Text

Proposed Tech Specs Re Engineered Safety Sys Which Responds to Fuel Handling Accident Conditions
	ML20087J545
Person / Time
Site:	River Bend
Issue date:	08/17/1995
From:	; ENTERGY OPERATIONS, INC.
To:	;
Shared Package
ML20087J540	List: ML20087J545; RBG-41728, LAR 95-04 to License NPF-47,revising TS Re Engineered Safety Sys Which Responds to Fuel Handling Accident Conditions;
References
	NUDOCS 9508210247
	Download: ML20087J545 (126)
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{{#Wiki_filter:e Secondary Containment' Isolation Instrumentation 3.3.6.2 L . ~Tabte 3.3.6.2 1'ipage 1'of 1) secondary Containment Isolation Instrumentation APPLICABLE

MODES AND -REQUIRED

                                                                       . 0THER           CHANNELS SPECIFIED          PER TRIP          SURVEILLANCE-      - ALLOWABLE
                              ' FUNCTION                             CONDITIONS-          SYSTEM           REQU*REMENTS            VALUE

1. Reactor vessel Water '1,2,3 2 sR '3.3.6.2.1 a: -47 inches Level - Low Low, Level 2 sa 3.3.6.2.2 st' 3.3.6.2.3 st 3.3.6.2.4 sR 3.3.6.2.5

            ' 2. Drywell Pressure - High                                 1,2,3              2             sR   3.3.6.2.1  s 1.88 paid SR 3.3.6.2.2 SR 3.3.6.2.3

[ SR 3.3.6.2.4 SR 3.3.6.2.5

3. Fuel Building ventilation (a) 1 SR 3.3.6.2.1 s .18 x 10 3 Exhaust Radiation- High SR 3.3.6.2.2 pc[/sec

                    - (1RMS*RESA)                                                                         SR   3.3.6.2.4 SR   3.3.6.2.5

4. -Fuel Building ventilation (a) 1 SR' 3.3.6.2.1 s 7.05 x 10 4 Exhaust Radiat10n- High SR 3.3.6.2.2 pcf/cc (1RMS*RE58) sa 3.3.6.2.4 st 3.3.6.2.5

5. Manual Initiation 1,2,3, 2 sa 3.3.6.2.5 NA (a)

(a) Duringmovementofrecentifijirradiatedfuelassembliesinthefuelbuildingforfuelbuilding.

                   -isolation.                                                                                                                  !

i i

,)

l l l RIVER BEND 3.3-61 Amendment No. 81 LATER 9508210247 950817 PDR P ADOCK 05000458 PDR

CRFA System Instrumentation 3.3.7.1 s Table 3.3.7.1-1 (page 1 of 1) Control Room Fresh Air System Instrisnentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FRON SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION A.1 REQUIREMENTS VALUE

1. Reactor vessel Water 1,2,3 2 3.3.7.1.1 B SR 2: -47 inches Level - Low Low, Level 2 SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4 SR 3.3.7.1.5

2. Drywell Pressure - High 1,2,3 2 C SR 3.3.7.1.1 s 1.88 psid SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4 SR 3.3.7.1.5

3. Control Room 1,2,3 3.3.7.1.1 1 D SR s 0.97 x 1D

Ventilation Radiation (a),(b) SR 3.3.7.1.2 pci/cc Monitors SR 3.3.7.1.4 SR 3.3.7.1.5 (a) During operations with a potential for t raining the reactor vessel. i (b) During c^a5 ."1!5".'"?! rd it ravemet.t of recenttpjf rradiated fuel assenblies in the primary or secondary containment. RIVER BEND 3.3-71 Amendment No. 81 LATER i t l (

I Primary Containment Air Locks 3.6.1.2 ' 3.6' CONTAINMENT SYSTEMS 3.6.1.2 Primary Containment Air Locks LCO 3. 6.1. 2: .Two primary containment air locks shall be OPERABLE. h APPLICABILITY: MODES 1, 2, and 3, . Duringmovementofpecentif3irradiatedfuelassembliesin the primary containment', t An. =.4. m.m, P. A.. D.C..A.I. T.

                       .       .                      . C. D A.T.
                                                                . T A.M. .C. ,

During operations with a potential for draining the reactor ; vessel (0PDRVs). l t i

ACTIONS ___..................____..........--NOTES------------------'------------------ >

1. Entry and exit is permissible to perform repairs of the affected air lock :

components. '

2. Separate Condition entry is allowed for each air lock.

3. Enter' applicable Conditions and Required Actions of LC0 3.6.1.1, " Primary Containment-Uperating," when air lock leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3.

CONDITION I REQUIRED ACTION' COMPLETION TIME A. One or more primary ------------NOTES------------ i containment air locks 1. Required Actions A.1, i with one primary A.2, and A.3 are not containment air lock applicable if both doors door inoperable. in the same air lock are inoperable and Condition C is entered.

2. Entry and exit is permissible for 7 days under administrative I controls if both air locks are inoperable.

l (continued) l RIVER BEND 3.6-3 Amendment No. B1 LATER

Primary Containment Air Locks 3.6.1.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.3 Restore air lock to 24 hours OPERABLE status. D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time of Condition A, AND B, or C not met in MODE 1, 2, or 3. D.2 Be in MODE 4. 36 hours E. Required Action and E.1 Suspend.. movement of Immediately associated Completion recently! irradiated Time of Condition A, fuel ~ asssmblies in B, or C not met during the primary movement of recently containment. irradiated fuel assemblies in the AND primary containmentr CORE ^.LTERATIONS, or E.2 Su: pend CORE Immed htcly OPDRVs. ALTERATIO % AND E.32 Initiate action to Immediately suspend OPDRVs. l l RIVER BEND 3.6-6 Amendment No. B1 LATER

Primary Containment-Shutdown 3.6.1.10 l

     -3.6. CONTAINMENT SYSTEMS 3.6.1.10 Primary Containment-Shutdown                                                      i LCO- 3.6.1.10        Primary containment shall be OPERABLE.

1 APPLICABILITY: During movement of Escintif! irradiated fuel ' assemblies in the primary containmenf, 0" Tin" COPl ."IIER^.I!Ofl, During operations with a potential for draining the reactor i vessel (0PDRVs). i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME' i A. Primary containment A.1 Suspend movement of Immediately inoperable. rscentlji! irradiated

fuel ~aisEmbliesin i the primary , containment. l AND < l A.2 Su:per.d CDP.E Int: dict:1y , ALTER ^.TIONS . AND l A.32 Initiate action to Immediately suspend OPDRVs. l 1 l 1 l l 1 l RIVER BEND 3.6-31 Amendment No. 84 LATER l l

SCIDs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Dampers (SCIDs) LCO 3.6.4.2 Each SCID shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of Fecently3 irradiated fuel assemblies in the fuel buildins"for ~ fuel building isolation. ACTIONS

  .....................___..__.......      .N0TES------------------------------------

1. Penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by SCIDs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected 8 hours penetration flow paths penetration flow path with one SCID by use of at least inoperable. one closed and de-activated automatic damper, closed manual damper, or blind flange. AND (continued) RIVER BEND 3.6-48 Amendment No. 81 LATER

SCIDs 3.6.4.2 ! l, CONDITION REQUIRED ACTION COMPLETION TIME

    . A.   (continued)          A.2         --------NOTE---------                            l

Isolation devices in l high radiation areas may be verified by ;

use of administrative ; means. ' Verify the affected Once per 31 days

                                           . penetration flow path is isolated.

B. One or more B.1 Isolate the affected 4 hours penetration flow paths penetration flow path with two SCIDs by use of at least inoperable. one closed and i de-activated -! automatic damper, ' closed manual damper, ; or blind flange. , C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND or B not met in MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours D. Required Action and -------------NOTE------------ associated Completion LC0 3.0.3 is not applicable. Time of Condition A ----------------------------- or B not met during 3 movement of FedEntl D1 Suspend movement of Immediately irradiated fuel ~ ~ ' Ncentlyj] irradiated assemblies in the fuel f6el"issemblies in building. the fuel building. l RIVER BEND 3.6-49 Amendment No. 81 LATER !

Fuel Building 3.6.4.5 3.6 CONTAINMENT SYSTEMS 3.6.4.5 Fuel Building LCO 3.6.4.5 The fuel building shall be OPERABLE. APPLICABJ TY: During movement of Fecsntly( irradiated fuel assemblies in the fuel building. ACTIONS l ____________.-----NOTE------------------------------------------ LC0 3.0.3 is not applicable. I l CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel building A.1 Suspend movement of Immediately inoperable. EscentljIirradiated fuel ass 5mblies in the fuel building. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.5.1 Verify fuel building vacuum is 2: 0.25 24 hours inch of vacuum water gauge. SR 3.6.4.5.2 Verify all fuel building equipment 31 days hatches and shield blocks are installed. SR 3.6.4.5.3 Verify each fuel building access door is 31 days closed, except when the access opening is being used for entry and exit.

1 RIVER BEND 3.6-55 Amendment No. 81 LATER

Fuel Building Ventilaticn System-Fuel Handling 3.6.4.7 3.6 CONTAINMENT SYSTEMS

 .3.6.4.7         Fuel Building Ventilation System - Fuel Handling LC0        3.6.4.7       Two fuel building ventilation charcoal filtration subsystems

' shall be OPERABLE and one shall be operating in emergency mode. APPLICABILITY: During movement of reeentlyjirradiated fuel assemblies in the fuel buildihg; '~' ACTIONS j

                     ....___..__..--------NOTE------------------------------------------

LCO 3.0.3 is not applicable. CONDITION REQUIRED ACTION COMPLETION TIME i A. One fuel building A.1 Restore fuel building 7 days i ventilation charcoal ventilation charcoal ; filtration subsystem filtration subsystem ' inoperable. to OPERABLE status. B. Required Action and B.1 Suspend movement of Immediately i associated Completion Feeentlyjirradiated Time of Condition A ' fuel ~aisemblies in not met. the fuel building. 0.3 Two fuel building ventilation charcoal filtration subsystems inoperable. O_E One fuel building ventilation charcoal filtration subsystem not in operation. RIVER BEND 3.6-58 Amendment No. 84 LATER

                  ~
     ,                                                                             'CRFA System

, 3.7.2 3.7 PLANT SYSTEM ! 3.7.2 Control Room Fresh Air (CRFA) System LC0 3.7.2 Two CRFA subsystems shall be'0PERABLE. i APPLICABILITY: MODES 1, 2, and 3, , DuringmovementofFecintlj31irradiatedfuelassembliesin the primary or"YsEbndafy containment, During CORE ^.LTERATIONS, During operations with a potential for draining the reactor vessel (OPDRVs). ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CRFA subsystem A.1 Restore CRFA 7 days inoperable. subsystem to OPERABLE status.. 1

          -B. Required Action and         B.1        Be in MODE 3.           12 hours Associated Completion                                                                 >

Time of Condition A AND not met in MODE 1, 2, ' or 3. B.2 Be in MODE 4. 36 hours (continued)

                                                                                                   'l RIVER BEND                                3.7-5                 Amendment No. El LATER 4

CRFA System 3.7.2

       -ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and ------------NOTE------------- associated Completion LC0 3.0.3 is not applicable. Time of Condition A ----------------------------- not met during movementofrecentli C.1 Place OPERABLE CRFA Immediately irradiated fuel' ~ ~ subsystem in assemblies in the emergency mode. primary or secondary containment, during 0_R CORE ALTERATION &r or during OPDRVs. C.2.1 Suspend movement of Immediately recentljjirradiated fuel Tssemblies in the primary and secondary containment. AND C.2.2 Suspend CORE I cdiately ALTERATIONS. . AND C.2.32 Initiate action to Immediately suspend OPDRVs. D. Two CRFA subsystems D.1 Enter LC0 3.0.3. Immediately inoperable in MODE 1, 2, or 3. (continued) RIVER BEND 3.7-6 Amendment No. 81 LATER

CRFA System 3.7.2 i ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. Two CRFA subsystems E.1 Suspend movement of Immediately ' inoperable during Eecentlyjirradiated ; movement of resently fdel~ assemblies in ; irradiated fuel ^ ^ the primary and j assemblies in the secondary ' primary or secondary containment. containment, during i CORE ALTERATIONS, or AND during OPDRVs. E.2 Suspend CORE !=cdi ately 1 ALTERATIONS. ; AND E.32 Initiate action to Immediately ) suspend OPDRVs. ; l l l l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.I Operate each CRFA subsystem for ;a: 10 31 days continuous hours with the heaters operating, l SR 3.7.2.2 Perform required CRFA filter testing in In accordance i accordance with the Ventilation Filter with the VFTP l Testing Program (VFTP). I 1 SR 3.7.2.3 Verify each CRFA subsystem actuates on an 18 months actual or simulated initiation signal, o (continued) RIVER BEND 3.7-7 Amendment No. M LATER

7A 9 Control Room AC SystGm 3.7.3 . c 3.71 PLANT SYSTEMS ! 3.7.3 Control. Room Air Conditioning (AC) System 6 LC0 3.7.3 Two control room AC subsystems shall be OPERABLE. APPLICABILITY: . MODES 1, 2, and 3, During movement of.Fecehtlfjirradiated fuel assemblies in ' the primary orsecbHdafy containment, ' During CORE ALTERATIONS, . During operations with a potential for draining the reactor vessel (0PDRVs). t ACTIONS i CONDITION REQUIRED ACTION COMPLETION TIME. i A. One control room AC A.1 Restore control room 30 days subsystem inoperable. AC subsystem to OPERABLE status. B. Two control room AC B.1 Verify control room Once per 4 hours' > subsystems inoperable, area temperature

                                                   . :s 104

F .

l AND j B.2 Restore one control 7 days - room AC subsystem to OPERABLE status. l

                                                                                                 'l l

C. Required Action and C.1 Be in MODE 3. 12 hours i Associated Completion 1 Time of Condition A or AND l B not met in MODE 1, l 2, or 3. C.2 Be in MODE 4. 36 hours j (continued) i RIVER BEND 3.7-9 Amendment No. 81 LATER

Control Room AC System 3.7.3 , \. ACTIONS (continued)- 4 CONDITION REQUIRED ACTION COMPLETION TIME' , D. Required. Action and .------------NOTE-------------

                -associated Completion  LC0 3.0.3 is not applicable.

Time of Condition A ----------------------------- not met during movement of re6enti D.1 Place OPERABLE Immediately irradiated fuel

~'

control-room AC < assemblies in the subsystem in primary or secondary operation. ! containment, during

    -            CORE ALTERATIONS, or  OR during OPDRVs.

D.2.1 Suspend movement of Immediately recentlyjirradiated fuel ~ assemblies in the primary and , secondary ; containment. ; ONE , 0.2.2 Suspend CORE I : diately

                                                     ^LTERATIONS.

AND D.2.32 Initiate action to Immediately , suspend OPDRVs. ! r (continued) . l 1 1 RIVER BEND 3.7-10 Amendment No. 81 LATER : l 1

.I

a Control Room AC System 3.7.3 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME. f E. Required Action and E.1 Suspend movement of Immediately associated Completion recentlyjirradiated Time of Condition B fuel 7sssemblies in not met during the primary and ; movementofrecentlj secondary -i irradiated f6el" ~" containment. assemblies in the : primary or secondary AND containment, during ' CORE ALTERATIONS, or E.2 Suspend CORE In:cdictcly during OPDRVs. ALTERATIONS. AND E.32' Initiate action to Immediately suspend OPDRVs. SURVEILLANCE REQUIREMENTS- ) SURVEILLANCE FREQUENCY ) I SR 3.7.3.1 Verify each control room AC subsystem has 18 months the capability to remove the assumed heat load. I i i RIVER BEND 3.7-11 Amendment No. 81 LATER I l

AC Sources--Shutdown 3.8.2

3.8 ELECTRICAL POWER SYSTEMS

      ,3.8.2   AC Sources--Shutdown LCO. 3.8.2       .The following AC electrical power. sources shall be OPERABLE:

a. One qualified circuit between the offsite transmission network and the onsite Class IE AC electrical power distribution subsystem (s) required by LC0 3.8.10,

                               " Distribution Systems--Shutdown"; and

b. One diesel generator (DG) capable of supplying one division of the Division I or IIL onsite Class IE AC electrical power distribution subsystem (s) required by LC0 3.8.10; and

c. One qualified circuit, other than the circuit in LC0 3.8.2.a. between the offsite transmission and the Division'III onsite Class IE electrical power distribution subsystem, or the Division III DG capable of supplying the Division III onsite~ Class IE AC electrical power distribution subsystem, when the

                             -Division III onsite Class IE electrical power distribution subsystem is required by LC0 3.8.10.

APPLICABILITY: MODES 4 and 5, Duringmovementofredentlys'irradiatedfuelassembliesin the primary contsinment or fuel. building. RIVER BEND 3.8-17 Amendment No. && LATER

I AC Sources-Shutdown 3.8.2 ACTIONS

                       ---------------------------------------NOTE-----------------------------------

LC0 3.0.3 is not applicable. CONDITION REQUIRED ACTION COMPLETION TIME l A. LC0 Item a not met. ------------NOTE------------- Enter applicable Condition and Required Actions of LC0 3.8.10, when any required division is de-energized as a result of Condition A. A.1 Declare affected Immediately required feature (s)

    ..                                                                                 with no offsite power available from a required circuit inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS. AND A.2.2 Suspend movement of Immediately Fecently] irradiated fuel ~assimb1ies in the primary containment and fuel building. ' AND A.2.3 Initiate action to Immediately I suspend operations with a potential for i- draining the reactor vessel (0PDRVs). AND (continued) i RIVER BEND 3.8-18 Amendment No. 81 L.ATER

6 AC Sources-Shutdown-3.8.2 ACTIONS' CONDITION REQUIRED ACTION. COMPLETION TIME-A. (continued) A.2.4 Initiate' action to Immediately restore required offsite power circuit to OPERABLE status. B. LCO Item b not met. B.1 Suspend CORE Immediately ALTERATIONS. AND B.2 Suspend movement of Immediately recently] irradiated fuer~issemblies in primary containment and fuel building. AND B.3 Initiate action to Immediately suspend OPDRVs. AND B.4 Initiate action to Immediately restore required DG to OPERABLE status. C. LC0 Item c not met. C.1 Declare High Pressure 72 hours

                                                                                            . Core Spray System and Standby Service Water                          I System pump 2C                                 l inoperable. '                                 l
                                                                                                                                                  )

RIVER BEND 3.8-19 Amendme.'t No. 81 LATER

 ^~
 }'r' DC Sources--Shutdown    ,

3.8.5.

              - 3.8 ' ELECTRICAL. POWER SYSTEMS 3,8.5 DC Sources--Shutdown LCO '3.8.51          The following shall be OPERABLE:

a. One Class IE DC electrical power subsystem capable of supplying one division of the Division I or II onsite Class 1E DC electrical power distribution subsystem (s) required by LC0 3.8.10, " Distribution Systems-Shutdown";

l

b. One Class 1E battery or battery. charger, other than the DC electrical power. subsystem in LC0 3.8.5.a, capable of supplying the remaining Division I or:II onsite Class 1E DC electrical power distribution subsystem (s) when required by LCO 3.8.10;'and

c. The Division III CC electrical' power subsystem capable of supplying the Otvision III onsite Class ~ IE DC electrical power d:stribution subsystem, when the Division III onsite Class 1E DC-electrical power distribution subsystem is required by'LC0 3.8.10.

APPLICABILITY: MODES 4 and 5, During movement of E'sEsnt]flirradiated fuel assemblies in the primary containmenf or fuel building. i l

                                                                                                      'l RIVER BEND                                  3.8-28              Amendment No. 84 LATER

DC Sources-Shutdown 3.8.5 ACTIONS _______.___............______________..N0TE----------------------------------- LC0 3.0.3 is not applicable CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare affected Immediately DC e.lectrical power required feature (s) subsystems inoperable. inoperable. 0.8 A.2.1 Suspend CORE Immediately ALTERATIONS. AND A.2.2 Suspend movement of Immediately recentlyll irradiated fuel assemblies in ' the primary containment and fuel building. AND A.2.3 Initiate action to Immediately suspend operations with a potential for draining the reactor ; vessel. AND A.2.4 Initiate action to Immediately restore required DC electrical power subsystems to OPERABLE status. l I 1 I RIVER BEND 3.8-29 Amendment No. 84 LATER

Inverters- Shutdown 3.8.8~ ; 3.8L ELECTRICAL POWER SYSTEMS I gE. 3.8.8 Inverters--Shutdown ' ! t c, , LCO 3.8.8 One Divisional inverter.shall be OPERABLE capable-of ! supplying one division of the Division I or II onsite Class' ' H 1E uninterruptible AC vital bus electrical power ,

;F                             distribution subsystem (s) required by LC0 3.8.10, g-                               " Distribution Systems-Shutdown".

