.

a '

gg g c. At least once per 40 monthkby disassembling at least one of each j of the above valves and performing a visual and surface inspection (4P) of all valve seats, disks and stens and verifying no unacceptable flaws or excessive corrosion. If unacceptable flaws or excessive corrosion are rund, all other valves of that type shall be inspected.

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1: tr.; r:0 ::'n!';; L^.:;;.

.( l PERRY - UNIT 1 3/4 3-97 Amendment No.18

_ _ _ _ _ _ _ _ _ mm. - '

r s -.

PY-CE1/NRR-1214 L l Attochment 3 !

Page 8 of 26 REACTOR COOLANT $Y$ TEN C .

IDLE RECIRCULATION LOOP $7ARTUP l

i l

1 LINITING CONDITION FOR OPERATION I

3.4.1.4 An idle recirculation loop shall not be started unless the :

temperature differential between the reacto'r pressure vessel steam space I coolant and the bottos head drain line coolant is less than or equal to l 100*F*, and:

j

a. When both loops have been idle, unless the temperature differential ;

between the reactor coolant within the idle loop to be started up and the colant in the reactor pressure vessel is less than or equal i

&M to 50' or

b. When only one loop has been idle, unless the temperature differential between the reactor coolant within e idle and operating recirculation

, g_ m loops is less than or equal to 50' nd the operating loop flow rav.e is less than or equal to 50% of rate floop flow.

( l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With temperature differences and/or flow rates exceeding the above limits, suspend startup of any idle recirculation loop. ;

$URVEILLANCE REQUIRENENTS 4.4.1.4 The temperature differentials and flow rate shall be detamined to be within the limits within 15 minutes prior to startup of an idle recirculation loop.

g *Below 25 psig, this temperature differential is not applicable. i PEARY - UNIT 1. 3/4 4-6

PY-CEI/NRR-1214 L Attcchment 3 Page 9 of 26 gv CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued)

l. Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 L,. The forwula to be used is:

[L, + L,, - 0.25 L,3 i eLi R, + Lam + 0.25 L.) where L, =

supplemental test result; L, a superimposed leakage; i

L,, = measured Type A leakage.

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

3. Requires the quantity of gas injected into the primary contain-ment or bled from the primary containment during the supple.

'nental test to be between 0.75 L, and 1.25 L,.

_d. Type B and C tests shall be conducted ith gas at P , 11.31 psig*, !

l DSLET1.- r at intervals no greater than 24 month except for tests involving: i b -

1. Air locks. '

g

2. Main steam line isolation valves.

3. Valves pressurized'with fluid from a seal system.

(' 4. All containment isolation valves in hydrostatically tested lines per Table 3.6.4-1 whicn penetrate the primary contairement, and y

5. Purge supply and exhaust isolation valves with resilient .

material seals,

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

f. Main steam line isolatiori valves shall be leak tetted at least once . ,

per 18 months. l

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

system may be excluded, subject to the provisio1s of Appendix J of 10 CFR 50 Section III.C.3, when determining the combined leakage rate :

provided the seal system af.d valves are pres',urized to at least i 1.10 Pa 12.44 psig, and the seal system c%pacity is adequate to ;

maintain system pressure for at least 30 days,

' b" E h. All containment isolation valm in hydrostatically tested lines per '

Table 3.6.4-1 which penetrate the yimary containment shall be leak hh r tested at least once per 18 month a s

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

) . interval extenston to the first refueling outage is g '

l for primary con tion valves listed in '

. , @ich are t identified in letter PY-CE!/ -

p ember 11. 1987) as r.eeding g,) -a plant outa e s one time te -

+he provisio?.s of

.0.2 are not applicable. %

PERRY - UNIT 1 3/4 6-5 Amendment No. 5.J9.22

PY-CEI/NRR-1214 L

.. . Attachment 3

- Page 10 of 26 CONTAINMENT SYSTEMS

['

DRYWELL AND CONTAINMENT PURGE SYSTEM LIMITJNGCONDITIONFOROPERATION 3.6.1.8 The drywell and containment purge 42-inch outboard (1M14-F040, F090) supply and exhaust isolation valves and the 18 inch supply and exhaust isolation valves (1 Pila-F190, F195, F200, F205) shall be OPERABLE and:

a. Each 42-inch inboard purge valve (1M14-F045, F085) shall be sealed closed,

b. Each 42 inch outboard purge valve (1M14-F040, F090) may be open limited to an opening angle of 50* or les or purge system operation

with such operation limited toO908 hnur per 365 days for reducing airborne activity and pressure control _

c. Each 24-inch (1M14-F055A, 8 and F060A, s; ano 36-inch (1M14-F065, F070) drywell purge valve shall be sealed closed.

d. Each 2-inch (1M51-F090 and F110) backup hydrogen purge system iso-lation valves may be open for controlling drywell pressure.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a. With 6 42-inch inboard drywell and containment purge supply and/or

