Text

Proposed Tech Specs,Providing Addl Changes for Full Power OL
	ML20210A981
Person / Time
Site:	Clinton
Issue date:	02/04/1987
From:	; ILLINOIS POWER CO.
To:	;
Shared Package
ML20210A955	List: ML20210A981; U-600817, Application for Amend to License NPF-55,providing Addl Changes to Clarify & Enhance Tech Specs Which Will Accompany Full Power OL for Plant.No Amend of Present Low Power License Requested.Changes Consistent W/Fsar as Amended;
References
	NUDOCS 8702090051
	Download: ML20210A981 (46)
	v • d • e

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_ TABLE 1.2 jgy OPERATIONAL CONDITIONS MODE SWITCH CONDITION AVERAGE REACTOR POSITION COOLANT TEMPERATURE

1. POWER OPERATION Run Any temperature

2. STARTUP Startup/ Hot Standby Any temperature

3. HOT SHUTDOWN Shutdown #'*** > 200*F

4. COLD SHUTDOWN Shutdown #'N'***

1 200*F

5. REFUELING

Shutdown or Refue1**'# $ 140*F TABLE NOTATIONS

The reactor mode switch may be placed in.the Run or Startup/ Hot Standby position to test the switch interlock functions, provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

N The reactor mode switch may be placed in the Refuel position while a single control rod drive Specification is being removed from the reactor pressure vessel per 3.9.10.1.

Fuel in the reactor fully tensioned or withvessel the headwith the vessel head closure bolts less than removed.

- See Special Test Exceptions 3.10.1 and 3.10.3.

- - The reactor mode switch may be placed in the Refuel position while a single i

control rod is being r=ghd7 provided that the one-rod-out interlock is OPERABLE.

ed g 8702090051 870204 PDR ADOCK 05000461 P PDR CLINTON - UNIT 1 1-11

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- TABLE 1.2

- OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR

- CONDITION POSITION COOLANT TEMPERATURE

1. POWER OPERATION Run Any temperature

2. STARTUP Startup/ Hot Standby Any temperature

3. HOT SHUTDOWN ,

Shutdown # *** > 200'F

4. COLD SHUTDOWN Shutdown # ## ,*** 1 200*F

5. REFUELING

Shutdown or Refuel ** # $ 140*F

The reactor mode switch may be placed in the Run or Startup/ Hot Standby -

position to test the switch in,terlock functions provided that the control rods are verified to remain fully inserted by a second licensed operator or '

other technically qualified member of the unit technical staff.

1. The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.9.10.1.

Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

- See Special Test Exception 3.10.3

- - The reactor mode switch may be placed in the Refuel position while a single control rod is being moved provided that the one-rod-out interlock is OPERABLE.

LA SALLE - UNIT 2 1-9

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TABLE 1.2 OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE

1. POWER OPERATION Run Any temperature

2. STARTUP Startup/ Hot Standby Any temperature

3. HOT SHUTDOWN Shutdown #'*** > 200*F

4. COLD SHUTDOWN Shutdown #'##'*** < 200*F

5. REFUELING

Shutdown or Refuel **'# < 140'F i

The reactor mode switch may be placed in the Run or Startup/ Hot Standby position to test the switch interlock functions and related instrumentation provided that the control rods are verified to remain fully inserted by a l

second licensed operator or other technically qualified member of the unit j

technical staff.

1. The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.9.10.1.

Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

- See Special Test Eiceptions 3.10.1 and 3.10.3.

- - The reactor mode switch may be placed in the Refuel position while a single control rod is being recoupled or withdrawn provided that the one-rod-out interlock is OPERABLE.

RIVER BEND - UNIT 1 1-11 L

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TABLE 1.2 OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR POSITION COOLANT TEMPERATURE

. CONDITION

1. POWER OPERATION Run Any temperature

2. STARTUP Startup/ Hot Standby Any temperature

3. HOT SHUTDOWN Shutdown #'*** >.200'F

4. COLD SHUTDOWN Shutdown #'##'*** $ 200'F

5. REFUELING

Shutdown or Refuel ***# $ 140*F 1

c

The reactor mode switch may be placed in the Run or Startup/ Hot Standby l

position to test the switch interlock functions and related instrumentation provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit l

technical staff. -

l I

1. The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per f Specification 3.9.10.1.

" Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

- See Special Test Exceptions 3.10.1 and 3.10.3.

l

- - The reactor mode switch may be placed in the Refuel position while a single control rod is being recoupled or withdrawn provided that the one-rod-out l

i interlock is OPERABLE.

i HOPE CREEK 1-11

I s Description of Change Figure 3.2.3-2 page 3/4 2-9 Replace this figure.with a new figure 3.2.3-2 Justification This figure was drafted incorrectly in that the vertical lines should be at 40% and 70% of Core Power and not the as-drawn 38% and 68%.

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FOR ATs2'd*F C-C' ALL CORE A GE-EXPOSURES

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1.0 0/ 20 40 60 80 100 120 CORE POWER (% of Roted) l Figure 3.2.3-2 Clinton MCPRp VerSus Power l

l CLINTON - UNIT 1 3/4 2-9 '

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~' ALL CORE AVERAGE l EXPOSURES I 1.1 0 20 40, 60 80'

. 100: 120' CORE POWE5 (% of Ratea)l Cilnton MCPRp Versus Power; Figure 3.2.3 2

CLINTON-ONIT li -

--. 3/42-9i - - - - - - - - - - - -

I e Description of Change Table 4.3.6-1, page 3/4 3-68 Change surveillance frequencies as indicated.

Justification On June 23, 1986 the BWROG submitted (BWROG-8623) a report (NEDC-30851P Supplement 1, Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation). This report was a supplement to the technical specification improvement analyses conducted for the BWR Reactor Protection System (RPS). The evaluation provides a basis for improvements to surveillance test intervals in the control rod block instrumentation Technical Specification.

