Text

Forwards Final Draft Tech Specs,Following Review,Including Comparison of Tech Specs W/Proposed Draft Std Specs,Fsar & SER for Listed Items
	ML20154R646
Person / Time
Site:	River Bend
Issue date:	05/13/1985
From:	Rubenstein L; Office of Nuclear Reactor Regulation
To:	Crutchfield D; Office of Nuclear Reactor Regulation
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Docket No. 50-458 I

MEMORANDUM FOR: Dennis Crutchfield, Assistant Ofrector for Safety Assessment, Division of Licensing -

FROM:

L. S. Rubenstein, Assistant Director for Core and

gO4, Plant Systems, Division of Systems Integration i

SUBJECT:

RIVER BEND STATION - REVIEW OF FINAL DRAFT COPY OF TECHNICAL SPECIFICATIONS The Auxiliary Systems Branch has reviewed those portions of the Final Draft Copy of the River Bend Technical Specifications which are in ASB's area of primary responsibility.

Our review included a comparison of the River Bend Technical Specifications with the proposed draft standard specifications, the River Bend Final Safety Analysis Report, and our Safety Evaluation Report for the following sections l

Technical Specification Subject 3/4.1.3.1 Control Rods 3/4.1.3.3*

Control Rod Scram Accumulators 3/4.1.5*

Standby Liquid Control System 3/4.7.4*

Remote Shutdown Monitoring Instrumentation and Controls 3/4.4.3*

Reactor Pressure Boundary Leakage

~

Detection Systems 3/4.4.7*

Main Steam Line Isolation Valves 3/4.4.10*

Main Steam Line Shutoff Valves 3/4.6.1.5*

MSIV Leakage Control System 3/4.6.1.10*

Penetration Valve Leakage Control System 3/4.6.5.5*

Shield Building Annulus Mixing System 3/4.6.5.6*

Fuel Building Ventilation 3/4.7.1.1*

Standby Service Water System 3/4.7.1.2*

Ultimate Heat Sink 3/4.7.2*

Main Control Room Air Conditioning System 3/4.7.8*

Area Temperature Monitoring 3/4.7.11*

Spent Fuel Storage Pool Temperature 3/4.9.6*

Refueling and Fuel Handling Platform 3/4.9.7*

Crane Travel - Spent and New Fuel Storage Transfer and Upper Containment Pools

Contact:

J. Ridgely X29566 B6 pJ33g124860123 PLETTIMS-511 p

PDR V

>: n W13 %;

z Technical Specification Subject 3/4.9.8*

Water Level - Reactor Vessel 3/4.9.9*

Water Level - Spent Fuel Storage and Upper Contain-ment Fuel Pools 3/4.9.12*

Inclined Fuel Transfer System 83/3.1.3 Bases - Control Rods 83/4.1.5*

Bases - Standby Liquid Control System 83/4.3.7.4*

Bases - Remote Shutdown Monitoring Instrumentation 83/4.4.3.1*

Bases - Leakage Detection Systems 83/4.4.7 Bases - Main Steam Line Isolation Valves 83/4.4.10*

Bases - Main Steam Shutoff Valves 83/4.6.1.5 Bases - MSIV Positive Leakage Control System 83/4.6.1.10 Bases - Penetration Valve Control System 83/4.7.1 Bases - Standby Service Water System 83/4.7.2 Bases - Main Control Room Air Conditioning System 83/4.7.8 Bases - Area Temperature Monitoring 83/4.7.11*

Bases - Spent Fuel Storage Pool Temperature 83/4.9.6 Bases - Refueling Platform 83/4.9.7*

Bases - Crane Travel - Spent and New Fuel Storage. Transfer and Upper Containment Fuel Pools 83/4.9.8* and Water Level - Reactor Vessel and 83/4.9.9 Water Level - Spent Fuel Storage and Upper Containment Fuel Pools 83/4.9.12*

Inclined Fuel Transfer System 5.6*

Fuel Storage 6.8.4*

(No Title) 6.9.2.b*

Special Reports t

- :s e >;<:u sysrvjpg s..

s IW4Y 13 UI 3

The

indicates those sections with which we have comments. Our comments include the re-writing of some forthe Technical Specifications in accordance with your instructions which were provided at the 1:00 p.m., April 23. 1985 meeting. Therefore, if the applicant provides the enclosed Technical Specifications as the River Bend Technical Specifications, they will be acceptable.

Technical Specification 3/4.7.1.2, Ultimate Heat Sink, includes a surveillance

'of the cooling tower basin water temperature. This aspect of this technical specification is an interim compensatory measure until the applicant has installed a permanent continuous monitoring systen by the first refueling outage. A license condition covering this monitoring system will be included in our next SSER input to DL.

L. S.1t enstein Assistant Director i

for Co and Plant Systems i

Division of Systems Integration j

Enclosure:

i As Stated cc w/ enclosure:

4 y

J. Ridgely D. Hquston cc w/o enclosure:

0. Parr J. Wilson W. Butler S. Stern C. Thomas h

L.,

s 7

. '. ;'e. ; (-

REACTIVITY CONTROL SYSTEMS c'

CONTROL RCD SCRAM ACCUMULATORS FlHL HAFT LIMITING CONDITION FOR OPERATION

~

3.1.3.3 All control rod scram accumulators shall be OPERABLE.

1, %

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 5*.

3 ACTION:

a.

In OPERATIONAL CONDITIONS 1 or 2:

1.

With one control rod scram accumulator inoperable, within 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s:

a)

Restore the inoperable accumulator to OPERABLE status, or b)

Declare the control rod associated with the inoperable accumulator inoperable.

Otherwise, be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

With more than one control rod scram accumulator inoperable, declare the associated control rods inoperable and:

a)

If the control rod associated with any inoperable scram accumulator is withdrawn, immediately verify that at least one control rod drive pump is operating by inserting at least one withdrawn control rod at least one notch or place the reactor mode switch in the Shutdown position.

~~

b)

Insert the inoperable control rods and disarm the associated directional control valves either:

1)

Electrically, or 2)

Hydraulically by closing the drive water and exhaust water isolation valves.

Otherwise, b,e in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In OPERATIONAL CONDITION 5*:

O 1.

With one withdrawn control rod with its associated scram accumu-lator inoperable, insert the affected control rod and disarm the associated directional control valves within one hour, either:

"At least the accumulator associated with each withdrawn control rod.

Not i

applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

RIVER BENO - UNIT 1 3/4 1-9 m 2 6 MS

,,c-_v---.,,

, - ~ - -,... - -. - -

,_y

.-----y

-

wpsey REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS LIMITING CONDITION FOR OPERATION ACTION:

(Continued) a)

Electrically, or b)

Hydraulically by closing the drive water and exhaust water isolation valves.

2.

With more than one withdrawn control rod with the as'oc.ated scram accumlator inoperable and with no control rod drive pump operating, immediately place the reactor mode switch in the Shutdown position.

The provisions of Specification 3.0.4 are not applicable.

c.

A-4.1.3.3 Each control rod scram accumulator shall be determined OPERABLE:

At least once per 7 days by verifying that the indicated pressure is a.

greater than or equal to M2 g0ig unless the centrol rod is inserted and disarmed or scrammed. A b.

At least once per 18 months by:

1.

Performance of a:

a)

CHANNEL FUNCTIONAL TEST of the leak detectors, and b)

CHANNEL CALIBRATION of the pressure detectors, and verifying an alarm setpoint of Mee psi on decreasing pressure.

920Z IW 2.

Verifying that each individual accumulator check valve maintains the associated accumulator pressure above the alarm set point for d Ad _gn:ta rw2 10 minutes, : u--;.. 3

.~ i.j m u:

m

r t'

with no control rod drive pump operating.

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RIVER BEND - UNIT 1 3/4 1-10

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REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 Two standby liquid control subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 5*.

ACTION:

a.

In OPERATIONAL CONDITION 1 or 2:

1.

With one subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days 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 .

(

2.

With both subsystems inoperable, restore at least one subsystem

)

to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 .

W b.

In OPERATIONAL CONDITION 5":

{

1.

With one subsystem inoperable, restore the inoperable subsystem k

to OPERABLE status within 30 days or insert all insertable con-trol rods within the next hour.

9 2.

With both subsystems inoperable, insert all insertable control k

rods within one hour.

A SURVEILLANCE REQUIREMENTS 4.1.fr-Each standby liquid control subsystem shall be demonstrated OPERABLE:

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that; f

a.

M,W N6

,s 1.

The temperature of the sodium pentaborate solytton 9 th:

tem

& 4: gaatar tt,m,.. we' t; 7C"T.- Is.h o$ Hy w 1. I T-l.

2.

The available volume of sodium pentaborate solution is within the limits of Figure 3.1.5'T for the percent weight concentra-tion determined once per 31 days per Specification 4.1.5.b.2.

3.