.(

- exhaust isolation valve (s) open or not sealed closed, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months close ard/or seal the 42-inch valve (s) or otherwise isolate the penetratton or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in CM D SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. With a 18-inch or 42 inch outboard drywell and containment purge supply and/or exhaust isolation valves inoperable or open for more than 3000 hours0.0347 days 0.833 hours 0.00496 weeks 0.00114 months per 365 days for purga system operation *, within four hours close the open 18- or 42-inch valve (s) or otherwise isolate the penetration (s) or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUT 00WN within the following 24' hours.

c. With a 24- or 36-inch drywell purge supply and/or exhaust isolation valve (s) open or not sealed closed, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months close and/or seal close the 24- or 36-inch valve (s) or otherwise isolate the penetra-tion, or be in at least H0T SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

d. With a drywell and containment purge supply and/or exhaust isolation valve (s) with resilient material seals having a measured leakage rate exceeding the limit of Surveillance Requirement 4.6.1.8.3 and/or

" Purge system operation shall be defined as any time that both 18-inch and the 42-inch outboard purge valves are open concurrently in either the supply or exhaust line.

,s..u,. >-s , -. , . . s -2, - -- - - - ,, >-- --- --- , - -

. .. - -- 2

,2 _ e A L - *- L 'A

...-1 . - . . . . . . . , - - -, - - . . . . .. , . ...- --. r.. --- --, .-.- -vr .--.

PERRY - UNIT 1 3/4 6-12

i

. j PY-CEI/NRR- 1214 L j Attachmsnt 3 j Page 11 of 26 j

. PLANT SYSTEMS BASES .

i SNU68ERS (Continued) j

' DELgrr, /pf g

1. Functionally test 10% of a type of snubber with an additional

? tested for each functional testing failure, or [

Za h T $ ( 2. Functionally test a sample size and detemine sample acceptance or rejection using Figure 4.7.4-1, or i

i

3. Functionally test a representative sample size and detemine sample !

acceptance or rejection using the stated equation. j Figure 4.7.4-1 was developed using "Wald's Sequential Probability Ratio l Plan" as described in " Quality Control and Industrial Statistics" by Acheson J. !

Ouncan.

Permanent or other. exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption !

is presented and, if applicable, snubber life destructive testing was perfomed I to qualify the snubbers for the applicable design conditions at either the com- j plation of their fabrication or at a subsequent date. Snubbers so exempted shall be itsed in the list of individual snubbers indicating the extent of the ;

exemptions.

The service life of a snubber is evaluated via manufacturer input and ;

information through consideration of the snubber service conditions M asso-

I ciated installation and maintenance records (i.e. , newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area,etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a perfomance evaluation in view of their age and operating conditions. These records will provide statis-tical bases for future consideration of snubber service life.

3/4.7.5 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak ,

testing, including alpha esitters, is based on 10 CFR 70.39(c) limits for ,

plutonium. This limitation will ensure that leakage from byproduct, source, i and special nuclear material sources will not exceed allowable intake values. '

Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a. ,

source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism, i.e., sealed sources within radiation monitoring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

i PERRY - UNIT 1 B 3/4 7-3

_ _ _ - - - - . _ . _- .- __ . _ . - - _ - . ---n_-_---

.4 +

", PY-CEI/NRR-1214 L j Attachment 3 i Page 12 of.26 ;

3/4.8 ELECTRICAL POWER SYSTEMS -)

h. 3/4.8.1 A.C. SOURCES A.C. SOURCES - OPERATING j LIMITING CONDITION FOR OPERATION l

i

3. 8.1.1 As a minimum, the following A.C. electrical power sources shall be ;

OPERA 8LE*

a. Two physically independent circuits between the offsite transmission !

network and the onsite Class 1E distribution system, and ;

s i

b. Three separate and independent diesel generators, each with: )

1. A separate day fuel tank containing a minimum of 225 gallons of l ftel for Div 1 and Div 2 and 204 gallons of fuel for Div 3, i 1

2. A separate fuel storage system containing a minimum of I 69,430 gallons of fuel for Div 1 and Div 2 and 34,424 gallons j of fuel for Div 3, and 1

3. A separate fuel transfer pump.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION: J l

a. With one offsite circuit of the above required A.C. electrical. power TMSElR'T"- e sources inoperable, demonstrate the OPERABILITY of the remaining A.C. ,

sources by performing Surveillance Requirement 4.8.1.1.1.a within I L 6,yg, pty I nour and at least4 hours thereafter. If either diesel generator Div 1 or Div 2 has not been successfully tested within the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , demonstrate its OPERABILITY by performing Surveillance Re- 1 quirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 for each such diesel gen- I erator separately within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the offsite circuit to OPERA 8LE status within Tt hours or be in at least HDT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and ir, COLD SHUTOOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. With either diesel generator Div 1 or Div.2 inoperable, d.aonstrate the OPERA 81LITY of the above required A.C. offsite sources by perform-ing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. If the diesel generator became inoperable ;

due to any cause other than preplanned preventive maintenance or j testing, demonstrate the OPERABILITY of.the remaining OPERABLE diesel generators by performing Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 separately for each diesel generator within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *;

