ML20077S440
| ML20077S440 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 01/18/1995 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20077S439 | List: |
| References | |
| RTR-NUREG-1434 NUDOCS 9501240095 | |
| Download: ML20077S440 (792) | |
Text
{{#Wiki_filter:- .. _ - _ . (2AR 93-NNQ l l
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1 1 l l RIVER BEND l l l CHAPTER 1 l ATTACHMENT 1: NOT USED ! ATTACHMENT 2: ITS - PSTS COMPARISON DOCUMENT 9501240095 DR 950118 ADOCK 05000458 i PDR f
(ux 93-iaD , ATTACHMENT 2 ITS - PSTS COMPARISON DOCUMENT REVISION 1 CHAPTER 1 REVISED PAGES 2A: MARKUP OF ITS : 2B: NOT USED
4 ( u e 93 1920 i ATTACHMENT 2A ITS - PSTS i COMPARISON DOCUMENT - REVISION 1 MARKUP OF ITS
' - GA K 92-/ +'II) ~
Definitions 1.1 1.1 Definitions CHANNEL CALIBRATION remaining adjustable devices in the etannel. l (continued) Men ver a spnstng e seent A repla c, the ext ' req red inflack e ss cal ration onsist of h < c aring e othe sensi entin nstalle senni elene s with h elene / The CHANNEL CALIBRA" ION may be per"omes oy means of any series of sequential, overlapping, or total ! channel steps so that the entire channel is calibrated. ' CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and states to other indications or status derived from independent instrument ' channels measuring the same parameter. CHANNEL FUNCTIONAL TEST A CH/.NNEL FUNCTIONAL TEST shall be[ eO ":' n ?"- r 's Ehe injection of a simulated- i oractualsignallntothechannelascloseto : the sensor as practicable to verify OPERABILITY, including required alars, interlock, display, and trip functions, and channel failure trips g ;
- b. Sistab e channel (e.g., ssure swi hes and switc contacts -the inj ction of a inulated ,
or a ual sign into-t channel as close to : the ensor as practicab e to verify OP BILITY, including quired al and trip efu ctions. g The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or
/ total channel steps so that the entire channel is tested.
CORE ALTERATION CORE ALTERATION shall be the movement of any g l, g,/ sources,treactivity control componentsg Oorw%
4 /s 3 _ v O J IZZrn inr:: ; ==== withinPthe reactor vessel wbt!c rv is a cans (de Hzm AkfERAT /provic!ed there are no fuel a sembifes in the associated core cell.
9 Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe . position. CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. T3ese cycle spectfic limits shall be determined for each reload cycle in accordance with Specification 0_.' '.1.',.1 Plant /D i operation within these limits s individual Specifications. addressed
,4.. in, @' 6 5 DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration !
of I-131 (microcuries/ gram) that alone would . produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134, . and I-135 actually present. The thyroid dose conversion factors used for this calculation shall i be those listed in fTable III of TID-14844,
/,c7 AEC, 1962, " Calculation of tance Factors for gj Power and Test Reactor Sit J li:::: i,, " -
g., S g,m,,,.-: g ;;;.: = ., :::_." , :; . . :
.:=th::: _.
t
- VERAGE E'shall be e average weighted i p portio /
1 D INTEG IONENER!Y .to the con entration each radi uci de in he /
\
reactor e olant at t time of s ling) of the sum of e average ta and g energia per 1 l
/ ,/y BI disint ration (i MeV) for I iodin s, with ha lives > SP ainute , making u topes, et r than w/ ,
at I ast 95% o the total niodine tivity in k /! l EMERGENCY CORE COOLING th coolant. The ECCS RESPONSE TIME shall be that time interval g/ 1 l SYSTEM (ECCS) RESPONSE from when the monitored parameter exceeds its ECCS l TIME initiation setpoint at the channel sensor until the ECCS equipment is capable of performing its ' safety function (i.e., the valves travel to their (continued)
.T,Te 1.1-3 Rev. O,^^/22/02[
l
h9FN) llcfinitiens 1.1 1.1 Definitions EMERGENCY CORE C0OLING required positions, pum SYSTEM (ECCS) RESPONSE their required values, petc.). discharge T'mes pressures shall include reach TIME diesel generator starting and sequence loading (continued) delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. Crmum.4 J) END OF CYCLE The E0C-RPT SYSTEM RESPONSE TIME shall be that RECIRCULATION PUMPtime interval from inttial [. ',f hTRIP h (EOC-RPT) SYSTEM RESPONSk he associated turbine stop valve ,!'r _- 9"' t- 'y 9 or ' TIME feseMahen the turbine control valve hy t r!!: e!! g .s _ns --__.. u .. u 6. _ _ _ . . . . C/ i_s__iEEEU:IbiFtocompletesuppressioioIthe ' electric are between the fully open contacts of ' the recirculation pump circuit breaker. The response time may be measured by anans of any series of sequential, overlappirg, or total eps so that the entire response time is measu,_
, I i e stta % )
ISOLATION SYSTEM The ISOLATION SYSTEM SE TIME shal1~be that RESPONSE TIME time interval from wh the monitored parameter
@c.: exceeds its isolation tpoint at the channel sensor until the isolation valves travel ta tb r
/.D reauired nositions.fTiuns J sha)f ine)(de (Msek' (gene /storMartpi' J/ '/ ltrhete anslicalde.g 1The anfsequente lostline response time may#1avC be measured by means-ef any series of sequential, '
overlapping, or total steps so that the entire l response time is measured. '
/ LEAKAGE LEAKAGE shall be:
- a. Identified LEAKAGE O 1. LEAKAGE into the drywell such as that from
! ) j',g pump seals or valve packing, that is captured and conducted to a sump or
- collecting tank; or.
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es,Ja . ,t w ai, pesx coda,med paauga. wa f p,)by m, atri,e f continuen) cal 4 [, 0 \ x , f5 1.1-4 Rev. O,9#34482 l
(I.48 9 " W K D Definitions 1.1 .- I.1 Definitions LEAKAGE 2. LEAKAGE into the drywell atmosphere from (continued) sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;
- b. Unidentified LEAKAGE All LEAKAGE into the drytell that is not identified LEAKAGE;
- c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;
- d. Pressure Boundarv LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, -
pipe wall, or vessel wall. INEAR HEAT GENERATION The LHGR shall be the heat generation rate per 1 RATE unit length of fuel rod. It is the integral of F y a (LHGR) the heat flux over the heat transfer area
; _ t: V associated with the unit length. g ;
LOGIC SYSTEM FUNCTIONAL A LOGI :MFUNCTIONALTESTsh)1beatest TEST of allulogic components (i.e., allurelays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuatec 6 to verify OPERABILITY. The ! LOGIC SYSTEM FUNCTIONAL TEST may be performed by l
/, O -
O . means of any series of sequential, overlapping, or i su de wc e u tal system steps so that the entire logic system !
? is tested. )
l
'I MAXIMUM FRACTION The NFLPD shall be the largest value of the 0F LIMITING fraction of limiting power density in the core. \
l O, POWER DENSITY (MFLPD) The fraction of limiting power density shall be 9-g the LHGR existing at a given' location divided by l _ the specified LHGR limit for that bundle type. l (continued) P,w Oul l BWR/.6 SIS l'.1-5 Rev. 0, 99/29f99-mi
{4e DefinitFons 97 /v7[) l 1.1 ! . 1.1 Definitions (continued) MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the coreAfor each Si class of fuel}t' The CPR is that power in the assembly that is calculated by application of the dppropriate Correlation (s) to Cause some point in the assembly to experience boiling transition divided by the actual assembly operating power,. MODE A MODE shall correspond to any one inclusive , combination of mode switch position, average 1 reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in +ha ==" tar vessel . or twe. opf.a AeluTV) OPERABLE-OPERABILITY FA system, subsys , co ,cnent, or device 1 i 03 shall be OPERA 8L n it is capable of perfoming its spectiied safety function (s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary l C5 cly . s ) equipment that are required for the system,
**) subsystem,*M, component, or device to perform {
its specified safety function (s) are also capable of perfoming their related support function (s).
-PuYS!CS TE5T M ICS TESTS sh e those tests perTrued to 7 asure the asental nuclear char ristics of ;
the react core and related ins ntation. These ts are:
', Described in Chapter 4, Initial Test ProgranQ of the F ;
- b. Authorized der the provisions 10 CFR .59; or
- c. Otherwise ap Commission. proved byAfi'e Nuclear Regulatory
,o PRES RE AND TE ERATURE IMITS
/ The P R is t unit ecific doc nt that PORT (P )
/ pro des th reactor essel press re and t eratu limits, neluding h tup and c old n ,
,e g, tes, f period the eu ent reactor essel fl nc These ssure and erature i's g,0
/ /2 7
/
(continued) Q.s GJ eWR/64IS 1.1-6 Rev. O, 7/""/%
(LAC 93-!v2t] (, o J r' lO 297 l 1.1 Definitions [ 4.] a PR 'URE all be de ruined f each uence n(riod in T ERA E LIMI S accordance ith Spec icati 11. =1 Plan R T PTLR) operatfor ithin t se op ting 1 ts is (coninued) address #,.inLC0 .4.11, RCS P ure an X Temperdture (P/ Limit ." RATED THERMAL POWER RTP shall be a total reactor core heat transfer ) (RTP) h rate to the reactor coolant of ja& MWt. REACTOR PROTECTION The RPS RESPONSE TIME shall be that ties interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS 4 TIME trip setpoint at the channel sensor until I de-energiration of the scram pilot valve j solenoids. The response time may be measured by 1 means of any series of sequential, overlapping, or l total steps so that the entire response time is ! measured. , I SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is suberitical or would be subcritical assuming that: -
- a. The reactor is xenon free; ,
- b. The moderator temperature is 68'F; and l
- c. All control rods are fully inserted except for the single control rod of highest reactivity I 40 worth, which is assumed to be fully withdrawn. '
/0/ M s With control rods not capable of being fully
, inserted, i the reactivity worth of these control ! l rods,must be accounted for in the determination of l (LSDMJ STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the i
~
testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, . channels, or other designated components are. l tested during n Surveillance Frequency intervals, i where n is the total number of systems, i subsystems, channels, or other designated components in the associated function. l
\
(continued) . Q.J. LJ l
=/5 S'S' 1.1-7 Rev. O, 00/20/02 I
(1dR 4L7-/YR/3 l l l l l I l RIVER BEND ! l CHAPTER 2 ! i l l l 1 l l l ATTACHMENT 1: NOT USED ATTACHMENT 2: ITS - PSTS COMPARISON DOCUMENT l
~. _ _ _ . . __ .
(4411 93-/4fd , ATTACHMENT 2 . ITS - PSTS COMPARISON DOCUMENT REVISION 1 CHAPTER 2 REVISED PAGES 2A: MARKUP OF ITS 2B: NOT USED l l 4
(1,AR [3-198t) i I l l l l ATTACHMENT 2A i i ITS - PSTS COMPARISON DOCUMENT i REVISION 1 MARKUP OF ITS 9
(L4 93*M h) SLs ; 2.0 2.0 . SAFETY LIMITS (SLs) 2.1 SLs - 2.3.1 Reactor Core SLs 2.1.1.1 With the reactor steam done pressure < 785 psig or core ' flow < 10% rated core flow: THERMAL POWER shall be s 25% RTP. 2.1.1.2 With the reactor steam done pressure a 785 psig and cove flow = 10% rated core flow: g MCPR shall be a 11.07? for two oo j) operation or = }1.0 Q for single operation. recirculationdh recirculatione 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel. ,
)
i 2.1.2 Reactor Coolant System Pressure SL _Y Reactor steam done pressure shall be pe+nse+n>ee 51325 psig.
]
2.2 SL Violations With any SL violation. the following actions shall be completed: 2.2.1 Within I hair, notify the NRC Operations Center, in accordance with 10 CFP 50.72. 2.2.2 Within 2 hours: 2.2.2.1 Restore compliance with all SLs; and a,D 2.2.2.2 Insert all insertable control rods. . A7a. 2.2.3 Within 24 hours, notify the C W ,h nager < T... ...,
\ * ** "7;;id;7.t-Le'.ee. 4;;;;' ;7.5C. C:ffs 0; r;Jt; ;r; A ra-ef'f ed f: Spee!'f estic 5.5.2, "[0ff:f t:8"*"% : dr' t") '
- 4) fe corporakt E4ecsds'Je respoMnk\t. ,a b (ny 4or overall p\ Ant nuclear safeh , ' -
- (continued) river BEAlb
~.;/5 STS ppf 2.0-1 Rev. O, 09/28/92
(LA2 93-NR Q SLs 2.0 2.0 SLs 2.2 SL Violations (continued) 2.2.4 Within 30 days, a Licensee Event Report (LER) shall be prepared p,0 pursuant to 10 CFR 50.73. The LER shall be suh=itted to the NRC ) jg - theGff n; re.n ; , :r 17R.' '- '- ri M _. 5.3.;; . ::M
- l f_i.: ;;rafaganager- _ E_ ant'4and
- - -: " - - - - = ;
/
f>O
"'~ '-{ ' [.'de., corp o rate tueudive. 't-e spn3ih d Qo r overall plant nuclear safe % .
/73 2.2.5 Operation of the unit shall not be resumed until authorized by the NRC.
i 1 I v 0
- s R w 6ed
= /; ';;; 2.0-2 Rev. O, 09/28/92
- , = . - -. - . - - . . -. - - {A 93 F/Ah : Reactor Core SLs B 2.1.1 BASES
~
l BACKGROUND Operation above the boundary of the nuclette boiling regime (continued) could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp : reduction in heat transfer coefficient. Inside the steam film, high cladding tem water (zirconium water)peratures reaction mayare take reached, place. and a cladding This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant. APPLICABLE The fuel cladding must not sustain damage as a result of SAFETY ANALYSES nomal operation and A00s. The reactor core SLs are estsblished to preclude violation of the fuel design criterion that an MCP is to be established, such that at least 99.9% of the fuel rods in the core would not be J 7,o expected to experience the onset of transition bgiling. ! 4 74, A V The Reactor Protection ystem set o er Octor Protection Systgpi (RPS) points Instrumentation"),in (LC0 3.3.1.1, combination with"cTr - <LCOs, are designed to prevent any I enticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THE L POWER O69 level that would result in reaching the MC SL Sb 2.1.1.la' Fuel Claddina Inteer' qv. hr:--! EMtAj g __ _ _ _ _ _ _ _ . m .... GE critical power correlations are applicable for all 'e 6.0 critical power calculations at pressures a 785 psig re . 4 76 ' ' ~ flows a 10% of rated flow. For operation at low pressures Qlow flows, another basis is used, as follows:
~
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be
> 4.5 psi. Analyses (3Ref. 2) show that with,a bundle flow of 28 x 10 lb/hr, bundle pressure ,
drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow
. with
> 28 a' x 10 4.l lb/hr. psi Full driving head scale ATLAS will betest data taken at pressures from 14.7 psia to 800 psia (continued)
BWR/6 STS B 2.0-2 Rev. O, 09/28/92
1 l' [ME %WR] Reactor Core SLs B 2.1.1 BASES ' l APPLICA8LE 2.1.1.3 Reactor Vessel Water Level (continued) SAFETY ANALYSES active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action. 3,o SAFETY LIMITS The reactor core SLs are established to protect the 4 76 integrity of the fuel clad barrier to the release of
- je cede - radioactive materials to the environs. SL 2.1.1.1 and pf,4,,y e/,fo va/c [ SL 2.1.1.2 ensure that the core operates within the fuel l
*gg s, ,,, /wa design criteria. SL 2.1.1.3 ensures that the reactor vessel l
les ,, ./ water level is greater than the top of the active irradiated aln g,3 g,, v m _......... ,_ . ...-.-.. .-.. _ _ - - -_ _ , _ l APPLICA81LITY SLs 2.1.1.1. 2.1.m.2 2.1.1.3 are anolicable in all own t ue dd e SAFETY LIMIT LZJ VIOLATIONS If any SL is violated, the NRC Operations Center must be notiff within I hour, in accordance with 10 CFR 50.72 (Ref. . LLZ t Exceeding an SL may cause fuel damage and create a potential for radioactive releases in ss of 10 CFR 100, " Reactor r/y Site Criteria," limits (Ref Therefore, it is required to insert all insertable con rods ar:i restore compliance with the SL within.2 hours. The 2 hour Completion Time ensures that the operators take prompt remedial action and
. also ensures that the probability of an accident occurring ;
riuring this period is minimal. 1 (continued) BWR/6 STS B 2.0-6 Rev. O, 09/28/92
4K93-/9'Rl} Reactor Core SLs B 2.1.1 BASES SAFETY LIMIT L2J - VIOLATIONS (continued) IfanySLisviolated,thefp.inw
.diew nuciear aiens enu ui..gr;pri.te
.itp'shall ;;r.i:r ::::
be notified ::tOct within
/ 24 hours. Tie 24 hour perios provides time for plant operators and staff to take the appropriate immediate action
#' " '"'ionmanagement.
(47, ~ _
=
~ ' '
, Generd Mov o ger Pfad OpidS A1Ae.- Vice 0'A**! '
oparti,w la2.d If any SL is violated, a Licensee Event Report shall be arenared and submitted within 30 days to the NR Y he i! T t h accoraancy witn 10 CFR 50.73 (Ref. c ,,t gg f e ofl 4 nll r so ac socmd fo 7'ke I Lza -- If any SL is violated, restart of the unit shall not connence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation. REFERENCES 1. 10 CFR 50 Appendix A GDC 10. h 2. NEDE-24011-P-A,9(latest approved revision). f -3. XN-N 524(A), Nevis10n i Nov. e r ib3. D 3 k 10 CFR 50.72. 'Cwreial 5/aeh?.54,0,<[
~
y f 10 CFR 100. /9# C " !' f [ 10 CFR 50.73. / > h BWR/6 STS B 2.0-7 Rev. O, 09/28/92 l I
(L $ 9.T- M Q RCS Pressure SL < 8 2.1.2 8 2.0 SAFETY LIMITS (SLs) 8 2.1.2 Reactor Coolant System (RCS) Pressure SL c 8ASES BACKGROUND h Theagainst SL onoverpressurization. reactor steam done pressure protects the>(RCSV In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphers. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. According to 10 CFR 50, Appendix A, GDC 14. " Reactor Coolant Pressure Boundary," and GDC 15, " Reactor Coolant System Design" (Ref. 1), the reactor coolant pressure boundary (RCPS shall be designed with sufficient margin to ensure that the) design , conditions are not exceeded dur<ng normal operation and ticipated operational occurrences [A00s).10er1 mut3 1 i ane z, he re tor vessel water i abov the to of the gdtive fu(fevei o be a is getuiI to prov)6e core soling k ca 111ty. 7 J During norb1 operation and A00s, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial ! operation when there is no fuel in the co Any further 00.5 hydrostatic testing with fuel in the core cone under m% i LC0 3.10.1, ' Inservice Leak and Hydrostati esting _ be Operation." Q---f components shall be pressure testFollowing inception of uni in accordance with the requirements of ASME Code, Section I (Ref. 3). Ov RCPS iration of the RCS could result in a breach of cladding, failure, fissiIf this- occurred ucts enuldinenuar conjunction th. with a fuel y, g inment atmosphere j7 06 - - - - - - - - D specified in 10 CFR m : e limits 100, " Reactor ]
, Site Criteria" (Re . 4p.
\
reduciq b nmbr ok gro+eADe $Arrier4 dup *5 4 fo remt roliwwe rekses b excadig At E
- Ik -.A (continued) 4WR/4-44 8 2.0-8 Rev. O, 09/28/92 !
1 l
h.42 93-Nk) RCS Pressure SL B 2.1.2 BASES SAFETY LIMIT LL2 VIOLATIONS (continued) Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CTR 100, " Reactor Site Criteria." limits 3 (Ref.4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours. The 2 hour Completion Time ensures that V operators takapt remediay1ictionA, an.f cdse e,rures dd Me. OCl,I j/ Qhas;/,yy J ,,, ace s,,9 ,ec g, ,p,,,, g,3 ,,g, j
-gs /, /
Ifhaany SL is vjolated. thalt neefriatMnioraufifaamarh'M 1Eif mar 1Hant and"the 1ty/5ha 1 be notified w' th1n 24 hours. The 24 hour period provides time for plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to the senior management. f&exe'd Mortger Plant Operddu a 10 t re. i VIEeV/es'Not-Ose'a4 m
), D / If any SL is violated, a Licensee Event Re mrt sh _be
# y, prepared and sa hitted within in dawn to t:e th se r t s n -
accordance w'th 10 crn 3v./3 (Ref. 5).2 g ~ gf yk (redoNsbaNabo be- .sub m f
-fo Ae g g ~
If any SL is violated, restart of the unit shall .not commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to nonnal operation. REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.
- 2. ASME, Boiler and Pressure Vessel Code, Section (fW w - ++w ,
(continued) BWR/6 STS B 2.0-10 Rev. O, 09/28/92
OAav.s-ino d RIVER BEND SECTION 3.0 l ATTACHMENT 1: CTS-PSTS COMPARISON DOCUMENT ATTACHMENT 2:ITS-PSTS COMPARISON DOCUMENT
( L e< c iu o ATTACHMENT 1 ITS - PSTS COMPARISON DOCUMENT REVISION 1 SECTION 3.0 REVISED PAGES 1 A: MARKUP OF CTS 18: NOT USED - l 1C: NOT USED l l
2aJ 4a - -- 4 * [AR93-ivRt) ATTACHMENT 1 A CTS - PSTS COMPARISON DOCUMENT REVISION 1 MARKUP OF CTS
93-/vah INSERT 1D LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. Exceptions to this Specification are stated in the individual Specifications. These exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered allow unit operation in the MODE or other specified condition in the Applicability only for a limited period of time. 9 LCO 3.O'{ i.s onfy cappbeobb hr edry 3* - In fo ce P10DL or o rber .spee;bh cowSYioo k t e. $ Aff buObYy '"
} />10DE S I,2,ou$ 2.
INSERT 9holvi R8V@@ @@[R5) _ MO .@-lL191_ 404 M
QAR 93-14th INSERT 2D SR 3.0.4 Entry into a MOl:E or other r,pecified condition in the Applicabilj ty of an LCO shall not be made unless the LCO's Surveillances have been met within their specified Frequency. This provision shall not prevent entry into - MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 30 L R '3. 0. 4 is coly ay bhalfe he coby iAb a noon oc sh, yac:h'f co.h,6 ,a ibe A y kha li k /y on t?? ODES /,2,a J 3, 1 i t INSERT ' RIVER BEND 3/4 0-2 (2) _ 10/1/93
CLAR 93TE D ATTACHMENT 2 l ITS - PSTS COMPARISON DOCUMENT l REVISION 1 SECTION 3.0 REVISED PAGES 2A: MARKUP OF ITS 2B: NOT USED
, . _ _ , _ _ , , _ . , , . . . . . , _,,.,y,- _ _ . ..,_m
hR 93-/VA3 ATTACHMENT 2A ITS - PSTS COMPARISON DOCUMENT REVISION 1 MARKUP OF ITS
-, m v-.
() A E 9 3 - N R 1 3 LCO Applicability 3.0 3.0 LCO APPLICA8ILITY 3.o 4 C2 sf L0 0. continued) specified conditions in the Appitcability that are required to comply with ACTI - - - g 4co 3.o.4 a u/7 Exceptions to this Specification arif stated in the ~
,p /,6 4 & ,,d,y individual Specifications. These exceptions allow entry
' a
- yogg , , . into MODES or other specified conditions in the
# 74 #
""pI# Ap licability when the associated ACTIONS to be entered a d' 4"'" 7' 4y/'d'I'VY al ow unit operation in the MODE or other specified
;,. noors 1,2 3.
condition time. in the Applicability only for a limited period of LC0 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perfom testing required to demonstrate its OPERA 8ILITY or the OPERA 8ILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERASILITY. LCO 3.0.6 When a supported system LC0 is not met solely due to a support system LC0 not being met, the Conditions and ! i
-=
Required Actions associated with this supported system are
'[g,f,lo bO not required to be entered. Only the support system LCO '
ilgt ' - ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, _( - additional evaluations and limitations may be required in s accordance with SpecificatTo
, Detemination Program (SFDP)@." ' Safety If a loss Function of safety function
[ is determined to exist by this program, the appropriate Conditions and Required Actions of the LC0 in which the loss ! of safety function exists are required to be entered. ) When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. (continued)
~ "/6 ST3- 3.0-2 Rev. O, C/22/2-RiverGed
I-(L4A 93-syft) SR Applicability 3.0 3.0 SR APPLICABILITY SR 3.0.3 (continued) When the Surveillance is perfomed within the delay period 4 and the Surveillance is not us.t. the LC0 must immediately be declared not met. and the annlicable condition (s) must he ' QueeAlately/upon philure to/ meet tid Survei/ lance.ontered / i l I SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LC0 shall not be made unless the ' Surveillances have been met within their spe::ified e% = Frequency. This provision shall not prevent g p
*wm MODES or other specified conditions in rH4 rf with m(ACT:0Nby - r" ~9 (the A pphe LhN -N4 I L a,,_r,q ;,,J %_ ca.o %)
boe et are pat 4: of a
}%%e & We A.
g9 .$ R 3 *
*l t.s 9 y cs,epb eab c hr enhy' / A
~ ~ ~
3l pyngg ,, ,gg ,p ee;f,Q co 0 h b in 1ke Ayllcd;$/y u n100ES /,2,a d 3, l
' I6 SU 3.0-5 Rev. O, O/20/^t l
_=--. . .. ._ - & 93-MQ LCO Applicanility B 3.0
.nus LC 0 ontinued)' assemblies in the associated fuel storage pool."
this LC0 can be applicable in any or all MODES. Therefore, and the Required Actions of LC0 3.7JS are not met wh MODE placing the 1,unit 2, in ora3shutdown there iscondition. no safety benefit to be g OP,~ The Required
' Jtion of LC0 3.7.2 of " Suspend movement ae fuel of irrad appropriate Required Action to complete actions of LC0 3.0.3.
individual Specifications.These exceptions are addressed in the .
.q cd ' - !
- _ _ , t LC0 3.0.4 )
' L(5W D 4'g , - LCO 3.0.4 establishes limitations on changes in MODES o
, b liu M +3 la.b other specified conditions in the Applicability when an LC is not set.
[ ',' It precludes placing the unit in a M M00E or other specified condition,when the following exis An%sM desire 3(
, , L'+, u ea, ge ed') j a. - - - - - - -
t i M g ;t; w Are' requirements ofisk LCnf "; O in ..s q -Q , i s _e 6.aa' A -et a % w w utc.c__a 36n:,.O ter m=tfm to be entere6 ; e ;. ; y kg y q- -- j b. - - - - -
~ , i Continued noncompliance with thedftCO requjresents i
-[
/~
' ' '4 4,pi;cMkh '
swould
" e rfresult in the unit being required to p; pe d
"" ^" ,
l Me erek_ J N "" to Actions. comp'y with the Required * **' " M - m " - - ! fait A e A r b M W 1 desire & % {e edered - Ccapliance with Required Actions that permit continued i
- c to MODE or other specified condition provide level of safety for continued operation.
This is without regard change. to the status of the unit before or, after the MODE Therefore, in such cases, entry into a MODE or j p/em,y an other specified condition in the Appliegbility may be made
"" *j#p v upg in accordance with the provisions of the Required Actions.
n The previsions of this Specificatica should not be or odesgedre,F interpreted as endorsing the failure to exercise the good n.u;,s ,. .A practice of restoring systems or components to OPERA 8LE states sefore sw.n.w *
/g l<d/h.
The previsions of LC0 3.0.4 shall not prevent changes in (7 -afo MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. Al provisions of LCO 3.0.4 shall not prevent changes in MODES (continued)
= l', ^J'r K.w %d B 3.0-5 Rev. 0, ^*/?*!^2
l (24R 93 !YAD LCO Aoplicability B 3.0 ) BASES I
& C2 L0 i or other specified conditions in the Applicability that ontinued) result f -
- hutdown. !
Exceptions to m u J.0.4 are stated in the individual Specifications. Exceptions may apply to all the ACTIONS or to a specific Required Action of a Specification. ) Surveillances ds .'ot have to be perfomed on the associated inoperable equipment (or on variables outside the specified limits), as permitttd by SR 3.0.1. Therefore, changing q NDDES or other specified conditions while in an ACTIONS y,0 Condition, either in compliance with LC0 3.0.4, or where an exception to LCO 3.0.4 is stated, is not a violation of s SR 3.0.1 or SR 3.0.4 for those Surveillances that do not i i have to be perfomed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY yu3er prior to declaring the associated equipment OPERABLE (or < dM variable within limits) and restoring compliance with the affected LCO. LCO 3.0.5 LC0 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been i removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with ' the applicable Required Action (s)) to allow the performance of SRs to demonstrate: 4
- a. The OPERASILITY of the equipment being returned to service; or
- b. The OPERABILITY of other equipment.