APPLICABILITY: MODES 4 and 5, During movement of EsEehtlylirradiated fuel' assemblies in i the primary cont'hinmeni or fuel building. ACTIONS j

            --..__.....____........__...._______..-NOTE----------------------__..------_-.                          4
            0 3.0.3 is not' applicable CONDITION                        REQUIRED ACTION             COMPLETION TIME                 :

A. One or more required A.1 Declare affected Immediately ! inverters inoperable, required feature (s) i inoperable. i OB i 1 A.2.1 Suspend CORE Immediately . ALTERATIONS. 1 AND i A.2.2 Suspend handlinglof Immediately J recently] irradiated fueT'issemblies in j the primary j containment or fuel I building. ! AND _ . , I(continued) RIVER BEND 3.8-36 Amendment No. 81 LA[ER

t 5 !. Distribution Syste2s--Shutdown 3.8.10- l t 3.8 ELECTRICAL. POWER SYSTEMS l 3.8.10 Distribution Systems--Shutdown LC0 3.8.10- The necessary portions of the Division I, Division II, and , Division III AC, DC, and Division I and II AC vital bus electrical power distribution subsystems shall be OPERABLE i to support equipment required to be OPERABLE. ! APPLICABILITY: MODES 4 and 5, ! During movement of Fedently! irradiated fuel assemblies in ' the primary containment or' fuel building. t i ACTIONS ____________________.------------------NOTE----------------------------------- ; LC0 3.0.3 is not appliccble ; l CONDITION REQUIRED ACTION COMPLETION TIME ! A. One or more required A.1 Declare associated Immediately ! AC, DC, or AC vital supported required { bus electrical power feature (s) ! distribution inoperable. l subsystems inoperable. . OR l A.2.1 ' Suspend CORE Immediately 4 ALTERATIONS. :) AND I J A.2.2 Suspend movement of Immediately l Fecentli} irradiated j f6el assemblies in ; the primary ; containment and fuel ! building.. : AND (continued) l RIVER BEND 3.8-41 Amendment No. 81 LATER l

72 i Enclosute 5 Revised ITS Bases L 4 h P 0 t i e g' h i 1 b Y

k Szcondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 2. Drywell Pressure-Hiah j SAFETY ANALYSES, (continued) LCO, and' ' APPLICABILITY steam and gas. This Function is not required in MODES 4 and 5 because the probability and consequences of these r events are low due to the RCS pressure and temperature limitations of these MODES. 3 and 4. Fuel Buildino Ventilation Exhaust Radiation-Hiah High secondary containment exhaust radiation is an l indication of possible gross failure of the fuel cladding. 1 The release may have originated from the primary containment -1 due to a break in the RCPB or the fuel building due to a ! fuel handling accident. When Exhaust Radiation-High is i detected, secondary containment isolation and actuation of ! the associated ventilation system are initiated to limit the j release of fission products as assumed in the USAR safety - analyses (Ref.1). The Exhaust Radiation-High signals are initiated from radiation detectors that are located on the ventilation exhaust piping coming from the fuel building ventilation. The signal from each detector is input to an individual , monitor whose trip outputs are assigned to an isolation j channel. i The Allowable Values are chosen to promptly detect gross failure of the fuel cladding. , The Exhaust Radiation-High Function is required to be OPERABLE during movement of recentif2 irradiated fuel assemblies in the fuel buildincj becalise the capability of detecting radiation releases due to fuel failures (due to fuel uncovery cr dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded. Duett 6 radi asst'issfdeEaihthi RFsnEti6rutii bnlyfeqsfredituii sol ate thetfGelibuildingEdurilngffuellhandlingiaccidentsMnfthejfuel buildingiinvolvingshandlingirecentif? irradiated l fuel 41eh fuelf thatDhasioccupiedipartioQaleriticaliresstodcofe WithinithelpreviousL11Tdays)1 , 5. Manual Initiation ' The Manual Initiation push button channels introduce signals (continued) RIVER BEND B 3.3-174 Revision No. O LATER

t i Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES i APPLICABLE into the secondary containment isolation logic that -i SAFETY ANALTSES, are redundant to the automatic protective instrumentation i LCO, and channels, and provide manual isolation- capability. There is APPLICABILITY no specific USAR safety analysis that takes credit for this (continued) Function. It is retained for the secondary containment ; isolation instrumentation as required by the NRC approved , licensing basis. ! There are four push buttons for the. logic, two manual - initiation push buttom per trip system. There is no j Allowable Value for this Function since the channels are i mechanically actuated based solely on the position of the l push buttons. Four channels of the Manual Initiation Function are available and are required to be OPERABLE in MODES 1, 2, and 3 and during movement of iFscihtlfl irradiated fuel asseinblies in the fuel buildihh, ~since these are the MODES and other specified conditions in which the Secondary Containment Isolation automatic Functions are required to be OPERABLE. Moving recentljiirradiated fuel assemblies in the fuel building (the only portion of secondary containment in which fuel can be handled) requires only that portion of the Manual Initiation Function associated with the fuel building to be OPERABLE. ACTIONS A Note has been provided to modify the ACTIONS related to I secondary containment isolation instrumentation channels. , Section 1.3, Completion Times, specifies that once a ' Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits l will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the 1 Condition. However, the Required Actions for inoperable ! secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel. (continued) RIVER-BEND B 3.3-175 Revision No. O LATER

CRFA System Instrumentation i B 3.3.7.1 BASES l APPLICABLE 1. Reactor Vessel Water level-low Low. Level ? SAFETY ANALYSES, LCO, and Low reactor pressure vessel (RPV) water level indicates that APPLICABILITY the capability to cool the fuel may be threatened. A low (continued) reactor vessel water level could indicate a LOCA, .and will automatically initiate the CRFA System, since this could be ' a precursor to a potential radiation release and subsequent radiation exposure to control room personnel. ; Reactor Vessel Water Level-Low Low, Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low, Level 2 ' Function are available (two channels per trip system) and are required to be OPERABLE to ensure that no single instrument failure can preclude CRFA System initiation. The Allowable Value for < the Reactor Vessel Water Level-Low Low, level 2 is chosen to be the same as the Secondary Containment Isolation l Reactor Vessel Water Level-Low Low, Level 2 Allowable Value ' (LC0 3.3.6.2). , i The Reactor Vessel Water Level-Low Low, Level 2 Function is required to be OPERABLE in MODES 1, 2, and 3 to ensure that the control room personnel are protected. In MODES 4 and 5, the probability of a vessel draindown event or of a LOCA, is , minimal. Therefore this Function is not required. In addition, the Control Room Ventilation Radiation Monitor Function provides adequate protection. ,

2. Drywell Pressure-Hiah 1

High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB). A high drywell i pressure signal could indicate a LOCA and will automatically ' initiate the CRFA System, since this could be a precursor to a potential radiation release and subsequent radiation i exposure to control room personnel. Drywell Pressure-High signals are initiated from four i pressure transmitters that sense drywell pressure. Four ! channels of Drywell Pressure-High Function are available l (two channels per trip system) and are required to be OPERABLE to ensure that no single instrument failure can preclude CRFA System initiation. (continued) RIVER BEND B 3.3-200 Revision No. O LATER

CRFA System Instrum:ntation l B 3.3.7.1 1 BASES APPLICABLE 2. Drywell Pressure-Hioh (continued) SAFETY ANALYSES, _ i LCO, and The Drywell Pressure-High Allowable Value was chosen to be

~ APPLICABILITY the same as the Secondary Containment Isolation Drywell Pressure-High Allowable Value (LCO 3.3.6.2). The Drywell Pressure-High Function is required to be t OPERABLE in MODES 1, 2, and 3 to ensure that control room personnel are protected during a LOCA. In MODES 4 and 5, ' the Drywell Pressure-High Function is not required since there is insufficient energy in the reactor to pressurize the drywell to the Drywell Pressure-High setpoint.

3. Control Room Ventilation Radiation Monitors :

The Control Room Ventilation Radiation Monitors measure radiation levels exterior'to the inlet ducting of the MCR. A high radiation level may pose a threat to MCR personnel; thus, a detector indicating this condition automatically signals initiation of the CRFA System. The Control Room Ventilation Radiation Monitors Function consists of two iniependent monitors. Two channels of l Control Room Ventiiation Radiation Monitors are available 1 and are required to be OPERABLE to ensure that no single I instrument failure can preclude CRFA System initiation. The l Allowable Value was selected to ensure protection of the control room personnel. The Control Room Ventilation Radiation Monitors Function is required to be OPERABLE in MODES 1, 2, and 3, and during CORE ALTERATIONS, operations with a potential for draining the reactor vessel (0PDRVs)r and movement of yess@y irradiated fuel in the secondary containment to ensure that control room personnel are protected during a LOCA, fuel handling event, or a vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g., CORE ALTERATIONS OPORVs), the probability of a LOCA orfuel'damageislow;thus[theFunctionisnotrequired. Al sEdue'i tbirad i oast issidecaWithi sIFssdtisii s ; 5hly requ'ireditolinitiateitheCRFESystemiduringMfuelihandling accidenthi svol ving s handl indecentlyl;i eridi stedl fueli (i[sh~ ~ ! fue16 th at% has s obcupi ed i part rofM a Nei t ical treactoK core

                                                                     ~" ~~^~~~

J MthinMhelprevious[11[d#sif "~ 1 (continued) j l RIVER BEND B 3.3-201 Revision No. O LATER

t Prirary Containment Air Locks B 3.6.1.2 BASES BACKGROUND DBA. Not maintaining air lock integrity or leak tightness (continued) may result in a leakage rate in excess of that assumed in the unit safety analysis.

APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary ,

containment is OPERABLE, such that release of fission " products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (La) of 0.26% by weight of the containment and drywell air per 24 hours at the calculated maximum peak containment pressure (Pa) Of 7.6 psig. This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air locks. ! i Primary containment air lock OPERABILITY is also required to l minimize the amount of fission product gases that may escape ' primary containment through the air lock and contaminate and pressurize the secondary containment. During plant operations in other than MODES 1, 2, and 3, the primary containment contains the fission products from a fuel handling accident (FHA ,Eis96Wi5ihahdll iFeEehilj Wradfitedifue14Ms?Mfuelf)thatshnlocsupie oft reviouss117 days criticaldreactoNeoreNithinitheip'tB~TiinTf ths7fGify"c6htiinssit'(REf!4Tl d5ses af the site boundary to within limits. The primary. containment air lock OPERABILITY assures a leak tight fission product barrier during activities with the unit shutdown. Primary containment air locks satisfy Criterion 3 of the NRC Policy Statement. LC0 As part of the primary containment, the air lock's safety function is related to control of containment leakage rates following.aDBA,anFHAiWY6Wini?hi6dl'ihjiFisint19 IFFidtitedlifueR(iisMfseRtha.tghisyoscupisdjast(iffs critic.al.ireactoracoreiithinithe?previ60sn11sdays)ior other g sd rentiVitj"dFWafeFliViT"bEUFiion." Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such events. (continued) RIVER BEND B 3.6-6 Revision No. O LATER

Prirary Centainment Air Locks : B 3.6.1.2 ! i BASES LC0 The primary containment air locks are required to be : (continued) OPERABLE. For each air lock to be considered OPERABLE, the .

 ,                    air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test,and both air lock doors must be OPERABLE. The                      i interlock allows only one air lock door to be open at a                 )

time. This provision ensures that a gross breach of primary ) containment does not exist when primary containment is a required to be OPERABLE. Closure of a single door in each air lock is sufficient to provide a leak tight barrier l following postulated events. Nevertheless, both doors are i kept closed when the air lock is not being used for normal i entry into and exit from primary containment. l APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of i these MODES. Therefore, maintaining OPERABLE primary containment air locks in MODE 4 or 5 to ensure a control volume is only required during situations for which significant releases of radioactive material can be postulated; such as during operations with a potential for , draining the reactor vessel (0PDRVs), &&g C0a.E i

                      ^LTEPf.TIONS, or during fuel movement of Fidestlyjirradiated fuel assemblies in't radioactive; decay"$Whe_               primary _containmeht".W0seltu imaryn contiiineentyiMSciun      areson         !

requipididertndifuelihandlingjinEthefprimaryscostainment ) irsvolviniihand' ingWdentlybirrad1lats6fueM(iMdfueQlst ] h'a slioccupied Preg ouji1Mda s9 gp{artinf

                                               ~ ~ ~ ra? cri tical ; reactors coreiwithinithe l
                                                         ~ - ~ ~ ~ ~ ~ ^ ~

ACTIONS The ACTIONS are modified by Note 1, which allows entry and i exit to perform repairs of the affected air lock component. l If the outer door is inoperable, then it may be easily accessed for most repairs. It is preferred that the air lock be accessed from inside primary containment by entering through the other OPERABLE air lock. However, if this is not practicable, or if repairs on either. door must be , performed from the barrel side of the door, then it is permissible to enter the air lock through the OPERABLE door, which means there is a short time during which the primary containment boundary is not intact (during access through the OPERABLE door). The ability to open the OPERABLE door, (continued) i RIVER BEND B 3.6-7 Revision No. O i.ATER !

I i Primary Containment Air Locks l B 3.6.1.2 ' BASES ACTIONS even if it means the primary containment boundary is (continued) temporarily not intact, is acceptable due to the low , probability of an event that could pressurize the primary ; containment during the short time in which the OPERABLE door l is expected to be open. After each entry and exit, the OPERABLE door must be immediately closed. Note 2 has been included to provide clarification that, for this LCO, separate Condition entry is allowed for each air 1 lock. ' l This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable air lock. Complying with the Required Actions may allow for continued operation, and a subsequent inoperable air lock is governed by subsequent-Condition entry and application of associated Required Actions. The ACTIONS are modified by a third Note, which ensures appropriate remedial actions are taken when necessary. Pursuant to LC0 3.0.6, ACTIONS are not required even if ' primary containment is exceeding its leakage limit. Therefore, the Note is added to require ACTIONS for ; LCO 3.6.1.1, " Primary Containment-Operating," to be taken in ' this event. The leakage limits of LC0 3.6.1.1 are only applicable in MODES 1, 2, and 3, therefore, the provisions of Note 3 apply only during MODES 1, 2, and 3. A.I. A.2, and A.3 With one primary containment air lock door inoperable in one ! or more primary containment air locks, the OPERABLE door must be verified closed (Required Action A.1) in each affected air lock. This ensures that a leak tight primary containment barrier is maintained by the use of an OPERABLE air lock door. This action must be completad within 1 hour. The l' hour Completion Time is consistent with the AC'i10NS of LC0 3.6.1.1, which requires that primary containment be restored to OPERABLE status within I hour. In addition, the affected air lock penetration must be isolated by locking closed the OPERABLE air lock door within the 24 hour Completion Time. The 24 hour Completion Time is considered reasonable for locking the OPERABLE air lock door, considering the OPERABLE door of the affected air lock is being maintained closed. (continued) RIVER BEND B 3.6-8 RevisionNo.OLATE3 ,

L N 1 ._' ~ Pri:::ary Containment Air Locks B 3.6.1.2 i I BASES i JACTIONS- A.I.'A.2. and A.3 -(continued)- ! Required Action: A'.3 ensures that-the affected air lock with , an-inoperable door has been isolated by the use of.a locked' t closed OPERABLE air lock door. This ensures that an . 1 4 s acceptable primary containment leakage boundary is ! maintained. .The; Completion Time of once per 31 days.is based on engineering judgment and is considered ' adequate in i

                                           . view of the low likelihood.of.a locked door being                                                             ?

mispositioned and other administrative controls. l Required Action A.3 is modified by a Note that applies'to j air lock doors located in high radiation areas and allows ! these doors to be verified locked closed by.use of : administrative controls._ Allowing verification by . i administrative controls is considered acceptable, since ; access to these areas is' typically restricted. ;Therefore, ! the probability of misalignment of the. door, once it has , been verified to be in the. proper position, . i.s .small. 1

                                           -The Required Actions have been modified by two Notes....

Note 1 ensures that only-the Required Actions and associated ~ j : Completion Times of Condition C are required if both_ doors- 1 in the air lock are. inoperable. With both doors in-the air- i lock inoperable, an OPERABLE door is not available to be ; closed. Required Actions C.1 and C.2 are the appropriate -l remedial actions. The exception of Note 1 does not affect ! tracking .the Completion Times from the initial entry into '! Condition A;.only the requirement toLcomply,with the l Required Actions. Note 2 allows use of the air lock for- , entry and exit for 7 days:under administrative controls if- l both have an inoperable. door. This-7. day restriction.begins ! when.the second airlock is discovered inoperable.

Primary containment entry mayl.be required to perform _ q Technical Specifications (TS) Surve111ances and Required . Actions, as well as other activities inside primaryl - j containment that are required _by TS or. activities that-. ~! support TS-required: equipment. This Note is not intended to 1 preclude performing other activities (i;e., non-TS-related i activities) if the primary containment was entered, using 1 7 the inoperable air lock,' to perform an allowed activity. ; listed above. 1The administrative controls required consist ! of the stationing of a dedicated individual to assure. ; closure of the'0PERABLE door except during the entry and j exit, and assuring-the OPERABLE door is relocked after -e ( fcontinued) , RIVER BEND B 3.6-9 RevisionNo.-0LATER -l r

       .        -       4    -          --   ,,+-...-.__c-.,---.,-.c.-             ...L...~.,        ..~ - - . - . . , , , - -                   .

3.- m - - . . - .

                                                                                                                                                               .-- - ~

Primary Centainment Air Locks

l B 3.6.1.2 , i BASES 1 > ' ACTIONS 'A.I. A.2.'and A.3 (continued) j i m completion of the containment l. entry and exit. This. ' allowance-is acceptable due to the 1ow probability of an . event that could pressurize the primary containment during l the short time that the OPERABLE door is' expected to be' open. j l B.1. B.2. and B.3

                                                                                                       .                                                                    I With an air lock interlock mechanism inoperable in one or.                                                                    I both' primary containment air locks, the Required. Actions and .                                                              !

associated Completion Times are consistent with those 1 specified'in Condition A. - 1 The Required Actions have'been modified by two Notes. Note 1 ensures that only the Required Actions and associated Completion Times of Condition C are required if.both doors ! in one air lock are inoperable. With'both' doors in the' air- l lock inoperable, an OPERABLE door is not available to be closed. Required Actions.C.1 and C.2 are:the appropriate remedial actions. Note 2 allows entry into and exit from the primary containment under the control of a: dedicated -1 individual stationed.at the air lock to ensure that only one ! door is opened at a time (i.e., the individual performs' the .i function of the interlock). . j Required Action B.3 is modified by a Note.that applies to-air. lock doors located in high radiation areas and allows these doors to be verified locked closed by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas. is typically. restricted. Therefore,- the probability of misalignment o.f the door, once. it has been verified to- be in the proper position, is- small. C.I. C.2. and C.3 With one or more air locks 1 inoperable for reasons other than-those described in Condition A or B, Required Action C.I. requires action to be .immediately initiated to evaluate. containment overall leakage rates using current air lock leakage test results. An evaluation is acceptable since it ; is overly conservative to immediately declare the primary . containment inoperable if both doors in an air loch have (continued) RIVER BEND- B 3.6-10 Revision No. O QTER _ __... _ . . . _ _ . .-. _ _ . . . _ , . _ , . . . . . . . - _ . . . . ~ . . , , ,- ,. ,_, ,.

4 Pricary Centainment Air Lccks-B 3.6.1.2

-[

BASES 4 ACTIONS C.I. C.2 and C.3 (continued) ;

 .              failed a seal test or if the overall air lock leakage is not          ;

within limits. In many instances (e.g., only one seal per door has failed) primary containment remains OPERABLE, yet only 1. hour (according to LCO 3.6.1.1) would be provided to restore the air lock door to OPERABLE status prior to requiring a plant shutdown. In addition, even with both doors failing the seal test, .the overall containment leakage rate can still be within limits. Required Action C.2 requires that one door in the affected primary containment air-locks must be verified closed. This Required Action must be completed within the 1 hour Completion Time. This specified time period is consistent with the ACTIONS of LC0 3.6.1.1, which require that primary containment be restored to OPERABLE status within 1 hour. Additionally, the air lock must be restored to OPERABLE status within 24 hours. The 24 hour Completion Time is reasonable for restoring an inoperable air lock to OPERABLE status considering that at least one door is maintained closed in each affected air lock. D.1 and D.2 If the inoperable primary containment air lock cannot be restored to OPERABLE status within the associated Completion Time while operating in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed. Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. E.1 and E.2 If the inoperable primary containment airlock cannot be restored to OPERABLE status within the associated Completion Time during operations with a potential for draining the reactor vessel TERATIONS, or during movement of fec_.(0PDRVs)r-daring CORE A'y in the entlj irradiated fuel assemblies primarycontiinmiht,]actionisrequiredtoimmediately suspend activities that represent a potential for releasing (continued) RIVER BEND B 3.6-11 Revision No. O LATER

a, ,

              '."  m Prfnary Centainment' Air Locks

B 3.6.1.2

                  ; BASES ACTIONS.      E.1 and E.2 L(continued) radioactive material, thus placing the unit in a Condition that minimizes risk. If applicab1:, C"E ALTEST!"S =d '

r movement of resih$jirradiated fue1' assemblies in'the primary contiinment must be immediately suspended. Suspension of these activities shall not preclude completion 4 of movement of a component to a safe ~ position. . Also, if.. applicable, action must be.immediately initiated to suspend. OPDRVs to minimize the probability of a vessel draindown'and' subsequent potential. for fission product release. ' Action. L must continue until OPDRVs'are suspended. SURVEILLANCE SR 3.6.1.2.1 REQUIREMENTS Maintaining primary containment air locks O' PERABLE requires compliance with the. leakage rate test requirements of 10 CFR 50, Appendix J (Ref. 2), as modified by approved exemptions when in MODES 1, 2, .and 3. -This SR reflects l the leakage rate testing requirements.with regard to air lock p leakage (Type B leakage tests). The acceptance criteria. (i.e., :s13,500 cc/hr for the combination of all annulus bypass _ leakage-paths that are required to be' meeting leak' tightness) ensures that the combined leakage rate of annulus bypass leakage paths is less than the specified leakage. rate. This provides assurance in MODES 1, 2, and 3 that the assumptions in the radiological evaluations are met. The leakage rate of each bypass leakage path is assumed to be the maximum pathway leakage (e.g., leakage through the air. lock: door with the-highest leakage) unless'the penetration is isolated by use of (for this- Specification) one closed ; and locked air' lock door. The leakage rate of the isolated . ; bypass-leakage path is assumed to be.the' actual' pathway. l leakage through the isolation devices' (e.g., air lock door). If both air lock doors-are closed, the actual leakage rate-is the lesser leakage rate of the two barriers (doors). This method of quantifying maximum pathway leakage-is.only- . to be'used for this SR-(i.e., Appendix J maximum pathway -l 1eakage limits used to evaluate Type A, B and C limits are to be quantified in'accordance with Appendix J). During irradiatedthefuel operational assembliesconditions of moving in the' primary M conti W'inment, C"I ALTEMTIC*S, or OPDRVS, the only annulus bypass path leakage required to be met:is through the two primary containment-airlocks; therefore the entire 13,500 cc/hr: limit can be (continued) RIVER BEND B 3.6-12 Revision No. O LATER

                                                                . Primary Containment-Air Locks B 3.6.1.2             l BASES I
                -SURVEILLANCE  SR   3.6.1.2.1   (continued)                                                 l REQUIREMENTS                                                                               ;

applied to the air locks. In these operational conditions the reactor coolant system is not pressurized and specific j primary containment leakage limits are not imposed. I However, due to the size of the air lock penetration, leakage limits are imposed to assure an OPERABLE barrier. In these conditions the leakage limits are not related to radiological evaluations, but only reflect engineering , judgment of an acceptable barrier. The periodic testing J requirements verify that the air lock leakage does not i exceed the allowed fraction of the overall primary -l containment leakage rate. The Frequency is required by 10 CFR 50, Appendix J, as modified by approved exemptions. . Thus, SR 3.0.2 (which allows Frequency extensions) does not ' apply. The SR has been modified by two Notes. Note:1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the~ event of a DBA. Note 2 has been added to this SR, requiring the results to be evaluated against the acceptance criteria of , SR 3.6.1.1.1 during operation in MODES 1, 2, and 3. This l ensures that air lock leakage is properly accounted for ir. ! determining the overall primary containment leakage rate. ! Since the overall primary containment leakage rate is only applicable in MODE 1, 2, and 3 operation, the Note 2 requirement is imposed only during these MODES. I SR 3.6.1.2.2 The seal air flask pressure is verified to be at a: 90 psig every 7 days to ensure that the seal system remains viable. , It must be checked because it could bleed down during or - following access through the air lock, which occurs i regularly. The 7 day _ Frequency has been shown to be

i acceptable through operating experience and is considered '

adequate in view of the other indications available.to , operations personnel that the seal air flask pressure is ! low. 1 (continued) : RIVER BEND B 3.6-13 Revision No. G LATER

Primary Containment Air-Locks B 3.6.1.2 , BASES SURVEILLANCE SR 3.6.1.2.3 REQUIREMENTS ' (continued) The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary containment pressure (Ref. 3), closure of either door will support primary containment OPERABILITY. Thus, the interlock' feature supports primary containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur. Due to the nature of this interlock, and given that the interlock mechanism is only challenged when the primary containment airlock door is opened, this test is only required to be performed upon entering or exiting a primary containment air lock, but is not required more frequently than once per 184 days. The 184 day Frequency is based on engineering judgment and is considered adequate in view of other administrative controls. SR 3.6.1.2.4 A seal pneumatic system test to ensure that pressure does not decay at a rate equivalent to > 1.28 psig for a period of 24 hours from an initial pressure of 90 psig is an effective leakage rate test to verify system performance. The 18 month Frequency is based on the fact that operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. USAR, Section 3.8.

2. 10 CFR 50, Appendix J.