"This test is required to be completed regardless of when the inoperable diesel (V

l Q;;s

[ generator is restored to OPERABILITY. The provisions of Specification 3.0.2 are not applicable.

l PERRY - UNIT 1 3/4 8-1 1

I l

_____~._._._.i_.________.__.____________________._________ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ , . _ , _

PT-CEI/NRR-12I4 L !

' ' Attachment 3' Pags 13 of 26 '

i

[' TABLE 4. 8.1.1. 2-1 . -

i OIESEL GENERATOR TEST SCHEDULE !

Number of Failures Number of Failures in in Last 100 Valid

.t.ast 20 Valid Tests * -Tests

Test Frequency 1 l

11 14 Once per 31 days j i

1 2"* 15 Once per 7 days ;

l 1

4 4

(I

Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, but determined on a per diesel generator basis.

For the purposes of-determining the required test frequency, the previous test failure count may be reduced to zero if a complete diesel overhaul # to like- I new conditicn is completed, provided that the overhaul . including appropriate g post-maintenance operation and testing, is specifically approved by the manu-facturer and if acceptable reliability has been demonstrated. The reliability '

c riterion -shall be the successful connletion of -14' consecutive tests in a k' {,)J,. murveil'inale teries. Ten of these tests stal ance Requiremen .1'......; t pin andaccordance 4.8.1.1.2.a.5, with the routine four tests, in accor-dance with the 184-day testin re trement of Surveillance Requirements

},gg,g,hr34.8.1.1.2.a.4an. f this critcrion is not satisfied during the 71rst series or f.ests, any alternate criterion to be used to transvalue the failure count to zero requires NRC approval.

- The test frequency shall be maintained until seven consecutive failure free i demands have been performed and the number of failures in the last 20 valid i demands has been reduced to Tess than or equal to one.

A one-time waiver to the requirement for performance of a complete diesel generator overhaul to like-new condition has been granted in order to rezero four control air related diesel generator failures-(valid failures Nos. 3 L through 6 which occurred on 8/11/86, 2/27/87, 3/17/87 and 10/15/87 respectively).

. PERRY - UNIT 1 3/4 8-10 Amendment flo. 12

,i",

PY-CEI/NRR-1214 L'

, Attachm2nt 3 Page 14 of 26 INSTRUMENTAT10N SASES

}.

MONITORING INSTRUMENTATIO1 (Continued) 3/4.3.7.4 REMOTE SHUTDOWN INSTRUMENTATION AND CONTROLS The OPERABILITY of the remote shutdown monitoring instrumentation and controls ensures that sufficient capability is available to pemit shutdown and maintenance of HOT SHUTDOWN of the unit from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.

3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that suffi-cient information is available on selected plant parameters to monitor and assess important variables following an accident. This capability is consistent with the recomendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident,"

December 1975 and NUREG-0737, " Clarification of TMI Action Plan Requirements,"

Novencer 1980. The CHANNEL CHECX for the Primary Containment Isolation' Valve Position consists of the verification that indication of valve position (open or closed) can be determined by the valve position lights in the control room. The CHANNEL CALIBRATION for the Primary Containment Isolation Valve Position consists ;

of the Position Indicator Test (PIT), which is conducted in accordance with Specification 4.0.5.

y, n. g N' 3/4.3.7.6 SOURCE RANGE MONITORS j The' source range monitors provide the operator with infomation of the status l of the neutron level in the core at very low power levels during startup and shut-down. At these power levels, reactivity additions shall not be made without this flux level information available to the operator. When the intermediate range monitors are on scale, adequatte infomation is available without the SRMs.and they '

can be retracted. ,

The SRMs are required OPERABLE in OPERA.TyAL CON 0iTION 2 to provide for rod block capability, and are required OPERABLE in 7CCMONDITIONS 3 and 4 to provide '

monitoring capability which provides diversity of protection 7to the mode switch interlocks.

3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM

' The'0PERABILITY of the traversing in-core probe system with the specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial gama flux distribution of the reactor core. With less than the specified complement of equipment, the spatial

.gama flux distribution of the reactor core can still be accurately represented by using replacement data from symetrical strings (LFRM locati s), provided the conditions specified in the LCO are met. ;

The TIP system OPERABILITY is denonstrated by nomalizi all probes (i.e. ,

detectors) prior to perfoming an LPRM calibration function. itoring core themal limits may involve utilizing individual detectors to monit selected areas of the hq, reactor core, thus all detectors may not be required t LE. The OPERABILITY of individual detectors to be used for monitoring is demonstrated by comparing the detector (s) output with data obtained during the previous LPRM calibrations.

l PERRY - UNIT 1 B 3/4 3-5 Amendment No. 25 +

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

PY-CEI/NRR-1214L !