The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to , perfom the allowed SRs. This Specification does not provide time to perform any other preventive or corrective maintenance. An example oi' demonstrating the OPERA 8ILITY of the equipment being returned to service is reopening a containment (continued)
?P 'S 5" 8 3.0-6 Rev. O,C/C/%
%e,%)
. ,. _ - . - . _ , . - . .-- . - - . -. . . = . . . -
4 OAK 93-NA'0 INSERT B6A LCO 3.0.4 is only applicable wher. ontering MODE 3 from MODE 4, MODE 2 from MODE 3 or 4, or MODE 1 from MODE 2. Furthermore, LCO ~ 3. 0. 4 is applicable when entering any other specified
.O condition in the Applicability only while operating in MODE 1, jf 2, or 3. The requirements of LCO 3.0.4 do not apply in MODES 4 and 5, or in other specified conditions of the Applicability
( (unless in MODE 1, 2, or 3) because the ACTIONS of individual
'- Spe.:ific tions suf ficiently define the remedial measures to be taken. [I some c es (e. .,...) t ese AunO S provid aNotej !
~
tat s "Whi this CO is n met, en into MODE o oth s ecifie cond ion in the App cabilit is
@phat pe. it d, unl s re equire nt ex icitly red to c ly wit CTIONS.
ecludin entry i o_a This t te J orj u/i ~ a t;he speci ed cogdition o the Appl abilit .y 9 5 f Y l 7 h INSERT RIVER BEND 2/'_ 0 ; g)O 10/1/g; cE
.-. G . 3 o- 6 .
e l ((4K W~/W0 l i LC0 Applicability 8 3.0 8ASES
# C2 L .
continued) declared inoperable or direct entry into Conditions and Required Actions for the supported systes. This may occur famediately other Required or after Action.some specified delay to perfore some Regardless of whether it is immediate or after some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions : and Required Actions shall be entered in accordance with
$1 - }Cf 'O'b'.
~~(S.S. @ -- &
l Specification ,
" Safety Function Determination Program" !
($FDP), ensures loss of safety function is detected and appropriate actions are taken. Upon failure to meet two or more LCOs concurrently, an evaluation shall be made to determine if loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory acf, ions may 6e identified as a result of the support system inoperability and corresponding exception to enterinjf ; supported system Conditions and Required Actions. 'ine- SFDP implements the requirements of LC0 3.0.6. Cross division checks to idecify a loss of safety function for those support systems 1. hat support safety systems are required. The cross division check verifies that the i supported systems of the redundant OPERA 8LE support systes are OPERA 8LE, thereby ensuring safety function is retained. If this evaluation determines that a loss of safety function - exists, the appropriate Conditions and Required Actions of the LC0 in which the loss of safety function exists are required to be entered. ' LCD 3.0.7 There are certain special tests and operations required to be perfomed at various times over the life of the unit. These special tests and operations are necessary to demonstrate select unit perfonaance characteristics, to perfom special maintenance activities, and to perfore special evolutions. Special Operations LCOs in Section 3.10 allow specified T5 requirements to be changed to pemit performances of these special tests and operations, which ', otherwise could not be performed if required to comply with the requirements of these TS. Unless otherwise specified, all the other TS requirements remain unchanged. This will (continued) 6 8 3.0-8 Rev. O, ^^/2"!^? - Se b.A
(TM 93 /vD SR Applicability ; B 3.0 8ASES i SR 3.0.3 Conditions be (continued) Surveillance. gin immediately upon the failure of the Completion of the Surveillance within the delay. period ! allowed by this Specification, or within the completion Time of the ACTIONS, restores compliance with SR 3.0.1. , si > SR ' 3.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability. , This Specification ensures that system and component OPERASILITY requirements and variable limits are met before entry into MODES or other specified conditions in the J Applicability for which these systems and components _ ensure sa_fe operation of the unit. pity spe to O' q a wf catio as p1 1 i
-fFKGT 1 %abi
.]
cg _, f 2,
! T384 A l The provisions of SR 3.0.4 shall not prevent changes in I MODES or other specified conditions in the Applicability l that are required to comply with ACTIONS.9 Tus(U GHB }
The precise requirements for perfomance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition (s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance. A Surveillance that could ' not be performed until after entering the LCO Applicability would have its Frequency specified such that it is not "due" d,o until the specific conditions needed are met. Alternately, s/ the Surveillance may be stated in the fors of a Note as not i required (to be met or performed) until a particular event,
' condition, or time has been reached. Further discussion of 3fssgr' the specific formats of SRs' annotation is found in gg jq Section 1.4, Frequency.
l 4WRMrSTS- 8 3.0-14 Rev. O,-0?/28/02 ! RIVE R 13 EA)b i
- - - ~
h 92-NR$ INSERT B14A
, wbhk dda k r-k \
However, in certain circumstances failing to meet an SR will i not result in SR 3.0.4 restricting a MO change or other i ! specified condition change. When a system, subsystem, s , division, component, device, or variable is inoperable or # outside its specified limits the associated SR(s) are not i required to be perfonned, Der, SR 3.0.167)urveillances do not i have to be performed on inoperable equipment *fr- When,/ equipment is inoperable, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be-performed is removed. Therefore, failing to perform the Surveillance (s) within the specified Frequency, on equipment that is inoperable, does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions in the Applicability. However, since the LCO is not met in this instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes. p The provisions of this Specification should not be interpreted jI ~~ as endorsing the failure to exercise the good practice of , restoring systems or components to OPERABLE status before ' entering an associated MODE or other specified condition in j the Applicability. INSERT B14B l In addition, the provisions of SR 3.0.4 shall not prevent ! changes in MODES or other specified conditions in the : Applicability that result from any un).t shutdown. ! INSERT B14C l fSR3.0.4 is only applicable when entering MODE 3 from MODE 4, j MODE 2 from MODE 3 or 4, or MODE 1 from MODE 2. Furthermore, M SR 3.0.4 is applicable when entering any other specified
- I -- - condition in the Applicability only while operating in MODE 1, 2, or 3. The requirements of SR 3.0.4 do not apply in MODES 4 and 5, or in other specified conditions of the Applicability (unless in MODE 1, 2, or 3) because the ACTIONS of individual f Specifications sufficiently define the remedial measures to be I taken.
l i INSERT RIVER BEND B 3.0-14 10/1/ 8 -
(
. l
**/ 0 l l
l l l l l l l RIVER BEND i l I SECTION 3.1 ATTACHMENT 1: CTS-PSTS COMPARISON DOCUMENT ATTACHMENT 2:ITS-PSTS COMPARISON DOCUMENT l 1
. . = ya9:-irai) l l l ATTACHMENT 1 : l iTS - PSTS . COMPARISON DOCUMENT l l REVISION 1 SECTION 3.1 REVISED PAGES ; l l l 1 1 A: MARKUP OF CTS j 18: NOT USED 1C: NOT USED l i
e l 1 ATTACHMENT 1 A CTS - PSTS COMPARISON DOCUMENT REVISION 1 MARKUP OF CTS
[La 9.1-Hi] l REACTIVITY CONTROL SYSTEMS l Terca,Ee._.l. ped 4
$URVEILLANCE REQUIREENTS (Continued) dhe cor<ed paw 6
- 2. Determining" that the available weight of Beren-10 is greater (R 3.17.E than or equal to 143 lbs, the percent weight concentration of sodium pentaborate in solution is equal Eo or less than 9.55 by weight, and the minimum required solution volume.
- 3. Verifying that each valve, manual, power operated or automatic, SR 11.7. G in the f'ow path that is not locked, sealed, or otherwise ,
secured in. position, is in its correct position,4 sR 3 1.7 1 4. Determining that the Staney Liquid Control System satisfies 5 i the follow'ng equation: (C)(E) 1 413 Where: - C = sodium pentaborate concentration, in weight percent, as determined per specification 4.1.5.b.2. E = Baron-10 enrichment, in atos percent **.
- c. Demonstrating that, when tested pursuant'to Specification 4.0.5, the SR J. l .7.7 .
ainieue I' low requirement of 41.2 pe per pump at a pressure of greater
, than or equal to 1220 psig is set.
- d. At least once per 18 months (during shuttewn]by;
- 1. ' Initiating one of the staney liquid control system loops,
.5U ll6 inc1 Wing an explosive valve, and verifying that a flow pat from the sumos to the reactor pressure vessel is availasle y I (pumpine dominera11ree water into the reactor vonsel.IK r replacement energe for sne explosive valve sha be from the LA3 same annufactured batch as the one fired or free another batch which has been certified by having one of that batch success-(fullyfired.fBothinjectionloopssnaitsetestesinzemonths.
j Sr3.l.1.5*This test shall also be performed anytise water or boron is added to the solution or when the solution temperature drops below 45*F. l Al I
- # 3'^ '
M **The b ron-10 enrichment of the solution sl.all be determined anytime boron is added to the solution.
/#h I3 AIVER BEND - UNIT 1 3/4 1-20 Amendeent No.
I l
(ik vi-no
insertable control rods in core cells containing one or . more fuel assemblies. U @nso$ cu%.wh E.3 Initiate action to I hour restore to @ OPERA 8LE status. ' f E4 Init te acti to 1 our ; res re one T A sub ystem to ERABLE t us. \1- . - IM E.5 nitiate tion to I hou restore e isolati valve associst inst tation to status in each [ econdary conta unent] s penet tion f1 path ~ % not olated. l l -ko TE - - - W Inisatt. actioSn l yyou( ^ ' g dwin each Exhy anfix2 a [ primaq conkim+ \ pa m n.'lh n A c \ airloch 7 m sLnak/a p ,p , \ ) coohol . . yg- ; \ _ \ 3,1 - R~a, 0.. J 9WIW6-646 3.1-3 Rev. O, 09/28/92 (T dx. 9 3-4 4 0 Control Rod OPERA 81LITY 3.1.3 ACTIONS C0fGIT10N REQUIRED ACTION COMPLETION TIME E-- ---WTE ' -- A. (continued) A.2 (No Ticab ; I s a o en e low r G .I setpoint (LPSP) of ~ N gpg the Rod Pattern control System 4A./-) Perfors SR 3.1.3.2 24 hours from and SR 3.1.3.3 for / dew,r ,/ each withdrawn c,,f,y,;,7,, ,4 OPfRABLE control rod. c,,wmg ,,W -rnwe twsd AE y,u 4a A.3 Perfors SR 3.1.1.1. 72 hours B. Two or more withdrawn 8.1 0 sars t assoc ted k ,, control rods stuck. . 7 - l \ .2 Be in MODE 3. 12 hours #40' (continued) t l 1 l i Qw Be-J sWRf6-SM- 3.1-8 Rev. O, 09/28/92 s? QA 9:-W20 Centrol Red Scram Times 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times LCO 3.1.4 a. No more than OPERA 8LE control rods shall be " slow,' in accordance with Table 3.1.4 1; and
- b. No b 'M-
- OPERA 8LE control ro slow" shall occupy locatio yg g ,,,j/y f, uoPEf442Jcw a ,wa
,,co.ma-M af? NN APPLICA8ILITY: MODES 1 and 2. g/<a/ ev8 d[a_ s fw< - 3
- 3. /
ACTIONS DA ColeITION REQUIRED ACTION COMPLETION TINE A. Requirements of the A.1 8e in M00E 3. 12 hours LC0 not met. SURVEILLANCE REQUIREMENTS ......................................N0Tg..................................... During single control rod scras time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator. l SURVEILLANCE FREQUENCY SR 3.1.4.1 ' Yerify each control rod scram time is Prior to within the limits of Table 3.1.4 1 with exceeding OBl reactor steam dose pressure a,J950P)sig. 40% RTP after 7 fuel movement . within the reactor pressure vessel (continued) 9.~ 6r.J m 3.1 13 Rev. O, 09/28/92 {} AR 92 * /Y51 ) / - 3.t.s ~' 3.1 REACTIVITY CONTROL SYSTDtS , , 6 3.1.6 kod PattehtontroD' h LCO 3.1.6 OPERABLE control rods shall comply with the requirements of th anked position withdrawal sequence (8PWS) .t.o @ APPLICABILITY: MODES 1 and 2 with THERMAL POWER W 4 RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TINE A. One or more OPERA 8LE A.1 --..--.-NOTE--------- control rods not in Affected control rods compliance with @ .V' 18PWSW may h bypassed in Rod Action Control System (RACS) in accordance with SR 3.3.2.1. . Nove associated 8 hours control rod (s) to correct position. 28 A.2 Declare associated 8 hours control rod (s) inoperable. (continued) kW/hd 3.1-19 Rev. O, 09/28/92 (UK 93*/'/50 ACTIONS (continued) @ COMITION REQUIRED ACTION COMPT.ETION TIME B. Nine or more OPERABLE B.1 --------NOTE--------- control rods not in Affected control rods compliance with may be bypassed in h SPWS}@- RACS in accordance with SR 3.3.2.1. r insertion only. Suspend withdrawal of Immediately control rods. i O 8.2 Place the reactor 1 hour mode switch in the shutdown position. l SURVEILLANCE REQUIRENENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERA 8LE control rods comply with ,BPWS M 24 hours O e 2 L.) /C :TO 3.1-20 Rev. O, 09/28/92 __ m _ =_ _ _ --- =- - - - - - - - NorE - . . - - - --. - - - - (1y 93.Hg; ) -fte erwieum re avire0 a erai/a ble s o/u vulame is c[efetmheD by the perfor/wsr same. 5LC 5ystem s ' ph SR 3.1.7.f. 3.1.7 ,3,ggj - - - - - - - - - - - - - ,, w -w _----- SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 erify available volume of sodium vpentaboratesolutionisQ [ ISM] W 1 % ,. d Dh4hourk (3/ '], -- & A*4o r erawIA.WM*VievM reQuiteH AvailaWe, .s o/u /ron vQ v*~rC _} SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours -- " .A "I[ y, [""I""" _ \ Aco._rs'.sgeT 224 ktWL) - SR 3.1.7.3 Ve/ify t erature f pump uction pin 24 ho s s withi the li ts of [F gure 3. 7-1]g f- Ps ~ SR 3.1.7.4 Verify continuity of explosive charge. 31 days SR 3.1.7.5 er fy he ra n 31 days lFg e 3 1.7-Fg M , /o 6eou Once within 0 6 /Y3 /b, uA +be jcerced werf gf7,"rfr'on coxcewkhM 0 4 h k o m p e d u /pra /e is added to / o/v60d E .:5 k~ d A 7', solution , JM Ja ,- a 1ha -; y we-<.z~.D a Daila 4/e so /sa v,/vme . / Once within / 24 hours after 7' l ' solution temperature is MI restoreb [ gu
- 1. -
h ON (continued) BWR/6 STS 3.1-22 Rev. O, 09/28/92 W, f M 93-/Vij) INSERT 22A SURVEILLANCE FREQUENCY .................... NOTE------------------------ Sodium Pentaborate Concentration (C), in weight percent, is determined by the performance of _ SR 3.1.7.5. Doron-10 enrichment (E), in atom percent, is determined by the performance of G('l j E# SR 3.1.7.9. Verify that the SLC System satisfies the following 31 days equation: (C) (E) = 413 INSERT C//l9'l RIVER BEND 3.1-22 10/1/91 4 (Jxis-ivil} SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY -+ / SR 3 1.' 1 Veri sodium entabora enrich nt is Prior o a .0) ato percent -10. addi ion to A (v - . SL tank -J I S K 3,t. 7 9 De.femOc 0: row - O cane 4ew7' Oxee wlfbdi ph' rke ,s ob boa. 2 V bh/'s a # 6I florost ib Ac0c0ck to tbe .so/a lsoA/ wcs BWR/6 STS 3.1-24 Rev. O, 09/28/92 ()M 93-/Y5) SOM B 3.1.1 8 3.1 REACTIVITY CONTROL SYSTEMS 8 3.1.1 SHUTDOWN MARGIN (SDM) 8ASES i 8ACKGROUNO SOM requirements are specified to ensure:
- a. The reactor can be made suberitical from all operating conditions and transients and Design Basis Events;
- b. The reactivity transients associated with postulated accident conditions are controllable within acceptable !
limits; and l
- c. The reactor will be maintained sufficiently l suberitical to preclude inadvertent criticality in the l shutdown condition.
These requirements are satisfied by the control rods, as described in GDC 26 (Ref.1), which can compensate for the reactivity effects of the fuel and water tem experienced during all operating conditions.perature changes APPLICABLE The control rod drop accident (CRDA) analysis (Refs. 2 g SAFETY ANALYSES and 3) assumes the core is subcritical with the highest g worth control rod withdrawn. Typically, the first control rod withdrawn has a very high reactivity worth and, should e the core be critical during the withdrawal of the first control rod, the consequences of a CRDA could exceed the M */ fuel damaes limits for a CRDA (see Bases for LCO 3.1.6, " Rod PatterngentrnP) . Also, SON is assumed as an initial condition for the control rod removal error during a refueling accident (Ref. 4). The analysis of this reactivity insertion event assumes the refueling interlocks are OPERA 8LE when the reactor is in the refueling mode of o mration. These interlocks prevent the withdrawal of more taan one control rod from the core during refueling. (Special consideration and requirements for multip e control rod withdrawal durin Operations LC0 3.10.g refuelingControl 6, Multiple are covered Rod in Special Withdrawal-Refueling.') The analysis assumes this condition is acceptable since the core will be shut down with the highest worth control rod withdrawn, if adequate 50M has been demonstrated. '(continued) {M, !" kJ B 3.1-1 Rev. O, 09/28/92 4 v-- m r a-w , , - ~ _-_ - (IM 93-/Q) Control Rfd OPERA 81LITY B 3.1.3 BASES APPLICA8LE the potential effects of reactivity insertion events caused - SAFETY ANALYSES malfunctions in +ha CR0 stem. gp J4 (continued) Qns vides assu e capa5T1fTy of inserting the entro1 MI that the assumptions for scram reactivity in the DRA and transient analyses are not violated. Since the SOM Ansures the reactor will be suberitical with the seconeossfcontrol rod withdrawn (assumed single failure), the add'tional failure of a second control rod to insert could invalidate the demonstrated SOM and potentially limit the ability of the CR0 System to hold the reactor suberitical. If the control. rod is stuck at an inserted position and becomes decoupled from the CRD, a control rod drop accident (CRDA) can possibly occur. Therefore, the requirement that all control rods be OPERA 8LE ensurus the CRD System can perform its intended function. The control rods also protect the fuel from daruge that could result in release of radioactivity. The limits protected are the MCPR Safety Limit (SL) (see Bases for LC0 3.2.2, ' MIN!ESI CRITICAL p0WER RATIO (MCPR)'), the 14 cladding plastic strain fuel design limit (see Bases for-LC0 3.2.17 " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLGHR)andLC03.2.3,'LINEARHEATGENERATIONRATE s./ (LH6R)'), and the fuel desage_ limit (see Bases for g6 LC0 3.1.6, td ontroy) during reactivity j - _ insertion e @ven _, The negative reactivity insertion (scram) provided by the CRD System provides the analytical basis for determination , of plant thermal limits and provides protection against fuel damage limits during a CRDA. Bases for LC0 3.1.4, l l LC0 3.1.5, and LC0 3.1.6 discuss in more detail how the SLs are protected by the CRD System. ~ Control rod OPERASILITY satisfies Criterion 3 of the NRC l Policy Statement. l \ l l LC0 OPERASILITY of an individual control rod is based on a 1 - combination of factors, primarily the scram insertion times, l the' control rod coupling integrity, and the ability to determine the control rod position. Accumulator OPERASILITY : . is addressed by LC0 3.1.5. The associated scram accumulator l status for a control rod only affects the scram insertion l (continued) p 044 l l L~f"O 8 3.1-13 Rev. O, 09/28/92 l ~. . . -- .. [Mi' 93 /MD Centrol Rod OPERAsILITY B 3.1.3 BASES ACTIONS A.I. A.2. and A.3 (continued) 4 g/ 2 hours is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to M. e, - / [ insert, and provides a reasonable amount of time to perfom the Requif ed Action in an orderly manner.fliolating thi-k ' condol rod Trom 'sdisi' p'r'evints diusa'ge'to the CRDM. N [p/ , { g icontrol rod can be isolated from scram by isolating tht, The
- T C B A - ihydraulic control unit from scram and normal m
p',CRD.f pressure, yet still maintain cooling wate 3,f Monitoring of the insertion capability for each withdrawn control rod must also be performed within 24 hours. f SR 3.1.3.2 and SR 3.1.3.3 perform periodic tests of the , Ns control rod insertion capability of withdrawn control rods, ' Testing each withdrawn control rod ensures that a generic I 6 & problem does not exist. The allowed Completion Time of. 1 " , 24 hours provides ~ reasonable time to test the control-rods, considering the potential for a need to reduce power i 7 ER L ow to perfore the tests. Required Action A.2 - f y d I 55
- J yAohatestateutne'riteutrament n nos'= ie h1 d_n g
'.. L b .__-jthe actua' low power setpoint (LPSP) of the . rod t i_ Qu - pattern controller (RPC), since the notch insertions may not be compatible with the requirements of rod. pattern control (LC0 3.1.6) and the RPC (LCO 3.3.2.1, ' Control Rod Block Instrumentation"). /'kn a m$ibih 1 b* "'# g7y" To a11ow continued operation with a withdrawn control rod stuck, an evaluation of adequate SOM is also required within M . d R '/ 72 hours. Should a 08A or transient require a shutdown, to 4aue C# / preserve the single failure criterion an additional control 20' du. Segwed, Y ,A,; s, rod would have to be assised to have failed to insert when f required. Therefore, the original SOM demonstration may not Ac/Ja dyr h ~4,o be valid. The SOM must therefore be evaluated g# "' # f f," "'^' measurement or analysis) with the stuck control (by rod at its stuck position and the highest worth OPERA 8LE control rod codes / rol u assumed to be fully withdrawn. .' * "") The allowed completion Time of 72 hours to verify SDM is adequate, considering that with a single control rod stuck 90WER h gre,/er i, a withdrawn position, the remaining OPERABLE control rods g i,y l v are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram. Even with the (continued) kWaua/5 Fi B 3.1-15 Rev. 0, 09/28/92 (LAR 93/4ED Control Rod OPERAt!LIiy 8 3.1.3 8ASES ACTIONS A.1. A.2. and A.3 (continued) postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach and maintain MODE 3 conditions (Ref. 7). g B.1 and B.2 1 , 40)' With two or more wit Wrawn control rods stuck, e- -t 9 _ s 'k - C : ' -- - - uld W!:^ ^ A ;;n _. ;;;xx . . m. . .. 7, g_, he plant + brought to HQDE 3 wi thin 12 houru.) l -d3 4 1l' M Ft M f56. K 0 2 :" C # R" j occurrence of more snan one control rod stuck at a withdrawn position increases the probability that the reactor cannot l be shut down if required. Insertion of all insertable l control rods eliminates the possibility of an additional ' failure of a control rod to insert. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. . c.1 and C.2 With one or more control rods inoperable for reasons other than bein continue,gprovided stuck in the the control withdrawn rodsposition, are fullyoperation inserted may within 3 hours and disarmed (electrically or hydraulically) within 4 hours. Inserting a control rod ensures the shutdown and scram capabi ities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disarmed b exhaust water isolation valves.y closing the drive Electrically, the water control and rods can be disarmed by disconnecting power from all four directional control valve solenoids. Required Accion C.1 is modified by a Note that allows control rods to be bypassed in the RAC5 if required to allow insertion of the inoperable t control rods and continued operation. SR 3.3.2.1. provides j additional requirements when the control rods are bypasse i to ensure compliance with the CRDA analysis. q l The allowed Completion Times are reasonable, considering the ! small number of allowed inoperable control rods, and provide 1 (continued) 2.y/' L "56. 1 B 3.1-16 Rev. 0, 09/28/92 --.,r- [4AW 93-WQ Control Rod OPERASILITY 8 3.1.3 8ASES ACTIONS (continued) l inoperable control rods exist, the plant must be brought to i a MODE in which the LC0 does not apply. To achieve this I status, the plant must be brought to MODE 3 within 12 hours. This ensures all. insertable control rods are inserted and , places the reactor in a condition that does not require the : active function (i.e., scram) of the control rods. The : number of control s pemitted to be inoperable when ' h d operating abov TP (i.e., no CRDA considerations) could l be more than the valuc specified, but the occurrence of a large number of inoperable control rods could be indicative ; of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours is reasonable, based on i operating erperience, to reach MODE 3 from full power conditions in an orderly manner and without challenging. l plant systems. 3.1.3.1 ' h a[x h SURVEILLANCE SR 6he REQUIREMENTS The position of each control rod must etemined, to i ensure adequate infomation on control tion is available to the operator for detemini ERA 8ILITY and controlling rod patterns. Control rod position may be detemined by the use of OPERA 8LE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The 24 hour Frequency of this SA is based on operating experieret related to expected changes in control rod position and the availability of control rod position indications in the control room. SR 3.1.3.2 and SR 3.1.3.3 control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. ; The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal. These Surveillances ar'e not requirsd when the actual LPSP of the RPC since the notch invertions m@ay not be compatible with the requiremen __- .- ~ W Or QM k gggjngd) B 3.1-18 Rev. 0, 09/28/92 (Tsc 93 sv4 0 Control Rod OPERA 8ILITY B 3.1.3 8ASES SURVEILLANCE SR 3.1.3.5 (continued) REQUIREMENTS performed anytime a control rod is withdrawn to the " full out' position (notch position 48) or prior to declaring the control rod OPERA 8tE after work on the control rod or CRD System that could affect coupling. This includes control l rods inserted one notch and then returned to the " full out' I position during the performance of SR 3.1.3.2. This ' Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and operating experience related to uncoupling events. REFERENCES 1, 10 CFR 50, Appendix A, GDC 26, GDC 27, GDC 28, and GDC 29. 2.U/SAR,Section/4.3.2.5.5[ @- 4.3.u/SAR,Section)4.6.1.1.2.5.3 'A/SAR, Section/5.2.2.2.37 @- 5. u/SAR, Section/15.4.17-3,/ 6.U/SAR,Sectic1/15.4.9[ C# 7. ED0-21231, " Banked Posit on Withdrawal Sequence," h ection 7.2, January 1977 6 'tOld B 3.1 20 Rev. O, 09/28/92 ()h 93-Mi0 Ccntrol R:d Scra3 Times 8 3.1.4 8ASES APPLICA8LE The scram function of the CRD System protects the MCPR SAFETY ANALYSES Safety Limit (5L) (see Bases for LC0 3.2.2, ' MINIMUM (continued) CRITICAL POWER RATIO (MCPR)'), and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, ' AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLH6R)," and LC0 3.2.3, ' LINEAR HEAT GENERATION RATE (LHGR)'), which ensure that no fuel damage will occur 1f these limits are not exceeded. Above 950 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prever.1 the - actual MCPR from becoming less than the MCPR SL .during the analyzed Ifmiting power transient. Below 950 psig, the scram function is assumed to perfom during the control rod drop accident (Ref. 6) and, therefore, also provides .3,I protection against violating fuel damage limits during gb06 s_ q = reactivity insertion accidents (see Bases for LC0 3.1.6, Rod PattenL@ tatJD')) . For the reactor vessel overpressure , ) protection ana' ysis, the scram function, along with the safety / relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits. , control rod scram times satisfy Criterion 3 of the NRC l Policy Statement. l LC0 The scram times s accompanying LCO)pecified are requiredintoTable ensure3.1.4-1 that the(inscram the I reactivity assumed in the DBA and transient analysis is set. To account for single failure and " slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster Clo than those assumed in tne design basis analysis. The scram I times Tave a margir4to allow up to 7.5% of the control rods @ (e.g.* S x 7.5% = D to have scram times that exceed the specified limits (i.e., " slow" control rods) assuming a t single stuck control rod (as allowed by LCO 3.1.3, " Control ' Rod OPERASILITY') and an additional control rod failing to scram per the single failure criterion. The scram times are 3 1 specified as a function of reactor steam done pressure to ! account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes (" pickup") when the index tube passes a specific location and then opens . (' dropout") as the index tube travels upward. Verification ' . of the specified scram times in Table 3.1.4-1 is accomplished through measurement of the " dropout" times. (continued) bIh4;^- -- B 3.1-22 Rev. O, 09/28/92 i _ 93-/VRI) , Control Red Scram Times , 8 3.1.4 SA$ES : LC0 To ensure that local scram reactivity rates are maintained (continued) ithin acceptable 1 its, no n ; " _.. :__ ;f "_T"= "tla=" caa hel y occupy 4 - . . (n w e le r & co s w a s o r n , e ocationA)S/J w,94 ,_ s(/aled
- u
"" *'( a M Table 3.1.4-1ismodifiedbytwoIlotes,whichstatecontro , rods with scram times not within the limits of the Table a #' c , red " slow" and that control rods with scram times g>9 oconds are considered inoperable as required by 91 frisEL Y Ea3 A - APPLICA81LITY In MODES 1 and 2, c. scram is assumed to function during transients and accidents analyzed for these plant conditions. These events are assumed to occur during startup and power operation; therefore, the scram function of the control rods is required during_ these MODES. In 4 jh MODES 3 and 4, the control rods are _m gg r: to De withdrawn t pesi r d ate, l . . . 4 4v. . .i ) 823 "Si P Mb- 1 * ; l adequate requirements for contro' rod scram capability during these conditions. Scram requirements in MODE 5 are conta' ned in '.C0 3.9.5, " Control Rod OPERASILITY-Refueling." i l ACTIONS M h- W5ffT When the requirements of this LC0 sre not met the plant 8 23 C-must be brought to a MODE in which the LC0 does not apply. l To achieve .his status, the plant must be brought to MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach M00E 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE The four SRs of this LC0 are modified by a Note stating that during a single control rod scram time surveillance, the CRD REQUIREMENTS pumps shall be isolated from the associated scram accumulator. With the CRD pump isolated (i.e., charging valveclosed),theinfluenceoftheCRDpum affect the single control rod scram times. p head During does not a full (continued) k Ms Osi!) s 3.1 23 Rev. O, 09/28/92 [4K 93-191Q s.I "~ ./'3T hk g/odal,[ 7 of INSERT B25A The limits for reactor pressures < 950 psig are established based on N - T--* ^f relaric ;...u toineeting the acceptance criteria at reactor pressures 2 950 psig. Limits for 2 950 psig are found in Table 3.1.4-1. If testing demonstrates the affected control rod does not meet these limits, but is within the 7 second limit of Table 3 .1. 4 - 1 Note 2, the control rod can be declared OPERABLE and " slow." 6 INSERT thhv RIVER BEND B 3.1-25 N- (44C 934VdD Control Rod Scram Accumulators 8 3.1.5 8 3.1 REACTIVITY CONTROL SYSTDIS 8 3.1.5 Control Rod Scram Accumulators BASES BACKGROUND The control rod scram accumulators are part of the Control Rod Drive (CRD) System and are provided to ensure that the control rods scram under varying reactor conditions. The control rod scram accumulators store sufficient energy to fully insert a control rod at any reactor vessel pressure. The accumulator is a hydraulic cylinder with a free floating piston. The piston separates the water used to scram the control rods from the nitrogen, which provides the required energy. The scram accumulators are necessary to scram the control rods within the required insertion times of LC0 3.1.4, " Control Rod Scram Times." APPLICA8LE The analytical methods and assuptions used in evaluating SAFETY ANALYSES the control rod scram function are presented in References 1, 2, 3, and 4. The Design Basis Accident (DSA) and transient analyses assume that a 1 of the control rods scram at a specified insertion rate. OPERASILITY of each I individual control rod scram accumulator, along with LC0 3.1.3, " Control Rod 0PERASILITY," and LC0 3.1.4, ensures that the scram reactivity assumed in the DBA and transient ; i analyses can be met. The existence of an inoperable ; accumulator may invalidate prior scram time measurements for 1 the associated control rod. l The scram function of the CRD System, and, therefore, the OPERABILITY of the accumulators, protects the MCPR Safaty Limit (see Bases for LC0 3.2.2, "MINIMim CRITICAL POWER RATIO (MCPR)') and the 1% cladding plastic strain fuel design limit (see Bases for LC0 3.2.1, " AVERAGE PLANAR
- LINEAR HEAT GENERATION RATE (APLMGR)," and LC0 3.2.3,
'LINEARHEATGENERATIONRATE(LHGR)),whichensurethatno ! fuel damage will occur if these limits are not exceeded (see i Bases for LC0 3.1.4). Also, the scram functior at low reactor vessel pressure (i.e., startup conditions) provides l protection against violating fuel design limits during l reactivity insertion accidents (see Bases for LC0 3.1.6, s/ god (continued) tIf f5 B 3.1-27 Rev. O, 09/28/92 1 l (/,Ad 93- hWT) Control nod scram Accumulators 8 3.1.5 .w BASEE APPL.1 CABLE Control rod scram accumulators satisfy Criterion 3 of the SAFETY ANALYSES NRC Policy Statement. (c.ontinued) LCO The OPERASILITY of the control rod scram accumulators is required to ensure that adequate scram insertion capability exists when needed over the entire range of reactor tressures. The OPERASILITY of the scram accumulators is used on maintaining adequate accumulator pressure. APPLICABILITY In MODES 1 and 2, the scram function is required for skceAereufe)r mitigation of OSAs accumulators mustand transients and, be OPERABLE therefore, to support the scram thefunction. scram 4 gg g, ;, In M00E5^ 3. 5,wc withdrawn and 4,opercontrol rods are Nto y n l es L .Iv.s/ sing sku&e ad a ivi df..it 1- hutd " an 3.1 ,1 g 4 cu b d d go'g g L in a+ _ od Wi rewal - d Ad - ." ch l g
- adequate requirements for contro rod scram l gNHej, 7/,js prog- accumulator OPERASILITY under these conditions. !