3. USAR, Table 6.2-1.

4. USAR, 15.7.4.

1 RIVER BEND B 3.6-14 Revision No. O LATER

i l Primary Containment-Shutdown l B 3.6.1.10 l i B 3.6 CONTAINMENT SYSTEMS ' B 3.6.1.10 Primary Containment-Shutdown i BASES I BACKGROUND The function of the primary containment is to isolate and contain fission products released from the Reactor Primary System following a Design Basis Accident (DBA) and to confine the postulated release of radioactive material to within limits. The primary containment consists of a steel lined, reinforced concrete vessel, which surrounds the Reactor Primary System and provides an essentially leak l tight barrier against an uncontrolled release of radioactive material to the environment. Additionally, this. structure provides shielding from the fission products that may be present in the primary containment atmosphere following accident conditions. ; The isolation devices for the penetrations in the primary containment boundary are a part of the primary containment leak tight barrier. To maintain this leak tight barrier for~ accidents during shutdown conditions:

a. All penetrations required to be closed during accident conditions are closed by manual valves, blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in LC0 3.6.1.3,

                          " Primary Containment Isolation Valves (PCIVs)";

b. Primary containment air locks are OPERABLE, except as provided in LC0 3.6.1.2, " Primary Containment Air Locks"; and

c. All equipment hatches are closed.

This Specification ensures that the performance of the primary accident containment, in the event of a fuel handling ? fuel (FHA)EinvolyitigThandl.isgWedentljIfpradlAted ( C eM fuelithati hasi occUpi ed / parthf) at criti ealifreactor: core yithin @ efprevious R days), Tnidieftint~cFiticili~ty, or reactor vessel draindown, provides an acceptable leakage barrier to contain fission products, thereby minimizing offsite doses. (continued) RIVER BEND B 3.6-50 Revision No. O LATER

Primary ' Containment-Shutdown l B 3.6.1.10 { i BASES (continued) APPLICABLE- The safety design basis for the primary containment is that l SAFETY ANALYSES it contain the fission p'roducts from a FHA?involviny hindli6gfecentlyhirpadiatedlfueli(iieMfuelithat3has,ious occupiedtpart;ofsa7criticalsreactorscore(withinithe? prev . . . . 111 days)? iWside" thiiriniiry~bontsinient~(Rsf.'2);^Tto^1imit~ ' " doses it the site boundary to within limits. The primary i containment performs no active function in response to this ) event; however, its leak tightness is required to ensure j that the release of radioactive materials from the primary ) containment is restricted to those. leakage rates assumed in I safety analyses. ) The FHA inside the primary containment is assumed to occur j only after 2: 80 hours since the reactor was last critical. ' The fission product release is, in turn, based on an assumed leakage rate from vent and drain valves with a ccabined flow rate of 70.2 cfm (based on an assumed 0.367 inch water gauge differential pressure). This assumed pressure reflects the fact that the FHA does not produce elevated containment pressures as is the case for the DBA LOCA. However, as an added conservatism, the analysis assumes a non-mechanistic additional leakage of 0.26% of the containment volume per

day. Primary containment satisfies Criterion 3 of the NRC Policy Statement. LC0 Primary containment OPERABILITY is maintained by providing a contained volume to limit fission product escape following a FHAif nVoliisg5 hshdli iIi M sifuel thatihasToccupied hrtsofsiWecintlyff hidistidWssM(l M sritisii nsecten sope thintths previ 6u sil lfdays[~ @o r~~ot hn^iih^in fisiliifed ~fiiEtiVi ty~'o level'sE0r5' ion. Compliance with this LC0 will ensure a , primary containment configuration, including equipment- : hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis. Since no credit is assumed for automatic isolation valve closure, and any leakage which would occur prior to valve closure is similarly not accounted for, all penetrations which could communicate gaseous fission products to the environment must remain closed. However, a limited number of primary containment penetration I vent and drain valves may remain opened, and the primary (continued) RIVER BEND B 3.6-51 Revision No. O LATER

Primary Containment-Shutdown B 3.6.1.10 BASES - LCO containment considered OPERABLE provided the calculated (contunded) leakage flow rate through the open vent and drain valves is less s 70.2 cfm. Leakage rates specified for the primary containment and air locks, addressed in LCO 3.6.1.1 and LCO 3.6.1.2 are not directly applicable auM ng the shutdown conditions addressed in this LCO. APPLICABILITY In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining an OPERABLE primary containment in MODE 4 or 5 to ensure a control volume, is only required during situations for which significant releases of radioactive material can be postulated; such as during operations with a potential for draining the reactor vessel (0PDRVs l f,LTERf,TIONS, or during movement of r), durir.g COREeceist191 irra , assemblies in the primary containment': 'Dueito"radiosctivs decayJprimary/containmentsisYonlydeq6] ped.Lduhingl fue1~ ^ h andl i ng i i n LtheD primary; con t ai nmenti i nvol.vi ng1handl i ng ~ recentlyJirradiated fuel (ile2,ifuelsthatihas? occupied:part offaicriticalfreactoricore withinithe[p.reviousil17 days). ' l Requirements for ECCS OPERABILITY during MODES 1, 2, and 3 i are discussed in the Applicability section of the Bases for l LC0 3.5.1. ! l ACTIONS A.1 and A.2 In the event that primary containment is inoperable, action is required to immediately suspend activities that represent a potential for releasing radioactive material, thus placing the unit in a Condition that minimizes risk. If applicable, CORE f,LTERf,TIONS cr.d movement of feceistly@ irradiated fuel assemblies must be immediately suspundedf~ Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until OPDRVs are suspended. (continued) RIVER BEND B 3.6-52 Revision No. O LATER

F Primary Containment-Shutdown B 3.6.1.10 : BASES (continued)

   ' SURVEILLANCE      SR 3.6.1.10.1 REQUIREMENTS                   .

This SR verifies that each primary containment penetration ; that could communicate gaseous fission products to the environment during accident conditions is closed. ~The SR helps to ensure that post accident leakage of radioactive gases outside of the primary containment boundary is within-design limits. The method of isolation must include the use - of at least one-isolation barrier that cannot be adversely affected by a single active _ failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, or.a blind flange. This SR does not require any testing or valve manipulation. Rather, it. involves verification, through a system walkdown, that the required valves are in.the correct position. The 31 day Frequency was chosen to provide added assurance that the valves remain in the correct positions. i The SR is modified by a Note stating that the SR is not . required to be met for vent and drain line pathways provided i the total calculated flow rate through open. vent and drain pathways is s 70.2 cfm. Administrative controls ensure , ,

that open vent and drain pathways will: (1) only be opened < to support leakage rate testing; (2) not exceed 12 valves; ' (3) require monitoring opened vent and drain valves, as well as the containment-to-auxiliary building differential pressure every 2 hours; and (4) assure at least one person l is assigned to each open penetration (Ref. 1). ; REFERENCES 1. NRC SER for TS Amendment #35, dated March 3, 1989.

2. USAR, Section 15.7.6.

l i i RIVER BEND B 3.6-53 Revision No. O LATER

Secondary Containment-Operating B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.'..i Secondary Containment-0perating BASES BACKGROUND The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA). In conjunction with operation of the Standby Gas Treatment (SGT) System, Fuel Building Ventilation System, and closure of certain vaives whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment. The secondary containment consists of the shield building, auxiliary building, and fuel building, and completely encloses the primary containment and those components that may be postulated to contain primary system fluid. This structure forms a control volume that serves to hold up and dilute the fission products. It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g., due to pump / motor heat load additions). To prevent ground level exfiltration while allowing the secondary containment to be designed as a , conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LC0 3.6.4.2, " Secondary Containment Isolation Dampers (SCIDs)," LC0 3.6.4.3,

                     " Standby Gas Treatment (SGT) System," LC0 3.6.4.4, " Shield Building Annulus Mixing System," and LC0 3.6.4.5, " Fuel Building Ventilation System."

The isolation devices for the penetrations in the secondary containment boundary are a part of the secondary containment barrier. To maintain this barrier:

a. All Auxiliary Building penetrations, Fuel Building penetrations and Shield Building annulus penetrations required to be closed during accident conditions are either:

(continued) RIVER EEND B 3.6-83 Revision No. O

Secondary Containment-Operating B 3.6.4.1 BASES BACKGROUND 1. Capable of being closed by an OPERABLE secondary (continued) containment automatic isolation signal, or

2. Closed by at least one manual valve, blind fit.nge, or deactivated automatic valve or damper, as applicable, secured in its closed position, except as provided in LC0 3.6.4.2;

b. All Auxiliary Building, Feel Building and Shield Building Annulus equipment hatches are closed and sealed;-

c. The Standby Gas Treatment System is OPERABLE, except as provided in LC0 3.6.4.3;

d. The Fuel Building Charcoal Filtration System is OPERABLE, except as provided in LCO 3.6.4.6; and

e. At least one door in each access to the Auxiliary Building, Fuel Building and Shield Building Annulus is closed, except for routine entry and exit of personnel and equipment.

APPLICABLE There are two principal accidents for which credit is SAFETY ANALYSES taken for secondary containment OPERABILITY. These are a LOCA(Ref.1)andafuelhandlingaccidentninVolvind handlinpecentiffirradisted;fuslT(i!sQfuel::lthathhas_. occupied)partiof?aicriticalEreactorJcore within5thelprevious containmlentperformsnoll1 days) in"the"fu61 active-function in response~6uildihi to each (Ref72) . of these limiting events; however, its leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage 3 paths and associated leakage rates assumed in the accident i analysis, and that fission products entrapped within the i secondary containment structures will be treated by the SGT System or Fuel Building Ventilation System prior to discharge to the environment. Secondary containment-operating satisfies Criterion 3 of the NRC Policy Statement. LC0 An OPERABLE secondary containment provides a centrol volume : into which fission products that bypass or leak from primary l (continued) RIVER BEND B 3.6-84 Revision No. O LATER l

Secondary Containment-Operating B 3.6.4.1 BASES LC0 containment, or are released from the reactor coolant (continued) pressure boundary components located in the shield ~' building, auxiliary building, or fuel building, can be diluted and processed prior to release to the environment. For the secondary containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained. APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY. In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary ccntainment OPERABLE 's not required in MODE 4 , or 5 to ensure a control volume, except for other situations l for which significant releases of radioactive material can ' be postulated, such as during movement of ficshtli i irradiated.fue1[(ifeh4fdelithit@df6cc01ed/partof?a ) critijal@eactorfcoreiwithiajhejprevjous)q1sday's) ! assedlics in ths f6el buiTding. Th6 fiiel bQilding i OPERABILITY during ressntlysirradiated fuel handling is addressed in LC0 3.6:4~ 6T" Fuel Building Ventilation Systems-Operating. " l J ACTIONS A.1 l If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours. The 4 hour Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal. B.1 and B.2 If the secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must (continued) RIVER BEND B 3.6-85 Revision No. O LATER

E' s m' L Secondary Containment 4ptrating B 3.6.4.1 BASES' ACTIONS. B.1 and 'B.2 (continued)

    '.               be-brought to~a MODE in which the LC0 does not apply. To            !

achieve this status,'the plant must be brought to at least . MODE 3 within 12 hours and to MODE 4 within 36 hours. The' ! allowed Completion Times are reasonable, based on operating , experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.4.1.1 .; REQUIREMENTS This SR' ensures that the shield building annulus, auxiliary building, and. fuel building boundary is sufficiently leak tight to preclude exfiltration under expected wind ' conditions. The 24. hour Frequency of this SR was developed- , based on operating experience related to sace.ndary - containment vacuum variations during the applicable MODES - and the low probability of a DBA occurring betwe1n surveillances. Furthermore, the' 24 hour Frequency is considered adequate in~ $ view of other indications available in the control room,- ' including alarms, to alert the operator to an abnormal ; secondary containment vacuum condition. SR 3.6.4.1.2 and SR 3.6.4.1.3 ' Verifying that secondary containment equipment hatches _and ; access doors are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying ; that' all such openings are closed provides adequate assurance that exfiltration from the secondary containment' i will not occur. In this application the term " sealed" has. ' no connotation of leak tightness. Maintaining secondary containment OPERABILITY requires verifying each door.in the .; access opening is closed, except when the access opening is being used for entry and exit. ' The 31 day Frequency for these SRs has been shown to be adequate based on operating experience, and is consioered adequate in view of the other ; controls on secondary cont.ainment access openings. ; (continued) 1

RIVER BEND B 3.6-86 Revision No. O LATER

e .

         y            w
                                                                          !S:condary Centainment-Oparating ;          i B 3.6.4.1 y
                    ' BASES                                                                                           '
                                                  -                                                                  1 SURVEILLANCE       SR   3.6.4.1.4 and'SR     3.6.4.1.6 REQUIREMENTS                                     -

j

         '..            (continued)      The SGT System exhausts the shield building annulustand auxiliary building atmosphere to'the environment through appropriate treatment' equipment. To ensure that all fission                 i products are treated, SR 3.6.4.1.4 verifies that the SGT System will -rapidly establish and maintain a pressure in the-shield building annulus and auxiliary building that is less than the lowest postulated pressure external to the secondary containment ~ boundary. This is confirmed by demonstrating that one SGT subsystem will draw down the shield building annulus and auxiliary building to a 0.5 and a 0.25 inches of vacuum water gauge in s 18.5_ and               -

s.13.5 seconds, respectively. This cannot be accomplished ; if the secondary containment boundary.is not intact. SR 3.6.4.1.6 demonstrates that each SGT subsystem can-maintain = 0.5 and' a 0.25 inches of vacuum water gauge for. I hour. The 1 hour test period allows shield building annulus and auxiliary building to be in thermal equilibrium at steady state conditions. Therefore,_these_two tests are used to ensure the integrity of this portion.of_the_

                                       -secondary containment boundary. . Since these_SRs are secondary containment tests, they need not'be performed with each SGT subsystem.       The SGT subsystems are tested on a STAGGERED TEST BASIS, however, to ensure'that in addition to the requirements of f.C0 3.6.4.3, either SGT subsystem will-               1 perform this test,       ]perating experience has shown these-components usually pass the Surveillance when performed at                 -i the 18 month Frequency.       Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

l 3 SR 3.6.4.1.5 and SR 3.6.4.1.7 l The Fuel Building Ventilation System exhausts the fuel j building atmosphere to the environment through appropriate ! treatment equipment. To ensure that all fission products are treated, SR 3.6.4.1.5. verifies that the Fuel Building - Ventilation. System will rapidly establish and maintain a pressure in the fuel building that is less than the lowest postulated pressure external to the secondary containment boundary. This.is confirmed by demonstrating that one fuel i building ventilation subsystem will. draw down' the fuel ! building to a 0.25 inches of-vacuum water gauge in s 12.5 seconds. This cannot-be accomplished if the secondary containment boundary is not intact. SR 3.6.4.1.7 ; demonstrates that each SGT subsystem can maintain (continued) RIVER BEND B 3.6-87 Revision No. G LATER-

S2cendary Containment-Operating s B 3.6.4.1 BASES l l SURVEILLANCE SR 3.6.4.1.5 and SR 3.6.4.1.7 (continued) REQUIREMENTS

a: 0.25 inches of vacuum water gauge for 1 hour. The 1 hour test period allows the fuel building to be in thermal equilibrium at steady state conditions. Therefore, these two tests are used to ensure the integrity of this portion of the secondary containment boundary. Since these SRs are secondary containment tests, they need not be performed with each SGT subsystem. The SGT subsystems are tested on a STAGGERED TEST BASIS, however, to ensure that in addition to the requirements of LC0 3.6.4.3, either SGT subsystem will perform this test. The 18 month Frequency is based on the need to perform this Surveillance under.the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at. the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. USAR, Section 15.6.5.

2. USAR, Section 15.7.4.

I i l l RIVER BEND B 3.6-88 Revision No. O LATER

q a SCIDs B 3.6.4.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.2 Secondary Containment Isolation Dampers (SCIDs) l BASES BACKGROUND The function of the SCIDs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Ref. 1). Secondary containment isolation within the time limits specified for those isolation dampers l designed to close automatically ensures that fission products that leak from primary containment following a DBA, , that are released during certain operations when primary ) containment is not required to be OPERABLE, or that take place outside primary containment, are maintained within the i secondary containment boundary. ; 1 The OPERABILITY requirements for SCIDs help ensure that an l adequate secondary containment boundary is maintained during- j and after an accident by minimizing potential paths to the ' environment. Isolation barrier (s) for the penetration are discussed in Reference 2. The isolation devices addressed , by this LCO are either passive or active (automatic). ! Manual dampers, de-activated automatic dampers secured in their closed position, check dampers with flow through the j damper secured, and blind flanges are considered passive ' devices. Check dampers and other automatic dampers designed to close without operator action following an accident are considered active devices. Automatic SCIDs close on a secondary containment isolation signal to establish a boundary for untreated radioactive , material within secondary containment following a DBA or other accidents. l Other penetrations are isolated by the use of dampers or valves in the closed position or blind flanges. APPLICABLE The SCIDs must be OPERABLE to ensure the secondary SAFETY ANALYSES containment barrier to fission product releases is established. The principal accidents for which the secondary containment boundary is required are a loss of coolant accident Ref. 1) and a fuel handling accident injo has g1viT ngthMd1 occupied ihg(yishtly0 parb oQalcri tidals reactop core WidiijsiMfse1 Previousglidays)'in' thi fEi1 lisilding (Rif'.~i'withintthe 3). The (continued) RIVER BEND B 3.6-89 Revision No. O LATER

SCIDs ; B 3.6.4.2 BASES APPLICABLE secondary containment performs no active function in SAFETY ANALYSES response to each of these limiting events, but the boundary (continued) established by SCIDs is required to ensure that leakage from the primary containment is processed by the Standby Gas ' Treatment (SGT) System and Fuel Building Ventilation System before being released to the environment. Maintaining SCIDs OPERABLE with isolation times within limits ensures that fission products will remain trapped 4 inside secondary containment so that they can be treated by the SGT System or Fuel Building Ventilation System prior to discharge to the environment. SCIDs satisfy Criterion 3 of the NRC Policy Statement. LC0 SCIDs form a part of the secondary containment boundary. The SCID safety function is related to control of offsite radiation releases resulting from DBAs. i The power operated isolation dampers are considered OPERABLE when their isolation times are within limits. Additionally, ; power operated automatic dampers are required to actuate on , an automatic isolation signal. ' The normally closed isolation dampers or blind flanges are considered OPERABLE when manual dampers are closed or open in accordance with appropriate administrative controls, automatic dampers are de-activated and secured in their closed position, or blind flanges are in place. The SCIDs covered by this LCO, along with their associated stroke times, if applicable, are listed in Reference 4. APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore, OPERABILITY of SCIDs is required. In MODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIDs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during movement cf recerit191 irradiated fuel assemblies. Movingrscintli irFadiatid fuel assemblies in the Primary Contaihment is (continued) RIVER BEND B 3.6-90 Revision No. O LATER

SCIDs ! B 3.6.4.2 i

i BASES l APPLICABILITY " Primary I (continued) addressed adequately Containment-Shutdown." in LC0 3.6.I.10,itMie~decsif~,TSCIDs DesiteNedisa i onlifreqdf red lto1 beTOPDIASLE dWagifuels hardling i ind ths~^~ ~j fuel; building inWivingihandlingirecentlyfirradiated?: feel

( i . e m fuel f thatihas ? occupied fpartiofiascri ticalireactorf : core;withinitheprevious111SdaysQ t Moving resentlyiirradiated fuel assemblies in the fuel ; building ~(th'e ohly portion of secondary containment in which fuel can be handled) will require only the SCIDs associated with the fuel building to be OPERABLE. 4

l ACTI0hs The ACTIONS are modified by three Notes. The first Note allows penetration flow paths to be unisolated i intermittently under administrative controls. These ! controls consist of stationing a dedicated operator, who is ! in continuous communication with the control room, at the i controls of the isolation device. In this way, the : penetration can be rapidly isolated when the need for secondary containment isolation is indicated. l The second Note provides clarification that for the purpose : of this LC0 separate Condition entry is allowed for each penetration flow path. This is acceptable, since the ; Required Actions for each Condition provide appropriate ! compensatory actions for each inoperable SCID. Complying I with the Required Actions may allow for continued operation, ! and subsequent inoperable SCIDs are governed by subsequent Condition entry and application of associated Required ; Actions. l The third Note ensures appropriate remedial actions are i taken, if necessary, if the affected system (s) are rendered l inoperable by an inoperable SCID. , 1 A.I and A.2

                                                                                                    /l l In the event that there are one or more penetration flow paths with one SCID inoperable, the affected penetration flow path (s) must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure.

Isolation barriers that meet this criteria are a closed and de-activated automatic damper, a closed manual damper or a blind flange. For penetrations isolated in accordance with Required Action A.1, the device used to isolate the

         ,                                                                            (continued)

RIVER BEND B 3.6-91 Revision No. O LATER

, SCIDs B 3.6.4.2 <

                 .                                                                    h BASES                                                                            ,
    ~ ACTIONS      'A.1 and A.2    (continued)

penetration should be the closest available device to secondary containment. This Required Action must be completed within the 8 hour Completion Time. The specified time period is reasonable considering the time required to c isolate the penetratien and the low probability of a DBA, ,

which requires the SCIDs.to close, occurring during this ; short time. l For affected penetrations that have been isol.ated in ' accordance with Required Action A.1, the affected i penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that secondary , containment penetrations required:to be isolated following ' an accident, but no longer capable of being automatically isolated, will be isolated should an event occur. This : Required Action does not require any testing or isolation device manipulation. Rather, it involves verification that l the affected penetration remains isolated. l Required Action A.2 is modified by a Note that applies to ; isolation devices located in high radiation areas and allows I them to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment once they have been verified to be in the proper position, is low. B.1 With two SCIDs in one or more penetration flow paths inoperable (Condition A is entered if one SCID is inoperable in each of two penetrations), the affected penetration' flow path must be isolated within 4 hours. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic damper, a closed r.anual damper, and a blind flange. The 4 hour Completion Time is reasonable, considering the time required to isolate the penetration and the low probability of a DBA, which requires the SCIDs to close, occurring during this short time.