At t achmant 3

"T" Page 15 of 26 eseft sCatt se esCtets i Assyt vtsstL REmo !

gust $8 LEW4L meestettatunt t

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Ci c '~~ nitiate LPCI and LPCS,

- se s.s ' ,

ase Start Div I and Div II Diesels, Contribute to ACS, and close htsivs j i

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so REACTOR .0 5EL WATER LEVEL k~f h Bases Figure B 3/4 3-1 NF PERRY - UNIT 1 B 3/4 3-8 ." ,

e - . - .. , . m..%-_ _ _ . _ . - ____.____________mm__._____.__._____.________ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ . _ . _ _ _ . _ _ _ _ _ _ _ _ . , _ _ _ . _ _ _ _ _ . _ . _

s- + .

PY-CE1/NRR-1214 L Attachment 3 1

Page 16 of 26 ,

700 = =

l Scale in Scale in inches inches stove above top of Active

( ves:e1 tero Fuel (TAF) ,

34,11ett*tt naareset a 400 * = tavat me. treat aeoit vesset geste % . -.

II#8 i "tatII (setISmi etevt t5F3

( meats) 750 = = g,3 ,,,,, tl,',

727 = ~363.5-vessel (tl ses.s b Flange a tal see.s se,s,g g ,,3 700 = = (38 S48 8 att.t (3) 493.3 gpg,3 (1I 30s.3 16.S 650 = = Main 644.3 " - 281 Steam Line 600 = = wide Narrow Range Range hf'-219. 5-espes.

( 564 6 -=201.1= Netr.81 =205.1-High Level Altra NO" "

y,ter ame tage ogge((7)

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contribute to ads, 3

reec-48603" E 124.15 2 '== *127. 8 *3a111*** # Cit. to " Pes. t ripclose r ecirc pumps Water 474.25 == 110. 75 = C o r e * *'t ">cs tene si eteest. LFMC5, RHR 5 pray t ' I ' * ** 1 ' ' * '"* ' ' 2 * ** shutoown cooling 4$0 = = Peteer, stetoe teetetten g,gg ,*

volvee eseeet nun en t.

seen teeting teetetten vetves oas matv's erevise stonet to enes.

400 = = (1) - 16.3-Initiate RHR (LPC1 Mode) and LPCs,

. (_

g 5. Start Div ! and Div !! Diesels, 3'y ",",0 i n . 7 Contribute to AOS, one close MSIV's 00 = = g,ggv, Fuel 250 = =

w 213.5- = -150 - -

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Recite I7h,3 g7 5 ' -

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( REACTOR VESSEL WATER t.EVEL Bases rigure B 3/4 31 I 3 3/4 }.8 PERP,Y UNIT 1

1

+.

PY-CE! /NRR- 1214 1.

Attochm2nt 3 Page 17 of 26

.. 6.0 ADMINi$TRATIVE CONTROLS !

.(!g .

6.1 RESPONSIBILITY !

Nudes.rbr pg j 6.1.1 The General Manager. Perry lant 0;;.i:_ Departme(n . sha 1 be i I responsible for overall unit ope tion and shall delegate in riting the succession to this responsibility during his absence.

6.1.2 The Shif t Supervisor or, during his absence from the control room, l a designated individual shall be responsible for the control room.connand function. A management directive to this effect. signed by the Vice i President - Nuclear shall be reissued to all station personnel On H an annual basis. m j 6.2 ORGANIZATION 6.2.1 0FFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation !

and corporate management, respectively. The onsite and offsite organiza- 1 tions shall include the positions for activities affecting the safety of I the nuclear power plant. I

a. Lines of authority, responsibility, and connunication shall be ,

established and defined from the highest management levels '

through intemediate levels to and including all operating

(' organization positions. These relationships shall be documented and updated, as appropriate, in the fom of organization charts. '

functional descriptions of departmental responsibilities and -

relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. -These requirements shall be documented in the USAR and updated in accordance with i

10 CFR 50.71(e).

Qukt kar Nf+b

b. The General Manager. Pl PerryA ant 0;;12.'.::.QDepartment 4 shall be responsible for overall unit safe operation an shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c. The Vice President - Nuclear 11 have corpsrate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable pedermance of i the staff in operating, maintaining, and pmviding technical support to the plant to ensure nuclear safety. '

l

d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independ-ence from operating pressures.

