Text

Requests Assistance in Investigating Allegation of Util Wrongdoing.D Houston Belief That Alleged Activities Occurred Based on Assessment of Util Performance Throughout Tech Spec Development Process.D Houston 850709 & 15 Memos Encl
	ML20212K411
Person / Time
Site:	River Bend
Issue date:	07/18/1985
From:	Harold Denton; Office of Nuclear Reactor Regulation
To:	Martin R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
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 8608220050
	Download: ML20212K411 (1)
	v • d • e

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[ ~%^ UNITED STATES

<h E 8 n NUCLEAR REGULATORY COMMISSION

E WASHINGTON, D. C. 20555

%...../ July 18, 1985 MEMORANDUM FOR: Robert D. Martin, Regional Administrator Region IV FROM: Harold R. Denton, Director Office of Nuclear Reactor Regulation

SUBJECT:

REQUEST FOR AN INVESTIGATION The purpose of this memo is to request your assistance in investigating an allegation of wrongdoing by Gulf States Utilities, the applicant for River Bend.

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By memo dated July 9,1985, Mr. Dean Houston, a Reactor Engineer in the Division of Licensing, NRR, informed us of an allegation regarding the quality of certain licensing documents. In Mr. Houston's presence, an individual with professional ties to the applicant stated that the applicant's staff is aware of many errors, discrepancies and problems in the River Bend Technical Specifications. He

' further stated that Gulf States Utilities' management had instructed their staff not to pursue corrections at this time since it would hold up issuance of a license. As Enclosure 1, I am providing a copy of Mr. Houston's July 9,1985 memo.

Mr. Houston believes the alleged activities likely occurred. This is based _

on his assessment of the applicant's performance throughout the Technical Specification development process. As Enclosure 2. I am providing a second

! memo on this subject from Mr. Houston dated July 15, 1985. This memo was written at our request to document relevant background information.

Mr. Martin Virgilio of the Technical Specification Review Group, DL, has been designated as office contact for this action. Mr. Virgilio can be reached at 492-8947.

H o d R. Denton ir to h

0 ~ e of Nuclea Rea to regulation

Enclosures:

I As stated cc: J. Taylor -

B. Hayes

" i 860E220050 860815 PDR FOIA PLETTIN85-511 PDR

8'p va 'o, UNITED STATES E' ., g NUCLEAR REGULATORY CO* 11SSION wass NcTon.o. c.20sss e .

l May 21, 1985

%, . . . . / m B

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 Houstoh'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 1U. The missed errors are circled on the enclosed pages. ,,

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

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

Enclesure:

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

[ A 8^ A M __

Q)YAU DdU f5 f*

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, a 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 fucl damage is calculated to occur if the limit -

is not violated. Because fuel damage is net 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 e 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 oigration from cladding perforation is just as measurable as that from use

.f. .

relateo cracking, the thermally caused cladding perforations signal a thres-i hold beyond which still greater thermal stresses may cause gross rather than

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incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would pr'bduce onset of transi-tion boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

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 28 x 0 s/hr, bund r 'intre drop is nearly independent of bundle power and ha a,'value of 3 5 p

hus, the bundle flow with a 4.5 psi driving head t ill greater tha 28 l as/hr. Full scale ATLAS test data taken at % M[ .s'c' '

pressures from 14.7 sia psia indicate that the fuel assembly critical power at this flow i ! ately 3.35 MWt. With the design peaking facters, this corresponds to a TH RMAL POWER of more than 50% of RATED THERMAL POWER.

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

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APR 2 61985 RIVER BEND - UNIT 1 B 2-1

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REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIES FINAL. DRMT .

LIMITING CONDITION FOR OPERATION 3.1.2 The reactivity equivalence of the difference between the actual R0D DENSITY and the predicted ROD DEN 5ITY shall not exceed 1% delta k/k. _

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION: O With the reactivity equivalence [ re ,e ceeding 1% delta k/k:

O

a. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months perfo an an is to determine and explain the cause G'3 of the reactivity differe ; operation may continue if the difference is explained and corrected.

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b. 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 .

SURVEILLANCE RE0VIREMEATS 4.1.2 The reactivity equivalence of the difference between the actual ROD DENSITY and the predicted ROD DENSITY shall be verified to be less than or equal to 1% delta k/k: _

a. During the first startup following CORE ALTERATIONS, and

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

{_ . .

.D p t,61985 RIVER BEND - UNIT 1 3/4 1-2

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REACTIVITY CONTROL SYSTEMS CONTROL ROD MAXIMUM SCRAM INSERTION TIMES l 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 00me '

Pressure (psic)* 43 29 13 950 G G 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 provided 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 M M .

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 1050 0.30 W l.40 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" control rods does not exceed 5. - ..

"For intermediate reactor vessel dome pressure, the scram time riterifis determined by linear interpolation at each notch position.

/

- - Q Q.

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

L/

REACTIVITY CONTROL SYSTEMS

}

ROD PATTERN CONTROL SYSTEM b U LIMITING CONDITION FOR OPERATION 3.1.4.2 The rod pattern control system (RPCS) shall be OPERABLE. I APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2"# .

ACTION: !

a. With the RPCS inoperable or with the requirements of ACTION b, below, l not satisfied and with, THERMALPOWERlessthanorlfequal f] *fb

1. n 20% RATED THERMAL POWER control rod movement shall not be ; xcept by a scram. y pJ

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 by 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 p' known to be untrippable, this inoperable control rod may be i bypassed in the rod gang drive system (RGDS) and/or the rod A 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

1) Electrically, or -

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

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

Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose cf 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 ggg

_ TABLE 3.3.7.2-1 )

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SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE

1. Triaxial Time-History Accelerogra hs

a. Reactor Bldg O'0"

' ~~

b. Reactor Bldg 0 2 1. 0 g 1 Shield Wall EL 232'0" 0 1 1.0 g

c. Reactor Bldg Orywell EL 151'0" I

d. 0 1 1.0 g I

' Free Field - Grade Level 0 2 1.0 g 1

~ '

2. Triaxial Peak Accelerographs .

a. Reactor Bldg SLCS Storage Tank

b. 0 2 10.0 g 1 Reactor Bldg - RHR Inj. Piping 0 1 10.0 g c Aux. Bldg Service Water Piping I 0 2 10.0 g 1

, 3. Triaxial Seismic Switc

a. Reactor B1dg Ma E 70' '

O.025 to 0.25 g I I")

4. Triaxial Response-Spec Recorders

a. Reactor Bldg Mat EL 70'0"

b. Reactor Bldg Floor EL 141'0" 012g 012g I I")

c. Auxiliary Bldg Mat EL 70'0" I

d. Auxiliary Bldg Floor EL 141'0" 012g 1 --

0i2g 1

(*)With reactor control room indication and annunciation .

I

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RIVER BEND - UNIT 1 3/4 3-71 @ t 6 885

/

FINA!. DRAFT TABLE 3.3.9-1 (Continued)

ACTION 150 - a. Withonechannelinoperagle,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 -lesT'than 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SNUTDOWN within

_ ~12 hours an SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 152 - Declare t sociated Containment Ventflation System inoperabl -

gg ACTION 153 - '1. With the number of OPERABLE channels one less than required by the Minfaum 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least .

STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Provided this does not actuate the system.

i

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

,_ FINAL DiiAFT I I i l l l I 1600 - ,

A - SYSTEM MYDROTEST LIMIT w TH FUEL #N VES$tt 3 - NONWuCLE AR NE A' TING Leust 1400 -

A 8 C C - NUCLE AR (CORE CRITICALI LausT fNf bl _

RE 7

g j B ASEO ON G.E CO. SWR LICEN$iNG TOPICAL MEPORT NEDO 2t?78A

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AFTgn eg#$HIFT l A*. B'. C'. - CORE B E LTU8+t1MITS A FTE A AN ASSUMED 44*F TEMP. SHIF T I* ~

I FAOM AN INITI AL PLANT RT

$ l OF 9'F. "

t jg i I vessel i I g

OtSCON.

TINUITY ' l l *

$ 1000 - LIMITS g 8

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$ / TI N 0 -

1 I a ex I I

E p ,I _

F

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l 1 j o _

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xv L' .h m 4 _

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NEW 10CFR50 (19831 7 APPENDIX G LIMITS A= -

c 3 5W&T " '

80LTup .

312 ps.g LIMIT A 70 *F + 8 C '

200 ~

I I f f f .

t 0

N 300 400 500 soo roe 8

MINIMUM RE ACTOm VE5SEL MET AL TEMPE A ATumt 7:

NOTE Cumvf 5 A.S & C ARE PREDICTED TO APPLY AS TME Llulls FOR 11 YE Ams 18.8 EFPV) OF OPER ATION. *

q$,

MINI '[MEA /

(TOR PRESSURE VESSEL b METAL ( TE PERATURE [

Figurei3.4.6.J-1

'} l RIVER BEND - UNIT 1 3/4 4-22 APR 2 6 ses

=1

= ...

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PLANT SYSTEMS l lim a3s-id 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 chemical subsystem by:release in any ventilation zone communi-1.

Verifying that the subsystem satisfie and bypass 1 akage testing acceptance criterithe of in )- la e penetra 0.05% and %he test procedure guii nce in e hess than latory Posi- X tionsC.S.a.C.S.candq.5.dofRegul Go X

2. March 1978,andthljsystemflowrateis4000cfm+10%..52, Revision X 2.

of a representative carbon sample obtained in Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2 March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978 for a methyl fodide penetration of less than 0.175%; and ,

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3.

Verifying a subsystem flow rate of 4000 cfm + 10% during sub-

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system operation when tested in accordance with ANSI N510-1975. U A d.

'Af ter every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 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 the laboratory testing criteria of Regulatory Position C.6.a of penetration of Guide Regulatory less than1.52, Revision 2. March 1978, for a methyl iodide 0.175%.

e. _

At least once per 18 months by:

1. .

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

i ..

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

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

. . , ~ -

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PLANT SYSTEMS l '

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 for the fu .

plans and shall before reviewed be randomly beginningselected th from the snubbers of each type ensure as far as practical that a ge.egt g ;, The review shall various configurations, operating environrepresentative of the y and capacity of snubbers of each type. Snubbers ments, range of size,-

same locations as snubbers placed in the test shall be ratested at ich failed the previous functional but shall not be included 5t time of theplan.

the sample next functional test y functional testing, additional sampling is requi I during the y ure of only one type of snubber, the functional testing df due results to fail- /

shall be reviewed at the time te detemine if additional samples '

should functional betesting.

Ifmited to the type c' snubber which has failed the 4) 88 snubbers shall be functionally tested.For each type t

g, Three (3) snubbers

- of each ance' crit type are allowed not meeting the functional test accept-is grea rt

( equal t 2 3, the numbkr o' snubbers that failed the test in additional sample of that type of snubber A-3) shall be functionally tested, where "A" is snubbers failed during the functional testX thethe of totr Iresen number tiveofj sample.

the funct 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 t .

f.

Functional Test Acceptance Criteria The snubber functional test shall verify that:

1)

Activation (restraining action) is achieved within the specified range in both tension and compression; 2)

For mechanical snubbers, the force required to initiate or main-tain motion directions of theand of travel; snubber is within the specified range in both 3)

For snubbers vous load specifically required not to displace under contin-displaceme,nt.the ability of the snubber to withstand Toad without meters other than those specified if those results to the specified parameters through established methods.

RIVER BEND - UNIT 1 3/4 7-13 APR18 W

' ~ ~ ~ ~ ~ ' ~ '

6 PLANT SYSTEMS HALON SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.6.3 i system shall be OPERABLE with the storage tanThe a on main c charge weight and 90% of full charge pressure.ks having at least 95% of full APPLICABILITY:

to be OPERASLE. Whenever equipment protected by the Halone systems is

_ ACTION:

a.

With the above fire watch patrol,required Halon system inoperable , establish an hourly b.

The provisions of Specifications 3.0.3 and 3.0.4 are not ap cable.

SURVEllLANCE REQUIREMENTS 4.7.6.3 a.

The above required Halon system shall be demonstrated :

O At least once per 31 da Cs operated or automatic, ys by verifying that each valve, manual, power

b. in the flow path is in its correct position.

At least onc and pressure.e per 6 months by verifying Halon storage tank weigh c.

At least once per 18 months by:

1. %

Verifying the system actuates, manuallya receipt of a simulated actuation signal '.omatically, upon Halon bottle initiator valve actuation,g a ctup1 Halon release.

burning may be excluded from the test),(an Vglectro-thermal link

2. ,

Performance no blockage. of a flow test through headerso assure and nozzles t l

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l RIVER BEND - UNIT 1 3/4 7-23 APR 2 6 W

. l ELECTRICAL POWER SYSTEMS l

i SURVEILLANCE REQUIREMENTS (Continued)

Jhe)s,e limits dKing this test.sta e,est.ecater voltage and frequenc Within 5 minutes after com- g (4.8.1.1.2.f.4.a)2)andb)2)*.pleting thig 24-7our test, perform Surveillan w

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9. it' ; G. he auto-connected loads to each diesel generator do not exceed 330 kw for diesel generator A and 8 and 2600 kw for diesel generator IC.

10. Verifying the diesel gentrator's capability to:

., a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restora-tion of cffsite power, .

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

11.

Verifying thatjwith the diesel generatorgoperating in a test N mode,and connected to its bus, a simuland ECCS actuation signal g., overrides the test mode by (1) returning the diesel generator tostandbyoperationgand(2)automaticallyenergizestheemer- X gency loads with offsite power.

12.

Verifying that the automatic load sequence timers are OPERABLE with the interval between each load block within i 10% of its design interval for diesel generators IA and 18. _

13.

Verifying that the following diesel generator lockout features -

prevent diesel generator starting only when required:

a)

ForDieselGeneratorsIAand18:[controlpower

1) Diesel control panel loss o

2) Starting air pressure below 50 psi.

[

3) Stop scienoid energized.

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

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" Ptsition.

"If Surveillance Requirements 4.8.1.1.2.e(4).a)2 and b)2) are not seisfactorily completed, it is not necessary to repeat the preceding 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test. Instead, the diesel generator may be operated at 3130 kw for diesel generators IA and 18 and 2600 have temperatures kw forstabilized.

diesel generator IC for one hour or until operating RIVER BEND - UNIT 1 3/4 8-7 APR 2 61985

RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS h -

~

LIMITING CONDITION FOR OPERATION

3.11.1.4 The quantity of radioactive material contained in any unprotected outdoor tritium and tank shall be dissolved orlimited to less entrained noblethan or equal to 10 curies, excluding gases.

APPLICABILITY: At all times.

ACTION:

a.

With the quantity of radioactive material in any of the above unprotected outdoor tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank; within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit; and describe >(

the events leading to this condition in the next Semiannual Radio- p<

active Efflue'nt Release Report.

~

b.,

The provisfons of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS I 4.11.1.4 The quanti of r loactive material contained in each of the above analyzing a represe tat Neunprotected outdoor tankjsha 1 he determined to be with g

7 days when radioact mple of the tank's contents at least once per -

erials are being added to the tank. M_,

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RIVER BEND - UNIT 1 3/4 11-6 !

APR 2 61985 '

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

\

RADIOACTIVE EFFLUENTS '

EE 4/t6 I.

w 3/4.11.2 GASEOUS EFFLUENTS Q"" d "

>: l DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1

effluents from the site to areas at and beyond the SITE BO 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 arems/yr to the skin, b.

For iodine-131, for iodine-133, for tritium, and for all r dionuclides in particulate form with half lives greater than 8 days: ss tha,n rt K orequalto1500arems/y&oanyorgan,

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APPLICABILITY: At all times.

- (

i aL & M w W p s

f ACTION:

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

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The se rate F

3 determined to be within th and parameters in the CM to noble gases in gaseous effluents shall be above limits in accordance with the methodology N dkp

)<

! 4.11.2.1.2 .

radionuclides in particuTate torm with half lives greater a /

nce with the me,thodology and parameters in theXX 00C tive samples analysis specifiedand perfoming in Table 4.11.2.1.2-1.

analyses in accordance with the sampling a e

l g. . e e

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

RADIOACTIVE EFFLUENTS ,

GASEOUS RADWASTE TREATMENT U 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:

n

/

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a. With GASEQUS RADWASTE TREATMENT (OFFGAS) SYSTE in operable for more than7 days,prepareandsubmittotheCommissinwithin3pdays,

' ~ pursuant to Specification 6.9.2, a Special Repor at M 1udes 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 yQ -

? ,ggum ! _

4.1 4 T,hA nstruments i specif 'd in the ODCM shall be M every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months wh j ver the main c;ondenser air ejectorEn operatio to ensure that the [X GASE0 5 RADWASTE TREATMENT (OFFGAS) SYSTEM is functioni . -

4 1

RIVER BEND - UNIT 1 3/4 11-13

1 5 TABLE 3.12.1-1

5i

=

RADIOLOGICAL ENVIR00 MENTAL MONITORING PROGRAM E

g Number of

, Representative Exposure Pathway Samples and Samp11ng and Type and Frequency k and/or Sample Sample Locations, Collection Frequency of Analysis b

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

i (DRI-DR40) either with two or more dosimeters or with one instrument for measuring and recording dose rate continuously, placed as follows:

o an i r ing P stations, one in

} eac set logi.tal sector in the y ge al ao the SITE BOUNDARY T

w (DR1- 6-an outer ring of stations, one in cach meteorological sector in ,

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

I Y

the balance of the stations <y-

{ (DR33-DR40) to be placed in -

i special interest areas such

[:. 3 as population centers, nearby residences, schools, and in 1 W 3 or 2 areas to serve as control g s tatiotas.