              ._                                                          (continued)

RIVER BEND B 3.6-92 Revision No. O LATER

SCIDs l B 3.6.4.2 ,

 . BASES
 . ACTIONS       C.1 and C.2 (continued)

If any Required Action and associated Completion Time cannot l be met, the plant must be brought to a MODE in which the LCO- , does not apply. To achieve this status,.the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 l within 36 hours. The allowed Completion Times are t reasonable, based on operating experience, to reach the ' required plant conditions from full power conditions in an i orderly manner and without challenging plant systems. ! D.I. D.2, and 0.3 If any Required Action and associated Completion Time cannot be met, the plant must be placed in a condition in which the LC0 does not apply. When applicable, movement of pecentl irradiated fuel assemblies in the fuel building muit~bs~y . immediately suspended. Suspension of this activity shall not preclude completion of movement of a component to a safe position. Required Action D.1 has been modified by a. Note stating that . LCO 3.0.3 is not applicable. Ifmovingrhentlyjirradiated ' fuel assemblies while in MODE 4 or 5, LC0"3".0 3 would not : specify any action. If moving pecentlyslirradiated fuel assemblies while in MODE 1, 2, &^3',^'"the fuel movement is independent of reactor operations. Therefore, in either 1 case,inabilitytosuspendmovementofFedsstlyj! irradiated fuel assemblies would not be a sufficie6FrEason to require a reactor shutdown. i SURVEILLANCE SR 3.6.4.2.1 REQUIREMENTS This SR verifies each secondary containment isolation manual damper and blind flange that is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the secondary containment boundary is within design , limits. This SR does not require any testing or damper manipulation. Rather, it involves verification that those SCIDs in secondary containment that are capable of being-mispositioned are in the correct position. j (continued) RIVER BEND B 3.6-93 Revision No. O LATER

SCIDs B 3.6.4.2 BASES SURVEILLANCE SR 3.6.4.2.1 (continued) REQUIREMENTS Since these SCIDs are readily accessible to personnel during normal unit operation and verification of their position is relatively easy, the 31 day Frequency was chosen to provide added assurance that the'SCIDs are in the correct positions. Two Notes have been added to this SR. The first Note applies to dampers and blind flanges located in high radiation areas and allows them to be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these SCIDs, once they have been verified to be in the proper position, is low. A second Note has been included to clarify that SCIDs that are open under administrative controls are not required to meet the SR during the time the SCIDs are open. SR 3.6.4.2.2 Verifying the isolation time of each power operated and each automatic SCID is within limits is required to demonstrate OPERABILITY. The isolation time test ensures that the SCIDs will isolate in a time period less than or equal to that assumed in the safety analyses. The Frequency of this SR is 92 days. SR 3.6.4.2.3 Verifying that each automatic SCID closes on a secondary containment isolation signal is required to prevent leakage of radioactive material from secondary containment following a DBA or other accidents. This SR ensures that each automatic SCID will actuate to the isolation position on a 1 secondary containment isolatian signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this SR to provide complete testing of the safety function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. (continued) RIVER BEND B 3.6-94 Revision No. O LATER

                                                                               % .v. I
                                      '                                       SCIDs    ;

B 3.6.4.2 1 ! BASES l l SURVEILLANCE SR 3.6.4.2.3 (continued) REQUIREMENTS

 ,                   Operating experience has shown these components usually pass the Surveillance when performed Lat the 18 month Frequency.

L Therefore, the Frequency was concluded to be acceptable from . l a reliability standpoint. ' REFERENCES 1. USAR, Section 15.6.5. i i USAR, Section 6.2.3. l 2. l

3. USAR, Section 15.7.4. !

4. USAR, Table 6.2-40.

l

                                                                                     -I 1

l l l i i l l 1 i l RIVER BEND B 3.6-95 Revision No. G LATER

A SGT System B 3.6.4.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.3 . Standby Gas Treatment (SGT) System BASES - BACKGROUND The SGT System is required by 10 CFR 50, Appendix A, GDC 41, ;

                        " Containment Atmosphere Cleanup" (Ref. 1). The function of the SGT System is to ensure that radioactive materials that leak from the primary containment into the secondary containment following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the t

environment. The SGT System consists of two fully redundant subsystems, each with its own set of ductwork, dampers, charcoal filter train, and controls. Each charcoal filter train consists of (components listed in order of the direction of the air flow):

a. A moisture separator;

b. An electric heater;

c. A prefilter;

d. A high efficiency particulate air (HEPA) filter;

e. A charcoal adsorber;

f. A second HEPA filter; and I

g. A centrifugal fan.

The SGT System serves as a backup non-ESF system to the ; Annulus Pressure Control System (APCS) during normal ! operation. Upon loss of the APCS, or upon an ESF signal (i.e., LOCA), the annulus air and air from the shielded ; compartments in the auxiliary building are automatically diverted through the SGT System filter trains. (continued) 1 I RIVER BEND B 3.6-96 Revision No. O LATER l

SGT System B 3.6.4.3 BASES BACKGROUND If the SGT System filter trains are not treating the annulus (continued) atmosphere or the exhaust air of the shielded compartments in the auxiliary building, the containment and drywell purge can be manually diverted through both SGT System filter trains. By utilizing both SGTS filter trains, a maximum of . 25,000 cfm of containment /drywell purge air can be processed ' by the filter trains. The SGT System is designed to maintain a negative pressure of at least 0.50 in W.G. in the annulus during post-LOCA operation. With the annulus at a negative pressure, any ' potential leakage is directed inward (away from the shield building). Therefore, if a primary containment DBA occurs, airborne radioactivity which exfiltrates the steel primary containment is collected and passed through a filter train of the SGT System before being released. The SGT System is also designed to maintain a negative i pressure of at least 0.25 in W.G. in.the Auxiliary Building. ' The moisture separator is provided to remove entrained water

in the air, while the electric heater reduces the relative humidity of the airstream. The prefilter removes large particulate matter, while the HEPA filter is provided to remove fine particulate matter and protect the charcoal from fouling. The charcoal adsorber removes gaseous elemental !

iodine and organic iodides, and the final HEPA filter is provided to collect any carbon fines exhausted from the charcoal- adsorber (Ref. 2). APPLICABLE The design basis for the SGT System is to mitigate the SAFETY ANALYSES consequences of a loss of coolant accident and fuel handling l accidents (Ref. 3). For all events analyzed, the SGT System-is shown to be automatically initiated to reduce, via filtration and adsorption, the radioactive material released to the environment. The SGT System satisfies Criterion 3 of'the NRC Policy Statement. LC0 Following a DBA, a minimum of one SGT subsystem is required to maintain the secondary containment at a negative pressure with respect to the environment and to process gaseous releases. Meeting the LCO requirements for two operable subsystems ensures operation of at least one SGT subsystem in the event of a single active failure. (continued) RIVER BEND B 3.6-97 Revision No. O LATER

b i

                                                                      .SGT Syst;m     !

B 3.6.4.3 BASES- (continued) APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to primary containment that leaks to secondary i containment. Therefore, SGT System OPERABILITY is required during these MODES. In MODES 4 and 5, the probability-and consequences of these events are reduced due to the pressure and temperature , limitations in these MODES. Therefore, maintaining the SGT .! System OPERABLE is not required in MODE 4 or 5. ! l l ACTIONS A.1 and A.2 With one SGT subsystem inoperable, action must be taken to verify that the OPERABLE SGT subsystem is not operating in i the primary containment purge flowpath. ! 4 Additionally, the inoperable subsystem must be restored to ) OPERABLE status within 7 days. In this Condition, the 3 remaining OPERABLE SGT subsystem is adequate to perform the ! required radioactivity release control function. However, i the overall system reliability is reduced because a. single j failure in the OPERABLE subsystem could result in the i radioactivity release control function not being adequately j performed. The 7 day Completion Time is based on : consideration of such factors as the~ availability of the .l OPERABLE redundant SGT subsystem and the low probability of- 1 a DBA occurring during this period. B.1 and B.2 l i If the SGT subsystem cannot be restored to OPERABLE status . within the required Completion Time in MODE 1, 2, or 3, the i plant must be brought to a MODE in which the LC0 does not i apply. To achieve this status, the plant must be brought to l at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant I conditions from full power conditions in an orderly manner ' and without challenging plant- systems. (continued) i RIVER BEND B 3.6-98 Revision No. O LATER

n SGT Systea B 3.6.4.3 l

                      .                                                                        l BASES (continued)

SURVEILLANCE SR 3.6.4.3.1' l REQUIREMENTS Operating each.SGT subsystem for = 10 continuous hours ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on (automatic heater cycling to maintain temperature) for 210 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system. SR 3.6.4.3.2 This SR verifies that the required SGT filter testing is i performed in accordance with the Ventilation Filter Testing i Program (VFTP). The SGT System filter tests are in accordance with Regulatory Guide 1.52 (Ref. 4). The VFTP i includes testing HEPA filter performance, charcoal adsorber , efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specified test frequencies , and additional information are discussed in detail in the < VFTP. l SR 3.6.4.3.3 This SR requires verification that each SGT subsystem starts upon receipt of an actual or simulated initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this SR to provide complete testing of the safety function. While this Surveillance can be performed with the reactor at power, operating experience has shown these' components usually pass the Surveillance when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. SR 3.6.4.3.4 This SR requires verification that the SGT filter cooling bypass damper can be opened and the fan started. This ensures that the ventilation mode of SGT System operation is (continued) RIVER BEND B 3.6-99 Revision No. O LATER

LSGT SystGm B 3.6.4.3 BASES - r

  .       . SURVEILLANCE   SR' 3.6.4.3.4     (continued)

REQUIREMENTS' , available. While this Surveillance'can be performed with- , the reactor at power, operating experience has shown these components usually pass the Surveillance when performed at : the 18 month Frequency, which is based on the refueling- . cycle. Therefore, the Frequency was concluded to be. ! acceptable from a reliability standpoint. ' REFERENCES 1. 10 CFR 50, Appendix A, GDC 41. i

2. USAR, Section 6.2.3.

3. USAR, Section 15.6.5. i

4. Regulatory Guide 1.52, Rev. 2. !

t

                                                                                          -f I

I I l RIVER' BEND B 3.6-100 Revision No. O LATER

Annulus Mixing System B 3.6.4.4 8 3.6 CONTAINMENT SYSTEMS B 3.6.4.4 Shield Building Annulus Mixing System BASES BACKGROUND The Shield Building Annulus Mixing System, in conjunction with the secondary containment and Standby Gas Treatment System, is required to ensure that radioactive materials that leak from the primary containment into the shield building annulus portion of the secondary containment following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment. Specifically, the Shield Building Annulus Mixing System provides thorough mixing of the iodine and noble gases leaking from the primary containment into the voiume between the steel containment and the shield building. The Shield Building Annulus Mixing System consists of two fully redundant subsystems, each with its own set of ductwork, dampers, and controls. The Shield Building Annulus Mixing System automatically starts and operates in response to actuation signals indicative of a LOCA. Following initiation, both shield Building Annulus Mixing fans start. APPLICABLE The design basis for the Shield Building Annulus Mixing SAFETY ANALYSIS System is to mitigate the consequences of a loss of coolant accident (Ref. 1). For the events analyzed, the Shield Building Annulus Mixing System is shown to be automatically initiated to reduce, via mixing, the quantity of radioactive material processed by the Sindby Gas Treatment System. This results in minimizing th release rates for radioactive material released to the environment. The Shield Building Annulus Mixing System satisfies Criterion 3 of the NRC Policy Statement. LC0 Following a DBA, a minimum of one shield building annulus mixing subsystem is required to adequately mix gaseous releases for processing by the Standby Gas Treatment System. Meeting the LC0 requirements for two operable subsystems ensures operation of at least one shield building annulus mixing subsystem in the event of a single active failure. (continued) RIVER BEND B 3.6-101 Revision No. O LATER

l'

3. Annulus Mixing Systea B 3.6.4.4 BASES (continued)

APPLICABILITY In MODES 1, 2, and 3, a DBA LOCA could lead to a fission product release to primary containment that leaks to secondary containment, including the annulus. Therefore, Shield Building Annulus Mixing System OPERABILITY is required during these MODES. , In MODES 4 and 5, the probability and consequences of a DBA LOCA event is reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Shield Building Annulus Mixing System OPERABLE is not required in MODE 4 or 5. ACTIONS A.1 With one shield building annulus mixing subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE shield building annulus mixing subsystem is adequate to perform the required radioactivity release mixing function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release mixing function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant shield building , annulus mixing subsystem and the low probability of a DBA occurring during this period. i B.1 and B.2 l If the shield building annulus mixing subsystem (s) cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. (continued) RIVER BEND B 3.6-102 Revision No. O LATER

i Annulus Mixing Systen B 3.6.4.4

                 .                                                                         t BASES (continued)-

SURVEILLANCE SR 3.6.4.4.1 ! REQUIREMENTS

 '                   Operating each shield building annulus mixing subsystem for a 15 minutes ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action.            >

The 31 day Frequency was developed in consideration of the ; known reliability of fan motors and controls and the ' redundancy available in the system. SR 3.6.4.4.2 This SR requires verification that each shield building ! annulus mixing subsystem starts upon receipt of an actual or simulated initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.S overlaps this SR to provide complete testing of the safety function. While this Surveillance can be performed with the reactor at power, operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency, which is based on ; the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. USAR, Section 15.6.5. i i 1

RIVER BEND B 3.6-103 Revision No. O LATER

Fuel Building B 3.6.4.5 8 3.6 CONTAINMENT SYSTEMS i B 3.6.4.5 Fuel Building 2 i BASES

     ,                                                                                                                         l BACKGROUND                                      The function of the fuel building is to contain, dilute, and hold up fission products that are released from a design basis accident. In conjunction with operation of the Fuel Building Charcoal Filtration (FBCF) System and closure of certain valves whose lines penetrate the fuel building, the fuel building is designed to reduce the activity level of the fission products prior to release to the environment.

The. fuel building is a structure that houses the spent fuel pool. This structure forms a control volume that serves to hold up and dilute the fission products. To prevent ground level exfiltration, the fuel building requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LC0 3.6.4.2, " Secondary Containment Isolation Dampers (SCIDs)," and LC0 3.6.4.7, " Fuel Building Ventilation System - Fuel Handling." APPLICABLE There are two principal accidents for which credit is SAFETY ANALYSES taken for the fuel building OPERABILITY. These are a LOCA, Rimieliinglhahdlisii r'and a Fuel Handling Accident (FHA)slithatihafeccupiedlaFt ecentlV91Wsdistedifsels(Usnifu p of Ta7 critical Treactsri core!within?thelp"eviounllCdays)' ~" inside"the f6kl~ buiTding^(Rif. "1)7^The fuel"tidilding performs no active function in response to these events; however, its leak tightness is required to ensure that the release of radioactive materials is restricted to those leakage rates assumed in the accident analysis. The fuel building satisfies Criterion 3 of the NRC Policy Statement. LC0 An OPERABLE fuel building provides a control volume into which fission products can be diluted and processed prior to release. For the fuel building to be considered OPERABLE, it must have adequate leak tightness to ensure the required vacuum can be established and maintained. I { (continued) l l RIVER BEND B 3.6-104 Revision No. O LATER

Fuel Building B 3.6.4.5 i BASES (c'ontin ed) ,

                                                                                                   .i APPLICABILITY.       In plant operating MODES 1,2, and 3, OPERABILITY of the fuel
                       ' building is addressed in LC0.3.6.4.1, " Secondary. Containment
                          - Operating." Regardless of the plant operating MODE,

' anytime recentlyjirradiated fuel,is being handled inithi fuelibsildingjthere is the potential for a FHAsinsolving handling 4recentif9Fradiatedifuelandthefuel"b611didg OPERABILITY ^is requiFed~to mftigate the consequences.jDu_e , tof rsdicadtiveidecaySthe3fssiibuilding? Mon 19freMed during?fuellhandlingLin?thekfueltbuildingUnvolwing. handling. recentlylirradi ccupied _ of; a s cri ticalc r.ated; eactor; core s fuelf(i tempreviyusy withirE the; fuelf thatihas@jydays)par 1

ACTIONS A.1 With the fuel building inoperable the plant must be brought ; to a condition in which the LCO does not apply since it is i incapable of performing its required accident mitigetion I function. To achieve this, receitiffirradiated fuel handling must be suspended immediateTy. Suspension shall not preclude completion of fuel movement to a safe position. , SURVEILLANCE SR 3.6.4.5.1 .! REQUIREMENTS This SR ensures that the fuel building boundary is sufficiently leak tight to preclude exfiltration under , expected wind conditions. The 24 hour Frequency of this SR l was developed based on operating experience related to fuel ' building vacuum variations during the applicable MODES and the low probability of a FHA %initWfselitsildihiifsolVind hahdlingf fscintlyii Fradi atedifuel$(i?eh ) occupied!partlofMcEiticaDreactsrMore%fnel@thatthas~~" Nithiniths

                                                                                ~~~p~~rssi6us    '   i 11[ days)'oEuWihfbsfWieii'IBFVei1TihesiT~~ '                          ~      '

Furthermore, the 24 hour Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal fuel building vacuum condition. SR 3.6.4.5.2 and SR 3.6.4.5.3 Verifying that fuel building equipment hatches and access doors are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not' occur. Verifying that all such openings are closed provides adequate assurance (continued)

   -RIVER BEND                               B 3.6-105                    Revision No. O LATER

                                                                            .3                    _

to:- Fue1LB uildings q, ! B 3.46.4.5 g j

       =m
                                          -                                                                                            ib1
                                                                                                                                        ,       i BASES                                                                                                        '

l i SURVEILLANCE. LSR 3.6.4.5.2 and SR~ 3.6.4.5.3 (continued)- j

                         . REQUIREMENTS
                                                                                                                                       ;j'     -

that exfiltration from'the fuel building will not occur. , Maintaining fuel building OPERABILITY. requires verifying ! j each door in'the access-opening is closed, except:when the. access opening islbeing used for entry and exit. . !

     ,                                                                                                                                         J The 31' day Frequency for these SRs has been shown to be                                        :

adequate based on operating experience,' and is considered. adequate:in view of the other indications of door and hatch. status thatiare available to the operator. REFERENCES 1. USAR, Chapter 15.- .. l

i i I i s I 1 RIVER BEND B 3.6-106 Revision No. O LATER !

 .                               .-     ,   ,  ...,-.c           -             _ , . .               -                           w e

f Fuel Building Ventilation' System-Opsrating B 3.6.4.6-B 3.6 CONTAINMENT SYSTEMS l B 3.6.4.6: Fuel Building Ventilation System-Operating BASES I l BACKGROUND The Fuel Building Ventilation System is required by j 10 CFR 50, Appendix A, GDC 41, " Containment Atmosphere , Cleanup" (Ref. 1). One function of the Fuel Building j Ventilation System is to ensure that radioactive materials : that leak from the primary containment into the Fuel ! Building portion of the secondary containment following a ' l Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment. The Fuel Building Ventilation System consists of two fully redundant fuel building ventilation charcoal filtration subsystems, each with its own set of ductwork, dampers,: charcoal filter train, and controls. Each charcoal filter train consists of (components listed in order of the direction of the air flow): ,

a. A moisture separator;

b. -An electric heater;

c. A prefilter;

d. A high efficiency particulate air (HEPA) filter;

e. A charcoal adsorber;

f. A second HEPA filter; and

g. A centrifugal fan.

The moisture separator is provided to remove entrained water in the air, while tha electric heater reduces the relative humidity of the airstream to less than 70% (Ref. 2). The prefilter removes large particulate matter, while the HEPA filter is provided to remove fine particulate matter and protect the charcoal from fouling. . The charcoal adsorber removes gaseous elemental iodine and organic iodides, and ) the final HEPA filter is provided to collect any carbon l fines exhausted from the charcoal adsorber. The Fuel Building Ventilation System automatically starts and operates the charcoal filtration subsystems in response ; to actuation signals indicative of a Fuel Handling Accident ( FHA)[i nitWfuelb bGildihgiiriiv61 Hn( hihd11ng7resentii ) irrsdiated)fde14(tyelpfuelithatihassoccupiedipartisf!i critical { reactor: coreiwithinithelprevionstilldays)~of 1.0CA. Foll6ki6g~initisfie6f tioth~fdel^buildinfiienfilation charcoal filtration fans start. (continued). RIVER BEND B 3.6-107 Revision No. O t.ATER

Fuel Building Ventilation System-Operating. B 3.6.4.6 ' BASES (continued) APPLICABLE . The design basis for the Fuel Building Ventilation SAFETY ANALYSIS System is to mitigate the consequences of a loss of coolant

accident (Ref. 2). (Additionally this sy mitigate the FHAtihitheifuelEbsildingfinv.olvingVhandlin stem functions to recently ?i rradi ated[fuelf heseWF th i sT fss t i on"a nd ~ g

Ap'plicable~ Safety Analysis is addressed in LCO 3.7.9, " Fuel Building Ventilation-System-Fuel Handling ") For the events , analyzed, the fuel building ventilation charcoal filtration -! subsystems are shown to be automatically initiated to reduce, via filtration and adsorption, the radioactive : material released to the environment. The fuel building charcoal filtration subsystems of the Fuel ; Building Ventilation System satisfies Criterion 3 of the NRC ~ Policy Statement. LCO Following a DBA, a minimum of one fuel building ventilation charcoal filtration subsystem is required to maintain the fuel building at a negative pressure with respect to the. , environment and to process gaseous releases. Meeting the LC0 requirements for two operable subsystems ensures operation of at least one fuel building ventilation charcoal filtration subsystem in the event of a single active failure. APPLICABILITY In MODES.1, 2, and 3, a DBA LOCA could lead to a fission I product release to primary containment that leaks to ,' secondary containment, including the fuel building. Therefore, fuel building ventilation charcoal filtration subsystem OPERABILITY is required during these MODES. In MODES 4 and 5, the probability and consequences of a DBA LOCA event is reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the fuel building ventilation charcoal filtration subsystems OPERABLE is not required in MODE 4 or 5, except for other situations under'which significant releases of radioactive material can ., bepostulated,suchasduringmovementofEecentii irradiated criti fuel (iieMfuelitha%h'as h13re_act.ortcore;withiniit.herpre$os v.ious M daisMpledfpart3 Q ; g g g. gm 7gggg Th"s^0PERABIl^ITY)"of the ' Fuel Building Ventilation System during 'rssstli'jirradiated fuel handling is addressed in LCO 3.6.4.7~,"~"FGel Building j Ventilation System-Fuel Handling." (continued) RIVER BEND B 3.6-108 Revision No. O LATER

Fuel Building Ventilation System-Operating ' B 3.6.4.6 BASES (continued) ACTIONS. .A_d ; With one fuel building ventilation charcoal filtration - subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status.within 7 days. In this condition, the remaining OPERABLE fuel building ventilation a charcoal filtration subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single i failure in the OPERABLE subsystem could result in the-radioactivity release control function not being adequately , performed. The 7 day Completion Time is based on i consideration of such factors as the availability of the : OPERABLE redundant fuel building ventilation charcoal filtration subsystem and the low probability of a DBA occurring during this period. B.1 and B.2 If the fuel building ventilation charcoal filtration subsystem cannot be restored to OPERABLE status within the ; required Completion Time, the plant must be brought to a 4 MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed l Completion Times are reasonable, based on operating ' experience, to reach the required plant conditions from full l power conditions in an orderly manner and without ; challenging plant systems. SURVEILLANCE SR 3.6.4.6.1 REQUIREMENTS Operating each fuel building ventilation charcoal filtration subsystem for a 10 continuous hours ensures that both subsystems are OPERABLE and 1. hat all associated controls are functioning properly. It also ensures that blockage, fan or ! motor' failure, or excessive vibration can be detected for ; corrective action. Operation with the heaters operating (automatic heater cycling to maintain temperature) for a 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was , developed in consideration of the known reliability of fan I motors and' controls and the redundancy available in the i system. (continued) m: RIVER BEND B 3.6-109 Revision No. 9 LATER

Fu 1 Building Ventilaticn System-Operating { B 3.6.4.6 l BASES' SURVEILLANCE SR 3.6.4.6.2 REQUIREMENTS (continued) This SR verifies that the required fuel building ventilation

   '                    charcoal filtration filter testing is performed in .                     l accordance with the Ventilation Filter Testing Program                  '

(VFTP). The fuel building ventilation charcoal filtration subsystem filter tests are in accordance with Regulatory ! Guide 1.52 (Ref. 3). The VFTP includes testing HEPA filter i performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated i charcoal (general use and following specific operations). Specified test frequencies and additional information are ], discussed in detail in the VFTP, j SR 3.6.4.6.3 l l This SR requires verification that each fuel building i ventilation charcoal filtration subsystem starts upon l receipt of an actual or simulated initiation signal. The ! LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this l SR to provide complete testing of the safety function. ' While this Surveillance can be performed with the reactor at i power, operating experience has shown these components ! usually pass the Surveillance when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from

7. reliability standpoint.