The heat tracing circuit is OPERABLE by determining the tempera-ture of the pump suction piping up to the first storage tank outlet valve to be greater than or equal to 70*F.

b.

At least once per 31 days by; 1.

Verifying the continuity of the explosive charge.

"Witn any control rod withdrawn. Not applicable to control rods removed per t

Specification 3.9.10.1 or 3.9.10.2.

RIVER BEN 3 - UNIT 1 3/4 1-19 pg 2 61985

nNAL DRAFT REACTIVITY CONTROL SYSTEMS

[B (f*o SURVEILLANCE REQUIREMENTS (Continued)

/

2.

Determininn that the ava weight of sodium pentaborate is greater than or equal t 424 bs and the percent weight concen-tration of sodium pentabor e in solution is within the limits of Figure 3.1.5-1 by chemical analysis."

~

3.

Verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position, Demonstrating that, when tested pursuant to Specification 4.0.5, the c.

minimum flow requirement of 41.2 gpa per pump at a pressure of greater than or equal to 42ftT psig is met.

// 90 d.

At least once per 18 months during shutdown by; 1.

Initiating one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch success-fully fired. Both injection loops shall be tested in 36 months.

2.

- Demonstrating that all heat traced piping 5:' ---

+4- **-- :

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g ( A -44 4r4 W 9 nnen is unblocked byt, p y lg A - W r M y y

--1,.,7 t h enr+4nn u r..

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=e, and then draining and flushing the piping -'-' ':

1-Q 4ert with domineralized water.7. + 9

~'wm

;;; ;-: _; - - +,-.--~

3.

Demonstrating that the storage tank heaters are OPERABLE by veri-fying the expected temperature rise for the sodium pentaborate solution in the storage tank after the heaters are energized.

"This test shall also be performed anytime water or boron is added to the solution or when the solution temperature drops below 70*F.

- This test shall also be performed whenever both heat tracing circuits have been found to be inoperable and may be performed by any series of sequential, over-C.

lapping or total flow path steps such that the entire flow path is included, t

i RIVER BEND - UNIT 1 3/4 1-20 APR 2 61985 i

l

. ze.x: n.y s.sp FINAL DRAF REACTIVITY CONTROL SYSTEMS j

f Figure 3.1.5-2

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9' J

@lO 14 i

X Z

N ACCEPTAB i

'N O ERATING GION A

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- TANK d

\\ LOW LEVEL l IA OVERFLOW ALARM 3542 I

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2-CO CENTRA N LINE JIfilli!11'll

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8 (5

2

\\\\

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=

6

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5 3000 4000 5000 NET VOLUME OF SOLUTION IN TANK (GALS)

FIGURE 3.1.5 b SODIUM PENTABORATE SOLUTION VOLUME / CONCENTRATION REQUIREMENTS i

RIVER BEND - UNIT 1 3/4 1-21 APR 2 6 #

n m

D TABLE 3.3.7.4-1 2<

REMDIE SHUIDOWN MONITORING INSTRUNENTATION

=

i m"

o MINIMUM INSTRUMENT REA000T CHANNEL 5 E

LOCATION

OPERA 8LE PER PANEL Q

1.

Reactor Vessel Pressure RSP1, RSP2 1

2.

Reactor Vessel Water Level RSP1, RSP2 1

3.

Safety / Relief Valve Demand Position.

RSP1, RSP2 (3) valves 1/ valve 4.

Suppression Pool Water Level RSP1, RSP2 1

5.

Suppression Pool Water Temperature RSP1, RSP2 R

1 6.

Drywell Pressure y

RSP1, PRS 2 1

0 7.

Drywell Temperature RSP1, RSP2 1

8.

RHR Systee Flow: Loop A RSP1 1

Loop B RSP2 1

Loop C RSP2 1

9.

RHR Hz Cooling Water Systee Flow: Loop A RSP1 Loop B RSP2 1

1.

RCIC Systen Flow 1

RSP1 1

II. RCIC Turbine Speed RSP1 1

1

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RSP1 - Remote Shutdown Panel Division I 8>/ h RSP2 - Remote Shutdown Panel Division !!

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FPdMF% DMFT TABLE 3.3.7.4-2 (Continued)

REMOTE SHUTDOWN SYSTEM CONTROLS MINIMUM CHANNELS OPERABLE DIV. 1 DIV. 2

22. RHR Outboard Shutdown Isolation MOV 1

NA (IE12*M0VF008)

23. RHR Inboard Shutdown Isolation MOV -

1 NA (IE12*MOVF009) 24 RHR Hx Flow to Supp. Pool MOV /

1 1

(IE12*MOVF011A, 8)

25. RHR Reactor Head Spray MOV 1

NA (IE12*M0VF023)

26. RHR Test Line MOV 1

1 (IE12*MOVF024A,8) f

27. RHR Hx Flow to RCIO MOV 1

NA (1E12*MOVF026A)

/

28. RHR Injection Shutoff MOV 1

1 (1E12*MOVF027A,8)

29. RHR Upper Pool Cooling Shutoff MOV f

1 1

(IE12*MOVF37A,8)

30. RHR Injection MOV-1 2(,)

(IE12*MOVF042A,8,C) f

31. RHR Hx Shell Side Inlet MCV 1

1 (IE12*MOVF047A. 8)

(

j

32. RHR Hx Shell Side Bypass MOV 1

1 (IE12*MOVF048A,8)

'33 RHR Discharge to Radwaste MOV

/

1 NA (1E12*MOVF040) 34 RHR Steam Isolation MOV e

1 1

(1E12*MOVF052A,8) f 35,- RHR Injection MOV 1

1 (1E12*MOVF053A,8)

36. RHR Pump Minimum Flow MOV 1

2(,)

/

(1E12*MOVF064A, 8 C) f

37. RHR Hx Water Discharge MOV 1

1 (1E12*MOVF068A,8)

38. Safety Re11ef Valves

/

3(,)

(1821*RVF051 C, G, 0)

39. SSW Pump

/

2(,)

(ISWP*P2A,2C,28,20)

40. Normal Service Water Isolation MOV /

1 1

(ISWP*MOV96A,8) j

41. SSW Cooling Tower Inlet MOV

/

1 1

(1SWP*MOV55A,8

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(a) - One per control equipment RIVER BEND - UNIT 1 3/4 3-79 asum da t

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2 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS i

g es E

CHANNEL CHANNEL y

INSTRlMENT CHECK CALIBRATION 4

E 1.

Reactor Vessel Pressure M

R w

2.

Reactor Vessel Water Level M

R 3.

Safety / Relief Valve Position M

NA 4.

Suppression Pool Water Level M

R 5.

Suppression Pool Water Temperature M

R 6.

Drywell Preasure M

R

,s

[

7.

Drywell Temperature M

R 8

8.

RtR System Flow: Loop A M

R Loop 8 M

R Loop C M

R 9.

RHR Hx Cooling Water System Flow: Loop A M

R

,a Loop B M

R U2 h

L 1). RCIC Systes Flow M

R.

"T1 IT' RCIC Turbine Speed muume M

R M

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0

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REACTOR COOLANT SYSTEM

['

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.3.1 The following reactor coolant system leakage detection systems shall be OPERABLE:

The drywell ata,csphere particulate radioactivity monfi:oring system, a.

b.

The drywell and pedestal floor sump drain flow monitoring Q l'.'$J 14w 4,r.'.~ a-J y'fd d a r-~p* flm "

b sy%,

y 4.

Either the drywell air coolers condensate flow rate monitoring system or the dryw11 atmosphere gaseous radioactivity monitoring system.

APPLICABILITY: OPERMIONAL CONDITIONS 1, 2 and 3.

ACTION:

m *-

w With only hue-of the above required leakage detection systems OPERABLE, t

cperation may continue for up to 30 days provided grab samples of the drywell

(

atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the i

required gaseous and/or particulate radioactive monitoring sys.ea is inoperable; otherwise, be in at least NOT 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 .

t SURVEILLANCE RE0ulREMENTS I

4.4.3.1 The reactor coolant sy6 tem leakage detection systems shall be demon-strated OPERABLE by:

/'

Drywell atmosphere particulate and gardous monitoring systems-a.

performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a CHANNEL FUNCTIONAL TEST at i4sst ence per 31 days and a CHANNEL CALIBRATION at least once per 16 months.

-A4.

b.

i7:!' r-; W ;-et! " : sump drain flow monitoring systems-performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION TEST at least once per 18 months.

Drywell air coolers condensate flow rate monitoring system-c.

performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.

RIVER BEND - UNIT 1 3/4 4-8

,,_,,.--w-

... ~.,. q..s o REACTOR COOLANT SYSTEM 4

OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE.

b.

5 gpm UNIDENTIFIED LEAKAGE.

c.

25 gpm total leakage veraged over any 24-hour period

,o.

d.