2.

. *e.

AIR 96RNi v.)

,.3

,K3

% Radiolodine and Samples from 5 locations (Al-AS): Continuous sampler Radiolodine cannister: ' *'I Particulates 3 samples (Al-A3) from close operation with sample collection weekly, or I-131 analysis weekly. W to the 3 SITE BOUNDARY locations, more frequently if y in different sectors,,0f the required by dust Particulate Sampler:

v

, highest calculated annual average loading. Gross beta radioactivity c -- ~"avel D/Q. aa-lysis following r~T7 -

TABLE 4.12.1-1 (Continued)

FINAL D%37 TABLE NOTATION 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 posterori (after the fact) limit for a particular measure-g nt. _ Analyses shall be performed in such a manner that the stated LLDs

)

f will be#unavoidablgsmall sample sizes, the presence of interfering '

s t lJunachievable.nuclides, or other uncontrollable circumstances may render these LLDs.

In such cases, the contributing factors shall be identi- _

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

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APPLICABILITY .

FIMLD8'saWl 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 CONDITIONS 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 CONDI-TIONS or other conditions are provided in the individual Surveillance Require-ments. Surveillance Requirements for Special Test Exceptions need only be perfomed when the Special Test Exception is being utilized as an exception to an individual specification.

4.0.2 The provisions of this specification provide allowable tolerances g for performing surveillance activities beyond those specified in the nominal .

surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations. The phrase "at least" associated with a surveillance frequency does not negate this allowable tolerance; instead, it pemits the more frequent performance of surveillance activities.

The tolerance values, taken either individually or consecutively over three

test intervals, are sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not significantly degraded beyond

[ !

i that obtained from the nominal specified interval. !

4.0.3 The provisions of thi ication set forth the criteria for detemination of compliance with ABILITY requirements of the Limiting p Conditions for Operation. Under th Leria, equipment, systems or components are assumed to be OPERABLE if th ass ted surveillance activities have been

[N satisfactorily perfomed within ified time interval. Nothing in this _

provision is to be construed as defining equipment, systems or components OPERABLE, when such items are lund or known to be inoperable although still .

meeting the Surveillance Requirements.

4.0.4 This specif at on ensures that surveillance activities associated ti i

with a Limiting Condit n fo Operation have been performed within the specified time interval prior to try nto an applicable OPERATIONAL CONDITION or other h 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 terms of this specification, 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 equipment into OPERABLE status.

APR 2 61965 RIVER BEND - UNIT 1 B 3/4 0,2

=$

REACTIVITY CONTROL SYSTEMS BASES ,

Il CONTROL RODS (Continued) 1 i Centrol rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature 8' provides the only positive means of determining that a rod is properly coupled and,therefore3this check must be performed prior to achieving criticality after jN com leting CORE ALTERATIONS that could have affected the control rod coupling integrity. The subsequent check is performed as a backup to the initial demonstration.

. In order to ensure that the control rod patterns can be followed andj X therefore,that other parameters are within their limits, the control rod . )(

position indication system must be OPERABLE.

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 I rod reactii vity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage i- 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 f rods are OPERABLE and not so frequent as to cause excessive wear on the system components.

! 3/4.1.4 ROD PATTERN CONTROL SYSTEM e h- h

. . s , , e The rod withdrawal limiter system input nowe i nal o inates from the .

first stage turbine pressure. When operating with t a bypa valves l open, this signal indicates a core power level which is les the true . i

, core power. Consequently, near the low power setpoint and high power

setpoint of the rod pattern control system, the potential exists for non-t conservative control rod withdrawals. Therefore, when operating at a

sufficiently high power level; there b a mall probability of violating 4 fuelSafetyLimitsduringadicensino-bas rod withdrawal error transient. 4 4 wX%.l?

To ensure that fuel Safety Limits are olated, this specification i prohibits control rod withdrawal when a biased power signal exists and core

,{ power exceeds the spectfied level.

4 Control rod withdrawal and insertion sequences are established to assure i thatthemaximumirhequenceindividualcontrolrodorcontrolrodsegmentswhich [ .

} are withdrawn at ary time during the fuel cycle could not be worth enough to j result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control i rod drop accident. The specified sequences are characterized by homogeneous, -

scattered patterns of control rod withdrawal. When THERMAL POWER is greater !

than 20% of RATED THERMAL POWER, there is no possible rod worth whiche if dropped at s enthalpy o pp/ jn, rate TGg e velocity limiter, could result in POWER is le c 7pm.

thaa or equal 20% of RATED THERMAL POWER provides adequate a peak @ gb iirin l control.

I RIVER REND - UNIT 1 8 3/4 1-3 APR 2 61985

/

REACTIVITY CONTROL SYSTEMS FIN 2*D"7#"N9 BASES ROD PATTERN CONTED'_ TIEM (Continued) f TheRPCSproviddau matic supervision to assure that out-of-sequence rods M M- @4 will not be w'thdraw o inserted.

?

The analysis of the rod the FSAR and the techniques of drop accident is presented in analysis are presented in a topical report Section /15.j of K t fa - f , and two supplement ,, - ..... _ - . . _ . .

The RPCS is also designed to automatically prevent fuel damage in the event of erroneous power operation.

rod withdrawal from locations of high power density during higher

A dual channel system is provided that, above the low power setpoint, -

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

3/4.1.5 ~ STANDBY LIQUID CONTROL SYSTEM

~

The standby liquid control syst the reactor from full power to a col , penor o t-free

des a backup shutdown, capability assuming that the for bringing withdrawn control rods remain fixed nThe ated power pattern. To mest this objective it is necessary to inject a tity of boron which produces a concen-C trationof660ppeinthereactorcoreinapproximatelyg90to120) minutes.

A minimum available quantity of 3542 gallons of sodium pentaborate solution

containing a minimum of 4246 lbs. of sodium pentaborate is required to meet a shutdown requirement of 3% Ak/k. Thereisanadditionalallowanceof(150 ppm - X in the reactor core to account for imperfect mixing and the filling of oth)er piping systems connected to the reactor vessel. The time requirement was l seleJe ed to override the reactivity insertion rate due to cooldown following ,

thegenen poison peak and the required pumping rate is 41.2 gpm. The minimum /

storage volume of thef 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 one of the redundant components inoperable.

1.

C. 1 ~)aone, 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 I to NEDO;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 APR 2 61985 RIVER BEND - UNIT 1 B 3/4 1-4

POWER DISTRIBUTION LIMITS 3 ,

BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derive from the established fuel cladding integrity Safety Limit MCPR of 1.06 and an analysis of abnormal operational transients.

For any abnormal o erating transient analysis d r K isen 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 Limit given in MCPR at any time Specification 2.2. during the transient assuming instrument trip

[

To assure that the fuel cladding integrity Safety Limit is not exceeded-during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to detemine which result in the largest reduction in CRITICAL POWER RATIO (CPR . The type of transients evaluated were loss of e

flow, increase in presrsure an)d power, positive reactivity insertion, and co temperature decrease.

The limiting transient yields the largest delta MCPR.

When added to the Safety Limit MCPR of 1.06, the rpquired minimum operating Itait MCPR of Specification 3.2.3 is obtained anW orm. 66 ire 3.2.3-1.

The power-flow map of Figure B 3/4 2.3-1 sho operation. -

4t'ypicalregionfo ^

plant O

/'

The evaluation of a given transient begins ;^... i.a . . . . nitial para- II

( metersshowninFSARTable15.0-2thatareinputtoaG[,oredynanicbehavior transient computer prog )(

described in NEDO-24154 y . The code used to evaluate pressurization eventr is 7 described in NE00-10802(2) and the program used in non pressurization events is The outputs of this program;along with the initial %

MCPR the$1ngle form the input for further analyses of the themalTy limiting bundle with /

NEDE-25149gannel. transient thermal hydraulic TASC code described in -

MCPR caused by The the principal transient. result of this evaluation is the reduction in The purpose of the MCPR g and MCPR, 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 of the than 100% of rated flow and power the required MCPR is the larger value MCPR f and MCPR, at the existing core flow and power state. The MCPR s y

that the 99.9% MCPR limit requirement can be assured.are established K to The MCPR s were calculated such that for the maximum core flow rate and thecorresponbingTHERMALPOWERalongth/105%of"ratedsteamflowgontroll s X thelimitingbundle'srelativepowerwasadjusted above the Safety Limit. ntilkheMCPRwasslightly Using thisFrelative und e power, the MCPRs were calcu- X lated at different points along the 105%-of ra 3 1 steam flow. control line /

corresponding to different core flows. ThecalcuTatedACPRa a gifen point of core flow is defined as MCPRf.

RIVER BEND - UNIT 1 ..B. T,/4 2-4 m.. Y,

\

FINAL gggq 3/4.3 INSTRUMENTATION BASES 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

b. Preserve the integrity of the reactor coolant system /(.

j

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

~

d. Prevent inadvertent critica .

This specificatibr, provides t e[a tir [n itions r efationnecessary j)(_

to preserve the ability of the sys em perform its inte e fWnction even t

during periods when instrument chan s may throut 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 C., channels A(A1) and C(A2) comprise one trip system and the logic channels B(BI) and D(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 channels (one irr,each logic channel) to monitor each parameter. The tripping -

K of a logic channel in each trip system will result in a reactor scram.

The sensurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are com-plated 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 channelresponsetimeasdefin(ed. total channel test measuremen , provided X demo Sensor response time verification may be such tests f

demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utili2ing replacement sensors with certified response times.

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

/

INSTRUMENTATION

8ASES 1 .

l 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness vf the instrumentation used 4

j 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. Someofthetripsettings)ghavetolerancesexpli'citly stated where both the high /

'ow values are crit' cal and may have a substantial !

j effect on safety. The se oin s f other instrumentation, where only the high I or low end of the setting a f rect bearing on safety, are established at i a level away from the nor mal ope ting range to prevent inadvertent actuation 6 4, )4(

ja of the systems involved. ,

t I :

Except for the MSIVs the safety analysis-does not address individual. sensor response times or the response times of the logic systems to which the sensors are connected. For 0 4. 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. pod r supply is lost and is restored by startup of the emergency diesel j generators. In this event, a time of 10 seconds is assumed before the valve i r starts to move. In addition to the pipe break, the failure of the D.C. opeR.-

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

kith the 10 second diesel startup. The safety analysis constders an allowable i

l jnventory loss in each case which in turn detemines the valve speed in conjunc-

.tton witn T.ne 10 second delay It follows that checking the valve speeds and g*

i the 10 second time for emergency power establishment will establish the response time for the isolation functions. However to enhance overall system reliability '

j and to monitor instrument channel response, time trends, the isolation actuation instrumentation response time shall be measured and recorded as part of the j ISOLATION $YSTEM 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- i i i ence between each Trip 5etpoint and the Allowable Value is equal to or less

{ .

than the drift allowance assumed for each trip in the safety analyses.

1

! 1 3/4.3.3 EMERGENCY C0RE C0OLING 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 j

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 j to send the actuation signal to more than one system at the same time.

) :

' Operation with a trip set less conservative than its Trip 5etpb bt'bu't t

within its specified Allowable Value is acceptable on the basis that the differ-once 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. I j RIVER REND - UNIT 1 8 3/4 3-2 1

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INSTRUMENTATION 8ASES Q ._

3/4.3.4 REC 1RCULAT!0N PUWP ~ RID AC'. * ?3N INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scras during an anticipated transient. The response of the plant to this postulated event falls within the envelope of ,

study events in General Electric Company Topical Report NEDO-10349, dated March 1971 and NED0-24222, dated December 1979, and Section 15.8 of the FSAR. ,

i The end-of-cycle recirculation pump trip (E0C-RPT) system is a part of I the Reactor reactor trip.

Protection System and is an essential safety supplement to the i- - The purpose of the EOC-RPT is to recover the loss of thermal ,

margin which occurs at the end-of-cycle. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add.

positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity. Each E0C-RPT system trips both recircu-lation pumps, reducing coolant flow in order to reduce the void collapse in g the core during two of the most limiting pressurization events. The two .

i events for which the EOC-RPT protective feature will function are closure of . :

the turbine stop valves and fast closure of the turbine control valves. -

.r-

-: A fast closure sensor from each of two turbine control valves provides

~

input to the EOC-RPT system; a fast closure sensor from each of the other two 3 turbine control valves provides input to the second EOC-RPT system. Similarly, a position switch for each of two turbine stop valves provhes input to one EOC-RPT system; a position switch from each of the other two stop valves provides input to the other E0C-RPT sptea. For each E0C-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure vaTves. of turbine control valves and a 2-out-of-2 logic for the turbine stop The operation of mither logic will actuate the E0C-RPT system and -_ -

trip both recirculation pumps.

Each E0C-RPT system may be manually bypassed by use of a keyswit ch dministratively controlled. The manual bypasses and the automat !

ating j m.ss at less than 40% of RATED THERMAL POWER are annunciated in th rol 1 The EOC-RPT system response time is the time assumed in the analysis between initiation of valve action and complete supp f the electric arc, i.e., 140 ms. Included in this stM are: the espo the time allotted for breaker arc suppression and t 'ne of the sensor, e time of the {

system logic.

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 Setpcint and the Allowable Value is equal to or less than the drif t allowance assumec for each trip in the safety analyses.

l l

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

l

~

,. INSTRUMENTATION

BASES l

3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is i

provided to initiate actions 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 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

! a less than the drift allowance assumed for each trip in the safety analyses.-

3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION l '

The control rod 11ock functions are provided consistent with the require-i sents 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.

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

l j

Operation with a trip set less conservative than its Trip setpoint but within its specified Allowable Value is acceptable on the basis that the i

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, j 3/4.3.7 MONITORINC INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION j

The OPERABILITY of the r fat [ n oring instrumentation ensures that; (1) the radiation levels are entin y I

individual channels; (2) the larm or u asured atic in the areas served by the 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 i

j The OPERASILITY of the seismic rin instrianen tiop ensures that sufficient capability is vai ble t t deterni he'eagnitude of a i seismic event and evalua respon e of those features important to safety.

1 This capability is require to permit comparison of the measured response to that used in the design basis for the unit. This instrumentation.ts; cons'istent

with the recommendations of Regulatory Guide 1.12 " Instr oentation for Earth-

quakes" April 1974.

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

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

l I

3/4.4 REACTOR COOLANT SYSTEM FFua .t. m &m .;

j BASES I .

I!

t 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor g coolant recirculation loop inoperable is [

prohibited until en evaluation orthe performance of the ECCS during one loop -

V operation has been performed. " z'M and detegM

~

sukofer us beenined to be acceptable. K An inoperable 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, J 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."

Jet pump failure can be detected by monitoring jet pump perfomance on a prescribed schedule for significant degradation. Recirculation loop flow ,

mismatch limits are in. compliance with ECCS LCCA 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 p 50'F of the reactor pressure vessel coolant temperatura 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 h differenc< 4e.g g re, eaterundue than 100*F. stress on the vessel would result if the temperature.pgad jh A \

3/4.4.2 / RELIEF VALVES - -

I j, l The safety valve function of the safety / relief valves (SRV) e to '

prevent the reactor coolant system from being pressurized abeve t Safeiy imit g of 1375 psig in accordance with the ASME Code. A total of 9 OPERA u. -..ty- h relief valves is required to limit reactor pressure to within ASME III allowable 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. .

I

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

Specification 4.0.5.

, 7 The low-low set s ensures that safety / relief valve discharges are N ..

minimized for a secon pening of these valves, following any overpressure transient. This is ac feved 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 containneht biowdown duty cycle is substan ially . reduced. Sufficient redundancy is provided for the low-low set ste such that failure of any one valve to open or close at 7 its reduced setpo n does not violate the design basis. '

APR 2 61985 RIVER SEND - UNIT 1 8 3/4 4-1 r

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

l 3/4.4.3.2 OPERATIONAL LEAKAGE

i The allowable leakage rates from the reactor coolant system have been b'ased on the predicted and experimentally observed behavior of cracks t r.ormally expected background leakage due to equipment design and h

' es. The

ection' n, p

capability of the instrumentation for detemining system leakag %e 1 lo considered. The evidence obtained from experiments suggests tha eakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE,the probability )( I is small that the imperfection or crack associated with such " leakage would grow rapidly. Ho ver, in all cases, if the leakage rates exceed the values specified '

or the 1 a 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 t % 31 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. Dubngsbutdownandrefuelingoperations,thetemperaturenecessary b

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

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 thalt the chlorides are not exceeding the limits. *- -

RIVER BEND - UNIT 1 8 3/4 4-2 APR 2 619o5

. /

3/4.7 PLANT SYSTEMS FliVL D"g BASES '

I 3/4.7.1 STANDBY SERVICE WATER SYSTEM I l

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 '

capacity of these systems, assuming a single failure, is consistent, pith the ;

assumptions used in the accident Qhinacceptablelimitsg [X 3/4.7.2 MAIN CONTROL ROOM AIR CONDITIONING SYSTEM The OPERABILITY of the main control room air conditioning system ensuras that(1)theambientairtemperaturedoesnotexceedtheallowabletemperature /

for continuous uty rating for the equipment and instrumentation cooled by this system an 2) the control room will remain habitable for operations per- X 1

sonnel during a following all design basis accident conditions. Continuous operation of the system with the heaters OPERABLE for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months during each %

31 day period,is suffic,ient to reduce the buildup of moisture on the adsorbers /-

, and HEPA filt'ers. The OPERASILITY of this system y in conjunction with control /~

room design provisionsi # s based on limiting the radiation exposure to personnel /

b, occupying the control room to $ ren or less whole body, or its equivalent.