I SR 3.6.4.6.4 l This SR requires verification that the fuel building l ventilation charcoal filtration filter cooling bypass damper ' can be opened and the fan started. This ensures that the ventilation mode of Fuel Building Ventilation Charcoal Filtration System operation is availa'le.o While this Surveillance can be performed with the reactor at power, l operating experience has shown these components usually pass ; the Surveillance when performed at the 18 month Frequency, ' which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. (continued) , i RIVER BEND B 3.6-110 Revision No. O LATER

7 n -

Fuel Building Ventilatien System-Operating _B 3.6.4.6

BASES (continued) i REFERENCES 1. 10 CFR 50, Appendix A, GDC 41.

2. USAR, Section 9.4.2.

3. Regulatory Guide 1.52, Rev. 2. ;

1 1 l l l i l RIVER BEND B 3.6-111 Revision No. O LATER !

7

    -                                                                                            t Fuel Building Ventilation-System-Fuel Handling          -

B 3.6.4.7 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.7 Fuel Building Ventilation System-fuel Handling BASES BACKGROUND The Fuel Building Ventilation System is required by i 10 CFR-50, Appendix A, GDC 41, " Containment Atmosphere Cleanup" (Ref. 1). The function of the Fuel Building l Ventilation System is to ensure that radioactive materials that escape from fuel assemblies damaged following a design basis Fuel Handling Accident (FHA)sinsidssthelfseldbuilding i nvol vi ng! handl i ng jecentlyli rFadi atedi fuel s (Me M fuel ithat has:occupiedpart;of?afcritical:/reactoricorewithinsthe ~

~~

previousJllidays)'"arefilteredandadsoIFbed' prior ^th exhausting'to the environment. The Fuel Building Ventilation System consists of two fully redundant subsystems, each with its own set of ductwork, dampers, charcoal filter train, and controls. l Each charcoal filter train consists of (components listed in ' order of the direction of the air flow):

a. A moisture separator;

b. An electric heater;

c. A prefilter;

d. A high efficiency particulate air (HEPA) filter; j

e. A charcoal adsorber; '

f. A second HEPA filter; and

g. A centrifugal fan with inlet flow control vanes.

The moisture separator is provided to remove entrained water j in the air, while the electric heater reduces the relative i humidity of the airstream to less than 70% (Ref. 2). The ! prefilter removes large particulate matter, while the HEPA ' filter is provided to remove fine particulate matter and protect the charcoal from fouling. The charcoal adsorber removes gaseous elemental iodine and organic iodides, and I the final HEPA filter is provided to collect any carbon fines exhausted from the charcoal adsorber. The Fuel Building Ventilation System automatically starts i and operates in response to actuation signals indicative of conditions or an accident that could require operation of the system. l (continued) RIVER BEND B 3.6-112 Revision No. O LATER

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

1 3 n Fuel Building Vtntilation System 4uel Handling

                                                                                                                                'i B 3.6.4.7               1 I

BASESL (continued)- !

                                                                                                                                  \
                ~ APPLICABLE         The design basis for. the Fuel Building Ventilation' System is                               l SAFETY ANALYSES. to mitigate the consequences of a. fuel handling accident                                    !

(Ref. 3 finvolwing?handlingire d atlyIDrediatodifsell Ce !

~*

fuelfth)t1Thasioccupied;partf6fsatcriticaUreactortcsre( a ~ ~y l withii(thstpreviousil11daysFF5 Fill"eilehti~nalized, . the l F0el;^ Building Ventilation System is shown to reduce, via ; filtration and adsorption, the radioactive material released l to the environment. Since the system is assumed to filter j all releases, with the analysis not. accounting for any delay 1 in system startup, at least one subsystem must be in .i operationwhilehandlingrecentlyjirradiatedfuel. ! The Fuel Building Ventilation System. satisfies Criterion 3 4 of the NRC Policy Statement. j 1 LC0 Following .a FHAsinsidsstEfsillbiiilldisillsbliiihg'ihandl.iM l redestlysi rradiatodi fue1 MiieMfeelethetihas%sceptedipairt of!alcriticallreactoricoreiwithinithe! reviedss11sdays P mifiinius'bf 6ni Fsel"Bhildihg'Vinfilit3ori~subsFiteE~ is .)', T~ , required to maintain the fuel building at.a negative- ) pressure with respect to the environment'and to process-gaseous releases. Meeting the LCO requirements for two-operable subsystems ensures operation of at least one Fuel Building Ventilation subsystem in the event of a1 single active. failure. Requiring'one subsystem to be in operation ensures'no releases occur that are not filtered'and adsorbed. APPLICABILITY In plant operating MODES 1,2 and 3, OPERABILITY of the fuel building is addressed.in LCO 3.6.4.1, " Secondary Containment

                                    -Operating." Re anytime resentl)gardless 4 irradiated  of fuel the plant is being    operating handledMODE,-

there is the potehtiaT~ fir'a FHAlis tdeithisfueE bkildla handlingWiciatly!iifsd1atodifsel"~ihd ~the" Fo BEilding" el" gYinvolviM -- ; VeritilitiBd"Systei*is7sq'GfFid^t6 mitigate the consequences.* 1' Des?tiWiiHsactliiieIdiiisiyMthifFsiiMgitMisijVsistilation 59stenitiscalyf requirididurt(1%ielshandlingtin$thes fuel: building fue15tha $hhisccupied)partsinvo1Wgshandl.ing?recenti tNishrJectetcor,e

~~

y{Q M thelp g [eusill(days %)3 (cirpster (continued) i RIVER BEND B 3.6-113 Revision No. O LATER

Fu 1 Building Ventilation System - Fuel Handling ! B 3.6.4.7 BASES (continued) l ACTIONS A.1 ! With one fuel building ventilation charcoal filtration subsystem inoperable, the inoperable subsystem must be , restored to OPERABLE status within 7 days. In this ! Condition, the remaining OPERABLE fuel building ventilation i charcoal filtration subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the ! radioactivity release control function not being adequately performed. The 7 day Completion Time is based on . consideration of such factors as the availability of the l 0PERABLE redundant fuel building ventilation charcoal : filtration subsystem and the low probability of a FH Minside the fuelebuildingtinVolving h'andlin %centl ~~~ yjirradiatef fuel' occurring"during this period ~~g"' l 3.1 and R.2 If the fuel building ventilation charcoal filtration subsystem cannot be restored to OPERABLE status within the required Completion Time the plant must be brought to a condition in which the LCO does not apply. Additionally, if both subsystems are inoperable or if the one required subsystem not in operation the system is incapable of performing its required accident mitigation function and the plant must be brought to a condition in which the LC0 does not apply. Toachievethis,rehentifjirradiatedfuel handling must be suspended immediately. Suspension shall not preclude completion of fuel movement to a safe position. SURVEILLANCE SR 3.6.4.7.1 REQUIREMENTS This Surveillance demonstrates that one fuel building ventilation charcoal filtration subsystera is in operation and filtering the fuel building atmosphere. The Frequency of 12 hours is sufficient in view of other visual and audible indications available to the operator for monitoring the fuel building ventilation charcoal filtration subsystem in the control room. (continued) 1 RIVER BEND B 3.6-114 Revision No. O LATER I l

Fu21 Building V;ntilation System - Fuel Handling B 3.6.4.7 BASES SURVEILLANCE SR 3.6.4.7.2 RF0VIREMENTS (continued) Operating each fuel building ventilation charcoal filtration

subsystem for a 10 continuous hours elsures that both subsystems are OPERABLE and that all associated controls are functioning properly. It als) ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters operating (automatic heater cycling to maintain temperature) for a 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system.

SR 3.6.4.7.3 l This SR verifies that the required fuel building ventilation charcoal filtration filter testing is performed in accordance with the Ventilation Filter Testing Program ! (VFTP). The fuel building ventilation charcoal filtration l filter tests are in accordance with Regulatory Guide 1.52 ) (Ref. 4). The VFTP includes testing HEPA filter I performance, charcoal adsorber efficiency, minimum system l flow rate, and the physical properties of the activated , charcoal (general use and following specific operations). l Specified test frequencies and additional information are ' discussed in detail in the VFTP. 1 SR 3.6.4.7.4 l This SR requires verification that each fuel building ventilation charcoal filtration subsystem starts upon l receipt of an actual or simulated initiation signal. The i LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this SR to provide complete testing of the safety function. While this Surveillance can be performed with the reactor at l power, operating experience has shown these components ' usually pass the Surveillance when performed at the 18 month Frequency, which is based on the refueling cycle, i Therefore, the Frequency was concluded to be acceptable from l a reliability standpoint. (continued) RIVER BEND B 3.6-115 Revision No. O LATER

Fuel Building Ventilation System - Fuel Handling B 3.6.4.7 BASES SURVEILLANCE SR 3.6.4.7.5 REQUIREMENTS (continued)' This SR requires verification that the fuel building ventilation charcoal filtration filter cooling bypass damper can be opened and the fan started. This ensures that the ventilation mode of Fuel Building Ventilation System operation is available. While this Surveillance can be performed with the reactor at power, operating experience i has shown these components usually pass the Surveillance when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the, Frequency was concluded to be acceptable from a reliability standpoint. 1 REFERENCES 1. 10 CFR 50, Appendix A, GDC 41. ,

2. USAR, Section 6.2.3.

3. USAR, Section 15.6.5. l

4. Regulatory Guide 1.52, Rev. 2.

RIVER BEND B 3.6-116 Revision No. O LATER

s Drywell B 3.6.5.1 B 3.6 CONTAINMENT SYSTEMS B~3.6.5.1 Drywell BASES - BACKGROUND The drywell houses the reactor pressure vessel (RPV), the reactor coolant recirculating loops, and branch connections of the Reactor Coolant System (RCS), which have-isolation valves at the primary _ containment boundary. The function of the drywell is to maintain a pressure boundary that channels steam from a loss of coolant accident (LOCA) to the suppression pool, where it is condensed. Air forced from the drywell is released into the primary containment through the suppression pool. The pressure suppression capability of the suppression pool assures that peak LOCA temperature and pressure in the primary containment are within design limits. The drywell also protects accessible areas of the i containment from radiation originating in the reactor core and RCS. To ensure the drywell pressure suppression capability, the drywell bypass leakage must be minimized to prevent overpressurization of the primary containment during the drywell pressurization phase of a LOCA. This requires periodic testing of the drywell bypass leakage, confirmation that: the drywell air lock is leak tight, OPERABILITY of the drywell isolation valves. The isolation devices for the drywell penetrations are a part of the drywell barrier. To maintain this barrier:

a. The drywell air lock is OPERABLE except as provided in LC0 3.6.5.2, "Drywell Air Lock";

b. The drywell penetrations required to be closed during accident conditions are either:

1. capable of being closed by an 0PERABLE automatic 1 drywell isolation valve, or l

2. closed by a manual valve, blind flange, or de-activated automatic valve secured in the ;

closed position except as provided in LCO 3.6.5.3, "Drywell Isolation Valves."; and

c. All drywell equipment hatches are closed and sealed. .

1 (continued) RIVER BEND B 3.6-117 Revision No. G LATER

Drywall B 3.6.5.1 j BASES BACKGROUND This Specification is intended to ensure that the (continued) performance of the drywell in the event of a DBA meets the l assumptions used in the safety analyses (Ref. 1). APPLICABLE Analytical methods and assumptions involving the drywell are SAFETY ANALYSES presented in Reference 1. The safety analyses assume that for a high energy line break inside the drywell, the steam is directed to the suppression pool through the horizontal vents where it h condensed. Maintaining the pressure suppression tapability assures that safety analyses remain valid and that the peak LOCA temperature and pressure in the primary containraent are within design limits. The drywell satisfies Criteria 2 and 3 of the NRC Policy Statement. LC0 Maintaining the drywell 0PERABLE is required to ensure that ) the pressure suppression design functions assumed in the safety analyses are met. The drywell is OPERABLE if the drywell structural intergrity is intact and the bypass ; leakage is within limits, except prior to the first startup i after performing a required drywell bypass leakage test. At ' this time, the drywell bypass leakage must be :s 10% of the drywell bypass leakage limit. APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to the primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of : these MODES. Therefore, the drywell is not required to be l OPERABLE in MODES 4 and 5. l ACTIONS A.1 l In the event the drywell is inoperable, it must be restored to OPERABLE status within 1 hour. The 1 hour Completion Time provides a period of time to correct the problem i commensurate with the importance of maintaining the drywell l OPERABLE during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring drywell OPERABILITY) occurring during periods when the ; I (continued) RIVER BEND B 3.6-118 Revision No. G LATER

L' Drywell

   ,                                                           %                                                                   B 3.6.5.1 BASES ACTIONS                                         Ad (continued)

drywell is inoperable is minimal. Also, the Completion Time is the same as that applied to inoperability of the primary containment in LCO 3.6.1.1, " Primary Containment 4)perating."

8.1 and 8.2 If the drywell cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed [ Completion Times are reasonable, based on operating experience,-to reach the required plant _ conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.5.1.1 REQUIREMENTS The seal air flask pressure is verified to be at at 75 psig every 7 days to ensure that the seal system remains viable. It must be checked because it could bleed down during or following access through the personnel door. The 7 day Frequency has been shown to be acceptable through operating experience and is considered adequate in view of the other indications available to operations personnel that the seal air flask pressure is low. l SR 3.6.5.1.2 A seal pneumath system test to ensure that pressure does not decay at r + ate equivalent to > 0.67 psig for a period of 24 hours from an ;nitial pressure of 75 psig is an ! effective leakage rate test to verify system performance. l The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant i outage and the potential for an unplanned transient if the t Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

                                                                                   ,                                              (continued)

RIVER BEND B 3.6-119 Revision No. O LATER

r; Drywell B 3.6.5.1 BASES SURVEILLANCE SR 3.6.5.1.3 i REQUIREMENTS

  '       (continued) ~The analyses in Reference 1 are based on a maximum drywell bypass leakage. This Surveillance ensures that the actual       1 drywell bypass leakage is less than or equal to the             '

acceptable A/V'Ic design value of 1.0 ft 2assumed in the safety analysis. As left drywell bypass leakage, prior to the first startup after performing a required drywell bypass ; leakage test, is required to be :s; 10% of the drywell bypass leakage limit. At all other times between required drywell leakage rate tests, the acceptance criteria is based on design A/V'Ic. At the design A/V'Tc the containment temperature and pressurization response are bounded by the ! assumptions of the safety analysis. The leakage test is performed every 18 months, consistent with the difficulty of performing the test, risk of high radiation exposure, and the remote possibility that a component failure that is not ! identified by some other drywell or primary containment SR { might occur. Operating experience has shown that these ' components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was 1 concluded to be acceptable from a reliability standpc11c. I SR 3.6.5.1.4 The exposed accessible drywell interior and exterior surfaces are inspected to ensure there are no apparent physical defects that would prevent the drywell from performing its intended function. This SR ensures that drywell structural integrity is maintained. The Frequency was chosen so that the interior and exterior surfaces of the drywell can be inspected in conjunction with the inspections of the primary containment required by 10 CFR 50, Appendix J (Ref. 2). Due to the passive nature of the drywell structure, the specified Frequency is sufficient to identify component degradation that may affect drywell structural integrity. REFERENCES 1. USAR, Chapter 6 and Chapter 15. l 1 RIVER EEND B 3.6-120 Revision No. O LATER ;

Drywell Air Lock B 3.6.5.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.5.2 Drywell Air Lock i BASES I BACKGROUND The drywell air lock forms part of the drywell boundary and provides a means for personnel access during MODES 2 and 3 during low power phase of unit startup. For this purpose, one double door drywell air lock has been provided, which maintains drywell isolation during personnel entry and exit from the drywell. Under the normal unit operation, the drywell air lock is kept sealed. The air pressure in the seals is maintained > 60 psig by the seal air flask and , pneumatic system, which is maintained at a pressure l a: 75 psig. The drywell air lock is designed to the same standards as the drywell boundary. Thus, the drywell air. lock must withstand the pressure and temperature transients associated with the rupture of any primary system line inside the drywell and also the rapid reversal in pressure when the steam in the drywell is condensed by the Emergency Core Cooling System flow following loss of coolant accident flooding of the reactor pressure vessel (RPV). It is also designed to withstand the high temperature associated with i the break of a small steam line in the drywell that does not result in rapid depressurization of the RPV. The air lock is nominally a right circular cylinder,10 ft in diameter, with doors at each end that are interlocked to prevent simultaneous opening. During periods when the , drywell is not required to be OPERABLE, the air lock interlock mechanism may be disabled, allowing both doors of the air lock to remain open for extended periods when frequent drywell entry is necessary. Each air lock door has been designed and tested to certify its ability to withstand a pressure in excess of the maximum expected pressure i following a Design Basis Accident (DBA). l The air lock is provided with limit switches on both doors that provide control room indication of door position. q l The drywell air lock forms part of the drywell pressure L' boundary. Not ;intaining air lock OPERABILITY may result in degradation of the pressure suppression capability, which is assumed to be functional in the unit safety analyses. (continued) RIVER BEND B 3.6-121 Revision No. 9 LATER

-ri , M5 . o : Drywell Air Lock B 3.6.5.2 BASES BACKGROUND The drywell air lock does not need to meet the requirements

           '(continued)    of 10 CFR 50, Appendix J (Ref. 1), since it.is not part of the primary containment leakage boundary. However, it is        :

o prudent to specify a leakage rate requirement for the drywell air lock. A seal leakage rate limit of s; 4.05 scfh ! and an air lock overall leakage rate limit of s 11.85 scfh, at 3.0 psid, have been established to assure the integrity of the seals. ' APPLICABLE Analytical methods and assumptions involving the drywell are SAFETY ANALYSES presented in Reference 2. The safety analyses assume that , for a high energy line break inside the drywell, the steam is directed to the suppression pool through the horizontal vents where it is condensed. Since the drywell air ~ lock is part of the drywell pressure boundary, its design and-maintenance are essential to support drywell OPERABILITY, which assures that the safety analyses are met. The drywell air lock satisfies Criterion 3 of the NRC Policy Statement. LC0 The drywell air lock forms part of the drywell pressure boundary. The air lock safety function assures that steam resulting from a DBA is directed to the suppression pool. Thus, the air lock's structural integrity is essential to the successful mitigation of such an event. The air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, air lock leakage must be within limits, and both air lock doors must be OPERABLE. The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of the drywell does not exist when the drywell is required to be OPERABLE. Closure of a single door in the air lock is necessary to support drywell 0PERABILITY following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for entry into and exit from the drywell. ' (continued) I RIVER BEND B 3.6-122 Revision No. O LATER

Drywell Air Lock B 3.6.5.1 . BASES (continued) APPLICA31LITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to the primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, the drywell air lock is' not required to be OPERABLE in MODES 4 and 5. i ACTIONS The ACTIONS are modified by Note 1 which that allows entry and exit to perform repairs on the_ affected air lock component. If the outer door is inoperable, then it may be easily accessed to repair. If the inner door is inoperable, however, then there is a short time during which the drywell boundary is not intact (during access through the outer ; door). The ability to open the OPERABLE door, even if it i means the drywell boundary is temporarily not intact, is ! acceptable due to the low probability of an event that could pressurize the drywell during the short time in which the , OPERABLE door is expected to be open. The OPERABLE door i must be immediately closed after each entry and exit. The ACTIONS are modified by a second Note, which ensures appropriate remedial actions are taken when necessary. Pursuer.L te LC0 3.0.6, ACTIONS are not required even if the drywell is exceeding its bypass leakage limit. Therefore, the Note is added to require ACTIONS for LC0 3.6.5.1 to be ) taken in this event. A.I. A.2, and A.3 With one drywell air lock door inoperable, the OPERABLE door must be verified closed (Required Action A.1). This ensures that a leak tight drywell barrier is maintained by the use of an OPERABLE air lock door. This action must be completed within 1 hour. The 1 hour Completion Time is consistent with the ACTIONS of LC0 3.6.5.1, "Drywell," which requires that the drywell be restored to OPERABLE status within I hour. In addition, the air lock penetration must be isolated by i locking closed the OPERABLE air lock door within the 24 hour : Completion Time. The Completion Time is considered reasonable for locking the OPERABLE air lock door, considering that the OPERABLE door is being maintained closed. . _ (continued) ' RIVER BEND B 3.6-123 Revision No. O LATER i

4 s - p E * [!? , . [;

                                                                                . Drywell' Air Lcck            j
                                                                                             -B 3.6.5.2--         i
                                                                                                                ]

[ BASES' .!

ACTIONS:: A.l'. A.2. -and A.3 (continued)'

Required Action A.3 verifies that the' air lock'has been .. isolated by the use of a locked and closed OPERABLE air lock : m ' door. This-ensures ~ that an acceptable drywell boundary is l maintained. -The Completion.' Time of.once-per 31 days-is . J based on engineering judgment and is considered adequate in -i view of.the low likelihood of a. locked door being mispositioned and other administrative controls that ensure ! that the.0PERABLE air lock door remains closed. j The Required Actions are. modified by two Notes. Note 1 ) ensures only the Required Actions and associated Completion j Times'of Condition ~ C are required if bcth doors in the air- i lock are inoperable. ..The exception of the Note does not- ! affect tracking the Completion Times from the initial entry . into Condition A; only the. requirement-to comply with the i) Required Actions. Note 2 allows use of the air. lockifor 1 entry and exit for 7 days under administrative controls.- l Drywell entry may be required to perform Technical Specifications (TS) Surveillances and Required Actions, as - well as other activities on equipment inside the:drywell that are required by TS or activities on equipment that support TS-required equipment. This Note is not intended to preclude performing other activities (i.e., non-TS-required activities) if the drywell was entered, using the inoperable ~ air lock, to perform an allowed activity listed above. The administrative controls required consist of the stationing of a dedicated individual to assure closure of the OPERABLE door except during the entry and exit,.and assuring the OPERABLE door is relocked after completion of the drywell entry and exit. This allowance is acceptable due to the low probability of an event that could pressurize the;drywell during the short time that the OPERABLE door _is expected to be open. L1. B.2. and B.3 a With the drywell air lock interlock mechanism inoperable, i the Required Actions and associated Completion Times consistent with Condition A are applicable. The Required Actions are modified by two Notes. Note I: ensures only the Required. Actions and associated Completion 1 Times of Condition C are required if. bothLdoors in the air -l lock are inoperable. Note 2 allows entry and exit into the 1 (continued) RIVER BEND B 3.6-124 Revision No. 9 LATER H a

                -.m-.
                                           ~

j s , -! i

                                                                             -Drywell-Air L5ck J

B 3.6.5.2' -l r

              -BASES                                                                                  '
             -ACTIONS         B.I. B.2. and B.3' (continued)
                                                                                                   ;i drywell under the control of a dedicated individual .                   '

stationed 'at-the air lock to ensure that only one door is : openet at a time. ! C.I. C.2. and C.3 .j E-With the air lock inoperable for reasons other than those described in Condition A or.B, Required Action C.1 requires- I action to be immediately initiated to evaluate drywell ; bypass leakage using current air ' lock. test results. An :- evaluation is acceptable, since it is overly conservative.to ! immediately declare the drywell inoperable if both doors in j an air lock have failed a seal test or the overall air lock ; leakage is not'within limits. In many instances.(e.g., only 1 one seal per door has failed),. drywell'rerains OPERABLE, yet ! only I hour (per LCO 3.6.5.1) would.be provided to restore j the air lock door to OPERABLE status prior to requiring a : plant shutdown. In addition, even with both doors failing i the seal test, the overall drywell leakage rate can still be j within limits. y Required Action C.2 requires that one door in'the drywell. air lock must be verified to be closed.- This Required Action must be completed within the I hour Completion Time. This specified time period is consistent with the ACTIONS of- , LCO 3.6.5.1, which requires that the drywell be restored to OPERABLE status within 1 hour. Additionally, the air lock must be restored to OPERABLE status within 24 hours. The 24. hour Completion Time is reasonable for restoring an inoperable air lock to OPERABLE status, considering that at least one door is maintained closed in the air lock. D.1 and D.2 If the inoperable drywell air lock cannot he restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE.in which the LCO does not apply. To achieve this status,.the plant must be brought to . 'I at least MODE 3 within~12 hours and to MODE 4 within 36 hours. The allowed Completion. Times are reasonable, j based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner q and without challenging plant systems. (continued)

            -RIVER BEND                          B 3.6-125                 Revision No. O LATER

B v

                                                                                                                    -1 Drywell Air Lock             i B 3.6.5.2          -i .