. Requirements for scram accumulators in MODE 5 are contained in LC0 3.g.5, " Control Rod OPERASILITY-Refueling.' ACTIONS QRukauc.w4/w The ACTIONS table is modified by a Note indicating that a , separate Condition entry is allowed for each control rod. 2,/ l This is acceptable since the Required Actions for each . *Ygot Condition provide appropriate compensatory action for each 7 piammuse8m control rod. Complying with the Required Actions C - ecE'd/ x may allow for continued operation and subsequent sumpammmaa 4 control rods governed by subsequent Condition entry and application of associated Required Actions. A.1 and A.2 fo00 With one control rod scram ac ator inoperable and the reactor steam done s g, the control rod may be declared " slow,' pressure asince the control rod will still scram at the reactor operating pressure but may not satisfy the . C, required scram times in Table 3.1.4-1=: = :=:. Required Action A.1 is modified by a Note, which clarifies g (continued) QweAM 9WRftMHS 8 3.1-28 Rev. O, 09/28/92 A Att CSI Red Pattern ontt ,/ l 83.1 REACTIVITY CONTROL SYSTDt5 %/ ~ { B 3.1.6 od Pa M- ' ' l BASES j l l BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod pattern controller (RPC) (LC0 3.3.2.1, " Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the opetrating range of all GI W "d" **"*"' "d' i"""# # ; * * * ' " " ' " * " pww effectively limit the potent'al amount of reactivity 1 ,,pt addition that could occur in the event of a control rod drop l pggg accident (CRDA). This Specification assures that the control rod patterns are consistentytntheassumptionsoftheCRDAanalysesof References g \ a+4 2 ) f-0yd C APPLICABLE The analytical methods and assumptions usadain evaluating j SAFETY ANALYSES the CRDA are summarized in References : r- -- 2 CRDA -l analyses assume that the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis. The RPC (LC0 3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions ; of the CRDA analysis are not violated. ' Prevention or mitigation of positive reactivity insertion events is necessary to limit the one deposition in the fuel, thereby preventing significant I damage, which could result in undue release of radioactivity. Since the failure consequences for U0, have been shown to be ins cant below fuel energy depositions of 300 cal /gm ( the fuel damage lim <t of 280 cal /gm provides a -Q ma safety from si nificant core dama ge.fwhich wou W result 'n release of rad osctivity (Refs. _--_ ) . Generic evaluat< ons (Reff.-W of a design basis CRDA (i.e., a CRDA resulting in a peak fuel energy deposition of 280 cal /ge) have shown that if the peak fuel enthalpy remains below 280 cal /ge, then the maximum reactor p re h will be less than the required ASME Code limits (Ref. the calculated offs ses will be well within the and required limits (Re (continued) ~ (2 ""d..g,s. .1 8 3.1-32 Rev. O, 09/28/92. -v -gw----w-- ~ QtC93 AID Rod Pattern ' ontt ,- .1.6 20 APPLICABLE Control rod patterns analyzed in fe 11ow the SAFETY ANALY banked po in withdrawal sequence (SPW5 cribed in (continued Reference The BPWS is applicable from the c ton of all cent s fully inserted to RTP (Ref For the BPW5, the control rods are required to be no in groups, with all control rods assigned to a specific group required to be within specified banked positions (e.g., between notches 08 and 12). The banked positions are defined to minimize the maximum incremental control rod worths without being overly restrictive d normal plant operation. The generic BPW5 analysis (Re also evaluated the effect of fully inserted, i ble control rods not in compliance with the sequence, to allow a limited number (i.e., eight) and distribution of fully inserted, inoperable control rods. Rod pattern control satisfies the requirements of Criterion 3 of the NRC Policy Statement. LC0 Compliance with the prescribed control rod sequences miniaires the potential consequences of a CRDA by limiting the initial conditions to those consistent with the BPWS. ' This LC0 only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LC0 3.1.3, ' Control Rod OPERASILITY," r.ansistent with the allowances for inoperable control rods in the SPW5. , APPLICA81LITY In MODES 1 and 2, wheniTN POWER is s M RTP, the CRDA is a Design Basis Acciden DRA) and, therefore, compliance ' with the assumptions of the safety analysis is required. ~ leen THERMAL POWER is > RTP, there is no credible control rod configuration that results in a control rod worth that could e the 280 cal /gm fuel damage limit during a CRDA (Ref In MODES 3, 4, and 5, since the reactor is shut nd only a single control rod can be withdrawn from a core celt containing fuel assemblies, adequate SOM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a singic control rod withdrawn. L 6,a (continued) BWRffr43s B 3.1 33 Rev. O, 09/28/92 (1493-!'/Rb Rod Pattern' Control ,j - alo06 BASE 3 (continued) ACTIONS A.1 and A.2 With one or more OPERA 8LE control rods not in compliance with the prescribed control rod sequence, action may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8 hours. Noncompliance with the prescribed sequence any be the result of " double notching," drifting from a control rod drive coolino water sient, leaking scram valves, or a power @ reduct<on to RTP before establishing the correct control rod patte The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence. When the control rod pattern is not in compliance with the prescribed sequence, all control rod movement should be stopped except for moves needed to correct the control rod pattern, or scram if warranted. Required Action A.1 is modified by a Note, which allows control rods to be bypassed in Rod Action Control System (RACS) to allow the affected control rods to be returned to their correct position. This ensures that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action A.2. OPERA 8!LITY of control rods is determined by compliance with LC0 3.1.3; LC0 3.1.4, ' Control Rod Scras Times'; and i LC0 3.1.5, ' Control Rod Scram Accumulators." The allowed Completion Time of 8 hours is reasonable, considering the ' restrictions on the number of allowed out of sequence control rods and the . low probability of a CRDA occurring during the time the control rods are out of sequence. 8.1 and R.2 If nine er more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be . suspended immediately to prevent the potential for further : deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn be i allowed position is allowed since, in general, insertionyondoftheir control rods has less impact on contro rod worth than withdrawals have. Required Action 8.1 is modified by a Note (continued) B 3.1-34 Rev. 0,09/28/92 l @< 9S'/W3 Rod Pattern ' ontr) g , . - .6 260 BASES ACTIONS B.1 and B.2 (continued) 3 that allows the affected contro rods to be bypassed in RACS in accordance with 5R 3.3.2.1 to allow insertion only. With nine or more OPERA 8LE control rods not in compliance with BPWS, the reactor mode switch must be placed in the shutdown position within 1 hour. With the reactor mode switch in shutdown, the reactor is shut down, and therefore does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence. SURVEILLANCE SA 3.1.6.1 REQUIRENENTS The control rod pattern is verified to be in compliance with I the BPWS at a 24 hour Frequency, enstring the assumptions of the CADA analyses are met. The 24 hour Frequency of this Surveillance ucs developed considering that the primary 4 check of the control red patters cam iance with the BPWS is . performed by the RPC (LC0 3.3.2.1). p The RPC provides control rod blocks to enforce the required control rod , and is required to be OPERA 8tE when operating at i
- 9. gg y of p.g ,,, y y g A
he\. G CSTAR C ' l'114dsk acerovd truiM. REFERENCES i g r.m.....,_ .. .=. t. ,~== m gi , 6 ;2...i ~.ti;:t h; !;-t- :," ~'~ i = C r r d '~' O Rd edTo 7k RM , Section 15.4.9. ts Ary> IRNtEG-0g79, NRC Safety Evaluation Report t T GE55AR II OWR /6 Nuclear Island Design, Docket ' No. 50-447,' Section 4.2.1.3.2 April 1983. ' 4. NUREG-0800, " Standard Review Plan," Section 15.4.9 ' Radiological Consequences of Control Rod Drop Accident (SWR)," Revision.2, July 1981. - (continued) a.a 34:!?! B 3.1-35 Rev. O, 09/28/92 (p,g 93-/YQl} Rod Patte q ontr .- .6 8ASES g REFERENCES (continued) h 10 CFR 100.11, " Determination of Exclusion Area Low Population Zone and Population Center Distance." NED0-21778-A, " Transient Pressure Rises Affected
- Fracture Toughness Requirements for Solling Water Reactors," December 1978.
ASME, Boiler and Pressure Vessel Code. NED0-21231 " nked Position Withdrawal Sequence," January 1977. 1 I %. G..) E M CIS B 3.1-36 Rev. O, 09/28/92 _ _ _ _ _ _ - - _ ~. . (UR 93-NRD INSERT B41A The requirements of 10 CFR 50.62 (Ref.-1) are met by the use ._ g,/*fd of a sodium pentaborate solution enr:.ched in the B-10 isotope _ \. (65 atom percent B-10). Enri hed a lium p tabor te so tion ' N ris de D mixin gran lar, nrich sod pen abora wit : wat r. otopi test on e gra lar ium ntab ate 4 ve ify e act 1 B- en chmen must e per rmed rior o a diti to e SL tan in or r to naure hat e pr er )Q, , -10 tom p centa e is eing ed. , SR 3.1.7.3 determines whether the sodium pentaborate i concentration, in conjunction with the B-10 enrichment, is within limits to meet the requirements of 10 CFR 50.62 (Ref. 1). SR 3.1.7.5 ensures that the parameters used in the ; detemination of sodium pentaborate concentration are within limits.3 This Surveillance requires an examination of the [ sodium pentaborate solution by using chemical analysis to ( ensure the proper weight of B-10 exists in the storage tank. , i SR 3.1.7.5 must be performed anytime boron or water is added i to the storage tank solution to establish that the weight of B-10 is within the specified limits. This Surveillance must ! be perfomed anytime the solution temperature is restored to . = 45'F, to ensure no significant boron precipitation occurred. l The 31 day Frequency of these Surveillances is appropriate because of the relatively slow variation of boron concentration between surveillances. A ;/ A /6 n L- u 4 u!& n/- ak & ; % ;, %p-p n a n p n A L A A b y +~x. INSERT RIVER BEND B 3.1-41 Cf M , i. t 93-/YRb SLC System B 3.1.7 BASES e-9 SURVEILLANCE REQUIREMENTS temperat SR 3.1. re verifi tion of t is piping
- 3. Howe r, if, in utred b I
rformin deters ed that t tempera re of thi piping ha below he specif SR 3.1.7. it]is fallen perfo d once w; einimum this Su eillance st be is hin 24 h rs after the pipin emperature) tored wittWn the If ts of Fi re 3.1.7 . / 'sa3.1.7.14{ Enriched sodfiss pen ahorate so tion is na granula , enriched sodium pen ahorate wit water. by mixin D @ tests the gran lar sodium I topic actua 5-10 enri t must ntaborate to verify e perfo prior to dition to tentage pe e SLCistanking in order used. t ensure th the prope 8-10 atom; r i REFERENCES 1. 10 CFR 50.62. i h 2. p/IAR, Sectio p 9.3.5.3 h m .SK 1/i 7, 9 g,; *S9 2xrNIref soSum pexhbotek afahNu O make ay mxiu gia .vub t, avrilbeh so. Sum pes /a boro lc wNb wafer, .[.sofopis fads ' ow t$e .s oSv , h i p cx ta A n A s o /u /; , y fo A/c,,Ja. adual a -io ennkant ,,,-f na pa, for,,.9 j -ebe. \ ONce wifbdi o7 y' bovn aber noro" i c ' \ dSc$ch fo sbe sobhow Y otArt 1'O eNw'c f j tbaf rbe C -/O ewnlNo<rew/ E schef valc. , Earkbt veuf AA04dv k fl u owly regwr,chwbes aorow h male.Y sixec occur earkira reof ebawye. c wavl cy any o ther process . k 0o.J If5 M B 3.1-43 Rev. O, 09/28/92 Q42 9:-/YD 50V Vent and Drain Valves B 3.1.8 BASES ACTIONS for each Condition provide ap (continued) for each inoperable SDV line.propriate compensatory actions Complying with the Required ,m Actions may allow for continued operation, and subsequent [ 3,7g 4 g pr inoperable 50V lines are governed by subsequent Condition entry and application of associated Required Actions. ' s/;// no/utA, yde. afedeA % t/ /;oe '"#f '3, 6//~/" " ' L.L ;,, 0 -r/u .s When one SDV vent or drain valve is perable in one or more ' y ec y ,,, ,/, /c p w a u 7lines, days.the valves must be restored to OPERA 8LE status within ) /,, 4 //m fy ce The Completion Time is reasonable, given the level of redundancy in the lines and the low probability of a j f'"'"f .jo ascram occurrin 1 i 8" # /*p " c4 - . _3;__ g during the time the valve (is')are inoperable __ --- -- r. The SDV is still isolable since , Reat /o r cc e tie redundant valve in the affected line is OPERABLE. ( ,,, j g / 6 4 h a /, During these periods, the single failure criterion , N V- be preserved, and a higher risk exists to allow reactor not i ( water out of the primary system during a scram.
- i 2,/ CI 40c b1 l If both valves in a Ifne are inoperable, the line must be isolated to contain the reactor coolant during a scras.
i When a line is isolated, the potential for an inadvertent almm;s trr.4iv]e scram due to high SDV level is increased. Required i Action 5.1 is modified by a Note that allows ca+rel5 ecure draining of the SDV when a line is isolated. periodic During these 4 * #"("# periods, the line may be unisolated under administrative control. This allows an ! -- A__ Mi;; i n O. .y accumulated water to .} to be drained, in preclude the line a ructor scram on sDV hi OC2. level.. This is acceptable, since can be !_- closed quic a scram occurs with the valve open. LJk^' # The 8 hour Completion Time to isolate the line is based on ( No e c c A o r , the low probability of a scras occurring while the line is not isolated and unlikelihood of significant CR0 seal leakage. ful If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LC0 (continued) B 3.1-46 Rev. O, 09/28/92 Q293->wd RIVER BEND SECTION 3.2 l l l ATTACHMENT 1: CTS-PSTS COMPARISON DOCUMENT i ATTACHMENT 2:ITS-PSTS COMPARISON DOCUMENT (ue owiD j I ATTACHMENT 1 l l ITS - PSTS COMPARISON DOCUMENT REVISION 1 SECTION 3.2 REVISED PAGES , 1 A: MARKUP OF CTS 1B: NOT USED 1C: NOT USED i 9 ATTACHMENT 1 A CTS - PSTS COMPARISON DOCUMENT REVISION 1 MARKUP OF CTS l l l l l . U bf'kl i t.L : 4 U 1 -dL 4 - 5 0C l Oc, Ur*V4 15:20 f*o.015 P.08/15 POWER DISTRIRITION LIMITS (zg 93.gn) 1___/4 . f . 2 APWI SETPOINIX /N LIMIT 7mt ra*ITIne FOR OPERATION '? ;8 ~
- b The AMBI flow biased sinelated thermal power-high scram trio setoojnt ($?
cind flens biW neutmn Flux-upscale cont":sl rod block trip setpcint (5,,))shall / be established according to the following relationsh1ps: RI l
- a. Two Recirculation loop Operation Trin Satnoint Allowable Value S $ (0.66W + 4as)T S 5 (0.4 W + 515)T g, Ma (Gs(0.66W+425)T S. s (0.68W + 455) %
%2 M b. Single Recirculation Loop Operation Trin Satnatat Allowable Value ' S s (0.66W t- 42.75)T S s (0.66W + 45.75)T S s (0.66W + 36.75)T S., s (0.66W + 39.7%Q where: 5 and S , are in percent of RATED THERMAL POWER, W = Loop recirculation flow n a percentage of the loop recirculation flow which produces a rated core flow of 84.5 million Ibs/hr. 1 3xPRTP+ t 1 T = gggg provided CMFLPD $ 0.6 x FRTP + 0.4, othe mise PRTP ^ T = CMPLPD T is applied only if less than or equal to 1.0. \ t \ g \ FRTP is the FRACTION OF RATED THERMAL POWER. ' i CMFLPD is the CORE MAXI 1Mt FilACTION OF LIMITING POWER DENSITY. s APPLICABILITY:,<0PE*fTITf1 - ='!%;; ". tdTNEf8tAL POWER is greater than or equal to 255 of RATED THEMAL.PUWER. 2. 0 d76 ACTION: \ ' fGhpWithflowthe biasedAPRM neutreeflowflux-upscale biased simulated contrei runtherisal nioca triopower-high scram trip set setpoint lassRnser- g as abo vative than the value sm in the Allowable Value coluen for Se 1 i detemined,15Ftiate corrective scuos witnin 15 m' nuteDand adjus cgad/or'S , Coda jf to be consistemt with the Trip 5stpoint value
- within 6Lhours or mduce THERMAL Ccw@ 11 POWER to less than 255 of RATB THERMAL POWER within the next 4 hours.'
MI / / With I < 1.0, rather than adjusting the APlWI setpoints, the APigt gain may be adjusted such that the adjusted APRM readings result in a calculated T > 1.0 124c when the APRM reading is substituted for FRTPJprovided that the adjusted APRM ). r'eading coes not exceed 1005 of RAIm ThrisiAL POWER, and a notice of thef M adjustaunt is posted on the reactor control panel. f ' 3/4 2-7 Amendment No. M r h-RIVER BEle - UNIT 1 i OCT-F/-1994 15:27 301 504 3861 P.08 , U$NPC TEL:?01-5C4-3861 Oc. 07'94 15:21 No.015 P.09/15 POWER DISTRIHUTION LIMITS SURVEILLAM r RIQUIRfMENTS " '# ' q, l ! 4.2.2 The FATP and CMLPD shall be detemined, the value of T calculated, an the most recent actual APM flow biased simulated themal power-high ser flow biased neutron flux-upscale control rod block trirser.potnts verified to be g within the above limits or adjusted, as required:
- a. At least once per 24 hours, , y 9{,
y b. Within 12 ! Lours afterdsenMive er ERMA1, POWER i :r:--- e* p pt m 3 RATED THEMAL POWER, and
- c. Initially and at least once per 12 Itours when the reactor is operat-
#'#'Y'E ing with T s 1.0. }- , g The provisions of Specification 4.0.4 are not appitcabhD #7N [ RIVER BEND - UNIT 1 3/4 2-7a Amendment No. Sh75 h F4UFULF:~SL P.ED {LAR 93-l'l7 0 1 ATTACHMENT 2 ITS - PSTS COMPARISON DOCUMENT REVISION 1 SECTION 3.2 REVISED PAGES 2A: MARKUP OF ITS 2B: NOT USED i i 6 1 ATTACHMENT 2A i l l ITS - PSTS COMPARISON DOCUMENT l REVISION 1 ) MARKUP OF ITS l l l l l (Z,4e 9:8 APRM Gain and Sstpsints {^;; .. ^' 3.2 POWER DISTRIBUTION LIMITS . 3.2.4 Average Power Range Monitor (APRM) Gain and Setpoints {^-tf r:?,' " LCO 3.2.4' / a. shall be7-dess th e er :;;:1 t: Fr::ti r. c7 ::TW, or y,p b. Each required APRM setpoint specified in the COLR shall A be made applicable; or k c. Each required APRM gain shall be ad;usted such that the s APRM readings c m e := : tf-- "*L :.? e ~ " E' result ma c aliulele.0 Yh /.O Jen Ne QM n*S~y 4 su6, f% tee Ar FerP, APPLICABILITY: THERMAL POWER = 25% RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Satisfy the 6 hours LCO not met. ' requirements of the ~ : LCO. B. Required Action and 8.1 Reduce THERMAL POWER 4 hours associated Completion to < 25% RTP. . Time not met. > 9 E O Bn/C STS- 3.2-4 Rev. O, 09/28/92 TI;l (L AR 9:-iviQ APRM Gain and Sstpoints '^;ti:n!)* _ 3.2.4 SURVEILLANCE REQUIREMENTS
- SURVEILLANCE FREQUENCY -
SR 3.2.4.1 4 VerifyCMs@ithir, li=iD f , I0nce within 4 '# 412hoursaft]r e = 25% RTP #76 NtR h . 24 hours thereafter m - ~ _ _ _ _ - _ aora __ _ _ _ _ _ _ _ _ 14+- re9eired -4o I.e met if SR 3.2.4.2. u so+u bd -fe ico J. 2.#, _Tfem b or e recjvicea,n ts , 3 S u r if f t t ' ^ ^2f4 FRiousNCV N 3.7-.9. 2. _ ~ _ AJoT( ,__ _, Neb Itiun'etd k g. rn e y i[ $Q 3.2.4.l is ss+s deJ 4, Aco .f. 2,4 , Item ._a. __requi,ea e d r. , __. _. ) K I ' ,5 s [er 'th A PRf6 sehmA4s _ .a,ua u.a or d a.14 ~~ ( h G_ - ~ n - R a w eeni m /r., STs 3.2-5 Rev. O, 09/28/92 i l QAC 9 ^ -Nf}) APLHGR B 3.2.1 i BASES l i i LC0 (continued) recirculationthe loops operating, the Ifmit is determined by cx i l:- =- : = ; . . x. - _ . . - . - _ . _ . - I / exposure dependent APLHGR limits. With only one ! recirculation loop in operation, in confomance with the h-requirements of LC0 3.4.1, " Recirculation Loops Operating," ' the limit is detemined by multiplying the exposure , x ' ' e.i M =;' mt. _ rAQ '- ' ct vahe#g~s--ngle' i recirculation) loop analys APPLICABILITY The APLHGR limits are primarily derived from fuel design @to occur at high power levels. evaluations and LOCA and trans Design calculations 6 and operating experience have shown that as power is , reduced, the margin to the required APLHGR limits increases. This trend continues down to the power range of 5% to ) 15% RTP when entry into MODE 2 occurs. When in MODE 2, the : Td intermediate range monitor (IRM) scram function provides ! 473 ' prompt thereby scram effactivel initiation during any significant transient, concern in MODE 2.y removing any APLNGR limit compliance Therefore, at THERMAL POWER levels 25% RTP, the reactor o the APLHGR limits; thus,perates this LCOwith substantial is not required.margin to ACTIONS M h If any APLHGR exceeds the required limi regarding an initial condition of the DBA and transient an assumption analyses may not be met. Therefore, prompt action is taken h to restore the APLHGR(s) to within the required limit's) such that the plant will be operating within analyzed conditions and within the design limits of the fuel rods. The 2 hour Completion Time is sufficient to restore the APLHGR(s) to Q witain probability a and is acceptable based on the low itsoflimitytransient ~ or 08A occurring simulta with the APLHGR out of specification. M If the APLHGR cannot be restored to within its required l - @ limit within the associated Completion Time, the plant must (continued) %. Ge J 9WRf6-SM B 3.2-3 Rev. O, 09/28/92 . _ .- - = _ . . - _ _ . GAR 93MDMCPR B 3.2.2 8ASES APPLICA8LE 1 SAFETY ANALYSES The MCPR operating limits derived from the transient (continued) analysis are dependent on the operating core flow and power state (MCPR, and MCPR,, respectively) to ensure adherence to (ntcf'Rp)] h 2 fuel with designfrequency moderate limits duringf ::: the worst transient MCPR limituare determined by steady state thermal hydraulic that occurs _ ' Flow dependent matho using " the three dimensional 8WR simulator code (Ref. . -- - "---- - -- - - -"--- ' ' - ~ ' Te? . '2. MCPRr curves are previoed based on the maximum #(p7 W -.' 6credible flow runout transient for Loop Manual On:: operation. 90: JL-- The result of a single failure or single operator error during Loop Manual operation is the runout of only one loop because both recirculanion loons are under indeoendent control. Jh:n L::p ".:: i : ;re.dene; ir.esm (aMew-54euttentous-runeet ef-both-loops;because-e-s4ngle ) -(con t rol-l e r-regu la tes-core-f4owy J Power dependent MCPR limits (MCPR ) are determined by the three dimensional 8 h transient code (Ref transient response imulator code and the one dimensional Due to the sensitivity of the nitial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, high and low flow "f , mentioned bypass powe.- level. y""" operating lim ,, g The MCPR satisfies Criterion 2 of e R olicy atement. LCO The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. larger of the MCPRr and MCPR, limits.The MCPR operati i APPLICABILITY The MCPR operating limits are primarily derived from ! transient levels. analyses that are assumed to occur at high power ! Below 25% RTP g J recirculation pump spee,dw= :::the reactor is operating at a slow } j G,i e r,c.nt s and the moderator void ratio is small.
