ML20209E497

Forwards Addl Final Draft Tech Specs W/Editorial Errors Not Identified by Util,Supplementing D Houston 850517 Memo

ML20209E497
Person / Time
Site:	River Bend
Issue date:	05/21/1985
From:	Benedict R NRC - TECH SPEC REVIEW GROUP
To:	Butcher E NRC - TECH SPEC REVIEW GROUP
Shared Package
ML20209E496	List: IA-85-511, Discusses Remarks Re Util Awareness of Errors,Discrepancies & Problems in Current Tech Specs.Caution Should Be Employed in Final Audit & Review of App A.Util Already Demonstrated Inability to Identify All Errors in Final Draft ML20209E489 ML20209E497 ML20212K372 ML20212K376 ML20212K388 ML20212K400 ML20212K411
References
FOIA-85-511 NUDOCS 8505310550
Download: ML20209E497 (39)

Text

__

/ 'o

• g UNITED STATES 4 8' n NUCLEAR REGULATORY CO' 11SSION

{ c WASHINGTON, D. C. 20555 o, a

....+

/ May 21, 1985 r

MEMORANDUM FOR: Edward J. Butcher, Group Leader Technical Specification Review Group, DL -

FROM: R. A. Benedict, Reactor Engineer Technical Specification Review Group, DL

~

SUBJECT:

MORE DEFICIENCIES IN REGARD TO GSU CERTIFICATION OF RIVER BEND TECHNICAL SPECIFICATIONS (FINAL DRAFT)

Further to Dean Hous't'on's May 17, 1985 memorandum to you on this same subject, there are an additional 38 pages in which editorial errors were not identified by GSU. The missed errors are circled on the enclosed pages.

'1 All told, 79 pages.out of 509 had editorial errors that GSU missed.

R. A. Benedict, Reactor Engineer Technical Specification Review Group, DL

Enclosure:

As stated cc: D. Crutchfield D. Houston S. Stern i

as 85@55u06 XA 39pp

l 2.1 SAFETY LIMITS BASES

2.0 INTRODUCTION

The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.06. MCPR greater than 1.06 represents a conservative margin relative

- to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive -

materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use f .,

relateo cracking, the thermally caused cladding perforations signal a thres-(. j - hold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transi-tion boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operatio7.

2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region i ially all elevation head, the core pressure drop at low power and flows )

ill be greater than 4.5 psi. Analyses show that with a bundle flow of j 28 x~ O hl s/hr, bund r re drop.is nearly independent of bundle power and ha a,value of 3 5 p ' ' us, the bundle flow with a 4.5 psi driving head ill greatertha[28 bs/hr. Full scale ATLAS test data taken at g' o .

pressures from 14.7 sia [N M. ~ g psia indicate that the fuel assembly critical power at this flow i or ately 3.35 MWt. With the design peaking f actors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. '

Thus, a THERMAL POWER limit of 25% of RATED THERMAL PCWER for reactor pressure below 785 psig is conservative.  ;

e a

RIVER BEND - UNIT 1 B 2-1 '

I

/l l

, .oEACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIES LIMITING CONDITION FOR OPERATION 3.1.2 The reactivity equivalence of the difference between the actual R00 DENSITY and the predicted ROD DENSIT) shall not exceed 1% delta k/k.

j APPLICABILITY: OPERATIONAL CONDITIONS I and 2.

ACTION:

With the reactivity equivalence re ceeding 1% delta k/k: .

a. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> perfo an an is_to determine and explain the cause #

of the reactivity differe  ; operation may continue if the difference

(?3 is explained and corrected.

b.

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

SURVEILLANCE REQUIRE'MENTS

( 4.1.2 The reactivity equivalence of the difference between the actual R0D DENSITY equal to 1% anddelta the predicted k/k: ROD DENSITY shall be verified to be less than or l a.

During the first startup following CORE ALTERATIONS, and b.

At least once per 31 effective full power days during POWER OPERATION.

e

{

RIVER BEND - UNIT 1 pt6M 3/4 1-2

. /

REACTIVITY CONTROL SYSTEMS CONTROL ROD MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION '

3.1. 3. 2 The maximum scram insertion time of each control rod from the fully withdrawn position, based on de-energization of the scram pilot valve solenoids as time zero, shall not exceed the following limits:

Maximum Insertion Times to Notch Position (Seconds)

Reactor Vessel Dome '

Pressure (psia)* 43 29 13 950 0.31 0.81 1.44 1050 0.32 0.86 1.57 APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. -

ACTION: -

a. With the maximum scram insertion time of one or more control rods exceeding the maximum scram insertion time limits of Specification 3.1.3.2 as deter-mined by Surveillance Requirement 4.1.3.2.a or b, operation may continue p.rovided that:

1. For all " slow" control rods, i.e. , those which exceed the limits of Specification 3.1.3.2, the individual scram insertion times do not exceed the following limits:

Maximum Insertion Times to Notch Position (Seconds)

Reactor Vessel Dome Pressure (psic)* 43 29 13 950 0.38 1.09 2.09 1050 0.39 1.14 2.22

2. For " fast" control rods, i.e. , those which satisfy the limits of Specification 3.1.3.2, the average scram insertion times do not exceed the following limits:

Maximum Average Insertion Times to Notch Position (Seconds)

Reactor Vessel Dome Pressure (psic)* 43 29 13 950 0.30 0.78 1.40 1050 0.31 0.84 1,53 3.

The sum of " fast" control rods with individual scram insertion times in excess of the limits of ACTION a.2 and of " slow" contr,ol rods does not exceed 5. -

• For intermediate reactor vessel dome pressure, the scram time 't oO determined by linear interpolation at each notch position. riteriA is ,

,g

/ R RIVER BEND - UNIT 1 3/4 1-6 PR 2 6 E

. V REACTIVITY CONTROL SYSTEMS 1

!)

ROD PATTERN CONTROL SYSTEM 8"'" " " "

LIMITING CONDITION FOR OPERATION 3.1.4.2 The rod pattern control system (RPCS) shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*# .

ACTION:

a. With the RPCS inoperable or with the requirements of ACTION b, below, not satisfied and with:

fo ]

1. THERMAL POWER less than ori qual n20% RATED THERMAL POWER control rod movement shall not be ; xcept by a scram.

Q p3

2. THERMAL POWER greater than 20% of RATED THERMAL POWER control rod withdrawal shall not be permitted.

b. With an inoperable control rod (s), OPERABLE control rod movement may continue bf bypassing the inoperable control rod (s) in the RPCS provided that:

1. With one control rod inoperable due to being immovable, as a result of excessive friction or mechanical interference, or

' known'to be untrippable, this inoperable control rod may be bypassed in the rod gang drive system (RGDS) and/or the rod action control system (RACS) provided that the SHUTDOWN MARGIN has been determined to be equal to or greater than required by Specification 3.1.1.

2. With up to eight control rods inoperable for causes other than addressed in ACTION b.1, above, these inoperable control. rods may be bypassed in the RACS provided that:

a) The control rod to be bypassed is inserted and the direc-tional control valves are disarmed either:

1) Electrically, or -

2) Hydraulically by closing the drive water and exhaust water isolation valves. i.

b) All inoperable controi rods are separated from all other inoperable control rods by at least two control cells in all directions.

c) There are not more than 3 inoperable control rods in any RPCS group.

• See Special Test Exception 3.10.2

1. Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RPCS prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.