BASES (continued) l SURVEIL' LANCE SR 3.'6.5.2.1 I REQUIREMENTS . . j This SR requires a test.be' performed to verify seal leakage ;

of the drywell: tiir lock' doors at 3.0 psid. A seal: leakage !

rate limit of.:s; 4.05 scfh has.been established to ensure i the integrity.of'the seals. .The Surveillance is only

                                                                  .                                                     j required to be performed once within 72 hours after each                                j closing. The Frequency of 72 hours is' based on operating:                               i experience.                                                                              ;
                                                                                                                      'i SR' 3.6.5.2.2                                                                         -i Every 7 days the drywell air lock seal air flask pressure is verified to be a: 75 psig to ensure that the seal system                    ..

I remains viable. It must be checked because it could bleed- ! down during or following access through.the air lock, which j occurs regularly. The 7 day Frequency has been shown to be 1 acceptable, based on operating experience, and.is considered ; adequate in view of the other indications to the plant 2 operations personnel that the seal air flask pressure is l low. l1 SR 3.6.5.2.3 The air lock door interlock is. designed to prevent simultaneous opening of both doors in the air lock. .Since both the inner and outer doors of the air . lock are designed to withstand..the maximum expected post accident drywell. pressure, closure of either door will support drywell OPERABILITY. Thus, .the door interlock feature- supports drywell OPERABILITY while the air lock is being used for personnel transit in and out of the drywell. Periodic testing and preventive maintenance of this interlock demonstrates that the-. interlock will function as designed and that simultaneous inner and outer door-opening will not inadvertently occur. Due to the nature of this interlock, and given that the interlock mechanism is only challenged- ! when a.drywell air lock door is opened, this test is only i required to be_ performed once every 18 months. The 18 month ' Frequency is based on the need to perform this Surveillance l under the reduced reactivity conditions that apply during a 1

                              - plant outage and the potential for violating the drywell l

boundary. Operating-experience has;shown these components I usually pass the Surveillance when performed at the 18 month (continued) , i RIVER BEND B 3.6-126 Revision No. 9.LATER-m - - . +,. +, , .-y --

                                                                -        y   --, --,,-,-,n              ,3    9 m , , , <

Drywell Air Lock B 3.6.5.2 l BASES SURVEILLANCE- SR 3.6.5.2.3 (continued) REQUIREMENTS , Frequency, which is based on the refu W ng cycle. . Therefore, the Frequency was concin i be acceptable from a reliability standpoint. The Surveillance is' modified by a Note requiring the Surveillance to be performed only upon entry into the drywell. SR 3.6.5.2.4 This SR requires a test to be performed to verify overall air lock leakage of the drywell air lock at 3.0 psid. The i 18 month Frequency is based on the need to perform this , Surveillance under the conditions that apply during a plant- ' outage and the potential for violating the drywell boundary. ; Operating experience has shown these components usually pass : the Surveillance when performed at the 18 month Frequency, 1 which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. This SR has been modified by two Notes. The first Note l indicates that an inoperable air lock door does not invalidate the previous successful performance of an overall air lock leakage test. This is considered reasonable, since either air lock door is capable of providing a fission product barrier in the event of a DBA. The Surveillance is modified by a Note requiring the air lock to be pressurized to 19.2 psid prior to performance of the overall air lock leakage test. The 19.2 psid differential pressure is the assumed peak drywell pressure expected from the accident analysis. Since the drywell pressure rapidly returns to a steady state maximum differential pressure of 3.0 psid (due to suppression pool , vent clearing), the leakage is allowed to be measured at ' this pressure. SR 3.6.5.2.5 This SR ensures that the drywell air' lock seal pneumatic system pressure does not decay at an unacceptable rate. The air lock seal will support drywell 0PERABILITY down to a pneumatic pressure of 75 psig. Since the air lock seal air l (continued) i RIVER BEND B 3.6-127 Revision No. O LATER l

i Drywell Air Lock-B 3.6.5.2, BASES SURVEILLANCE SR '3.6.5.2.5 (continued)~ REQUIREMENTS flask pressure is verified in SR 3.6.5.2.2 to be a: 75 psig,

       ,                                          a decay rate s 0.67 psig over. 24 hours is acceptable. . The 24 hour interval is based on engineering' judgment, considering that there is no postulated DBA where the drywell is still pressurized 24 hours after the event. The.

18 month Frequency is based on the.need to perform this Surveillance under the conditions that apply during aLplant outage when.the. air. lock OPERABILITY is not required.' Operating experience has shown that these components usually pass the Surveillance when performed.at the 18 month ~ Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES '1. .10 CFR 50, Appendix J.

2. USAR, Chapters 6 and-15.

a i l l l i. RIVER BEND B 3.6-128 Revision No. O LATER-

Orywall Isolation Valves B 3.6.5.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.5.3 Drywell Isolation Valves BASES BACKGROUND The drywell isolation valves, in combination with other accident mitigation systems, function to ensure that steam and water releases to the drywell are channeled to the suppression pool to maintain the pressure suppression function of the drywell. The OPERABILITY requirements for drywell isolation valves help ensure that valves are closed, when required, and isolation occurs within the time limits specified for those isolation valves designed to close automatically. Therefore, the OPERABILITY requirements support minimizing drywell bypass leakage assumed in the safety analysis (Ref. 1) for a DBA. Typically, two barriers in series are provided for each penetration so that no credible single failure or malfunction of an active component can result in a loss of isolation. The isolation devices addressed by this LC0 are either passive or active (automatic). Manual valves, de-activated automatic valves secured in their closed position, check valves with flow through the valve secured, and blind flanges are considered passive devices. Check valves and automatic valves designed to close without operator action following an accident, are considered active devices. The Drywell Vent and Purge System is a high capacity system with a 24 inch line, which has isolation valves covered by this LCO. The system supplies filtered outside air directly to the drywell through two primary containment isolation valves (PCIVs) and two drywell isolation valves called drywell purge isolation valves. The drywell air is exhausted through a line also containing two drywell purge isolation valves via both divisions of the SGTS or a low volume c"rge through 1HVR-FLT6/FN14. 1HVR-FLT6/FN14 may also be H to , circulate the drywell atmosphere. The system is used to remove trace radioactive airborne products prior to personnel entry. The Drywell Vent and Purge System is not used in MODE 1, 2, or 3; therefore, the drywell purge isolation valves are sealed shut during power operation. (continued) RIVER BEND B 3.6-129 Revision No. O LATER

l l Drywell Isolation Valves l B 3.6.5.3 BASES ! BACKGROUND The drywell purge isolation valves fail closed on loss of (continued) instrument air or power. The drywell purge isolation valves j ,, are fast closing valves (approximately 4 seconds). ; i a APPLICABLE This LC0 is intended to ensure that releases from the core SAFETY ANALYSES do not bypass the suppression pool so that the pressure suppression capability of the drywell is maintained. Therefore,'as part of the drywell boundary, drywell isolation valve OPERABILITY minimizes drywell bypass leakage. Therefore, the safety analysis of any event requiring isolation of the drywell is applicable to this LCO. The limiting DBA resulting in a release of steam, water, or , radioactive material within-the drywell is a LOCA. In the analysis for this accident, it is assumed that drywell isolation valves either are closed or function to close within the required isolation time following event initiation. The drywell isolation valves and drywell purge isolation valves satisfy Criterion 3 of the NRC Policy Statement. LC0 The drywell isolation valve safety function is to form a part of the drywell boundary. The power operated drywell isolation valves are required to have isolation times within limits. Power operated automatic drywell isolation valves are also required to actuate on an automatic isolation signal. Additionally, i drywell purge valves are required to be closed. The normally closed isolation valves or blind flanges are considered OPERABLE when, as applicable, manual valves are closed or opened in accordance with applicable administrative controls, automatic valves are de-activated and sicured in their closed position, check valves with flow through the valve secured, or blind flanges are in place. The valves covered by this LC0 are included (with their i associated stroke time, if applicable, for automatic valves) in Reference 2. (continued) RIVER BEND B 3.6-130 Revision No. O LATER

Drywell Isolaticn Valves B 3.6.5.3 BASES (continued) APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to the primary containment. In MODES 4 and 5, the probability and consequences of these events are

    '                                                          reduced due to the pressure and temperature limitations in these MODES. Therefore, the drywell isolation valves are not required to be OPERABLE in MODES 4 and 5.

ACTIONS The ACTIONS are modified by four Notes. The first Note allows penetration flow paths, except for the 24 inch purge valve penetration flow paths, to be unisolated intermittently under administrative controls. These - controls consist of stationing a dedicated operator, who is in continuous communication with the control room, at the controls of the valve. In this way, the penetration can be rapidly isolated when a need for drywell isolation is indicated. The second Note provides clarification that for the purpose of this LCO separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable drywell isolation valve. Complying with the Required Actions may allow for continued operation, and subsequent inoperable drywell isolation valves are governed by subsequent Condition entry and application of associated Required Actions. The third Note requires the OPERABILITY of affected systems to be evaluated when a drywell isolation valve is inoperable. This ensures appropriate remedial actions are taken, if necessary, if the affected system (s) are rendered inoperable by an inoperable drywell isolation valve. The fourth Note ensures appropriate remedial actions are taken when the drywell bypass leakage limits are exceeded. Pursuant to LC0 3.0.6, these ACTIONS are not required even when the associated LCO is not met. Therefore, Notes 3 and 4 are added to require the proper actions are taken. A.I and A.2 With one or more penetration flow paths with one drywell isolation valve inoperable, the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. i (continued) l RIVER BEND B 3.6-131 Revision No. O LATER

,o , i

Drywell .Iselation Valves -

[ B 3.6;5.3

-i

BASES '

                               -ACTIONS                      :

A.1 and A.21 (continued)l 1 o Isolation' barriers that meet this criterion.are a closed and de-activated automatic valve, a closed manual valve, a blind; . flange, and a check-valve with flow through the valve - secured. In this condition, the remaining:0PERABLE_drywell - l isolation valve is adequate to. perform the isolation. l function. .However, the.overall: reliability is reduced-because.a single failure in the'0PERABLE drywell isolation valve could result in a loss of drywell-isolation. The s 8' hour Completion Time is acceptable,2 since if the drywell ; design bypass leakage A/v'Tc of 1.0 ft were exceeded,

                                                               . ACTIONS Note 4 will ensure appropriate conservative actions                                 q are implemented. In addition, the Completion Time is                                          :
                                                              - reasonable, considering the time required to isolate the                                      ,

penetration and the relative importance of; supporting ~ drywell OPERABILITY during MODES 1, 2,- and 3.

                                                                                                                          ~        '

For affected penetration flow paths that have been! isolated j in accordance with Required Action A.1, the.affected ; penetrations must be verified to be . isolated on a periodic . l basis._ This is necessary to ensure that drywell penetrations that are required to be isolated following an-accident, and are no longer capable of being-automatically ' isolated, will be isolated should an event occur. This : Required Action does not require any.. testing or device > manipulation; rather, it involves verification that those devices outside drywell and capable of potentially being > mispositioned are in the correct position. Since these devices _.are inside primary containment, the time period

   .                                                            specified as " prior to' entering MODE 2 or 3 from MODE 4, if not performed within the. previous 92 days," is based on engineering judgment and;is considered reasonable in view of the inaccessibility'of the devicesland'other~ administrative controls that will ensure that misalignment is an unlikely possibility. Also, this Completion Time is consistent with the Completion Time specified for PCIVs in LC0 3.6.1.3,
                                                                " Primary Containment Isolation Valves (PCIVs)."

Required Action A'.2 is modified by 'a No'e' that applies to j isolation devices . located in high radiation areas and allows' them to be. verified by use of administrative controls.

                                                             ~ Allowing verification by administrative controls 11s                                           i considered acceptable, since access to these areas is'
                                                                                                                                              ~

typically restricted. Therefore, the probability-of misalignment, once they have been verified to be in the proper position, is low. (continued) .- RIVER BEND B 3.6-132 Revision No. O LATER-a d- _e-____m__ ---_._ _ .-,., , = _ - y , g p ,m- 3%,t- --y- se y--

y ~

                                                                                     'Drywell Isolatien Valves                 :

B 3.6.5.3 I n ~ i BASES ~

i ACTIONS M :

                        -(continued)                                     ..
                                              ~With one or more penetration flow paths with two drywell                       1
          ,.                                    isolation' valves inoperable, the affected penetration flow -                 .i path must be isolated. The method of' isolation:must. include                    ;

the use of at least one isolation barrier that cannot be i adversely affected by a. single active failure. . Isol ation barriers that meet this criterion are a closed and ' de-activated automatic valve, a closed mann! valve, a blind - ! flange, and a check valve with flow through'the valve ; secured. The 4 hour Completion Time is acceptable, since ifE a the drywell design bypass leakage Agl of 1.0' ft were 8 exceeded, ACTIONS Note 4 will ensure appropriate i conservative. actions are implemented. The Completion Time ;

                                              .is . reasonable, considering the time required to isolate the                  -i penetration, and'the probability of a DBA, which requires                        j the drywell isolation valves to close, occurring during this                     t short time is very low.                                                         j i

C.1 and C.2 l If any Required Action and associated Completion Time;cannot ; be met, the plant must be placed in a MODE in which the LC0 , does not apply. To achieve this status, .the plant must' be .! brought to at least MODE 3 within 12 hours and to MODE'4 within 36 hours. The allowed. Completion Times are reasonable, based on operating experience, to reach the , required plant conditions from full power conditions in an ; orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.5.3.1 i REQUIREMENTS Each 24 inch drywell purge isolation valve is required to be verified sealed closed at 31 day intervals. This . .

Surveillance is required since the drywell purge isolation ' valves are not qualified to close under accident conditions. This SR is designed to ensure that a gross breach-of drywell is not caused by an inadvertent or. spurious drywell purge , isolation valve opening. Detailed analysis of these 24 inch " drywell purge valves failed to conclusively. demonstrate their ability to close during a LOCA in time to support-drywell OPERABILITY. Therefore, these valves'are required to be in sealed closed position during MODES 1, 2, and 3. These 24 inch drywell purge valves thatare sealed closed'. must have motive power to the valve operator removed. This

                                                                                                         -(continued) y RIVER BEND                                      B 3.6-133                   Revision No. O LATER

I Drywell Isolation Valves : B 3.6.5.3 f BASES ' SURVEILLANCE SR 3.6.5.3.1 (continued) i REQUIREMENTS . can'be accomplished by de-energizing the source of electric ;

 '                     power or removing the air supply to the valve operator. In        i this application, the term " sealed" has no connotation of leakage within limits. The Frequency is based on purge valve use during unit operations.

SR 3.6.5.3.2 This SR ensures that the primary containment /drywell hydrogen mixing isolation valves are closed as required or, if open, open for an allowable reason for limited periods of time. This SR has been modified by a Note indicating the SR is not required'tc be met when the primary containment /drywell hydrogen mixing inlet or outlet valves are open for pressure control, ALARA or air quality considerations for personnel entry, or Surveillances or special testing of the hydrogen mixing system that require the valve to be open. The 31 day Frequency is consistent-with the valve requirements discussed under SR 3.6.5.3.1. SR 3.6.5.3.3 ' This SR requires verification that each drywell isolation manual valve and blind flange that is required to be closed during accident conditions is closed. The SR helps to ensure that drywell bypass leakage is maintained to a minimum. Due to the location of these devices, the Frequency specified as " prior to entering MODE 2 or 3 from l MODE 4, if not performed in the previous 92 days," is appropriate because of the inaccessibility of the devices and because these devices are operated under administrative controls and the probability of their misalignment is low. 4 Two Notes are added to this SR. The first' Note allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable since access to these areas is typically restricted during MODES 1, 2, and 3. Therefore, the probability of misalignment of these devices, once they have been verified to be in their proper position, is low. A second Note is included to clarify that the drywell ; isolation valves that are open under administrative controls are not required to meet the SR during the time that the devices are open. (continued) RIVER BEND B 3.6-134 Revision No. O LATER

3 Drywall Isolation Valves i

                                                                                                  .B 3.6.5.3'                 ,

BASES

SURVEILLANCE SR 3.6.5.3.4 ; L REQUIREMENTS . ' ' (continued) Verifying that the isolation time of each power operated and i

     ..                            each automatic drywell. isolation valve is within limits:is                                '

required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a' time period.less 3 than or equal to that assumed in the safety analysis. . The ' isolation time and Frequency of.this SR'are in accordance. L with the Inservice Testing Program. .j I ! SR 3.6.5.3.5 Verifying that each automatic drywell isolation valve closes s on a drywell isolation signal is required'to prevent bypass , leakage from the drywell1following a DBA. This.SR ensures each automatic drywell isolation valve will- actuate to its isolation- position on a drywell isolation signal'. .The LOGIC . SYSTEM FUNCTIONAL TEST in SR 3.3.6.1.6 overlaps this SR to t provide complete. testing of the safety function. The. . 18 month Frequency is based on'the need to' perform this ! Surveillance under the conditions that apply during a plant 5 L outage and the potential for an unplanned transient if the . _ Surveillance were performed with the reactor at power, since. ! isolation of penetrations would_ eliminate cooling water flow. ' and disrupt the normal operation of many critical 3 components. 0perating experience has shown these components l usually pass this Surveillance when performed at the 18 month Frequency. .Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. ' l REFERENCES 1. USAR, Section 6.2.4.

2. USAR, Table 6.2-40.

i i l ! r i RIVER BEND B 3.6-135 Revision No. O LATER

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

Drywell Pressure B 3.6.5.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.5.4 Drywell Pressure BASES BACKGROUND Drywell-to-primary containment differential pressure is an assumed initial condition in the analyses that determine the primary containment thermal hydraulic and dynamic loads during a postulated loss of coolant accident (LOCA). If drywell pressure is less than the primary containment airspace pressure, the water level in the weir annulus will increase and, consequently, the liquid inertia above the top vent will increase. This will cause top vent clearing during a postulated LOCA to be delayed, and that would increase the peak drywell pressure. In addition, a negative drywell-to-primary containment differential pressure could result in overflow over the weir wall. The limitation on negative drywell-to-primary containment differential pressure ensures that changes in calculated peak LOCA drywell pressures due to differences in water level of the suppression pool and the drywell weir annulus are negligible. The limitation on positive drywell-to-primary containment differential pressure helps ensure that the horizontal vents are not cleared with normal weir annulus water level. APPLICABLE Primary containment performance is evaluated for the entire SAFETY ANALYSES spectrum of break sizes for postulated LOCAs. Among the inputs to the design basis analysis is the initial drywell internal pressure (Ref. 1). The initial drywell internal pressure affects the drywell pressure response to a LOCA (Ref.1) and the suppression pool swell load definition (Ref. 2). Additional analyses (Ref. 3) have been performed to show that if initial drywell pressure does not exceed the negative pressure limit, the suppression pool swell and vent clearing loads will not be significantly increased and the probability of weir wall overflow is minimized. Drywell pressure satisfies Criterion 2 of the NRC Policy Statement. (continued) RIVER BEND B 3.6-136 Revision No. O LATER

Drywell Prassure B 3.6.5.4 BASES (continued) LCO. A limitation on the drywell-to-primary containment , differential pressure of a: -0.3 psid and :s; 1.2 psid is ' required to ensure that suppression pool water is not-forced ' over the' weir wall, vent clearing does not occur during . normal operation, containment conditions'are consistent with : the safety analyses, and LOCA drywell pressures and pool swell loads are within design values. ' l APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of ' radioactive material to the primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining the drywell-to-primary containment differential pressure limitation is not required in MODE 4 or 5. ACTIONS A.1 With drywell-to-primary containment differential pressure- i not within the limits of the LCO, it must be restored within ! I hour. The Required Action is necessary to return operation to within the bounds of the safety analyses. The - 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.5.1, "Drywell," which requires that the drywell be i restored to OPERABLE status within I hour. B.1 and B.2 ' If drywell-to-primary containment differential pressure cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. . To achieve this status,-the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in.an orderly manner and without challenging plant systems. i (continued) RIVER BEND B 3.6-137 Revision No. O LATER

b

                                                                                                                              -l
                                                                                                                                 )

Drywell' Pressure l Si B 3.6.5.4 j ~ BASES.(continued) > SURVEILLANCE SR 3.6.5.4.1L !

                - REQUIREMENTS                                                                                                   !

This SR provides assurance that the limitations on- i

                                      'drywell-to-primary containment differential pressure stated '                              '

in the LCO are met. The 12 hour Frequency of this SR was ; developed, based-on operating experience related to trending i of drywell pressure variations during the applicable MODES ! and to assessing proximity to the specified LCO differential : pressure limits. Furthermore, the 12 hour Frequency is. i considered adequate'in view of other indications available . ' in'the control room, including alarms, to alert the operator !