- crer % =iv- in iw Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs. ;
(continued) %e< Or-.I ;;a/6 $TS-B 3.2 6 Rev. O, 09/28/92 (L4R G:-
- ( t '- l APRM Gain and Setp31nts 40pM SURVEILLANCE REQUIREMENTS b
SURVEILLANCE FREQUENC SR 3.2.4.1 QVerifyMs 6fthi : @ WY; f ,y,; (Once within r y.gousaf j gp * ,6 6 51 24 hours thereafter + ao ra - .- -- ._ -. 7/o+ re9 mired -4o be m e+ if SR 3.2.4.2 a sa+u Eed 4r- ico ,3,2.//, .y+ , , L,c recpiceA,n+3, Su r ir f t 4.@AJ f.d Flyo w g 3.2 . N.1 - - - - . - /VoT( ,,_ ,_ N4M rt%un'etd kt mej i[ $g
- b. 2.h. I 11 JEM 4 ed , / CO
- 3. 2.4 , Item a. r equicea ca 47 r
l fjf ' fer-Ik A PEN) se-fp A4s ,,- -. A .s . c e. ,J ,,wd L h. ec.t e +ed *" l M-AB A A A A A RwBed m " /0 CTS 3.2-5 Rev. O, 09/I (lkC 9 ^ && APLHGR B 3.2.1 BASES LC0 (continued) recirculation loops operating, the limit is determi_ned by ' the c or i i rr = f : = m c. =- --_ . -- c _ _ : = / exposure dependent APLHGR limits. With only one recirculation loop in operation, in confomance with the % requirements of LC0 3.4.1, " Recirculation Loops Operating," the limit is determined by multipl a vah M.de: :pendent APLHGR limit by 5;; :x r;ying MM ,J 2.=. r: r z r 22: i :: determined b:r a specific
- n. therrrr exposure
~~ gsingle' recirculation)loopanalysisqueez:@. i 3 APPLICABILITY The APLHGR limits are primarily derived from fuel design $to occur at high power levels. evaluations and LOCA and transie , i and operating experience have shown that as power isDesign c reduced, the margin to the required APLHGR limits increases. I This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs. When in M00E 2, the ; 18 intemediate range monitor (IRM) scram function provides prompt scram initiation durir.g any significant transient- { e73 ',. thereby effectively removing any APLHGR limit compliance l' concern in MODE 2. Therefore, at THERMAL POWER levels 25% RTP, the reactor o the APLHGR limits; thus,perater this LCO with substantial is not required.margin to ACTIONS M , h If any APLHGR exceeds the required limit regarding an initial condition of the DBA and transient an assumption analyses may not be met. Therefore, prompt action is taken Q to restore the APLPGR(s) to within the required limit's) such that the plant will be operating within analyzed conditions and within the design limits of the fuel rods. The 2 hour Completion Time is sufficient to restore the APLHGR(s) to Q witsin probability and is acceptable based on the low itsoflimitgtransient a or D8A occurring simultan with the APLHGR out' of specification. M If the APLHGR cannot be restored to within its required h limit within the associated Completion Time, the plant must _~ (continued) % Ge J 9WRMr-SM B 3.2-3 Rev. O, 09/28/92 Q A R 9 3 *'4' 0 MCPR ; B 3.2.2 , BASES APPLICABLE SAFETY ANALYSES The MCPR operating limits derived from the transient ' (continued) analysis are dependent on the operating core flow and power state (MCPRe and MCPR,, respectively) to ensure adherence to fuel design limits during the ~~ g with moderate frequency r:= worst transient - that occurs ~.' Flow dependent (#fC Pfd] MCPR limits.are determinec by steady state thermal hydraulic matho ; using the three dimensional BWR simulator code (Ref. .f : - :n ::- :: : :=: , ""2 12 . r:r_1 tr r - - ! MCPRe curves are prov:ided based on the maximum ; '3' # lp7 W --@ credible flow runout transient for Loop Manual (=0 ": LB operation. The result of a single failun or sin operator error during Loop Manual operation is the trunou% of i only one loop because both recirculauion loons are under independent control. m; Ln; ".= i g: retie..e; ~ .. N (alle ~(controller-regulates 4re #10w.Jtimu+teneous rer.e.1 of-so . h tr fen code Ref transient response to nitial core fl u o e ens "' d t# o sional 1 "1 power ; levels below those at which the turbi pnn osun and turbine control valve fast cl ' bypassed, high and low flow --- gsun scram trips are provided for mentioned hypass power level,operating betweeni57RTPandthepreviously - MCPRp . . (, tD The MCPR satisfies Criterion 2 of e R olicy atement. I LCO ! The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. larger of the MCPRr and MCPR, limits.The MCPR operat ' APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power ! levels. Below 25% RTP the reactor i ; g j recirculation pump spee,dcT:= ::: ::::s operating crvm - at - a' slow n its; /Gidii s.u ;wrp and the moderttor void ratio is small . Surveillance '67 thermal limits below 25% RTP is unnecessary - due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs. (continued) ! $Y.0..m, o .hb . . , - B 3.2-6 Rev. O, 09/28/92 P _ _. . - _ - -- - ~~~ ^ ~~ . 3 , APRM Gain and Setpoints B 3.2.4 , ' B 3.2 POWER DISTRIBUTION LIMITS 3 B 3.2.4 Average Power Range Monitor (APRM) Gain and Setpoints BASES BACKGROUND The OPERA 8ILITY of the APRMs and their setpoints is an d ' initial condition of all safety analyses that(I EEi% rod insertion upon reactor scram. Applicable GOCs are GDC 10, " Reactor Design"; GDC 13. " Instrumentation and Control"; GDC 20, " Protection System Functions"; and GDC 29 " Protection against Anticipated Operation Occurrences" (Ref. 1). Th s LCO is provided to require the APRM gain or APRM flow biased scram setpoints to be adjusted when operating under conditions of excessive power peaking to maintain acceptable margin to the fuel cladding integrity Safety Limit (SL) and the fuel cladding 1% plastic strain limit. FThe condi on of e ssiv powe peak gisdgernine by' the ra of the ctual ower eaki to the initi power b yg pea g at RTP This atio s equ the-l 6 MA '" c limiti MFLPD o th ract n of to th ratio RT (FRTP where F P 1 s the me ured T L POWER d ided by he RT.. Exces ive i power p,esking ex 'sts wh d) / -
- /
b MFLP .s79 ,g' F indicatinji that is not decr sing oporti atelylo 1 the overd'l po educ on, or onvers y, th power ! peakiruf is in easing. To mai ain ma gins s flar t those atIJ7Pcondi ons, th excess e pow peaki g is / - compensate by gain djustne on t APRM or adjustment o'f. ' ,the APRM etpoints Either of the adju ments s effecti ly the resu as na taini g MFLPD ess an / or eq to FRT and thu maint ns RTP margin for A R (and PR. / The normally 1 elected APRM setpoints position the scram above the e,,per bound of the normal power / flow operating region t'.at has been considered in the design of the fuel t rods. The setpoints are flow biased with a slope that approximates the upper flow control line, such that an i approximately constant margin is maintained between the flow biased trip level and the upper operating boundary for core flows in excess of about 45% of rated core flow. In the range of infrequent operations below 45% of rated core flow, (continued) { ,$,,YS B 3.2-12 Rev. O, 09/2'8/92 i (MKD-!vtQ INSERT B12A The condition of high power peaking is determined by the ratio of the-actual power peaking to the limiting power peaking at Rated Thermal Power (RTP) . This ratio is At any off-presented as a APRM setpoints T-factor (T). ,- rated power conditions T is equal to the Fraction of Rated 4,7 Thermal Power (FRTP) divided by the Core Maximum Fraction of Limiting Power Density (CMFLPD). FRTP is the measured CMFLPD is the limiting
- 1g thermal power divided by the RTP.
g linear heat generation rate (LHGR) divided by the rated LHGR limit. High power peaking exists when: FRTP CMFLPD As power is reduced with the design power distribution maintained, CMFLPD is reduced in proportion to the reduction t in power. However, if power peaking increases above the design value, the CMFLPD is not reduced in proportion to the reduction in power. Under these conditions, the APRM gains are adjusted upward or the APRM When theflowreactor biased isscram setpoints operating with are reduced accordingly. peaking less than the design value, it is not necessary to modify the APRM gains or flow biased scram trip setpoints. Adjusting the APRM gains or setpoints is equivalent to ! maintaining CMFLPD less than or equal to FRTP. I t + s INSERT B 3.2-12 10/27/94 RIVER BEND (bd9&ivRI) APRM Gain and Setpointo p );p f T '1 /. 0 rafber 1b"" 8 3'2'# 8ASES s})u,fkg de,APM x&xl~4 y, y),f7 R, g - yj 4f ~, LC0 { ' xMaM/,) /
- c. A the APRM gains e riNhe(APRM 25'-="~'
(continued) g :::er ;r.er. Icsp; := "n=. This Condition is to account for tne reduction in margin to the fuel 7 -- cladding integrity SL and the fuel cladding 1% plastic jru,i,/a3 mo// strain limit. ff "'f_" 40 * "!", f g/ t l ' FLPD is the ratio of the limiting LHGR to the LHGR limit 7 for the specific bundle type. As power is reduced, if the AF//51 ren,4, u design power distribution is maintained, MFLPD is reduced in " proportion to the reduction in power. However, if power i * # peaking increases above the design value, the NFLPD is not l#" L9 74f reduced in proportion to the reduction in power. Under these conditions, the APRM gain is adjusted upward or the T" ggy 1 ' APRM flow biased scram setpoints are reduced accordingly. , dgW When the reactor is operating with peaking less than the ' design value, it is not necessary to modify the APRM flow ,f biased scram setpoints. Adjusting the APRM setpoints is equivalent to maintaining (S; gain ;;22or:::= :$ # ii ? /,0 $pn! t; raTD as stated in the LCO. s \ ; ',, For compliance with LCO Item b (APRM setpoint adjustment) // or Item c (APRM gain adjustment), only APRMs required to be j OPERABLE per LCO 3.3.1.1, " Reactor Protection System (RPS) '# 7' g Instrumentation," are required to be adjusted. In addition, each APRM may be allowed to have its gain or setpoints ' adjusted independently of other APRMs that are having their / s gain or setpoints adjusted. I V4 I APPLICABILITY The @limit, & APRM gain adjustment, or APRM flow biased scram and associated setdowns are provided to ensure that I the fuel cladding integrity SL and the fuel cladding 1% I plastic strain limit are not violated during design basis transients. As discussed in the Bases for LCO 3.2.1 and l LCD 3.2.2 sufficient margin to these limits exists below , 25% RTP and, therefore, these requirements are only necessary when the plant is operating at t 25% RTP. \ \ ACTIONS L,1 'D If the APRM gain or set soints are not within limits while Q. / 4 /+%'4e ;;FUL nas wu...AJ' "El the margin to the fuel cladding integrity SL and the fuel cladding 1% plastic strain limit (continued) I((C B 3.2-15 Rev. O, 09/28/92 {/_Af93vd/) APRM Gain and Setpoints B 3.2.4 BASES ' ACTIONS M (continued) . may be reduced. Therefore, prompt action should be taken to store h m 7f2'io within its required limit or make c,3 within the assumed margin of the safety analyses. The 6 hour Completion Time is normally sufficient to restore ' either the MFLPD to within limits or the APRM gain or setpoints to within limits and is acceptable based on the low probability of a transient or Design Basis Accident occurring simultaneously with the LCO not met. M If the APRM gain or setpoints cannot be restored to within their required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours. The allowed Com on operating experience, pletion Time is reasonable, based to reduce THERMAL POWER to g < 25% systems.RTP in an orderly manner and without challenging plant SURVEILLANCE \ SR 3 REQUIRENENTS kb bot.2.4.1 ed SR 3.1. 4. '2. 1fonpared tomCDis a. APRMre uired to be calculated "=- M 5:-- - 'and gain or setpoints 'to ensure that Th jeeSRs3 safety the reactor'1s operating within the assumptions of the -j analysis h m : - e required only to determine b ppropriate gain or setpoi nd is not intended to be a ' ~ 1 Mr)#o f CHANNEL FUNCTIONAL TEST fo e APRM gain or flow biased ay'.~ 3 c6 f i
- e a 4ya f , neutron flux scram circuit #--- - -~~ -- '
^ ^ - " ~ " 6 The 24 hour Frequen is chosen to coincide with the .e ' g g determination of other thermal limits, specifically those for the APLHGR (LCO 3.2.1). The 24 hour Frequency is based ..f ' on both engineering judgment and recognition of the slowness \ ('s. of changes in power distribution during normal operation. s a; The 12 hour allowance after THERMAL POWER t 25% RTP is s achieved is acceptable given the large inherent margin to operating limits at low power levels. -- 3 % %;,fy,Q f 51E3.2.4.1 %aesopp- beveh s.,u m ,g - m =". pg7 x i 1, ' y , s ie ra V c 6 gea r p~e, A&l& uF i 0.veo 0J (continued) 44p%tt1 eq u k.). . . - - Ta/C STS- Q # B 3.2-16 Rev. O, 09/28/92 (EM 93-/& t APRM Gain and Setpoints B 3.2.4 8ASES (continued) R'fFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GOC 13, GDC 20, and GDC 29. h 2. hR, = xx . A Choke- 4, Appendix dB, h 3. R, 6 Chaphe IT Append lv/58.3 3 . ,7 - aeA up S -rh o,,ac .e xa : by.u, %-a, (-//yAaA .1%Mihn ia .91W er Q f,,,," %%h 4 h l B 3.2-17 Rev. O, 09/28/92 (isa e iaD RIVER BEND SECTION 3.3 ATTACHMENT 1: CTS-PSTS COMPARISON DOCUMENT ATTACHMENT 2:ITS-PSTS COMPARISON DOCUMENT [4x vs-naD l ATTACHMENT 1 ITS - PSTS COMPARISON DOCUMENT REVISION 1 SECTION 3.3 REVISED PAGES 1 A: MARKUP OF CTS 18: DISCUSSION OF CHANGES 1C: NO SIGNIFICANT HAZARDS CONSIDERATIONS l _.4, - ,---- fueoa-,vW ATTACHMENT 1 A CTS - PSTS . I COMPARISON DOCUMENT REVISION 1 MARKUP OF CTS l 1 i l l TABLE 3.3.2-1 (Continued) h ISOLATION ACTUATION INSTRtSIENTATION VALVEGROUPS MINIDRS4 APPLICARLE m OPERATED RY OPERARLE CHA8 RIELS OPERATIONAL TRIP FINICTION SIGNAL *** PER TRIP SYSTEM (a) COISITION ACTION gg REACTOR CORE 150 TAT 1000 q f5Kfitf3VETITT55EATY0es (continue (d);
- 1. Main Steam Line Tunnel Temperature Timer 2 1 1.2,3 X F/H J. RHR Equipment Room Ambient Temperature - High 2 1 1,2,3 .M"F/H
- k. RHR Equipment Room a Temperature - High 2 1 1,2,3 47 F/H R
- 1,2,3 , N F /rt
- 1. RHR/RCIC Steam Line Flow - 2 -
1
- ' High
- m. Drywell Pressure - High 3 I8) 1 1,2,3 ,p r[p
- n. Manual Initiation 2 II 1 1,2,3 5Z RHR SYSTEM 1504AT1000 .
- a. RHR Equipment Area Ambient
' Temperature - High 5, 14 2 1,2,3 # F/H ti. RHR Equipment Area a . Temperature - High 5, 14 2 1,2,3 .3eF/H .
- c. Reactor Vessel Water 3e" F/8 [%%I b Level - Low, tevel 3' 5, 14 2 1,2,3 ~
- d. Reactor Vessel Water Level -
Ql, S) - - Q* ' ( '21,? 2 ./ b Low Low tow, Level 1 10 2 1, 2, 3 30'F/H s L %v > '3.'Z. Vr d sif .B 0AK93-!1N.I) g.3 3 3 ,, ; (SR 5 3. (, . l .D ~,.I. .f.Ae.# D LCo 11 " < ,,. ;,,e . ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
- 1. PitIMARY CONTAIMENT ISOLATION 3
- a. Reactor Vessel Water Level - Low Low Level 2 ,
b. c. Drywell Pressure - High 7[<10(N) 7_ J 10 8 Containment Purge Isolation Radiation - High(b) g ;g i 4,1ff
- 2. MAIN STEAM LINE ISOLATION y
- a. Reactor Vessel Water Level - Low Low Low -
J.aval 1 7-71.0 */< lo(a)..
- b. ~
4 3h'g c. Main Steam Line Radiation - HiahW Main Steam Line Pressure - Low <l.0"/iiu'h-) < l.0 "/ 10
- d. Main Steam Line Flow - High d e. Condenser Vacuum - Low 3_J0.5*/I10")'""*)
7-m C * - f. Main Steam Line Tunnel Temperature - Hiah NA , _g. Main Steam Line Tunnel a Temperature - H_ich '7A08 NA3 ~ h. Main Steam Line Area Temperature - High (Turbine Bldg) NA 47f j I
- 3. SECONDARY CONTAIMENT ISOLATION
[f *; j a. b. Reactor Vessel Water Level - Low Low Level 2 Drywell Pressure - High G !<10(*) h, g t c. Fuel Buildina Ventilation Exhaust Radiation - High(b) 6-'m 510kl '
- d. Reactor Building Annulus Ventilation Exhaust Radiation - High(b) N 1
NA ) 7' 1 4. REACTOR WATER CLEANUP SYSTEM ISOLATION g! _ g, /\ a. A Flow - High 1 10(a) \
- b. A Flow Timer m I c. Equipment Area Temperature - Hiah NA /f g G, Equipment Area a reaperature - High ,g p
e. f. Reactor Vessel war,er Leyal - Low Low Level 2 Main Steam Line Tunnel Ambient '1 10 jg Temperature - Hiah NA I a, main S* - Line Td el a T e erature - High T i h. SLCS Initiation MA l
- 5. REACTOR CO M ISOLATION C0OLING SYSTEM ISCLATION
) a. b. RCIC Steam Line' Flow - High RCIC Steam Line Flow-High Timer lh10(*)# t c. RCIC Steam Supply Pressure - Low 7 1_10 ,
- d. RCIC Turbine Exhaust Diaphragm Pressure - High im i e. RCIC Equipment Room Ambient T erature - Hiah NA
. RCIC Fau1- t Rc s a T = erature - Hiah MP
- g. Main Steam Line Tunnel Ambient Temperature - Hiah 4A t
Yh. Main Steen Line Tunnel a Temperature - High. j RIVER BEND - UNIT 1 3/4 3-24 ' (?AR 92-/YAT.,) LCO 3.3 . t., t ~ L4 c 2 1 '. 2. p 'l :.3.2-? (Paau g ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)# - 1. Main Steam Line Tunnel Temperature Timer N *. RHR Equipment Room Ambient Temperature - High NA J. 7' 3'b'I dk. RHR Equipment Room & Temperature - Mign ~ n5> NA isfr p l 1. RHR/RCIC 5 team Line Ficw - Mign Drywell Pressure - High NA 'g e. '\ NA
- n. Manual Initiation s 6. RHR SYSTEM ISOLATION y jg
\ RHR Eculosent Area Ambient Temperature - High NA t , a. RHR Equi-A .t Area A T= rature - man j c. Reactor Yessel Water Level - Low Level 3 p <,10 , l I
- d. Reactor Vessel Water Level - Low Low Low < 10(*)
Level 1 ~ h- ~ /g
- e. Reactor Vessel (RHR Cut-in Persissive) NA Pressure - High /
- f. Drywell Pressure - High NA jg NA 2 /8 g
. 7. MANUAL INITIATION ~ gl C f ase 4 t(a) Isolation system instrumentation response time specified includes the c Q generator starting and sequence loading delaysf g ! I (b) Radiation detectors are exempt from response time testity.1) Response time /, s 11 be measured from detector output or the input of tw first electronic l LI Qomponent in the channel.f f " Isolation system instrumentation response time for MSIVs only. No diesel 7 generator delays assumed. 4/8 !;;i"icr. e,eb '. ;tre-teti^a **=aansa time for associated v= rx m . - J xM h ~ solation system instrumentation response time specified for the Trip @Functionactuatineeachvalve_groupshallbeaddedtoisolationtime,. _: u ; .; " for valves in each valve group to obtain j 6 Te,1.e 10N SYSTEM RESPONSE TIME for each valve. j ## Time delay of 45-47 second.s. #ffTime delay of 3-13 seconds. L 6 RIVER SEND - UNIT 1 3/4 3-25 , Takle 33.2.1-1 % Nkup . 6 Tc.bk 3.3.b-1J ; g CONTROL 200 SLOCK INSTALSENTATION SURVEILLANCE REQUIRE 9ENTS ,, S R I. T. t . s . x m CMieEL 2W'Y h OPERATION E -- E CHA881EL FUNCTIONAL CHAfGEL i CONDITIONS IN tellCN i
- TRIP FLAICTICII CHECK TEST CALIBRATI SURVEILLAfeCE REQUIMS ,
- 1. 800 PATTERN C011TA8L SYSTEM -
s L Q - a. Low Power Setpoint (b e) gg g g,Q l @ G b e @ 4y 1, 2 l l
- b. High Power Setpoint f tj 4 { l
- 43 __ _ g, I ' 1** j l
- 2. Apen
- a. Flow Blased Ilectron Flux .
$ scale WQ SA(g) ) b. Insperative 90 4 NA S/U((b) S/U b)'K Q O fA I 1, 2, 5 b)
- c. Downscale NA SA
- d. Itsutron Flum - Upscale, Startup NA S/U(b)
S/!I( g'# 4 h SA I 2, 5
- 3. 50MCE RAMIE IENIITORS g
- a. Betector met full in
- b. $acale NA 10 4 S/U ,W IIA SA 2,
2, 5 5 k
- c. Insperattwo NA S/U b)'W NA 2, 5 g
c
- d. hale NA S/U(b)*W S/U( ,W SA 2, 5 t4 ,
- 4. IIITENESIATE SAIIGE IIDelliets k
- a. Betector not full in - NA S/U .W 10 4 2, 5 h!
g b. $ scale IIA S/U(b),W SA 2, 5 e c. Insperattwo 11 4 S/U(b),W IA 2, 5 4p
- d. Sewascale 10 4 S/U ,W SA 2, 5 ;
=> 5. Stampe DI N VOLISIE h y a. Water Level-High 80 4 E 1. 2, 5* 9 ;
- 6. REACTOR C00LABIT SYSTEN RECIRCULATIcel FLOW W
' E4 RECEIVED S/u(b) $g(s) g
- t
- a. upscale NA ,_ ,
MAY 131988 _ i (Li G 9 3 - /Y n i j REACTIVITY CONTROL SYSTEMS @ 3. 3. 2, l R00 PATTERN CONTROL SYSTEM , , LIMITING CONDITION FOR OPERATION 1 ACTION (Continued) / The position and bypassing-of an inoperable control rod (s) is verified by a second Ifeensed operator or other technically ' M 33. 2* I* g - qualified member of the unit technical staff. - SURVEILLANCE REQUIREMENTS SR 3. 3.2. l . . 0.1.t.: "The RPCS shall be demonstrated OPERA 8LE by verifying the OPERA 81LITY of the: S R J.J. 2.1.5' /. . Rod pattern controller functions when THERMAL POWER is less than the low power setnointJ5y.selectina ano attempting to move an inntatte (E6ntrol rod: / . (After withdrawaI)of the first insequence control rod.or gang pq for eacn reactor startup. ssoona)therodinhibitmodeisautomatically initiat at , 6R 3 3 2 I- f.he RPC5 'ow power setpoint, 20 + 15. - 03 of RATED THERMAL ' P ER, during power reduction.__J
- 3. The first time only that a banked position, N1, N2, or N3, is
, reached during startup or during power reduction below the RPCS low power setpoint. f SR 3.3. 2. I. 5 ,3,y, 4. t Rod withdrawal limiter functions _when THERMAL POWER is greater than or g g ecual to the low nower setnaintIby selecting ano attempt 1ng to move a j (restricted control rod in excess of the allowaple distance: L A2. i $R. 3.3. 2. l. I ff - . seachpowerrangfabovetheRPCSlowpowersetpointTsentered " fR 5.12.l. Mol'E@auring a power increas d crease.
- 92. 4 S R 1 3.2..l. I
/. At least once per days while operation continues within a given power range above the RPCS low power setpoint. SR. J. 3. t.. l.7 2NSca.T' RIVER BENO - UNIT 1 '0/0 1-10 3/4 J -64 m W. 0 M 93 /Ydt ) Le_o 3.3. l. 2 REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) b Performance of a CHANNEL FUNCTIONAL TEST: fOt* $ 7'3.l.lf
- 1. @ithin24hourspriortothestartofCOREALTERATIONS)and
- 2. cat lecst once per ~-
lc. Verifying that the channel count rate is at least 0.7 cps *: (1. Priortocontrolrodwithdrawal}h 5031.2.4 2. (Prior to Mt least once per 12 hours during CORE ALTERATIONS, ana
- 3. At least once per 24 hours, except that:
y .,
- a. During fuel unloading, the required count rate may be permitted to be less than 0.7.cys". -
o , L b. Prior to and during fuel loading, until sufficient fuel 5R 33.1.2 4 has been loaded to maintain at least 0.7 eps", the requirer g count rate may be 4chieved by: . a) Use of portable external source, or ~~ Li b) Loading up to 2 fuel assemblies in cells containing inserted control rods around an SRM. >
- d. Verifying within 8 hours prior to and at least once per 12 hours duringthetimeanycontrolrodiswithdrawnNthattheRPScircuitry]l
" shorting links" have been removed unless adequate shutdown margin ) L2 has been demonstrated per Specification 3.1.1. j n Em6T 5 C. 3.3. l . h . ' 2. :. l. 2 .//D ls~R 3.3. l.2 4 }Providedsignaltonoiseratio3,2,otherwiseuse3.0 cps. ot requbed for control rods removed per Specification 3.9.10.1 or 3.9.10.2] RIVER BEND - UNIT 1 3/4 0 ? .3/+ 3 -96 m s.*- _a_a 2 , ATTACHMENT 1B CTS - PSTS COMPARISON DOCUMENT REVISION 1 DISCUSSION OF CHANGES l QAK 93-iv2D l I i DISCUSSION OF CHANGES ; CTS: 3.3.1 - RPS INSTRUMENTATION l TECHNICAL CHANGE - LESS RESTRICTIVE I (continued) l L.4 Applicability has been modified to only require RPS functions to be OPERABLE in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. Control rods ) withdrawn from a core cell containing no fuel assemblies do not I affect the reactivity of the core and therefore are not I required to be OPERABLE with the capability to scram. Provided all rods otherwise remain inserted, the RPS functions serve no purpose and are not required. In this condition the required SHUTDOWN MARGIN (LCO 3.1.1) and the required one-rod-out interlock (LCO 3.9.2) ensure no event requiririg RPS will occur. The ACTIONS for inoperable equipment in MODE 5 are also revised 1 to be consistent with the proposed Applicability. Since all I control rods are required to be fully inserted during fuel movement (LCO 3.9.3) , the proposed applicable conditions cannot be entered while moving fuel. The only possible core alteration is control rod withdrawal which is adequately addressed by the proposed ACTION. L.5 The main steam line radiation monitor (MSLRM) scram function has been removed from the Technical Specifications based on the guidelines provided by General Electric NEDO-31400A, " Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor. " This technical report provided the results of generic evaluations which indicated the MSLRMs are unnecessary to ensure compliance with the radiation dose guidelines of 10 CFR Part 100. EOI has confirmed that the e analyses presented in NEDO-31400A are applicable to RBS and bound the results of analyses for RBS. Additionally, the MSLRM is not credited for a reactor scram initiation for any design basis event. Finally, the reliability assessment of the i g,3, /,I elimination of the scram function on reactivity control failure q frequency and core damage frequency indicate a net improvement , in safety. l s EOI confirms that, upon disabling of the MSLRM scram function, 's the conditions identified in the NRC's SER for NEDO-31400A will be implemented as identified below. With the implementation of the technical report (NEDO-31400A) guidelines, the main steam line radiation monitor and offgas radiation monitor alarm ! setpoints will be standardized above the nominal background I dose rate to provide the indication of need for a prompt sample I of the reactor coolant to determine possible contamination I levels in the plant reactor coolant and the need for additional ! corrective action. Any significant increases in the levels of radioactivity in the main steam lines will be expeditiously controlled (by procedure) to limit both occupational doses and environmental releases. /0 9Y RIVER BEND 6 10/1/0 W 4 QAR 93 M h i DISCUSSION OF CHANGES' ' CTS: 3.3.2 - ISOLATION ACTUATION INSTRUMENTATION TECHNICAL CHANGE - MORE RMSTRICTIVE * (continued) M.2 ' This consnent numoer is not used for this station. ; 'M.3 The addition of Condition H provides specific guidance of the ACTIONS if the citrrent- ACTION cannot be met. Currently this would require ent:ry into LCO ,3.0.3 which allows one hour to initiate the shutd.own; _under Condition H, shutdown is required, but the one hour to initiate the shutdown is omitted. M.4 An additional requirement is included as Required Action J.3 f M ' that requires action be initiated to restore the prin try M ~ , containment boundary if the RHR Shutdown Cooling System is not isolated or the instrumentation is not restored to OPERABLE ' status. , M.5 The proposed Applicability has added a requirement for Reactor l Vessel Water Level - Low Level 3 instrumentation to be OPERABLE ' for the RHR System isolation logic. Since this is not presently a requirement, this change is more restrictive.: 1 M.6 This comment number is not used for this station. ! TECHNICAL CHANGE - LESS RESTRICTIVE ! " Generic" t LA.1 Testing of the response time is provided by a specific SR and i is an integral part of the OPERABILITY of certain ' instrumentation channels. Details of the methods for ! performing this and other surveillances and Required Actions l are relocated to the Bases and procedures. The design features ! and system operation which dictate the methods are described in , the USAR. Additionally, changes to the Bases- will be i controlled by the provisions of the proposed Bases - Control Process in Chapter 5 of the Technical Specifications. ; LA.2 System design and operational details -have been relocated to ! the Bases and procedures. Trip setpoints are an operational ;
- - detail that is not directly related to the OPERABILITY of the !