1 gpm leakage at a reactor coolant system pressure of 1 2 5 : 15-psig from any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within a.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

With any reactor coolant system leakage greater than the limits in b and/or c, above, reduce the leakage rate to within the limits within

(

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With any reactor coolant system pressure isolation valve leakage greater c.

than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least

~

two other closed manual, deactivated automatic or check

valves, 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 SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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Which have been verified not to exceed the allowable leakage limit at the last refueling outage or after the last time the valve was disturbed, whichever is more recent.

t RIVER BEND - UNIT 1 3/4 4-9 W

,, c. v, 9....

e REACTOR COOLANT SYSTEM -

g. ;,

SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:

Monitoring the drywell atmospheric particulate radioactivity at least a.

once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , ada b.

Monitoring the Z 4

.,: ;;;; = i M--.

least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

..:' sump flow rates at Monitoring the drywell air coolers condensate flow rate at least once c.

per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and d.

Monitoring the reactor vessel head flange leak detection system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5 and verifying the leakage of each valve to be within the 1

specified limit:

At least once per 18 months, sad-a.

b.

Prior to returning the valve to service following maintenance.

repair or replacement work on the valve which could affect its leakage rate j

-NERbf T gj n4.

4 e n t applicable for entry into 0N T p,6 % aei..q ~H.V se'tDows./u.%. q w -.44 rA~l f 1 & A w 40

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0(

,,w a RIVER BEND - UNIT 1 3/4 4-10

~~

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d' a

C FIM. DRAFT TABLE 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES Al M" SYSTEM VALVE NUMBER y pp m T

$NCTION 9: =

t,-

a) LPCS 1E21*A0VF006 LPCS Injection IE21"MOVF005 g,yrf LPCS Injectidn-~ ~

b) HPCS 1E22*A0VF005 e./, r-HPCS Injection IE22*MOVF004 HPCS Injection c) RCIC IE51*A0VF065 f.,

q., y.

RCIC Head Spray 1E51*MOVF013 RLIC Head Spray d) RHR 1E12*MOVF023

c. grf RHR Head Spray

~

1E12*A0VF041A LPCI A Injection IE12*MOVF042A LPCI A Injection IE12*A0VF0418 LPCI B Injection 1E12*MOVF0428 LPCI B Injection 1E12*A0VF041C LPCI C Injection 1E12*MOVF042C t'

LPCI C Injection 1E12*MOVF009 Shutdown Cooling A & B Suction 1E12*MOVF008 i Vff Shutdown Ccoling A & B Suction

(

IRHS*V240 Shutdown Cooling A & B Suction e

(

=

xp t ' 4:

RIVER BEND - UNIT 1 3/4 4-11 m

4 "iwiP-eeWWFWW*C ONE

MMeSe'W a*bW8New

,s.p y -

(

REACTOR COOLANT SYSTEM 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION Twomainsteamlineisolationvalves(MSIVs)permainsteamlineshallbe 3.4.7 OPERABLE with closing times greater than or equal to 3 and less_1han or equal to 5 seconds.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a.

With one or more MSIVs inoperable:

1.

Maintain at least one MSIV OPERABLE in each affected main steam line that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , either:

a)

Restore the inoperable vahe(s) to OPERA 3LE status, or b)

Isolate the affected main steam line by use of a deacti-vated MSIV in the closed position.

(.

2.

Otherwise, 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 .

2

' t,..... = 4 v 4 i >

v4 +..wu..

fi A sv.

e+

~77*f-=;

(

SURVEILLANCE REQUIREMENTS l

4.4.7EEachoftheaboverequiredMSIVsshallbedemonstratedOPERAELEp erifying full closure between 3 and 5 seconds when tested pursuant to Specification 4.0.5.

: ;--"h i an s o f 4- ' :-. e..,. 0., o i =

'" P"^

.r1mese r,,,& 3 t - --+ y '-t: 0" C ' " Z ",; Z ".:: ".^,,' u? -- 2 p,.., _

,,ie surveleiance :,,

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l l

=

.--#e,.---

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- - - - * - - " * ' * * * * ~ ~ ~ * " ~

.... : a::..py y.

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(

REACTOR COOLANT SYSTEM 3/4.4.f0MAINSTEAMLINE499am*99MVALVES LIMITING CONDITION FOR OPERATION CMM.

dd*Q

[ WW }

3.4.7 4we main steam line '-

'-+#, valve [+Enie per main steam line shall be OPERABLE.M th :'::in; ti :: ; --n.

^.........,..'. 1; : :... N, _.....

,. 1

%: :::--t.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

S WithoneormoreMS/Vsinoperaole:

a.

~

1.

.;ic.L..... ' ::i :::

" :" ::: AL :

- - - - " " ' * " - * ' - :.kc o u...~.. ; f - - ' - * = =_mw.h $$

4-ither:

a)

Restore the inoperable valve (s) to OPERABLE status /, orSO4W 5

b)

Isolate the affected main steam line by use of a deacti-vated MSIV in the closed position.

2.

Otherwise, 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 .

2 m, ~.... _.

... r........ :,. -- -- ;; ~

SURVEILLANCE REQUIREMENTS J

4.4.7 EachoftheaboverequiredMS$sshallbedemonstratedOPERABLEby verifying full closure i;t :- ? : 2 : :::.... when tested pursuant to Specification 4.0.5.

: ;--"4 inns of (n-^':-.....,.0.,.....v..,. R m.. k q_.._.4':

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CONTAINMENT SYSTEMS MSIV LEAKAGE CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.5 Two independent main ste n positive leakage control system (MS-PLCS) divisions shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With one MS-PLCS divisiot inoperable, restore the inoperable division to OPERABLE status within 36/ days 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 SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.1.5 Each MS-PLCS division shall be demonstrated OPERABLE:

By performing Surveillance Requirement 4.6.1.10.a.

a.

A L s O c. ~f-e h 4 % % M e d T b.

At least once per -31 days by ::r.L';,t;j-p :::r 0"E'f:IL: b v operating the compresser Icaded for at least 15 minutes.

c.

During each COLD SHUTDOWN, if not performed within the previous 92 days, by cycling each remote, manual and automatic motor operated valve through at least one complete cycle of full travel.

4.

At least onca per 18 months by performance of a functional test which includes simulated actuation of the division throughout its operattag' sequence, and verifying that each automatic valve actuates to ~its correct position and that 8.513 psid sealing pressure is established in each steam line.

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CONTAINMENT SYSTEMS P3El 5 qu yg.A R{

(

H; s.

PENETRATION VALVE LEAKAGE CONTROL SYSTEM

A8 LIMITING CONDITION FOR OPERATION 3.6.1.10 Two independent penetration valve leakage control system (PVLCS) divisions shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With one PVLCS division inoperable, restore the inoperable division to OPERABLE status within 7 days 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 SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REOUIREMENTS 4.6.1.10 Each PVLCS division shall be demonstrated OPERABLE:

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying division PVLCS accumulator a.

pressure greater than or equal to 101 psig.

b.

During each COLD SHUTDOWN, if not performed within the previous

(

92 days, by cycling each remote, manual and automatic motor operated valve through at least one complete cycle of full travel.

At least once per 18 months by:

c.

1.

Performance of a functional test which includes simulated actuation of the system throughout its operating sequence, and verifying that each automatic valve actuates to its correct l

position and that a sealing pressure greater than or equal to 21 psig is established in each sealing valve, and 2.

Leakage from valves-ipped with the PVLCS will be included in computation of 0.

La.

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C,0NTAINMENT SYSTEMS SHIELD BUILDING ANNULUS MIXING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.5 Two independent Shield Building Annulus Mixing subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one Shield Building Annulus Mixing subsyste6 inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, 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 SHUTDOWN w f

~.1 A b.

u p su c osa a,a.s

'kA ~****"+'"'L A M e-4 ~ co n N u M U

^ ' .7 '" ' 'N'+' ' '

d

t..

t uM " %

Ait w :.,) s.4 An-as.

SURVEILLANCE REQUIREMENTS 4.6.5.5 Each Shield Building Annulus Mixing subsystem shall be demonstrated OPERABLE:

(

64 A u.::t.-At least once per 31 days by initiating, from the control room, a.

J a-44 wbNes. system operates for at laast 15 minutesp=r.GW Muf

'.,... - sub I

I i

l L.

RIVER BEND - UNIT 1 3/4 6-59 APR 2 6 m 9

e

...... _S l CONTAINMENT SYSTEMS FUEL BUILDING VENTILATION LIMITING CONDITION FOR OPERATION 3.6.5.6 Two independent Fuel Building Ventilation Charcoal Filtration wh-systemsshallbeOPERABLE,andinOPERATIONALCONDITION]oneoperatingint 4

g ergency mode M -

^

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *.

ACTION:

With one Fuel Building Ventilation Charcoal Filtration subsystem a.

inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or:

1.

Ia OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUT 00hN 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 .

2.

In Operational Condition

, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

The provi-sions of Specification 3.0.3 are not applicable.

(

b.

