ML20113D823

Proposed Changes to Tech Specs Discussed During 850117 Meeting

ML20113D823
Person / Time
Site:	Wolf Creek
Issue date:	01/18/1985
From:	STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM
To:
Shared Package
ML20113D788	List: ML20113D783 ML20113D823
References
NUDOCS 8501230241
Download: ML20113D823 (18)

Text

'

T TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SEIPOINTS n=

SENSOR A

TOIAL ERROR

,',7 FUNCTIONAL UNIT ALIOWANCE (TA)

Z (S)

TRIP SEIP0lNT ALLOWABLE VALUE e

12.

Reactor Coolant Flow-tow 2.5 1.77 0.6

>90% of loop

>89.2% of loop 5

-A design flow **

Uesign flow"*

6'

13. Steam Generator Water 23.5 21.18 2.51

>23.5% of narrow

>22.3% of narrow Level low-low range instrument range instrument span span 14.

Undervoltage - Reactor

7. 5 1.3 0

210578 volts A.C.

110355 Volts A.C.

Coolant Pumps 15.

Underfrequency - Reactor 3.3 0

0 157.2 Hz 157.1 Hz m

Coolant Pumps

. ed 16.

Turbine Trip m

$<io, oo 534. A uma.a a.

' Low Fluid Oil Pressure N.A.

N.A.

N.A.

1504.02 psig 1534.75 psig g

b.

Turbina Stop Valve N.A.

N.A.

N.A.

31% open

->1% open Closure P

17.

Safety Injection Input N.A.

N.A.

N.A.

N.A.

N.A.

D from ESF "AD H

$U*EEEA*o*@g PDR

s REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three baron injection flow paths shall be OPERABLE:

The flow path from the Boric Acid Storage System via a boric acid a.

transfer pump and a centrifugal charging pump to the Reactor Coolant System, and b.

Two flow paths from the refueling water storage tank via centrifugal charging pumps to the Reactor Coolant System.

APPLICABILITY:

MODES 1, 2, and 3.*

ACTION:

With only one of the above required boron injection flow paths to the Reactor -

Ccolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANOBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% ak/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REOUIREMENTS 4.1.2.2 'At least two of the above required flow paths shall be demonstrated OPERABLE:

At least once per 31 days by verifying that each valve (manual, a.

power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position; b.

At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal; and

'At least once per 18 months by verifying that the flow path required c.

by Specification 3.1.2.2a. delivers at least 30 gpm to the Reactor Coolant System.

• The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERA 8LE status within 4 nours prior to the temperature of one or more of the RCS cold legs exceeding 375'F Wn*& # 85 0 " C-of i

WOLF CREEK - UNIT 1 3/4 1-8

..st

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REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two centrifugal charging pumps shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.*

ACTION:

With only one centrifugal charging pump OPEF.ABLE, restore at least two cen-trifugal charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and barated to a SHUTDOWN MARGIN equivalent to at least 1% ak/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

i SURVEILLANCE REOUIREMENTS 4.1.2.4 At least two centrifugal charging pumos shall be demonstrated OPERABLE by verifying, on recirculation flow, that the pump develops a differential pressure of greater than or equal to 2400 psid when tested pursuant to Specification 4.0.5.

• The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provi he centrifugal charging pump is restored to OPERABLE status within 4 ours rio to the temperpture of one or more of the RCS cold legs exceeding,375*F, A c-

& u '**5 m t-OY WOLF CREEK

' UNIT 1 3/4 1-10

TABLE 3.3-4 (Continued) g ENGINEERED SAFETY FEMURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

ng

' TOTAL SENSOR TRIP p

FUNCTIONAL UNIT ALLOW.NCE (IA)

Z ERROR (S)

SETPOINf.

ALLOWABLE VALUE

6. Auxiliary Feedwater (Continued) zU

2) -Start Turbine-N Driven Pumps 23.5 21.18 2.51 1 23.5% of 1 22.3% of narrow range narrow range instrument instrument e.