,( This limitation"A",

19 of Appendix is consistent 10 CFR Part 50. with the requirements of General Design eriet?Cr{< /C g

, 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure f

' adequate core cooling in the event of reactor isolation from its primary heat _

sink and the loss of feedwater flow to the reactor vessel .Y actuation of any of the Emergency Core Cooling System equ#without 1pment. The RCICrequiring

system is conservatively required to be OPERA 8LE 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. l 1

The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, i

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 OPERASILITY of the HPCS system htifiesthespecified14dayout-of- N service period. / A RIVER BEND - UNIT 1 8 3/4 7-1 26M l i

80 -

- - - -*/***e. e mme e,* ane w e e- e.e * , * - - e... + . ee em e- - **a * *

e-- e emme- - - * * * * * * *

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

...,.,==_.,,-******=,- .,, - ~ - , , ,

/

FIN Ka. k ~TipJ i

3/4.8 ELECTRICAL POWER SYSTEMS ,

BASES '

i 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES and DNSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient pow will be available to supply the safetyJeJated equipment l'equired for the safe X I shutdown of the facility and.E W the mitigation and control o accident condi-tions within the facilit . he minimum specified independent and redundant %

A.C. and D.C. power sot es a d distribution systems satisfy the requirements of General Design Crit tri417bfAppendix"A"to10CFR50.

The ACTION requiri

,p specified for the levels of degradation of the sur r esources provide restriction upon continued facility operation commen-iS .aee- th the level of degradation. The OPERABILITY of the power sources 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.

.C.-

d acci en ower sources and associated distribution systems OPERABLE during 6y L7pf) ,

conditions 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.

( 1974.

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 When diesel generator 1A or 18 is inoperable, there is an additional ACTION requirement to ver fy that all required systems, subsystems, trains, components 1A or 18 as a source and devices f " emergency, are also powerat OPERABLE. depend on the remaining M OPE is intended to provide assurance that a loss of offsite power event will notThis requiremen result -

perio a complete loss of safety function of critical systems during the esel generator 1A or 18 is inoperable. The term verify as used in this c textymeans to administrative 1y check by examining logs o,r other ,M><

information to determine if certain components are out-of-service for mainte-nance or other reasons. 1 ments needed to demonstrate the OPERABILITY of the component.It does ass The OPERABILITY of the minimum specified A.C. and D.C. power sources and ated distribJtion systems during. shutdown and refueling ensures that the facility extended can beand time periods maintain

( d in the shutdown or refueling condition for Y is available for monitori g and maintaining the unit status.ufficient instrumentation Y an 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 tower Supplies", March 10, 1971, and Regulatory Guide 1.108, " Periodic Test.ing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants" Revision 1, August 1977. i t

i !

APR 2 61985 :

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

. = ~ *

-== nee *~e- m -~ ~ ~- . ..<p=or+=...= ~ n==ve-* ~ * * ~ * ~ ~ ~ ~ * * * * * * ~ * * ' ~ ~ ~

ELECTRICAL POWER SYSTEMS = ;

BASES I

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES -

l Containment electrical penetrations and penetration conductors are protected i 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 er fuse eliability by testing at least g dg one representative sample of each manufac1 urer) b and of circuit breaker.asuffor f I fuse. Each manufacturer's molded case anc met case circuit breakers andG %

fuses are grouped into representative sample which are e en tested on a rotating % l basis to ensure that all breakers and fuses are tested. If a wide variety K exists within any manufacturer's bran of circuit breakers g or fuses, it is K necessary to divide that manufacturer's breakers.ajdforfuses into groups and, y-treat each group as a separate type of breaker or fusf for surveillance purposes. K The reactor protection system (RPS) electric power monitor ng_ assemblies provide redundant protection to the RPS,and other systems

~

- dceive power V from the RPS buses,by acting to disconnect the RPS from the power source cir- K cuits in~the pressfnce of an electrical fault in the power supply.

O O

. e APR 2 61985 l

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

I REFUELING OPERATIONS -

.=

BASES 3/4.9.7 CRANE TRAVEL - SPENT AND NEW FUEL STORAGE, TRANSFER AND UPPER CONTAINMENT FUEL POOLS l

The restriction on movement of loadsjin excess of the nominal weight of a o

T fuel load this assembly 2 ver9 other is droppe fuel assemblies the activity release willTn the pools bedimited t t in the event to ensures tha/at contained 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% lodine 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.

3/4.9.16 CONTROL R0D 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 provides for the core to remain subcritical

' i he SHUTDOWN MARGIN specification with only one control rod fully withdrawn.

3/4.9.11 RESIOUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal loop be OPERABLE .

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 reactor pressure vessel below 140'F as required during REFUELING, and(2) sufficient coolant circulation would be available , $

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 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 sjel flange, a large heat sink is available for core cooling. Thus, in the ev nt@ f ilure of the operating RHR loop, .

adequate time is provided'to initi te al rnate methods capable of

decay heat 6 g {

removal or emergency procedures to he core. .- I i

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

y

~

3/4.10 SPECIAL TEST EXCEPTIONS FEE'~# 'dd "' emry 4 BASES 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY /DRYWELL INTEGRITY

! The requirements for PRIMARY CONTAINMENT INTEGRITY and ORYWELL INTEGRITY -

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

3/4.10.2 ROD PATTERN CONTROL SYSTEM In order to perform t required in the technical specifications it i, is necessary to bypass se ene restraints on control rod movement. The,f(

g

/ additional surveillance requi nts ensure that the specifications on heat

generation rates and sh down r requirements are not exceeded during the i

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- SHUTOOWN MARGIN DEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not 1

occur. These additional restrictions are specified in this LCO.

(F- 3/4.10.4 RECIRCULATION LOOPS This special test exceptiog permits conditions and is required tQerfor&[EeIpeactor tain startup criticality under no flow and PHYSICS TESTS while 4

at low THERMAL POWER levels. [

3/4.10.5 TRAINING STARTUPS This special test exception permits training startups to be performedy with the reactor vessel depressurized at low THERMAL POWER and temperature, e with 1 while shutdown controlling RCS temperatg8 cooling mode.4- c-de- 4fn%gne pHR subsystem aligned in the IM contaminated water discharge to the 'l radiosctive waste disposal system.

.O ' .en e

i .

APR 2 s 19g5 RIVER BEND - UNIT 1 8 3/4 10-1 i-. ~

l RADI0 ACTIVE EFFLUENTS BASES

,sge .. gU e- -

/4 11 1-SE ( nti ) f fC V' j aa@$5% b 11 uid effluents are consistent with the methodology provided in Regulato 3uide1.109,"CalculationofAnnualDosestoManfromRoutineReleases[

Reac A M ix 1," Revision 1, October 1977 andEffluents for the Purpose of Ev luating Complianc f egulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases N

for the Purpose of Implementing Appendix I," April 1977.

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 -

l* 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 = W 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 i Section II. A of Appendix I,10 CFR Part 50, for liquid effluents.

3/4.11.1.4 E UID HOLDUP TANKS -

The tan 1 ted in this Specification include those unprotected outdoor o tanks th a the tank ot surrounded by liners, dikes, or walls capable of holding

- nts 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 ~

tanks provides assurance that f in the event of an uncontrolled release of the ,/

tanks' contents, the resulting concentrations would be less than the limits of 10 CD 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 GASEQUS EFFLUENTS 3/4.11.2.1 DOSE RATE

! h l

This specification is provided to ensure that y dose # at any time at and X beyond the SITE SOUNDARY,from gaseous effluents will'be within the annual dose X l

limits of 10 CFR Part 20'to UNRESTRICTED AREAS. The annual dose limits are the Table doses associated II, Column with the concentrations of 10 CFR Part 20, Appendix B,

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 outside the SITE SOUNDARY, to annual average concentrations exceeding the limits spettfied'in Appendix B. Table II of 10 CFR Part 20. O C CT" 7.>i. 2 T 'f? . For MEMBERS Y OF THE PUBLIC who may at times be within the SITE SOUNDARY, the occupancy of RIVER BEND - UNIT 1 B 3/4 11-2 l

FIN E h m.. p. l 8

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 lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting f tation operation. This monitoring program implementsSection IV.B.2 f Appa'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 expecwd asis of the effluent measurements and the modeling of the environmental n'[

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

The initially specified monitoring program will be effective for at least the first three years of cosmercial 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

, 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 &

particular measurement.

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

in NASL Procedures Manual, NASL-300 (revised annually); C for Qualitative Detection and Quantitative Determinatioh urrie, L.

Application A., " Limits to Radio-

[

chemistry," Anal. Chem. and Hartwell, J. K.,

for Radicanalytical Counting 40,586-93(1968);htlanticRichfieldHandfordCompany Techniques," " Detection Limits' f Report ARH-SA-215 (June 1975).

I i- --

RIVER BEND - UNIT 1 8 3/4 12-1

l .

FINAL cp' /

RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4/12.2 LAND USE CENSUS g g his specification is provided to ensure that changes in the use of areas q M h 4 j beyond the SITE BOUNDARY are identified and that modifications to the /

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 aerial shall survey,is be 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 -

ainimum 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: 1) 20% of the garden was used for grqwing broad leaf ugetation (i.e. , simila(r to lettuce K and cabbage), and(2) gegetation yieldfbt 2 kg/m2, y

.12f3 (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 i

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.

APR 2 61999 RIVER BEND - UNIT 1 B 3/4 12-2

. 5.0 DESIGN FEATURES flld{ DT a._

I 5.1 SITE i EXCLUSION AREA 1

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 RADI0 ACTIVE GASEOUS'AND LIQUID EFFLUENTS 5.1.3 )(

will allow identification of structures and release points "Informationregardingradioac 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 torispherical dome. Inside nd at the bottom of the containment is a reinforced concrete drywell compose steel head. Primary con reent contai vertical right cylinde roximately 20 feet nd a ,

waterin nMN h ^"htppression poolf riected to th 11 through a series of horizontal

. A. ')

, t vents. The primary containment has a minimum net free air voltme 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. - I

b. Maximum internal temperature:

1. Drywell 330*F.

2. Suppression pool 185'F. ,

l

c. Maximum externalgt ginternal differential pressure:

1. Drywell 20 ps d. [

2. Primary Containment 0.6 psid. -

SECONDARY CONTAINMENT 5.2.3 The secondary containment consists of the shield building, the 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 7985

" ~

DESIGN FEATURES .l#Al DdMT 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 j l I Y g 5.6. . e spent fuel storage racks are designed and shall be maintained with:

a. A k,ff m' '-M r 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. j

b. A fuel assembly minimum centesto-center storage spacing of 7 in.

within rows and 12.25 in, betw^een^ rows in the Low Density Storage '

[

Racks. _

c. AJuelassembivminimumcenterpogeent stora acing of 6.28 i X Cwith a neutron noison saterisi petwee sto tb

_ Density Storage Racks e aceyintheHigh The storage of spent fuel in the upper contal ent fuel storage pool is prohibited during normal operation.

^

, 5.6.1.2 The K for new fue for the first core loadin Y

.(. spent fuel stoke racks sha1/be administrative 1y contrbstored led to not dryexceed in the j 0.98 when optimum moderation (foam, spray, fogging, or small droplets) is

assumed.

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 storage facility, could exceed 0.98.

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.

l l 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT l 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

.o

/

FINAL. DRAFT !

ADMINISTRATIVE CONTROLS 1

~

6.3 UNIT STAFF. QUALIFICATIONS 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 Programs Guide 1.8, September who shall meet or exceed the qualifications of Regulatory 1975. The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all lice'nsees.

6.4 TRAINING 6.4.1 A retraining and replacement training program for th< up taff shall be maintained under the direction of the Manager-Administra ion p h 1 meet exceed the requirements and recommendations of-Section 5.5 t AN /ANS 3.1-1978 or and Appendix A of 10 CFR.Part 55 and the supplemental requirem ts specified in Sections A and C of_Entlosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience.

_ 6.5 REVIEW AND AUDIT

6.5.1 FACILITY REVIEW COMMITTEE (FRC)

(..

FUNCTION 6.5.1.1 The FRC shall function to advise the Plant Manager on all matters related to nuclear safety.

COMPOSITION ~

6.5.1.2 The FRC shall be composed of the: -

Chairman: Assistant Plant Manager-Technical Services Member:

Member: Assistant Plant Manager-Operations, Radwaste and Chemistry Assistant Plant Manager-Maintenance and Material Member: General Operations Supervisor Member: Supervisor-Radiological Program Member: Reactor Engineering Supervisor ALTERNATES '

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.

MEETING FREQUENCY ~

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

i RIVER BEND - UNIT 1 6-7 APR 2 6195LE l

ADMINISTRATIVE CONTROLS FINAL DRAFT l 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 s'afety;

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

e. - Investigation of ell violations of the Technical Specifications, includino the preparation and forwardi_no of reports covering evaluation

]*[Jhg*]hr # nd recomendations to prevent recurrence, to thMuclear Review Board; y

X.

Review of all REPORTABLE EVENTS;

g.

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

q/ tem 'that may be included in this review are NRC inspection reports.

QA its/ surveillance reports of operating and maintenance activities, '

_g audit results, and American Nuclear Insurer (ANI) inspecticn results;

h. Performance of special reviews, investigations, or analysesjand reports- K.

thereonI 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:

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.

Providewrittennotificationwithin24hourstotheVicefresident-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 ;

~ ~

~

ADMINISTRATIVE CONTROL FlMit DRMT 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:

1.

/\ [

Shall be submitted to the Commission in the Semiannual Radioactive EffluentReleaseReportfortheperiodinwhichthechangeg/

made. This submittal shall 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 solid 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.

x. 6.14 0FFSITE DOSE CALCULATION MANUAL (00CM) 6.14.1 The DDCM shall be approved by the Comission prior to implementation.

6.14.2 Licensee initiated changes to the 00CM:

y 1.

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

nade effective. This submittal shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be hwitheachpagenumbered V and provided with an approval and a pox, together with appro-priate analyses or evaluations ju ig the change (s); M 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 hb

~

e of 'ctive upon review and acceptance pursuant to Speci- d ficatio 6. 2.

RIVER BEND - UNIT 1 6-23

l m J g TABLE 3.3.7.8-1 (Continued) 1 FIRE DETECTION INSTRUMENTATION ..

OTAL INSTRUMENTS OPERABLE" l INSTRUMENT LOCATION FLAME ^ SMOKE HEAT M) M (x/y)

I. CONTROL BUILDING ZONE (Continuec) ,

E

. 0/9 17/0 PGCC PANEL MODULE, EL 136'0" 17/0 SD-143 C/9 PGCC PANEL MODULE, EL 136'0" 8/0 5D-144 0/8

- 50-145 PGCC PANEL MODULE, EL 136'0" 0/8 8/0 t-5D-146 PGCC PANEL MODULE, EL 136'0" 14/0 'g PGCC PANEL MODULE, EL 136'0" 0/12 :

i 50-147 0/12 18/0 PGCC PANEL MODULE, EL 136'0" 14/0 I 50-148 0/10 '

l 50-149 PGCC PANEL MODULE, EL 136'0" 15/0 :

PGCC PANEL MODULE, EL 136'0" 0/9 l 50-150 0/10 10/0 '

50-151 PGCC PANEL MODULE, EL 136'0" 8/0 l

- PGCC PANEL MODULE, EL 136'0" - 0/8 5D-158 10/0 F i

NON PANEL MODULE AREA NORTH, EL 135'0" 10/0 -

50-152 l

50-153 50-154 MON PANEL-AREA SOUTH, EL 135'0'"

GENERAL AREA, EL 136' 84/0 yr l

50-162 REMOTE SHUTOOWN PANE 1/0 EL 98'0" ,

r 50-163 REMOTE SNUTDOWN P IV 1/0

^ EL 98'0" -

FD-26 CHARC0AL FILTER 1HVC*FLT38, i EL 115'0" 1/0 &

FD-27 CHARC0AL FILTER 1HVC*FLT3A, c. .

EL 115'0" 1/0 ._

I II. REACTOR BUILDING *

~ ZONE .

6/4 MO SD-57 # CONTAINMENT AREA, EL 114'0" ANNULUS AREA, EL 186'3" N& 2Vo SD-102 17/0 .