RIVER BEND - UNIT 1 3/4 1-17 APR 2 6198s

TABLE 3.3.7.2-1 j

_ SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE 1.

Triaxial Time-History Accelerogra hs

{

a. Reactor Bldg 70'0" '

~~~

b. Reactor Bldg Ex Shield Wall 0 1 1. 0 g 1' EL 232'0"

c. Reactor Bldg Orywell EL 151'0" 0 1 1. 0 g 1

d. 0 1 1.0 g 1 s

Free Field - Grade Level 0 1 1.0 g 1

2. Triaxial Peak Accelerographs -

a. Reactor Bldg SLCS Storage Tank

~

0 t 10.0 g

b. Reactor Bldg - RHR Inj. Piping 1

c. 0 1 10.0 g I Aux. Bldg Service Water Piping 0 1 10.0 g 1

3. Triaxial Seismic Switch

a. Reactor . Bldg Ma Eh 70' '

O.025 to 0.25 g I I")

g 4. Triaxial Response-Spec Recorders

a. Reactor B1dg Hat EL 70'0"

b. Reactor Bldg Floor EL 141'0" Ot2g 1(*)

c. Auxiliary Bldg Mat EL 70'0" 012g 1

d. Auxiliary Bldg Floor EL 141'0" 012g 1 012g 1 I")With reactor control room indication and annunciation. '

I 8

RIVER BEND - UNIT 1 3/4 3-71 (P4 2 6 885

I FINA!. DRAFT  :

TABLE 3.3.9-1 (Continued) I ACTION 150 - a. Withonechannel.inoperagle,placetheinoperablechannel in the tripped condition within one hour or declare the associated system inoperable.

b. With more than one channel inoperable, declare the associated system inoperable.

ACTION 151 - a. With the number of OPERABLE channels one -lesrthan required by the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERABLE status within 7 days or be )

in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and COLD dr:UTDOWN l within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 1 9

b. With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels requirement, restore at least one of the inoperable channels to OPERABLE status ~

within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTOOWN within

.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> an SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 152 - Declare t e so lated Containment Ventilation System inoperabl.e gg ACTION 153 - 1.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERABLE status within 7 days or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2. With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels requirement, restore at least one of the inoperable channels to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least .

STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.  ;

1. Provided this does not M tuate the system. l 1

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RIVER BEND - UNIT 1 3/4 3-110 l

.1

. 00 FINAL DilAFT I i i i I l 1600 - _

A - Sv$ TEM HYOROTEST LIMIT wtTH FUE L IN VE$$E L e - NONWUCLE AR HE ATING LIMIT 1400 - 8 C - NUCLE AR (CORE CRIT Call LIMIT N NI Ul 8 ASE D ON G.E. CO. 8WR LICENSING LT LIN E l I '

l AFTER A*. 8'. C*.- CORE BE LT1.WMit$ AFTE R

{ ago $HIFT

% f AN ASSUMED 44*F TEMP.sMir7

$ FROM AN INITIAL PLANT RT o 1200 -

g l og gog, # -

t

$ VE$$E L j, l I

l I

>- OfSCON.

d TIN UITY ' l l .

N 1000 -

LIMIT $ g I i

y I

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8 i G

I I I

$ a00 - I l I g

E r p I _

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. GOO -

S f _

N EW 10CFR50119831 7 APPENDIX G LIMIT $

400 -

( ~ h I hy

• 50LTUP - 312 ps.g LIMIT A 70*F =>.

O ~

200a '"*

I l 1 l l .

]

O 100 200 300 400 500 600 700 MINIMUM RE ACTOR VESSEL MET AL TEMPER ATURE l'F

)

l NOTE CURVE $ A.8 & C ARE PREDICTED TO APPLY A$ THE Limits FOR 11 YEAR $ (8.8 EFPV) OF OPER ATION.

• MINI l

) l vs, h j EATORPRESSUREVESSELMETALfTEMPERATUREREACTORVESSELPRESSURE Figurei3.4.6.J-1 8

A [ ,

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l RIVER BEND - UNIT 1 3/4 4-22 APR 2 6 m I 1

, _ . , - , , - ~

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PLANT SYSTEMS

.1 9 p- p..

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SURVEILLANCE REQUIREMENTS (Continued) c.

At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, cating withfire the or chemicalby:release in any ventilation zone communi-subsystem 1.

Verifying that the subsystem satisfie the in e penetration and bypass 0.05% and 1 akage testing acceptance [criteriegof lless than

'ihe test procedure guidance in Re tions C.S.a C.S.c and .5.d of Regulalerv Gui lator9 Posi-K 2, March 1978, and th .52, Revision stem flow rate is 4000 cfm + 10%. X 2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatcry Guide 1.52, Revision 2 March 1978, meets the laboratory testing criteria of Regulatory ,

f Position C.6.a of Regulatorf Guide 1.52, Revision 2, March 1978 for a methyl iodide penetration of less than 0.175% rand

~

3.

Verifying a subsystem flow rate of 4000 cfm + 10% during sub- l system operation when tested in accordance with ANSI N510-1975. \'

A d.

"After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by within 31 days after removal that a laboratory a ton C.6.b of Regulatory Guide 1.52. Revision 2, March 1978 , meets Regulatory Guide 1.52 Revision 2the laboratory testing criteri . .

penetration of less than 0.175%. , March 1978, for a methyl iodide e.

At least once per 18 months by:

1. .

Verifying that the pressure drop across the combine HEPA filters and charcoal adsorber banks is less than 7 inches aterhuge X while operating the subsystem at a flow rate of 4 00 cfm + 10% .

1 f

e RIVER BEND - UNIT 1 3/4 7-6 APR 2 61985

.. c.-

. ~ ~ ~ - ~ - ' ~ ~ ~ ~ ~'

. _ . - . +

PLANT SYSTEMS g .

SURVEILLANCE REQUIREMENTS (Continued)

Testing equipment failure during functional testing may invali-date that day's testing and allow that day's testing to resume anew at a later time, providing all snubbers tested with the The-representative sample selected fu the fu .

plans and shall be reviewed randomly before beginning selected the from the snubbers of each type ensure as far as practical that g egti g The review shall a;_yrepresentative of the )(

and capacity of snubbers of each type.various configura Snubbers placed in the same locations as snubbers which failed the previous functional

- . test shall be retested at but shall not be included timesample the of theplan.next functional test  ;<

If yduring the y:

functional testing, additional sampling is required f due to fail- p ure of only one type of snubber, the functional testing -

'results shoyld be limited to the type of snubber which ha functional testing.

4) 88 snubbers shall be functionally tested.For each type Three (3) snubbers of.each ance crit type are allowed not meeting the functional test accept-C' is grea r 3, the number of snubbers that failed the test equal t 2 in additional sample of that type of snubber A-3) shall be functionally tested, where "A" is thethe of tot rI number of) snubbers failed during the functional test X resen ative sample.

the functi test of the resample, an additional sam snubbers of the same type shall be functionally tested. The all snubbers of that type have been functionally te ,

f.

Functional Test Acceptance Criteria i

i The snubber functional test shall verify that:

1) i Activation (restraining action) is achieved within the specified

! range in both tension and compression; j

2) 2 For mechanical snubbers, the force required to initiate or main-directions of travel; andtain motion of the snubber is within the 3)

For snubbers uous load specifically required not to displace under contin-displaceme,nt.theabilityofthesnubbertowithstandlyadwithout Testing methods may be used to measure parameters indirectly or p meters other than those specified if those results can be correlated to the specified parameters through established methods.

RIVER BEND - UNIT 1 3/4 7-13 APR 2 6125

. , . . ~ ~ ~ ' * ~ ~ ~

,,- - , , . , , , - n

- , - + - - - - - ,

j PLANT SYSTEMS HALON SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.6.3 system shall be OPERABLE with the storage tankThe main contr charge weight and 90% of full charge pressure. s having at least 95% of full APPLICABILITY:

to be OPERABLE. Whenever equipment protected by the Halon systems is r'equired ACTION:

a.

With the patrol.

fire watch above required Halon system inoperable, establish an .

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable .

SURVEILLANCE REQUIREMENTS 4.7.6.3 The above required Halon system shall be demonstrated OPERABLE a.

At least once per 31 days by verifying that each valve, manual , power operated or automatic, in the flow path is in its correct position.

b.

At least and once pressure. per 6 months by verifying Halon storage tank weig c.

At least once per 18 months by:

1. .