                                      .to an abnormal 'drywell pressure condition.                    ,

j RFFERENCES 1. 'USAR, Section 6.2.1. l

2. USAR, Section 3.8.

3. -USAR, Section.6.2.1.1.3. ,

I i i i

                                                                                                                               .j
                                                                                                                                 \

l ~ RIVER BEND .B 3.6-138 ' Revision No. O LATER i

 .-.     -   :.           -      -- -           . . .     .    - .-        -- - -             . -, -          -     - - - - < -- l

Drywell Air Temperature B 3.6.5.5 B 3.6 CONTAINMENT SYSTEMS , B 3.6.5.5 Drywell Air Temperature o BASES BACKGROUND The drywell contains the reactor vessel and piping, which add heat to the airspace. Drywell coolers remove heat and ; maintain a-suitable environment. The drywell average air temperature affects equipment OPERABILITY, personnel access, , and the calculated response to postulated Design Basis Accidents (DBAs). The limitation on drywell average air temperature ensures that the peak drywell temperature during ' a design basis loss of coolant accident (LOCA) does not exceed the design temperature of 330*F. The limiting DBA for drywell atmosphere temperature is a small steam line break, assuming no heat transfer to the passive steel and concrete heat sinks in the drywell. APPLICABLE Primary containment performance for the DBA is evaluated for' SAFETY ANALYSES the entire spectrum of break sizes for postulated LOCAs inside containment (Ref. 1). Among the inputs to the design basis analysis is the initial drywell average air temperature. Increasing the initial drywell average air temperature could change the calculated results of the design bases analysis. The safety analyses (Ref.1) assume an initial average drywell air temperature of 145'F. This limitation ensures that the safety analyses remain valid by maintaining the expected initial conditions and ensures that the peak LOCA drywell temperature does not exceed the maximum allowable temperature of 330*F. The consequence of exceeding this design temperature may result in the degradation of the drywell structure under accident loads. Equipment inside the drywell that is required to mitigate the effects of a DBA is designed and qualified to operate under environmental conditions expected for the accident. Drywell average air temperature satisfies Criterion 2 of the NRC Policy Statement. LCO If the initial drywell average air temperature is less than or equal to the LC0 temperature limit, the peak accident temperature can be maintained below the drywell design (continued) RIVER BEND B 3.6-139 Revision No. O LATER

Drywell Air Temperature B 3.6.5.5 BASES ; LCO temperature during a DBA. This ensures the ability of the (continued) drywell to perform its design function. l APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to the primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining drywell average air temperature within the limit is not required in MODE 4 or 5. ACTIONS A.1 When the drywell average air temperature is not within the limit of the LCO, it must be restored within 8 hours. The .i Required Action is necessary to return operation to within ' the bounds of the safety analyses. The 8 hour Completion Time is acceptable, considering the sensitivity of the analyses to variations in this parameter, and provides j sufficient time to correct minor problems. B.1 and B.2 ! If drywell average air temperature cannot be restored to within limit within the associated Completion Time, the - plant must be brought to a MODE in which the LC0 does not l apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times *are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.5.5.1 REQUIREMENTS Verifying that the drywell average air temperature is within the LC0 limit' ensures that operation remains within the limits assumed for the drywell analysis. Drywell air temperature is monitored in all quadrants and at various j elevations. Since the measurements are uniformly i l (continued) l RIVER BEND B 3. 6-140 Revision No. O LATER l

Drywell Air Temperature , B 3.6.5.5 EASES 1 3.6.5.5.1

     . SURVEILLANCE   SR              (continued)

REQUIREMENTS l distributed, an e-i+hmetic average is an accurate I representation of actual drywell average temperature. l The 24 hour Frequency of the SR was developed based on operating experience related to variations in-drywell average air temperature variations during the applicable MODES. Furthermore, the 24 hour-Frequency is considered l adequate in view of other indications available in the control room, including alarms, to alert the operator to an i abnormal drywell air temperature condition. REFERENCES 1. USAR, Section 6.2. l

I I i l

                                                                .                        )

l l l i l

                                                                                       'I l

RIVER BEND. B-3.6-141 Revision No. O LATER I i

CRFA System B 3.7.2 BASES APPLICABLE and 15 (Refs 3 and 4, respectively). The isolation mode SAFETY ANALYSES of the CRFA System is assumed to operate following a loss (continued) of coolant. accident, main _ steam line _ break, fuel handling accident involving handling!recently: irradiated 1 fuel (i.e., fuel that has i occupied p' art fof aferiticaltreactor; core within the previous,11cdays),"and control' rod ' drop ~ accident. The' radiological doses' to control room personnel as a result of the various DBAs are summarized in Reference 4. No single active or passive failure will cause the loss of outside or recirculated air from the control room. The CRFA System satisfies Criterion 3 of the NRC Policy Statement. LCO Two redundant subsystems of the CRFA System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem. Total system failure could result in a failure to meet the dose requirements of GDC 19 in the event of a DBA. The CRFA System is considered OPERABLE when the individual components necessary to control operator exposure are OPERABLE in both subsystems. A subsystem is considered OPERABLE when its associated:

a. Fan is OPERABLE;

b. HEPA filter and charcoal adsorber are not excessively restricting flow and are capable of performing their filtration functions; and

c. Heater, demister, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

In addition, the control room boundary must be maintained, : including the integrity of the walls, floors, ceilings, ductwork, and access doors. 1 APPLICABILITY In MODES 1, 2, and 3, the CRFA System must be OPERABLE to control operator exposure during and following a DBA, since i the DBA could lead to a fission product release. ' In MODES 4 and 5, the probability and consequences of a DBA are reduced due to the pressure and temperature limitations l in these MODES. Therefore, maintaining the CRFA System (continued) RIVER BEND B 3.7-11 Revision No. O LATER ! 1 l 1

i i l CRFA System ; B 3.7.2 E BASES 1 APPLICABILITY OPERABLE is not required in MODE 4 or 5,.except for the- l (continued) following situations under which significant radioactive ' releases can be postulated:

a. During operations with a p.o,tential .for draining the reactorvessel(0PDRVs);[andy

b. Ort;; COES ^'.I5"".II^"$; Zd eb. During movement of redstifjl irradiated fuel assemblies '

i. H in the primary or seEbndaFy containment.T6 0seito radiosti &EdeEayfitheICRFAT$ystindi sTsn1)f requi fidi:ts " L beLOPCRA8LEideringifuekhsadlingiinv61ving; tia fuehthatthisfaccup handling"ied recentlyJi_rradiatodifeelR(ico%peMthin3hilpreviousill; PartToffaicriticalsreactsr days)7 i i ACTIONS A.1 ' With one CRFA subsystem inoperable, the inoperable CRFA - subsystem must be restored to OPERABLE status within 7 days. With the unit in this condition, the remaining OPERABLE CRFA l subsystem is adequate to perform control room radiation i protection. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in loss of CRFA System function. The 7 day Completion Time-is based on the low probability of a DBA occurring during this time period, and that the remaining subsystem can provide the required capabilities. B.1 and B.2 ! In MODE 1, 2,- or 3, if the inoperable CRFA subsystem cannot be restored to OPERABLE-status within the' associated -i Completion Time, the unit must be placed in-a MODE that minimizes risk.. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit' systems. 1 (continued) l RIVER BEND B 3.7-12 Revision No. 0 LATER

                -     .    ...     .           -          -._       . .-.    -     .    .-         c.

l CRFA System _ ! ~ B 3.7.2 ' BASES ACTIONS C.I. C.2.1; anW C.2.2. : d C.2.3

 ~*
         -(continued)                                                                                   .

The Required Actions:of Condition-C are modified by a Note i indicating _that LC0 3.0.3 does not apply. If moving 1 recently or' 3,~the$fuel irradiated fuelisassemblies movement independentwhile in MODE 1,. 2, of reactor. I operations. Therefore, inability to suspend movement of recintlijirradiated fuel assemblies. is not sufficient reason ; tF require a reactor shutdown. ' During movement of i;idintly] irradiated fuel-assemblies in I the primary or secondaFf~ containment, hM ; CORE i ALTEPf.TIONS, or during OPDRVs, if the inoperable CRFA subsystem cannot be restored to OPERABLE status within the: j i required Completion Time, the OPERABLE CRFA subsystem may be-placed in the emergency mode. This action ensures that the remaining subsystem is OPERABLE, that no failures.that would prevent automatic actuation will. occur, and that any active failure will be readily detected. An alternative to Required Action C.1 is to immediately , suspend activities that present a potential for releasing. t radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. 8 Ifapplicable,COREALTEPf.TIONE::dmovementofFeiently irradiated fuel assemblies in the primary and seE6iidaF9" containment must be suspended immediately. Suspension of " these activities shall not preclude completion of movement of a component to a safe position. Also,' if applicable, j actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue'until the OPDRVs are suspended. D.1 1 If both CRFA subsystems are inor. cable in MODE 1, 2, or 3, the CRFA System may not be capab a of performing the intended function and the unit is in a condition outside of the accident analyses. Therefore, LC0 3.0.3 must be entered immediately. 1 , i fcontinued) i RIVER BEND B 3.7-13 Revision No. O LATER-

i l CRFA System ; B 3.7.2 ' l BASES ACTIONS E.lr~a'nd E.2. =d E.3 I (continued) During movement of fedntly] irradiated fuel assemblies in the primary or secondary containment, d # ng CORE

                  ^LTERf.TIONS, or during OPDRVs, with two CRFA subsystems inoperable, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

If applicable, CORE ?,LTEPf,TIONS =d movement of redestli l irradiated fuel assemblies in the primary and sedoridari ' containment must be suspended immediately. Suspension of these activities shall nod preclude completion of movement of a component to a safe pasition. If applicable, actions must be initiated immedf 2ely to suspend OPDRVs to minimize the' probability of a vessel draindown and subsequent potential for fission product release. Actions must. continue until the OPDRVs are suspended. 1 SURVEILLANCE SR 3.7.2.1 ' REQUIREMENTS This SR verifies that a subsystem in a standby mode starts I on demand and continues to operate. Standby systems should ' be checked periodically to ensure that they start and ; function properly. As the environmental and normal l operating conditions of this system are not severe, testing l each subsystem once every month provides an adequate check l on this system. Monthly heater operation dries out any i moisture accumulated in the charcoal from humidity in the j ambient air. Systems with heaters must be operated for ! a 10 continuous hours with the heaters energized to demonstrate the function of the system. Furthermore, the . 31 day Frequency is based on the known reliability of the I equipment and the two subsystem redundancy available. (continued) l 1 i l RIVER BEND B.3.7-14 Revision No. O LATER I i

Control Room AC System B 3.7.3 BASES (continued) LCO Two independent and redundant subsystems of the Control Room AC System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem. Total systeci failure could result in the equipment operating temperature exceeding limits. The Control Room AC System is considered OPERABLE when the individual components necessary to maintain the control room temperature are OPERABLE in both subsystems. These components include the~ cooling coils, fans, chillers, compressors, ductwork,~ dampers, and associated instrumentation and controls. The heating coils are not required for Control Room AC OPERABILITY. APPLICABILITY In MODE 1, 2, or 3, the Control Room AC System must be OPERABLE to ensure that the control room temperature will not exceed equipment OPERABILITY limits. In MODES 4 and 5, the probability and consequences of a Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Control Room AC System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated: a. During operations reactor vessel with a p@otential for draining the (0PDRVs);ja

b. During CDP.E ALTERATIONS; =d e6. During movement of FiEshtlyjirradiated fuel assemblies in the primary or sefoiidhy containment.K0ddita fadioictiWdeEiyRthiiEC66tFsilR6is?ACISistealisifshQ requi sodito) bei 0PERABLEduringifueVhahdli%M hvol ving iie g fuel?that!has handlingtrecentlynirrediathds; fuel!(3oj(hithinj[tW ~

occipiedjpartfofd)jchiticaljdachr; pregiousg1[ days ACTIONS i A.1 With one control room AC subsystem inoperable, the , inoperable control room AC subsystem must be restored to OPERABLE status within 30 days. With the unit in this condition, the remaining OPERABLE control room AC subsystem (continued) RIVER BEND B 3.7-18 Revision No. O LATER

                                                                             -   ^

m .

                                                                                        .l Control Room AC System     J B 3.7.3      ,
                                                                                        .I BASES                                                                             i i

I ACTIONS .A_d (continued) l l 1s adequate to perform the control room air conditioning

function. However, the overall reliability is reduced + because a single failure in the OPERABLE subsystem could result in loss of the control room air conditioning , function. The 30 day Completion Time is based on the low i probability of an event occurring requiring control room ; isolation, the consideration that the remaining subsystem : can provide the required protection, and the availability of ! alternate cooling methods, i B.1 and B.l If bot'.a control room AC subsystems r,n inoperable, the l Contris Room AC System may not be capable of performing its : inter.ded function. Therefore, the control room area ' temperature is required to be monitored once per 4 hours to ; ensure that temperature is being maintained low enough that : equipment in the control room is not adversely affected. ; With the control room temperature being maintained within ; the temperature limit, 7 days is allowed to restore a ' control room AC subsystem to OPERABLE status. These Completion Times are reasonable considering that the control : room temperature is being maintained within. limits, the low probability ~ of an event occuring requiring control room , isolation, and the availability of alternate cooling ! methods. > C.1 and C.2 In MODE 1, 2, or 3, if the control room area temperature ' cannot be maintained s 104*F or if the inoperable control room AC subsystem cannot be restored to OPERABLE status-within the associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status the unit must be placed in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating l experience, to reach the required unit conditions from full ! , power conditions in an orderly manner and without ; challenging unit systems. , l (continued) i RIVER BEND B 3.7-19 Revision No. O LATER

y, --

                                                                                                           -i
                                                                       ' Control _ Room AC System'          !

B 3.7.3 BASES

                                                                                                         'I
  ^

ACTIONS D.1. D.2.1.7sd D.2.2. ::d D.2.3 _ (continued) , C The Required Actions of Condition C are modified by a Note ; indicating that LCO 3.0.3 does not' apply. ; ' If moving recentlyiirradiated fuel assemblies while in i MODE 1, 2,'^or"37 the fuel movement is independent of reactor ; operations. Therefore, inability to suspend movement of- ! rech tly3 irradiated fuel assemblies is not sufficient reason ! to"requife a reactor shutdown. During movement'of foc~entif] irradiated fuel assemblies in , the primary or secondary ~containmentr-dring C^RE ; ALTESTI^"S, or during OPDRVs, if Required Action A.1 cannot. be completed within the required Completion Time, the " OPERABLE control room AC subsystem may be placed immediately in operation. This action ensures that.the remaining i subsystem is OPERABLE, that no failures that would prevent l actuation will occur, and that any' active failure will be .i readily detected. l An alternative to Required Action D.1 is to immediately l suspend activities that pre.mnt a potential ~ for releasing radioactivity that might require.. isolation of the control- ) room. This places the unit in a condition that minimizes i risk. ! J Ifapplicable,COREALTEST!^"!:dmovementoffecstly 1 irradiated fuel assemblies in the primary and seE6hdafy ' ! containment must be suspended immediately.-. Suspension of l these activities shall not preclude completion of movement' i of a component to a safe position. Also, ;if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended. 1 E.1Jiind E.2. : d E.3 During movement of 'recstlMirradiated fuel assemblies in the primary or secondsfy~ containment, d:H :; CORE ALTESTIC"S, or during OPDRVs if.the Required Action and associated Completion Time of Condition B is not met, action must be taken to immediately suspend activities that present a potential for releasing radioactivity that might require _j l (continued) l RIVER BEND B 3.7-20 Revision No. O LATER j

m . 6 a

                                                                                               -i Control. Room AC System        j B 3.7.3-     _j BASES f
                                                                                               ~.

ACTIONS E.1#said E.2. ' rd E.3 (continued) E isolation of the control room. This places-the-unit-in a : condition that minimizes risk.' l If applicab10, C'*E ^!.TE""T"*S-and handling: of recently ! irradiated fuel in the primary or secondary containment must j be suspended immediately. Suspension of these activities j shall' not preclude connletion of movement of: a component to ; a safe position.- Also, if applicable, actions must be 'i initiated immediately to suspend OPDRVs to minimize the ! probability of a vessel draindown and subsequent potential i for fission product release. Actions must continue until l the OPDRVs are suspended.

                                                                                                .l SURVEILLANCE  SR    3.7.3.1 REQUIREMENTS                                                                            -

This SR verifies.that the heat removal capability of the - .j system is sufficient to remove the. control room heat load ! assumed in.the safety analysis. The~SR consists of a' l l -combination of testing and calculation.. The 18 month i frequency.is appropriate since significant degradation'of. ! the Control Room AC System is not expected over this time : period. ! I REFERENCES 1. USAR, Section 6.4. -i l

2. USAR, Section 9.4.1.

l b [$a M E$ N N Il! l l i i i i l

                                                                                               .J RIVER BEND-                         B 3.7-21                 Revision No. O LATER

 >     3                       t j
             ,         1 s:                                                                   .

AC Sources-Shutdown ' ! B 3.8.2- ! B 3.8 ELECTRICAL POWER SYSTEMS ., B 3.8.2 AC Sources-Shutdown ! BASES i BACKGROUND A description of the AC sources"is provided in the Bases for- .! LCO 3.8.1, "AC Sources-Operating." ! APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 ! SAFETY ANALYSES and 5 and during movement of recent1H~ irradiated fuel ! assemblies in the primary contsinment or fuel building i ensures that: J

a. The unit can be maintained in the t 'own or !

refueling condition for extended peruds; j

b. Sufficient instrumentation and control-capability is i available for monitoring and maintaining the unit -

status; and '

c. Adequate AC electrical power is provided to mitigate .

events postulated during shutdown, such as an j inadvertent draindown of the vessel or a-fuel handling - ; acci Dusi. dent thise1Vi% head (teg peeentiflFNidlitedifd j required;te;miti AfteRheedMagiaccidentsstwoohring - i handlin. gMrecent1 iiriFediatedsfusl44Mgf'oeMthatihas !

                                              'occupieUpark Pr*4885Hidayj             ){pitjMlihicotejjit%F"                         !

In general, when the unit is shut down the Technical ! Specifications (TS) requirements ensure that the unit has , the capability to mitigate the consequences of. postulated ' accidents. However, assuming a single failure and . concurrent loss of all offsite or loss of all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are , analyzed .in MODES 1, 2, and 3 have no specific analyses in MODES 4 and 5. Worst case bounding events are deemed not credible in MODES _4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and the corresponding stresses-result in the probabilities of occurrence significantly reduced or eliminated, and minimal consequences. These deviations from DBA analysis assumptions and design (continued) 1 RIVER BEND B 3.8-34 Revision No. O LATER_

AC Sources-Shutdown B 3.8.2 BASES APPLICABLE requirements during shutdown conditions are allowed by the o SAFETY ANALYSES LCOs for required systems. (continued) During MODES 1, 2, and 3, various deviations from the analysis assumptions and design requirements are allowed within the ACTIONS. This allowance is in recognition that certain testing and maintenance activities must be conducted provided an acceptable level of risk is not exceeded. During MODES 4 and 5, performance of a significant number of required testing and maintenance activities is also required. In MODES 4 and 5, the activities are generally planned and administratively controlled. Relaxations from typical MODE 1, 2, and 3 LCO requirements are acceptable during shutdown MODES based on:

a. The fact that time in an outage is limited. This is a risk prudent goal as well as utility economic consideration.

b. Requiring appropriate compensatory measures for certain conditions. These may include administrative controls, reliance on systems that do not necessarily meet typical design requirements applied to systems credited in operating MODE analyses, or both.

c. Prudent utility consideration of the risk associated with multiple activities that could affect multiple systems.

d. Maintaining, to the extent practical, the ability to perform required functions (even if not meeting MODE 1, 2, and 3 OPERABILITY requitements) with systems assumed to function during an event.

In the event of an accident during shutdown, this LC0 ensures the capability of supporting systems necessary to avoid immediate difficulty, assuming either a loss of all offsite power or a loss of all onsite (diesel generator (DG)) power. The AC sources satisfy Criterion 3 of the NRC Policy Statement. LC0 One offsite circuit supplying onsite Class 1E power distribution subsystem (s) of LC0 3.8.8, " Distribution Systems-Shutdown," ensures that all required loads are (continued) RIVER BEND B 3.8-35 Revision No. O LATER

J, jij _ q I 'f:l AC Sources-Shutdown

                                                                                                                     .B 3.8.2-     .it
                  -BASES'                                                                                                          d LCO        .          .
                                              ,  powered from offsite power. An'0PERABLE DG, associated'with (continued)             a Division I or Division'II Distribution-System Engineered-                        -;

Safety Feature (ESF) bus required OPERABLE by LCO 3.8.10, ensures a diverse power source is available to provide .i electrical power support, assuming a loss of the offsite- 1 circuit. Similarly, when the high pressure core spray j (HPCS) is required to be OPERABLE,.a separate offsite j circuit to the. Division III Class 1E onsite electrical power i distribution subsystem, or an OPERABLE Division III-DG, "! ensure an additional source of power for.the HPCS. This additional . source for Division III is not necessarily j required to be connected to be OPERABLE. Either the circuit l required by LCO Item a, or a circuit required to meet LCO i Item c may be connected, with the second source available ' for connection. Together, OPERABILITY of the required i offsite circuit (s) and DG(s) ensure the availability of j sufficient AC sources to operate the plant in a safe manner - -; and to mitigate the consequences of postulated events during .i shutdown (e.g.. fuelreactor recentlRMQtedifiej, handling accidentsQiiviMWlWM).M vessel dFifKdswn The qualified offsite circuit (s) must be capable of maintaining rated frequency and voltage while connected to. their respective ESF bus (es), and accepting required loads during an accident. Qualified offsite circuits are those that are described in the USAR and are part of the. licensing basis for the plant. The offsite. circuit consists of incomina breaker and disconnect.to the' respective preferred statics service transformers IC and ID, the IC and ID prefe m d station service transformers, and the respective circuit path including feeder breakers to all 4.16 kV ESF buses required by LCO 3.8.10. The required DG must be capable of starting, accelerating to rated speed and voltage, and connecting to its respective ESF bus on~ detection of bus undervoltage, and accepting required loads. This sequence must be accomplished within , 10 seconds for DG 1A and DG 18 and 13 seconds for DG 10. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and must , continue to operate until offsite power can be restored to ! the ESF buses. These capabilities are required to be met from a variety of initial conditions such as: DG in standby-j j with the engine hot and DG in standby with the engine at- ! ambient conditions. Additional DG capabilities must be demonstrated to meet required Surveillance, e.g., capability 1 i (continued) i RIVER BEND B 3.8-36 Revision No. O LATER <

                 ~
   .                                                                                            l l

AC Sources-Shutdown i B 3.8.2 ! l BASES I L LC0 of the DG to revert to standby status on an ECCS signal i (continued) while operating in parallel test mode. Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG . OPERABILITY. In addition, proper load sequence operation is an integral part of offsite circuit and DG OPERABILITY since its inoperability impacts the ability to start and maintain I energized any loads required OPERABLE by LC0 3.8.10. l It is acceptable for divisions to be cross tied during shutdown conditions, permitting a single offsite power circuit to supply all required AC electrical power-distribution subsystems. As described in Applicable Safety Analyses, in the event of-an accident during shutdown, the TS are designed to maintain the plant in a condition such that, even with a single failure, the plant will not be in immediate difficulty. APPLICABILITY The AC sources required to be OPERABLE in MODES 4 and 5 and i during movement of irecentipsirradiated fuel assemblies (i M d fselfthstthitfoceupied jaft!6fisistitiss1}fisct65 coreiwithinithep#Foiids"asI0rance or' fuel building that:viousilifduysMi K"thiTfliii

a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel; i

b. Systems needed to mitigate a fuel handling accident l insolvjnpeseitly?}jfadjsted]fjiel are available; ;

c. Systems necessary to mitigate the effects of events ;

that can lead to core damage during shutdown are I available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold' shutdown condition or refueling condition. .