Safety Injection -

span span Start Motor-Driven Pumps

.See Item 1. above for all Safety injection Trip Setpoints and Allowable Values.

f.

Loss-of-Offsite Power-y Start Turbine-Driven Pump N.A.

N.A.

N.A.

N.A.

N.A.

[

g.

Trip of All Main Feed-m water Pumps - Start Motor-Driven Pumps N.A.

N.A.

N.A.

N.A.

N.A.

h.

Auxiliary Feedwater q-g Pump Suction Pressure-P ai.so e2 tow (Transfer to ESW)

N. A.

N.A.

N.A.

1 MrS6 psia 1 20.53 psia b

7. Automatic Switchover to Containment Sump D

a.

Automatic Actuation Logic and Actuation N

lM!g;g Relays (SSPS)

N.A.

N.A.

N.A.

N.A.

N.A.

-.g.g

'b.

RWST Level-Low-Low 3.4 1.21 1.86

> 36% of

> 35.1% of

^*!

Instrument Tnstrument Coincident with span span Safety Injection See Item 1. above for Safety Injection Trip Setpoints and Allowable Values.

TABLE 3.3-4 (Continued) 5 ENGINEERF0 SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

ng TOTAL SENSOR TRIP ALLOWABLE g

FUNCTIONAL UNIT ALLOWANCE (TA)

Z ERROR (S)

SETPOINT VALUE

[

8. Loss of Power

=U a.

4 kV Undervoltage H

-Loss of Voltage N.A.

N.A.

N.A.

1 83V (120V 1 74.7V (120V Bus)

Bus) w/Is w/1 + 0.2, -0.Ss delay delay b.

4 kV Undervoltage

-Grid Degraded

, os.9 y soa. 3 v' Voltage N.A.

N.A.

N.A.

> IG M V 1 10 W5V (120V Bus)

T120V Bus) w/119 1 11.6s delay w/119s delay

9. Control Room Isolation

,s a.

Manual Initiation N.A.

N.A.

N.A.

N.A.

N.A O

b.

Automatic Actuation Logic and Actuation Relays (SSPS)

N.A.

N.A.

N.A.

N.A.

N.A c.

Automatic Actuation M

Logic and Actuation Q

Relays (BOP ESFAS)

N.A.

N.A.

N.A.

N.A.

N.A.

g j

d.

Phase "A" Isolation See Item 3.a. above for all Phase "A" Isolation Trip Setpoints and Allowable Values.

D

10. Solid-State Load Sequencer N.A.

N.A.

N. A.,

N.A.

N.A.

g

11. Engineered Safety Features Actuation

~

I System Interlocks H

a.

Pressurizer Pressure, P-11 N.A.

N.A.

N.A.

5 1970 psig 5 1979 psig b.

Reactor Trip, P-4 N.A.

N.A.

N.A.

N.A.

N.A.

p)n m's 1 ",'

n3 e

=

k i ){

N "4 3 33 6 #J s J

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGhAL AND FUNCTION RESPONSE TIME IN SECON05 3.

Pressurizer Pressure-Low a.

Safety Injection (ECCS) i 29(1)/12( )

1)

Reactor Trip i2 2)

Feedwater Isolation 17

$)

Phase "A" Isolation i 2(5) 4)

Auxiliary Feedwater 1 60 5)

Essential Service Water 1 60(1) 6)

Containment Cooling 1 60(1) 7)

Component Cooling Water N.A.

8)

Emergency Diesel Generators i 14(6) 9)

Turbine Trip N.A.

4.

Steam Line Pressure-Low a.

Safety Injection (ECCS) i 24(3)/12(4) 1)

Reactor Trip i2 2)

Feedwater Isolation

<7 3)

Phase "A" Isolation 2(5) 4)

Auxiliary Feedwater 1 60 5)

Essential Service W

< 60(1) 6)

Containment Cool ng "....