Verifying the system actuates, manually n a receipt of a simulated actuation signal ;omatically, upon Halon bottle initiator valve actuation, a ctuAl Halon release, burning may be excluded from the test), lectro-thermal link p , g 2.

no blockage. Performance of a flow test through headers and nozz RIVER BEND - UNIT 1 3/4 7-23 APR 2 61%

ELECTRICAL POWER SYSTEMS 5

SURVEILLANCE REQUIREMENTS (Continued) state r e q r voltage and frequency shall be maintained within eElimitsdKingthistest. Within 5 minutes after com-pleting thig 24-rour test, perform Surveillance Requirement

4. 8.1.1. 2. pr, 4. a )2 ) and b)2)*. ,

p

9. N it ' ; tt.;. he auto-connected loads to each diesel generator do not exceed for diesel generator330 kwIC.for diesel generator A and B and 2600 kw 10.

Verifying the diesel generator's capability to:

a)

Synchronize with the offsite power source while the generator is loaded with its emer tion of offsite power, .gency loads upon a siniulated restora- -

b)

, Transfer its loads to the offsite power source,"and c) Be restored to its standby status.

11.

Verifying that ywith the diesel generator 4 operating in a test mode and connected to its bus, a simulafed ECCS actuation signalN

, overrides the test mode by (1) returning the diesel generator tostandbyoperationfand(2)automaticallyenergizestheemer- X gency loads with offsite power.

12.

Verifying that the automatic load sequence timers are OPERABLE with designthe i.7terval interval between for diesel eachIAload generators block within ! 10% of its and 18.

13.

Verifying that the following diesel generator lockout features -

prevent diesel generator starting only when required:

a)

ForDieselGenerators1AandIB:[

1) 2)

Diesel control panel loss o control power.

[

Starting air pressure below 50 psi.

3) Stop solenoid energized.

4)

Diesel engaged). in the maintenance mode (includes barring device

. 1

5) Overspeed trip device actuated.  !

6)

Generator backup protection lockout relay tripped.

b) For Diesel Generator IC:

1) Diesel generator lockout relays not reset. ~

2)

Diesel engine mode switch not in "AUT0" pgsition.

"If Surveillance Requirements 4.8.1.1.2.e(4).a)2 and b)2) are not satisfactorily completed, it is not necessary to repeat the preceding 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test. Instead, IB and 2600 kw for diesel generator IC for one hour or u temperatures have stabilized.

RIVER BEND - UNIT 1 3/4 8-7 '

APR 2 61985

' ~ ~ ~ ~ ~ * ~~

RADI0 ACTIVE EFFLUENTS

_ LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4

outdoor tank shall be limited to less than or eThe quantity of radioa tritium and dissolved or entrained noble gases. qual to 10 curies, excluding APPLICABILITY

At all times.

ACTION:

a.

With the quantity of radioactive material in any of the above i

s unprotected outdoor tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank; within t

i 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit; and describe ><

p<

the events active Effluent leading Releaseto Report.

this condition in the next Semiannual Radio-b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable .

SURVEILLANCE REQUIREMENTS 4.11.1.4 The quanti unprotected outdoor tank sh analyzing a represe atthe iloactive t material contained in each of the ll be determined to be within the above limit by above 7 days when radioac mple of the tank's contents at least once per rials are being added to the tank. -

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e RIVER BEND - UNIT 1 3/4 11-6 i APR 2 61985 A

- - , , _, .,, --, .,. -y.,,-w.c , - . 4,,-- e e -- -----.-m --

RADI0 ACTIVE EFFLUENTS I,1 S f, ' ' , ,

3/4.11.2 GASEOUS EFFLUENTS

.L- u " " "

• d ,a DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 ,

effluents from the site to areas at and beyond the SITE B Figures 5.1.1-1 and 5.1.3-1) shall be limited to the following:

a. For noble gases:

body and less than or equal to 3000 aress/yr to the skin b.

For todine-131, for iodine-133, for tritium, and for all ra ionuclides in particulate form with half lives greater than 8 days: ess than or equal to 1500 arems/yr to any organ. jK g

APPLICABILITY: At all times. b i

a(L M cr- M p

_ ACTION:

With the dose rate (s) exceeding the above limits, without estore the delay r release rate to within the above limit (s).

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The se rate to noble gases in gaseous effluents shall be determined to be within th and parameters in the CM above limits in accordance with the methodology NM 4.11.2.1.2 ,

The dose rate,due to iodine-131, iodine-133, tritium, and all radionuclides in particuTate form with half lives greater than 8 /

ga days in nce with the me,thodology and parameters in the -

X 00 tive samples analysis and specified performing analyses in accordance ng in Table withand the sampli 4.11.2.1.2-1.

I e

RIVER BEND - UNIT 1 3/4 11-7 APR 2 61985 I

RADI0 ACTIVE EFFLUENTS ,

GASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.4 The GASEOUS RADWASTE TREATMENT (OFFGAS) SYSTEM shall be in operation.

APPLICABILITY: Whenever the main condenser air ejector system is in operation.

ACTION: '

p, n l'

a. With GASEOUS RADWASTE TREATMENT (OFFGAS) SYST in operable for more than 7 days, prepare and submit to the Comissi n within 5 days,

' pursuant to Specification 6.9.2, a Special Repor ludes the following information: ~

1. Identification of the inoperable equipment or subsystems and the reason for inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action (s) taken to prevent b.

recurrence.

[

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS QQ **

T w&+ Cow I

4. .I 4 T Q truments specif d in the ODCM shall beMevery 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />sj

[M

~

wh ver the main condenser air ejector is in operationyto ensure that the GASE0 5 RADWASTE TREATMENT (OFFGAS) SYSTEM is functioning. -

RIVER BEND - UNIT 1 3/4 11-13

5

< TABLE 3.12.1-1 E

e, RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAN

• m g

^ ,

Number of

! , Representative ,

g Exposure Pathway Samples and Sampilng and i and/or Sample Sample locations, Type and Frequency

m s Collection Frequency of Analysis

• 1. DIRECT RADIATION 40 routine monitoring stations Quarterly , Gamma dose quarterly.

(OR1-OR40) either with two or more dosimeters or with one instrument for measuring and i

i recording dose rate continuously, placed as follows:

o an i r ing P stations, one in

} eac met logi.tal sector in the g ge al ao the SITE BOUNDARY 7

(DRI- 6-an outer ring of stations, one in each meteorological sector in '

the 6- to 8-km range from the site (DR17-DR32);

i the balance of the stations M

i (DR33-OR40) to be placed in special interest areas such N

as population centers, nearby

3 residences,-schools, and in 1 W_

1 or 2 areas to serve as control

. stations. g

2. AIR 80RNE

'PJ

.".; J.3 Radiciodine and Samples from 5 locations (Al-AS): Continuous sampler Radiof odine Cannister: '* $

Particulates i

l' [ 3 samples (Al-A3) from close operation with sasyple collection weekly, or I-131 analysis weekly. W to the 3 SITE BOUNDARY: locations, more frequently if

' h y in different sectors, of the required by dust Particulate Sampler:

T , highest calculated annual average loading. Gross beta radioactivity.

j e ~'~"* vel ' 0/Q. a " lysis following y w i

U h _ _

FINAL DRen TABLE 4.12.1-1 (Continued)

TABLE NOTATION It should be recognized that the LLD is defined as an a priori (before the fact) limit represerting the capability of a measurement system and not as an a posterori (after the fact) limit for a particular measure-e nt. Analyses shall be performed in such a manner that the stated LLDs willbehnavoidabigjsmallsamplesizes,thepresenceofinterfering nuclides, or other uncontrollable circumstances may render these LLDs aunachievable. In such cases, the contributing factors shall be identi- _

fled and described in the Annual Radiological Environmental Operating h Report pursuant to Specification 6.9.1.7.

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e RIVER BEND - UNIT 1 3/4 12-12 6 Sg3

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l APPLICABILITY E BASES 4.0.1 This specification provides that surveillance activities necessary ,

to ensure the Limiting Conditions for Operation are met and will be performed '

during the OPERATIONAL CCNDITIONS or other conditions for which the Limiting Conditions for Operation are applicable. Provisions for additional surveillance

, activities to be performed without regard to the applicable OPERATIONAL CON 01-