50-104 # CONTAINMENT AREA, EL 186'3" 4/+ Vo 5D-117 # CONTAINMENT AREA, EL 162'3" 13/0 ',

50-119 # CONTAINMENT AREA, EL 141'0" 2/0 50-156 # CONTAINMENT AREA, EL 95'9" .

FD-13 #RECIRC PUMPS

DRYWELL, EL 70'0"

& 98'0" 2/0 t

(x/y):

x is number of Function A (early warning fire detection and notifi-cation only) instruments.

y is number of Function 8 (actuation of fire suppression syst' ems and early warning fire detection).

fThe fire detection instruments located within the Containment are no.t. required to be OPERA 8LE during the performance of Type A Containment Leakage Rate Tests.

3/4 3-89 APS I 8 25 l RIVER BEND - UNIT 1 i

O

TABLE 3.3.7.10-1 (Continued)

TABLE NOTATION ,

' ACTION 100 - With the number of channels OPERABLE 1ess than required by the Minimum Channels OPERABLE requirement, effluent releases may ,

continue for up to 14 days provided that prior to initiating a release:

ut

a. At least two independent samples are analyzed in accordance

'with Specification 4.11.1.1.1,and

b. At least two technically qualified aeopers of the facility staff independently verify the release rate calculations and discharge line valving;

( f I

I Otherwise, suspend release of radioactive affluents via this -

- pathway. _

ACTION 101 - With the nurter of channels OPERA 8LE less than required.by the Minfaum-Cha.inels OPERA 8LE requirements, effluent releases via

this pathway may continue for up to 30 days provided that, at

_ least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples arie collected and analyzed r for gross radioactivity (beta or gamma) at a limit of detection i

of at least 10 7 alcrocuries/al.

ACTION 102 - with the number of channels OPERABLE less than required by the -

Minimus Channels oprs*"- quirement, effluent releases via /

h thispathwaymayfontinue/3orupto30daysprovidedtheflow j

~

rate is estimateo .6 ... 5 once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. i Pump curves generated in situ any be used to estimate flow.

.j e

9 i

e e I

i i

l AIVER SEND - UNIT 1 3/4 3-96 .

4

, ,-_-._,.__.w. , -. _- - , . -- . , , . _ _ . , _ ,__ y - _., _.._,._ ,_,_m 7 __ , , . , ,

TABLE 4.3.7.10-1 (Continued)

TABLE NOTATION FINAL DRAFT (1) The CNANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway occurs if any of the following conditions axists:

1. Instrument indicates measured levels above the alarm / trip setpoint.

2. Circuit failure.

(2)

FUNCTIONAL TEST shall also demonstrate that control room alarm ~

annfiation ecurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint. I

2. Circuit failure. ,

3.

Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

1 (3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or

- using standards that have been obtained from suppliers that participate in sensurement assurance activities with NBS. These standards shall permit (

calibrating the system over its intended range of energy and measurement i range. For subsequent CHANNEL CALIBRATION, sources that have been related (

to the initial calibration shall be used. ;

(4) CHANNEL CHECK shall consist of verifying indication of flow during periods ;

of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on

.. days which continuous, periodic, or batch releases are made. i 1

O AIVERBEND-UNIT 5 3/4 3-98 6 2 8 IS85

l TABLE 4.3.7.11-1 (Continued) e TABLE NOTATIONS *

At all times.

During main condenser offgas treatment systes operation.

C..;c4 ge,eder. :" th: =' nadnnr :!r :l=t:r.

(1[)annflatiohcurs if atty of the following conditions exists:" -"'"" /y FUNC 1.

1 Instrument. indicates asasured levels above the alars setpoint.

2. Circuit failure.

3.

, . Instrument indicates a downscale failure.

4. Instroent controls not set in operata mode.

(2)

The initial CHANNEL CALIBRATION shall be performed using one or more of i the reference standards certified by the National Sureau of Standards or i

- using standards that have been obtained from suppliers that participate -

A., in measurement assurance activities with ISS.

' calibrating range. the 'systee over its intended range of energy and esasurementThe i to the initial calibration shall be used.For subsequent CHANNEL CALIBR (3)

The CMANNEL containing a nominal: CALIBRATION shall include the use of standard gas samples

~~1.

One volume percent. hydrogen, balance nitrogen, and -

2.

Four volume percent hydrogen, balance nitrogen.

l l

l e .

RIVER BEND - UNIT 1 3/4 3-105 EIIE

I

, INSTRUMENTATION

, 3/4. 3. 9 PLANT SYSTEMS ACTUATION INSTRUMENTATION = *

d"" "

~~

LIMITING CONDITION FOR OPERATION s i

~

3.3.9 The plant systems actuation instrumentation channels shown in Table 3.3.9-1 shall be OPERABLE with their trip setpoints set consistent with the ,:

values shown in the Trip Setpoint column of Table 3.3.9-2.

APPLICABILITY: As shown in Table 3.3.9-1.

ACTION:

_, ' s. With a plant system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values '

column of Table 3.3.9-2, declare the channel inoperable and take -

the ACTION required

, Table 3.3.9-1. , ;

b. With one or mor take the ACTION

' stems actuation instrument channels inoperable, d by Table 3.3.9-1. '

SURVEILLANCE REOUIREMENTS i

^.

4.3.9.1 Each plant system actuation instrumentation channel shall be t - demonstrated OPERABLE by the performance of the CHANNEL CHECK. CHANNEL FUNCTIONAL ,

TEST and CHANNEL CALIBRATION operations for the OPEPJ,TIONAL CONDITION 5 and at '

the frequencies shown in Table 4.3.9.1-1.

4.'3.~9.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of -

all channels shall be performed at least once per la months. '

e

(

! RIVER 8END - UNIT 1 3/4 3-108 APR261!ES

FliM1. DRAF REACTOR COOLANT SYSTEN -

3/4.4.2 SAFETY VALVES .

SAFETY / RELIEF VALVES

_ LIMITING CONDITION FOR OPERATION The safety valve function of at least 5 of the following olio " <es 3.4.2.1 ce

the relief other than those valve function satisfying the of at least safety 4 additional valve function requirement valves ans of a i .tb '

I OPERABLE with the specified lift settings:

l Function Setooint* (esic) :

Number of Valves i.

Safety U65 1 1% '

7 Safety n 80 1 1% - h l 5 1190 1 1% * '

4 Safety

. Relief - n 03 1 15 psig 1

Relief n 13 1 15 psig ,

8 i Relief n23 2 15 psig ;

. - 7 l

The scoustic monitor for each OPERABLE valve shall be OPERABLE. . . . -

l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

r' ACTION: L J

a. With the safety and/or relief valve function of one or more of the above required safety/ relief valves inoperable, be in at least HOT SHUTDOWN p.

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

b.

With one or more safety / relief valves stuck open, provided that suppres-sion pool average water tasperature is less than 105'F, close the stuck

' open safety / relief valve (s); if suppression pool average water temperature.-

is 105'F or greater; place the reactor mode switch in the Shutdown position '

c. With one or more safety / relief valve acoustic monitors inoperable, restore j

the inoperable monitor (s) to OPERA 8LE status within ,

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

i '

i 5

.I

?

"Ine Itft setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

J p5 6@ ,

l 3/4 4-5

~

RIVER BEND - UNIT 1 me

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

l REACTOR COOLANT SYSTEM

( 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION I

3.4.5 The specific activity of the primary coolant shall be limited to: I

! a. Less than or equal to 0.2 microcuries per gram DOSE EQUIVALENT I-131, and ~ ~ ~ ~

b. Less than or equal to 2004 microcuries per gram.

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

l

~

ACTION:

a. In OPERATIONAL CONDITIONS 1, 2 or 3 with the specific activity of the primary coolant;

)

t -

1. Greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram, operation may

- continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provided that the cumulative opera-1 ting. time under these circumstances does not exceed 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months in any consecutive 12-month period. With.the total c a ulative

~ operating time at a primary coolant specific activity greater i, than 0.2 microcuries per gram DOSE EQUIVALENT I-131 exceeding i

500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months in any consecutive six month period, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days indicating the neber of hours of operation above this limit. The provisions of Specification 3.0.4 are not appli-2 .. cable. -

2. Greater tha'n 0.2 microcuries per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or for more than 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months cumulative operating time in a consecutive 12-month period, or greater than 4.0 microcuries per gram, be in at least NOT SHUTDOW with the main steam line isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3. - Greater than 2004 mice : . ^ -

gram, be in at least NOT j SHUTDOW with the mai ne solation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I

b. In OPERATIONAL CONDITIONS 1, 2, 3 or 4, with the specific activity of the primary coolant greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 or greater than 200/E microcuries per gram, perform l the sampling and analysis requirements of Ites 4a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to I within its limit. A REPORTABLE EVENT shall be prepared and sub-mitted to the Commission pursuant to Specification 6.5.1. This report'shall contain the results of the specific activity analyses 4

L RIVER BEND - UNIT 1 3/4 4-16 ps t 6 95 l

..m., . . . - , - - _ - . . . _ . . _ . . - , . _ _ _ _ _ . , _ , , , . . _ _ , - ____-,-,,_.___.._-.__.,...m-

i

, l l

REACTOR COOLANT SYSTEM

\ 3/4.4.7 MAIN STEAM LINE !$0LATION VALVES LIMITING CONDITION FOR OPERATION 3.4.7 Two sain steam line isolation valv OPERABLE with c7osing times greaterMSIV than o. 3pand

(.-^..steam line shall be k to 5 seconds.

.sern/s s,s than or equal g APPLICABILITY: OPERATIONAL CONDITIONS 1. 2 and 3.

ACTION:

7 a. With one or more MSIVs inoperable:

1.

Maintain at least line that is open one MSIV and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,OPERABLE either: in each affected m a) Restore the inoperable valve (s) to OPERABLE status, or b)

Isolate the affected main steam line by use of a deacti-f .

wated MSIV in the closed position.

2.

s 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 SNUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . .

b.

The provisions of Specification 3.0.4 are not applicable.

$URVEILLANCE REQUIREMENTS _

4.4.7 verifying4.0.5.

Specification full closure between 3 and 5 seconds when test for entry into OPERATIONAL CONDITIONS 2 or 3 provided th performed and prior to entrywithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months CONDITION into OPERATIONAL after reaching

1. a reactor steam pressure of 6 RIVER BEND - UNIT 1 3/4 4-25 M I' r

. PLAN 1 SYSTEMS IARD FIRE NYORANTS AND MYDRANT N05E MOUSES LIMITING COND] TION FOR OPERATION l t

3.7.6.5 z Table 3.7.6.5-1 shall be OPERABLE.The yard fire hydrants a e houses shown in APPLICAB]LITY:

hycrants is required to be OPERABLEWhenever equipment in the ACTION:

e ya c fire a.

With one or more of the yard fire hydrants or l hose houses shown in Table 3.7.6.5-1 inoperablassociated hy

,, , sufficient additional lengths of 21/2 inch diameter in an adjacent OPIAASLE hydrant hose house t he, wi a ose located the improtected area (s) if the inoperable o provide fi service to b.

t.therwise provide the additional .

hose w

, The provisions . < of Specifications 3.0.3 and 3 0 4 ..

j are not applicable.

$URVEiLbNCE REQUIREMENTS ..

C- .

Z 4.7.6.5 Each of the shown in Table 3.7.6. yard fire hydrants and associated hydrant h

a. 51 shall be demonstrated OPERABLE: ,

At least once per 31 days by visual inspection ,

b.

house to assure all esquired equipment is at thof the hydran -

e hose house.

At least once per 6 months, by visually inspecti .

fire hydrant and verifying that the hdrant that the h drant is not damaged.

ng eachba rre.

yard is dry and c.

,At least once per 12 months by:

1. (

Conducting a hose hydros or at least 50 psig abov st at a pressure of 150 psig pressure, whichever is gr..... . he maximum fire main operating

2. ,

3. Replacement of all degraded gaskets s. in coupling Performing a flow check of each hydrant .

\

i l

RIVER SEND - UNIT 1 i 3/4 7-27 APR281lP

~ --- ,-.., .-.,_, n.- ,- ..,,,-,-,,,,,.-,___,,,,,__-,,.,_,,,,_,,,_,n.. - -

fp"'91.M q*p ELECTRICAL power SYSTEMS A l

SURVEILLANCE REQUIREMINTS (Centinued) l I

6. Simulating a loss of offsite power in conjunction with an ECCS I actuation test signal, and:

a) For divisions I and II: 1

1) Verifying deenergization of the emeYfe'ni:y busses and --

load shedding from the emergency busses. -

2) Verifying the diesel generator starts on the auto-start signal, energizes the emergency busses with permanently { .

connected loads within 10 seconds, energizes the auto- ,

, connected ene4down loads through the 4eed sequencing logic and operates.for greater than or equal to .

5 minutes while its generator is loaded with the emergency loads. After energization, the steady . state voltage and frequency of the emergency busses sh4A1 be F saintained at 4160 1 420 volts and 60 2 3 Hz during g this test. .,

For division III:

b) , **h

, s,3,

1) Verifying de energization of the emergency bus.

i

2) Verifying the diesel generator starts on the auto-

\ p e rm...,

ly * *-a~ *a '

A start signal, energizes the emergency bus with its J.. J w.*L..,

10 s a w .,#c, ; S:d: c,r.; the auto-connected emeageney loads n'tM-e.se-pe s 10 rect-f: and operates for greater than or equal to 5 minutes while its generator is loaded with the

.- emergency loads. After energization, the steady state ~

' voltage and frequency of the emergency bus shall be ,

maintained at 4160 2 420 volts and 60 2 3 Hz during

this test.

7. Verifying that all automatic diesel generator trips are automatic-ally bypassed upon E ctuation signal except: .

a) For divisio engine overspeed and generator differential en .

b) For divisi ngine overspeed and generator differential current.

8. Verifying the diesel generator operates for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The diesel generators shall be loaded to 3130 kw for diesel generator 1A and 18 and 2600 kw for diesel generator 1C.. The generator voltage and frequency shall be 4160 s 420_ volts and 60 2 3 Hz within 10 seconds after the start signal; the steady 1

(

RIVER BEND - UNIT 1 3/4 8-6 gpp261# I l

TABLE 4.8.1.2.2-1 F $ */du \ ".4.j*h' .

s' e

_ DIESEL GENERATOR TEST SCHEDULE i

Numb f Failures in Las Valid Tests

_ Test Frecuency 1

L 11 ~

Atleastonceper31 days 2

At least once per 14 days

.. . 3 4

At least once per 7 days "

>4 1

- At least once per 3 days I

I.

t

. Criteria for cetermining number of failures and number of valid Regulatory Guide 1.108, Revision 1 tests shall be in accordance w j \. 100 tests are determined on a per n,uclear unit basis. August 1977, where the la For the purposes of this test schedule, only valid tests conducted after "last 100 valid tests."the DL issuance date shall be included in the computat

,, made at the 31 day test frequency. Entry into this test schedule shall be ~

J l

APR 2 61995 RIVER BEND - UNIT 1 3/4 3.g

. i TABLE NOTATION N =

, continued s i b

is the standard deviation of the background counting rate or of eth counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 x 10' is the number of disintegrations per minute per microcurie, f Y is the fractional radiochemical yield, when applicable. :

I A is the radioactive decay constant for the particular radionuclide "and ,

At for plant collection andeffluents is the elapsed time between the midpoint of ' sample time of counting. L Typica'1 values of E. V, Y, and at should be used in the calculation 6

i 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.

f b - A batch release is the discharge of liquid wastes of a discrete volume . Prior mixed to assure representative sampling.to sampling for analy c - The sivelyprincipal gammaradionuclides:

are the following esitters for which the LLD specification applies exclu-i j - Mo-99, Cs-134, Cs-137 Ce-141, and Ce-144.Mn-54, Fe-59, co-58, Co-60, 2n-65 only these nuclides are to be considered. This list does not mean that Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be . . .

pursuant to specification 6.9.1.99. analyzed and reported in the Seal d - A composite s 6 on tional to th of sampling w,quan)iM}in liquid waste which theand discharged quantity in which theof method Ifquid sampled ifquids released.,L ,. results in a specimen that is representative of the 1

i i

APR 2 61985 RIVER SEND - UNIT 1 3/4 11-3

RADI0 ACTIVE EFFLUENTS

. ~

( _ LIQUID RADWASTE TREATMENT SYSTEM, LIMITING CONDITION FOR OPERATION -

i 3.11.1.3 3

radioactive materials in liquid astese when prior e the to their projected organ in a 31 doses due toFigure day period. the liquid 5.1.3-1) effluent, would exceed to UNRESTRICTE the see 0.06 arem to y or 0.2 arem to any APPLICABILITY: At all times.

_ ACTION:

a. \

, With radioactive in excess of liquid waste being discharged without t

, grithin 30 d bove limits, prepare and s reatment and includes th svant to Specification 6. to the Commission ing information- pecial Report that

)f

_ 1. I treatment, identification of anyorinoperable

~

out subsystems, and the reason for the inoperability, .

if 2.