instrumentation. The Allowable Value is the required j limitation for the parameter and this value is retained. i Details relating to system design and operation. (e.g., ; commonality with RPS, bypasses, specific valves or valve groups i affected, etc.) are also unnecessary in the LCO and have been : i relocated to the Bases and procedures. The design features and ' system operation are also described in the USAR. Changes to , l the Bases will be controlled by the provisions of the proposed l Bases Control Process in Chapter 5 of the Technical l Specifications. j j 9D u/'?f RIVER BEND 15 ' M ^ l' " # i J,4.3 cod ^ U. Y new.(2 h Yr f a;e.- (L4f( 93*/YAb' l b , Av ' h s0f./re;. ol 14.'l oN >aye / ir DISCUSSION OF CHANGES CTS: 3.3.2 - ISOLATION AC."TUATION INSTRUMENTATION { TEC'HNICAL f'HANGE - LWAS DRATRICTIVE (continued) , LA.3 This comment number is not used for this station. LA.4 This' comment number is not used for this station. LB.1 This coment number is not used for this station. LB.2 The note is clarified to provide direct indication of the ; intent of the current wording. Providing "at least one : OPERABLE channel in the same trip system ... monitoring that parameter" is' intended to assure that the trip capability of * . that function is maintained. However, it does not provide this i assurance for all logic system designs. The proposed Note is ; based on previously conducted reliability analyses (NEDC-31677-P-A and NEDC-30851-P-A, Supplement 2). ; i " Specific" l L.1 The Frequency for the Channel Functional Test of the SLCS 'initiation logic is extended from every 6 months (i.e., every 92 days on a STAGGERED TEST BASIS) to 18 months. This function is manually initiated and the testing is very similar to the + testing ~of the manual initiation function. The 18 months is based on the potential for an unplanned transient due to the - loss of the reactor water cleanup . system flow when . the j surveillance is performed with the reactor at power. In ! addition, operating experience at other plants has shown these components usually pass the surveillance when performed at this frequency. ! L.2 This comment number is not used for this station. \ ~ L.3 This comment number is not used for this station. L.4 The Required Action if the Required Action and associated Completion Time of Conditions A or B are not met is proposed to , allow isolation of the affected main steam line (s) . Some conditions may affect the isolation logic for only one main i steam line. In these cases, it-is not necessary to require a i shutdown of the unit; rather, isolation of the affected line- l
- returns the system to a status where it can perform the i
- , remainder of its isolation function, and continued operation is {
- allowed (although it may be at a reduced power level.) )
l \ l l RIVER BEND 16 vhhv 10/1/ C , . -. .- - , . . - - . .x . (UK 93 MI) l DISCUSSION OF CHANGES CTS: 3.3.2 - ISOLATION ACTUATION INSTRUMENTATION TECHNICAL CHANGE - LESS RESTRICTIVE (continued) L.5 The main steam line radiation monitor (MSLRM) scram and MSL isolation functions have been removed from the Technical Specifications based on the guidelines provided by General / Electric NEDO-31400A, " Safety Evaluation for Eliminating the , Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation i Monitor." This technical report provided the results of i generic evaluations which indicated the MSLRMs are unnecessary to ensure compliance with the radiation dose guidelines of 10 I CFR Part 100. EOI has confirmed that the analyses presented in 3 3, kV NEDO-31400A are applicable to RBS and bound the results of analyses for RBS. Additionally, the MSLRM is not credited for i a reactor scram initiation or MSL isolation for any design 47 , basis event. Finally, the reliability assessment of the , elimination of the scram function on reactivity control failure frequency and core damage frequency indicate a net improvement in safe' y.
- EOI confirms that, upon disabling of the MSLRM scram function, the conditions identified in the NRC's SER for NEDO-31400A will be implemented as identified below. With the implementation of the technical report (NEDO-31400A) guidelines, the main steam line radiation monitor and offgas radiation monitor alarm setpoints will be standardized above the nominal background dose rate to provide the indication of need for a prompt sample of the reactor coolant to determine possible contamination levels in the plant reactor coolant and the need for additional corrective action. Any significant increases in the levels of radioactivity in the main steam lines will be expeditiously controlled (by procedure) to limit both occupational doses and environmental releases.
L.6 The ACTION to isolate all main steam lines is a sufficient ACTION with the referenced functions inoperable and will require being in MODE 2 to avoid a scram. The requirement to be in MODE 2 is therefore implicit and is deleted. The time allowed to isolate the associated main steam lines is extended from 6 hours to 12 hours. The additional time is provided to allow for more orderly power reduction. Since 36 hours are otherwise provided to exit the applicable MODES, this short extension is still well within any reliability or availability assumptions. L.7 This comment number is not used for this station. RIVER BEND 16 g, m/rk 24/1/^3 &R 93-/Y& DISCUSSION OF CHANGES CTS: 3.3.2 - ISOLATION ACTUATION INS _RUMENTATION TECHNICAL CHANGE - LESS RESTRICTIVE I (continued) I L.8 The manual initiation function is not assumed in any accident or transient analysis in the USAR and isolation of the penetration due to loss of this function within one hour is considered overly conservative. The time allowed to isolate the line is extended to 24 hours to provide time for adequate preparation to accommodate the effects of isolating the penetration. (Note that the time allowed to isolate the penetration was reduced from 48 hours to 24 hours for other manual initiation functions. See discussion of change M.1.) L.9 This comment number is not used for this station. L.10 This comment number is not used for this station. l L.11 Options are provided that would allow isolation of the af fected ' lines, or for secondary containment, either operation of "the i SGTS or a declaration of an inoperable SGTS. These ACTIONS conservatively compensate for the inoperable status of the instrumentation through restoration of the single failure capability or through providing the required instrumentation actuation function. Therefore, providing this option does not impact safety. L.12 This comment number is not used for this station. L.13 This comment number is not used for this station. L.14 This comment number is not used for this station. L.15 The SLCS is not required during MODE 3 since no control rods can be withdrawn and adequate SHUTDOWN MARGIN prevents I criticality under these conditions. This is consistent with the Applicability requirements for the SLCS. L.16 This comment number is not used for this station. l <3, % / L.17 This comment number is not used for this station. ; /3 L.18 The proposed change deletes primary containment isolation instrumentation and secondary containment isolation instrumentation response time testing where the required G, N1 response time corresponds to the required diesel start time. d3I The bases for this specification is also revised consistent . with deleting this testing. These changes are revised based on guidance provided in Generic Letter 93-05, Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements I for Testing During Power Operation, September 1993. l As described in the current Technical Specifications bases, I /0/Y/9"f RIVER BEND 17 -it/.L/93 1 k-93-HeQ' DISCUSSION OF CHANGES CTS: 3.3.2 - ISOLATION ACTUATION INSTRUMENTATION TECHNICAL NANGE - Lung punTRICTIVE (continued) with exception of the Main Steam line Isolation Valves (MSIVs), , individual sensor response times or the response times of the logic systems ~to which the sensors are - connected are not addressed in the safety analysis. For D.C. operated valves, a 3 second delay is assumed before the valve starts to move. For A.C. operated valves, it is assumed that the A.C.' power supply is lost and is restored by startup of the emergency diesel generators. In this event, a time of'10 seconds is assumed before the valve starts to move. In addition to the pipe break, the failure of the D.C. operated valve is assumed; thus , the signal delay (sensor response) is concurrent with the 10 second diesel- startup. The safety analysis considers an allowable reactor coolant inventory lose in each case which in turn determines the valve speed in conjunction with the 10, second delay. It follows that checking the valve speeds and the 10 second time for emergency power establishment will establish the response time for the isolation functions. Thus, the signal delay (sensor response) is concurrent with the 10 second diesel startup. Since typical response times are 7 measured in fractions of a second, the chance is remote that a channel's response would degrade to the point where it exceeds - / the 10-second diesel start time without a noticeable failure. { Therefore, the proposed changes are justified. t I ( \ _ s g , p,/ - \ $$$ s \ g , % G. 9 #3[ r Mf9'bY ' RIVER BEND 18 40/M f @ NkD DISCUSSION OF CHANGES CTS: 3.3.6 - CONTROL ROD BLOCK INSTRUMENTATION ,3 ) b I hI TECHNICAL CHANGE - LESS RESTRICTIVE (continued) g L.6 ITS SR 3.3.2.1.9 allows control rods to be bypassed and s repositioned under the direction of a second licensed operator N or other qualified member of the technical staff. This allowance is currently only applicable (per CTS 3.1.4.2 ACTION b) for insertion of the bypassed control rod. CTS Special Test Exception 3.10.2 allows a bypassed control rod to be withdrawn to its prior position to support special tests, including scram time tests. The ITS has expanded this flexibility to allowing the control rod to be bypassed with the positioning of the control rods controlled by other Specifications. Namely, LCO 3.1.4 does not allow slow or stuck control rods to occupy adjacent locations. With respect to the Control Rod Drop Accident (CRDA), which is only of concern below the Low Power Setpoint (LPSP) of the Rod Pattern Control System (RPCS), LCO 3.1.6 requires OPERABLE control rods to be in compliance with the Banked Position Withdrawal Sequence (BPWS) analysis while LCO 3.1.3 Condition D requires inoperable control rods to either comply with the BPWS analysis or be separated by at least two core cells. This ensures that the control rods remain within the patterns assumed for the CRDA analysis. With respect to the Rod Withdrawal Error (RWE) accident, which is only applicable above the LPSP of the RPCS, adequate separation of the control rods is assured by the thermal operating limits as discussed in CTS 3.1.3.1 DOC L.1. As stated above, the ITS has expanded the flexibility for bypassing and moving bypassed control rods. Implicit in the requirement that movement of the bypassed control rods be performed under the direction of a second licensed operator or other technically qualified member of the staff is that the positioning be in conformance with applicable safety analysis. When operating below the LPSP of the RPCS, the applicable analysis is the CRDA. Compliance with this analysis is ensured by conformance with the generic BPWS analysis or a specific BPWS analysis for the evolution. Similarly, when operating above the LPSP of the RPCS, the applicable analysis is the Rod Withdrawal Error (RWE) analysis. Movement of bypassed control rods must be in conformance with the generic RWE analysis or with a special analysis to ensure that the conclusions of the RWE analysis remains supported. The above controls ensure that positioning and movement of bypassed control rods remain within the bounds of previous analysis. RIVER BEND 32 8/29/94 n (J.AR 93-/vfD DISCUSSION OF CHANGES . CTS: 3.3.7.6 - SOURCE RANGE MONITORS TECHNICAL CHANGE - LRAS RESTRICTIVE (continued) L.6 As identified in DOC R.1 for CTS 3.3.6, the requirements for the SRM Control Rod Block Functions have been Relocated from the Technical Specifications to licensee controlled documents as they do not satisfy the NRC Policy Statement technical specification screening criteria. As a result, as further supported by the Bases for ITS LCO 3.3.1.2, the function of the SRMs, as required by the ITS is to provide the operators with information on neutron flux levels at low power levels and during shutdown conditions, including-CORE ALTERATIONS. The ITS does not require that the SRMs provide any trip functions, only indication. These functions are adequately tested during
- MODE 5 operations via SR 3.3.1.2.1 which provides for a CHANNEL CHECK at least.once per 12 hours with adequate count rate (per SR 3.3.1.2.4) and by SR 3.3.1.2.7 which provides for a CHANNEL CALIBRATION at least once per 18 months. Since the SRMs do.not provide any trip functions and there are no additional functional requirements for the SRM channels while in MODE 5 than there are in MODE 2, 3, or 4, performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days as provided by NUREG-
/ 1434 SR 3.3.1.2.6 provides sufficient assurance of the / functional capability of the required SRM channels. I \ 1, 3 ,/, 2 $/o -f RIVER BEND 46 g/;/A _w.. -_a i.- ., a a__n.- _ . ~ _ a - . _ ATTACHMENT 1C CTS - PSTS COMPARISON DOCUMENT REVISION 1 NO SIGNIFICANT HAZARDS CONSIDERATIONS , \ r w mw w -m-- - --- (MR 93-N& g,$k d #34 NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.3.2 - ISOLATION ACTUATION INSTRUMENTATION i aL18" CHANGE I Entergy Operations Inc., has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. O, The following evaluation is provided for the three categories of the g, W O I -significant hazards consideration standards: -
- 1. Does the change involve a significant increase in the probability or consequences of an accident. previously evaluated?
Individual sensor responee times for the affected instruments are not addressed in the accident analysis. A 10 second time delay is. assumed prior to isolation valve movement. Thus the signal delay (sensor response) is concurrent with the 10 second diesel startup. Since typical response' times are measured in fractions of a second, the chance is remote that a channel's , response would degrade to the point where it exceeds the 10 second diesel start time without a noticeable fail 6re. Therefore, this change will not involve a significant increase in the probability or consequences of any accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change introduces no new mode of plant operation and it does not involve physical modification to the plant. -> Response time testing on the affected instruments is only intended to enhance system reliability and to monitor-instrument channel response time trends. The response times are concurrent with the 10 second diesel startup. Therefore it , does not create the possibility.of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The instrumentation response times are concurrent with the 10 second diesel generator startup time assumed in the accident analysis. The proposed changes do not impact the safety L analysis or a margin of safety. Therefore, the proposed change 'does not result in a significant reduction in the margin of safety. e /0 W RIVER BEND 23 10/1/9 9 (5A2 9319$ 4,$.9 A jff NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.3.6 - CONTROL ROD BLOCK INSTRUMENTATION \ "L6a CHANGE Entergy Operations Inc., has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. i . The following evaluation is provided for the three categories of the ' significant hazards 7onsideration standards
- 1. Does the change involve a significant increase in the i probability or consequences of an accident previously evaluated?
This change will allow control rods to be bypassed to allow , movement other than insertion. CTS 3/4.1.4.2 allows inoperable control rods to be bypassed to allow them to be inserted and disarmed. However, CTS 3.10.2 allows OPERABLE control rods'to be bypassed to allow performance of scram time and other , testing. Because the proposed change continues to require these movements to be performed in conformance with the control rod drop and rod withdrawal error analyses, as applicable, this ~ change does not significantly increase the probability of a previously, analyzed accident nor does it significantly increase the consequences of a previously analyzed accident. , 2. Does the change create the possibility of a new or different ! kind of accident from any accident previously evaluated? ! The proposed change introduces no new mode of plant operation .' (since it is currently allowed by CTS 3.10.2) and it does not involve physical modification to the plant. Therefore, it does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of '
safety? This change does not involve a significant reduction in a , margin of safety since the proposed change continues to require , these movements to be performed in conformance with the control ; rod drop and rod withdrawal error analyses, as applicable. e 4 k /0 Y RIVER BBND 36 '^^O i (1AK 93-mD? ,3,1I.2 g ja NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.3.7.6 - SOURCE RANGE MONITORS I "L6" CHANGE \ \ Entergy Operations Inc., has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
This change will increase the Channel Functional Test interval for the Source Range Monitors (SRMs) while operating in MODE 5. There are no accidents that are initiated by the failure of the SRMs. Therefore, this change does not increase the probability of a previously analyzed accident. Further, the ITS does not require the SRMs to perform any trip functions, only indication (see CTS 3.3.6 comment R.1). These functions are assured by the performance of a Channel Check every 12 hours per SR 3.3.1.2.4 and a Channel Calibration per SR 3.3.1.2.7 every 18 months. Therefore, this change does not significantly increase the consequences of a previously analyzed accident.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change introduces no new mode of plant operation and it doe.9 not involve physical modification to the plant. Therefore it does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
This change does not involve a significant reduction in a margin of safety since, as stated above, the SRMs are not required to perform any trip functions, only provide indication. These functions are adequately assured by the performance of a Channel Check every 12 hours per SR 3.3.1.2.4 and a Channel Calibration per SR 3.3.1.2.7 every 18 months. RIVER BEND 49 10/1/93 (fM 91/4tl) ATTACHMENT 2 ITS - PSTS , COMPARISON DOCUMENT REVISION 1 SECTION 3.3 REVISED PAGES 2A: MARKUP OF ITS - 2B: DISCUSSION OF CHANGES -a _
- 1 1
ATTACHMENT 2A . ITS - PSTS COMPARISON DOCUMENT REVISION 1 MARKUP OF ITS Qg 93 MIlI} S M Instrumentation 3.3.1.2 SURVE!LLANCE REQUIREMENTS (continued) ' SURVEILLANCE FREQUENCY SR 3.3.1.2.4 --------.---.-.---NOTE- ----..--.-...-..- Not rsquired to be met with less than or , squal to four fuel assemblies adjacent to , the 5 5 and no other fuel assemblies in the associated core quadrant. Verify count rate'is: 12 hours y during CORE
- a. m D!.or cos # tt ri- :? t r'#-J '-
ALTERATIONS _ e F" a * ;?:1% or @ b. a 0.77Ennwithasignal,tonoise E i i ratio a C*a::" 24 hours i r - SM 3 Fn im e C"fl_ _ _-i E;-:iiC-- . s aa n . D .3.1.2.- 7 geis) , SR 3.3.1.2 l/ --...-..-------.--NOTE----.----------.--- Not required to be performed until h 12 hours after I ms on Aange 2 or below. 4g,,; Perform CHAf51EL FL5ICTIONAL TEST. 31 days t . 3 - s}SR 3.3.1.2 1 ..----------------NOT -.--.-------.----. ' @Neutrondetectorsareexcluded. Perform CHAf81EL CALIBRATION. % 18 months f ry,y}.4.eqirej -b 6 peJarmel until N 12 hours aZ4e r TRtis n V.aq7-(_ortoeleu.). BidR/6 STS 3.3 12 Rev. 0,09/28/92 fha $ 9 ~14$ SRM Instrumentation 5.3.1.2 T ate 3.3.1.2 1 (sese 1 of il Seurse Range seantter Instruonntation AMLICABLE IEBER OR OTNER ag atagg amVEILLAeCE PimCTim segCIPIB CSSiflest CasseELs aseytageuTS 1 Source som seenitor 2(e) j l SR 3.3.1.2.1 88 3.3.1.2. SR 3.3.1.2 88 3.3.1.2 - x 3,4 2 sa 3.3.1.2.3 M 3.3.1.2. I W 3.3.1.2 i M 3.3.1.2 s 5 2(b),(c) ER 3.3.1.2.1 N 3.3.1.2.2 [' / "'. J J P M 3.3.1.2.6 / so st 3.3.1.2 se 3.3.1.2 'a (e) With IEts en Benge 2 er below. , . (b) etty one WI shersiot le reesired to be SWABLE edring spiret off tead er rotead esten the fusted reglen lastudes enty that me detester. (s) Special eswebte detesters ser he med in plass of SWts if eenmested to normal SWI sf rcuite. I I l l l BWR/6 STS 3.3-13 Rev. O, 09/28/92 l l14 93-nRI) Centrol Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.2.1.4 ....-....-........N0TE.--..--............ Not required to be perfo until I hour I after THERMAL POWER is s k RTP in ' MODE 1. Perfore CHANNEL FUNCTIONAL TEST. M92[ days h . . 3.;7 l O W,iN) SR 3.3.2.1.5 Calibrate the low power se t( The 184 days Allowable Value shall be t RTP and 42. s p35 RTP. @o; SR 3.3.2.1.6 Verify the RWL high power Function is not days b f. ssed when THERMAL POWER is RTP. @ sic ~L t 2. t. ~1 pe rder e c HAnhCL CAMBRAT'lan). 184 da ys y 7,g,2,,,' y ..................N0TE....-.-............ Not required to be performed until I hour ,- after reactor mode switch is in the e shutdown position. Perfors CHAfflEL FUNCTIONAL TEST. N18 nths h l ~ SR 3.3.2. .)9 Verify the bypassing and movement of Prior to and control rods required to be ssed in during the > Rod Action Control System ( ) ya movement of second licensed operator or other control rods .1 lified moaber of the technical staff. bypassed in p;I _ ^' SI - n hc. odor. ram s4aySed'elw I y ___ l l BWR/6 STS 3.3 16 Rev. O, 09/28/92 i 93-t't& Centrol Rod Block Instrumentation 3.3.2.1 Tatte 3.3.2.1 1 (pose 1 of 1) Centret Red sleet Instrumentatten Ap*LICABLE Nets OR ofugt SptCIFIS AsallaB MmCT!au SLAvt!LLANCE' 03BITleus CRAmuRLS ageWitEMENTS
- 1. and Pettern Centret System
- s. Red wittedroel Limiter hb.3.2 ..,
A (a) [ h 2 I 3R 3.3.2.1.1 ON f(b) 2 5m 3.3.2.1.2 H *:.H g\ k g
- b. Red pottern centrolter 1(c),2 2 4at't'.1.33 3.3.2 s,.3
. t
- 2. Reester made setteh =Ihuteenn Peef tfen b a 3.3.2.1.4 4" !:!:!:':g' g
i' / (d) 2 3 3.3.21.f e /\ /t (.9 .'. , I / (a) T m PmEE > IMit ATP.
- I (b) T m PmER p ; RTP and s U+ ' 'I i (c) With T m Pa m & 5 TP.
20 f (G enester made auf tet shutdeun positten. / ~3.:*2-l bS / BWR/6 STS - 3.3 17 Rev. O, 09/28/92 ftAR 93-!W$ PM Instrumentation 3.3.3.1 3.3 INSTRtMENTATION 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation LC0 3.3.3.1 The PAM instrumentation for each Function in Table 3.3.3.1-1 shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTIONS ....................................-N0TES----..-.-........-..................
- 1. LCO 3.0.4 is not applicable.
- 2. Separate Condition entry is allowed for each Function.
COMITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore required 30 days with one required channel to OPERA 8LE channel inoperable. status. Required Action and 2.2.:./ B. B.) Initiate actioy lunediately #/7 associated Completion c n w h ' Time of Condition A [f[?f"'; "yg fi at n ~- , not met. 5 .2 . U Sesed. , C. 6.--.'.. M TE.'...... l C.1 Restore one required 7 days I Not pplicab to channel to OPERABLE [hy itor] status. .f ... .... ,. . . . .. .. . . .J One or more Functions with two required - channels inoperable. (continued) BWR/6 STS 3.3 18 Rev. O, 09/28/92 (L AR 93- h20 PAM Instrumentation 3.3.3.1 ACTIONS (continued) COMITION REQUIRED ACTION COMPLETION TIME 6 Two [ ired hydrogen [.1 Restore one required 72 h rs ' channels hydrogen nitor] inop able. channel h ( monito status. OPERA 8LE D, i Required Action and f.1 Enter the Condition Immediately associated Completion referenced in Time of Condition C Table 3.3.3.1-1 for @ not met. the channel. @ As required by Be in MODE 3. 12 hours Required Action . and referenced
- Table 3.3.3.1-1.
As requitad by iL Initiate actio Immediately Required Action .1 nce it p and referenced ica on - g _ Table 3.3.3.1-1. >-9 5 .2.c g _ \ w \ \ _. g - O ll$l$N [/ A ' suf nW a speal{ Sepotf. , i c, 2, 7, "17 BWR/6 STS 3.3 19 Rev. O, 09/28/92 (4Af 91 -/YPQ PAM Instrumentation 3.3.3.1 SURVEILLANCE REQUIREMENTS ........ b. ..............N0TE-.--..-.--........................ 2 T~::: 5: r a ts each Function in Table 3.3.3.1-1. ..............................................................................)i SURVEILLANCE FREQUENCY i-m , SR 3.3.3.1.1 d Perform CHANNEL CHECK. 31 days r __vorc____ _ _ _ _ _ _ _ . . l ' Applerdle Er s*el h *e+*% k yam
- 3 33 A-1 4 i r e.J rr,s 10* Jil.
~ - , l R 3 3!37 . ~ ~ ~ - - - ~ , 4 FerfoEn't)WWeiEL CALIBRT . 1 18 months i i t b . \ l - - - - IJ OTO - - ~ " " " - ~ ~ ~ ~ ~ ~ ~ t S.S SI I O^ h Appb4W Oor b Nar \0a,J 11 U Ia*!8! I 3 3 I~ I Jg2 __ _ __ ._ .__ . < \ i s 'A u a , , , > t%% cmmet. :au awna. , 92 Lys ' P4 l l l l l I 1 I i 1 swr /6 STS 3.3-20 Rev. O, 09/28/52 (L49 9.3-/YRI] PAM Instrumentation 3.3.3.1 febte 3.3.3.1 1 (pege 1 of 1) Post Accident scenitoring Instrumentation C M IT!cNS REFtBENCED FRel PuncT!0N REGUIRED Resu! y cnAsufLS ACTIou .1 %:: / 1 Reactor steen Dame Pressure 2 .,r* { 2. J seester Veneet Water Level - Wde base 2 y (E ,i.f. Beuc.for Venel W4te Leel Suppression Poet Water Level Nel Cone a F 2 pL (o.k." Drywell Presour' 2 #E 9.Y Prisary CenteInuent Aree medietien 2 4erp (:ruu n ur yy erywnt" ~- -m kr*: Pala%
- AQg
##per 2 P'i,y*g'ow) f r _ j /44 iPc'iv resttf an hs,d;[ch M #( 3,34-l (v. Wide sange n,diron rim
- 10. Drywett N, &n o ' tyser
/ / / 2 / / r) ~ f 2 .P' 6 ' , / / 11.CentefrumentNia Anotyser 2 .P E
- 7. hl' Priaery Centeinennt Preneure *
. 2 *6 f 6 4 SLppreseien Poet esperature 2 AE 3 f(e) Not reedir'ed for feetetle[vetwee these tad penetretfan en poth le foole'ted tur et least one ' 4 . cleoed and de activoted luutenstic / volve, e annual volve, fnd flenge, or volve with flow ;;,, J the velve -- - 75 h.44rf Itenf tering each /f;;" 1 = c r---- _ __ rp wo sectori) g sevleuerde mete: Tebte 3.3.3.1 1 sheLL he amended for $esh plant es necessory to tlots 1. Att top 4Letery Guldn'1.M Type A instroent and ] M f 2./ Att Regulatory evide 1. W Category 1, non- A instrumente epoc led in the plant's aegulatory ,'l Cufde 1. W 5efetyp volustian Aspert. ,/ _ - - ~ - - - NJ$ regartre h h r io b4 vab'd3 *vbesC 0JSOsll*_ JOdof6 $" O 0 **' pab d n'olde0. M O /y one pos:/,6 ;,,L /,,',o c.la,,nel a ,cy,,,'a 0 4 ,pa,,d,,/,ah I' L p a6 -,M on/, aac ccar,/ ,an ,,,A L:n A J. t k\ ,1.1 */, 7 BWR/6 STS 3.3-21 Rev. O, 09/28/92 [ TAR 9:- /Y & ATWS.RPT fnstrumentation 3.3.4.2 3.3 INSTRUMENTATION 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS.RPT) Instrumentation LCO 3.3.4.2 Two chaanels per trip system for each ATWS.RPT instrumentation Function listed below shall be OPERABLE:
- a. Reactor Vessel Water Level-Low Low, Le"a 2; and S b. Reactor Steam Dome Pressure-High.