60(1) 7)

Component: Cooling Water N.A.

8)

Emergency Diesel Generators i 14(6) 9)

Turbine Trip N.A.

'b.

Steam Line Isolation i A a")

1 WOLF CREEK - UNIT 1.

3/4 3-30 L..

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TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 5.

Containment Pressure-High-3 a.

Containment Spray 5 32(1)/20(2) b.

Phase "B" Isolation 1 31.5 6.

Containment Pressore-High-2 Steam Line Isolation 5 % aI 7.

Steam Line Pressure-Negative Rate-Hicn Steam Line Isolation 5 i A "I 8.

Steam Generator Water Level-High-High a.

Turbine Trip 52.5 b.

Feedwater Isolation 57 9.

Steam Generator Water Level - Low-Low Start Motor-Driven Auxiliary a.

Feedwater Pumps 5 60 b.

Start Turbine-Oriven Auxiliary Feedwater Pumps 1 60 10.

Loss-of-Offsite Power

. Start Turbine-Driven Auxiliary Feedwater N.A.

Pumps 11.

Trip of All Main Feedwater Pumps Start Motor-Driven N.A Auxiliary Feedwater Pumps

't 1

WOLF CREEK - UNIT 1 3/4 3-31 c

Specification:

Table 3.3-5, items 4b, 6, 7 Justification:

The response times for steamline isolation were changed from 7 seconds to 2 seconds with a note reflecting the fact that the time does not include valve closure time.

This change is required for the sane reasons that necessitated deleting the testing requirements of speci-fication 3.6.3 for the main steam isolation valves (MSIV's); namely these valves cannot be stroke tested prior to entering mode 3 if an accurate evaluation of the stroke time is desired.

Note that the applicable modes for steamline isolation in Table 4.3-2 are 1, 2, and 3 (and specification 3.6.3 is applicable in modes 1, 2, 3, and 4). The MSIV vendor has stated that a reliable determination of stroke time cannot be obtained until normal operating tenperature and pressure are reached in the steam generator, and these conditions are not achievable until mode 3.

The testing of the circuitry necessary for steam line isolation will still be accomplished in accordance with the requirenents of specification 3.3.2 prior to entry into mode 3.

The MSIV's are stroke time tested in accordance with specification 3.7.1.5 and Wolf Creek's ASME section XI Pump and Valve Program (as -

invoked by specification 4.0.5).

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' ' L" REACTOR COOLANT SYSTEM OVERPRESSURE _ PROTECTION SYSTEMS LIMITING CONDITION FCR OPERATION 3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:

Two residual heat removal (RHR) suction relief valves each w a.

Setpoint of 450 psig 3%, or b.

Two power operated relief valves (PORVs) with Setpoints which do not exceed the limit established in Figure 3.4-4, or The Reactor Coolant System (RCS) depressurized with an RCS vent of c.

greater than or equal to 2 square inches.

APPLICABILITY:

MODE 3 when the temoerature of any RCS cold leg is less than or equal to 368 F, MODES 4 and 5, and MODE 6 with the reactor vessel head on.

ACTION:

own!

ud od inere m bl -

a.

With h::

t tha t gPORV.or t.we RHR suction relief valvet 9MMM9tE, 4

either restore two PORVs or two RHR suction relief valves to O status within 7 days or depressurize and vent the RCS through at least a 2 sauare inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, b.

With both PORVs and both RHR suction relief valves inoperable, depressurize and vent the RCS through at least a 2 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

In the event the PORVs, or the RHR suction relief valves, or the RCS-c.

vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days._ The report shall describe the circumstances initiating the transient, the effect of the PORVs or the RHR suction relief valves any corrective action necessar,y to prevent recurrence.or RCS vent (s) d.

- The provisions of Specification 3.0.4 are not. applicable.