Action (s) taken to restore the inoperable equipment t status, and o OPERABLE 1 3. *

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

a recu The provisions of Specifications 3.0.3 .. and 3applicable. 0 4 are not

[URVEILLANCEREQUIREMENTS

' ' 4.11.1.3. Doses due to liquid releases to UNRESTRICTED AREA 3

at least once per 31 days in accordanceS shall the ODCM. witharameters bethe methodolog projected in 1

k j RIVER BEND - UNIT 1 3/4 11-5 APR 2 81985

j. -

7 e"-ISOLATION ACTUATION INSTRUMENTATION FINAL DRAFT 3 /4. 3. 2 _

4 LIMITING CONDITION FOR OPERATION l 3.3.2 The isolation actuation instrumentation ith the values shown TEM RESPONSE channels show shall be OPERABLE with their trip setpoints set consisten l

TIME as shown in Table 3.3.2-3.

I APPLICABILITY:

As shown in Table 3.3.2-1. i f ACTION:

a.

With an isolation actuation instrumentation channel trip setpoint I.I less of Table conservative 3.3.2-2, declare than the the channel value shownuntil inoperable in the Allowable the channel is restored to OPERABLE status with its trip setpoint adjusted

. consistent with the Trip 5etpoint value.

LE channels less than required by the Minimum M

b. With the number to System requirement for one trip system, I r

~

CPERABLE Channel hanne)(s) and/or that trip system in the tripped place the inopera The provisions of Specification 3.0.4 ;

j condition

within one hour.

are not applicable.

c.

With the number of OPERA 8LE channels less than required by the M _

OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one

' hour and take the ACTION required by Table 3.3.2-1.

l-I -

~ i-i . .

.- c i

! * .I

\ i d condition where this ,

An inoperable channel need not be placed in the tr ppeIn these cases, theby inoperable channel J

would cause the Trip Function to occur.shall be restored to CPERAS !,

4 Table 3.3.2-1 for that Trip Function shall be takent d tac" on if this would cause :

j **The trip system need not be placed in the tripNWhen a trip system o nithb slaced in th the Trip Function to occur.

j condition without causing the Trip Function to occur, have piece d'

the the' trip sy the most inoperable channels in the tripped condition; 4

. . . f condition.

l la i

k _

apn i ss -

3/4 3-10 RIVER BEND - UNIT 1 .

, - - , - - - . ,-. - -,n- , ,.--+,-,,s-, , - - -, , , . , - ~ - . , - , , - , m-,

i e' q.

" Mg g

> TABLE 3.3.2-1 (Continued) a das Wdafd[p

, ISOLATION ACTUATION INSTRUMENTATION ACT]ON i 1

ACTION 20 -

Be in the within at least next 24 NOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHU hourt.

ACTION 21 -

Close the affected system isolation valve (s) within one hour or:

a. In OPERATIONAL CONDITION 1, 2, or 3, 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 corn m i m wy within the following 24 hou Swjs/ f

b. InOperationalConditionh,suspendCOREALTERATIUM5, s handling of irradiated fuel in the primary containment and operations with a potential for draining the reactor vessel.

ACTION 22 -

Restore the manual initiation function to CPERABLE status within 48 andhours in COLD orSHUTDOWN be in at least NOT within the SHUTDOWN following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

~ '

ACTION 23 -

Se in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least NOT SHUTDOWN within 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 .

ACTION 24 -

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

IM ACTION 25 -

Establish SECONDARY CONTAINHENT INTEGRITY - OPERATING wit ore hour.

ACTION 26 -

standby gas treatment Aaed'systenf Fwel $..qyav$ operating 9 within, leis)en9 se Restore the manual initiation function to OPERABLE status l within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or 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 .

ACTION 27 -

Close the af'fected system isolation valves within one hour and -

~

declare the hffected system inoperable. '

T0 -

f:

21 ! ^ :a"" Ca".! M u' !"'!:*.!'" 5 th; :t: e , ;::

t ::* :nt :;;r:t h; c'at' :n: t. Initiate and maintain annulus sixing sytten with the reactor building annulus exhaust to at least one operating standby gas treatment train within.

I hour. .

ACTION 46 -

E Lock the affected system isolation valves closed within one hour and declare the affected system inoperable.

\

RIVER BEND - UNIT 1 3/4 3-17 1

E

TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME RESPONSE TIME (Secones)n .;

TRIP FUNCTION

1. PRIMARY CONTAINMENT ISOLATION

< 10I ")

a. Reactor Vessel Water Level - Low Low, Level 2 < 10

b. Drywell Pressure - High g < 10 Containment Purge Isolation Radiation - High 'E

c. NA

d. Manual Initiation i

2. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water Level - Low Low Low, < 1.0 */< 10(*)a=

l Level 1 I 1.0 8 8 10(8)== I

b. Main Steam Line Radiation - High(b) -

l I 1.0 */7 10((a) ,

' c. Main Steam Line Pressure - Low 7 0.5

B 10 "')==

d. Main Steam Line Flow - Ni h

- ~

e. Condenser Vacuum - Low y.yy4,] RA NA Main Steam Line Tunne~1 n.,

f. NA

g. Main Steam Line Tunnel A Temperature - High NA

h. _ Manual Initiation

~

3. SECONDARY CONTAINMENT ISOLATION

< 10f ")

a. Reactor Vessel Nater Level - Low Low, Level 2 I 10 .

h b.

a c. Drywell Pressure - HighFuel Building Ventilation Exhaust 1 10 Radiation - High(b) ,

d. Reactor Building Annulus ) I Ventilation &ames Exhaust Radiation - High $ ID ") t' NA

. e. Manual Initiation s u

i 4. .-

REACTOR WATER CLEANUP SYSTEM ISOLATION l

' < 10(*)## - $

a. A Flow - High WA

b. A Flow Timer NA i c. Equipment Area Temperature - High NA

d. Equipment Area A Temperature - High g)

< 10 t

e. Reactor Vessel Water Level - Low Low, Level 2 I

f. Main Steam Line Tunnel Ambient NA Temperature - High NA U" ' '- Tuntiel A Temperature - High

g. Main NA

h. SLC6niti)[atio9 NA f

1. Manual a n . 5.eu on

5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION RCIC Steam Line Flow - High < 10(*)'##

a. EA

b. RCIC Steam Line Flow-High Timer < 10(*) '

c. ACIC Steam Supply Pressure - Low RA

d. RCIC Turbine Exhaust Diaphragm Pressure - High NA

e. RCIC Equipment Room Ambient Temperature - High NA

f. RCIC Equipment Room A Temperature - High NA

g. Main Steam Line Tunnel Ambient Temperature - High NA

h. Main Steam Line Tunnel A Temperature - High RIVER SEN0 - UNIT 1 3/4 3-24 APft 2 619E5 i

I

' ~ -- . . .- .. _

r' -

( .

TA8tt 4.3.2.1 _1 (Continued) .

$ 150 TAT 10N ACTUATIDN INSTRUMENTATION SURVElltANCE REQUIREMENTS W "

CHANNEL i OPERATIONAL e.

FUNCTIONAL CHANNEL CONDITIONS IN WHICH CHANNEL T TEST CALIBRATION SURVElllANCE REQUIRED CHECK TRIP FUNCTION -

g 3. SECONDARY CONTAINNENT 150 tail 0N R ICI 1, 2, 3 l U a. Reactor Vessel Water M '

tevel - Low tow, level 2 O p 11 Pn ssun tw 5

5 M M 1,2,3

b. b ,,, ,,-- ,, -- R 1, * ' --

'M e

c. Fuel Building Asee venuiauonr >

d. ' Reactor Building Annulus^ '

Ventilation Ex R 1,2,3 e.

Radiation - MI Manual Initiation k[ 5 MA Mg ,)

M MA 1,2,3 ,

j w 4. REACTOR WATER CLEANUP SYSTEM ISOLATION R 1, 2 3 N 5 M

'

a. A Flow - High, 1, 2, 3 w NA M Q a b. A Flow Timer w

c. Equipment Area Temperature - M R 1,2,3 5

High ,

d. Equipment Area R 1, 2, 3 5 M a Temperature - High

e. Reactor Vessel Water 5 M RI "I 1,2,3 level - Low tow, tevel 2 1,2,3 smyg

f. Main Steam Line Tunnel Ambient M R 5

Temperature - High .

1, 2, 3 - -

g. Main Steam line Tunnel A Temperature - High 5 M R 3, 7. 3 M

h. StCS Initiation NA M(b) ,,

F N(a)

l. Manual Initiation MA yg 3, g, 3 D

- :xt g .

m

=al p 7 'ir

,- j- 3-I

( .

. Yg' TABLE 3.3.6-2 *-

5 W CONTROL ROD BLOCK INSTRUMENTATION SETP0fMTS

a. TRIP FUNCTION TRIP SETPOINT

/.M oE ' Alt 0WA8tE VALUE

1. ROD PATTERN CONTROL SYSTEM e

e a. Low Power Setpoint 27.5 i TED TifERMAL POWER 27.5

M RATED THERMAL 4

}

w 2.

b. High Power Setpoint 62.5 ATED THERMAL h)WER POWE 62.5 1 .

ATED THERMAL POWER

/ K APRM

a. Flow Blased Neutron Flux .

- Upscale .

< 0.66 W + 42%* '

b. Inoperative RA

< 0.66 W e 45%*

c. Downscale HA

d. g% of RATED THERMAL POWElI, Neutron Flux - Upscale - 13% of RATED THERMAL POWER Startup i 12% of RATED THERMAL POWER 114% of RATED THERMAL POWER

3. SOURCE RANGE MONITORS w a. Detector not full In MA 1 b. Upscale < 1 x 105 cps NA w c. Inoperative NA

< 1.6 x 105 cp, g d. Downscale 1 0.7 cps RA 1 0.5 cps **

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in NA ,

N4

b. Upscale 1 108/125 division of full < 110/125 division of full scale scale

c. Inoperative NA

d. NA Downscale > 5/125 division of full scale -> 3/125 division of full scale M

5. SCRAM DISCHARGE VotUME

a. Water Level-High -

< 18 Inches < 22 inches E3

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW

a. Upscale

ga i 108% of rated flow $ 111% of rated flow .

i O

to

= *The The tripAverage setting of thisPower Range function must be maintained Monitor rod block in accordance withfunction Specificationis3.2.'varied as a funct

- Provided signal to noise' ratio is > 2, otherwise setpoint of 3 cps and allowable 1.8 cps.

5 * "T1

~ _ TABLE 4.3.7.2-1 A SEISMICMONITORINGINSTRUMENTATIONSURVEILLANCEREQUI

CHANNEL -

CHANNEL FUNCTIONAL ~

INSTRUMENTS AND SENSOR LOCATIONS CHECK CHANNEL TEST

1. CALIBRATION Triaxial Time-History Accelerographs

a. I Reactor 81dg 70'0"

b. M Reactor Bldg hield Wall M SA R EL 232'0" SA R Y

c. ;

d.

Reactor 81dg. Drywell EL 151'0" M -

7 Free Field-Grade Level SA R M SA R

2. Triaxial Peak Accelerographs

- a. Reactor 81dg. SLCS Stora

b. Reactor Bldg. - RHR Inj.ge TankNA NA NA

~

c. Piping NA R

~

Aux. 81dg. Service Water Piping NA R '

NA R 3.

Triaxial Seismic Switches

s. Reactor Bldg. Mat El 70'0" M(a) SA 4 R Triaxial Resportse-Spectrum Recorders a.

(.*- 4 b.

Reactor 81dg. Mat EL 70'0' Reactor 81dg. Ficor EL 141'0" M SA R

c. Auxiliary 81dg. Mat EL 70' NA SA R

d. NA Auxiliary Bldg. Floor EL 41 0" I NA NA R NA R xcept seismic trigger.

AFi,1i25 L '

RIVER SEND - UNIT 1 3/4 3-72 !

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 tAND USE CENSUS i

I LIMITING CONDITION FOR OPERATIO C - _' O 3 82.L 4

0. 2. 2 '

A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of garden" of greater the nearest than milk50 m2animal, thet nearest residence and the nearest (500 ft ) producing broad leaf vegetation.

APPLICABItITY: At all times. '

ACTION:

a.

With a land use census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, identify the new location (s)

- in the to next Seminannual Specification 6.9.1.8.Radioactive Effluent Release Report, pursuant '

b.

- With land use census identifying a location (s) that yields a calcu ated dose or dose commitment (via the same exposure pathway 20 percent-greater than at a location from which samples are curre)ntly being

.f- obtcined in accordance with Specification 3.12.1, add the new location (s)

! to the radiolo2ical envirbnnental monitoring program within 30 days.

The sampling location (s), excluding the control station location, 1 -

having same the lowest o ure pau calculated dose or dose commitment (s), via the

, may be deleted from this monitoring program aftithetober 3 the year in which this land use census was

.' conduu... T.....nt to Specification 6.9.1.8, identify the new

.. location (s) in the next Seminannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table -

for the CDCM reflecting the new location (s).

c.

The provisions of Specificatior.s 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.2 i

least once per 12 months using n that i' formation that willThe land use ;

results, such as by a door-to-door survey, aerial survey, provide the best or by consulting i local agriculture authorities. The results of the land use census shall be i included in the Specification Annual Radiological Environmental Operating Report pursuant to 6.9.1.7.

" Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the site boundary in each of two different direction sectors f

with the highest predicted D/Qs in lieu cf the garden census. Specifications t for broad leaf vegetation sampling in Table 3.12.1-1, 4c shall be followed.

including analysis of control samples.

RIVER BEND - UNIT 1 APR 2 61985 3/4 12-13 a

~

INSTRUMENTATION 8ASES N _

3/4.3.4 REC 1RCULATION PUMP RID AC'.* ~ ~ MN INSTRUMEN12 TION

The anticipated transient without scram (ATWS) recirculation pump trio system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The i response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NE00-10349, dated March 1971 and NED0-24222, dated December 197S, and Section 15.8 of the FSAR.

The end-of-cycle recirculation pump trip (EOC-RPT) system is a part of reactor trip.Protection System and is an essential safety supplement to the the Reactor

- - The purpose of the E0C-RPT is to recover the loss of thermal

~ margin which occurs at the end-of-cycle. The physical phenomenon involvet! is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control .

rods add negative scram reactivity. Each EOC-RPT system trips both recircu-the core during two of the most limiting pressurization The two events.la events for which the EOC-RPT protective feature will function are closure of ,

- the turbine stop valves and fast closure of the turbine control valves. -

A fast closure sensor from each of two turbine control valves provides

, 4 input to the EOC-RPT system; a fast closure sensor from each of the other two

\ turbine control valves provides input to the second EOC-RPT system.

EOC-RPT system; a position switch from each of *.he other t provides input to the other E0C-RPT synes. Foi each EOC-RPT system, the l'

sensor relay contacts are arranged to form a 2 out-of-2 logic for the fast closure vaTves. of turbine control valves and a 2-out-of-2 logic for the turbine stop _

trip both recirculation pumps.The operation of either logic will actuate the E - -

Each EOC-RPT dministratively system may be manually bypassed by use of a keyswitch controlled. The manual bypasses and the autoriat ating m.ss at less than 40% of RATED THERMAL POWER are annunciated ir, th rol between initiation of valve motion and complete arc, i.e. ,140 ms.

supThe EOC-RP the electric Included in this stM are: the espo 'me of the sensor, the time allotted for breaker arc suppression and t l

system logic. e time of the -

) within its specified Allowable Value is acceptable on the basis difference between each Trip Setpcint and the Allowable Value is ecual to or

! less than the drif t allowance assumec for each trip in the safety a'nalyses. ,

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

l l -

ar+

4eOTE: SCALF IN INCHES g gpgggggt .,, g AS 3' SEL ZERO CATE EL NOMENCLATURE k

ME10HT ASLVE N O. VESSEL ZERO (IN.) READING

.) 800 - C 572.s23 (8)

(7) sse.42 52 3s.

/\

, (4) 851.42 30.3 7509 - (3) 529.52 . g .g 722.75 YESSEL - (2 475.12 45.5 e FLANGE =

375.12 i

145.5 1 700- -

l 850 - - MAIN

- 536.5- STEAM ,

LINE INSTRUMENT

$00 - - ZERO g b"--Pr" 572.82(8) ) "2 52(g) 0-

[,3 g,, -

52 TRIP RP 8)

COTTOM OF STEAM i

550 - rQ@j HPCS. RCIC ,Mi L M YER SKIRT -529.52(3) ALAng . g,gg33

- _ _RF. ACTOR SCR AL FEED 483'5500:

WATER --475.12[2) CORE 45.5(2) CCfNFIRMATORY 65 ADS TRIP

. . .. SPRAY # INITIATE RCIC, HPCS;

.. 450 - - TRIP RECIRC. PUMPS I

~

4;s.56 ~~ gn0 - _ $V$f/& JE b4bd4Mb z<Mu g

C ~!g!:6 (' -160 .145.5(1) 350 -- 358.56 INITIATE RHR AND LPCS,

. O START DIESEL, INITIATE

, ,r[ #, j id ADS AND CLOSE MSIV's y l 'qff's,".

pf M sco- -

ACTIVE q

FUEL 250- -

=

208.56 200- m 206.56

. d RECIR C OUTLET 1ss.5 -171.5 ',"'yLEo

' 7 NOZ2LE 150 - -

100 - -

I

50 - -

Bases N gure 8 3/4 3-1 REACTOR VESSEL WATER LEVEL - .-

)

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

RADIOLOGICAL ENVIRONMENTAL MONITORING ilNAIDyp7 .

a BASES 3/4/12.2 LAND USE CENSUS l Jo a his specification is provided to ensure that changes in the use of areas ra beyond the SITE SOUNDARY are identified and that modifications to the o ogical environmental results of this census. monitoring program are made if required by the from aerial survey or from consulting with local agricultural authoritie shall be used.