APPLICABILITY: MODE I. ACTIONS ____ ......___..___........____...... NOTE----------.--...---..-..---....-...__ Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels / A.1 Restore channel to 14 days inoperable. OPERABLE status. 93 9,2, v, ;7 A.2 ...__---NOTE--.------ g7 Not applicable if inoperable channel is the result of an inoperable breaker. l Place channel in 14 days trip. (continued) e BWR/6 STS 3.3-29 Rev. O, 09/28/92 GI b=l Yed an0Oly~eh 'l N A *& J / Primary Cintainment Isolation Instrumentation g , 7. f, , - _ 3.3.6.1 ; ApS '% ~ 3.3 INSTRWENTATION anf by^'d \ s 3.3.6.1 Primary Containment Isolation Instrumenta pg // LCO 3.3.6.1 f The primary containment isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERABLE. APPLICA8ILITY: According to Table 3.3.6.1-1. ACTIONS .---.--......---------.----------..--NOTE------------------------------------- Separate Condition entry is allowed for each channel. C0fGITION REQUIRED ACTION CONPLETION TINE l A. One or more required A.1 Place channel in 12hAursfor channels inoperable. trip. Functions 2.b.ffgSe. p - L5iEl d6 - @ 5.{( ' Q PH 83 \ /; =. :. s .i 24 hours for 's ' Functions other , i than Fu tions 2.b,( S A ag HM c . -- : S eEO 5.e[f J 5n)? B. One or more aut B.1 Restore attusMh I hour Functions with y. .rr ., , 6 isolation M isoisuon capability not capability. maintained. i l t I (continued) BWR/6 STS 3.3-48 Rev. O,09/28/92 ; 1 I ~ hR955 Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS (continued) CONDITION REQUIRED ACTION ' COMPLETION TIME H. As required by H.1 Se in MODE 3. Required Action C.1 12 hours and referenced in gg Table 3.3.6.1-I. H.2 8e in MODE 4. 36 hours Required Action and ! associated Completion Time of Condition F or G not met. Occio4eD I. As required by I.1 .- Declare #Jtandby I hour Required Action C.1 and referenced in giquid fontrol (5yeeer . inoperAle. Table 3.3.6.1-1. g, B - I.2 Isolate the Reactor 1 hour Water Cleanup System. J. As required by J.1 Initiate action to Required Action C.1 Ilunediately restore channel to and referenced in OPERABLE status. Table 3.3.6.1-1. J.2 Initiate action to Imediately isolate the Residual
- Heat Removal (RHR)
Shutdown C clin - ~7.3.6./ 5 steep sufwa $ros 13 7 , de een for veneli (continued) EMEW g. 3.0.t
- L COA ]' ' ' sur BWR/6 STS 3.3-50 Rev. O, 09/28/92 l
(LA2 93-IVR D INSERT 50A QR - J. (continued) J.3.1 Initiate action Immediately to restore primary containment to
- OPERABLE status.
AND J.3.2 ------NOTE------ Entry and exit is permissible under administrative control. Initiate action Immediately to close one door in'each , primary containment air lock. l 1 l l ~l t . j l f RIVER BEND INSERT 10/7/94 3.3-50 j] Ai G-/& Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS .......... .......................... MOTES---.---.------.--------------..--... . , j/ 1. p.Refer :P to+ Table Ei m 3.3.64 1 to determine which SRs apply for each t f 2- h JVunction. g . s. &- qf 2. When a channel is placed in an inoperable status solely for perfomance of s required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function '] maintains p f : ; s.t : C ..a isolation ca ............p................................pability. ................................ SURVEILLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK. 12 hours SR 3.3.6.1.2 Perfom CHAfWlEL FUNCTIONAL TEST. 992 bays g 3.3.6.1.3 Calibrate the trip unit. N92[ days N@l SR 3.3.6.1.4 Perfom CHAfWtEL CALIBRATION. 92 days SR 3.3.6.1.5 Perfom CHANNEL CALIBRATION. '918[ months h SR 3.3.6.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST. 918[ months h SR 3.3.6.1.7 s--.-NOTE-- 4------r'---- O ,Radiationde'tectorsma 9...:....y......-....ybeexcluded.------ .j. . . . . . 4. . . . . ., . .) /J 4---- ~ ~ ~ ~ Verify the ISOLATION SYSTEM RESPONSE TINE N18 months on is within limits. ,N l M,,,, .%" a STAGGERED h ,fful&> e 4,> b/ms TEST BASIS BWR/6 STS 3.3 52 Rev. O, 09/28/92 ,d C: /"i Primary Containeert Isolaticn Instrumentation 3.3.6.1 P \ ~ fette 3.3.6.1 1 (pose 1 of 6) Primary Centelement4teetetten Instrumentation j 3#f?7'.f g }),Q /'oC'l $ $lc cr}e .. ~~ APPLICABLE CtBSITIONS is3ES OR Result 9 REFERENCED OTER CuAmugLS PAst SPECIFIS PER TRIP RESJite SaVEILLANCE RAICTION ALLIRMSLE COSITIONS SYSTEN ACTION C.1 REGUIRBENTS VALUE
- 1. Main steen Line leotatien e.
-14 7 Reester vessel Water 1,2,3 0 38 3.3.6.1.1 - . ;1 Level - Lev Low Law, ,f'I23 SR 3.3.6.1.2 3 inenes Level 1 M sg 3.3.6.1.3 -- - sa 3.3.4.1.5 ER 3.3.6.1.6 m 3.3.6.1.7
- b. mein steen Line Preneure - Low 1
[h E 3R 3.3.6.1.1 se 3.3.6.1.2 m(57)Yis h 3.3.6.1.3 at 3.3.6.1.5 se 3.3.6.1.6 SR 3.3.6.1.7
- c. main steen Line 1,2,3
[per"" h a sa 3.3.4.1.1 "" J e a. @ <4 ii! ir SR 3.3.6.1.5 !-IIs' F5'd e9 3'> I'"' 9 P> ^? l st 3.3.6.1.6 f ISS r sd , Lehe c. E 3.3 6.1.7 f 169 s+J s L.._'ef
- d. Canutenser vemam-Lee 1,2(a) , III' 8 SR 3.3.4.1.1 2G
/' >< > @ <J ii! ir "' arm inchesZD M 3.3.6.1.5 st 3.3.6.1.6 14 %.5
- e. Noin steen Tisvis' 1,2,3 ar" D st 3.3.4.1.1 s 19Mt*F 4 '3.g Tempweture - NIsh St 3.3.6.1.2 -- --
at 3.3.6.1.5 .* /d g at 3.3.4.1.6 g / , T me at naviet 1 ,3 i D SR 3. .1.1 1946t* , D ffer t at 3 .6.1.2 e ~"' i & '*** d 3:!tl f l f (g.' norumt Initietiem 1,2,3 4 3R 3.3.4.1.6 ue # Primary Centelement l /} 2. leetettem , ' e. Roseter vessel Water 1,2,3 N st 3.3.6.1.1 a f4het \ Lovet -Law Law, Level 2 /' 3.3.6.1.2 hg"t Iruhee g , a s - 3.3.6.1.3 2M ~ 1-(F.s' > :.J \ yJ?f ! u.a.ng t g u 3,e 3 q= % 5. e >
- . 7
\' _1- - h 3 3 6. 1 LG (cent f rtwed) \ ' f) Alu repatr_N 4e or.op/c 143 cruavraleS cfry atf , sol /ri , . g (e) with.er Tyurbine.Isten vel d net etened. __ % I op NA Sr .- M g @<en. l 2.3 ; D A 146.3 *E ' T.v fi d wSm.a a m... Td y%.. (CJ . 'f 5 40 l / f X Te y.n. w - H l Fwn. r. 7 g ---- a r t 146.3 'F [ h,g rr4. pr.; u,3.i.(Ei,Il.JiQ T.,, bc. , ,iL3 y -p g g,, 3 y , s , it Air ie-pe ro t,e - H , 2,3 3 - t.- MS L m e.d.,. 5er.,.4.. l -- "'* f g3o aF N
- 2. O B d ' d ' " A' u T.,,. 3, e - 1, 2,3 3.3-53 Rev. O, 09/28/92 Q _ _ _ . _
(L.11< G: -Ivq) PH eary Containment Isolation Instrumentation 3.3.6.1 Teatte 3.3.6.1*1 (pose 2 of 6) Prieory Cmtelnennt leetetten Instrumentation APPt.lCASLE CIMITIWS usts et SHU!am REFgatuCe OTat caAsuRLs Feet EPECIFIS PER TRIP ReeJIggD SEVE!LLANCE ALLOW 48tf PLAICTIS CRESITIst $f8798 ACTIS C.1 AEEl!RSENTS VALUE ~ 2. (h ~ f
- b. Drywett Pressure-nigh 1,2,3 u et 3.3.4.1.1 s se
/, m 3.3.6.1.2 e- ' 3.3.6.1.3F a 3.3.6.1.5 _ . . c,. m 3.3.4.1.6f GL." O - ~ t2 C .
- c. a ter Wesset Water 1,2,3 (23 / F st .3.4.1.1 a t 152.51 /
L L - Les Lou Law, m .3.4.1.2 inches ( 1 (ICCS (3 3.3.6.1.31 Ivielene 1 and 2)
- 3.3.6.1.5 3.3.6.1.6 3.3.6.1.7
/ , /
- d. Drywott Prenevre-Nigh 1,2,3 ,/ CII F M 3.3.6.1.1 s (1. 3 pef [
(OCCs Divial 1 a 3.3.4.1.2 and 2) Em 3.3.6.1.31 a 3.3.6.1.5 a 3.3.4.1.6 a 3.3.4.1.7 ~ _ -4
- e. A ter veneet unter 1,2,3 tal F a 3.3.6. 1 t ( 43.83 ovel-Lee Low, Lovet a 3.3.4 .2 innhen 2 (IgbCE) Em 3.3. 1.31 5 3.3 .1.5 a 3. 6.1.6 a 3.6.1.7
- f. Drywett n - Nish 1,2,3 tal F 3.3.4.1.1 s (1 I pois (subCs) 3.3.4.1.2 3.3.6.1.31 a 3.3.6.1.5 a 3.3.4.1.6 m 3.3.6.1.7
] D N[ temation- Afgh @@O :E 3.3.6.1.5 i:l i i:1 @)@ m I.1 4.1.6 l @ = ' ' 2 " L' " - l ((b)) / 123 .- t E 3.3.6.1.1 s 14.01 es/hr / a 3.3.4.1.2 / / 3.3.4.165 t m a 3.3.6.1.4 - l 5 3.3.4.1.7 l (sent s,uded) ' g, , ((b) During .c .,,, d 's,.ALTERAft . m a ,et e isies. forM o.f irredt reini,. fust see m tise m re.ste, ,eseei. (prisery or f j ry contairument). i l l ' Q 4 Q f -1, iW4 A c,an /> .fpa/ ,2,/, /,,, g;q l BWR/6 STS J.3-54 Rev. O, 09/28/92 l [M 9.? Nil) i Primary Containment Isolation Instrumentation 3.3.5.1 fatte 3.3.6.1 1 (pose 3 ef 6) Primary Centeineent teetetten Instruentation APPLICABLE CaelT10e8 mass om egeules agFa geCED ofeER CNAmeELs Feel SPECIFisD PER TRIP R9EJIRED REVEILLAeCE ALLOW 45LE FtmCT!ce Caelfles sYSTM ACTim C.1 R9eulBSEeft q 3,b*[ VALUE - f 2. Primary Centefnennt ' / S' Isoletten (continued) ( 49 I d -
- f. merest Inittetten 1,2,3 s sa 3.3.6.1.6 mA
- 3. Reester Core Isoletten Casting (SCIC) System feeletten gg
- e. SCIC steen Line Flem - #Ish 1,2,3 y
J1 F 3R 3.3.6.1.1 a 3.3.4.1.2 s #69 inches meter h I-4M 3.3.6.1.3 F gi . m 3.3.6.1.5 at 3.3. C.Ep.L - c . n 3,/ # 73 / D. RCIC Steen Line Fles 7fme DeLsy / J1,2,3T / J1[ F M 3.3.4.1.3 3 3.3.4.1.4 a seconde ens 43D sesenes /, a 3.3.6.1.6 ( 13
- c. aCIC Steam 1,2,3 F agt P _ - ,eg .epty Line 3.3.6.,1.1
/>1 . .&. 3 - @r".Md. i:lii:! ' E "p * *:S
- d. actc furtine Emhoust 0iephrege 1,2,3 hF st 3.3.6.1.1 s (20f B 3.3.6.1.2 /
pets h Prosaure - N f sh d= '3*'>+ Osir'In3.3.4.1.5 m 3.3.4.1.6 gg
- e. RCIC Equipment teen 1,2,3 J1 61 F M 3.3.4.1.1 s $486&'F Ashtent / sa 3.3.4.1.2
'J. N' l iesperature - N f gh a 3.3.6.1.5 s E 3341 Y _g pIk h // n s
- f. eCIC test DiffgPentfet asam 1,2, 3.3.6.
5 3.3. 1 1.2 sflast'P T ,JL' .6:i:ifm,,< main Steen Line fumet 1,2,3 _J1[BI F M 3.3.6.1.1 s 1999t'F Antient 7 m 3.3.4.1.2 Yesysroture - aigh a 3.3.6.1.5 m 3.3.6.1.6 .,s.__ ! ~ __ . . _ _ . . __ I'*"'i""*d3 y_:& A/>a tuu,;cf -lo is.'Viile rI 9, 2. (r ass en? le A hy~e// v la/a~ A~cl' n - - g BWR/6 STS 3.3-55 Rev. O, 09/28/92 f B 93-N#1) Primary Containment Isolatten Instrumentation 3.3.6.1 Table 3.3.6.1 1 (page 4 of 6) Primary Centeltuunnt testation Instrsamentation APPLICA4LE CIMIT!WS seBES OR REFlatsCEB OTMR ROWIRS FtWI spOCIFIS CilAMELS Pke temitsp smVE!LLANCE FimCT!all CB SITIgus ALLa male Tft!P 878791 ACTIM C.1 ageuleSENTS vatut
- 3. ACIC Systes feetation (eentirased) _
- h. p tese no Turvie ,2,3 ,,11 $1 m 6.1.1 6 e D1 1 -
s ( M F" T e - Nigh M .3.6.1. e 3.3.6. 5 3.3.6. 4[ ,,b noin steen Line Twviet 1,2,3 d[M,i ;- Temperature fleer ft)#h F M 3.3.4.1.2 a 3.3.6.1.4 h M 3.3.4.1.6 I I n 4 ' am Esaissent Reen Assient fauparature eigh 1,2,3 O' ;-- IhhF . M 3.3.4.1.1 sa 3.3.6.1.2 s *F h g ' a 3.3.4.1.5 ) s _ _ _ _ _ _ m 3.3.6.1.6 [1,2,3 g ii i lETa'-,/ ' /
- t. Re geefalant Respr b" / ;: _' @7 lit F l.'.:i:
/! 3.3.6A.1%y'4.1f! [" ' F g 3 > M2 /J1[h g 4 8 aCIC/tm Stems Line 1,2,3 flew - Nigh F M 3.3.4.1.1 sitShinehen gi II 5 3.3.6.1.2 meter _ n tm 3.3.6.1.35f a 3.3.6.1.5 \ l M 3.3.6.1.6 ' \ ,,f g pryisett Preneure-afgh g 1,2,3 / (1) h F M 3.3.4.1.1 s st 3.3.4.1.2 - - soldd l 3.3.6.1.3Q B 3.3.6.1.5 , i s \ pg 3 3.3.6.1.6 i - i ' .' moreset Inf tietien 1,2,3 J2f G st 3.3.6.1.6 .M ~l mA f- I - 4 teoster Water Cteams (asW) Systes leetetten gg f (1[h
- e. Differentfet Flem-sid 1,2,3 F M 3.3.6.1.1 s 10pt som f at 3.3.6.1.2 --
, ~ ' N 3.3.6.1.5 ' , ( C,2@ l m_ 3.3.4.1.4 we.1 0Bl A Y ?ld;Jr 4 ~
- b. Difforentiet Flem fleur 1,2,3
/[@J1 F st 3.3.6.1 3 M 3.3.4.1.4 M 3.3.6.1.6 s 1886 eenende )
- c. nWCW meet 1,2,3 1 F ER 3.3.4.1.1 gestament seen M 3.3.4.1.2 Tessereture - Nigh SR 3.3.6.1.5 .
m 3.3.6.1.6 (continuse) BWR/6 STS 3.3-56 Rev. O, 09/28/92 . ~ . .. - @k C-NPD Primary Containment Isolation Instrumentation 3.3.6.1 Teste 3.3.6.1 1 (pose 5 of 6) Primary Cente,reent Isoletten tretrusettetten APPLICA8LE CMITitbl8 mmES OR REFtesmCED OTIER RGWIRS Fatpl SPECIFIS CMauRLs PER ageW!aED RAWEILLAnte FWCTitel Attomatg Cine!T!cus TRIP 57879 ACTim C.1 Atmlastuft vatyg ,\ O 4 AlEN System leetetten al} (eentirtaad) -- g s' d. EWtu, meet Edad DHf a tot
- 1. 3
/ J1 h F st 3.3.6.1. 3.3.6.1 s _r 3.3.6 .5 Tessorsture Nish ** 1 1.6 1 hautuPupessee 1,2,3 F e 3.3.6.1.1 s 44ael'F Temperature - mish "pAir'_"*^ - - M 3.3.6.1.2 se 3.3.6.1.5 m 3.3.4.1.4 y, set,/ 7' rWe' /il!g[4 w 3.3.4.1.6 ~' W 6 . RE3A velve uset teen 1,2,3 11 8 (8,! F M 3.3.4.1.1 s f9ett*F feequereture - Wish # gl M 3.3.6.1.2 ' N 3.3.4.1.5 , a 3.3.4.1.4 Yff'"/ 'Y /"Y ' /il;If$lff'<? . d .4.i.6 % .si .te. u t T e.. i,2,3 /ir'@ , 3.3.6.i., 3.3.6.1.2 , n!, @ ". i:liid ,u ./f~'/ ( ')' "'~f.*! ' ; i.62; . .3... . 3.3.6.1. / [h seerter vesset Water 1,2,3 F st 3.3.6.1.1 a Levet . Law Lem, Levet 2 gg i e 3.3.4.1.2 inshee E 3.3.4.1.3( 3.3.6.1.5 . OW m _ 3.3.6.1.y 7. { . ) / j jf fSta $ Liepsid Centret 1,2 (IT I E 3.3.6.1.6 N6 ' Syste Initistien I ,,, IRitI8 tim 1,2,3 8 E 3*3*0+1 0 #A Rw.u De a,4I.ic E**ms T perebra - N'9k ),2,3 i F ---+ f Il45bett i R wcu Re, c'i-9 T.-k. f f i 1.1 l n i I F ---> f II+ 5F % ,...sel.H + r,, , ~ ~ ~ ~ _ _ - _. Gt, 3.i:,il BWR/6 STS 3.3-57 ft1:Llr.}l ( st23615 I Rev. O, 09/28/92 (it if @AR 93 !/Al} \ l Primary Containment Isolation Instrumentation j 3.3.6.1 , i l Table 3.3.6.1 1 (pose 6 of 6) Priomry containment Isoletten Instrumentatten - 1 APPLICASLE Com !T10NS ntnis OR REFERENCED OTNER asculatD FROM l sPECIFIED CNANNELs PER RfeUIRED REVEILLANCE ALLOWASLE ' g FUNCTION Com 1710Ns TRIP SYSTEM ACTION C.1 REGUIRSE!!TS VALug ] UW_) 1
- 5. A _ . : ;; 4 system I
/ Isotsiitten h, f pf,j
- e. RNR Esalpeent Room Ambient
/ 2,3 " -- - - n - F sa 3.3.6.1.1 st 3.3.6.1.2 s (# 4)*F Temperature - Nigh st 3.3.6.1.5 g sa 3.3.6.1.6 ~ , [' ' $ ',3, g (T ret -N [ 3.3.6. teacter vesset Water ,5 ( J sa 6.1.1 tI b Lovet - Law, Love o h gsa sa ,3.,3.6.1.2 . 6. . .,u inches sa 3.3.6.1.5 f J t'As ,s-aQb,'f
- d. Reecter steam Dems Pressure - Nigh 1,2,3 h F st 3.3.6.1.1 s 15 st 3.3.6.1.2 le h f m 3.3.4.1.3 k n 3.3.6.1.5 st 3.3.6.1.6 f, y
- e. Drywet t Pressure - Nish 1,2,3 y F M 3.3.6.1.1 s &&.45) psiff n a 3.3.6.1.2 h
& f aa3.3.6.1.3 3.3.6.1.5 R h, msN,ta/.LH /, p,3 2 S. ~ ^ , ,'3 f(, >< ( ~ - - (g) on y one trip system reesired in Mtsts 4 and 5 with RNR shutdeun Coeting system integrity maintained. ,' {C-) Wlfb reae.iot sicom o{ cme pressate greafer + bon or eg,re / -fp f{e y }/ g c e,t. j,7 m (G.i ser~ru,ve p< essa <c . (S) u/e'tk rea e 3 fe. e c?wa ,oresssore les1 1barr tkt: R/Y[ cef-it1pertwijside, pres 1are . (c. :.c.. / l . w __ -_ _q 7--~ ~,/ ,e I, 2, 2 ') F f a' '3.; o 6. I I y 9,9
- e :. 2. s t. 2 ;,g a .
7 s i 0 3 4. /. 3 s y 2 3. b .I. J' s a 2. *3. * * *
- O 3. 3. &.
4d s, r C , bdo, Vesbalve 1,2, 2 a f in 3 3. 61 ' ? -/47 ) - 4eial- fo, k lo SR 3 * & ' }^/ebe s Y( fl'l , 32 3. L (p. /. 3 b'"l l .a a.2. s.t. T ) ,- sf 3. L & to & r ~ ( ~ ; nj y/# 8WR/6 STS 3.3-58 Rev. O, 09/28/92 (h +'-/Y/Q S condary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.6.2.2 Perfom CHANNEL FUNCTIONAL TEST. t{92 fdays @ Calibrate the trip unit. k92fdays / -. 3.3.6.2.3 ._ SR 3.3.6.2.4 Perform CHANNEL CALIBRATION. T[18[ months SR 3.3.6.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. t{18fmonths h /h *I , [ 3,3, [2.6 ......../.......N0TE.......... Radiation /.' detectors,maybeexcluded. A Verif the 150 ION SYST RESPONSE ihE is w hin limi . k18 a ST months ERED - h CB / T f ' t . BASIS / 1 l l l 1 BWR/6 STS 3.3-61 Rev. O, 09/28/92 I I (L AI 9:-'WD _ Secondary Containment Isolation instrumentation 3.3.6.2 febte 3.3.6.21 (pose 1 of 1) Seeerudery Centainment Isoletten Instrumentatte APPt!CASLA semes As agelatD ofst CmAmuRLS SPECIFIS Pga Talp amVEILLANCE ALLthmsLE FUNCTION COMIT!aNS STtTEM M etBSENTS VALUE 1 Reacter Vessel lister Levet -Lou Law, Level 2 h 1,2,3,9 ^- E 3.3.6.2.1
- M 3.3.4.2.2 a--
gc.ASEM3.3.4.2.31 3.3.6.2.4 ,,. s ^ '* SR_3.3.5.g _% g,gg \ 1,2,3 ' J2>'Oa' = >2*
- 2. Dryuen t Pressure - nigh M 3.3.4.2.2 1
= '+ ** e's ' / & " !:li!: P - _ _ 2..o) 1 M i i; n 4 ~"= ' E Fuel ting Ares 1,2,3 m 3.3.4.2.1 s (4.01 am/hr ' ventil ten Emaist ((e),(D31 M 3.3.6.2 assi en-Nigh atsh M 3.3.6. 4 EE 3.3. .5 CM 3.3 .2.63 h 1 -- E 6. F 4 anuRing Ares 1,2,3, (23 5 3 .4.2.1 s (35 et/hr Emmat ((o),(b)) E .3.4.2.2 Radiatien-Nigh El B 3.3.4.2.4 3 3.3.4.2.5 s 3.3.4.2.41 lemust InitietIen 1,2 M 3.3.4.2.5 uA V ;= -te)-purice operati - ; e e -.. iet fu C:.-: , t ^ - - ] __t _ (a) .Hyr Durtre newesent of Irrediated fust senaabtise in the - - . .c Eel buiIJ.L 3-t 3. Fu el $d fel (s) \ 4 i '} s s x lo) .u C, fuc, n ws+ h.Yeok n -uvahoe Il Ree3
- RESA)
INjbdt*f.'h , N I --* i 7of vto%Ci/cc O Rms " RE5 0 ) f 5 K 3.5.6.2.lD S R 3.1.6.2.2. S R 3.1. 6, 2.4 l l s a 5.5. 6. 2.s'j . t l l t 1 i BWR/6 STS 3.3-62 Rev. O, 09/28/92 i i LAA 93=/Ydt) h J CRFA d ys(tem Instrument 3.3.7.1 teMe 3.3.7.1 1 (pose 1 of 1) h trot Rose Freep Air yttee Instrumentation APPL 1 CARE CIBSITI13s NSES OR AGGUIRED REFERENCE OflER CNelAELS Peel SPECIFla PER ft!P ageslage navEILLANCE ALLenastg MAICTION CIESITIGIS SYSTWI ACT!n A.1 AguileggNTS v4 Lug
- 1. seester veneet Water 1,2,3,
/ g . nT' ~ 3 (23 a 3,3,7.1.1 t la48.43 Inanes . c, ;) L m L = Law Law, L m t 2 - 3a 3.3.7.1.2 U , w, y laM3.3.7.1.31 * ,y y 3.3.7.1.6 - - at 3.3.7.1.5 - - . ~w
- 2. Drywelt Pressure - NI$ 1,2,3 III C ga 3.3.7.1.1 3 pegg a 3.3.7.1.2 %
- 3.3.7.1.3W g m 3.3.7.1.4 } [ - m 3.3.7.1.5 _
- 3. Centret seen ventitetten Radiation 1,2,3, g D m 3.3.7.1.1 m monitore gg a 3.3.7.1,3 og7, gg.gg; g,, -
(a),(b) .1. m 3.3.7.1.4 _ m 3.3.7.1.5 (a) Swing operations with a potentist for estning the reester veneet. (b) Durigy of irradiated fuel sesseMles in thebleery er eseandery sentelruent[ f_wc Actreenotn Jduri,-{_g i 1
- 2. 3.7, /
93/ se BWR/6 STS 3.3 76 Rev. O, 09/28/92 I ll Ad G* -l'/& \ RPS Instrumentation B 3.3.1.1 l l . ) BASES N" E the containment by minimizing the energy that must be [( TATW AUAl%absorbed 1 following a LOCA. l Leo , c ed "N RPS instrumentation satisfies Criterion 3 of the NRC Policy : Statement. Functions not specifically credited in the \ (CoM'd ) accident analysis are retained for the e l I w ,- JR rem".; - --@ RPS as required by the NRC approved ! , ' ' IIcensing basis. { D / . The OPERA 8ILITY of the RPS is dependent on the OPERABILITY ) '-m I of the individual instrumentation channel Functions [910 specified in Table 3.3.1.1-1. Each Function must have a j ) ( required number of OPERABLE channels per RPS trip system, with their setpoints within the specified Allowable Value, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. , Each channel must also respond within its assumed response 1 time. , Allowable Values are specified for each RPS Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal ! setpoints are selected to ensure that the actual setpoints l do not exceed the Allowable Value between successive CHANNEL l CALIBRATION $. Operation with a trip setpoint less l conservative than the nominal trip setpoint, but within its ! Allowable Value, is acceptable. A channel is inoperable if l its actual trip setpoint is not within its required l Allowable Value. i Trip setpoints are those predetensined values of output at I which an action should take place. The setpoints are l
- compared to the actual process parameter (e.g., reactor )
vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined, accounting for the i remaining instrument errors (e.g., drift). The trip ! setpoints derived in this manner provide adequate protection because instrumentation uncertainties; process effects, calibration tolerances, instrument drift, and severe (continued) BWR/6 STS B 3.3-3 Rev. O, 09/28/92 (LM 93-14 0 RPS Instrumentation 8 3.3.1.1 8ASES APPLICABLE 1.b. Intemediate Ranae Monitor-Inon SAFETY ANALYSES, LCO, and This trip signal provides assurance that a minisium number of APPLICABILITY IRMs are OPERA 8LE. Anytime an IRM mode switch is moved to (continued) any position other than " Operate," the detector voltage drops below a preset level, or a module is not plugged in, an inoperative trip signal will be received by the RPS unless the IRM is bypassed. Since only one IRM in each trip system may be bypassed, only one IRM in each RPS trip system may be inoperable without resulting in an RPS trip signal. This Function was not specifically credited in the accident 31 l analysis, but it is retained for theG_.eiel; r;n -;a, q ) ,p ' ] %;.licensing . . ., m tiasts. @ RPS as required by the NRC approved Six channels of Intermediate Range Monitor-Inop with three l channels in each trip system are required to be OPERA 8LE to ) ensure that no single instrument failure will preclude a * ' scram from this Function on a valid signal. Since this Function is not assumed in the sa'fety analysis, there is no Allowable Value for this Function. This Function is required to be OPERABLE when the Intermediate Range Monitor Neutron Flux-High Function is required. 2.a. Averaae Power Ranae Venitor Neutron Flux-Hich. Setdown The APRM channels receive input signals from the local power range monitors (LPIR) within the reactor core to provide an ind< cation of the power distribution and local power changes. The APM channels average these LPRM signals to prov'de a continuous indication of average reactor power free a few percent to greater than RTP. For operation at low power (i.e., MODE 2), the Average Power Range Monitor, Neutron Flux-High, Setdown Function is capable of . generating a trip signal that prevents fuel damage resulting from abnormal operating transients in this power range. For most operation at low power levels, the Average Power Range Monitor Neutron Flux-High, Setdown Function will provide a secondary scraa to the Intermediate Range Monitor Neutron Flux-High Function because of the relative setpoints. With (continued) BWR/6 STS B 3.3-6 Rev. O, 09/28/92 d , ( C -/@ r RPS Instru1entation 8 3.3.1.1 BASES
- APPLICA8LE 2.d. Averaae Power Ranae Monitor-Inon (continued) ,
SAFETY ANALYSES, LCO, and APRM has too few LPRM inputs (< 11), an inoperative trip ; APPLIC481LITY signal will be received by the RPS, unless the APRM is bypassed. Since only one APRM in each trip system may be bypassed, only one APRM in each trip system may be inoperable without resulting in an RPS trip signal. This < 1 ,l.I Function was not specifically credited in the accident ! analysis, but it is retained for thed--P ns.. ee..s. -- *
- p. ;
' q #e- tty e' PORPS as required by the NRC approved licensing basis. ree channels of Average Power Range Mcnitor-Inop with channels in each trip systes are required to be OPERA 8LE to ensure that no single failure will preclude a scram from this Function on a valid signal. There is no Allowable Value for this Function. . This Function is required to be OPERABLE in the MODES where the APRM Functions are required.