With both Fuel Building Ventilation Charcoal Filtration subsystems inoperable or with one not operating in the emergency mode in Opera-tional Condition *, suspend handling of irradiated fuel in the sec-ondary containment, CORE ALTERATIONS or operations with a potential for draining the reactor vessel. The provisions of Specifica-tion 3.0.3. are not applicable.

SURVEILLANCE RECUIREMENTS i

4.6.5.6 Each Fuel Building Ventilation Charcoal Filtration subsystem shall be demonstrated OPERABLE:

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in OPERATIONAL CONDITION *, by verifying a.

one Fuel Building Ventilation Charcoal Filtration System operation.

b.

At least once per 31 days by initiating, from the control room, flow l

through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters OPERABLE.

"Wnon irradiated fuel is being handled in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

RIVER 3END - UNIT 1 3/4 6-61 g26g

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- - - - ~ ~ ~ - ~ ~ - ' ~ - ~ - - -

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ue-FINE DEFT 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS STANDBY SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.1 k

with each se:y: tea comprised of:At least two independent standby service water (

.;t,;,

M/

Two OPERABLE SSW pumps, and a.

b.

An OPERABLE flow path capable of taking suction from the standby cooling tower basin and transferring the water through the RHR heat exchangers, ECCS pump room seal coolers, and associated coolers and pump heat exchangers, d;erel gewb eMe^s, shall be OPERABLE:

In OPERATIONAL CONDITION 1, 2 and 3, two subsystems.

a.

b.

In OPERATIONAL CONDITION 4, 5 and*, the subsystem (s) associated with 3.5.2, 3.9.11.1, 3.9.11.2 and 3.8.1.2. systems and components require APPLICABILITY:

i

,k OPERATIONAL CONDITIONS 1, 2, 3, 4, 5 and *.

ACTION:

aliya *d d' 8"A }&

b

.U sw % *A at; 4 4 44-ONk

_ a.

In OPERATIONAL COWDITION 1, 2 or

/ *W" Ut, f falf With one SSW 3

$tes inoperable, kstore the inoperable M a# '

P'

2 yas to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 SHUTDOWN fl.

within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . gy,.4 vf 4 sw p--f r,a afes-W With both SSW t;;.inoperablefbeinatleastHOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN d ithin the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

7 7' u.

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. e nw..;L..,. : : f i; i--#" -i m 1 I 9 1 ca si:

_ :: :.,. a.

,._ ;; sau..

"When handling irradiated fuel in primary or secondary containment.

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w PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) t I

_ ACTION:

(Continued)

ADrDATTON4L rnNnTTTON a n.

i 33 Ta

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__a SURVEILIANCE REQUIREMENTS 4.7.1.1 shall be demonstrated OPERABLE:At least the above required standby serv Ae of'::(s) 2 :, _

p awensgyys % 3/asa{ d,m A A d*c path that is not locked, sealed or otherwise secu a.

i in its correct position.

is b.

At least once per 18 months during shutdown by verifying that:

l 1.

Each automatic valve :;-- '-'

-<-+;

'....,.., ~_. _

-'#-*y - ' t:d position;on a normal service water low pressure signal actuates to the correct

.,r-.,

EcLaks Iss 4me pump in each :2:,p : starts on a normal service water 2.

_t low pressure signal, and 1sof 3.

Each pump in each '"'n :t._ starts on a manual control signal from the main control room.

I i

RIVER BEND - UNIT 1 3/4 7-2 i,

1 o

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..+.

2 PLANT SYSTEMS.

ULTIMATE HEAT SINK 3

g QMITINGCONDITIONFOROPERATION

3. 7.1. 2 The standby cooling water storage basin shall be OPERABLE with level, USGS datum, andA minimtm basin water level at or above e a.

b.

An average basin water temperature of less than or equal to 82*F o AJM3 w.%c4el Two OPERABLE, cooling tower fan cells (5 fans per cell) per div c.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, 5 and *.

ACTION:

With the requirements of the above specification not satisfied With the basin water level less than 108'6" MSL or the temperatu wad ^

a.

greater than 82*F, then declare the SSW system inoperable ano take a

re the Action required by Specification 3.7.1.1.

b.

hith 71 f.o w i

anyonefancellinopedble OPERABLE status within # :,; or, be in at least HOT SHUTDOWN restore the inoperable fan cell to i

f f

the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and COLD SHUTDOWN within the next

! (

c.

With.one fan cell f - f ' '-,

T:."J.".: :n'-..

.... '..,. inoperable, m M n.. ', -.

...e j

be in at least HOT SHUTDOWN within w

the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the n i

_...vo w.w. wnui A vn

- " ' ' ' * " ~

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', ' ".' ' ! '.". 2, " K' G * " " ' ' " " * ' ~ ".

"When hanoling irradiated fuel in primary or secondary containment 4

i 1

RIVER BEND - UNIT 1 3/4 7-3 APR 2 61985

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PLANT SYSTEMS ULTIMATE HEAT SINK h,

SURVEILLANCE REQUIREMENTS OPEh8E: he standby cooling tower and water storage basin shall be dete 1

,7-q.;

g,

-,.~w..

u-

-v-u_.

u.4.,

,=+=4--

,o o,

__e O_

At least one per 31 days by starting the cooling tower fans in each cell from the control room and operating the fan ra least 15 minutes.

5 y n :f ;sy n bar;u sA IA 'k b c

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RIVER SEND - UNIT 1 3/4 7-4 APR 2 61985 l

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ll51.

FINAL Dggl PLANT SYSTEMS e

3/4.7.2 MAIN CONTROL ROOM AIR CONDITIONING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2 handling unit /ffiter train subsystems shall be OPERABLET APPLICABILITY:

All OPERATIONAL CONDITIONS and *.

~

ACTION:

In OPERATIONAL CONDITION 1, 2 or 3 with one main control roo a.

conditioning subsystem inoperable, restore the inoperable subsys to OPERABLE status within 7 days or be in at least HOT SDUTDO i

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 n

rs.

b.

In OPERATIONAL CONDITION 4, 5 or *:

1.

With one main control room air conditioning air handling / filter train subsystem inoperable, restore the inoperable subsystere to OPERABLE status within 7 days or initiate and maintain operatio of the.0PERABLE subsystem in the emergency mode of operation.

2.

With both main control room air conditioning air handling / filter train subsystems inoperable, suspend CORE ALTERATIONS,

(.

of irradiated fuel in the primary est secondary containment and operations with a potential for d aining the reactor vessel.

The provisions of Specification 3.0.3 are not applicable in Operatio c.

Condition *.

SURVEILLANCE REQUIREMENTS 4.7.2 Each main control room air conditioning subsystem shall be demonst OPERABLE:

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the control room air a.

temperature is less than or equal to 104*F.

b.

At least once per 31 days on a STAGGERED TEST BASIS by initia i

from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters OPERABLE.

Shen irradiated fuel is being handled in the primary containment or Fu l Eufiding.

e i

RIVER BEND - UNIT 1 3/4 7-5 APR 2 61985

......_.--.-.7.,

" ' ~

T~ '

.b.$.i PLANT S STEMS

.v~~q h

6 SURVEILLANCE REQUIREMENTS (Continued) t At least once per 18 months or (1) after any structural mainten c.

x.-

on the HEPA filter or chercoal adsorber housings, or (2) follo e

painting, fire or chemical release in any ventilation zone commu cating with the subsystem by:

1.

and bypass leakage testing acceptance crite 0.05% and uses the test' procedure guidance in Regulatory P tions C.S.a. C.5.c and C.S d of Regulatory Guide 1.52

2. March 1978, and the system flow rate is 4000 cfm + 10%

, Revision.

2.

of a representative carbon sample obtained 1

Regulatory Position C.6 b of Regulatory Guide 1.52 March 1978, meets the laboratory testing criteria o,f Regul Revision 2, Position C.6.a of Regulatory Guide 1.52, Revision 2 for a methyl fodide penetration of less than 0.175%; and, March 19 i

3.

Verifying a subsystem flow rate of 4000 cfm +

~.M,' during sub-system operation when tested in accordance wit /s A'll N510-1975.

d.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifyi within 31 days after removal that a laboratory

('

g ton C.6.b of Regulatory Guide 1.52, Revision 2, March 1978 Regulatory Guide 1.52, Revision 2the laboratory testing crite

, meets penetration of less than 0.175%., March 1978, for a methyl iodide At least once per 18 months by:

e.

I 1.

Verifying that the pressure drop across the combi and charcoal adsorber banks is less than 7 inch's HEPA filters while operating the subsystem at a flow rate of

terMauge

.f O cfm + 10%j e M "#)

4s. 4 p Rik pmts ek f it IAs r %

.t. ;e4,A g 4J ~% pn,w key m o rt > & JE9A 4%G V y v.eLa # W-(

i RIVER BEND - UNIT 1 3/4 7-6 s

APR 2 61985

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/

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.

test signals, the subsystem automatically sw 4

at a positive pressure of > 1/8 inch (ateremerger.cy mod outside atmosphere during s,ubsystem o)perat on at a flow rat uge relative to the less than or equal to 4,000 cfm:

a)

LOCA, e d b)

Local air irtake radiation monitors - High s)

/ J ;<.h k Lm % J tnw, 3.