Appendix I to 10 This CFR census Part 50.satisfies the requirements of Section IV.B.'3 of than 50 m2 Restricting the census to gardens of greater 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.

this minimum garden size, the following assumptions were made: To determine

1) 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/m2, 3/4.12/3 INTERLAB6RATORY COMPARISON PROGRAM g Program is provided to ensure that independent checks on th accuracy of the measurements of radioactive material in environmental sample

( s 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.

e APR 2 s 1995 RIVER BEND - UNIT 1 8 3/4 12-2

m Enclosure 2 Errors of Comission em W

j I _ _ , _ , _ ,- - - - - _ -- - - - .

FIN 21 ELECTRICAL 90WER SYSTEMS P 7~ 7 SURVEILLANCE REQUIREMENTS (Continued) .

1 ,

3. 5 Verifying the diesel generator capability to reject a load of 3130 kwCfor generator diesel without generators IA and 18 and 2600 kw for tripping.

The generator voltage shall not exceed 4784 volts for diesel generators IA and 18 or 5824 vo!

for diesel generator 1C during and following th load on.

rejecti .

4. , ,

j Simulating a loss of offsite power by itself, and:

a) For divisions I and II:

1)

Verifying deenergization of the emergency busses and load shedding from the emergency busses.

2)

_ Verifying the diesel generator starts on the auto-start connected loads within 10 seconds, e connected and operates for greater than or equal to 5 m i , while its generator is loaded with the : L if: ~

After energization, the steady state voltage andloads.

frequency at 41601420of the emergency busses shall be maintained volts and 60 2 3 Hz during this test.

( b) For division III:

1)

Verifying de energization of the emergency bus.

! 2)

.. p.n.,

signal, Verifying energizes the the diesel m generator starts on connected loads withi with the permanently y

,i greater than or equal (w0+seconosyd : operates for 9 ,

' is loaded with the Nt"n loads.: N _.. while its generator After energization bus shalleb' maintained atthe steady state voltage an 4160 1 420 volts and

! 60 2 M Hz during this test. '

. I

5. l

{

Verifying that on an ECC5 actuation test signal, without loss l of offsite power, the diesel generator starts on the auto start 5signal minutes.and operates on standby for greater than or equal to 2 420 volts and 6013 Hz within 10 seconds -start after' th i

be maintained within these limits during this tes y .

l' ,rg# #) , a. , , w a a. ...~. + a 1...:<

a + aa u v . - 1.s. , _

\

t(

j AIVER 8END - UNIT 1 s 3/4 8-5 k i

APR 2 6 25

_w --ww-""*'" _e-.- *'"NT ' ,_ ,v. -,.r - ~ " " * ' ' ~

! 7'T 3 .M

1 ELECTRICAL POWER SYSTEMS K I

i

( $URVEILLANCE REQUIREMENTS (Continued)

~

1 6.

Simulating actuation a loss test of offsite signal, and: power in conjunction with an ECCS a) For divisions I and II:

1) ,

Verifying deenergization of the emerge ~ncy busses and load shedding from the emergency busses.

, 2)

Verifyirig the diesel generator starts on the auto-start signal, energizes the emergency busses with permanently P' - connected loads within 10 seconds, energizes the auto-connected et ^.in loads through the W sequehcing

{ logic and operates for greater than or equal to i 5 minutes while its generator is loaded with the emergency loads.

i

. After energization, the steady. state voltage and frequency of the emergency busses shAl be saintained at 4160 1 420 volts and 60 2 3 Hz duririg this test.

b) For division III:

"f,,fa

, i, , ,,

1)

t Verifying de energization of the emergency bus.

\ pu ..,4/y . .-~ + . 4 2) Verifying the diesel generator starts on the auto-

le. Jr w ,fl..-, 10 s m .,4r g M: d: start signal, energizes the emergency bus with its M cc ;;,4 rd: the auto connected : : ;;;n;, loads Wh';

! eu e-1.e s and operates for greater than or equal to

[ j 5 minutes while its generator is loaded with the emergency loads.

~

After energization the steady state I h/WW voltage and frequency of the emergenc,y bus shall be L ya:ew/

maintained at 4160 2 420 volts and 6013 Hz during '

this test.

7.

, ally bypassed uponECaVerifying that all automatic diesel generator ctuation signal except:

i a) For division i ngine overspeed and generator differential even .

b) For divisi n i current. engine overspeed and generator differential 1 8.

Verifying the diesel generator operates for at least 24 hou The diesel generators shall be loaded to 3130 kw for diesel _rs.

generator IA and 18 and 2600 kw for diesel generator IC. The i -

generator voltage and frequency shall be 4160 1 420. volts and 1

.~ 60 1 3 Hz within 10 seconds after the start signal; the steady ,

i RIVER BEND - UNIT 1 3/4 8-6 APR 2 61980

! l 1

1

. l i

FROM ATTACHMENT B TO GSU LETTER OF MAY 6, 1985. !

TECENICAL CRANGE REQUESTS ,

DESCRIPTION OF CHANGE / JUSTIFICATION:

28)TS 3.7.6.2 - Deleted Railroad Bay, a No sprinkler systems are identified for the railroad bay as there is no safety related equipment located in this area. _

---> 29) TS 4.7.6.3.a - Delete, g 4 p,f.g, There are no valves in the flow path of any PGCC subsystem.

Jf df' 7 30) TS 3/4.7.6q 4 Table 3.7.6 1 - Added footnote *.

4 Reflects River Bend design. - .

31) TS Table 3.7.1 Add items and revise temperatures. .

Additional item have been identified for inclusion and corrections to temperatures from review of EnvironmentaT Design Criteria.

Added 3.7.10-2, revised the Technical

32) TS 3/4.7.10 - Table l

Specification accordingly and also revised Table 3.7.10-1.

These changes make the Technical Specification consistent with FSAR Section 2.5.

33) TS 3/4.7.11 - Add new Specification.

This Specification is provided to address SER requirement in 9.1.3 page 9-6.

Action c, 4.8.1.1.2.f.4.b.2, 34).TS 3/4.8.1, 3.8.1.1 4.8.1.1.2.f.6.b.2, 3.8.1.2 Action b, 3.8.2.1 Action b, 3.8.2.2 Action b, and 3.8.3.1 Action b.2, - Addition of C SSW pump.

4 Revisions reflect the. powering of standby service water pump 1 ISWP*P2C and it's auxiliaries from the HPCS diesel (Div III).

35) TS 3.8.3.1.b.1 and 3.8.3.2.b.2 - Added panel 1ENB*PNLO4A. f I

i SER open item 13, Added in conjunction of outstanding Safe / Alternate Shutdown Design Modification. . ..

~

l Page 6 of 7

~ ' - - - ^ " * - - - - . _y .. _ __ _ _

.....,_i )

_E;"

8 9

<r e.

l'

, ali o

I $.f = .

[ ll!

@! s

5![

8 aDi E n!

- E .6 'I , ,

N}P si '

" Ill IE I,L

\f .

\

i g . .

ik~~i

\w f.

h .

- D i> e g _, .

in

- ian -

,g

! 4

] -

l 1.

l L 8'

\ i

\ .

l il s-~ o 4g

, g - -

game

==. e D D-= gi '

I -. iC J- .

- ' e t

' \ _

ICMp * -

l I a l \,Ba S)j n

l-

~.

Edward J. Butcher May 17, 1985 i

L

! same submittal. GSU properly proposed an identical change with the clause to be inserted after 10 seconds. As proposed i on 3/4tor.

opera 8-5. the revision makes no sense and would confuse the i i

I (2) In Attachment B of the GSU submittal of May 6,1985. Item 29

' requests a deletion of a surveillance requirement because -

"There are no valves in the flow path of any PGCC subsystem."

In past discussions, GSU has resisted this requirement on the i

basis that the valves did not have a position indicator although. in fact, the valves do have a trip indicator. A i copy of FSAR Figure 9.5-13 is enclosed which shows numerous solenoid operated valves as well as a couple of check valves in the flow path. Therefore, the GSU statement of "no valves j .. . in the flow path" appears to be a false representation.

4 b (3) Also, in Attachment B. Item 30 refers to adding a footnote to TS 3/4.7.6.4. Table 3.7.6.4-1. This is in error as the proposed j

footnote was identified with TS 3/4.7.6.5. Table 3.7.6.5-1.

In addition to the deficiencies noted above. I would like to comment briefly on other uncertainties associated with the River Bend Tech Spec review.

GSU has submitted a listing of 55 areas in the FSAR that need revision to support Tech Spec sections. Amendment 19 to the FSAR was delivered on May 14, 1985 and only 12 of these areas were addressed. Therefore, in the other 43 areas, the NRR Technical Reviewer has not seen the necessary documentation to support the current Tech Spec section or a proposed revision to a section. There is also the potential for additional FSAR i revisions resulting from the reviewer's evaluation. This lack of timely information will impact the accelerated schedule for issuance of the Tech l Specs with the River Bend license in June.1985. _

There seem to be some values in the FSAR and Tech Specs that are constantly being changed. For example, the DBA activity release to the environment following a LOCA (used for containment Tech Spec review) were revised in i ~ Amendment 18 to the FSAR dated April 1985 and revised again (increased)

{ in Amendment 19 on May 13. 1985. In the Tech Specs. GSU has pro a

the water level for the Ultimate Heat Sink be 112'4" (2nd Draft) posed that 108"6"

[ (Final Draft) and 111'10" (current revision). Changes of this frequency l would indicate that the utilities review process has not settled down.

All of the above matters should be given due considerations when discussing i i

comitments and completion schedules for the River Bend Tech Specs.

Original signed by M. Dean Houston. Reactor Engineer Technical Specification Review Group cc: Division of Licensing - .

D. Crutchfield TSR :DL Distribution .

' R. Benedict DHouston:jc Docket File TSRG File

$. Stern 5/rJ/85 i

l l l

l . - - - . - . - . . - . - . - - - , - - - - - _ . . - - - . - - - , - - - .

v TABLE 3.3.3-2 (Continued)

=;!

E ENERGENCY CORE COOLING SYSTEM ACTUATION INSTRUNENTATIO em g TRIP FUNCTION g ALLOWABLE '

TRIP SETPOINT 7

c C. DIVISION /DIRIP SYSTEM 1.

VALUE

)(\

HPCS SYSTEM

~

-4 a.

w Reactor Level 2) Vessel Water Level - (Low Low. >-45.5 inches

b.

~~

i

>-47.7 inches Drywell Pressure - High '

c. Reactor Vessel Water Level - High Level 8 < 1.68 psig .< 1.88 psig

d. Condensate Storage Tank Level - Low 7 52 inches

7 54.2 inches

e. Suppression Pool Water Level - High I ?4 f rS::a* o nor.hes i 3 9 r=5= " --4.5 mc.he d ' .

3 Pump Discharge Pressure - High - = */.0i ,4 46,sa < +-= g. o mc.n e s C,,,,

f.

g. HPCS System Flow Rate - Low 2 45pg "_ "20 l p;t; i en' --":<g

h. Manual Initiation : Saa g 201sgpm -] 145 p:!g-inc n:ta ; 2 14 0 p g' NA g NA w 2 scoggy -

R D. LOSS OF POWER

.:c r in 1. DivisionfDend O h a. 4.16 kw Emergency Bus Undervoltage (Sustained Undervoltage)## a. 4.16 gif Basis -

3 g 3'3 2 3 2970 voltstime 2970 1 148 volts 4-4 sec. j2,Q.^55::.timedelay delay i-

b. 4.16 kw Emergency Bus Undervoltage y ,$3 (Degraded Voltage) a. 4.16 kw Basis -

13740 volts bOI 4 % 4 40 sec. &o!.klesse 3740),J87 S^J.055 volts

_. time delay -

time delay (w/o [?60.05'm time delay LOCA) '

J t o.3 ---+9 sec. time 3t o.33 Sy

+

delay (w/o LOCA) "Tg 4. ';

i "t . .

i.s e

m"amm

,wm m

. F"" 15

o i

.h C 3::=7 "v1 maand

sam.-> auea9as,s-Ja-- ---ar--3h*MA6dd3 -A4- b **----a--e'- -- -e- - - " - - - - ' - a *-m- - - - - - M>+A A s " --+'-d tup 050RE 4

,.o

/ r A 2 h ~

q g og .t - '

% ;J 8 ~k i b r1 i b ' q$ .

!fff !

4 t

~e g-

\a S t if .! Jj 11,JH)6s)*; @) Y 'h g b

! p9g N

. ,. l.. .

xfBh ,h i

\ '

b -.

\ .

- h -* i

2 2 !

i j g\

j* ER *

.c s- J-l s!

i f.

k'f.?!N"*g f ,

sig 2: 14_

E2

? g' S' j ii n i Ct d a ) Shi h

i .

" 9 a

! I- V'l n (4 4 e

'g.b:.

x l Eg ". 2# 3a .1 #

E a ~

d.i

- a

= . . .

l I

! l j l' . I, 1.

!s!)ja;

. 1t L 4

I, sn ih fg,!3 i 1l e N

. . 1 Iy ] IIl!,! N,35g 2 .

                                                       !s Ea,mm*E'd*<                                                                                               e
 . _ - _ .             7_ E_ "[ $                          . ._                               _ . _ _ _ _                                 . _ _ . . _                                                    ___                     _

EMERGENCY CORE COOLING SYSTEMS FIN 1'" D h i

       \        SURVEILLANCE REQUIREMENTS I

4.5.1 ECCS division 1, 2 and 3 shall be demonstrated OPERA 8LE by:

a. At least once per 31 days for the LPCS, LPCI and HPCS systems:

1. Verifying by venting at the high point vents that the system piping from the pump discharge valve to the sy'steTfsolation valve is filled with water.

2. h ri n hat each valve, manual, power operated or automatic

                                      ... . , . sow path that is not locked, sealed, or otherwise secu, red in position, is in its correct position.                                                      i b.

Verifying ths.c, when tested pursuant to Specification 4.0.5, each: .

                        .      1, LPCS pump develops a flow of at least 5010 gpm with a pump differential pressure greater than or equal to 281 psid.

I - 2. LPCI pump develops a flow of at least 5050 gpm with a pump differential pressure greater than or equal to 100 psid. 3. t HPCS pump develops a flow of at least 5010 gpm with a pump

  }f differential pressure greater than or equal to 399 psid.

Y c. For the LPCS, LPCI and HPCS systems, at least once per 18 months, performing a system functional test which includes staulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in

              .-              the flow path actuates to its correct position. Actualinjec-                                      -

' tion of coolant into the reactor vessel may be excluded from this test. - d. For the HPCS system, at least once per 18 months, verifying that the suction is automatically transferred from the condensate storage l tank to the suppression pool on a condensate storage tank low water level signal and on a suppression pool high water level signal, and verifying that the NPCS system will automatically restart on Reactor { Vessel Water Level - Low Low, Level 2. i I 1 APR 2 61985 RIVER BEND - UNIT 1 3/4 5-4 ap ! -- - - - --- - ~ ~'~ ~- '- ~ '~~ ~ ~ ~ ~

_ CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION (Continued) d. ! The combined leakage rate for all penetrations shown in Table 3.6.1.3-as annulus bypass leakage paths exceeding 13,500 cc/hr, or-e.

The combined leakage rate, for all valves shown in Jable .3.6.4-1 to
be equipped with PvtCS, exceeding 170,000 cc/hr, or f.

The measured combined leakage rate for all containment isolation valves i in hydrostatically tested lines per Table 3.6.4-1 which penetrate the primary containment exceeding 1 gpa times the total number of such valves, restore:

a. .

      ,                           .        The overalland applicable,        integrated leakage rate (s) to less than 0.75 La as b.

l c. to Type 8 and C tests to less than or equal to 0.60 L The measured leakage rate to less than 340 scfh for each of the valve o groupings identified in 3.6.1.3.c.1, 3.6.1.3.c.2, and 3.6.1.3.c and d. The combined leakage rate for all penetrations shown in Table 3.6.1.3

e. as annulus bypass leakage paths to less than or equal to 13,50 The combined leakage rate, for all valves shown in Table 3.6.4-1 to be equipped with PVLCS, to less than or equal to

f. ' 170.000 cc/hr, and ,

! The combined leakage rate for al1C *C d. ; -)ntainment isola-tion valves in hydrostatically teswu ...m. ,m. isoie 3.6.4-1 which the total number of such valves, penetrate the primary containm prior to increasing reactor coolant systes temperature above 200'F. SURVEILLANCE REQUIREMENTS ' t 4.6.1.3 The primary containment leakage rates shall be demonstrated at the follow-ing test schedule and shall be determined in conformance with the criteria s 1 in Appendix J of 10 CFR 50 using the methods and provisions of ANSI M . 3 a. Three Type A Overall Integrated Containment Leakage Rate tests shall psig, during each 10 year service period.be conducted at 40 2 The third test.of each set shall be inservice conducted during the shutdown for the 10 year plant inspection. s RIVER BEND - UNIT 1 3/4 6-4 1 se

CONTAINMENT SYSTEMS

me a N MSIV LEAKAGE CONTROL SYSTEM id) w{

LIMITING CONDITION FOR OPERATION 3.6.1.5 divisions shall be OPERABLE.Two independent main steam positive leakage contro APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. - - ~ ~ ACTION: With one MS-PLCS division inoperable, restore the inoperable division to

 , a.,    .