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TABLE 4.3.6-1 .

b CONTROL R0D BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS "i

3

. CHANNEL OPERATIONAL

c- CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH 55 TRIP FUNCTION CHECK TEST CALIBRATION (3) SURVEILLANCE REQUIRED

[ 1. R0D PATTERN CONTROL SYSTEM

a. Low Power Setpoint NA S/

) d)(e) g(f) 1, 2

b. RWL High Power Setpoint NA S/U(g) ) (f)

2. APRM

a. Flow Biased Neutron Flux -

. Upscale NA- S/U ,P q SA 1 j b. Inoperative NA q NA t, 2, 5

c. Downscale . NA S/U(b),V S/U 9 bA 1 o d. Neutron Flux - Upscale, Startup NA S/UIb)',MS q SA 2, 5 A 3. SOURCE RANGE MONITORS T a. Detector not full in NA S/U(b) W

, NA 2, 5 2, 5 8 b. Upscale NA S/U ,W SA l c. Inoperative NA S/U(b),W NA 2, 5

d. Downscale NA S/U ,W SA 2, 5

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in NA S/U(b) W

, NA 2, 5 I b. Upscale NA S/U ,W SA 2, 5 l c. Inoperative NA S/U(b),W NA 2, 5

d. Downscale NA S/U ,W SA 2, 5

5. SCRAM DISCHARGE VOLUME

a. Water Level-High S # c4 R II) 1, 2, 5*

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW ,

a. Upscale NA S/U(b) /, SA 1 ',

7. REACTOR MODE SWITCH

a. Shutdown Mode NA R NA 3, 4

b. Refuel Mode NA R NA 5

4

, e o Description of Change

^

Table 3.3.7.12-1, page 3/4 3-102 Change Table 3.3.7.12-1, Item 2.a. and b. ACTION from 126 to 121, i ..

Justification ACTION 126 is not consistent with the Illinois Power Company submittal or the NRC Standard Technical Specifications for Radiological Effluents (SRETS). As discussed with.NRC Messrs B. Siegel and W. Mienke, ACTION 126 is commonly used only with purge systems._ It is not the intent (as shown by i other licensed BWR's) of_the NRC to restrict the release from this pathway, but only to ensure the release is properly monitored. Therefore, ACTION 121

is correct and similar to the ACTION at other licensed BWR's (Grand Gulf, Susquehanna and LaSalle) and similar to other CPS-TS for monitoring instrumentation which monitors plant gaseous effluents i.e., Table 3.3.7.12-1, Item 1.a. and b.

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TABLE 3.3.7.12-1 i n C RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

! 'i E

MINIMUM INSTRUMENT CHANNELS OPERABLE APPLICABILITY ACTION i 5

1. Station HVAC Exhaust PRM

[ *

a. High-Ra,ge Noble Gas Activity I 121 l

1 Monitor

121

b. Low-Range Noble Gas Activity 1 Monitor

122

c. Iodine Sampler 1

122

d. Particulate Sampler 1

$ e. Sample Flow-Rate Measuring 1'

123 w Device O Effluent System Flow Rate 1

123 S f.

Measuring Device

2. Standby Gas Treatment System Exhaust PRM

a. Medium-Range Noble Gas 1

- M 12. l Activity Monitor

b. Low-Range Noble Gas 1

- g 12t Activity Monitor

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c. Deleted ,

2

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Description of Change Table 3.3.9-2, page 3/4 3-113 Change the Trip Setpoint and Allowable Value for Trip Function 1.b.,

Containment Pressure - High, to 22.3 psia and 23.4 psia respectively.

Justification During the latter part of CY 1982, the NRC staff requested several Operating-License applicants, with General Electric nuclear steam supply systems, to document the methodology used to establish the protective system actuation instrumentation setpoints in plant Technical Specifications.

Illinois Power Company's Clinton Power Station (CPS) participated in the Instrument Setpoint Methodology (ISM) program sponsored by the Boiling Water Reactors Owners Group (BWROG). The BWROG authorized General Electric Co. to perform a validation calculation of the Technical Specification operating limits. Based on the analysis performed, ISM calculations for the Residual Heat Removal (RHR) Containment Spray initiation instrumentation indicate that the loop accuracy is significantly affected by high environmental temperatures. The inaccuracy of the loop requires lowering the setpoint of the affected instruments, 1E12-N062 A-D and 1E12-N662 A-D, as specified in the marked-up Technical Specification.

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O N C' D TABLE 3.3.9-2 n

C; PLANT SYSTEMS ACTUATION INSTRUMENTATION SETPOINTS 5

E

, ALLOWABLE e TRIP FUNCTION , TRIP SETPOINT VALUE 4 5

1. CONTAINMENT SPRAY SYSTEM

]

a. Drywell Pressure-High < 1.68 psig < 1.883sig

b. Containment Pressure-High

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2d422,5 < M. g 22.4 l

c. Reactor Vessel Water Level-Low Low Low, level 1 5-145.5 in.*

I-147.7 in.

~

d. Timers -

1. Loop A, Loop B 10.17 min.

1 10.10 5 10.23 min.

2. Loop B only < 90 sec. < 90.6 sec.

t i 2. FEEDWATER SYSTEM / MAIN TURBINE' TRIP SYSTEM j w a. Reactor Vessel Water Level-High, Level 8 5 52.0 in.* 5 52.6 in.

1 w 3. SUPPRESSION POOL MAKEUP SYSTEM i O l 0 a. Drywell Pressure-High 5 1.68 psig 5 1.88 psig I

b. Reactor Vessel Water Level-Low Low Low, Level 1 1 -145.5 in.* > -147.7 in.

c. Suppression Pool Water Level-Low Low 1 37 9/16 in.** 2 29 in.**

d. Suppression Pool Makeup Timer 5 25 minutes 5 30 minutes

e. SPMS Manual Initiation NA NA

f. SPMS Mode Switch Permissive NA NA , e l .