WOLF CREEK - UNIT'l' 3/4 4-34 N

i- ~-

1 EMERGENCY CORE COOLING SYSTEMS J

3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350'F LIMITING CONDITION FOR OPERATION i

3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

One OPERABLE centrifugal charging pump, a.

b.

One OPERABLE Safety Injection pump, One OPERA 8LE RHR heat exchanger, c.

d.

One OPERABLE RHR pump, and An OPERABLE flow path capable of taking suction trom the refueling e.

water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY:

MODES 1, 2, and 3.*

ACTION:

With one ECCS subsystem inoperable, restore the inoperable subsystem a.

to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at-least HOT STANOBY within the ' text 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and_ submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

' *The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pumps and the Safety-Infection pumps declared inoperable pursuant to Specification 4.5.3.2 provided the centrifugal charging pumps and the Safety Injection pumps are restored to OPERABLE status within 4. hours rior to the temperature of one or more of the RCS cold legs exceeding 375'F, ecAce. me5 b t.

01' WOLF CREEK

-UNIT l' 3/4 5-3

- - -- ' 1 - -- -----

(f.?

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• P ~ f E;%% Ltdn. ;f b EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) h.

By performing a flow balance test, during shutdown, following com-pletion of modifications to the ECCS subsystems that alter'the subsystem flow characteristics and verifying that:

1)

For centrifugal charging pump lines, with a single pump running:

a)

The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 346 gpm, and b)

The total pump flow rate is less than or equal to-G&& gpm.

ssu 2)

For Safety Injection pump lines, with a single pump running:

a)

The sum of the injection ilne flow rates, excluding the hignest flow rate, is greater than or equal to akse gpm, anc 454 sua b)

The total pump ficw rate is less than or equal to 66G gpm.

i By performing a flow test, during shutdown, following completion of modifications to the RHR subsystems that alter the suesystem flow characteristics and verifying that the RHR puma lines, with a single pump running:

1)

.The sum of the injection line flow rates is greater than or equal to 3800 gpm, and 2)

The total pump flow rate is less than or equal to 5500 gpm.

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BNAL*1RAb" EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2.

4.5.3.2 All centrifugal char above allowed OPERABLE pumps,ging pumps and Safety Injection pumps, except the shall be demonstrated inoperable" by verifying that the motor circuit breakers are secured in the open position within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, after entering MODE 4 from MODE 3 prior to the temperature of one or more of the RCS cold legs decreasing bei therea f ter.

325 F and at least once per 31 days Jh a.04.c < c we.s f. <2 t.,

"An. inoperable pump may be energized for testing or for filling accumulators pr videc the cisenarge of tne pump nas been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.

WOLF CREEK - UNIT 1 3/4 5-8

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.m EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A minimum contained borated water volume of 394,000 gallons, b.

A boron concentration of between 2000 and 2100 ppm of boron, A minimum solution temperature of 37*F, and c.

d.

A maximum solution temperature of 100 F.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

.With the RWST inoperable, restore the. tank to OPERA 8LE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY withing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN within the following 30-hours.

(ge nest SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

At least once per 7 days by:

a.

E1)

Verifying the contained borated water volume in th'e tank, and 2)

' Verifying the boron concentration of the water.

b.

At least onc2 par 2?'heur: y varifying the RW37 tempera ure wnen

~

the outside air temperature is either less than 37*F or greater than 100 F.

S

-,n-

l TABLE 3.6-1 (Continued)

CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQ 9 IRED (Seconds) 8.

Hand-0perated and Check Valves - (Continued)

P-98 KB V-002 Breathing Air Supply C

N.A.

to Rx Bldg.

P-6/

KC V-478 Fire Protection C

N.A.

Supply to RX Bldg P-5/

SJ V-111 Liquid Sample from A,C N.A.

PASS to RCDT 9.

Other Automatic Valves P-1 AB-HV-11***

Mn. Stm. Isol.