- 3. Reactor Vessel Steam na== Pressure-Hiah An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and i THERMAL POWER transferred to the reactor coolant to !
increase, which could challenge the integrity of the fuel cladding and the RCPS. No specific safety analysis takes direct credit for this Function. However, the Reactor J Vessel Steam Dome Pressure-High Function initiates a scram O ;' q for transients that result 8 Fin a pressure increase, counteracting the pressure increase by rapidly reducing core / power. For the overpressurization protection analysis of g*g g'y >, '\ N Reference 2, the reactor scram t--===' - """ "-" > 6 (6 the analyses conservatively assume scram on , A bd the Average Pouer Range Monitor Fixed Neutron Flux-High
- t- - uma r'-"-1 signal) signa , not tnet = = ---
4'"# $ 6( --- 1 @, along with the S/RVs, limits the peak RPV pressure to ~ less than the ASME Section III Code limits. High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure-High Allowable Value is (continued) B 3.3-11 Rev. O, 09/28/92 BWR/6 STS (he 9:-/w O RPS Instrumentation 8 3.3.1.1 , BASES APPLICABLE 6. SAFETY ANALYSES, Main Steam Isolation valve-closure (continued) LCO, and close. In MODE 2, the heat generation rate is low enough so APPLICA8ILITY that the other diverse RPS functions provide sufficient protection.
- 7. Drvwell Pressure-Hioh High pressure in the drywell could indicate a break in the RCPS. A reactor scram is initiated to minimize the y possibility of fuel damage and to reduce the amount of g one being added to the coolant and the drywell. The Drywe.1 Pressure-High Function is a secondary scram signal to Reactor Vessel Water Level-Low, Level 3 for*LOCA events gl inside the drywell. This Function was not specifically
[,3 4~ (' 0 'i credited in the accident analysis, but it is retained for the 4 3mii .. _.. -- i- r: @ r. m # M PS as_reenired ._ ) by the NRC approved icensing basis. .,g gqQg;p ' High drywell pressure signals are initiated from four. O79 pressuretransmittersthatsensedrywellGrossure. The Allowable Value was selected to be as low as possible and be indicative of a LOCA inside primary containment. Four channels of Drywell Pressure-High Function, with two channels in each trip system, are required to be OPERA 8LE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is ; required in MODES 1 and 2 where considerable energy exists in the RCS, resulting in the Ifmiting transients and ~ accidents. 8.a. h. Scram Discharoe Vol- Water Level-Hioh ! \ The SDV receives the water displaced by the motion of the CRD pistons during a reactor scram. Should this volume fill to a point where there is insufficW volume to accept the i displaced water, control rod inser' 4 would be hindered. ! Therefore, a reacter scram'is initived when the remaining free volume is stik sufficient to 4sommodate the water ; from a full core scram. However, even though the two types ' of Scram Discharge Volume Water Level-High Functions are an ' input to the RPS logic, no credit is taken for a scram initiated from these Functions for any of the design basis I (continued) I l BWR/6 STS B 3.3-15 Rev. O, 09/28/92 I I [f 9 :-tvrD i i RPS Instrumentation . B 3.3.1.1 BASES APPLICABLE 11. Reactor Mode Switch-Shutdown Position SAFETY ANALYSES, LCO, and The Reactor Mode Switch-Shutdown Position Function provides APPLICABILITY signals, via the manual scram logic channels, that are (continued) redundant to the automatic protective instrumentation ' channels and provide manual reactor trip capability. This Function was not specifically credited in the accident l /,y]b analysis, but it is retained for the cy;;-a! i m::=y 2 M .__ _ A S RPS as required by the NRC approved j ' Ifcensing basis. The reactor mode switch is a single switch with four channels, each of which inputs into one of the RPS logic channels. . There is no Allowable Value for this Function since the-channels are mechanically actuated based solely on reactor mode switch position. Four channels of Reactor Mode Switch-Shutdown Position Function, with two channels in each trip system, are ' available and required to be OPERABLE. The Reactor i Mode-Switch Shutdown Position Function is required to be ' OPERABLE in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel i assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
- 12. Manual Scram The Manual Scram push button channels provide signals, via the manual scram logic channels, to each of the four RPS logic channels that are redundant to the automatic
, protective instrumentation channels and provide manual 7 ,'gf reactor trip capability. This Function was not specifically credited in the accident analysis, but it is retained for \ # d ']l he RPS as required There is one hanual Scram push button channel for each of the four RPS logic channels. In order to cause a scram it is necessary that at least one channel in each trip system be actuated. ; i (continued) ) l BWR/6 STS 8 3.3-19 Rev. O, 09/28/92 []/M %=NI RPS Instrumentation B 3.3.1.1 BASES 4 SURVEILLANCE SR 3.3.1.1.1 (continued) REQUIREMENTS CHANNEL CHECK ispa comparison of the parameter indicated on J one channel to a similar parameter on other channels. It is f- based on the assumption that instrument channels monitoring the same parameter should read approxilantely the same value. OCA Significant deviations between the instrument channels could r be an indication of excessive instrument drift on one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are deterisined by the plant staff based l on a combination of the channel instrument uncertainties, including indication and readabilit If a channel is h outside the aggl> criteria, it may b. an indication that the instrument has drifted outside its limit. . I The Frequency is based upon operating experience that ) T demonstrates channel failure is rare. I ne-peMermancej-e j 5-- NI N b N C Ia k U?i [ The CHANN5[ CHECK supplements less forma', but more frequent, checks of , channels during normal operational use of the displays associated with the channels required by the LCO. l J , SR 3.3.1.1.2 ; To ensure that the APRMs are accurately indicating the true ' core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. LC0 3.2.4, ' Average Power Range Monitor (APRM) Sain and Setpoints " allows the APWis'to be reading greater than actiial THERMAL POWER to compensate for localized power peaking. When this adjustment is made, the requirement for the APRMs to g [C-_100p J indicate _within 2% RTP of calculated power is modified to require the Ar,m. to indicateW.;.. Z , "") of calculated NFLPD. The Frequency of once Ter 7 days is based on minor '3 changes in LPRM sensitivity, wtich could affect the APRM l Lqr , reading between perforacnces of SR 3.3.1.1.8. I "' ' . }# A h is provided that Gonrequires the SR to be meth at ' / E'S \ a 25% RTP because it is difficult to accurately AR5iEM t//' core THERMAL POWER a heat balance when < 25% RTP. At ' ] l cAe e n% ETPj _ F:~ Vakkk APkM covsis4mk wik dW1 ;hh of (continued) BWR/6 STS B 3.3-2E Rev. O, 09/28/92 l h' A 9'? k l RPS Instrumentation ' 8 3.3.1.1 . BASES SURVEILLANCE SR 3.3.1.1.10 ! REQUIREMENTS (continued) The calibration of trip units provides a check of the actual i trip setpoints. The channel must be declared inoperable if , the trip setting is discovered to be less conservative than ' the Allowable Value specified in Table 3.3.1.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be ' readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days for'SR 3.3.1.1.10 is based on the reliability analysis of Reference 9. SR 3.3.1.1.11 G SR 3.3.1.1.13m b [ se s.N.O 47 p ' v _- _ A CHANNEL CALIBRATION is a complete check of the instrument i loop and the sensor. This test verifies the channel responds to the measured g rameter within the necessary range and accuracy. CHAf00EL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrationst ----J... ..:. ::d ;;;.,;t:0 :::r M itz,: sT3 , alejr---tP-- - t he ---'--ne(consistent with the plant specific setpoint methodology. J . J_.....; ....;i .,. hr T s ,, g irrat;d ;I r_ -
- ht--2 -!ti M e--+4^== a' + ha - -= *--i -3
~~ o , ; ,1,1 s l 1 Uf ifT'tne as r. zusyvi % ::t eithi ^'t; . q .i... l Allowable ue, t ant specifi 4dtpoint met ology ma , : s be rev , as a riate, if history an 1 other Md - pe t inf tion indic a need for.1 revisio Th t l dyb Map s point 1 be left s consistent wffh the assum ef-M ca- ;;t ;'; .t :;:-ifh20treint C9-' d. ptions is--noteiFd neutron detectorsvare 4 4exciuseub rtimue Tram % dnQ ower.E ' CALIBRATION because of the difficulty of simulating a.
- r. meaningful signal. Changes in neutron detector sensitivity M66e,N are compensated for by performing the 7 day calorimetric ,
l'L,,.6,...t Lib alibration (SR 3.3.1.1.2) and the : 000 20/T LPRM l A calibration against the TIPS (5R 3.3.1.1.8) J The Frequency uisd..a_J.. r1 m of SR 3.3.1.1.114 1s based upon the assumptio MT'3EmiiD g; ; . . _ . ___J
- ;- - - --- -s of he magnitude hi.h.1. I, d -h h2NSEPJ B29 A h
[Le sg 3.3. i. i. m rfD (c'"ti""'d) _ -_v 8WR/6 STS B 3.3-2g Rev. O, 09/28/92 r' 43 & S M Instrumentation B 3.3.1.2 ! BASES i SURVEILLANCE SR 3.3.1.2.1 and SR 3.3.1.2.3 (continued) REQUIREMENTS more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.1.2.2 -- q h, To provide adequate coverage of potential reactivity changes in the core, one SRM is required to be OPERABLE in the quadrant where CORE ALTERATIONS are being performed, and the ! other OPERASLE SM must be in an adjacent quadrant / Note l states that this SR is required to be met only during CORE ! ALTERATIONS. It is not required to be met at other times in ' MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a review of plant logs to-ensure that SRMs required to be OPERABLE for given CORE - l ALTERATIONS are, in fact, OPERABLE. In the event that only 1 one SRM is required to be OPERABLE, per Table 3.3.1.2-1, l ,Q, p footnote (b). only the a. portion of this SR is required. ' Note 2 clarifies that+the three OC1(w 4 J irements can be met by ,J the same n = ;-- OPERABLE S The 12 hour Frequency is based upon operat ng experience and supplements operational controls over refuelt [.M 2 include steps to ensure that the b activities, required by which the LCO a // are in the proper quadrant. 1 SR 3.3.1.2.4 l This Surveillance consists of a verification of the SRM instrument readout to ensure that the SRM reading is greater than a specified minimum count rate. This ensures that the detectors are indicating count rates indicative of neutron flus levels within the core. Verification of the signal to noise ratio also ensures that the detectors are inserted to a norssi operating level. In a fully withdrawn condition, the detectors are sufficiently removed from the fueled . region of the core to essentially eliminate neutrons from reaching the detector. Any count rate obtained while fully withdrawn is assumed to be " noise" only. With few fuel assemblies loaded, the SRMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient " source" material, in the fors (continued) BWR/6 STS B 3.3-40 Rev. O, 09/28/92 l l - -~ -- - ._- . _ - - . _ ._. _ - _ . ()4 c:-/YeD SRM Instrumentation 8 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.4 (continued) : REQUIREMENTS ' of irradiated fuel assemblies, to establish the minimum ' count rate. ' i To accomplish this, the SR is modified by a Note that states that the count rate is not required to be met on an SRM that
- has less than or equal to four fuel assemblies adjacent to the SAM and no other fuel assemblies are in the associated core quadrant. With four or less fuel assemblies loaded around each SAM and no other fuel assemblies in the associated quadrant, even with a control rod withdrawn the i configuration will not be critical. i The Frequency is based upon channel redundancy and other information available in the control room, and ensures that the required channels are frequently monitored dile core i reactivity changes are occurring. When no reactivity !
changes are in progress, the Frequency is relaxed free
- 12 hours to 24 hours. l
, a.%7 --j SR 3.3.1.2.5 7 ER 2_2.LLEM ( e/D Performance of a CHAfWlEL FUNCTIONAL TEST demonsjtates the \ ' l associatedchannelwillfunctionproperly.15W3.3.1.2.5<s} frequi nuvEf 5, and the day Ffoq y ensures that the ' fchan s) OPERA 8LE while ore activ ty c nges could be in j s. is 7 day F uenc is asona le, based on s o at< ex ene and othe Surv illan es (s h as a s s E th ensu p f ioni bet n I TUT b SR . 3.1. 6i requi in 2 ith I on ange 2 or i ' los in 3 nd 4. Sinc core acti ity c ange / not ras y tak place _, the F y has n uten ed / mfres 7 days o 31 ys. Jihe 31 day Frequency is based ch operating experience and on other Surve111ances (such as CHAfstEL CHECK) that ensure proper functioning between CHAISIEL FUNCTIONAL TESTS. l The Note to the Surveillance allows the Surveillance to be i delayed until entry into the specified condition of the i Applicability. The SR aust be performed in MODE 2 within
- 12 hours of entering MODE 2 with IRMs on Range 2 or below.
The allowance to enter the Applicability with the 31 day l (continued) BWR/6 STS B 3.3 41 Rev. O, 09/28/92 6/4 9:-/yd SRM Instrumentation B 3.3.1.2 [a , 2 t.2 8ASES ( wp ] J2 SURVEILLANCE SR 3.3.1.2.5 G sa ~ 3.t r 2 O (continued) ~ REQUIREMENTS Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERA 8LE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances. . . ; /] ,g . . SR 3.3.1.2 d WO - Perfomance of a CHANNEL CALIBRATION verifies the perfomance of the SRM detectors and associated circuitry.. The Frequency considers the plant conditions required to perfom the test, the ease of pe'rforming the test, and the , likelihood of a change in the system or component status. i The neutron detectors are excluded from the CHANNEL ggggp7/ CALIBRATION because they cannot readily be adjusted. The y detectors are fission chambers that are designed to have a '89EA relatively constant sensitivity over the range, and with an accuracy specified for a fixed useful life. REFERENCES None. BWR/6 STS B 3.3-42 Rev. O, 09/28/92 QA4 hiQ Control Rod 81cck Instrumentation B 3.3.2.1 BASES , ~bD APPLICA8LE SAFETY ANALYSES, 1.b. Rod Pattern Controller (continued) x f@.fy p' LCO, and compliance with BPWS are specified in LC0 3.1.6, ' od r ANLICABILITY Pattern " gt ,p Q The Rod Pattern Controller Function satisfies Criterion 3 of the NRC Policy Statement. Since the RPC is a backup to w OCE o>erator ontrol of control rod sequences, only a single h bE) ctanne required OPERA 8LE L (Ref. 6). However, the RPC is designed as a dual channel system and will not function without two OPERABLE channels. Required Actions of LC0 3.1.3, ' Control Rod OPERASILITY," and LC0 3.1.6 may necessitate bypassing individual control rods in the Rod Action Control System (RACS) to allow continued operation with inoperable control rods or to allow correction of a control rod pattern not in compliance with the BPWS. The individual control rods may be bypassed as required by the conditions, and the PC is not considered inoperable . provided SR 3.3.2.1 is set. Compliance with the BPWS, and therefere OPERASILITY of the po o RPC, is required in MODES 1 and 2 with JHERNAL POWER .- s 34% RTP. Men THERMAL POWER is > MR RIr, there is no possible control rod configuration that results in a control - rod worth that could exceed the 280 cal /ge fuel damage limit during a CRDA. In MODES 3 and 4, all control rods are required to be inserted in the core. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SOM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.
- 2. Reactor Mode Switch-Sjiutdown Position During M00ES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be suberitical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch-Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking i control rod withdrawal, thereby preserving the assumptions of the safety analysis.
The Reactor Mode Switch-Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement. i l (continued) BWR/6 STS 8 3.3-46 Rev. O, 09/28/92 l O 9 *'O Control Rod Block Instrumentation B 3.3.2.1 BASES i ACTIONS A d (continued) inoperable if individual control rods are bypassed in the RACS as required by LC0 3.1.3 or LC0 3.1.6. Under these conditions, continued operation is allowed if the bypassing of control rods and movement of control rods is verified by a second licensed operator or other qualified member of the technical staff per SR 3.3.2.1 . c.1 and C.2 ^ If one Reactor Mode Switch-Shutdown Position control rod withdrawal block channel is inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function. Required Action C.1 and Required I {, 9> #/Wl Action C.2 are consistent with the normal action of an > OPERABLE Reactor Mode Switch-Shutdown Position Function to , maintain all control rods inserted. Therefore, there is n'o distinction between Required Actions for the Conditions of one or two channels inoperable. In both cases (one or both W# gje y 's4,,)D channels inoperable), suspending all cent ismediately, and feestately*fu 1 insert withdrawal s 1 insertable Av' control rods in core cells contai ing one or more fuel assemblics will ensure that the core is subcritical, with adequate SON ensured by LC0 3.1.1, "$NUTDolet MARGIN (50M)." l ' Control rods in core cells containing no fuel assembifes do not affect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more. fuel assemblies are fully inserted. , I 3 SURVEILLANCE ~ ( Reviewer's Note: Certain Frequencies 'are based on approved'P-REQUIREMENTS topical reports. In order for a consee to use these / Frequencias, the licensee must ify the Frequencies as Gl required'by the staff SER for topical report. f _ ! As noted at the beginning of the SR the SRs.for each l l Control Rod Block < nstrumentation Function are found in the l SRs column of Table 3.3.2.1-1. The Surveillances are also modified by a Note to indicate that when a channel is placed in an inoperable status solely forl@ performance of required Surveillances, en (continued) BWR/6 STS B 3.3-48 Rev. O, 09/28/92 [ , hA/ c?3 /Q 5/ nfia Codts f Io.0: nsy ala c: g ,, ,,:, 9 L, p., , , 2 .;, ,i ,o ,.,,,, Control Rod 81ock Instrumentation ,,,% haal $< f l.,<r-l dry t c in aq &,, ;,,A a p:!~4 & wd( <; yo"kW orfwcnk 2x1 I / ic,a,u, &c.l ow,.4 Q - .r- / & .. e:n ~ :-) - SURVEILLANCE SR 3.3.2.1 continued) / 5:Y e 12 REQUIREMENTS control rod or correction of a control rod pattern not in i compliance with BPWS. With the contro1 rods bypassed in the N ( RACS, the RPC will not contro1 the movement of these bypasses control rods. FTu ensure the proper bypassing and , u ,o movement of those affected control rods, a second licensed g,,4,, ,,e operator or other qualified member of the technical staff .g;f/ ,g /,c,, /c _ must verify the bypassing and movement of these control Ny,,,j, . u .( y rods? Compliance with this SR a11ows the RPCpto be OPERABLE with these control rods bypassed. j REFERENCES 1. .Sectiony7.6.1.7M
- 2. h , Section/ 15.4.2 [ i ,
h, '-
- 3. NEDE-24011-P-AQ), " General Electrical Standard Application for Re'oad Fuel."Ir;;; . nr-vm;ji (th) q>prmGwissen --
Q tesei, = D - : = ' !. " ; ' : ' " ~ ~ ^
- 4. ' Modifications to the Requirements for Control Rod D Accident Mitigating Systems," BWR Owners Group, Jul 1986.
- 5. MED0-21231, " Banked Position Withdrawal Sequence,"
January 1977.
- 6. NRC SER, Acceptance of Referencing of Licensing Topical Report NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel, Revision 8 Amenht 17," December 27, 1987.
- 7. NEDC-30851-P-A, " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"
October 1988. BWR/6 STS B 3.3-51 Rev. O, 09/28/92 QAR93m)7 PAM Instrumentation B 3.3.3.1 BASES 2. APPLICABLE
- Detemine the potential for causing a gross breach of SAFETY ANALYSES the barriers to radioactivity release; ,
(continued) , Detemine whether a gross breach of a barrier has occurred; and Initiate action necessary to protect the public and to obtain an estimate of the magnitude of any impending threat. The plant specific Regulatory Guide 1.97 analysis (Ref. 2) documents the process that identified Type A and Category I, non-Type A, variables. PAM instrmntation that meets the definition of Type A in Regulatory Guide 1.97 satisfies Criterion 3 of the NRC Policy Statement. Category I, non-Type A, instrumentation is retained in the Technical Specifications (TS) because it is intended to usist operators in minimizing the consequences of accidents. Therefore, these Category I, non-Type A, variables are important for reducing public risk. _ c[ , at /w t) LCO LCO 3.3.3.1 requires two OPERABLE channels for all but one Function to ensure no single failure prevents the operators from being presented with the information necessary to detemine the status of the unit and to bring the unit to, and maintain it in, a safe condition following that accident. Furthennore, provision of two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information. More an two annels y be {daterequire at some its if he Re latory ide 1. analyr s g ned that ilure o one cident nitori g chann OGI resul s in info tion igui (e.g. the re ndant I , dis ays disa ee)tha could ead o defea or toj [fa to acc lish a quire safety /pi stors'unctio .]r The exceptionhthe two channel requirement is primary G.13l containment isolation valve (PCIV) position. In this case, s/ the important information is the status of the 1rimary containment penetrations. indicatobr for each eye =d%his is suffi f;#~Q (continued) BWR/6 STS B 3.3-53 Rev. O, 09/28/92 QsR 93 MD - PAM Instrumentation B 3.3.3.1 BASES d-3 LCO l redundantly verify the isolation status of each solable (continued) penetration either via indicate status of theh val e(4 and prior knowledge of passive valve or via system boundary 3 ', status. If a normally " - PCIV is known to be closed and h .3,I deactivated, position in M ion is not needed to determine /E status. Therefore, the position indication for valves in this state is not required to be OPERA 8LE. T.rumgr of'/6h Listed below is a discussion of the specified instrument 12.S /N Functions listed in Table 3.3.3.1-1, in the accompanying 4/g r .
- 1. Reactor Steam Dome Pressure Reactor steam done pressure is a Category I variable provided to support monitoring of Reactor Coolant System (RCS) integrity and to verify operation of the Emergency Core Cooling Systems (ECCS). Two independent pressure transmitters with a range of 0 psig to 1500 psig monitor pressure. Wide range recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
Reactor Vessel Water Level [ ((ygd Reactor vessel water level is a Category I variable provided to support monitoring of core cooling an to verify operation of the ECCS. The wide rang water level channels provide the PAM Reactor Vessel Water Level Function. The wide range water level channels usature from /IT nes nel h JMEfT f3fyg y (/thac efuel skirt I ide ran water 1 tj o'w isfeeasured by two th Indepenc nt a m erentlal pressure transuitters. The output from these channels is recorded on two independent pen recorders. These recorders are the primary indication used by the operator during an accident. Therefore, the PAM ' Specification deals specifically with this portion of the instrument channel. l (continued) ! BWR/6 STS B 3.3-54 Rev. O, 09/28/92 l 1 1 [1AR 9.3-/v/?;) , INSERT B54A two inches above the top of active fuel to 218 inches above the top of active fuel. The fuel zone water level channels overlap with the wide range channels and measure down to the bottom of the active fuel. Both the wide range and the fuel zone water levels are I 1A/ SERT $3598 In a)$ lion, 4lok (c) o h $$le 3. 3. 3. /-/ rejoiEeJ osvly onc j]osllron irrhea /<o'n her tbose pare'/ral<oY w /Eb ouly ban ose pos:/<in ikScal<a piou,2.0 h h Cov/rof R00M, \ \ 9,33.I d/ F l INSERT RIVER BEND B 3.3-54 @' ~ (LAC 93-MNQ l PAM Instrumentation ! B 3.3.3.1 ) BASES ' LCO Primary Containment Area Radiation (Hiah Ranoel Econtinued) 1 operators in determining the need to invoke site emergency plans. h ~Ar % g?==t, inary containment area radiation (high 4 range) PAM instrumentation consists of 9-e fe? ?c--hg a Eg
- 6. Drvwell S= level ,r A Drywell s evel is a C gory I va able pro ded or verifica n of ECCS fu ions that erate to intain R S i integr y.