1 The AC power requirements for MODES 1, 2, and 3 are covered in LC0 3.8.1. (continued) RIVER BEND B 3.8-37 Revision No. O LATER i

l AC Sources-Shutdown B 3.8.2 i BASES. (continued). ACTIONS The ACTIONS are modified by.,a Note._ indicating that LC0 3.0.3 I does not apply. If moving recentl Wirradiated fuel . assemblies while in M0M 1,~2'~ ~oF3',' the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of reEentifjirradiated fuel assemblies is not sufficient reason ~to~fequire reactor shutdown. A.1 An offsite circuit is considered inoperable if it is not available to one required ESF division. If two or more i ESF.4.16 kV buses are required per LCO 3.8.10, division (s) ; with offsite power available may be capable of supporting l sufficient required features to allow continuation .of CORE ALTERATIONS, fuel movement, and operations with a potential i for draining the reactor vessel. By the allowance of the J option to declare required features inoperable which are not l powered from offsite power, appropriate restrictions can be : implemented in accordance with the required feature (s) LCOs' ACTIONS. Required features remaining powered from offsite power (even though that circuit may be inoperable due to failing to power other features) are not declared inoperable by this Required Action. - 1 A.2.1. A.2.2. A.2.3. A.2.4 B.1.-B.2. B.3.'and B.4 l 1 With the offsite circuit not available to all required ) divisions, the option still exists to declare all required i features inoperable. Since this option may involve undesired administrative efforts, the allowance for sufficiently conservative actions is made. With the required DG inoperable, the minimum required diversity of AC power sources is not available. It is, therefore required to suspend CORE ALTERATIONS, movement of Hiehtlf,(irradiated fuel assemblies in the ~ primary containment ~~6Ffdel building, and activities that could potentially result in inadvertent draining of the reactor vessel. Suspension of these activities shall not preclude completion of actions to establish a safe consarvative condition. These actions minimize probability of the occurrence of postulated events. It is further required to initiate ; action immediately to restore the required AC sources and to 4 continue this action until restoration is accomplished in order to provide the necessary AC power to the plant safety l

             .       systems.                                                            .

(continued) l RIVER BEND B 3.8-38 Revision No. O LATER

AC Sources-Shutdown B 3.8.2 BASES' l-ACTIONS A.2.1. A.2.2. A.2.3 A.2.4. B.I. B.2. B.3. and B.4

  ,                   (continued)                                                       t The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. -The        .

restoration of the required AC electrical power sources should be completed as quickly as possible in order to ! minimize the time during which the plant safety systems may be without sufficient power. Pursuant to LCO 3.0.6, the Distribution System ACTIONS are , not entered even if all AC sources to it are inoperable, resulting in de-energization. Therefora, the Required Actions of Condition A have been modified by a Note to i indicate that when Condition A is entered with no AC power ' to any required ESF bus, ACTIONS for LCO 3.8.10 must be immediately entered. This Note allows Condition A to provide requirements for the loss of the offsite circuit whether or not a division is de-energized. LC0 3.8.10-provides the appropriate restrictions for the situation involving a de-energized division. C.1 j When the HPCS is required to be OPERABLE, and the additional required Division III AC source is inoperable, the required diversity of AC power sources to the HPCS is not available. Since these sources only affect the HPCS, the HPCS is ! declared inoperable and the Required Actions of the affected Emergency Core Cooling Systems LC0 entered. In the event all sources of power to Division III are lost, ! Condition A will also be entered and direct that the ACTIONS i of LC0 3.8.10 be taken. If only the Division III additional required AC source is inoperable, and power is still i supplied to HPCS, 72 hours is allowed to restore the l additional required AC source to OPERABLE. This is reasonable considering HPCS will still perform its function, absent an additional single failure. SURVEILLANCE SR 3.8.2.1 REQUIREMENTS SR 3.8.2.1 requires the SRs from LC0 3.8.1 that are necessary for ensuring the OPERABILITY of the AC sources in (continuedl RIVER BEND B 3.8-39 Revision No. O LATER i

s: AC Sources--Shutdown B 3.8.2

     . BASES
     . SURVEILLANCE                  SR  3.8.2.1' (continued).

REQUIREMENTS other than MODES 1, 2,:and 3. SR 3.8.1.7 is not required to be met since only one offsite circuit is required'to be OPERABLE. SR 3.8.1.16 is not required to be met because the required OPERABLE DG(s) is not required to undergo periods 1 of being synchronized to the offsite circuit. SR 3.8.1.19 is excepted because starting independence is not required with the DG(s) that is not required to be OPERABLE; Refer ~ to the corresponding Bases for.LCO 3.8.1 for a discussion of each SR. This SR is modified by a Note. -The reason for the Note is-to preclude requiring the OPERABLE DG(s) from being paralleled with the offsite power network or otherwise rendered. inoperable during the performance of SRs, and . preclude de-energizing a required 4.16 KV.ESF bus or disconnecting a required offsite circuit during performance of SRs. With limited AC sources.available,'a single event could compromise both the required circuit.and the DG. It is the intent that these SRs must still be capable of-being met, but. actual pstfot nance'is not-required during periods when the DG.is rego <d to be OPERABLE. REFERENCES None. d i RIVER BEND B 3.8-40 Revision No. O LATER

DC Sources-Shutdown B 3.8.5 B 3.8' ELECTRICAL POWER SYSTEMS

       -B _3.8.5_ - DC Sources-Shutdown l        BASES BACKGROUND                                                                                        A description of the DC sources is'provided in the Bases for LCO 3.8.4, "DC Sources-Operating."

APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the USAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume that Engineered Safety Feature systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the diesel generators, emergency auxiliaries, and control and switching during all MODES of operation. The OPERABILITY of the DC subsystems is conristent with the initial assumptions of the accident analyscs and the requirements for the supported systems' OPERABILITY. The OPERABILITY of the minimum DC electrical power sources during MODES 4 and 5 and during movement of Eetb6tli irradiated fuel assemblies in the primary confalhment or fuel building ensures that: i

a. The facility can be maintained in the shutdown or refueling condition for extended periods; '

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate DC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the accidentsinifolVidhandliridif_ vessel _or a. fuel handling ic6ntifBeradlitsdifneR DueltiT{ radioactive?decafM OCislectricalk powerli sionly requiredstoimitigateifueljhandlinglaccidentsiinvolving

                                                                                                                .handl ing grecentlyd rradi_atedifueln ( D ess fuelt that! has occupiedi p~artij off a icri ticalireactor? core: withi ni~~'the '
                                                                                                                                                            ' ~ ~ ~ ' " ~ ~

prev _i_ous. ~I_l. ids.ys)I~~~~~~~~'~~

                                                                                                                                    -                                                   i The DC sources satisfy Criterion 3 of the NRC Policy Statement.

(continued) RIVER BEND B 3.8-59 Revision No. O LATER

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

r : -

f , (

                                                                                                                   -DC Sources-Shutdown'         l c.
                                                                                                                                       'B 3.8.5    :

,y t r .

                    ., BASES:. . .

(continued). LCO. One DC electrical. power subsystem consisting of one battery, -)

            ,,                                             one battery charger, and the corresponding control equipment and interconnecting cabling supplying. power to the .                               $

associated bus within the division,- associated with Division- 1 I or.II onsite Class 1E DC electrical power distribution; 1 subsystem (s) required by LCO 3.8.10, " Distribution Systems- 1 Shutdown" is required to be OPERABLE. Similarly, when thel ; High Pressure Core Spray (HPCS) system'is required to be- ; OPERABLE, the Division'III DC. electrical power subsystem .: associated with the Division III onsite Class 1E DC . i electrical power distribution subsystem required to be OPERABLE by LC0 3.8.10 is required to be OPERABLE. In addition to'the preceding subsystems required to be

                                                          .0PERABLE, a; Class 1E battery or. battery charger and the-                            d associated control equipment and interconnecting cabling .

capable of supplying power to the remaining-Division I or II-onsite Class IE DC electrical power distribution subsystem (s), when portions of both Division I and II DC electrical power distribution subsystems are' required to be . OPERABLE by LCO 3.8.10. This ensures the availability of ' sufficient DC electrical power. sources to. operate the unit: in a safe manner and..to-mitigate the consequences of postulated events during shutdown-(e.g., fuel handling . accidents linisWiiijliisieht,)p{f~hidniedifsilRand

                                                                                                        ~ ^ ~ ~ ~ ~ ~ ' ^ ~

inadvertent reactorvenel*dfasd6Wn)".T' APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES ~4 and 5 and during movement-of y ~pi[siiilyjirradiated fuel assemblies in the primary containment or fuel buildirg provide assurance that: a Required features to provide, adequate coolant. inventory makeup are available for the irradiated-fuel assemblies in the core'in case of an inadvertent draindown of the reactor vessel;. -l

b. . Required features needed to mitigate a fuel handling accident liW4WihMecentlyisifidMtedifiifel are '

available? eWistiiHT , psiiersiiCoe(Dueltodadhectivei ly$egiairedit 464 hanGliti" accidentnt'aveMaglhand(46 lt$stdiesi partiof((MgigahWjifolii491MM;eccepWii -elMM dayj); ] (continued)

                   ~ RIVER' BEND                                                       B 3.8-60                       Revision No. 9 LATER
                                                    --'                                                                     -y

DC Sources-Shutdown B 3.8.5 BASES APPLICABILITY c. Required features necessary to mitigate the effects of (continued) events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The DC electrical power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.4. ACTIONS The ACTIONS are modified by a Note indicating that LC0 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require reactor shutdown. A.1, A.2.1. A.2.2, A.2.3 and A.2.4 If more than one DC distribution subsystem is required according to LC0 3.8.10, the DC subsystems remaining OPERAGLE with one or more DC power sources inoperable may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, fuel movementMnVolVing recentlyiippadiited! fuel, and operations with a"p6tsntisl fof dFiinincTthfreidtd vessel. By allowing the option to declare required features inoperable with associated DC , power source (s) inoperable, appropriate restrictions are l implemented in accordance with the affected system LCOs' ACTIONS. In many instances this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made susper.d CORE ALTERATIONS, movement of ifebentlys(i.e., to irradiated fuel assemblies, and any activities thit~c6'Uld* result in inadvertent draining of the reactor vessel). Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required DC electrical power subsystems and to continue this action until restoration is i { (continued) RIVER BEND B 3.8-61 Revision No. O LATER

9: 6 E DC Sources--Shutdown -

B 3.8.5 ; t BASES. l i ACTIONS . accomplished in order to provide the necessary-DC electrical !

  , _           (continued)-      power to the plant safety _ systems.

The Completion Time of.immediately is consistent with the - required times'for actions requiring prompt attention. The l restoration of the required DC electrical. power subsystems ! should be completed as quickly as'possible in order to minimize the time during which the plant-safety systems may j be without sufficient power. l SURVEILLANCE SR 3.8.5.1 - REQUIREMENTS SR 3.8.5.1 requires performance.of all Surveillances required by SR 3.8.4.1 through SR 3.8.4.8. . Therefore, see , the corresponding Bases for LC0 3.8.4 for a discussion of-each SR. - This SR is modified by a Nate. The reason for the Note is to preclude requiring the OPERABLE DC sources ^from being ! discharged below their capebility to provide the required l _ power supply or otherwise rendered inoperable during the ; performance of SRs. It is'the intent that these SRs must ! still be capable of being met, but actual performance is not ! required. , REFERENCES 1. USAR, Chapter 6.

2. USAR, Chapter 15. -

l i RIVER BEND B 3.8-62 . Revision No.'0 LATER i

Inverters-Shutdown B 3.8.8 i B 3.8 ELECTRICAL POWER SYSTEMS l B 3.8.8 Inverters-Shutdown BASES BACKGROUND A description of the inverters is provided in the Bases for LC0 3.8.7, " Inverters-Operating." ; APPLICABLE The initial conditions of Design Basis Accident (DBA) and SAFETY ANALYSES transient accident analyses in the USAR, Chaper 6 (Ref.1) and Chapter 15 (Ref. 2), assume Engineered Safety Feature ; systems are OPERABLE. The DC to AC inverters are designed to provide the required capacity, capability, redundancy, and reliability to ensure the availability of necessary power to the Reactor Protection System and Emergency Core Cooling Systems instrumentation and controls so that the fuel, Reactor Coolant System, and containment design limits are not exceeded. The OPERABILITY of the inverters is consistent with the initial assumptions of the accident analyses and the l requirements for the supported systems' OPERABILITY. The OPERABILITY of the minimum inverters to each AC vital I busdurtngMODES4and5,andduringmovementofpedently irradiated fuel assemblies in the primary containment"or fuel building ensure. i.ha+:

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability are available for monitoring and maintaining the unit status; and

c. Adequate power is available to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident W61sthg%sadliMIpscentlyEirradiatadifieMi!Dudits padioactiveldecaydAC?and?0Cfelectficalipower31sion19 vequiredjtoimitigatelfuelshand)ingsaccidentsiinvolving' handling!recently!irradiatedifuel t(i .eMfdell thatthas occupied'partfoffa?critica1geactor%coreiwithin?the~

                                                                 " ~ ~ ~ ~ ~

previo~us[111dayg~'~~~~ > I (continued) i l l RIVER BEND B'3.8-74 Revision No. O LATER

j 1

                                                                           . Inverters-Shutdown B 3.8.8 BASES APPLICABLE      "The inverters were previously identified as part of the SAFETY ANALYSES  Distribution System and, as such, satisfy Criterion 3 of the                l
     ,        (continued)    NRC Policy Statement.

l i LCO .The inverters ensure the availability of electrical power. for the instrumentation for systems required to shut down . the reactor and maintain it in a safe condition after an : anticipated ~ operational occurrence or. postulated DBA~. .The : battery powered inverters provide uninterruptible supply of l AC electrical power to the AC vital buses even if the 4.16 kV safety buses are de-energized. OPERABLE inverters require the associated AC vital bus be powered by the l inverter through inverted DC voltage from.the required Class ' IE battery, or from an internal AC source via a rectifier with the battery available as backup. Thi. (nsures the-availability of sufficient inverter power sources to operate j the plant in a safe manner and to mitigate the consequences- l e.g., fuel handling -l ofpostulatedeventsduringshutdown(diffueijandinadvertent recshtl l accidents reactor jdvolVing[idown)~^.~~flirradf" vesseT'~ dea n

                                                             ~ " ~

ate : l l APPLICABILITY The inverters required to be OPERABLE in MODES 4 and 5 and ) assemblies in the primary containkint~^6F T]uel buildingalso a ! provide assurance that: l l

a. Systems to provide adequate coolant inventory makeup i are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel; b.

Systems needed to mitig_ fuenthathaiioccup'isHpartieffa ate a fuel handl tica1feactorp ggjpfhle[p@oug}}ipaq) are@availabl[e;

c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

Inverter requirements for MODES 1, 2, and 3 are covered in LC0 3.8.7. J (continued) RIVER BEND B 3.8-75 Revision No. O LATER

N~ Inverters-Shutdown B 3.8.8 BASES (continued) ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 l does not apply. Ifmovingrecentif3irradiatedfuel assemblies while in MODE 1, 2, ^oF3, the fuel movement is independent of reactor. suspend movement of recoperations. ently irradiatedTherefore, inabilityisto fuel assemblies notsufficientreasonto^ req]uirereactorshutdown. A.I. A.2.1. A.2.2. A.2.3, and A.2.4 If two divisions are required by LC0 3.8.10, " Distribution Systems-Shutdown,".the remaining OPERABLE inverters may be capable of supporting sufficient required feature (s) to allow continuation of CORE ALTERATIONS fuel movement insolVin # ki'th~d^)g?hindlingireEentlyfivradiktid, si, and operations o t en t i kl" fo F d ra i n i 6i~thi^Yeic t^6F^Ve s s el . By the allowance of the option to declare required feature (s) ) inoperable with the associated inverter (s) inoperable, i appropriate restrictions are implemented in accordance with I the affected required featura(s) of the LCOs' ACTIONS. In i many instances, this. option may involve undesired. ' administrative efforts. Therefore, the allowance for , sufficiently conservative actions is made (i.e., to suspend i CORE ALTERATIONS, movement of %6stly} irradiated fuel assemblies in the primary contiinmst and fuel building, and ! any activities that could result in inadvertent draining of the reactor vessel). Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required inverters and to continue this action until restoration is accomplished in order to provide the necessary inverter power to the plant safety systems. Notwithstanding performance of the above conservative Required Actions, the unit is still without sufficient AC vital power sources to operate in a safe manner. Therefore, action must be initiated to restore the minimum required AC vital power sources and continue until the LCO requirements are rostored. The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required inverters should be completed as (continued) RIVER BEND B 3.8-76 Revision No. O LATER

ry' , l Inverters--Shutdown B 3.8.8 BASES i 1 ACTIONS quickly as possible in order to minimize the time the plant ' (continued)- safety systems may be without power or powered from a constant voltage source transformer. SURVEILLANCE SR 3.8.8.1 ! REQUIREMENTS This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and AC vital buses energized from the inverter. The verification of proper voltage and frequency o!stput ensures that the required power is readily available for the instrumentation connected to the AC vital buses. The 7 day Frequency takes into account the redundant capability of the inverters and other indications available in the control room that alert the operator to inverter malfunctions. REFERENCES 1. USAR, Chapter 6.

2. USAR, Chapter 15.

l I i RIVER BEND B 3.8-77 Revision No. O LATER

f:.. ,

                                                                                               .r Distribution Systems-Shutdown B 3.8.10 83.8 ELECTRICAL POWER SYSTEMS 8 3.8.10 Distribution Systems-Shutdown-
    ~

t BASES BACKGROUND A description of the AC, DC, and AC vital bus electrical power distribution systems is provided in the Bases for LC0 3.8.9, " Distribution Systems-0perating." ' APPLICABLE The initial conditions of Design Basis Accident and : SAFETY ANALYSES transient analyses in the USAR, Chapter 6 (Ref. 1) and ' Chapter 15 (Ref. 2), assume Engineered Safety Feature (ESF)

systems are OPERABLE. The AC, DC, and AC vital bus 4 electrical power distribution systems are designed to l provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power'to. )

ESF systems so that the fuel, Reactor Coolant System, and ' containment design limits are not exceeded. The OPERABILITY of the AC, DC, and AC vital bus electrical power distribution system is consistent with the initial assumptions of. the accident analyses and the requirements for the supported systems' OPERABILITY. The OPERABILITY of the minimum AC, DC, and AC vital bus i electrical power sources and associated power distribution subsystems during MODES 4 and 5 and during movement of recentlyjirradiatedfuelassembliesintheprimary contaihment or fuel building ensures that: j

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit

                     ,,           status; and

c. Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident i rivelvingihindl ii@NebentiffissdiatidIfeeQdiDielts radioactiseTdecay d ACF and !0CielectricaRpoweri t sNiilf requ.iredJtojattjgatelf M lhandp N iaccfdentsd n M ils j (continued) ;

RIVER BEND B 3.8-89 Revision No. 9 LATER

f . l Distribution Systems-Shutdown B 3.8.10 BASES APPLICABLE h'aidlingMecent19]iWadiatedifselS iialdfue17that?has i SAFETY ANALYSESE occupied part offi?criticalsreacto6('coreNithinithe~

                                                                              "' ~ ^ ~

[previousillfd_ays) M " ~ ~ '

                                                                  ' ~ ' ' '

(continued)- - The AC and DC electrical power distribution systems satisfy Criterion 3 of the NRC Policy Statement. LC0 Various combinations of subsystems, equipment, and components are required OPERABLE by other LCOs, depending on the specific plant condition. Implicit in those requirements is the required OPERABILITY of necessary support required features. This LCO explicitly requires energization of the portions of the electrical distribution system necessary to support OPERABILITY of Technical Specifications' required systems, equipment, and components-both specifically addressed by their own LCOs, and implicitly rcquired by the definition of'0PERABILITY. Maintaining these portions of the distribution system energized ensures the availability of sufficient power to operate the plant in a safe manner to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents involVii iM Ebstl ~~~~ and inadvertent reactor vessel'd@riindon)B~hadiat . APPLICABILITY The AC, DC, and AC vital bus electrical power distribution subsystems required to be OPERABLE in MODES 4 and 5 and during movement of recsntly$ irradiated fuel assemblies in the primary containment or fuel building provide assurance that:

a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel;

b. Systems needed to mitigate a fuel handling accident inslving"rebentlilirradiktedifuel are availableT(Due to radi6activeidecayMAC?and50CielsctHealIpowerbis onlyrrequiredito; mitigate'fuelihshdling! accidents"

, i nvol v i ng ; handl ingirecentlyiirradi ated 4 feelithati h'as occupi ed -;part( of t a2 cri ticalkeactodcorej ithi.nj the" previous)ll! days); (continued) ) RIVER BEND B 3.8-90 Revision No. O LATER I l

y . Distribution Systems--Shutdown B 3.8.10 BASES APPLICABILITY c. Systems necessary to mitigate the effects of events (continued) that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit.in a cold shutdown or refueling condition.

The AC, DC, and AC vital bus electrical power distribution subsystem requirements for MODES 1, 2, and 3 are covered in LC0 3.8.9. ACTIONS The ACTIONS are modified by a Hote indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require reactor shutdown. A.I. A.2.1. A.2.2. A.2.3 A.2.4 and A.2.5 Although redundant required features may require redundant divisions of electrical power distribution subsystems to be OPERABLE, one OPERABLE distribution subsystem division may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, FisiitlyliFridistid? fuel movement, and operations with a potintill*f6F~dFlidihg" the reactor vessel. By allowing the option to declare required features associated with an inoperable distribution subsystem inoperable, appropriate restrictions are implemented in accordance with the affected distribution subsystem LCO's Required Actions. In many instances, this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS, movement of Fiseitly]jirradiated fuel assemblies in the primary contilhmeht' and fuel building and any activities that could result in inadvertent draining of the reactor vessel). Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and DC electrical power distribution subsystems and to continue this action (continued) RIVER BEND B 3.8-91 Revision No. O LATER

f. ;

Distributicn Systems--Shutdown' l

B 3.8.10 ,

                  ' BASES                                                                >

q ACTIONS A.I.~'A.2.1. A.2'.2. A.2.3. A.2.4. and A.2.5 (continued) .; I until restoration is accomplished in order to provide the necessary power to the plant safety' systems. 9

Notwithstanding performance of the above conservative-Required Actions, a required residual. heat removal-shutdown .

j cooling (RHR-SDC) subsystem may be inoperable. In this- j case,: Required Actions A.2.1 through A.2.4'do not adequately. : address the concerns relating to coolant circulation and :

                                   " heat removal . Pursuant to LCO 3.0.6, the RHR-SDC ACTIONS                 "

would not be' entered. Therefore, Required Action A.2.5 is ! provided to direct declaring RHR-SDC inoperable, which ; results'in taking the appropriate RHR-SDC ACTIONS. j l The Completion Time of immediately is consistent'with the- 1 4 required times for actions requiring prompt attention. tThe ~ restoration of the required distribution . subsystems should be completed as quickly as possible in' order to minimize the, j! time the plant safety systems may be without. power.. i 1 i SURVEILLANCE SR 3.8.10.1 REQUIREMENTS .

                                  -This Surveillance verifies that the required;AC, DC, and AC                          i vital bus electrical power distribution subsystems' are                             (

, functioning properly, with the buses energized. The i verification of- proper voltage availability on the required a buses ensures that=the required power.is readily available j for motive as well as control functions for critical system .j loads connected to these buses. The 7' day Frequency takes ; into account the. redundant capability of the. electrical . power distribution subsystems, as well as other indications a available-in the control room that alert the operator to ! subsystem malfunctions. j j 2 REFERENCES 1. USAR, Chapter 6. . i

2. USAR, Chapter 15.

a 1 i RIVER BEND B 3.8-92' Revision-No. O LATER. !

'l

\i

                                                       -                                     , , . .__~____ ____}}

ML20087J545

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