A 5t/A P-2 A'l-HV-14 * *

• Mn. Stm. Isol.

A I do P-3 AB-HV-17***

Mn. Stm. Isol.

A 5 NA P-4 AB-HV-20***

Mn. Stm. Isol.

A 5NA P-5 AE-FV-42***

Mn. FW Isol.

A FN4 P-6 AE-FV-39***

Mn. FW Isol.

A 5 #4 P-7 AE-FV-40* *

• Mn. FW Isol.

A 5 d4

-P-8 AE-FV-41***

Mn. FW Isol.

A g NA

-P-9 BM-HV-4**

SG Blowdn. Isol.

A 10 P-10 BM-HV-l**

SG Blowdn. Isol.

A 10 P-11 BM-HV-2**

SG Blowdn. Isol.

A 10 P-12 BM-HV-3**

SG Blowdn. Isol.

A 10

.,.. ; s io,,, e f--Speei f tration-3:0: 4-are not -appHcable.

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we m.se<etene s. wme,-u a spa.V u t oo s u. 3 do net. o ppi, insted the. <cevements of s p.u.4,u t..:,w y

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WO CREEK COOLING L AKE Meteorological Tower 1.

The exclusion-restricted H

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CirCie centered around Unit 1 6

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M A,N D AM F AS 10 8j J

FINAL DRAFT ADMINISTRATIVE CONTRO S FUNCTION (Continued) g.

Mechanical and electrical engineering, and h.

Quality assurance practices.

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.

COMPOSITION 6.5.2.2 The NSRC shall be composed of M : et lent the fc ll-ap Chairman:

Manager Nuclear Services Member:

Manager Nuclear P? ant Engineering Member:

Manager Quality Assurance (Home Office)

Member Director Nuclear Operations Member:

Mant.ger Licensing =f Radiclegical Suv.c.e s Meraber:

Vice President-Engir.cering Mertber:

Manager Nuclear Safety

@ Son.I m bus and v',ce, g,we n % k..pp nte d ny the Cmac%n.

ALTERNATES a

6.5.2.3 All alternate members shall be appointed in writing by tne NSRC 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.

MEETING 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 a/3 a( N.

6.5.2.6 The quorum of the NSRC necessary for the performanc of the NSRC review and audit functions of these Technical Specification shall consist of the Chairman or his designated alternate and at least fette-NSRC members including alternates.

No more than a minority of the quorum shall have line resnonsibility for operation of the Unit.

WOLF CREEK - UNIT 1 6-10

L H%LDWT ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued)

Technician) or personnel continuously escorted by i'

exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less i

than 1000 mR/h, provided they are otherwise following plant radiation r ote

.tfon procedures for entry into such high radiation areas.

accompanied by one or more of the following: group of individ Any individual or A radiation monitoring device which continuously indicates a.

the radiation dose rate in the area, or b.

~

A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them, or An individual qualified in radiation protection c.

procedures with a radiation dose rate monitoring device, who is responsible for providing positive

~ *e control over the activities within the area and j o?

shall perform periodic radiation surveillance at t

the frequency specified by the Site Health Physicist

--J l.

in the RWP.

6.12.2 In addition to the requirements of Specification 6.12.1, to personnel with radiation levels greater than 1000 mR/h at 45 rea accessible the radiation source or from any surface which the radiation pe ett m(

in.) fro.

provided with locked doors to prevent unauthorized entry, and es shall maintained under the administrative control of the Shift Super isor on d e k ys shal e

and/or health physics supervision.

Doors shall remain locked e ring periods of access by personnel under an approved RWP which shall specify the d i

individuals in that area. rate levels in the immediate work areas and the maximu i

In lieu of the stay time specification of the RWP, be made by personnel' qualified in radiation protection positive. exposure control over the activities being performed within the area.

For individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located within large areas, such as PWR con-tainment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device.

WOLF CREEK - UNIT 1 6-23

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