Verifying that the heaters' dissipate 23 1 2.3 Kw when tested in accordance with ANSI N510-1975.

f.

y,A., %,.g.J Mi&7,^ - (ying that the HEPA filter banks::t ':;

t' ~After each complete or partial replacement of a HE verif y

m e f.

.,~ J... % ~ wo -

.t,wg w h M f.e,/uaccordance with ANSI N510-1975 f

.....3

.-s,...s-

't:.., ' '........ : ;-~ 2.% W m

rate of 4000 cfm 10%.

while operating the system at a flow After each complete or partial replacement of a charcoal adsoroer g.

bank by verifying that the charcoal adsorber banks::t'-;; W <e

' ;';;..- -.. A m w'S M N '.

(

f';..,?:-1::'"[.u.,j,n accordance w D_ s ou

.m

. o, om--....

a genated hydrocarbon refrigerant test ga dt Mb10-1975) fee a halo-at a flow rate of 4000 cfm : 10L while operating tne system ggy a 4AJ i"f'$

6 e

RIVER BEND - UNIT 1 3/4 7-7 APR 2 61985

~ ~

~~~~

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.w PLANT SYSTEMS g

3/4.7.8 AREA TEMPERATURE MONITORING

_ LIMITING CONDITION FOR OPERATION 3.7.8 within the limits indicated in Table 3.7.8-1.The temperature of e

APPLICABILITY:

Whenever the equipment in an affected area is required to be OPERABLE.

ACTION:

With one or more areas exceeding the temperature limit (s) shown in e 3.7.8-1:

For more than eight hours, prepare and submit a Special Report a.

the Commission pursuant to Specification 6.9.2 within the next 30 providing a record of the amount by which and the cumulative time ys the temperature in the affected area exceeded its limit and a sis to demonstrate the continued OPERABILITY of the affecte 16, b.

By more than M*F,. in addition to the Special Report required ab within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either restore the area to witnin its tereperature lim or declare the equipment in the affected area inoperable.

SURVEILLANCE REQUIREMENTS 4.7.8 dstermined to be within its limit at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months RIVER BENO - UNIT 1 3/4 7-31 APR 2 61985 ann M '

7

.. R.a.. :.1 FIMi. D2iiFT

(':

TABLE 3. 7. 8-1 AREA TEMPERATURE MONITORING AREA TEMPERATIIRE LIMIT (*F) 1.

Auxiliary Buildina a.

LPCS area b.

RHR A pump room 122 c.

RCIC pump room 122 d.

RHR B pump room 122 e.

RHR C pump room 122 f.

HPCS pump room 122 g.

MCC area (West) 122 h.

MCC area (East) 122 122 1.

Main steam tunnel (north) 122 j.

Standby gas treatment rooms k.

114 Annulus mixing fan area 122 2.

Diesel Generator Control Rooms a.

Diesel Generator IA b.

Diesel Generator 1B 104 c.

Diesel Generator 1C 104 104 3.

Control Buildino Standby switchgear room 1A a.

j b.

Standby switchgear room 18 104 104 c.

Division I battery room 70 d.

Division II battery room 70 e.

Division III battery room f.

Inverter IA room 70 g.

Inverter IB room 104 e

h.

Inverter 1C room 104 104 y, sw 6 Sa.de r wL 0"t WW 2.

3 s. SJ A ff W' l

l l

i RIVER BEND - UNIT 1 3/4 7-32 1

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7 FINAL *BY'h REFUELING OPERATIONS 4

3/4.9.6 REFUELING AN0 d > FUEL HAND _ LING PLATFORM LIMITING CONDITION FOR OPERATION 3.9.6 for handling fuel n;;; d ice or control rods.The refueling ander fuel e

b % L)e, APPLICABILITY: Duringhandlingoffuel.I

$ or control rods.

ACTION:

With the requirements for refueling andJu> fuel handling platform OP not satisfied, suspend use of any inoperable refueling platform equipmen operations involving the handling of control rods and fuel ::;--i?ie:

placing the load in a safe condition.

after A "= M *f SURVEILLANCE REOUIREMENTS L af4r 4.9.6 Each refueling ande r fu control rods or fuel t':^-t!!::

andling platform hoist used for handling of prior to the start of such operations with that hoist by:shall be demonstrat C

before the load exceeds 1200 pounds. Demonstrating opera a.

b.

Demonstrating operation of the overload cutoff on the frame mounted and monorail mounted auxiliary hoists when the load exceeds pounds.

500 2 50 M4^ t*a Demonstrating operation of the uptravel M c" a4ce' M c

- 2; ' w e.

c.

e..e + s w c---,4' fuel assembly to 8 feet, 6 inches below the water level.H ;ts whe d.

i.rk le A Demonstrating operation of the downtravel sect.e.6.1 M + 4 on the main hoist when grapple hook down travel reaches 4 inches below fuel assembly handle.

Demonstrating operation of the slack cable cutoff on the main hoist e.

when the load is less than 50 2 10 pounds.

f.

Demonstrating operation of the loaded interlock on the main hoist when the load exceeds 485 2 50 pounds.

Demonstrating operation of the redundant loaded interlock on the g.

main hoist when the load exceeds 456k 2 50 pounds.

Ll81 RIVER SEND - UNIT L APR 2 61985 3/4 9-8

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

M-

=3?

._ s FINAL DRAFT 1

REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL-SPENT AND NEW FUEL STORAGE, TRANSFE FUEL POOL 5 LIMITING CONDITION FOR OPERATION a;.

3.9.7 runtlice-in the spent or new fuel storage, transfer pool racks.

APPLICABILITY:

6 Alo or upper containment fuel pools.With fuel n r:Mcpin the spent or new fuel sto ACTION:

With the requirements of the above specification not satisfied load in a safe condition.

, place the crane The provisions of Specification 3.0.3 are not appli-cable.

SURVEILLANCE REQUIREMENTS 6

4W 4.9.7.1 Thefuelbuildingcraneloadsshallbeveriff equal to 1200 pounds before travel over fuel = n;;; m; to weigh less than or fuel storage pools and the lower transfer pools.

in the spent or new

~

4.9",7.2 than or equal to 1200 pounds before travel over fuel rn-t'M in the uppThe reactor building polar cran transfer and containment fuel pools.

er 3

atu c

1 RIVER BEND - UNIT 1 APR 2 61585 3/4 9 9 e

... :.--.42.:~

REFUELING OPERATIONS'

)

C WATER LEVEL - REACTOR VESSEL.

3/4.9.8 LIMITING CONDITION FOR OPERATION 3.9.8 pressure vessel flange.At least 23 feet of water shall be maintained over th reactor bm 110

_ APPLICABILITY:

reactor pressure vessel while in OPERATIONAL CONDIT being handled are feradiated or the fuel ---"i he seated within th p pt3 M ::::,l i o vessel are feradiated.

w aTo e reactor ACTION:

lsmE Withtherequirementsoftheabovespecificatio[notsatisfied operations involving handling of fuel

--"P

, suspend all reactor pressure vessel after placing all fuel M::e!bs and control rods i or control rods within the a safe condition.

n

%c) {.o 1

SURVEILLANCE REQUIREMENTS I

4.978 The reactor vessel water level shall be determined to b i

minimum required depth witnin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at leas per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during handling of fuel reactor pressure vessel.

1 S ; or control rods within the l m ellu 9

l l

l 1

APR 2 61o.25 RIVER BEND - UNIT 1 3/4 9-10

~~ m,

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FINAL DRAF REFUELING OPERATIONS

(

3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE AND UPPER CONT LIMITING CONDITION FOR OPERATION 3.9.9 At least 23 feet of water shall be maintained over the top of irradiated fuel c n:-il k_ - $ d in the spent fuel storage and upper containment fuel po racks.

'i *

J 6

Abs APPLICABILITY:

Whenever irradiated fuel !::: ilie:

or upper containment fuel pools.

are in the spent fuel storage ACTION:

gW With the requirements ofAhe above specification not satisfied, suspend all movement of fuel ""-'- and crane operations with loads in the spent fuel storage or upper containment fuel pool areas, as applicable after placing the fuel est:: k; and crane load in a safe condition.

Specification

.0.3 are not applicable.

The provisions of G-b l21 SURVEILLANCE REOUIREMENTS 4.9.9 The water level in the spent fuel storage and upper containment fuel pools shall be determined to be at least at its minimum required depth at least once per 7 days.

6 O

e APR 2 61985 RIVER BEND - UNIT 1 3/4 9-11 ma.

S""

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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 provide that:

The access door and floor plugs of all rooms through which the tr a.

system penetrates are closed and locked, b.

All access iteterlocks and palm switches are OPERABLE.

The blocking valve located in the fuel butiding IFTS hydraulic pow c.

unit is OPERABLE.

3.t '-e r + r:y;q a J < % 437 d I;p :4IM p

d.

IFTS Le OPERABLE.--' , carriage position vindicatori ^.

m.

on,s.

^ ^ : : ^ :.: ^ " ' ' ' * -

All keylock switches which provide IFTS access control-transfer sy e.

lockout are OPERABLE.

An A,Gq f.

Tee,warnifig lights outside of the access doors are OPERABLE APPLICABILITY,: When the IFTS containment blank flange is removed.

ACTION:

With the requirements of the above specification not satisfied operation with the IFTS at either terminal point.

, suspend IFTS The provisions of Specifi-cation 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.9.12.1 are in areas immediately adjacent to the IFTS tube and tha and floor plugs to rooms through which the IFTS tube penetrates are closed and locked.

4.9.12.2 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, verf fy that:Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the operation of IFTS an Asg Fm An & ~~3 WsAid

." ' - : t IFTS a.

eva se OPERABLE.: " "jcarriage position indicator.r::. _....

,- H

,a

""'7/ " L :.

An m.,Gi b.

The,wartnng lights outside of the access doors are OPERABLE I

RIVER BEND - UNIT 1 APR 2 61985 3/4 9-18

, p..

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= =.

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.... =

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REFUELING OPERATIONS

[

SURVEILLANCE RECUIREMENTS

(:

.me_

--.a

.......m C.

All access interlocks and palm switches are OPERA 3LE.

h The blocking valve in the Fuel Building IFTS hydraulic power unit is OPERADLE.

g All keylock switches which provide IFTS access control-transfer sys lockout are OPERABLE.

I i

t I

RIVER BEND - UNIT 1 3/4 9-19 APR 2 61985 re.e. -*

e ** s et" * * * *** * ; ** 6

a

_ _., _.. _ _. _ _ _ _ _ _ _ _ _.. _ _ _. - - - - - i

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~