h l,

PERRY - UNIT 1 6-1 Amendment No. 13 l

V _

% 0 8 PY-CEI/NRR-1314 1. j Attochment 3 Page 18 of 26 ,

l y ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION l

6.2.3.1 The ISEG shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources of unit design and operating experience infomation, including units ef simil-ar design, which may indicate areas for improving unit safety. The IhEG shall <

make detailed recomendations for revised procedures, equipment modifications. I maintenance activities.4 operations activities, or other means of improving unit safety to the Director Nuclear Engineering Departmen l

COMPOSITION 9ereg hAEk 6.2.3.2 The !SEG shall be composed of at least five, dedicate.d, full-time !

engineers or technically oriented individuals located onsite. Each shall have either (1) a bachelor's degree in engineering or related science and at least i 2 years professional level experience in his field, at least 1 year of which i experience shall be in the nuclear field, or (2) equivalent work experience as described in Section 4.1 of ANSI /ANS 3.1, December 1981. ,

r RESP 0NSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of unit

( activities to provide independent verification

that these activities are per-formed correctly and that human errors are reduced as much as practical. ,,

RECOROS 6.2.3.4 Records of activities perfomed by the ISEG shal be prepared, main-tained, and forwarded each calendar month to the Director (Nuclear Engineering Department. l 6.2.4 SHIFT TECHNICAL ADVISOP, 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Shif t Supervisor in the areas of thermal hydrauli-s, reactor engineering, ,

and plant analysis with regard to 'afe operation of the unit. The Shift Tecnnical Advisor shall have a Nchelor's degree oc. equivalent in a scientific .

or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room.

6.3 UNIT STAFF QUALIFICATI QS1 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifica-tions of ANSI MIB,1-1971 for comparable positions, except for the Plant Health Physicist wno shall meet or exceed the qualifications of Regulatory Guide 1.8, l

. (p~.

September 1975. The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 flRC letter to all licensees. ,

'Not responsible for sign-off funct, ion.

PERRY - UNIT 1 6-7 Amendment No. 22

g + .

PY-CEI / NRR- 1214 L

. Attachannt 3 Page 19 of 26 ADMINISTRAT!vE CONTROLS 6.4 TRAINING 6.4.1 A re raining and replacement training program for the unit staff shall be maintaine(d under the direction of the Perry Training Section Manager, and l shall meet od exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix A o' 10 CFR Part 55 and the supplemental require-ments specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience.

6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORCl s

Nuclear hower 6.5.1.1 The PORC shalp^ unction g R., ^to: .advise the. ,General s en. , I;. Manager, II;..; 7.;T... .;;' &;ee.Perry t Plant- g

^^'

Q_ ^g;. ;t'.;r.; Depart:nent,. .

- n n ('"" } . on all ma tters M....ated to nuclear safety.

d COMPOSITION 6.5.1.2 The PORC shall be comoosed of the:

Nu.\ ear

( Chainnan: p,D1 Director, Perry "';.c.; T.Q...' neefing Vice-Chainaan/ Member: Manager)0perations Section Department

_ Manager,b';;M'n! Err-- wSq<Ae.ns,qoep'q Sec.Q Member:

Member:

keg ' Manager W4aintenance Sectioh Reactor Engineer Member: Manager, Radiation Protection Section Member: Plant Health Physicist Member: Manager, instrumentation and Control Section Member: Manager, Licensing and Compliance Section ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PORC Chairman to serve on a temporary basis, however, no more than two alternates j shall participate as voting members in PORC activities at any one time.

MEETING FREQUENCY

6. 5.1. 4 The PORC shall meet at least once per calendar month and as convened by the PORC Chairman or his designated alternate.

PERRY - UNIT 1 6-8 lcendment No. 22

p 3=:,,

F3 g, t-PY-CEI/NRR-1214L

,c.,

Attechmant 3 Page 20 of 26 p's

.W

, ADMINISTRATIVE CONTROLS QUORUM

6. 5.1. 5 The quorum M tm FiW. necessary for the perfomance of the PORC responsibility and at. thor..s o tvisions of these Technical Specificationf, shall consist of the Chm.r. or his designated alternate and at least four members including alternatts. l

,, RESPONSIB'!LITIES 1

6. 5. l'. 6 The PORC shall be responsible for:

a. Review of all Administrative Procedures;

b. ' eview of the safety evaluations for (1) proposed procedures /

...,tructions, (2) changes to procedures / instructions, equipment, systems or facilities, and (3) tests or experiments performed under the provisions of 10 CFR 50.59 to verify that such actions do not 'i constitute an unreviewed safety question; -i

c. . Review of proposed procedures / instructions and changes' to procedures /

instructions, equipment, systems or facilities which involve an- .

unreviewed safety question as defined in 10 CFR 50.59;

d. Review of proposed tests or experiments which-involve an unreviewed safety question as defined in 10 CFR 50.59; -

e. Review of proposed changes to Technical Specifications or the i Operating License;

f. Investigation of all violations of the Technical Specifications- !