'. TIONS or other conditions are provided in the individual Surveillance Require-ments. Surveillance Requirements for Special Test Exceptions need only be performed when the Special Test Exception is being utilized as an exception to 1 an individual specification.

4.0.2 The provisions of this specification provide allowable 5,olerances for performing surveillance activities beyond thnse specified in the nominal surveillance interval. These tolerances are necessary to provide oper6tional flexibility because of scheduling L performance considerations. The phrase "at least" associated with a surveillance friquency does not negate this- '

allowable tolerance; instead, it permits the more frequent performance of surveillance activities.

The tolerance values, taken either individually or consecutively over t4No.

• test intervals, are sufficiently restrictive to ensure that the reliability [

associated with the surveillance activity is not significtntly degraded beyond

_ that obtained from the nominal specified interval.

- 4.0.3 The provisions of th fication set forth the criteria for determination of compliance with ABILITY requirements of the Limiting Conditions for Operation. Under th are assumed to be OPERABLE if th Leria, equipment, systems or components f1M'[ gb satisfactorily performed within ass c ted surveillance activities have been ified time interval. Nothing in this provision is to be construed as defining equipment, systems or components OPERABLE, when such items are found or known to be inoperable although still .

meeting the Surveillance Requirements.

4.0.4 shis specif' at on ensures that surveillance activities associated ti with timeaintervalLimitingprior Condit' to n fo Operation havo been performed within the specified h try nto an applicable OPERATIONAL CONDITION or other specified applicability ition. The intent of this provision is to ensure that surveillance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation.

Under the teres of this "ptcification, for example, during initial plant startup or following extended plant outage, the applicable surveillance activ-ities must be performed within the stated surveillance interval prior to placing or returning the system or eculpment into OPERABLE status.

m APR 2 61965 RIVER BEND - UNIT 1 8 3/4 0-2

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F.EACTIVITY CONTR0'. SYSTEMS BASES CONTROL RODS (Continued)

Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and,therefore3this check must be performed prior to achieving criticality af ter  %

com)leting CORE ALTERATIONS that could have affected t.he control rod coupling integrity. The subsequent check is performed as a backup to the initial t

demonstration.

In order to ensure that the control rod patterns can be followed and X therefore that other parameters are within their limits, the control rod # X position $ndicationsystemmustbeOPERABLE. .

The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawal is j

less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.

The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.

3/4.1.4 ROD PATTERN CONTROL SYSTEM

. . s ,

The rod withdrawal limiter system input powe

, e i nal)o inates from the r h- h first stage turbine pressure. When operating with t a bypa valves open, this signal indicates a core power level which is 'es the true core power. Consequently, near the low power setpoint and high power setpoint of the rod pattern control system, the potential exists for non-conservative control rod withdrawals. Therefore, when operating at a sufficiently high power lav small probability of violating fuel Safety Limits during licensino-bas rod withdrawal error transient. a &m .5?

To ensure that fuel Safety Limits are not v olated, this specification h prohibits control rod withdrawal when a biased power signal exists and core power exceeds the specified level.

Control rod withdrawal and insertion sequences are established to assure that the maximum ir@,equence individual control rod or control rod segments which h are withdrawn at ariy' time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event'of a control rod drop accident. The specified sequences are characterized by hontogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER it greater than 20% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at rate e[ velocity limiter, could result in a peak s

enthalpy o pp/7pm.Qn, c

POWER is le than or equal t ht? iring the RPCS to be OPERABLE when THERMAL pp 20% of RATED THERMAL POWER provides adequate control.

l RIVER BEND - UNIT 1 8 3/4 1-3 APR 2 61985

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REACTIVITY CONTROL SYSTEMS FINdamND=rr g BASES l

ROD PATTERN CONTE 01 M SIEM (Continued) - jg TheRPCS/provid au matic supervision to assure that out-of-sequer>ce rods M[4 will not be w'thdraw o inserted. I

?

The analysis of the rod drop accident is presented in Section /15. j of the FSAR and the techniques of g analysis are presented in a topical repbrtk h f;. - f , and two supplementsj-.....____......

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 setpoir.t.

restricts the withdrawal distances of all non-peripheral control rods. This restriction is greatest at highest power levels.

3/4.1.5 STAND 8Y LIQUID CONTROL SYSTEM The standby liquid control syst .'p the reactor from full power to a col , enor l-freedes shutdown, a backup capability assuming that the for bringing withdrawn control rods remain fixed n he 'ated power pattern. To meet this

[ objective it is necessary to inject a tity of boron which prodqces a concen-( tration of 660 ppm in the reactor core in approxi.nately g90 to 120$ minutes.

A minimum available quantity o' 3542 gallons of sodium pentaborate solution containing a minimum of 42461)s. of sodium pentaborate is required to meet a shutdown requirement of 3% Ak/k. Thereisanadditionalallowanceof(150 ppm )<

in the reactor core to account for imperfect mixing and the ffiling of o&}r piping systems connected to the reactor r?ssel. The tiene requirement vu c

seleJed to override the reactivity insertion rate due to cooldown tallowing ,

theg enon poison peak,and the required purping rato is 41.2 gpm. The minimum /

storage volume of th( solution is established tc allow for the portion below the pump sv: tion that cannot be inserted. The temperature requirement is necessary to ensure that the sodium pentacorate remains in solution.

With redundant pumps and explosive injectior, 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 one of the redundant components inoperable. I

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

2. C. J. Paone, R. C. Stirn and R. M. Young, Supplement 1 to NED03 10527, July 1972

3. J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2, " Exposed Cores,"

Supplement 2 to NEDO-10527, January 1973 l

I RIVER BEND - UNIT 1 0E f B 3/4 1-4 l

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POWER DISTRIBUTION LIMITS

h. a Y BASES Q~{O 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as spe:ified in Specification 3.2.3 are derive from the established fuel cladding integrity Safety Limit MCPR of 1.06 and an analysis of abnormal y' operational transients. For any abnormal o erating transient analysis M_r 7c tsen with the initial condition of the reactor being at the steady state ' K operating limit, it is required that the resulting MCPR does not decrease below the Safety setting i.initinMCPR given at any time Specification 2.2. during the transient assuming instrument trip

' [

To assura that the fuel cladding integrity Safety Limit is not exceeded during any anut.f,nated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which_ result in the largest reduction in CRITICAL POWER RATIO

. flow, increase in pressu(CPR). The type of transients evaluated were loss of temperature decrease. re and power, positive reactivity insertion, and coolant .

The limiting transient yields the largest delta MCPR.

When added to the Safety Limit MCPR of 1.06, the rpquired minimum operating limit MCPR of Specification 3.2.3 is obtained ane or m .,tw in i re 3.2.3-1.

The power-flow map of Figure B 3/4 2.3-1 sho 4 t'ypical region 6of lant opcration.

' The evaluation of a given transient begin

^

h LE Rem initial para-

n. W

/'

( metersshowninFSARTable15.0-2thatareinputtoaGE[redynamicbehavior transient computer prog )(

described in NE00-24154 y . The code used to evaluateand the program used in non pre described in NE00-10802(2) The outputs of this progr pressurization events is MCPR form the input for further analyses of the thermal limiting along with the initial  %