- - - ~ - - -

c REACTIVITY CONTROL SYSTEMS BASES

~

', /

ROD PATTERN CONTROL SYSTEM (Continued)

~

- a The RPCS provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or in;e.rted.

The analysis of the n d drop accident is presented in Section (15.

) of I

the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.

The RPCS is also designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during higher power operation.

A dual channel system is provided that, above the low power 56.tpoint, restricts the withdrawal distances of all non-peripheral control rods.

This restriction is ' greatest at highest power levels.

3/4.1.5 STAWBY LIQUID CONTROL SYSTEM The staidby liquid control system prevides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods rema,in fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of baron wAich produce,s a concen-(

tratior. of 660 ppm in the reactor core in app oximately f90 to 120 rminutes.

[

A minimum available cuantity of 3542 gallons of. sodium pentatorate solution containing a minimum of"4046 lbs. of socium peataborate is required to rpeet J,

shutdown requirement of 3% Ak/k. There is an additional allowance of 4150rppm in the reactor core to account for imperfect mixing and the filling of other piping. systems connected to the reactor vessel. The time requirement was selected to override the reactivity insertion rate due to.cooldown following the l'enon poison peak and the required pumping rate is 41.1. gpm. The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted. The temperature requirement is necessary to ensure that the sodium pentaborate remains in solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with ore of the red.:ndant components inoperable.

l 1.

C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NE00-10527, March 1972 2.

C. J. Paone, R. C. Stirn and R. M. Young, Supplement 1 to NEDO-10527, July 1972 3.

J. M. Haun, C. J. Paone and :t. C. Stirn, Addendum 2, " Exposed Cores,"

Supplement 2 to NED0-10527, January 1973 APR 2 61985 RIVER BEND - UNIT 1 8 3/4 1-4

~

- -.. ----- n.

. 1 r$..:..

INSTRUMENTATION i

(q J

BASES

.s MONITORING INSTRUMENTATION (Continued) 3/4.3.7.3 METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the meteorological monitoring instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to eval-uate the need for initiating protective measures to protect the health and safety of the public.

This instrumentation is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs," February,1972.

3/4.3.7.4 REMOTE SHUTDOWN MONITORING INSTRUMENTATION 4 g MO The OPERABILITY of the remote shutdown monito n i trumentation ensures that sufficient capability is available to permit f)fg. '

locations outside of the co/ =l room. wN

-... o.a.. a

. ;^ T ",;.07: ^

' t - m ntro This capability is requirid 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 sufficient information is available on selected plant parameters to monitor and assess important variables following an accident.

(This capability is consistent with the recommendations 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 THI Action Plan Requirements," November 1980).

3/4.3.7.6 SOURCE RANGE MONITORS 2

The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown. At these power levels, reactivity additions shall not be made without this flux level information ava~ilable to the operator. When the intermediate range monitors are on scale, adequate information is avail-able without the SRMs and they can be retracted.

3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY 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 neutron flux distribution of the reactor core.

APR 2 61985 RIVER BEND - UNIT 1 B 3/4 3-5 4

..-. ~

r REACTOR COOLANT SY5 TEM BASES

~

. 7. e 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required' by this specification are provided to monitor and detect leakage from the reactor coolant pressure teundary. These de.tection systems are consistent with the recommendations of Regulatory Guide 1.45, "Repctor Coolant Pressure Boundary Leakage Detection

,,4 ~. Systems", May 1973. 5-

- -m--

^ #ar '% 6, A r.vr, w.La. r. 4 M. 4',, g a c,.w.u -,,r u. t.- w %, 4.%.%,p :., -,. ~ -..r..c:w, 3/4.4.3.2 OPERATIONAL LEAKAGE neWi J io) cc/u.

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 considered. The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly.

However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action.

(-

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross i

valve failure and consequent intersystem LOCA.

Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

3/4.'4.4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant.

Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is not as great when the oxygen concentration in the 3

coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION.

During shutdown and refueling operations, the temperature necessary

.for stress corrosion to occur is not present so a 0.5 ppm concentration of chlorides is not considered harmful during these periods.

I Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions ~. When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits. With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.

1 RIVER BEND - UNIT 1 B 3/4 4-2 APR 2 61985 l

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'd::'.lla,L, PLANT SYSTEMS BASES 3/4.7.8 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause loss of its OPERABILITY.

.3/4.7.9 MAIN TURBINE BYPASS SYSTEM The main turbine bypass system is required to be OPERABLE consistent with the assumptions of the feedwater controller failure analysis in FSAR Chapter 15, 3/4 7.10 STRUCTURAL $ETTLEMENT Structural settlement limitations are imposed and required to be verified so as to preserve the assumptions made in the static design of the major safety related structures, y%g 3 /q. -). t 8 TPkT h w% f @ L- %"P

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APR 2 6 %

RIVER BEND - UNIT 1 8 3/4 7-5 l

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FINA.I yp REFUELING OPERATIONS n

BASES 3/4.9.7 CRANE TRAVEL - SPENT AND NEW FUEL STORAGE, TRANSFER AND UPPER FUEL P0OLS bb The restriction on movement of 1 s in excess of the nominal weight of a fuel 2:: 21y over other fuel ="e 5 in the pools ensures that in the event this load is dropped 1) the activity release will be limited to that contained,*a af1 g

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and 2) any possible distortion of fuel in the storage racks

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will not result in a critical array.

activity release assumed in the safety analyses.This assumption is consistent with the 3/4.9.8 and 3/4.9.9 3TORAGE AND UPPER CONTAINMENT FUEL POOLSWATER LE is available to remove 99% of the assumed 10% iodine ga the rupture of an irradiated fuel =etWy This minimum water depth is consis-tentwiththeassumptionsofthesafety\\ana.

lysis.

hw b le 3/4.9.10 CONTROL ROD. REMOVAL These specifications ensure that maintenance or repair of control rods or of inaavertent criticality. control rod drives will be performed under conditions The requirements for simultaneous removal of more

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than one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn.

3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal loop be OPERABLE demo,nstrated ard that an alternate method of coolant mixin ensures that 1) sufficient cooling capacity is available during REFUELING, and 2) sufficient coolant circulation wou through the reactor core to assure accurate temperature indication and to distribute and prevent stratification of the poison in the event it becomes necessary to actuate the standby liquid control system.

The requirement to have two shutdown cooling mode loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures of residual heat removal capability.that a single failure of the operating lo With the reactor vessel head removed and 23 feet of water above the reactor vessel flange, a large heat sink is available for core cooling. Thus, in the event a failure of the operating RH3 loop, adequate time is provided to initiate alternate methods capable of decay heat removal or emergency procedures to cool the core.

APR 2 61985 RIVER BEND - UNIT 1 B 3/4 9-2 y,,-

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i REFUELING OPERATIONS BASES y

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Hn.r; 3/4.9.12 INCLINED FUEL TRANSFER SYSTEM 4'

The purpose of the inclined fuel transfei system specifi l

control personnel access to those potentially high radiation a cation is to adjacent to the system and to assure safe operation of th 1.

reas immediately e system.

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ll d.C FINAL DEFT DESIGN FEATURES 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure.5.1.1-1..

5.6 FUEL STORAGE CRITICALITY

5. 6.1.1 The spent fuel storage racks are designed and shall be maintained with:

A k,ff equivalent to less than or equal to 0.95 when flooded with a.

unborated water, including all calculational uncertainties and biases as described in Section 9.1 of the FSAR.

b.

A fuel assembly minimum center to center storage spacing of 7 in, within rows an

.25 in. between rows in the Low Density Storage Racks A th eqw mbW%4. pool.

5" A fuel assembly minimum center to center storage spacing of 6.25 in.

c.

with a neutron poison material between stored spaces in the High Density Storage Racks A

rped M A -4 A;16 '~ +6 A..t L.!/le,t J

i The storage of syssut fuel in the upper containment fuel storage pool is prohibited during n :' :p __~- _ o N4* N A M* W

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5.6.1.2 The K for new fuel for the first core loading stored dry in the

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spentfuelstokeracksshallbeadministrative1ycontrolledtonotexceed 0.98 when optimum moderation (foam, spray, fogging, or small droplets) is assumed.