*See Bases Figure B 3/4 3-1.

l ** Instrument zero is 727'-0" as1.

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Description of Change Specification 3/4.3.10, page 3/4 3-115 Amend note

to read:

"The STS may be periodically taken out of the fully automatic mode of operation for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the purpose of performing.

surveillance testing and . preventive or corrective maintenance to satisfy technical specification requirements for those components the STS is designed to monitor."

Justification The SELF TEST SYSTEM (STS) of the Nuclear System Protection System (NSPS) is i a testing and surveillance system designed to automatically and continuously monitor the NSPS. The NRC staff has approved the STS for performing surveillance testing. required by the Clinton Plant Technical Specifications (CPS-TS). Each of the four divisional NSPS control room panels contain a self-test controller (STC) consisting of a microprocessor executing firmware program designed to perform the NSPS circuit monitoring function of the STS.

Because the STC's are also used to augment the required Technical Specification'surveillances,-the STS NSPS monitoring function is interrupted while performing the surveillance. However, use of the STS is not. limited to performing surveillance requirements of the CPS-TS. The primary purpose of the STS is to improve the availability of the NSPS by optimizing the time to detect and determine the location of a failure (s) in the system. Failure locations are traced to the module or printed circuit (PC) card level, which is the established increment of field replacement. An STC stops its test sequence upon detection of a fault. Plant personnel can communicate with the STS through the plant computer via the diagnostic terminal. This allows technicians to quickly isolate a fault detected by the STS to the PC card level for replacement with a spare card. Upon replacement of a faulted card, STS is used es the primary piece of maintenance and test equipment that has the unique capability to test the replaced / repaired PC card once installed in the NSPS panels. In summary, the STS is used for functions other than surveillance testing which should not be restricted during normal plant operation.

Changing the time from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provides greater flexibility to perform detailed " trouble shooting" and preventive or corrective maintenance and surveillance testing prior to returning STS to automatic operation. In

-addition, the additional time provides greater flexibility in operation of the facility by allowing operator interaction / communication with the STS outside of the ACTION time limits of the Limiting Condition for Operation.

Describing the additional capabilities of performing preventive or corrective maintenance clarifies, to the operator, that the design purpose of STS is not restricted during normal operation. Since STS does not provide automatic initiation of systems used to mitigate the consequences of an accident or prevent / control the release of radioactivity to members of the public, there is no affect on the safety of operation of the plant.

INSTRUMENTATION 3/4.3.10 NUCLEAR SYSTEM PfiOTECTION SYSTEM - SELF TEST SYSTEM LIMITING CONDITION FOR OPERATION 3.3.10 The SELF TEST SYSTEM (STS) of the Nuclear. System Protection System shall be OPERABLE and operating in the fully automatic mode.* ,

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5.

ACTION:

a. Wit.h the.STS not operating in the fully automatic mode initiate corrective action within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to restore the STS to automatic operation for the maximum number of divisions available. -

} b. If the STS is not restored to fully automatic operation within 30 days, be in at leas't HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS:

~

4.3.10 Status indications of the STS shall be obtained at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , whenever the STS is operating in the fully or partially automatic mode.

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be periodically taken out of the fully automatic mode of operation f *The for up STS to may*

2 hours for the purpose of performing surveillance testingato satisfy technical specification requirements for those components the STS is designed to monitor. J

\ QQ prevedbe or correck CLINTON - UNIT 1 3/4 3-115 /+f441/mvane' I

l 1 - -_-- . , . . _ . - , _ _ - - , . . . _ . - - _ - - _ _ _ __ , _ _ , _ _ _ _ . _ - . . . _ , .

Description of Change Specification 4.4.1.1, page 3/4 4-2 Change paragraph number 4.4.1.1 to 4.4.1.1.1 Justification Typographical error

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS SURVEILLANCE REQUIREMENTS 4.4.1.131 Each reactor coolant system recirculation loop flow control valve l shall be demonstrated OPERABLE at least once per 18 months by:

a. Verifying that the control valve fails "as is" on loss of hydraulic pressure at the hydraulic control unit, and

b. Verifying that the average rate of control valve movement is:

1. Less than or equal to 11% of stroke per second opening, and

2. Less than or equal to 11% of stroke per second closing.

4.4.1.1.2 When total core flow is less than 45% of rated flow with two coolant system recirculation loops in operation and THERMAL POWER is greater than the limit specified in Figure 3.4.1.1-1, establish a baseline APRM and LPRM*

neutron flux noise value within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of entering this operating region unless baselining has previously been performed in the region since the last CORE ALTERATION, and

a. Determine the APRM and LPRM* noise levels at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

b. Determine the APRM and LPRM* noise levels within 30 minutes after the completion of a THERMAL POWER-increase of at least 5% of RATED THERMAL POWER.

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Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

CLINTON - UNIT 1 3/4 4-2 r .--

Description of Change Specification 4.4.4.c., Page 3/4 4-15 Delete "for up to 31 days" Justification Typographical error

REACTOR C00LhNT SYSTEM CHEMISTRY LIMITING CONDITION FOR OPERATION (Continued) s 3.4.4 ACTION (Continued):

2. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.4 The reactor. coolant shall be determined to be within the 'specified chemistry limit by:

a. Measurement prior to pressurizing the reactor during each startup, if not J

1 performed within the previous 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

b. Analyzing a sample of the reactor coolant for:

1. Chlorides at least once per:

a) 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and b) 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months whenever conductivity is greater than the limit in Table 3.4.4-1.