thedingle bundle with /

NEDE-25149gannel. transient thermal hydraulic TASC code described in MCPR caused byThe the principal transient. result of this evaluation is the reduction in The purpose of the MCPR f and MCPRp of Figures 3.2.3-1 and 3.2.3-2 is to define operating limits at other than rated core flow and power conditions.

At less than 100% of rated flow and power the required MCPR is the larger value of the MCPRf and MCPR p at the existing core flow and power state. The MCPR s f

are that the established 99.9% MCPR to limit protect the core requirement canfrom inadvertent core flow increasesjsuch be assured. [

thecorrespon8fngTHERMALPOWERalongth/105%oThe MCPR s were ca the limiting bundle's relative power was adjus;t rated steam-flowg ontrol line, X above the Safety Limit. Using thise relati ntilkheMhRwasslightly power, the MCPRs were calcu- X lated at different points along the 105%-of r 3 team flow control line corresponding to different core flows. The calcu ated F;.PRzat a giten point /

of core flow is defined as MCPRf .

O f

RIVER BEND - UNIT 1 B 3/4 2-4

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FINAL 0%ET 3/4.3 INSTRUMENTATIOy RASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a. Preserve the integrity of the fuel cladding '

X j

b. Preserve the integrity of the reactor coolant system j /C

c. Minimize the energy which must be adsorbed following a loss-of-coolant accident, and .

d. Prevent inadvertent critica .

g This specification provides t e[m tir to preserve the ability of the sys em td performits inte during periods when instrument chan t

itions f e ation necessary e f nction even [

s may St out of ser decause of main-tenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of four logic channels. The logic channels A(A1) and C(A2) comprise one trip system and the logic channels B(B1) and 0(B2) comprise the other trip system for determining compliance with technical specifications. Placement of either logic channel of a trip system in the tripped condition places the trip system in the tripped condition. The trip systems;  !

as defined above,are independent of each other. There are usually four instrument )C channels (one 16 each logic channel) to monitor each parameter. The tripping of a logic channel in each trip system will result in a reactor scram.

The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are com-pleted within the time limit assumed in the safety analyses. No credit was taken for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or channel response time as defin(ed. total channel Sense responsetest timemeasuremen verification may , provided be [

such tests de demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utiliting replacement sensors with certified response times.

e 5

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

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INSTRUMENTATION BASES 1

l 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used '

to mitigate the consequences of accidents by prescribing the OPERABILITY require-ments, trip setpoints and response times for isolation of the reactor systems.

When necessary, one channel may be inoperable for brief intervals to conduct required surveillance. Some of the trio settings g have tolerances explicitly stated where both the hig ow values are critical and may have a substantial / l effect on safety. The se s f other instrumentation, where only the high or low end of the setting a frect bearing on safety, are established at a level away from the normal ope ting range to prevent inadvertent actuation *

of the systems involved. 4 a

)

Except for the MSIVs, the safety analysis does nat address individual sensor response times or the response times of the logic systems to which the sensors are connected. For 4.C. operated valves, a 3 second delay is assumed before the valve starts to move. For A.C. operated valves, it is assumed that the A.C. power supply is lost and is restored by startup of the emergency diesel generators. In this event, a time of 10 seconds is assumed before the valve starts to nove. In addition to the pipe break, the failure of the D.C. opeR.- ><

gated valve is assumed; thus the signal delay (sensor response) is concurrent 7 )(

With the 10 second diesel startup. The safety analysis considers an allowable dtionwiththe10seconddelay. inventory It followsloss thatin each case which in turn determines the Cp the 10 second time for emer0ency power establish checking the valve speeds and

)P time for the isolation functions. However, to enhance overall system reliability and to monitor instrument channel response time trends, the isolation actuation instrumentation response time shall be measured and recorded as part of the ISOLATION SYSTEM RESPONSE TIME.

Operation with a trip set less conservative than its Trip setpoint but within its specified Allowable Value is acceptable on the basis that the differ-ence between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the ,

OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the differ-ence between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety anal,ses.

RIVER BEND - UNIT 1 B 3/4 3-2

INSTRUMENTATION

• BASES <

3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION {

i The reactor core isolation cooling system actuation instrumentation is

{

provided to initiate actions to assure adequate core cooling in the event of l reactor isolation from -its primary heat sink and the loss of feedwater flow to i the reactor vessel without providing actuation of any of the emergency core  !

cooling equipment.

I Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.-

3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION -

The control rod block functions are provided consistent with the require-ments of the specifications in Section 3/4.1.4, Rod Pattern Control System, Section 3/4.2, Power Distribution Limits and Section 3/4.3, Instrumentation.

The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION I i

The OPERABILITY of the r iat on oring instrumentation ensures that; g M (1) the radiation levels are entinu* y asured in the areas served by the individual channels; (2) the larm or au atic action is initiated when the radiation level trip setpoint d; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63 and 64.

3.4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monijpring instrumen atiop ensures that sufficient capability is avai ble t rushpt2) deterni he* magnitude of a seismic event and evaluat respon e of those features important to safety.

This capability is require to permit comparison of the measured rasponse to that used in the design basis for the unit. This instrumentation is consistent with the recommendations of Regulatory Guide 1.12 " Instrumentation Yor Earth-quakes", April 1974.

I RIVER BEND - UNIT 1 8 3/4 3-4

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3/4.4 REACTOR COOLANT SYSTEM FIMLDEJT BASES 3/4.4.1 RECIRCULATION SYSTEM f

Operation with one reactor M ' coolant recirculation loop inoperable is prohibited until an evaluation rthe performance of the ECCS during one , loop operation has been performed.

[

' ' M andVdeterstined to be acceptable.

K suk ofenuilon us kun An inoperabl4 jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident,

' increase the blowdown area and reduce the-capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.

1 Jet pump failure can be detected by monitoring jet pump performance on a

, prescribed schedule for significant degradation. Recirculation loop flow mismatch limits are in compliance with ECCS LOCA analysis design criteria. -

The limits will ensure an adequate core flow coastdown from either recircu-lation loop following a LOCA. '

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50*F of each other

' +

prior to startup of an idle loop. The loop temperature must also be within 50*F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Since the coolant in the bottom of t essel is at a lower temperature than the coolant in the upper regions of differenci re, undue stress on the vessel would result if the temperature.

g eater than 100*F. h (t M /<.

3/4.4.2 5 / RELIEF VALVES #8 8

The safety valve function of the safety / relief valves (SRV) p s.;.c to '