1 5.6.1.3 Provisions shall be taken to avoid the entry of sources of optimum moderation (foam, spray, fogging, or small droplets) to preclude that K,ff for new fuel, stored in the new fuel storaga facility, could exceed 0.98.

j ORAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 95'.

CAPACITY pQ,, :L.. y l

5.6.3 The spent fuel storage pool in the fuel building is designed nd shall be maintained with a storage capacity limited to no more than fuel j

assemblies.

M 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7.1-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7.1-1.

RIVER BEND - UNIT 1 5-6 APR 2 61985 l

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FINAL DRAFT ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) u~~

2.

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

b.

In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1.

Training of personnel, 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

c.

Post-accident Samplino A program which will ensure the capability to obtain and analyze reacter coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident l

conditions. The program shall include the following:

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1.

Training of personnel,

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2.

Procedures for sampling and analysis, and 3.

Provisions for maintenance of sampling and analysis equipment.

I d.

Biofouling Prevention and Detection A program, approved by the NRC Staff prior to introduction of river water to the systems, which will ensure the procedures to prevent biofouling of safety-related equipment, assure detection of Corbicula in the intake embayment and the Mississippi River at the River Beno Station site, and monitor and survey safety-related equipment to ce-tact biofouling. Changes to this program will be submitted to and i

approved by the NRC prior to implementation.

6.9 REPORTING REOUIREMENTS I'

ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.

STARTUP REPORT

6. 9.1.1 A summary report of plant startup and power escalation testing shal.1 be submitted following (1) receipt of an Operating License, (2) amendment to RIVER BEND - UNIT 1 6-15 gggg l

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FINAL DRAFT ADMINISTRATIVE CCNTROLS SPECIAL REPORTS M.

6.9.2 Special reports shall be submitted in the following manner:

Special reports shall be submitted tc the Regional Administrator of a.

the Regional Office of the NRC within the time period specified for each report.

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Special rep rts in rega,rd to Corbicula,will be submittedrin accord-.M s h -r dance with the settlement agreement cated October 10, 1984.

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6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.2 The following records :; hall te retained for at least 5 years:

Records and logs of unit operation covering time interval at each a.

power level, b.

Records and logs of principal maintenance activities, inspections, f

repair, and replacement of principal items of equipment related to

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nuclear safety.

c.

All REPORTABLE EVENTS d.

Records of surveillance activities, inspections, and calibrations required by these Technical Specifications.

Records of changes made to the procedures required by Specification e.

6. 8.1.

f.

Records of radioactive shipments, g.

Records of sealed source and fission detector leak tests and results.

h.

Records of annuhl physical inventory of all sealed source material of record.

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Records of analyses required by the radiological environmental monitoring program.

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Records of emergency drills and exercises.

6.10.3 The following records shall be retained for the duration of the unit

, Ooerating License:

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a.

Records and drawing changes reflecting unit design modifications made

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to systems and equipment described in the Final Safety Analysis Report.

RIVER BEND - UNIT 1 6-20 APR 2 3 B35

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/pa aecoq'o, UNITED STATES E

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WASHINGTON, D. C. 20555

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May 13, 1985

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MEMORANDUM FOR: Dennis M. Crutchfield, Assistant Director for Safety Assessment Division of Licensing Thomas M. Novak, Assistant Director for Licensing Division of Licensing 9;O U FROM: Don H. Beckham, Acting Decuty Director Division of Human Factors Safety

SUBJECT:

REVIEW 0F RIVER BEND TECHNICAL SPECIFICATIONS DHFS has reviewed Section 3/4.10 and Sections 6.1 through 6.8.3 of the final draft Technical Specifications for River Bend Unit 1. Our comments are provided in Enclosure 1. is a copy of the pages from the draft Technical Specifications that we recomend should be changed,' with the changes noted. i Subject to inclusion of the corrections noted, DHFS concurs witn issuance of the Technical Specifications for River Bend Unit 1. ,(() D W Y &. Don H. Beckham, Acting Deputy Ofrector j Division of Human Factors Safety cc: E. Butcher 1 t=, ._qF m s % 1 % f/ / 00 va o+m Juy z y lO, 8 '7

11'. I)?n??; 1. Section 6.1.2 - The words in lines 1 and 2 of this section which read "or during his absence from the control room, a designated individual" should be enclosed in parentheses. Reason: To be consistent with the Standard Technical Specifications. 2. Section 6.2.2.a - The words "on duty" in the first line should be hyphenated. Reason: To be grammatically correct and to be consistent with the Standard Technical Specifications. 3. Section f.2.2.f - The words " health physicists" in lines 3 and 4 should be deleted and replaced with the words " radiation protection technicians." Reason: To be consistent with the applicants' title for these individuals, as used in Section 6.2.2.c and in Figure 6.2.2.-l. 4. Section 6.2.3.2 - Change the final clause in this section to read, "at least 1 year of which experience shall be in the nuclear field." Reason: To improve the grammar and to be consistent with the wording of the Standard Technical Specifications. t ,..-.. -.... -.....,. -.. ~ - - -... - - - - -. L

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? 5. Section 6.4.1 - Insert a comma in the second line after the words Manager-Administration. Reason: To be grammatically correct and to be consistent with the Standard Technical Specifications. 6. Section 6.5.1.6.e - Insert the words "Vice President - R8NG and the" in the last line between the words "the" and " Nuclear." 9 Reason: We have customarily required these FRC investigation reports to be furnished to the utility individual at.the level of the Vice President - RBNG. Such a change also would make this section consistent with the wording of the Standard Techn'ical Specifications.* 7. Section 6.3.5.7 - Change the lead words of this section to read, "The NR8 shall be responsible for the review of:" Reason: The R8 need not itself perform the reviews. The revised wording is consistent with the Standard Technical Specifications. 8. Section 6.6 - Change the title of this section to read "Re[ortable Event j Action." Reason: To be consistent with the Standard Technical Specifications. f

4._. + E

C FINAL. DRAFT 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

6.1. 2 The Shift Supervisor (or during his absence from the control room, a Amanagementdirectiv)etothiseffect,signedbytheSeniorViceP X X River Bend Nuclear Group shall be reissued to all. station personnel on an annual basis. 6.2 ORGANIZATION OFFSITE 6.2.1 be as shown on Figure 6.2.1-1.The offsite organization for unit management and te UNIT STAFF 6.2.2 The unit organization shall be as shown on Figure 6.2.2-1 and: Each on,-duty shift shall be composed of at least the minimum shift a. X crew composition shown in Table 6.2.2-1; b. At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in OPERATIONAL CONDITION 1, 2 or 3, at lea'st one licensed Senior Operator shall be in the control room; A Radiation Protection Technician" shall be on site when fuel is in c. the reactor; d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limitec to Fuel Handling who has no other concurrent responsibilities during this operation; A site fire brigade of at least five members shall be maintained on e. site at all times". The fire brigade shall not include the Shift Supervisor, the Shift Technical Advisor, the Cor. trol Operating Foreman, nor the two other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency; and "The Raoiation Protection Technician and fire brigade composition may be less than i the minimum requirements for a period of time not to exceed 2 hours, in orcer i to accommodate unexpected absence, provided immediate action is taken to fill l the required positions. t ( I RIVER BEND - UNIT 1 6-1 1 526m 1 1

F g as ADMINISTRATIVE CONTROLS G l l UNIT STAFF (Continued) f. Administrative procedures shall be developed and implemented to limit radkfien prdet ism l f the workino hours of unit staff who perform safety-related functions fgehnicians j Y (e.g., licensed Senior Operators, licensec operators,"te:!*2 ;5 ::- 'r~ gese44, auxiliary operators, and key maintenance personnel). Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a nominal 40-hour week while the unit is operating. the event that unforeseen problems require substantial amounts ofHowever, in overtime to be used, or during extended periods of shutdown for refue-basis the following guidelines shall be followed: ling, majo 1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time. 2. An individual should not be permitted to work more than 16 hours in any 24-hour period, nor more than 24 hours in any 48-hour period, nor more than 72 hours in any seven day period, all excluding shift turnover time. 3. A break of at least eight hours should be allowed between work periods, including shift turnover time. 4. Except during extended shutdown periods, the use of overtime should be considered an an individual basis and not for the entire staff on a shift. Any deviation from the above guidelines shall be authorized by the Plant Manager or either one of the Assistant Plant Managers or the Supervisor-Radiological Programs, or higher levels of management, in basis for granting the deviation.accordance with established procedures Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager or his designee to assure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized. I I I RIVER BEND - UNIT 1 6-2 AP 2 6 585 l I n