12 hours and in COLD SHUTDOWN within the following 24 hou ,

SURVElltANCE REQUIREMENTS 4.6.1.5 Each MS-PLCS division shall be demonstrated OPERABLE: s. By performing Surveillance Requirement 4.6.1.10.a.

b. At least, once aa-by verifying compressor CPERASILITY by operating th{ompres(Poaded for at least 15 minutes." '

c. ) During each COLD SHUTDOWN, if not performed within the previous 92 days, by cycling each remote, manual and autamatic motor operated valve through at least one complete cycle of full travel. d. At least once per 18 months by performance of a functional test which includes simulated actuation of the division throughout its operating'

sequence, and verifying that each automatic valve actuates to its i

correct in each position steam line. and that 8.5 2 3 psid sealing pressure is established A a k art, 2 i M RIVER BEND - UNIT 1 i 3/4 6-10 l e

CONTAINMENT SYSTEMS

                                                                                             . fl*                   -

x DRYWELL BYPASS LEAKAGE - " ~ d5 ' LIMITING CONDITION FOR OPERATION l l 3.6.2.2 Drywell bypass leakage shall be less than or equal to 10% of the minimum acceptable A//E design value of 1.0 ft.2 ,l APPLICABILITY: When DRYWELL INTEGRITY is required per Specification 3.6.2.1. ACTION: With the drywell bypass leakage greater than 10% of the minimum acceptable A//E design value of 1.0 ft.1, restore the drywell bypass leakage to within the limit prior to increasing reactor coolant system temperature above 200ff. SURVEILLANChREQUIREMENTS 4.6.2.2 The drywell bypass leakage rate test shall be d at least once

f. per 18 months at an initial differential pressure of 3. s dtheA//E

   - !               shall be calculate'd from the measured leakage. One d
 .                                                                                                        lock door shall                          s
                   remain open during the drywell leakage test such that each drywell door is leak tasted during at least every other leakage rate test.

a. If any drywell bypass leakage test fails to meet the specified limit, the schedule for subsequent tests shall be reviewed and approved by the Commission. If two consecutive tests fail to meet the limit, a test shall be performed at least every 9 months until two consecutive! tests meet the limit, at which time the 18 month test schedule may be resumed. ! b. The provisions of Specification 4.0.2 are not applicable. I O km RIVER BEND - UNIT 1 1PR 2 6 SE5 3/4 6-19

                                                                       , _ _ . . _ _ . -     .._.___,,_r.

CONTAINMENT SYSTEMS " LIMITING CONDITION FOR OPERATION (Continued) . t z ACTION: (Continued) B

2. With the suppression pool average water temperature greater than:

a) 95'F for more than 24 hours and THERMAL POWEA--greater than ' 1% of RATED THERMAL POWER, be in at least HOT SHUTOOWN within 12 hours and in COLD SHUTOOWN within the next 24 hours. b) 110'F place the reactor mode switch in the Shutdown position

   -                             and operate a; least one residual heat removal loop in the suppression pool cooling mode.                             -
~
                 . 3. With the suppression pool average water temperature greater than             -

120'F, depressurize the reactor pressure vessel to less than 200-psig within 12 hour

                                                      ~                                       )

c. With only one suppressic water level indicator OPERABLE

 ~                   and/or with fewer than eignt suppression pool water temperature          '

indicators, one in each of AWeight locations, OPERABLE, restore (,, the inoperable indi + to OPERABLE status whithin 7 days or g

verify suppressio i t w ater level and/or temperature to be V within the limits . ce per 12 hours. 4

d. With no suppressio - water level indicators OPERABLE and/or with fewer than sev pression pool water temperature indicators, covering at least seven locations, OPERABLE, restore at least one

         *,          water level indicator and at least six water temperature indicators            -

to CPERABLE status within 48 hours or be in at least HOT SHUTOOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. - SURVEILLANCE REQUIREMENTS . 4.6.3.1 The suppression pool shall be demonstrated OPERABLE:

a. By verifying the suppression pool water volume to be within the limits at least once per 24 hours,

b. At least once per 24 hours, in OPERATIONAL CONDITION 1 or 2, by verifying the suppression pool average water temperature to be less than or equal to 95'F, except:

1. At least once per 5 minutes, during testing which adds heat to the suppression pool, by verifying the suppression pool average water temperature less than or equal to 105'F.

APE 2 6 E5 R!vER BEND - UNIT 1 3/4 6-28

         .                                                                                                                                            l
                               .                                                                                                                      1 m              ,                           mi
      ~

( CONTAINMENT SYSTEMS , SECONDARY CONTAINHENT AUTOMATIC ISOLATION DAMPERS LIMITING CONDITION FOR OPERATION m 3.6.5.3 The secondary containment ventilation system automatic isolation - dampers shown in Table 3.6.5.3-1 shall be OPERABLE with isolation times less than or equal to the times shown in Table 3.6.5.3-1. - - - - APPLICABILITY: As shown in Table 3.6.5.3-1.

      .               ACTION:

With one or more of the secondary containment ventilation system automatic isolation dampers shown in Table 3.6.5.3-1 inoperable, maintain at least one ' isolation damper OPERABLE in each affected penetration that is open, and within 8 hours either: .

a. Restore the inoperable damper (s) to OPERABLE status, or .

b. Isolate each affected penetration by use of at least one deactivat'ed automatic damper secured in the isolation position, or

c. Isolate each affected penetration by use of at least one closed manual valve or blind. flange. . #

Tk psas.u s ei spre,Mk. s c.g or< ru t*Mt s:. htL , Otherwise, in OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours and in COL OWN within the followin (cauaWH/74Gb) g 24 hours. Otherwise, in Operational Conditio spena nanaling of irradiated

               *                                       -te'--t , 00aE ^.'_Tiaf.?! OMS -f ar--ethn; d t'              :      -

Al t,.id.ag ; fuel in the :::r-e y e,;

                              ;t: ::;;   y ,,;;7,g;       ,; ;;;7 .;::::1. The provisions of Specifica-tion 3.0.3 are not applicable.                                                               -

SURVEILLANCE REQUIREMENTS 4.6.5.3 Each secondary containment ventilation system automatic isolation damper shown in Table 3.6.5.3-1 shall be demonstrated OPERABLE:

a. Prior to returning the damper to service after maintenance, repair or replacement work is performed on the damper or its associated actua'.or, control or power circuit, by cycling the damper through at least one complete cycle of full travel and verifying the specified isolation time.

nen irraciated fuel is being handled in the $!d : b'=:.i:-ffu#^0

                      -cesoma u ..;;; :<--                     <

p: t ui e r e:<-4 -- -- - 2 m e RIVER BEND - UNIT 1 3/4 6-52 SPS 2 6 Wi l

i CONTAINMENT SYSTEMS FUEL BUILDING VENTILATION

LIMITING CONDITION FOR OPERATION 4

3.6.5.6 Two independent Fuel Building Ventilation Charcoal Filtration sub-emergency mode.be OPERABLE, and in QPERATIONAL, CONDITION,", one operating in the systems shall ~

                                                                                    '. . w,. , 4                . _ .

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 end ". ACTION: -

  ^ -
   ~                              a.

With one Fuel Building Ventilation Charcoal Filtration subsystem" inoperable, restore the inoperable subsystem to OPERABLE status l within 7 days, or: 1. In CPERATIONAL CONDITION 1, 2 or 3 be in at least HOT SHUTOOWN withinthenext12hoursandinCOLDSHUTDOWNwithinthafollhing 24 hours.

     .                                                                                                                                       ?

2. In Operational Condition *

                                                                                       , suspend handling of irradiated fdel in the ;;;: : ; -aa+ * ;nt , !^"! ^.,' T "".TMMS

! --f :;; r;t i: .;

            '               Fuct O*.(4.,g s fons of Specification 3 0 3 are n t - ^^^: The                      ti:' ':-provi-f::t 'n;       tr.. r;::t:r ;;

I

                                                                                  . .          o applicable.

b. With both Fuel Building Ventilation Charcoal Filtration subsystems inoperable or with one not operating in the emergency mode in Opera-Feel*~b. . tional Condition *, suspend handling of irradiated fuel in the eee-ildq W'- 'fr:i:'n; it;'u...t. r.;;t;r?:=raa'

                                                                          .LT
                                                                           .:::;h
                                                                                  =I= :: :;;r:t?: : efth : .~ a m ei                              ~

The provisions of Specifica-tion 3.0.3. are'not appifcable. l SURVEILLANCE REQUIREMENTS t. 4.6.5.6 1 Each Fuel Building Ventilation Charcoal Filtration subsystem shall be demonstrated OPERABLE: .. 4...

                                                                                             /          u a.

At least once per 12 hours in OfERATIONAL CDNDITION'

hv verifying one Fuel Building Ventilation Charcoa) Filtrat ystem eration.

b. At least once per 31 days by initiating, from the coni.ror room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 hours with the heaters OPERABLE.

                        "when irractated fuel is being handled in thefa:- :mj dw/                         dong
                         *^^: f LT:^"'^"; :-f :;; :t '-" "'t' : ;:t:-t! ' y ::mt;i --: r t :-f Vn-  9;ir.in; th: 7;^;3                 - ".r-RIVER BEND - UNIT 1                                3/4 6-61 485 g g ggge

CONTAINMENT SYSTEMS ' i SURVEILLANCE REQUIREMENTS (Continued)

a. Manual initiation from th_e control room, and

b. Simulated automati signal.

4. Verifying that the filter cooling bypass dampers-een be manually opened and the fan can be manually started.

5. Verifying that the heaters dissipate 3,49 kw when tested in accordance with ANSI' N510-W1980.

f. After each complete or partial replacement of a HEPA filter bank Ay verifying that the HEPA filter bank satisfies the inplace penetration and bypass leakage testing acceptance criterion of less than 0.05% in * '~ accordance with ANSI N510-4W5 while operating the system at a flow rate of 10,000 cfm 2 10%. /9fo - p

                          - g. After each complete or partial replacement of a charcoal adsorber ?                        I bank, by verifying that the charcoal adsorber bank satisfies the e inplace penetration and bypass leakage testing acceptance criterion of less.than 0.05% in accordance with ANSI N510 1995 for a halo-

'b ) . genated hydrocarbon refrigerant test gas whilefhperating the system at 'a flow rate of 10,000 cfm i 10%- firo 1

                                                                                                             -                1 l

RIVER BENO - UNIT 1 3/4 6-63 he g , g I ! 1 I

l PLANT SYSTEMS T s ULTIMATE HEAT SINK k$ I p, ~. ' 7 = :r. d' SURVEILLANCE REQUIREMENTS

4. 7.1. 2 OPERABLE: The standby cooling tower and water storage basin ermined shall be a.

                -->           At least ea+  water  once per 24 hours by verifying the basiriE:$               :     =

te ; U e g. to be within their limits. At less

 .s cell fr            r 31 days by starting the cooling tower fans in each e control room and operating the fan for at least 15 minut f b. pwin3 k months af June 4rmsk SetMen, bekar, tat.
                            .heves of l500 ud Itcc verify 1As. basin wahr kynna/vnal
                   ^         slua m 9E

1 2 ' MSL. (attrwomately gn4 ele.wlim)

                                                                                                      +c
 ~

be behw Hs his t : s (*

t. wAlbelow los+TS*r enu- p 79 whur tha. pu. vie s u.ashg 2

af Itasi enu. pin .2y hours whm tk. pr >nes A.t.Ad<'ny Was yre.4484 % ?S*s' ~ ( RIVER SEND - UNIT 1 3/4 7-4 APA 2 61985 1

          ---                                               ~

M c , rt a [lsa an. 8 g UNITED STATES ENCLOSURE 2

          ;;               'j                NUCLEAR REGULATORY COMMISSION WASWNGTON D. C. 20555 k . . . . + p#                                       May 17, 1985 MEMORANDUM FOR: Edward J. Butcher, Group Leader Technical Specification Review Group Division of Licensing FROM:

M. Dean Houston, Reactor Engineer

 ,,                                     Technical Specification Review Group Division of Licensing

SUBJECT:

DEFICIENCIES IN REGARD TO GSU CERTIFICATION OF RIV TECHNICAL SPECIFICATIONS (FINAL DRAFT) By letter dated May 2,1985. Gulf States Utilitics (GSU) was requested to review by May 13,the final draft of River Bend Technical Specifications and submit 1985, draft plant. accurately reflects the FSAR, SER and as-built configuration By letter dated May 6,1985. GSU submitted their certification, under oath Included in theirandsubmittal affimation, were:of (1) the final draft of River Bend Tech Specs. identified editorial changes 173 items), (2) proposed revisions to the SER (7 items changes to the Tech Specs (38 items), (3) propos(ed to be revised with some propos)ed FSAR revisions (55 areas).and (4) id comission, have been identified.In my review of their submittal, numerous - not identified by GSU are presented in Enclosure 1. Examples of editorial errors that we not detected are circled.from the GSU submittal, some with their markup, and th headings, non-existent trip signals, nomenclature, etc.These errors take many words would have cancontributed be properly interpreted, to operator many of the other unidentified errorsWhile misspe confusion. the GSU review process was less than thorough.to be a complet errors in the final draft were not detected during their review.Approximately This is 20% of the an unacceptable level by any standard fication at licensing must be improved. and their review process for certi-Two of these are editorial errors and one is based on a p statement regarding their as-built plant. (1) Tech Specs as shown.GSU proposed changes to page 3/4 8-5 of the River Be inserted betwecn 10 and seconds.As proposed, the change was to be On page 3/4 8-6 of the 969 o S003.60 xh .nyp-

Edward J. Butcher May 17, 1985 I same submittal GSU properly proposed an identical change. with the clause to be inserted after 10 seconds. As proposed on 3/4 8-5, the revision makes no sense and would confuse the operator. (2) In Attachment B of the GSU submittal of May 6,1985, Item 29 requests a deletion of a surveillance requirement because -

                      "There are no valves in the flow path of any PGCC subsystem."

In past discussions, GSU has resisted this requirement on the basis that the valves did not have a position indicator although, in fact, the valves do have a trip indicator. A copy of FSAR Figure 9.5-13 is enclosed which shows numerous solenoid operated valves as well as a couple of check valves

   > -                in the flow path. Therefore, the GSU statement of "no valves       "

in the flow path" appears to be a false representation. (3) Also, in Attachment B. Item 30 refers to adding a footnote to TS 3/4.7.fu4, Table 3.7.6.4-1. This is in error as the proposed footnote was identified with TS 3/4.7.6.5, Table 3.7.6.5-1. In addition to the deficiencies noted above, I would like to comment briefly on other uncertainties associeted with the River Bend Tech Spec review. GSU has submitted a listing of 55 areas in the FSAR that need revision to support Tech Spec sections. Amendment 19 to the FSAR was delivered on May 14, 1985 and only 12 of these areas were addressed. Therefore, in the other 43 areas, the NRR Technical Reviewer has not seen the necessary documentation to support the current Tech Spec section or a proposed l revision to a section. There is also the potential for additional FSAR revisions resulting from the reviewer's evaluation. This lack of timely information will impact the accelerated schedule for issuance of the Tech _ Specs with the River Bend license in June, 1985. There seem to be some values in the FSAR and Tech Specs that are constantly being changed. For example, the DBA activity release to the environment i following a LOCA (used for containment Tech Spec review) were revised in l Amendment 18 to the FSAR dated April 1985 and revised again (increased) in Amendment 19 on May 13, 1985. In the Tech Specs, GSU has pro the water level for the Ultimate Heat Sink be 112'4" (2nd Draft) ,posed108"6"that (Final Draft) and 111'10" (current revision). Changes of this frequency would indicate that the utilities review process has not settled down. All of the above matters should be given due considerations when discussing commitments and completion schedules for the River Bend Tech Specs.

n. A MA M. Dean Houston, Reactor Engineer. .