prevent the reactor coolant system from being pressurized above tkSafe9y imit 3 of 1375 psig in accordance with the ASME Code. A total of 9 OPERAB C -. ty-relief valves is required to limit reactor pressure to within ASME III allowable h

values for_the worst case upset transient. Any combination of 4 SRVs operating in the relief mode and 5 SRVs operating in the safety mode is acceptable.

Demonstration of the safety-relief valve lift settings will occur only j during shutdown and will be performed in accordance with the provisions of Specification 4.0.5. ,

7

'^'

The low-low set minimized for a secon s ensures that safety / relief valve discharges are pening of these valves, following any overpressure N

transient. This is ac leved by automatically lowering the closing setpoint of

l. 5 valves and. lowering he opening setpoint of 2 valves following the initial opening. In this way the frequency and magnitude of the containment blowdown duty cycle is substan ially , reduced. Sufficient redundancy is provfded for the low-low set ste such that failure of any one valve to open or close at its reduced setpoin does not violate the design basis. X

, APR 2 61985 RIVER BEND - UNIT 1 8 3/4 4-1 l

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REACTOR COOLANT SYSTEM g

BASES 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 boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems", May 1973.

3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been. based on the predicted and experimentally observed behavior of cracks i ~ es. The capability of the instrumentation for detemining system he ;ectionttM-leakagnormal w 3 1 Lo considered. The evidence obtained from experiments suggests tha eakage  %

somewhat greater than that specified for UNIDENTIFIED LEAKAGE,the probability  ;<

is small that rapidly. Ho the imperfection or crack associated with such ' leakage would grow or the 1 a ver, in all cases, if the leakage rates exceed the values spe.:ified located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shu own o allow further investigation and corrective action.

C The lance Requirements for RCS pressure isolation valves provide rQbb h added assurance of valve integrity thereby . reducing the probability of gross 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 coolant is low

• thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION. b Dubng sbutdown 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.

t 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.

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

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3/4.7 PLANT SYSTEMS Flff M r n aJ BASES 1

3/4.7.1 STANDBY SERVICE WATER SYSTEM The OPERABILITY of the service water system and ultimate heat sink ensure that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling l capacity of these systems, assuming a single re, is consistent, pith the X assumptions used in the accident y ;withinacceptablelimitsg K 3/4.7.2 MAIN CONTROL ROOM AIR CONDITIONING SYSTEM

~

The OPERABILITY of the main control room air conditioning system ensu es that(1)theambientairtemperaturedoesnot,exceedtheallowabletemperature /

for continuous duty rating for the equipment and instrumentation cooled by thissysteman(2)thecontrolroomwillremainhabitableforoperationsper- K sonnel during ahd f611owing all design basis accident conditions. Continuous operation of the system fwith the heaters OPERABLE for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during each  %

31 day period# is sufficient to reduce the buildup of moisture on the adsorbers %

and HEPA filters. The OPERABILITY of this system y in conjunction with control /--

room design provisions f is based on limiting the radiation exposure to personnel %

occupying the control room to 5 rem or less whole body, or its equivalent.

M This limitation is consistent with the requirements of General Design Cr' erien f(

( 19 of Appendix "A", 10 CFR Part 50. '

3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vesselj without requiring .y actuation of any of the Emergency Core Cooling System equipment. The RCIC system is conservatively required to be OPERABLE whenever reactor pressure exceeds 150 psig. This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring the RCIC system.

The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2 and 3 when reactor vessel pressure exceeds 150 psig because RCIC is the primary non-ECCS source of emergency core cooling when the reactor is pressurized.

With the RCIC system inopera e adequate core cooling is assured by the i OPERABILITY of the HPCS system 'jkitifies the specified 14 day out-of- [ l service period. / A I

l APR 2 61985 RIVER BEND - UNIT 1 8 3/4 7-1

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3/4.8 ELECTRICAL POWER SYSTEMS ,

FMAL DONm

~

BASES 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES and ONSITE POWER DISTRIBUTION SYSTEMS '

1 The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient p will be available to supply the safety e ted equipment iequired for the safe X shutdown of the facility and the mitigation and control o accident condi-tions within the facilit he minimum specified independent and redundant  %

A.C. and D.C. power so es a d distribution systems satisfy the requirements of General Design Crit rigl7 hf Appendix "A" to 10 CFR 50.

(3

~

The ACTION requiri. 11 A specified for the levels of degradation of the er sources provide restriction upon continued facility operation commen-sur te w th the level of degradation. The OPERABILITY of the power sources iS .aee-an nsis ent with_the initial condition assumptions of the safety analyses X ace {b ed upon maintaining at least Division I or II of the onsite A.C.

d .C.

ower sources and associated distribution systems OPERABLE during acci en conditions 6g~L) $

coincident with an assumed loss of offsite power and single failure of the other onsite A.C. or D.C. source. Division III supplies the high pressure core spray (HPCS) system only.

The A.C. and D.C. source allowable out-of-service times are based on Regulatory Guide 1.93, " Availability of Electrical Power Sources", December 1974. When diesel generator 1A or 18 is inoperable, there is an additional ACTION requirement to ve fy that all-required systems, subsystems, trains, components and devices at depend on the remaining OPERABLE diesel generator J$

1A or 18 as a source emergency power, are also OPERABLE. This requirement is intended to provide assurance that a loss of offsite power event wil.1 not result perio a complete loss of safety function of critical systems during the iesel generator 1A or 1B is inoperable. The term verify as used in this c textymeans to administratively check by examining logs o,r other ,/

information to determine if certain components are out-of-service for mainte- K nance or other reasons. It does not mean to perform the surveillance require-ments needed to demonstrate the OPERABILITY of the component.

The OPERABILITY of the minimum specified A.C. and D.C. power sources and ass ated distribution systems during shutdown and refueling ensures that the facility can be maintain d in the shutdown or refueling condition for  %

extended time periods and (

is available for monitori g and maintaining the unit status. sufficient instrumentation Y and The surveillance requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guide 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Supplies", March 10, 1971, and Regulatory Guide 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants", Revision 1, August 1977.

l APR 2 61985 RIVER BEND - UNIT 1 B 3/4 8-1 7_ _. - .---.7

ELECTRICAL POWER SYSTEMS d.

BASES 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors are protected by demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers by periodic surveillance.

The surveillance requirements applicab lower voltage circuit breakers and fuses provides assurance of breaker ao fuse eliability by testing at least -

g one representative sample of each manufac1 ure ri fuse. Each manufacturer's molded case anc me b and of circuit breaker.an d or fuses are grouped into representative sample case circuit breakers and6fr-  %

basis to ensure that all breakers and/gp fuses are tested.which are variety If a wide em tested on a rotat exists within any manufacturer's branc of circuit breakers syefor fuses, it is  % l necessary to divide that manufacturer's breakers gor ses into groups and K treat each group as a separate type of breaker or fus for surveillance purposes.

>L ,

'K '