including the preparation and fomarding of reports covering evalua- 1 tion and recommendations to prevent recurrence to the Vice President -

Nuclear Gqnd to_ the Nuclear Safety Review Committee; j

g. Review of all REPORTABLE EVENTS;

h. Review of the plant Security Plan and Security Contingency Instruc-tions; I

1. Review of the Emergency Plan and implementing instructions; l

j. Review of changes to the PROCESS CONTROL PROGRAM, the OFFSITE DOSE- 1 CALCbuTION MANUAL, and Radweste Treatment Systems; 4 k. Review of any accidental, unplanned or uncontrolled radioactive release including the preparation of reports covering evaluatSn.

recomendations, and disposition of the corrective action to prevent recurrence and the forwarding o hese reports:to the General ( <

pMr becs Manager, Perry & Plant On n 9 1 epartment%"Z6;r n; Trnr4 ~

r-hP - q o _ "; ; "':n 'rH r' 2:;r^ m '"""} . the Nuclear Safety Review .

_ # Comittee and the Vice President - Nuclear Group; '

1. Review of Unit operations to detect potential hazards to nuclear safety; h m. Investigations or analysis of special subjects as requested by the Chairman of the Nuclear Safety Review Connittee; and

n. Review of the Fire Protection Program and implementing procedures, t PERRY - UNIT 1 6-9 Amendment No. 22 ,

.( g .g )

PY-CEI /NRR- 1214L s &yL+4 Attcchment 1 Page 21 of 26 E ADMINISTRATIVE CONTROLS -

RESPONSIBILITIES (Continued) l 6.5.1.7 The PORC shall:-

pO)

, ]l

a. Reconenend in writing to the General Manager ATTO:/0ir;c. . .?7 approval or disapproval of items considered under Specifications

6. 5.1. 6a. through e. , h. , i. , J. , and k. , - above prior to their 3 implementation; j b.- Render determinations in writing with regard to whether or not each item considered under Specifications 6.5.*1.6b. through e. ,

above, constitutes an unreviewed safety question; and c Provide _ written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice President .-

C S. nuclear Geowp I

ment.between theand theand PORC Nuclear Safety the General ReviewPerry Manager, Conunittee 4 Plant' of disagreeAudear # !

D- 0;;r:ti:-- Department; however, the General Manager, Perry 6TTantW\e.c r l N O *--W : : Ocoartment, shall have responsibility for resolution of ,

such disagreements pursuant- to. Specification 6.1.1 above.-

RECORDS' 6.5.1.8 The PORC shall maintain written minutes of each PORC meeting that.

(.. 'at a minimum, document the results of all PORC activities perfonned under the ,

responsibility provisions of these Technical Specifications.. Copies shall be

. provided to the Vice President - Nuclear .and the Nuclear Safety Review Commi ttee.-

6.5.2 NUCLEAR SAFETY REVIEW ComITTEE (NSRCI ,

, ' FUNCTION ,

6.5.2.1 The NSRC shall function to provide independent review and audit of

. designated activities in the areas of:

a. Nyclear power plant operations,

b. Nuclear engineering,

c. Chemistry and radiochemistry,

d. Metallurgy..

e. Instrumentation and control,

,. f. Radiological safety,

g. Meenanical and electrical engineering,

h. Quality assurance practices and administrative controls, and

i. Nondestructive testing.

The NSRC shall report to and advise the Vice President - Nuclear on those #

areas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8.

. PERRY - UNIT 1 6-10 Amendment No. 22

gs, PY-CEI/NRR-1214 L e^

Attachment'3-Page 22 'of 26 ADMINISTRATIVE CONTROLS

' COMPOSITION-6.5.2.2 _ The membership' of the NSRC shall be composed of at least eight person-nel appointed by the Vice President - Nuclear to provide collective ex-perience and competency in the following areas:

Nuclear power plant operations Nuclear engineering

, Chemistry and radiochemistry Hetallurgy-Nondestructive testing Instrumentation and control ,

Radiological safety Mechanical and electrical engineering Administrative controls .and quality assurance practices The Chairman, appointe by the Vice President - Nuclear , shal1~have 10 years of power plant expeiknce, of whica 3 years sL311 be nuclear power plant experience. The NSRC members : hall hold a bachelorc' degree in an engineering or physical science field, or equivalent experience, (nd a minimum of 5 years of technical experience of which a minimum of 3 years shall be in one or" more of the disciplines in Specificatici 6.5.2.1. Ceapetent alternates may be designated in. advance and consulta.its any be used for in-depth expertise if desired by the committee.

( ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by-the NSRC 4-

. Chairman to serve on a temporary basis; however, no more than two alternates shall participate as' voting members in NSRC activities at any one. time.

CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the NSRC Chairman to provide expert advice to the NSRC.