cm FINAL DRAFT l ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERI'NG GROUP (ISEG) 1 FUNCTIOj 6.2.3.1 The ISEG shall function to examine unit operating characteristics, NRC issuances, induttry advisories Licensee Event Reports, and other sources of unit design and operating experience information, including units of simi-lar 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 Vice President - Safety and Environment. COMPOSITION S 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time of wk,%) or related scionee and at least 2 years professional levelEach engineers located onsite. t experience in his l'h8N,bt fielo, at least 1 yea]experiencepn the nuclear nela. RESPONSIBILITIES 3 6.2.3.3 The ISEG shall be responsible for maintainirq 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.

I RECOR05 3 6.2.3.4 Records of activities performed by the ISEG shall be prepared, main- { tained, and forwarded each calendar month to the Vice President - Safety and i Environment. 6.2.4 SHIFT TECHNICAL ADVISOR 4 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor (ngineering, and plant. analysis with regard to safe operation of the unit. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineer-ing 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. 1 For the dual role position shown in Table 6.2.2-1, *.he Shift Technical Advisor Ch shall have a bachelor's degree or shall have completed all technical courses required for the degree in a scientific or engineering discipline and shall have received all of the training for the normal STA position described above. Fot responsible for sign-off function. ?. l RIVER BEND - UNIT 1 6-6 M 2 6 SIS - - - - ~ ~ - - - - 4 r

ADMINISTRATIVE CONTROLS QUORUM

6. 5.1. 5 The quorum of the FRC necessary for the performance of the FRC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated al no more than two alternates, ternate and(foud memcers including t N3 e4%// 6e re6ced /e f$rrta me bses W f det o'.

RESPONSIBILITIES

6. 5.1. 6 The FRC shall be responsible for:

Review of all plant general administrative procedures and changes a. thereto; b. ?* view of all proposed tests and experiments that affect nuc' ear safety; Review of all proposed changes to Appendix A Technical Specifications; - c. d. Review of all proposed changes or modifications to structures, compo-nents, systems or equipment that affect nuclear safety; visfruilerie e. Investigation nf all violations of the Technical Specifications, 4s 4 s.J h includino the preparation and forwardino of rep ets covering evaluation and recoemendations to prevent recurrence, to the Nuclear Review Board; - f. Review of all REPORTABLE EVENTS; Review of u'it operations to detect potential hazards to nuclear safety, , g. n items that may be included in this review are NRC inspection reports, QA audits / surveillance reports of operating and maintenance activities, NRS audit results, and American Nuclear Insurer (ANI) inspection results; - h. Performance of special reviews, investigations, or analyses and reports thereon as requested by the Plant Manager or the Nuclear Review Board; and i. Review of initial start up testing phase start up procedures and revisions.

6. 5.1. 7 The FRC shall:

Recommend in writing to the Plant Manager approval or disapproval of a. items considered under Specification 6.5.1.6.a. through d. prior to their implementation. b. Render determinations in writing with regard to whether or not each item considered under Specification 6.5.1.6.a. through e. constitutes an unreviewed safety question. Provide written notification within 24 hours to the Vice President - c. RBNG and the Nuclear Feview Board of disagreement between the FRC and the Plant Manager; however, the Plant Manager shall have rescon-sibility for resolution of such disagreements pursuant to Specifica-tion 6.1.1. t RIVER BEND - UNIT 1 6-8 APR 2 61985 +

.ms*** FINA!. DRAFT ADMINISTRATIVE CONTROLS MEETING FREQUENCY 6.5.3.5 year of unit operation following fuel loading and at least on thereafter. QUORUM 6.5.3.6 and audit functions of these Technical Specifications shall c Chairman or the Vice Chairman and at least six NRB members including no more than two alternates. responsibility for oper1 tion of the unit.No more than a minority of the quorum shall h REVIEW bs respend61s for the review of : 6.5.3.7 The NRS shall, read s The safety evaluations for (1) changes to procedures, equipment, or a. systems; and (2) tests or experiments completed under the provision of 10 CFR 50.59 to verify that such actions did not constitute an unreviewed safety question; b. Proposed changes to procedures, equipment, or systems which involve i an'unreviewed safety question as defined in 10 CFR 50.59; Proposed tests or experiments which involve an unreviewed safety c. question as defined in 10 CFR 50.59; d. Proposed changes to Technicai Specifications or this Operating License; Violations of ccdes, regulations, orders, Technical Specifications, e. license requirements, or of internal procedures or instructions having nucicar safety significance; f. Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety; g. All REPORTABLE EVENTS: h.. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; and 1. Reports and meeting minutes of the FRC. AUDITS 6.5.3.8 Audits of unit activities shall be performed under the cognizance of the NRB. These audits shall encompass: RIVER BEND - UNIT 1 6-11 APR 2 61985

3. A'DMINISTRATIVE CONTROLS

'8 RECORDS (Continued)

Minutes of each NR8 meeting shall be prepared, approved, and forwarded a. to the Senior Vice President - ABNG within 14 days following each meeting. b. Reports of reviews encompassed by Specification 6.5.3.7 shall be within 14 days following completion of the review, prepared, approv Audit reports encompassed by Specification 6.5.3.8 shall be forwarded - c. to the Senior Vice President - RBNG and to the management positions responsible for the areas audited within 30 days after completion of the audit'by the auditing organization. 6.6 REPORTABLE EVENT Ae rie v_ 6.6.1 The following actions shall be taken for REPORTABLE EVENTS: The Commission shall be notified and a report submitted pursuant to a. the requi.rements of 10 CFR 50.73 and b. Each REPORTABLE EVENT shall be reviewed by the FRC and the results of this review shall be submitted to the NRB and the Plant Manager. - 6. 7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated: The NRC Operations Center shall be notified by telephone as soon as a. possible and in all cases within 1 hour. 4 The Senior Vice Presicent - R8NG and the NRB chairman (or personnel acting for their function) shall be notified within 24 hours. b. A Safety Limit Violation Report shall te prepared. The report shall be reviewed by the FRC. This report shall describe (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. The Safety Limit Violation Report shall be submitted to the Commission, c. the NRB, and the Senior Vice President - R8NG within 14 days of the violation, d. Critical operation of the unit shall not be resumed until authori:ed by the Commission. 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written proceoures shall be established, implemented, and maintainec covering the activities referenced below: t RIVER BEND - UNIT 1 6-13 hPR 2 0 E$ t ,, w - gee. g a pu ew*'*' 3._-, 7

w ADMINISTRATIVE CONTROLS STARTUP REPORT (Continued) the license involving a planned increas wer level, (3) installation of fuel that has a different design or h been manufactured by a different fuel supplier, and (4) modifications th may have significantly altered the nuclear,, thermal, or hydraulic p formance of the unit.

6. 9.1. 2 The startup report s 11 address each of the tests identified in the Final Safety Analysis Report nd shall include a description of the measured N

valuesoftheoperatingconditionsorcharacteristicsobtainedduringthetest program and a comparison-of these values with design predictions and specifica-tions. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report. 6.9.1.3 tion of the startup test program, (2) 90 days following resumption whichever is earliest. ment of commercial power operation, or (3) 9 months following initia If the startup report does not cover all three events (i.e, initial criticality, completion of startup test program, and resumption at least every 3 months until all three events have been completed.or co ANNUAL RE' ORTS P 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

6. 9.1. 5 Reports required on an annual basis shall include:

A tabulation on an annual basis of the number of station, utility a. and other personnel (including contractors) receiving exposures gr, eater than 100 mrem /yr and their sissociated man-rem exposure according to work and job functions * (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance waste processing, and refueling). The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements. Small exposures totalling less than 20% o.' the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole-body dose received from external sources should be assigned to specific major work functions; b.' Documentation of all challenges to safety / relief valves. l "This taoulation supplements the requirements of $20.407 of 10 CFR Part 20. RIVER BEND - UNIT 1 6-16 y265 .}}

ML20154R646

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