The reactor protection system (RPS) electric power monitor n assemblies provide redundant protecticn to the RPS,and other systems '

ceive power V from the RPS buses,by acting to disconnect the RPS from the power source cir- K cuits in the prese(nce of an electrical fault in the power supply.

C .

S l

RIVER BEND - UNIT 1 B 3/4 8-3

. -~. .

M REFUELING OPERATIONS BASES FINAL DaYt ?

3/4.9.7 CRANE TRAVEL - SPENT AND NEW FUEL STORAGE, TRANSFER AND UPPER CONTAINMENT FUEL POOLS The restriction on movement of loadsjin excess of the nominal weight of a T fuel assemblyI over other fuel assemblies T~n the pools, ensures that,in the event M this load is droppe 9 the activity release will bedimited to tifat contained 7' in 123 fuel rod $ any possible distortion of fuel in the storage racks X will not result n a critical array. This assumption is consistent with the activity release assumed in the safety analyses.

3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL -SPENT FUEL STORAGE AND UPPER CONTAINMENT FUEL POOLS The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. This minimum water depth is consis-tent with the assumptions of the safety analysis.

3/4.9.10 CONTROL ROD REMOVAL These specifications ensure that maintenance or repair of control rods or control rod drives will be performed under conditions that limit the probability of inadvertent criticality. The require ents fpr simultaneous removal of more C than one control rod are more stringent ' i he 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 OP'ERABLE .

and in operation or that an alternate method capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation ensuresthat(1)sufficientcoolingcapacityisavailabletoremovedecayheat Y and maintain the water in the rea:: tor pressure vessel below 140*F as required duringREFUELING,and(,2)sufficientcoolantcirculationwouldbeavailable 6 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 liqui.d 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 that a single failure of the operating loop will not result in a complete loss of residual heat removal capability. '

the reactor vessel head removed and 23 feet of water above the reactor es el flange, a large heat sink is available for core cooling. Thus, h the ev nt f ilure of the operating RHR loop, adequate time is provided to initi te3al rnate methods capable of Necay heat %g

t. ,

removal or emergency procedures to 6 he core. 7 APR 2 61985 RIVER BEND - UNIT 1 B 3/4 9-2 n..-.,.... ....u. . . . . . .. . ..- ---.. . ...- ..

~ . - - -- -- -

) l

~# }

3/4.10 SPECIAL TEST EXCEPTIONS "" ' M E BASES 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY /ORYWELL INTEGRITY The requirements for PRIMARY CONTAINMENT INTEGRITY and DRYWELL INTEGRITY

• are not applicable during the period when open vessel tests are being performed during the low power PHYSICS TESTS. -

3/4.10.2 R00 PATTERN CONTROL SYSTEM In order to perform t required in the technical specifications it is necessary to bypass se g

additional surveillance requi Jene restraints on control rod movement. The f(

lentsjensure that the specifications on heat

• generation rates and sh down b i requirements are not exceeded during the '

period when these tests a performed and that individual rod worths do not exceed the values assumed in the safety analysis.

3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur. These additional restrictions are specified in this LCO.

3/4.10.4 RECIRCULATION LOOPS This special test exceptio 4 -

conditions and is required t# gpermits peactor criticality under no flow at low THERMAL POWER levels. ferfort'yeftain startup and PHYSICS TESTS while [

3/4.10.5 TRAINING STARTUPS This special test exception permits training startups to be performed, with the reactor vessel depressurized at low THERMAL POWER and temperaturej while controlling RCS --M temperat nefHRsubsystemalignedinthe shutdown cooling mode.'- - ATn9T g with g & contaminated water dis h radioactive waste disposal system. c arge to the

[

APR 2 61985 RIVER BEND - UNIT 1 B 3/4 10-1

l RADIOACTIVE EFFLUENTS '

BASES

{

ja"M'yim%.e 7" .s e7

.ma443pP3 i b liquid effluents are consistent with the methodology provided in w us Regulator 3uide 1.109, " Calculation of Annual Doses to Man from Routine Releases ac Effluents for the Purpose of E luating Compliance with 10 CFR Part 50, v l Ap ix I," Revision 1, October 1977f an egulatory Guide 1.113. " Estimating b

Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

l

~ 3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM '

The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This speci-fication implements the requirements of 10 CFR Part 50.36a, General Design I Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The g :: M limits governing the use of appropriate portions of the liquid radwaste treatment system were [

specified as a suitable fraction of the dose design objectives set forth in Section II. A of Appendix I,10 CFR Part 50, for liquid effluents.

3/4.11.1.4 A VID HOLOUP TANKS -

.i .

The tan I ted in this Specification include those unprotected outdoor n h.%

tanks th t a ot surrounded by liners, dikes, or walls capable of holding '

the tank ents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

Restricting the quantity of radioactive material contained in the specified l tanks provides assurance that j in the event of an uncontrolled release of the ,/

tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water  !

supply and the nearest surface water supply in an UNRESTRICTED AREA.

3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE h

This specification is provided to ensure that t,p dose #at any time at and X beyond the SITE BOUNDARY 3 from gaseous effluents wilP be within the annual dose X limits of 10 CFR Part 2U to UNRESTRICTED AREAS. The annual dose limits are the Table doses associated II, Column with the concentrations of 10 CFR Part 20, Appendix 8,

1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outs:Ide the SITE BOUNDARY, to annual average concentrations exceeding the limits spetified in Appendix B, Table II of 10 CFR Part 20. 0 0 C " ^ i. C T 'f " . For MEMBERS N OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of RIVER BEND - UNIT 1 B 3/4 11-2 l

=

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring program required by this specifi-cation provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lea'd to the highest potential radiation exposures of HEMBERS OF THE PUBLIC resulting f tation operation. This monitoring program implementsSection IV.B.2 f Appe'ndi I to 10 CFR Part 50 and thereby supplements the radiological l efflyuent onitoring program by verifying that the measurable concentrations  %

of radio tive materials and levels of radiation are not higher than expect.ed r( E-asis of the effluent measurements and the modeling of the environmental ,

exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring.

The initially specif_ied monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms.of the lower limits of detection (LLDs). The LLDs required by Table 4.12.1-1 are considered optimum for routine environmental measurements

C, in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a, posteriori (after the fact) limit for a particular measurement.

Detailed discussion of the LLD, and other detection limits, can be found 1

' in HASL Procedures Manual, HASL-300 (revised annually); C for Qualitative Detection and Quantitative Determinatioh urrie, L. A. , to Application " Limits Raoio-chemistry," Anal. Chem. and Hartwell, J. K. " Detection Limits )<,

for Radioanalytical Counting 40,586-93(1968);ktlanticRichfieldHandfordCompany Techniques,"

Report ARH_-SA-215 (June 1975).

RIVER BEND - UNIT 1 B 3/4 12-1 APR 2 61985

4 1

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l:

RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4/12.2 LAND USE CENSUS 4 g his specification is provided to ensure that changes in the use of areas q M h

, beyond the SITE BOUNDARY are identified and that modifications to the 4

radiological environmental monitoring program are made if required by the /

results of this census. The best information from the door-to-door survey, or from consulting with local agricultural authorities from be shall aerial survey,is used. Th census satisfies the requirements of Section IV.B.3 of/

Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made:

garden was used for gr ing broad leaf ugetation (i.e. , simila(r,1) to20% of' the lettuce K andcabbage),and(2) 'egetation yield'be 2 kg/m2 , y NTERLABORATORY COMPARISON PROGRAM p

equirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environment monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10~CFR Part 50.

i 9 l l

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2 APR 2 e ;ggg R'.VER BEND - UNIT 1 8 3/4 12-2

. 5.0 DESIGN FEATURES Fnfal. D.v d, ,

5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.

LOW POPULATION ZONE 5.1. 2 The low population zone shall be as shown in Figure 5.1.2-1.

MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASE005 AND LIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluen >(

will allow identi[ication of structures and release points " #js, definition y of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figures 5.1.1-1 and 5.1.3-1.

5.2 CONTAINMENT PRIMARY CONTAINMENT 5.2.1 The primary containment is a steel structure composed of a vertical right cylinder and a totispherical dome. Inside d at the bottom of the containment is a reinforced concrete drywell compose steel hpad. Primary conta[rwent contai C'fuppression pool $dlitiected to the vertical right cylinder and a ,

roximately 20 feet wateren nMNX C-) 11 through a series of horizontal '%

vents. The primary containment has a minimum net free air volume of 1,190,000 cubic feet. The drywell has a minimum net free air volume of 236,000 cubic feet.

DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment and drywell are designed and shall be maintained-for:

a. Maximum internal pressure:

1. Drywell 25 psig.

2. Primary Containment 15 psig. -

b. Maximum internal temperature:

1. Drywell 330*F.

2. Suppression pool 185*F.

c. Maximum externalgto; internal differential pressure:

1. Drywell 20 psid. [

2. Primary Containment 0.6 psid.

SECONDARY CONTAINMENT

• 5.2.3 The secondary containment consists of the shield building, t'he auxiliary building and the fuel building. Secondary containment has a minimum free volume of 2,278,000 cubic feet.

RIVER BEND - UNIT 1 5-1 6 1985

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 gf 5.6. . e spent fuel storage racks are designed and shall be maintained with:

a. A k,77 @ ' M O less than or equal to 0.95 when flooded with j )<

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 g storage spacing of 7 in..

within rows and 12.25 in. between rows in the Low Density Storage

/

Racks. -

c. A_Juel assembly minimum centerpogcent Cwith a neutron noison materni aer. wee a s acing of"6.28 i t X-storger ace in the High it, Density Storage Racks. y T The storage of spent fuel in the upper contai ent fuel storage pool is prohibited dur.ing normal operation.

N 5.6.1.2 The K for new fuel for the first core loading spent fuel stoke racks shal/be administrative 1y controistored dryexceed in the Y led to not 0.98 when assumed.

optimum moderation (feam, spray, fogging, or small droplets) is 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 new fuel, stored in the new fuel storage facility, could exceed 0.98.

for DRAINAGE 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 5.6.3 The spent fuel storage pool in the fuel building is designed and shall '

be maintained with a storage capacity limited to no more than 2680 fuel assemblies.

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 kPR 2 61985

FINAL DRAFT ADMINISTRATIVE CONTROLS 6.3 UNIT STAFF.QU LIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifica-tions of ANSI /ANS 3.1-1978 for comparable positions, except for the Supervisor-Radiological Guide Programs 1.8, September 1975.who shall meet or exceed the qualifications of Regulatory The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees.

6.4 TRAINING 6.4.1 A retraining and replacement training pro be maintained under the direction of the Managergram for tht u p taff shall Administra p exceed the requirements and recommendations of Section 5.5 t ion 4sh 1 meet or [y and Appendix A of 10 CFR Part 55 and the supplemental requirem AN /ANS 3.1-1978 ts specified in Sections A and C_of Enclosure 1 of the March 28 , 1980 NRC letter to all experience. and shall include familiarization with relevant industry operational licensees, i

6. 5 REVIEW AND AUDIT 6.5.1 FACILITY REVIE'W COMMITTEE (FRC) l FUNCTION 6.5.1.1 related to nuclear safety.The FRC shall function to advise the Plant Manager on a COMPOSITION 6.5.1.2 The FRC shall be composed of the:

Chairman:

Member:

Assistant Plant Manager-Technical Services Member: Assistant Plant Manager-Operations, Radwaste and Chemistry l Me:tber:

Assistant Plant Manager-Maintenance and Material l General Operations Supervisor Mnber: Supervisor-Radiological Program Fember: Reactor Engineering Supervisor ALTERNA1ES 6.5.1.3 All alternate members shall be appointed in writing by the FRC Chairman a to serve on a temporary basis; however, no more than two alternates shall partici-pate as voting members in FRC activities at any one time. l MEETING FREQUENCY

• 6.5.1.4 .

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

RIVER BEND - UNIT 1 6-7 W 2 61!ES 1

1 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 alternate and four members including no more than two alternates.

RESPONSIBILITIES '

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

a. Review of all plant general administrative procedures and changes thereto;

b. ~

Review of all proposed tests and experiments that affect nuclear safety; c.

Review of all proposed changes to Appendix A Technical Specifications; d.

Review of_all proposed changes or modifications to structures, compo-nents, systems or equipment that affect nuclear safety; e.

Investigation of all violations of the Technical Specifications, N*including the preparation and forwarding of reoorts covering evaluation K.

%{'6 andrecommendationstopreventrecurrence,tothNuclearReviewBoard; /,

Review of all REPORTABLE EVENTS; g.

' of unit operations to detect potential hazards to nuclear saf l .y.

4/temsthatmaybeincludedinthisreviewareNRCinspectionreports, .ng QA its/ surveillance reports of operating and maintenance activities , *)

audit results, and American Nuclear Insurer (ANI) inspection results; h.

Performance of special reviews, investigations, or analysesfand reports N thereon as requested by the Plant Manager or the Nuclear Review Board [ K and j

1. Review of initial start up testing phase start-up procedures and revisions.

6.5.1.7 The FRC shall:

a.

Recommend in writing to the Plant Manager approval or disapproval of 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.

c.

Providewrittennotificationwithin24hourstotheVicepresident-RBNG and the Nuclear Review Board of disagreement between~the FRC and the Plant Manager; however, the Plant Manager shall have respon-sibility for resolution of such disagreements pursuant to Specifica-tion 6.1.1.

RIVER BEND - UNIT 1 6-8 'APR 2 61985

1 ADMINISTRATIVE CONTROL 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation.

6.13.2 Licensee-initiated changes to~the PCP:

n [

1.

Shall be submitted to the Commission in the Semiannual Radioactive EffluentThis made. Release Report submittal shallfor the period in which the changeg/

contain:

was K

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

b. A determination that the change did not reduce the overall. confor-mance of the solidified waste product to existing criteria for soli ( wastes; and

c. Documentation of the fact that the change has been reviewed and found acceptable pursuant to Specification 6.5.2.

2.

Shall become effective upon review and acceptance pursuant to Speci-fication 6.5.2.

6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by'the Commission prior to implementation.

6.14.2 Licensee initiated changes to the ODCH:

1.

/

Shall be submitted to the Commission in'the Semiannual Radioactive Effluent Release Report for the period in which the change % was

• made effective. This submittal shall contain: K

a. Sufficiently deta, led information to totally support the rationale for the change without benefit of additional or supplemental information. Inf<rmation submitted should consist of a package of those pages of the 00CM to be and provided with an approval and a i jwith each page numbered V pox, together with appro-priate analyses or evaluations ju @ig the change (s); M k Q

b.

A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and

c. Documentation of the fact that the change has been reviewed and found acceptable pursuant to Specification 6.5.2.  :

2. Shall b e eff ctive upon review and acceptance pursuant to Speci-ficatio 6. 2. hN RIVER BEND - UNIT 1 6-23

.