2. Conductivity at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3. pH at least once per:

a) 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and b) 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months whenever conductivity is greater than the limit in Table 3.4.4-1.

c. Continuously recording the conductivity of the reactor coolant, or, when*

the continuous recording conductivity monitor is inoperable fer up te

' 31 d:y:, obtaining an in-line conductivity measurement at least once per:

' 1. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months in OPERATIONAL CONDITIONS 1, 2, and 3, and

2. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at all other times.

d. Performance of a CHANNEL CHECK of the continuous conductivity monitor with an in-line flow cell at least once per:

1. ' 7 days, and

2. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months whenever conductivity is greater than the limit in

~

Table 3.4.4-1.

CLINTON - UNIT 1 3/4 4-15

- - , - -- ,w- - - - - - - - ,,,---gw--- _m yyy , -- ,,w ,, - , , ,_ ,,,,,e,-,, , - ,- , - -- - - , , - ,,,,, - , , - - - - --,,-,__,-,,w-gm-ew-

Description of Change Specification 4.6.2.2, page 3/4 6-15 Delete paragraph 4.6.2.2.b. and incorporate paragraph 4.6.2.2.a. into one paragraph 4.6.2.2.

Justification In accordance with discussions with the NRC's Mr. B. Siegel, the NRC has been granting exemptions and/or removing from plants' Technical Specifications the provisions of paragraph 4.6.2.2.b. The NRC requirement is not to require a plant shutdown to perform this test since it is not required by the regulations. This change is consistent with the NRC's policy and similar BWR Technical Specifications.

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CONTAINMENT SYSTEMS ~ #-

DRYWELL BYPASS LEAKAGE '

, +

1 LIMITING CONDITION FOR OPERATION v R -

3.6.2.2 Drywell bypass leakage shall be less than or equal'to'10% of the minimum acceptable A/ 8 design value of. 1.18 ft2, APPLICABILITY: When DRYWELL INTEGRITY is required per Specification '3.6.2.1.

ACTION:

With the drywell bypass leakage greater than 10% of the minimum acceptable ,

A/8 design value of 1.18 ft ,2restore the drywell bypass leakage to within the limit prior to increasing reactor coolant system terperature above 200'F. .'

~

j SURVEILLANCE REQUIREMENTS' 4.6.2.2 The drywell bypass leakage rate test shall be conducted at least once l per 18 months at an initial differential pressure of 3.0 psi and the A/8 shall be calculated from the measured leakage. One drywell airlock door shall remain open during the drywell leakage test such that each drywell door is leak tested during at least every other leakage rate testy U CIf any drywell bypass leakage test fails to meet the specified limit, the schedule for subsequent tests shall be reviewed and approved by the Com-mission. If two consecutive tests fail to meet the limit, a test shall be performed at least every 9 months until two consecutive tests meet the limit, at which time the 18 month test schedule may be resumed,

b. The pr;vi:i;n: Of Sp;;fff;;ti;n 4.0.2 ;r; n;t ;ppli;;bi;. #

9 F

I CLINTON - UNIT 1 3/4 6-15

4- .

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De,sdriptioti of Change

a. - _,: ,

SpLcification 4.6.3.1.c.3, page 3/4 6-25 a :he

Mov'e'"with the water high temperature alarm setpoint set for 1.43*F" right and align properly with item c.3 indentation'.

Justification Typographical' error, s

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CONTAINMENT SYSTEMS SUPPRESSION POOL SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.1 (Continued) ,

c. By verifying sixteen suppression pool water temperature instrume'ntation channels, at least two channels in each suppression pool sector, OPERABLE by performance of a:

1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

2. CHANNEL FUNCTIONAL 1EST at least once per 31 days, and .

3. ' CHANNEL CALIBRATION at least once per 18 months, -

with the water high temperature alarm setpoint set for < 93*F. -]pr l

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, CLINTON - UNIT 1 3/4 6-25 l

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- Description of Change Table 3.6.4-1, page 3/4 6-32, Item 13), valve 1E51-F031 and page 3/4 6-33, Item 18), valve IE51-F064.

Add

to Isolation Signal B and R and note

to the bottom of the page as follows:

A single manual isolation switch (R) isolates outboard steam supply line isolation valve (F064) and the RCIC putop suction from suppression pool valve (F031) only following a manual or automatic Reactor Vessel Water Level 2 (B) RCIC system initiation.

Justification This change provides clarification and consistency between Table 3.3.2-1, Table Notations, note 1) on page 3/4 3-18 and Table 3.6 4-1.

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TABLE 3.6.4-1 (Continued) ,

n CONTAINMENT ISOLATION VALVES -

R "i

@ MAXIMUM SECONDARY

, APPLICABLE ISOLATION CONTAINMENT TEST c- VALVE PENETRATION ISOLATION OPERATIONAL TIME BYPASS PATH PRESSURE

?! NUMBER NUMBER SIGNAlt CONDITIONS (Seconds) (Yes/No) (psig)*

[AutomaticIsolationValves(Continued)

7) RHR Shutdown Cooling 14 ' 1, 2, 3 No 9. 0 1E12-F008 A,S,T,X,R 54 IE12-F009 A,S,T,X,R 54

8) RHR A To Fuel Pool Cooling 15 1,2,3 No' 9.0 1E12-F037A A,S,T,L R 95

9) RHR B To Fuel Pool Cooling . 16' 1, 2, 3 No 9.0 w 1E12-F0378 A,S,T,L R 95 i i m 10) RHR A/LPCS Test Line 18 1,2,3 No 9.9 J, IE12-IO24A L, U 117

" 1E12-F011A L, U 33 1E21-F012 L, U 90

11) RHR C Test Line 19 1,2,3 No 9.9 IE12-F021 L, U 123

12) RHR B Test Line 20 1, 2, 3 No 9.9 -

1E12-F0248 L, U 117 1E12-F0118 L, U 33

13) RCIC Suction 28 1,2,3 No 9.9 1E51-F031 V,S,T,X,( ( 48 E, F

' l

14) HPCS Test Line 33 1,2,3 No 9.9 IE22-F023 B, L 68

'

isc isotaan

' A Vasingle maisua/

/s/e ( F064.) a Adisotafan

/h e RCZC swife4 pump cx)

.:Tu hotate.s oaRward cfinn -/r' em .:et/f NS$sfea ri supply son pc/ ya /ve (Fa31) on/y

, fo//owiny a man aat or aa famahh Reaclor Vesse/ loafer Level 2 ( B.) RC.IC sys/em i,siiia fier. ,

TABLE 3.6.4-1 (Continued) c .

n .