- EETING FREQUENCY 6.5.2.5 The NSRC shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per 6 months-.thereafter.

QUORUM' 6.5.2.6 The quorum of the NSRC necessary for the performance of the NSRC review and audit functions of these Technical Specifications shall consist of

. the Chairman or his designated alternate and at least four but not less than one-half of. the NSRC members including alternates. No more than a ainority of the quorum shall have line responsibility for operation of the unit.

h l l

l PERRY - UNIT 1- 6-11

.- - - - - - . . - - - - - -- ..- ~ - - - - - - -

/ "-- . ,

g.

'PY-CEI/NRR-1214 L

^ ' Attachm nt 3 a i: *

. Pag 23 of 26 y

j -ADMINISTRATIVE CONTROLS I ?; .. . ;

H@ = AUDITS (Continued)

~

ll d. The performance of activities required by the Operational Quality l ~ Assurance Program to meet the criteria of Appendix 8,10 CFR Part L

t-50, at least once per 24 months; i

g e'. . The fire protection programmatic controls including.the implementing procedures at least= once per 24 months by qualified licensee QA-

. personnel; -

3

f. The fire protection equipment and program implementation at least-once per 12 months utilizing either a qualified corporate licensee

.I fire protection enginnr(s) or an outside independent fire protection ;

consultant. An outsi independent fire protection consultant shall be utilized at least t M third year;

g. The radiological environmi.._

monitoring program and the results thereof at least once per 12 k 'hs; l

h. The OFFSITE DOSE CALCULATION MANUAL. and implementing procedures at I least once per 24 months;

1. The PROCESS CONTROL PROGRAN and implementing procedures at least once per 24 months;-

( j. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per

. .12. months; and -

k. ~Any other area of unit operation considered appropriate by the NSRC

.or the Vice President - Nuclear M ,

um01 -

% 6.5.2.9 ' Records of NSRC activities shall be prepared, approved, and~

distributed as indicated below; a.. Minutes of each NSRC meeting shall be prepared, approved, and

' forwarded to the Vice President - Nuclear within 14 days following each meeting. .

b. Reports of reviews encompassed by Specification 6.5.2.7 shall be prepared, approved, and forwarded to the Vice President - Nuclear

% T:q within 14 days following completion of the review.

c. - Audit r? ports ucompassed by Specification 6.5.2.8 shall be forwarded to the Vice President - Nuclear to the management positions a responsible for the areas audited within days after completion of .i y the audit by the auditing organization.

l y

j l

1 PERRY - UNIT 1 6-13 1

a n.

-l A : .v w l PY-CEI/NRR- 1214L

, 3 '.i Attachment-3 Page 24 of 26

ADM!NISTRATIVE CONTROLS 1

6. 5.3' TECHNICAL REVIEW AND CONTROL l ACTIVITIES 6.5.3.1 Activities which affect nuclear safety shall be conducted as follows;

a. Procedures / instructions required by Specification 6.8 and other procedures / instructions which affect plant-nuclear safety, &it !

changes thereto, shall be prepared, reviewed and approved. Each !

,such procedure / instruction or procedure / instruction change shall be reviewed by a qualified individual (s) other.than the individual (s) which prepared the procedure / instruction or pcocedure/ instruction change, but who may be from the same section as the individual (s) j which prepared the procedure / instruction or procedure / instruction change. Instructions shall be approved by appropriate management i i

g Sec.tlort Jappropriatelmanager@r , "!rt 2=::' personnel as designated by the in writing by.j T r t:- ":tt s General '

Manager, {^^;, ;c4 r. "L .. x., ^^*" shall approve ndministrative Procedurer.1.gppg ,

b. - Proposed modifications to plant structures, systems and components '

that affect nuclear safety shall be reviewed by individuals desig- ,

, . h"]jated oy the Director,4Nelear modification shall be raviewed Engineering Department.

by a qualified individual Each such (s) other [

4 than the individual (s) which designed the modification, but who ' -

may be from the same section as the individual (s) which designed h

the modif uations. Proposed modifications to plant structures, systems and components that affect nuclear safety shall be reviewed by PORC and approved prior to implementation by the General " '

Hanager,

""^"l"i:==: . ""*" Q ,

PPD

c. Proposed ~ tests na.d experiments which ' affect plant nuclear safety shall be prepared, reviewed, and approved. Each such test or. a experiment shall be reviewed by a qualified individual (s) other ;

than the individual (s) which prepared the proposed test or experi-ment. Proposed tests and experiments shall be approved before implementation by e General Manager, ^^^^l?irxt:r ""'" g l

hP l, d.

Sections responsible for reviews, including cross?b -disciplinary j reviews, performed in accordance with Specifications 6.5.3.la. and ,

i

%pgg the General Manager,R"^^. ;c th; Cf nt6.5.3.1c. --EEE , shall be designated in writina

r

; i t.