Technical Specification Review Group Division of Licensing cc: D. Crutchfield R. Benedict S. Stern

ENCLOSURE . FINAL DRAFf i TECHNICAL SPECIFICATIONS RIVER BEND - UNIT 1 Markup Pages From o, . G5U Submittal of ' 5/6/85 Errors r.ot identified by GSU are circled and noted in margin with X. l Not Intended To Be Complete . April 26, 1985

SAFETY LIMITS . BASES 2.1.3 REACTOR COOLANT SYSTEM PRESSURE

The Safety Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity ;

of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel Code 1971 Edition, including i Addenda through Summer 1973, which permits a maximum pressure transient of 110%, 1375 psig, of design pressure,1250 psig. The Safety Limit of 1325 psig, as measured by the reactor vessel steam done pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The pressure Safety Limit is selected to be the lowest transient overpressure allowed by as . the ASME Boller and Pressure Vessel Code, Section III, Class I. - 2.1.4 REACTOR VESSEL WATER LEVEL - With fuel in the reactor vessel during periods when the reactor is shut sideration aust be given to water level requirements due to the effect eca eat. If the water level should drop below the top of the active irradi-

1 during this period, the ability to remove decay heat is reduced.

    -       This reduction in cooling capability could lead to elevated cladding tempera-f          tures and clad perforation in the event that the water level became less than two-thirds of the core height. The Safety Limit has'been established at the

( top of the active irradiated fuel to provida a point which can be monitored and also provide adequate margin for affe::tive action. e en e . O

                                                                                  -         APR t 81985 RIVER BEND - UNIT 1                         8 2-5                                '

e

               -               LIMITING SAFETY SYSTEM SETTINGS
                 .             BASES
                                                                                                                   ~

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Co the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods when they are tripped. The trip setpoint for each i 17 gallons of water. scram discharge volume is equivalent to a contained volume of app

10. Turbine Stop Valve-Closure . l The turbine stop valve closure trip anticipates the pressure, neutron

           .                 flux, and heat flux increases that would result from closure of the stop ,

valves. With a trip setting of 5% of valve closure from full open, the maintained during the worst case transient. resultant increase in heat flu 11. Turbine Controt Valve Fast Closure, Trio 011 Pressure-Low The turbine control valve fast closure trip anticipates the pressure neutron flux, and heat flux increase that could result from fast closure o,f

     /. -

ailure of the turbine bypass valves.rbine control. valves due to loa The Reactor Protection System (. imustes a trip when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 20 milliseconds after the start of control valve fast closure. This is achieved by the action of the fast

acting solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves.

logic input to the Reactor Protection System.is sensed by press This trip setting, a slower clofure time, and a differect valve characteristic from that of the turbine stopstop the valve, combine to produce transients which are very similar to that for valve. of the Final Safety Analysis Report. Relevant transient analyses are discussed

12. Reactor Mode Switch Shutdown Position The reactor mode switch Shutdown position is a redundant channel to the automatic reactor tripprotective capability. instrumentation channels and provides additional manual i

13. Manual Scram -

i instrumentation channels and provides manual reactor tr! e e RIVER BEND - UNIT 1 8 2-9 pft t 61585 w

a- . REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) -

4. No " slow" control rod, " fast" control rod with individual scram inser- ,

tion time in excess of the limits of ACTION a.2, or otherwise inoperable i control rod occupy adjacent locations in any direction, including the diagonal, to another such control rod. i Otherwise, be in at least HOT SHijTDOWN within 12 hours.

b. With a " slow" control rod (s) not satisfying ACTION a.1, above:

1. Declare the " slow" control rod (s) inoperable, and

2. Perform the Surveillance Requirements of Specification 4.1.3.2.c at least once per 60 days when operation is continued with three or -

more " slow" control rods declared inoperable.

                         ~       '
   ~

Otherwise, be in at least HOT SHUTDOWN within 12 hours.

c. With the maximum scram insertion time of one or more control rods exceed-in~g the maximum scram insertion time limits of Specification 3.1.3.2 as

_ determined by $pecification 4.1.3.2.c, operation may continue provided that: ..

f. ' ". 1. " Slow" control rods, i.e., those which exceed the limits of

( ~- . Specification 3.1.3.2, do not make up more than 20% of the 10% sample of control rods tested. 2. Each of these " slow" control rods satisfies the limits of ACTION a.1.

3. The eight adjacent control rods surrounding each " slow" control rod are:

a) Demonstrated through seasurement within 12 hours to satisfy the ' maximum scram insertion time limits of Specificati.on 3.1.3.2, and b) OPERABLE.

4. The total number of " slow" control rods, as determined by Specifica-tion 4.1.3.2.c when added e sum of ACTION a.3, as determined by Specification 4.1.3.2. an/b does not exceed 5.

k Otherwise, be in at least HOT SHUTDOWN within 12 hours. '

d. The provisions of Specification 3.0.4 are not applicable.

M N - Artt 2 6 385 RIVER BEND - UNIT 1 3/4 1-7 - l l e

              -ama                 a s m-    -       *A            --+a* 4-*~m=   - -a2-   -- 4-A-   a   J_    4:----- Je-REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS                                                  a        *                ' - -

LIMITING CONDITION FOR OPERATION ACTION: : (Continued) - a) Electrically, or b) Hydraulically by closing the drive water and exhaust water isolation valves. With nor X---y2. scram cu@ato ne withdrawn control rod with the associated operating, immediately place the reactor mode switch in t [_#

Shutdown position.

c. The provisions of Specification 3.0.4 are not. applicable. " 4.1.3.3 .Each. control rod scram accumulator Thall be determined O a. At least once per 7 days by verifying that the indicated pressure is and disarmed cr scrammed. greater than or equal to 1520 psig unless the

         .                b. At least once per 18 months by:

(, .

1. Performance of a: a) CHANNEL FUNCTIONAL TEST of the leak detectors, and b) CHANNEL CALIBRATI'ON of the pressure detectors and ve

                   '                             an alarm setpoint of 1520 psig on decreanng p,ressure.rifying

2. '

Verifying th'at each individual accumulator check valve maintains _ the associated greater accumulator pressure above the alarm set point for than or equ operating pressure,al to 10 minutes, starting at normal system l with no control rod drive pump operating. RIVER BEND - UNIT 1 3/4 1-10 Wt6E i i

              =         *                                                           .

INSTRUMENTATION FINAL DRAFT METEOROLOGICAL MONITORING INSTRUMENTATION N LIMITING CONDITION FOR OPERATION . 3.3.7.3 Theoefteorologicalmonitoringinstrumentationchannelsshownin Table 3.3.7.3-1 shall be OPERABLE.

                 "' '""" T TY: At all times.

k ith one or more meteorological monitoring instrumentation channels ' inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant.to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrumentation to OPERABLE status. ? - b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. o ~ SURVEILLANCE REQUIREMENTS

b. - 4.3.7.3 Each of the above required meteorological monitoring instrumentation

   '            channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.3-1.

em G* e e . *

                                                                                          -     ..        j
             ' RIVER BEND - UNIT 1                     3/4 3-73                       APR 2 s 1985

4 . . p.. .A. _ w. . ._.._..u .a - _ - .- -- - g X FINAL DRAFT e t t *

                                   .=                       %

t -- t gik , - - - - --- --- -

E' y . M G , g _

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                    . M          . 2 2 2                    2 2           "

g E' G G G G G

                                                                                       't
                         ~                  .     .       .           .

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                .i                                                                                              h"GG2C                G 4.

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                                        .             s e
                                                                              -      *3 e
                                                                                                 .            33_                                    **

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                                                                                                                                                    .w=
              ,                  8< A        a     et          4 i           s      n'        d                                                      r=

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                                                                                                                     -                            i .G G RIVER BEND - WIT 1                                      3/4 3-77 APR 2 s a65
                                                                                                                       .                   ..                 N*

_ m ,

\ .

A FINAL. DWT _ TABLE 3.3.7.4-2 REMDTE SHUTOOWN SYSTEM CONTROLS

                                                     .                                MINIMUw rumuurti e r e 1.

GIV. J.Z-Q Olv. Jh RCIC Suction from CST MOV 1 NA (IE51*MOVF010)

2. RCIC Injection Shutoff MOV 1 NA (1E51*MOVF013)

3. RCIC Min. Flow to Suppression 1 NA

4. Pool MOV (1E51*MOVF019)

_ , , RCIC Test Bypass to CST MOV (1E51*MOVF022) 1 NA

5. RCIC Gland Seal Air Compressor "

      ~                                  (1E51*PC002C)                            -

1 NA

6. RCIC Pump Suction from Suppression Pool MOV (1E51*M0VF031) 1 e

7. RCIC Steam to Turbine MOV F M-(IE51*MOVF045) 1 NA

8. RCIC Turbine tube Oil Cooling MOV (IE51*MOVF046) 1 NA '

9. '

RCIC Test Bypass to CST MOV F M-1 4 (\ (IE51*MOVF059)

10. RCIC Steam Supply Inboard Isolation NA 1 NA MOV(1E51*MOVF063)

11. RCIC Steam Supply Dutboard Isolation 1 NA MOV(IE51*M0VF064) en eee.

12. RCIC Turbine Exhaust to

                             *~                                     a Pool                  1            NA MOV(1E51*MOVF068)

13. _, i RCIC Steam Line Warausi Line Isolation 1 NA ' l

14. MOV(1E51*MovF076) i RCIC Vacuum Breaker Outboard Isolation 1 NA

        ..              15.           MOV(1E51*MOVF077)

RCIC Vac um Breaker Inboard Isolation 1 NA MOV(1 VF078)

16. RCIC Turbine Flow Controller 1 (IC61*FICR001) NA

17. RCIC Turbine Trip & Throttling MOV

                     .                                                                     1            NA 1 81      W.            (IE51*MOVFC002)

RHR Pump (1E12"PC002A,28,2C) 1 2(,)

              ~18      M.

RHR Hx Shell Side Outlet MOV 1 (IE12*MOVF003A. 8) 1 st Str. RHR Pump Suction MOV (1E12*MOVF004A, B; 1E12*M0VF105) 1 2(,) o M. RHR Shutdown Cooling MOV 2g,) f

                   %                 (IE12*M0VF006A. 68)                                              NA                           '

2. RCic Tureme L g ig Set u 4 S w',4c k i

                     .(a) One per control equipment                                                   y4 RIVER BEND - UNIT 1                            3/4 3-78 t                                                                                                                           _--

                                                                                                                                             ,                              l
                                                    -TABLE 3.3.7.4-2 (Continued)

REMOTE SHt1TDOWN SYSTEM CONTROLS

MINIM 58 PW8WWF {lv. LE} FC87]" DIV. C AS.e; RHR Outboard Shutdown Isolation MOV 1 NA (IE12*MOVF006) 2.y H. RHR Inboard Shutdown Isolation MOV 1 NA NN/ gec/%y (IE12*MOVF009) med.. ,A u .M. RHR Hx Flow to % g[ col MOV 1 fm,, 1 I (1E12*MOVF011A, 3) u fS. RHR Reactor Head Spray MOV (1E12*MOVF023) 1 NA M7/M I~I u M. RHR Test Line MOV 1 (1E12*MOVF024A, 8) 1 gy l tr N. RHR Hx Flow to RCIC MOV (1E12*MOVF026A) 1 MA J-E u M. RHR Injection Shuteff MOV 1 (1E12*MOVF027A, B) 1

      ~                                                               -

3 49. RHR Upper Pool Cooling Shutoff MOV 1 1 (IE12*MOVF37A. 8) 3: De. RHR Injection MOV 2(,)

                                                                                                                                         .1 (1E12*MOVF042A, 8. C)

11. M. RHR Mx Shell Side Inlet MOV (IE12*MOVF047A, 8) 1 1

   -(          37 M.        RHR Mx Shell Side Bypass MOV (1E12*MOVF048A,8) 1    .      1
      . \ sy DS. RHR Discharge to Radwaste MOV 1           NA (1E12*MovF040) sr 34. RHR Steam Isolation MOV                                                                                            1            1 (IE12*MOVF052A,B) st 35. RHRInjectionMOV                                                                                                     1
                        ~

(1E12*MOVF053A,8) 1 n Mr. RHR Pump Minimum Flow MOV 1 2(,) (1E12*MOVF064A, 8, C) Sr 37. ' RHR Nx Water Discharge MOV 1 (1E12*MOVF068A,B) 1

        . . Si 36. Safety Relief Valves                                                                                                3g,)         3(,)

(1821*RVF051, C G, D) 4 35F. SSW Pump (1SWP"P2A, 2f.T 28, 2D) 1+N 2(a) att 46. Normal Service Water Isolation MOV 1 1 (1SWP"MOV96A,8) 5:144. SSW Cooling Tower Inlet MOV 1 1

                          . (ISWP*MOV55A, B)

(a) (One per control equipment N Ssw pap iswa = P2c as a Divis .o IT c. RIVER BEND - UNIT 1 3/4 3-79 emea4. Lal w,,1 h ! 61985 is gr.,;4 g. O L

o-

                                                     -                                            _ __    _      ..                      ~                                                                             -

e' , f . i . m

                                                                                                            '_TA8tE 4.3.7.5-1
             .m

= ACCIDENT MONITORING INSTRUMENTA'l0N i

i SURVEfttANCE REQUIREMENTS '

              .3 m

E INSTRUMENT CHAML. CHANNEL APPLICA8tt

_ CHECK OPERATIONAL

1. ' CALIBRATION Reactor Vessel Pressure CONDITIONS z 2. M Reactor Vessel Water Level .

R 1, 2 Z a. Wide Range w b. Fuel Zone M R M 1, 2

3. Suppression Pool Water level ,

R 1, 2 4 Suppressipn Pool Water Temperature M ' R M 1,2,1 S. 7. ' .,. . , 0 .t . . - ..^. ^. 0 , b . . ^ m m " ,, R 1,2,f

6. Primary Containment Pressure : =,;

7 Drywell Pressure . M R M 1, 2

8. Drywell Air Temperature R 1, 2

9. M R Drywell and Primary Contatnment Sjd.;;;;. Concentration M 1, 2 Analyzer and Monitor Q* 1, 2 R* 10. Safety / Relief Valve Position Indicators M l

Y *11. Area Radiati d 7. L ., 0,...^. x,...^,0. R 1, 2

                        *?. ": ":'u n t "x"'*:t':: C tx ;t "_ '^ s ' ;** ^ = 4                                      .                       4                                             4-I          !?. "
                                  ..,".,__J: - *:d" ;ff -? " xd!' ; ^ :: " *:::                                                                                                                                               I    2, 3 I'        !" ;2, 0 :            W r"- ;; 07t s !:t;_;0 " :.!ts ,                                                     "                                                                                         !, ?, 3 '

4 1

15. O--0::

                                  . __ ._. n , _ _ E x 2 ":i;;;t: 5 '?d' ; " nt : !:t: ;t                                                                                             :                                           1, 2, 3 I5.      ' --1':: ":' ?d' ;
                                                                       "x'::-'" ;;' ; ?-- ; trix;t                                                                                                                                1, ? , 3-
                                -.__._m.__

1, 2.-3

'                      On Fusing sample gas containing:

a. One volume percent hydrogen, balance nitrogen. %

emmes M Four volume percent hydrogen, balance nitrogen. , i p g

                    **withhe                CipMMEL cal!8 RAT 10N shall consist of an electron {c calibration an installed or portable gamma source.

y

         ~n            R

{ w High range 2:t': ;21 monitors. gms 3acame , c. g Gg f . p,, ,y co,& ,,,,,I A m

            --                                                                                *        '                                                                                                                                   s?aD m

on..it Am R G3 "ini w n i,z. , M

                                                                                                                                 , _ _ _ _ ,                               ~.-      ~ * *                          =                '

H. ANNE PLETTINGER M 56 Villa Rose Drive BATON ROUGE, LOUISIANA 70806 504 343-9333 CDITIFID t%IL - RE11RN RECE[PT REQUETED July 17, 1985 'b #

                                                                                  ,-M.j @ 4 c.t Mr. Jams Feltm, Director                                                   F0% A ~tS-3l}

Office of Admnistration Ig %J ., (( U. S. Nuclear Pegulatory Omnission Washmgton, D. C. 20555 River Bend Station, Unit 1 Gulf States Utilities Co. Docket No. 50-458 Dear Mr. Feltcm lhis is a "Fmedcri of Informtim Act" request pursuant to 5 U.S.C. Sec. 552, et seq. and 10 C.F.R., Part 9, Subpart A. Please serx1 copies of all infoutntion (incitrimg but not limited to: letters and teleem mssages frua Gulf States Orpanv (GSU) to the NBC, meting notes, internal staff rences and letters and teleccn messages frm the NRC to GSU) regarding the Technical Specifications for River Bend, thit 1. In the unlikely event that access and c: pies are denied to any part of the requested records, please describe the deleted unterial and specify the statutory basis for the denial as well as your reasons for believing that tin alleged statutory justification applies in this instance. Also, please stata separately your reascns for not invoking your discretionary powers to release the requested doctrents in the public interest. I tc,mst that you waive any applicable fees as disclosure mets the statutory standard in that it wourt be "in the public interest because furnishing the information can be considerni as prinnrily Lwfiting the general public." 5 LK:S 552 (a)(4)(A). I am a landowner in the vicinity of River Beni Station and I have very closely followed this docket since early 1975. Should you have any questions about my request for waiver of any applicable fees, please call me . Sinmrely, fhl.4LC H. Anne Flettinger o cn n ni a ., n r3 - vvv i wp u ~ u LJ

                                  })    .

.}}

ML20212K411

Contents

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2.0 INTRODUCTION

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ML20212K411

Text

SUBJECT:

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2.0 INTRODUCTION

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