CONTAINMENT ISOLATION VALVES

.I C

'i l E MAXIMUM SECONDARY s APPLICABLE ISOLATION CONTAINMENT TEST i c VALVE PENETRATION ISOLATION OPERATIONAL TIME BYPASS PATH PRESSURE y NUMBER NUMBER SIGNALt CONDITIONS (Seconds) (YES/NO) (psig)*

l Automatic Isolation Valves (Continued)

15) Supp. Pool Cleanup Suction 34 1,2,3 Yes 9.9 ISF004 B. L. R 84

16) RCIC 41 1, 2, 3 No 9.0 1E51-F077 L, V# , 21

.i 1 17) RHR Head Spray 42 1,2,3 No 9.0 1E12-F023 . A,S,T,X,R 39

, R

18) RCIC Steam Supply 43 1,2,3 No 9. 0 IE51-F063 V,S,T,E F 41 i

1 i 1E51-F064 V,S,T,Ri d,XE,F,X 41 l

O 1E51-F076 V,S,T,E,F,X 8

19) RCIC Turb Vac Bkr Line 44 1, 2, 3 No 9.0 1E51-F078 L, V,,, 27

20) Main Steam Drain Line 45 1,2,3,#(II 'Yes 9.0 IB21-F016 C,D,E,G,H,J, 26 U, X, F, R i 1321-F019 C,D,E,G,H,J, 26 U, X, F, R A Single manual iso /ab swHeh (R) isolafes ou:lboard afeam supply Ade isolafion un /v c (R%1) and & Rczc pump .sucRon {' rom suppresubn pos/

va IVe ( F031 on fy fol/ocoing a manua/ or aulamaNe /?ea c{oe Veze/ Waler Level 2 (B))RC.Zc sy.s/cm iniNa/dn.

Description of Change

' Table 3.6.4-1, Item 67), page 3/4 6-51.

1 Add note (a) to the APPLICABLE OPERATIONAL CONDITIONS column.

L Justification l

Typographical error.

1 i-i 1

6 4

1 J

e

'l i

TABLE 3.6.4-1 (Continued) n CONTAINMENT ISOLATION VALVES G

t

$'i . MAXIMUM SECONDARY APPLICABLE ISOLATION CONTAINMENT TEST c VALVE PENETRATION ISOLATION OPERATIONAL TIME BYPASS PATH PRESSURE 5

-i NUMBER NUMBER SIGNAtt CONDITIONS (Seconds) (YES/NO) (psig)*

H Test Connections, Vents and Drains (Continued)

! 64) Drywell Pressure NA 1,2,3(a) NA 2

1CM076 No 9. 0 151 ICM077 203 l

65) Reactor Pressure 151 NA 1,2,3(a) NA No 9.0 ICM072 .

1CM073 g 66) Reactor Pressure 160 NA 1,2,3(a) NA No 9.0

ICM074 T ICM075

= (M

67) Equipment Hatch 1 NA 1,2,3 NA-l No 9. 0 I ICM099

4. Other Isolaticn Valves i

i

~

1) Main Steam Line C 5 1, 2, 3 NA No 9. 0 IE32-F001J NA 1821-F098C(c) gg i

l e

i

Description of Change Table 3.8.4.2'1, page 3/4 8-49 Add valve 1E51-C002E to this table.

Justification The line circuit for valve 1E51-C002E contains a thermal overload (TOL) bypass device and is safety related. The TOL is continuously bypassed in both the open and closed directions.

1 4

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l

-.-r -- - ,-,,e w -~ . , - .e., - , . .

r--- --

r - - - - .- - . ---- , wr--, -- - - - . - - , n- - -

i TABLE 3.8.4.2-1 (Continued)

MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION VALVE NO. BYPASS DIRECTION SYSTEM (S)AFFfCTED IE51-F059 Continuous Open/Close Reac Core Isol Cool IE51-F063 Continuous Open/Close Reac Core Isol Cool IE51-F064 Continuous Open/Close -

Reac Core Isol Cool IE51-F068 Continuous Open/Close Reac Core Isol Cool 1E51-F07G Continuous Open/Close Reac Core Isol Cool 1E51-F077 Continuous Open/Close Reac Core Isol Cool IE51-F078 Continuous Open/Close Reac Core Isol Cool 1E51-F095 Continuous Open/Close Reac Core Isol Cool

) I658- Coo 26 IFC007 CordmucuS

. Continuous Open/Clov Close RMc Core TSol Cool Fuel Pool, Cool & Clean l

1FC008 Continuous Close Fuel Pool Cool & Clean IFC011A Continuous Open/Close Fuel Pool Cool & Clean IFC011B Continuous Open/Close Fuel Pool Cool & Clean 1FC015A Continuous Open/Close Fuel Pool Cool & Clean 1FC0158 Continuous Open/Close Fuel Pool Cool & Clean IFC016A Continuous Close Fuel Pool Cool & Clean 1FC0168 Continuous Close Fuel Pool Cool & Clean 1FC024A Continuous . Close Fuel Pool Cool & Clean 1FC0248 Continuous Close Fuel Pool Cool & Clean IFC026A Continuous Open/Close Fuel Pool Cool & Clean .