The individual (s) performing the review shall me+ ar exceed the l 3 b ;

D

' qualification requirements of appropriate sectionts) of ANSI .118.1-

, 1971;

? e. Each review shall include a determination pursuant to 10 CFR 50.59 of i ,

L whether or not the potential for an unreviewed safety question exists.

If such a potential does exist, a safety evaluation per 10 CFR 50.59 gg W to determine whether or not an unreviewed safety question is involved shall be performed. Pursuant to 10 CFR 50.59, NRC approval of items involving unreviewed safety questions shall be obtained prior to implementation; and PERRY - UNIT 1 6-14 Amendment No. 22 a .

% .7* f. .

~

PY-CEI/NRR-1214L

..* ' Attachmant 3 Page 25 of 26 ADMINISTRATIVE CONTROLS th* ACTIVITIES ' (Continued)

f. The Plant-Security Plan and Emergency Plan, and implementing instruc-tions, shall be reviewed at least once per 12 months. ":: n ::::: a >

D

h: ;:: :: '2: ' ;'. r ::t' ; ' :tr.:t'. :- :h:'. ' i: :;; :.:: t; r..

"^;......,T...'.... ~.,,_..^.,..'...^.,..^;....,^...,":....

I

_ y M ' :' ?^^: t. ::~c ; :;r':t;.- Recommended changes to the P' ansD$ hall, be reviewed pursuant to the requirements of S_pecifica-h.

( v o ^,on ti 6.5.1.6 and approved

. ._.,. _. . ._ _- OepartmenQ.

by the General Manager,

- - -

Perry &PlantNdedM 3 ;;;;;;': :. NRC approval shall be obtained, as appropriate.

6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and a report submitted pursuant to the. requirements of Section 50.73 to 10 CFR Part 50,'and

b. Each REPORTABLE EVENT shall be reviewed by the PORC and the results -

N D ofcr;.

C the review submitted to the NSRC and the Vice President - NucleaQ

( 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following act; .:s shall be taken in the event a Safety Limit is violated: .

4. The NRC Operations Center shall be notified by telephone as-soon as.

possible and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months - The Vice President.-

Nuclep and the NSRC shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .. j

b. A Safety Limit Violation Report shall be prepared The report shall be reviewed by the PORC. Thisreportshalldescrkbe-(1) applicable circumstances preceding the violation, (2)_ effects of the violation upon unit components, systems, or structures, and (3) corrective action taken to prevent recurrence. -

c. The Safety Limit Violation Report shall be submitted to the Commission, '

NS C, and the Vfce President - Nuclear n 30 days of the Critical operation of the unit shall not be resumed until authorized c d.

by the Connission.

6.8 PROCEDURES / INSTRUCTIONS AND PROGRAMS l

6.8.1 Written procedures / instructions shall be established, implemented, and

p. maintained covering the activities referenced below:

Q$

a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33. Revision 2, February 1978.

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6.8- -PROCEDURES / INSTRUCTIONS ANO PROGRAMS (Continued) ,

b. The appliceble procedures required to implement the requirements; of NUREG-0737 and supplements thereto.

c. Security Plan implementation.

dL Emergency Plan impisaentation. .

- e. PROCESS CONTROL PROGRAM implementation.

f. OFFSITE DOSE CALCULATION MANUAL implementation,

g. Radiological Environmental Monitoring Program implementation.

.h . Fire Protection Program implementation.

6.8.2 Each administrative procedure of Specification 6.8.1, and changes-

-thereto. shall be reviewed by the PORC nd shall be approved.by the General.

hPpb'. J Manager,* PP00, ; .i 2; Sir;n:r. ""? prior to implementation. All-

' procedures / instructions shall be reviewed periodically as ret forth in .

administrative procedures.

, . - 6.8.3 Temporary chances. Temporary changes to procedures / instructions which'

'do not,enange the intent of the approved procedures / instructions shall be

+{?-

app *oved ifor implementation by two members of the plant management staff, at -

least one of whom holds a Senior Operator license. These temporary changes

!shall be: documented. The temporary changes shall. be approved by the original-approval authority within-14 days. For changes to procedures / instructions

-which may involve a change in intent of the pr: 'ures/instructionsithe original approval authority shall approve the .. ..ga prior _to implementation.

6.8.4 The following programs shall be established, implemented, and maintained:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside '

containment that could contain highly radioactive fluids during.a serious transient or. accident to as'iow as practical levels. The systems include the HPCS, RHR, RCIC, LPCS, feedwater leakage control -

system, the hydrogen analyzer portion of Combustible Gas Control, and post-accident sampling systems. The program shall include the following:

1. . Preventive maintenance and periodic visual inspection requirements, and

2. Integrated leak test iequirements for each system at refueling cycle intervals or less.

I V

PERRY - UNIT I 6-16 Amendment No. 22

ML20064A831

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