IFCO26B Continuous Open/Close Fuel Pool Cool & Clean l 1FC036 Continuous Close Fuel Pool Cool & Clean 3FC037 Continuous Close Fuel Pool Cool & Clean 1FP050 Continuous close Fire Protection IFP051 Continuous Close Fire Protection 1FP052 Continuous Close Fire Protection IFP053 Continuous Close Fire Protection 1FP054 Continuous Close Fire Protection 1FP078 Continuous Close Fire Protection 1FP079 Continuous Close Fire Protection 1FP092 Continuous Close Fire Protection 1G33-F001 Continuous 'Close React Wtr Cleanup

IG33-F004 Continuous Close React Wtr Cleanup

! IG33-F028 Continuous Close React Wtr Cleanup i 1G33-F034 Continuous Close React Wtr Cleanup I 1G33-F039 Continuous Close React Wtr Cleanup 1G33-F040 Continuous Close React Wtr Cleanup i IG33-F053 Continuous Close React Wtr Cleanup 1G33-F054 Continuous Close React Wtr Cleanup i

l 1HG001 Continuous Open H2 Recombining L IHG004 Continuous Open/Close H2 Recombining 1HG005 Continuous Open/Close H2 Recombining 1HG008 Continuous Open/Close H2 Recombining l

l CLINTON - UNIT 1 3/4 8-49 t

Description of Change Specification 3/4.9.12 page 3/4 9-19 Change note ** to read as follows:

"The blocking valve in the fuel building IFTS hydraulic

. power unit and the liquid level indicator are not required to be OPERABLE for purposes of these specifications until prior to off-loading of irradiated fuel."

Justification Letter from B. L. Siegel to F. A. Spangenberg dated December 11, 1986.

REFUELING OPERATIONS 3/4.9.12 INCLINED FUEL TRANSFER SYSTEM LIMITING CONDITION FOR OPERATION -

3.9.12 The inclined fuel transfer system (IFTS) may be in operation provided that:

, a. The access doors" of all rooms through which the transfer system penetrates are closed and locked.

b. All access door interlocks are OPERABLE.

c. The blocking valve located in the fuel building IFTS hydraulic power unit

'is OPERABLE.**

l

d. At least one IFTS carriage position indicator is OPERABLE at each carriage position and at least one liquid level sensor is OPERABLE.**

e. Any keylock switch that provides IFTS access control-transfer system lock-out is OPERABLE.

. APPLICABILITY: When the IFTS containment blank flange is removed.

i ACTION:

With the requirements of the above specification not satisfied, suspend IFTS operation with the IFTS at either termTnal point. The provisions of Specifi-cation 3.0.3 are not applicable. -

SURVEILLANCE REQUIREMENTS 4.9.12.1 Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to the startup of the IFTS, verify that no t personnel are in areas immediately adjacent to the IFTS and that all access doors to rooms through which the IFTS penetrates are closed and locked.

4.9.12.2 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the operation of IFTS and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, verify that':

a. All access door interlocks are OPERABLE.

b. The blocking valve in the Fuel Building IFTS hydraulic power unit is OPERABLE.**

c. At least one IFTS carriage position indicator is OPERABLE at each carriage position and at least one liquid level indicator is OPERABLE."*

l

Includes removable shields.

- The blocking valve in the fuel building IFTS hydraulic power unit and the liquid level indicator are not required to be OPERABLE for the purposes of these

... . specifications until after fu:1,10 ding, but b _fer:

1,r m ., ,mi ,r , __ m ,___ _ _ _ _ . . . ., u. ____ __

_ _ _ _x:::

. . _ding

_ . _ _ _ _ , 5% O f .*.

m_.2

. . . O O d/ ,

CLINTON - UNIT 1 3/4 9-19

t I Description of Change BASES 3/4.4.3 page B 3/4 4-3 Revise paragraph 3/4.4.3.1 as shown.

Justification The purpose of this revision is to update the BASES to the described revision to the FSAR as contained in letter U-600809 from F. A. Spangenberg to Dr. W. R. Butler.

t i

4 2

d I

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1

1 i

REACTOR COOLANT SYSTEM i

j BASES

3/4.4.1 RECIRCULATION SYSTEM (Continued) the recirculation flow control failures on increasing and decreasing flow are

presented in Sections 15.3 and 15.4 of the FSAR respectively.

2 The required surveillance interval is adequate to ensure that the flow control i valves remain OPERABLE and not so frequent as to cause excessive wear on the system components.

I 3/4.4.2 SAFETY / RELIEF VALVES .

i i The safety valve function of the safety / relief valves (SRV) operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1375 psig in accordance with the ASME Code. A total of 11 OPERABLE safety-i l

relief valves is required to limit reactor pressure to within ASME III allowable i

values for the worst case upset transient. Any combination of 5 SRVs operating

' in the relief mode and 6 SRVs operating in the safety mode is acceptable.

- Demonstration of the safety-relief valve lift settings will occur only during 4

shutdown and will be performed in accordance with the provisions of Specifica-j tion 4.0.5.

! The, low-low set system ensures that safety / relief valve discharges are minimized

! for a second opening of these valves, following any overpressure transient.

l This is achieved by automatically lowering the closing setpoint of 5 valves and l

lowering the opening setpoint of 2 valves following the initial opening. In l

this way, the frequency and magnitude of the containment blowdown duty cycle is i substantially reduced. Sufficient redundancy is provided for the low-low set system such that failure of any one valve to open or close at its reduced set-point does not violate the design basis.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE l

l 3/4.4.3.1 LEAKAGE DETECTION SYS Meef de /ded The RCS leakage detection systems required by this specification are provided to monitor and det et leakage from the reactor coolant pressure boundary. These detection systems r; ;;n:ht;nt with th: 7:: r :nd:t hn;"of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973/;

hev provide the ability to measure leakage from fluid systems in the drywell.

andare e*nsisfent wI He fhe recommendefions a f A Ats2 w 7.es, *.% ,, den / 4 e Q A/

"d/" 4 C**/*# Prenu, e A,,/.cy A c.4 I 3/4.4.3.2 OPERATIONAL LEAKAGE Defedan, " /48 2 The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also con-sidered. The evidence obtained from experiments suggests that for leakage some-what greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow CLINTON - UNIT 1 8 3/4 4-3

l Description of Change Figure 5.1.1-1, page 5-2 and Figure 5.1.3-1, page 5-4. '

Change the UNRESTRICTED AREA Boundary as shown.

Justification In accordance with the Offsite Dose Calculation Manual (ODCM) the UNRESTRICTED AREA is being revised to existing physical landmarks to provide :

for farming in certain areas surrounding the facility.

P i

a

. - - , . , , - , - ~ - - - - - . . . ~ , . . - - - .-_,-,,,,.n, , , . , _ _.o.c-n..n.--- -

,,.,,,n.._,,..., yr m , . .n, .-, -, . ---- ..,,- - , n..- .~

l , ,

f/

/

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Npy '

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.c c:n:,wns '

yaq @ / o ,,

** ) ' '

@caranoust

' rune noust-

//

. )

N '=

-g. s a[iSloi n M /

e!"tI's LIQUID O \

,/, ,,, I I EFFLUENT V s V s ARGE

,"o'7s"*/=' '

s wa POINT- ~

' . N'

/4 '/

e N et o

""E D.?;g,51C'L ,1

- .- - / N

'//Y//7W1///

3

// 1 l AIRBOR N E

/ .

I EFFLUENT DISCHARGE POINT LEGEND ELEVATION: 934 f f . MS L SITE BOUNDARY M UNRESTRICTED AREAS

'[ SUN'eUNg*ur"Sr0.E A RESIDENCE l

NOTE ,

The area in the lake between the bouys and the exclusion area boundary ji is unrestricted at this time, but will be controlled if plant effluent

i j condit4ons warrant closure.

Figure 5.1.1-1 Exclusion Area CLINTON - UNIT 1 5-2

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

I w

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i es i C l-

! 5 N

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! I, I

LEGEND

) ! E '

STE DOUNDARY I

/- , F77777/7A UNRESTRICTED AREA l - , ' , SPECIA4. UNftESTRlCTED

. AREA (See Note 1).

"" ,/ M AREA WITHIN SITE

! SOUNDARY 000T OW9ED me -

)."47

,4 "'

T L",'*, f .

.e.se.m (,..) BY IPC.

l

's F

I'

/ .s.mi..e m .

/

i

,. h//////r / /M ,, ,

/

appg g lA ,' / f

'I'l#

j /'/ l *l l/

l f i ,

,ii, /

/i s , /, .

, .., .en,

'f:.*:::,*.0 : ',*.: :"ll **.".:'f.2."*. #", ".". ';;::fr'

//' Nl,A_

b - - - - - - -

+ ,

e t ... . .. . , i ni . e. , .. ...m.

. /. />, ,

- 3.t -.i . , in .. o c , ei i e ==a % . ...

l 4 t . .. . in a... e , ...n . .

j g . .. .. i 6, soo et issL' H u .

. at m.. .. m. i.

.l r

Figure 5.1.3-1 Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Effluents l

1 j

Description of Change Figure 6.2.2-1, page 6-4 Change " Director" Plant Operations to " Assistant Manager" Plant Operations and delete

to the Manager - Clinton Power Station position.

Justification The Assistant Manager - Plant Operations position was created on October 17, 1986 and was reported to the NRC. This change took place after issuance of the low-power operating license.

The

to the Manager - Clinton Power Station is a typographical error. In accordance with the FSAR, there is no requirement for this position to hold an SRO License.

e MANAGER- ,

CLINTON POWER STATION p

h

.--e b$55s' tG Ai Nrube r AS$1STANT MANAGER CLINTON POWER

, $ ST ATION TECHNICAL

@ A DVISOR e

I g i I

ASSISTANT MANAGER ASSISTANT MANAGER d

Q PLANT MAINTENANCE STARTUP H

DIRECTOR PLANT "

AM DIRECTOR PLANT RADIATION PROTECTION DIRECTOR PLANT PLANT SUPPORT OPERATIONS MAINTENANCE TECHNICAL. SERVICES SUPERVISOR-PLANT RAOIATION PROTECTION .

i SUPE RVISOR- SUPERVISOR - SUPERVISOR - SUPERVISOR- SUPERVISOIF-SUPE RVISOR- ~

PLANT - MAINTENANCE MECHANICAL - RADIOLOGICAL -

NUCLEAR SECURITY OPERATIONS PLANNING MAINTENANCE ENGINEERING A

SU PERvlSOR- SUPERVISOR- SUPERVISOR

ADMNI.

SUPERVISOR- ~

' PLANT ELECTRICE -

RADIOLOGM - -

SUPERVISOR OPERATIONS COMPUTER MAINTENANCE OPERATIONS i SUPPORT SUPERVISOR- SUPE RVISOR- SUPERVISOR. SUPERVISOR-SUPERVISOR-_ ~

RADWASTE C81 -

RADIOLOGICAL -

SYSTEMS COMPLlANCE

' MAINTENANCE . SUPPORT t

< s.

SUPERVISOR- -

SUPERVISOR- SUPERVISOR-CHEMISTRY

' R ADIOLOGICAL - TESTING & -

ENVIRONMENTAL SCNEDULING e POSITION REQUIRES SUPERVISOR-i AN SRO LICENSE FIRE -

PROTECTION i

l Figure 6.2.2-1 Unit OnSite Organization l .

i l

i

. .o Description of Change Specification 6.2.3.4 page 6-6 Change Titles as indicated.

Justification Typographical Error. Titles should be consistent with Specification 6.2.3.1.

ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION 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 information, including units of similar

design, which may indicate areas for improving unit safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities, or other means of improving unit safety to the Manager-Licensing and Safety, Director-Nuclear Safety, and to

! members of the Nuclear Review and Audit Group (NRAG).

COMPOSITION 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time i engineers located on site. Each shall have a bachelor's degree in engineering or related science and at least 3 years' professional level experience in his field, at least 1 year of which experience shall be in the nuclear field.

RESPONSIBILITIES

The ISEG shall be responsible for maintaining surveillance of unit 6.2.3.3 activities to provide independent verification

that these activities are performed correctly and that human errors are reduced as much as practical.

i RECORDS 6.2.3.4 Records of activities performed by the ISEG shall be prepared, main-tained, and forwarded each calendar month to the Director-Nucleap Safety W j ing'n;;r'n; "n:1y:hT to the Manager-Lelear ';teti;n En-in;; ring and to members of the NRAG. Lietorstng and .%.fe.G 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support i

to the Shift Supervisor in the are'sa of thermal hydraulics, reactor engineering, and plant analysis with regard to safe operation of the unit. The Shift Technical Advisor shall have a bachelor's degree or equivalent in scientific or engineering discipline and shall have received specific training in the j 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.

I .

"Not responsible for signoff function.

CLINTON - UNIT 1 6-6 L

. c .

Description of Change Specification 6.7.1, page 6-14 Change' paragraph 6.7.1.d to provide the required reports to all parties within 30 days of the violation.

Justification Changing the 14-day limit to a 30-day limit is considered an operational enhancement. The 14-day limit was an internal restriction for scheduling provisions only. The 30-day limit is consistent with the requirements of 10CFR50.72 and 73.

.s .

ADMINISTRATIVE CONTROLS TECHNICAL REVIEW (Continued)

c. When required by 10 CFR 50.59, a safety evaluation to determine whether or not an unreviewed safety question is involved shall be included in the If it is determined that an procedure or the procedure change review.unreviewed safety quest to support that decision will be prepared and submitted to the FRG for review. Pursuant to 10 CFR 50.59, NRC approval of items involvir.g unre-viewed safety questions shall be obtained prior to the Power Plant Manager approval for implementation.

d. Written records of reviews performed in accordance with Item 6.5.3.1.a.

above, including recommendations for approval or disapproval, shall be prepared and maintained.

l 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTA8LE EVENTS:

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

b. Each REPORTABLE EVENT shall be reviewed by the FRG, and submitted to the NRAG and the Vice President.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. In accordance with 10 CFR 50.72, the NRC Operations Center shall be noti-l fled 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 'after l The Vice President and the NRAG shall

> the violation has been determined.

be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , l

\

b. A License Event Report shall be prepared in accordance with 10 CFR 50.73.

The report shall be re-

c. A Safety Limit Violation Report shall be prepared.This report shall de viewed by the FRG.

l preceding the violation, (2) effects of the violation upon unit components, systems, or structures, and (3) corrective action taken to prevent recurrenc h ll be submitted to the Comm ssion,

d. The Safety Limit Vio the Report NRAG,s and the Vice President withiny[ d l withi,. 00 @ =d L}ation a 3

the violation.

e. Critical operation of the unit shall not be resumed until authorized by the Commission.

. t 6-14 CLINTON - UNIT 1

- . ... - .. - = . . _ . . .- . . - _-. . ..- .. -. . - . _ - .

b i

Description of Change Specification 6.12.2, page 6-23 ,

Provide additional information for this Specification and add Specification

j. 6.12.3 t Justification l Upon issuance of the operating license, the NRC inadvertently left out portions of Specification 6.12.2 and 6.12.3.

3 i

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o *.

Insert to 1.6.12.2 under the administrative control of the Shif t Supervisor on duty and/or the Radiation Protection Supervisor. Doors shall remain locked except during periods of access by personnel under an approved RWP.* For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months a dose in excess of 1000 mrem ** that are located within large areas, such as the containment, where no enclosure exits for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted, and a flashing light shall be activated as a warning device.

6.12.3 In addition to the requirements of Specifications 6.12.1 and 6.12.2, for individual areas accessible to personnel such that a major portion of the body could receive in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months a done in excess of 3000 mrem,* entry shall require an approved RWP which will specify dose rate levels in the immediate work area and require that stay times shall be established.

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

Radiation protection personnel or personnel escorted by radiation protection personnel shall be exempt from the RWP issuance requirement for fields of less than 3000 mrem per hour during the performance of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.

- Measurements made at 18 inches from sources of radioactivity.

CLINTON - UNIT 1 6-23a

a ~, o ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued)

1. Records of the service lives of all snubbers including the date at which the service life commences and associated installation and maintenance records.

m. Records of analyses required by the radiological environmental monitoring program that would permit evaluation of the accuracy of the analysis at later date. This should include procedures effective at specified times and QA records showing that these procedures were followed.

6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consis-tent with the requirements of 10 CFR Part 20 and shall be approved, maiatained, and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR Part 20, each high radiation area in which the intensity e of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)." Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

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.

c. A health physics qualified intiividual (i.e., qualified in radiation pro-tection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Radiation Protection Supervisor in the RWP.

6.12.2 In addition to the requirements of Specification 6.12.1, areas access-ibletopersonnelwithradiationlevelssuchthatamajorportionofthebody shall be provided with could receive in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months a dose greater than 1000 mrema llockeddoorstopreventunauthorizedentry,andthekeysshallbemai DseH &Hachni

Radiation protection personnel or personnel escorted by radiation protection personnel shall be exempt from the RWP issuance requirement for fields of less than 3000 mrem per hour during the performance of their assigned radiation pro-tection duties, provided they are otherwise following plant radiation protec-tion procedures for entry into high radiation areas.

- Measurements made at 18 inches from sources of radioactivity.

CLINTON - UNIT 1 6-23 s

ML20210A981

Text

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