F this plant, e drywell s level P instrument ion Aonsists of t following: , /
- 7. D 11 Drain S- vel D 11 drain sump vel is a Cat ory I vari le provid detect breach the RCP8 and or verific fon and 1 g tem surveillan of ECCS func ons that o rate to ntain RCS integrity l ~
For this ~ 1, @ A ant, the drywe drain s evel PAM j / ,/ _ inst tation consist of the following: _] i 7 p a -. e " Primary Containment Isolation' Valve (PCIV) Position A ' PCIV position is provided for verification of containment y Eg integrity. In the case of PCIV position, the important infomation is the status of the containment penetration Pv'[' # '3 % y r Th Lcc_ :: f :: c,a mition incert= m.---- -- = CL l 1 ^C T." . This is sufficient to verify redundantly the T tion status of ach isolable penetration via indicated dorab) s t_us of the" valve a prior knowledge of passive valve or system coundary sta us. If a penetrati i g Y[M ' isol t position indication forau gt-gv the assoc ated penetration flow path is not needed to determine W$)i the status. Therefore, the position indication for valves in an q isolated penetration is not required to be OPERABLE. p.TEU , (25 t (continued) BWR/6 STS B 3.3-56 Rev. O, 09/28/92 l l____ (1Af 93 /MO INSERT B56A two high range containment area radiation signals transmitted from separate radiation elements and continuously recorded and displayed on two control room recorders. The recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
- 8. Drvwell Area Radiation (Hich Range)
Drywell area radiation (high range) is a Category I variable provided to monitor for the potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke site emergency plans. Drywell area radiation (high range) PAM instrumentation consists of two high range drywell area radiation signals transmitted from separate radiation elements and continuously recorded and displayed on two control room recorders. The recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel. INSERT B56B The Lc0 requires one channel of valve position indication in the control room to be OPERABLE for each automatic PCIV in a containment penetration flow path, i.e. , two total channels of PCIV position indication for a penetration flow path with two ' automatic valves. For eontainment penetrations with on1 one automatic PCIV having control room indication, Note @ r requires a single channel of valve position indicaticn to be j OPERABLE. I I INSERT B56C l by at least one closed and de-activated automatic valve, ) closed manual valve, blind flange, or check valve with flow through the valve secured, \ \ s e < g ,9. ,< * .:A / INSERT O RIVER BEND B 3.3-56 W O/1/y3 1 ([ M 9 & /YAl i PAM Instrumentation 8 3.3.3.1 l BASES .d 1, F/o hu/ h. ' , LCO W)f LE Primary Containment Isolation Valve (PCIV) Position 7 continued) l { I w f .# 1 [@ 0,15 kl.ndthe PCIV position PAM instrumentation - , /_ consists of q: fe!!=t t - 4 l gg7 gq _ 1
- 9. Wide Rance Neutron Flux
! Wide range neutron flux is a Ca,tegory I variable provided to l verify reactor shutdown. / l For'this plant, wide range /eutron flux PAM instrument'a tion l consists of the following. . / ' \s~ / -/ ^ i
- 10. 11. Drywell and Containment Hydrocen e b;::-7 Analyzer ' '
Drywell and containment hydrogen an~d analyzers are h "PeCategoryconcentration I instrumentsconditions provided that to detect high ahydrogen potential ori b represent for containment breach. This variable is also important in verifying the adequacy of mitigating actiorts. l G. tr.f 5 da the drywell and containment hydrogen 4W- I qB1 ,- 6 analyzers PAM instrumentation consists of alp #l / - fer =L.a 3 p ; ~ -1.D#6)LT' G O A l h Primary Containment Pressure Primary containment pressure is a Category I variable provided to verify RCS and containment integrity and to i verify the effectiveness of ECCS actions taken to prevent containment breach. Two wide range primary containment i pressure signals are transmitted from separate pressure transmitters and are continuously recorded and displayed on two control room recorders. These recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel. (continued) BWR/6 STS B 3.3-57 Rev. O, 09/28/92 l @k - 93 !YR INSERT B57A containment and drywell hydrogen concentration signals transmitted from two separate hydrogen analyzers and recorded on two two-pen recorders in the control room. One pen records the hydrogen concentration and one pen records the sample point on each of the two independent recorders. Measurement ' capability is provided over the range of 0 to 10 percent hydrogen concentration using a sample drawing system. ' ZA/SEG7* G 5 ?t'$ 1 ikrbrhaaI ,Qcs,fiab iuhes /ro'c fo- closcA / k e ' l conflo[topa.s der eseI auks 1a cou{sinNwf /$ ola froN Vabic as he.tc/i beh 'a USdl( i l .6eedro'~ 7 5 ( j?a-{cie m a 2 } , \ \ g, '.t. & Jl l U INSERT i RIVER BEND B 3.3-57 -10/1/92 (LAR 93+/E0 PAM Instrumentation B 3.3.3.1 BASES , a l ACTIONS function of the instruments, the operator's ability to (continued) diagnose an accident using alternate instruments and tuthods, and the low probability of an event requiring these instruments. , A Note has also been provided to modify the ACTIONS related C , to PAM instrumaatation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, ' h ,?i,M subsequentl3359, subsystems, components, or variables expressed ~in the Condition, discovered to be inoperable or not within limits, 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 Condition. However, the Required Actions for inoperable PAM instrumentation channels provide appropriate compensatory measures for separate inoperable functions. As such, a Note has been provided that allows separate , Condition entry for each inoperable PAM Function. Ad When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERA 8LE status within 30 days. The 30 day Completion Time is based on operatina exnerience and takes into account h the remaining OPERA 8LE(channe M(or'3in the case of a Function that has only one required channer, other non-Regulator Guide 1.97 instrument channels to monitor the Function)y, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the ' S12# low probability of an event requiring PAM instrumentation ' l 47/ during this interval. l i 4 Q prefare an0 El ~E j 5 " b'Y ^ ef e" If a channel has not been restored to OPERA 8LE status in refer / 30 days, this Reauired Action specifie initiation of _ E actions 5 ordan with s Ificat n5J.2.c/Spftia. e ~s," ich uires a ritten por+' 7 ;;:10 r g,
- h. a eve ,- att- o to be buitdic;the report amunars me resu' ts of tne root cause evaluation of mu.. mis d /s/ N '
the inoperability and identifies proposed restorative-actions.f This Action is appropriate in lieu of a shutdown Ae. S / N'%b y *f s4// ae. .,a fs,W,V;;&pm n,yf /0 u % t o.'/ e , W ,$ N Ap M e,y/mg C,,,Rk's B. (continued) M _ , - - s- ' BWR/6 STS B 3.3-59 Rev. O, 09/28/92 i 6.4R 93-!'/E') ' PAM Instrumentation B 3.3.3.1 BASES ACTIONS (Ph (continued) Condition referenced in the Table is Function dependent. Each time an inoperable channel has not met any Required Action of Condition C C- O. e; ;;;:te r ;;7/and th associated Completion Time has expired, condition s entered for that channel and provides for transfe to the appropriate subsequent Condition. 1 For the majority of Functions in Table 3.3.3.1-1, if any Required Action and associated Completion Time of L P '+ Condition C MODE in whic@h the LCD does not Thisapply.is is done bynot met, the p placing the plant in at least MODE 3 within 12 hours. , I The allowed Completion Times are reasonable, based on ! operating experience, to reach the required plant condition ! from full power conditions in an orderly manner and without challenging plant systems. { 6,3, s. / i j ,7 h Aleve! Since alternate means of monitoring C5 cte- = esse! - =$ primary containment area raatation nave ne#ete# " developed and tested, the Required Action is not to shut i g~Q@T== g ,,e own the o' ant but rather to :M ;;. ; ;; ; . . .. . ; .. ;L aul."D These(' alternate means may be temporarily installed if the normal PAM channel cannot be t.- ; j 'a d, ' N A d '
- restored to OPERABLE status within the allotted time. The report provided to the NRC should discuss the alternate peepau. cm8 sufJ seans' used, describe the degree to which the alternate means
" P[ "A#'f are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a l i Cfde A/RC- schedule for restoring the nonnal PAM channels. { i l , SURVEILLANCE The following SRs appl ach PAM ins ntation Function REQUIREMENTS in Table 3.3.3.1-1 etc,p / a, ,, /ed - Sf :- ; y < - $,0 fifatf 2 oY OC, subist L sfeso in yggof gaa . * !3' o,;4 e unro.y .,L O*/' /<O F, M An oSaA,,G ' (continued) J BWR/6 STS B 3.3-61 Rev. O, 09/28/92 , i ( .42 93-/VN].) PAM Instrumentation 8 3.3.3.1 BASES (continued) ! REFERENCES 1. Regulatory Guide 1.97, " Instrumentation for , Light-Water Cooled Nuclear Power Plants to Assess l Plant and Environs Conditions During a ollowing an Accident," (catehrdevg'qa d j ?.e eu.6 m so,
- 2. [ Plant specific de:mu-(e.g.. FSA". 8C egel:tery f l sutde-1rs7, SER 1etter).] ga e 33. y 7,a.,2,
%k . %h. ...a. % % ,u m a cs se I s, . p . ;.3 :. . .t , i Edu Es. J N h* , bil l , " daf,J ,,jve, 30,AT6. 23.'I - - f. I [ .Sedhv p BWR/6 STS 8 3.3-63 Rev. O, 09/28/92 ll4 % - /%'l} Remote Shutdown System B 3.3.3.2 BASES SURVEILLANCE SR 3.3.3.2.1 (continued) REQUIREMENTS verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is criteria it may be an indication that the outside sensor or the@ignal the s proces, sing equipment has drifted Og . . . outside its limit. As specified in the Surveillance, a CHANNEL CHECK is only required for those channels that are "'**NY'" _ g{g f; y Lp Y *
- N 1 .. 1 5 " 5~ -b- hd b g 7 ' ' '" ' ~ ~
2 <$ . :.1 The Frequency is based upon plant operating experience that 47 demonstrates channel failure is rare. , i / I SR 3.3.3.2.2 SR 3.3.3.2.2 verifies each required Remote Shutdown System y ,. c transfer switch and control circuit perfoms the intended , J, ,,,y ,,Ap,- l.O.o Q- function. This verification is performed from the remote / / 1',,,, / ,,,s ,,e a ,<x shutdown panel and local 19. as appropriate. $ This will f '"" ensure that if the contro' room becomes inaccessible, the l l ,a/deo I 73 f arPb "b ' j plant can be placed and maintained in MODE 3 from the remota shutdown panel and the local control stations. However, l i l r,,, j , ,,,, . ea a this Surveillance is not required to be perfonned only ; t c,~ ~>v'/'j#y' during a plant outage. Operating experience demonstrates ' that Remote Shutdown System control channels usually pass y the Surveillance when perfomed at the 18 month Frequency. SR 3.3.3.2.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies the channel respo.nds to esasured parameter values with the necessary range and accuracyy r The 18 month Frequency is based upon operating experience , , and is consistent with the typical industry refueling cycle. > w --- s / \/c,lve poch LG*5 are tydwled A ckueel pAhrwM 5 j Wyalepdel determind durig npednewe of oder va\ve cuNeillmv5 (continued) BWR/6 STS B 3.3-68 Rev. O, 09/2 J2 (AA[ O!YRD 4 R' emote Shutdown System 1 8 3.3.3.2 BASES (continued) i 1 REFERENCES 1. 10 CFR 50, Appendix A, GDC 19. h ' G . ~ g e.S . 6 b e a l T q e h e ne A M an ~ % N - ~ ~ . 1 { )$ f.c.:.2 s 47 t BWR/6 STS B 3.3-69 Rev. O, 09/28/92 i E0C-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE SAFETY ANALYSES, TCY Fast Closure. Trin Oil Pressure-Low (continued)
- LCO, and APPLICABILITY Closure, Trip 011 Pressure-Low in anticipation of the r -
transients that would result from the closure of these valves. / The E0C-RPT decreases reactor power and aids the >' d reactor scras in ensuring that the MCPR SL is not exceeded ( 2 2 d during the worst case transient. / / . /en 4]' e ' Fast closure of the TCVs is determined by measuring the EHC fluid pressure at each control valve. There is one pressure O '? : ". Y /'. transmitter associated with each control valve, and the ) Tm /e- %du E, signal from each transmitter is assigned to a separate trip channel. ! The logic for the TCV Fast Closure. Trip 031 1'L ;,g m ,y#agn ! Preswn-Low Function is such that two or more TCVs must be ve/v" " " j ' M "'" - closed (pressure transmitter trips) to produce an E00-RPT. t t /,s y A Twem This Function must be enabled at THERMAL POWER a 40% ATP. \ 90 % ,;' _'.c This is nornelly accomplished automatically by pressure transmitters sensing turnine first stage pressiiM ( rnw a .: channels of TCV Fast Closure. Trip 011 Pressure Low, Ts Four w with-two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an E0C-RPT from this Function on a valid signal. The TCV Fast Closure, Trip 011 Pressure-Low Allowable Value is selected high enough to detect imminent TCV fast closure. This protection is required consistent with the analysis, whenever the THERMAL POWER is a 40% RTP with any recirculating pump in fast speed. Below 40% RTP or with recirculation Dome Pressure pumps in slow speed, the Reactor Vessel Steam High and the APRM Fixed Neutron Flux-High Functions of the RPS are adequate.to maintain the necessary safety margins. The turbine first stage pressure / reactor power relationship for the setpoint of the automatic enable is identical to that described for TSV closure. ACTIONS Revi 's Note: C ain LC0 C etion Times based on appro to Q, the ines, pical rts. In o er for a lice ee to use the li ensee must ju tify the Comp 1 tion Times as &; C ired by the taff Safety E aluation Repo (SER) for the ] (_ t pical report _f A Note has been provided to modify the ACTIONS related to EOC-RPT instrumentation channels. Section 1.3, Completion (continued) BWR/6 STS B 3.3-74 Rev. O, 09/28/92 }M 9 -'61_) l EOC-RPT Instrumentation 8 3.3.4.1 BASES i l SURVEILLANCE SR 3.3.4.1.6 (continued) REQUIREMENTS accident analysis. The EOC-RPT SYSTEM RESPONSE TIME acceptance criteria are included in Reference 6. A Note to the Surveillance states that breaker interruption S A.f time may be assumed from the most recent performance of SR 3.3.4.1.7. This is allowed since the time to open the d?' ' contacts after energization of the trip coil and the are M. suppression time are short and do not appreciably change, ' due to the design of the breaker opening device and the fact t !hy }d <- that the breaker is not routinely cycled. / / y /j',~ ' , . . '. ,j 4: , EOC-RPT SYSTEM RESPONSE TIME tests are conducted on an /M '~ r' " 18 month STAGGERED TEST BASIS.* Response times cannot be 7 determined at power because operation of final actuated {U 3 . u m ,'e'yM# ., devices is required. Therefcre, m :: :: r.) Frequency is S, , g ,g J 4 f7VMc consistent with the typical industry refueling cycle and is Cy C,/ x , c n y>/**- . based upon plant operating experience, which shows that random failures of instrumentation components that cause ;g,..,/ b e Q ' _ ,, , / serious response time degradation, but not channel failure, are infrequent occurrences. , // / # ~ , ' s t (' p' e w / i -lo + S '< ' Sg 3,3,4,1,7 ,. , se y' :e . of aem 73,, pg This SR ensures e-;g_,:= g . that __ the 7 RPT breaker interruotion time W < -^/ a = ^--^-l is provided to the EOC-RPT 5T5 TEM RESPONSE TIME test. The 60 month !
- /^ j
' ',,,y . Frequency of the testing is based on the fficulty of / performin y 'e ny /b .' ,., breakers.g the test and the reliability f the circuit , i.2asra.ej Tr.73 AT 7M .T.' ~ REFERENCES 1. h ,(j %-"-- h.:[ - t~~ "" f r: _ ..iei i un bph
- 2. @AR, SectionNs.2.2r b3.@, Sectionsf {is.t.ihis.t.2[andjis.i.3[S
- 4. h , Sections ,andf7:5MR.
'T_.2.zs ss.:.G ~3s.2 M (continued) BWR/6 STS B 3.3-80 Rev. O, 09/28/92 hM 9 - 14 PL) ATWS-RPT Instrumentation B 3.3.4.2 BASES ACTIONS of the Condition continue to apply for each additional (continued) failure, with Completion Times based on initial entry into the Condition. However, the ftequired Actions for inoperable ATWS-RPT 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 ATWS-RPT instrumentation channel. f / A.1 and A.2 / With one or more channels inoperable, but with ATWS-RPT I capability for each Function maintained (refer to Required Action B.1 and C.1 Bases), the ATWS-RPT System is capable of , perfoming the intended function. However, the reliability and redundancy of the ATWS-RPT instrumentation is reduced, / such that a single failure in the remaining trip system P ' ,~,' 9, p could result in the inability of the ATWS-RPT System to (*7 perfom the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to , OPERABLE status. Because of the diversity of sensors ( available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse ( Functions, and the low probability of an event requiring the initiation of ATWS-RPT, 14 days is provided to restore the N ' inoperable channel (Required Action A.1). Alternately, the inoperable channel may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accomundate a single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an l inoperable breaker, since this may not adequately compensate ' for the inoperable breaker (e.g., the breaker may be l inoperable such that it will not open). If it is not ! desirable to place the channel in trip (e.g., as in the case I _ where placing the inoperable channel would result in an ' h--@fd5 GRPT) .or if the inoperable channel is the result of an inoperable breaker, Condition D aust be entered and its Required Actions taken. (continued) BWR/6 STS B 3.3-86 Rev. O, 09/28/92 (k.4QC - /M0 ATWS-RPT Instrumentation B 3.3.4.2 BASES ACTIONS Ed (continued) Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable. untripped channels within the same Function result in the Functior not maintaining ATWS-RPT trip capability. A Function is considered to be maintaining ATWS-RPT trip capability when sufficient channels are OPERABLE or in trip such that the ATWS-RPT System will generate a trip signal from the given Function on a valid signal, and both recirculation pumps can i be tripped. This requires two channels of the Function in l g4 3 the same trip system to each be OPERABLE or in trip, and the four motor breakers (two fast speed and two LFMG) to be ~, 7 OPERABLE or in trip.
- l. The 72 hour Coupletion Time is sufficient for the operator l 1
vgr7 to take corrective action (e.g., restoration or tripping of l / g-),gM channels) and takes into account the likelihood of an even.t requiring actuation of the ATW5-RPT instrumentation during this period and the fact that one Function is still maintaining ATWS-RPT trip capability. \ l \ l l N f.d l 1 Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within both Functions result in both Functions not maintaining ATWS-RPT trip capability. The description of a Function meintaining ATWS-RPT trip capability is discussed in the Bases for Required Action 8.1, above. The I hour Completion Time is sufficient for the operator to take corrective action and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period. D.1 and D.2 ~ With any Required Action and associated Completion Time not met, the plant must be brought to a MODE or other specified condition in which the LC0 does not apply. To achieve this ) status, the plant must be brought to at least MODE 2 within ' 6 hours (Required Action D.2). Alternately, the associated recirculation pump may be removed from service since this (continued) BWR/6 STS B 3.3-87 Rev. O, 09/28/92 ]/y; % Wll) ECCS Instrumentation B 3.3.5.1 BASES BACKGROUND Low Pressure Coolant Iniection Subsystem (continued) minimum flow return line valve is opened. The valve is automatically closed if flow is above the minimum flow setpoint to allow the full system flow assumed in the analyses. y __ TheRHRtestline[suppressionpoolcooling) isolation (HF r --- ;-- _: .____ ..-ian-m valves (which are also PCIVs) are closed on a LPCI iniciation signal to allow full system flow assumed in the accident analysis and maintain i ' containment isolated in the event LPCI is not operating. N The LPCI subsystems monitor the pressure in the reactor ! M'1 vessel to ensure that, prior .to an injection valve opening, the reactor pressure has fallen to a value below the LPCI - subsystem's maxi design pressure. The variable is dat monitored _by undant transmitters per Division, whi,ch art, in turn, connected to four trip units. The outputs-of the four Division 2 LPCI (loops B and C) trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic. The Divis'on 1 LPCI (loop A) receives its signal from the LPCS logic, which uses a similar one-out-of-two taken twice logic. Hiah Pressure Core Sorav System The HPCS System may be initiated by either automatic or ! manual means. Automatic initiation occurs for conditions of ' Reactor Vessel Water Level-Low Low, level 2 or Drywell Pressure-High. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic for each variable. The HPCS System initiation signal is a sealed in signal and must be manually reset. The HPCS pump discharge flow is monitored by a flow transmitter. When the pump is running and discharge flow is low enough that pump overheating may occur, the minimum flow return 1 ne valve is opened. The valve is automatically closed if flow is above the minimum flow setpoint to allow full system flow assumed in the accident analyses. The HPCS test line isolation valve (which is also a PCIV) is closed on a HPCS initiation signal to allow full system flow (continued) BWR/6 STS 8 3.3-94 Rev. O, 09/28/92 (2 y ;2- W Q ECCS Instrumentation B 3.3.5.1 BASES 1 APPLICABLE h.1.f. 2 SAFETY ANALYSES, Low Pressure Coolant Iniection and Low LCO, and Pressure (continued)Tore Sorav P- Discharae Flow-Low (9voass) APPLICABILITY LC0 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.
- b. 2.k. Manual Initiation The Manual Initiation push button channels introduce signals into the appropriate ECCS logic to provide manual initiation capability and are redundant to the automatic protective instrumentation. There is one push button for each of the two Divisions of low pressure ECCS (i.e., Division 1 ECCS, LPCS and LPCI A; Division 2 ECCS, LPCI B and LPCI C).
e .Id l - The Manual Initiation Function is no sumed in any o %* accident or transient analyses in th . However the- ' -) Function is retained forG5ereli ,;C.2x, n.e . ..n, .. ;., ei r ~ the low pressure ECCS function as requirea by the NRC in.the plant licensing basis. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons. Each channel of the Manual Initiation Function (one channel per Division) is only required to be OPERA 8LE when the associated ECCS is required to be OPERABLE. Refer to LC0 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems. Hiah Pressure Core Sorav System 3.a. Reactor Vessel Water Level-Low Low. Level 2 me Low RPV water level indicates that the capability to cool the fuel may be threatened. too far, fuel damage could result.Should RPV water level decrease Therefore, the HPCS OCA System and associated DG level above the top of the activeinitiated at Level 2 to maintain fuel. The Reactor Vessel Water Level-Low Low, Level 2 is one of the Functions assumed to be OPERA 8LE and capable of initiatin the transients analyzed in References 1 and 3. g HPCS The Reactor during Vessel Water Level-Low Low, Level 2 Functh,n associated l with HPCS is directly assumed in the analysis of the l recirculation line break (Ref. 2). The core cooling (continued) BWR/6 STS 8 3.3-104 Rev. O, 09/28/92 . - . _ . = - _ (M4 9:'-ND i ECCS Instrumentation B 3.3.5.1 BASES t APPLICA8LE 3.f. 3.a. HPCS P- Discharae Pressure-Hiah (Rvnass) and SAFETY ANALYSES, HPCS System Flow Rate-Low (Bvnass) (continued) LCO, and APPLICABILITY of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding comparature remains below the ; limits of 10 CFR 50.46. , one flow transmitter is used to detect the HPCS System's flow rate. The logic is arranged such that the transmitter causes the minimum flow valve to open, provided the HPCS pump discharge pressure, sensed by another transmitter, is ; high enough (indicating the pump is operating). The logic will close the minimum flow valve once the closure setpoint is exceeded. (The valve will also close upon HPCS discharge pressure decreasing below ths setpoint.) pump The HPCS System Flow Rate-Low and HPCS Pump Discharge Pressure-High Allowable Value is high enough to ensure that pump flow rate is sufficient to protect the pump, yet low' , enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core. The HPCS < Pump Discharge Pressure-High Alloweble Value is set high enough to ensure that the valve will not be open when the pump is not operating. One channel of each Function is required to be OPERABLE when the HKS is required to be OPERABLE. Refer to LC0 3.5.1 and ; LCO 3.5.2 for HPCS Applicability Bases. 3.h. Manual Initiation , I The Manual Initiation push button channel introduces a signal into the HPCS logic to provide manual initiation l capability and is redundant to the automatic protective ! inst'rumentation. There is one push button for the HPCS 1 System. 4,5.JJ The Manual Initiation Function is no sumed in any ( j3l accident or transient analysis in th * . However, the ' d Function is retained for s..i.ll ; r .:r r : : t r~itY c'M ' the HPCS function as required by the MRC in the plant licensing basis. There is no Allowable Value for this Function since the channel is mechanically actuated based solely on the position of the push button. One channel of the Manual (continued) BWR/6 STS 8 3.3-109 Rev. O, 09/28/92 l (Mtsur-/WI) ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 4.o. 5.f. ADS Bvoass Timer (Hioh Orywell Pressure) SAFETY ANALYSES, (continued) LCO, and APPLICA8ILITY is chosen to be short enough so that there is still time after depressurization for the low pressure ECCS subsystems to provide adequate core cooling. Four channels of the ADS Bypass Timer Function are only required to be OPERA 8tE when the ADS is required to be , CPERA8LE to ensure that no single instrument failure can ' preclude ADS initiation. Refer to LCO 3.5.1 for ADS Applicability Bases. 4.h. S.a. Manual Initiation The Manual Initiation push button channels introduce signals into the ADS logic to provide manual initiation capability and are redundant to the automatic protective . instrumentation. There are two push buttons for each ADS trip system (total of four). . %g ,1i *~ The Manual Initiation Function is not used in any accident or transient analyses in th . However, the _ PI s'j Function is retained for @ : =11 n n er --r e r: :'tv :C_ , the ADS function as required by the NRC in the plant ! licensing basis. There is no Allowable Value for this Function since the . channel is mechanically actuated based solely on the position of the push buttons. Four channels of the Manual Initiation Function (two channels per ADS trip system) are , only required to be OPERABLE when the ADS is required to be OPERA 8LE. Refer to LC0 3.5.1 for ADS Applicability Bases. l j ACTIONS Revi r's Note: Certai LCO Completion Times are based on' ) app ved topical report . In order for a icensee to use i th times, the licens must justify the ampletion Times as ired by the staff Safety Evaluation eport (SER) for the . ical report. _j A Note has been provided to modify the ACTIONS related to ECCS instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent g subsystems, components, or variables -~ (continued) l BWR/6 STS 8 3.3-114 Rev. O, 09/28/92 l l hM % ~/W l) ICCS Instrumentation 8 3.3.5.1 BASES ACTIONS F.1 and F.2 (continued) ,,5 d , Osi7 - channel to OPERABLE status if both HPCS and RCIC are ypugf OPERA 8LE. If either HPCS or RCIC is inoperable, the time is pnA shortened to 96 hours. If the status of HPCS or RCIC 7 J' w. changes such that the Completion Time changes fion 8 days to w- 96 hours, the 96 hours begins upon discovery of HPCS or RCIC inoperability. However, total time for an inoperable, untripped channel cannot exceed 8 days. If the status of N HPCS or RCIC changes such that the Completion Time changes \ from 96 hours to 8 days, the ' time zero" for beginning the 8 day " clock" ins upon discovery of the inoperchle, y untripped channe,. '.If the inoperable channel cannot be restorva to OPERAsLE status within the allowable out of service time, the channel must be placed in the tripped i condition per Required Action F.2. Placing the inoperable ! channel in trip would conservatively compensate for the- ! inoperability, restore capability to accommodate a single
- l failure, and allow operation to continue. Alternatel it is not desired to place the channel in trip (e.g.,y, if as.in the case where placing the inoperable channel in trip would result in an initiation), Condition H must be entered and its Required Action taken.
G.1 and G.2 , Required Action 6.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within similar ADS trip system Functions result in automatic initiation capability being lost for the ADS. Automatic initiation capability is lost if either (a) one Function 4.c channel and one Function 5.c channel are inoperable, (b) one or more Function 4.e channels and one or more Function 5.e channels are inoperable, (c) one or more Function 4.f ! channels and one or more Function 5.e channels are inoperable, or (d) one or more Function 4.s channels and one l or more Function 5.f channels are inoperable. Inthissituation(lossofautomaticinitiationcapability), the 96 hour or 3 day allowance, as applicable, of Required Action G.2 is not appropriate, and all ADS valves must be declared inoperable within 1 hour after discovery of loss of ADS initiation capability in both trip systems. The Note to ; Required Action G.1 states that Required Action 6.1 is only applicable for Functions 4.c, 4.e, 4.f. 4.g 5.c 5.e, (continued) j BWR/6 STS 8 3.3-123 Rev. O, 09/28/92 -(hdi9:-/9'b) \ INSERTN 12 m Sinc s not required t e OPE BLE at reactor s done pressurW50 psig,[ RC C will not be considered inoperable below this pressure regardless of the actual status capability of RCIC. I f# p1 ( I l l 1 ) 1 INSERT l RIVER BEND B 3.3-123 10/1/93' i AX % /'/ ECCS Instrumentation B 3.3.5.1 BASES ACTIONS G.1 and G.2 (continued) ; and 5.f. Required Action G.1 is not applicable to Functions 4.h and 5.g (which also require entry into this Condition if a channel in these Functions is inoperable), since they are the Manual Initiation Functions and are not assumed in any accident or transient analysis. Thus, a , total loss of manual initiation capability for 96 hours or 8 days (as allowed by Required Action G.2) is allowed. The Completion Time is intended to allow the operator time l ' to evaluate and repair any discovered inoperabilities. This ' Completion Time also allows for an exception to the normal l ' time zero" for beginning the allowed outage time " clock." ! For Required Action G.1, the Completion Time only begins : opon discovery that the ADS cannot be automatically 1 initiated due to inoperable channels within similar ADS- trip i ' system Functions, as described in the paragraph above. The i I hour Completion Time from discovery of loss of initiation g pj p capability is acceptable because it minimizes risk while: allowing time for restoration or tripping of channels. I I' >f Because of the diversity of sensors available to provide i f initiation signals and the redundancy of the ECCS design, an l allowable out of service time of 8 days has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable ' JA" ' ' r channel to OPERABLE status if both MpCS and RCIC are N )jh- OPERA 8LE (Required Action G.2). If either HPCS or RCIC is inoperable, the time is reduced to 96 hours. If the status ( of HCPS or RCIC changes such that the Completion Time changes from 8 days to 96 hours, the 96 hours begins upon discovery of HPCS or RCIC inoperability. However, total . time for an inoperable channel cannot exceed 8 days. If the status of HPCS or RCIC changes such that the Completion Time changes from 96 hours to 8 days, the " time zero" for beginning the 8 day " clock" begins upon discovery of the 3 ir. operable channel. If the inoperable channel cannot be restored to OPERA 8LE status within the allowable cut of service time, Condition H must be entered and its Required Action-taken. The Required Actions do not allow placing the channel in trip since this action would not necessarily result in a safe state for the channel in all events. i I 1 (continued) c BWR/6 STS B 3.3-124 Rev. O, 09/28/92 h/ 9;-/vr /]- \, / NSERT B134A / N -- N, Since RCI(is not required to be OPERABIE at reactor stea'm \ done pressQres < 150 psig, RCI ill not be considered inoperable beloMwi_s pressur egardi 'as of the actual status , or capability of RCI'. , \ 3<. , , i i 1 F i INSERT ' RIVER BEND B 3.3-124 10/1/93 ~. -. - _ _ _ . _ - . - - - _ _ - (LM 93 -N/I ) RCIC System Instrumentation 8 3.3.5.2 , BASES !
- APPLICA8LE 5. Manual Initiation SAFETY ANALYSES, LCO, and The Manual Initiation push button switch introduces a signal APPLICA81LITY into the RCIC System initiation logic that is redundant to l (continued) the automatic protective instrumentation and provides manual initiation capability. There is one nush button for the RCIC System.
The Manual Initiation Function is no sumed in any <8 C-Ale 3 Lecident or transient analyses in th . However. the Function is retained for er -:!! n ::::::y ::: ::P :r: tty $ e - ,I -ltheRCICfunctionasrequiredbytheswu.Intheplant licensing basis. There is no Allowable Value for this Function since the channel is mechanically actuated based solely on the position of the push button. One channel of Nanual Initiation is required to be OPERA 8LE when RCIC is required ! to be OPERA 8LE. / -3 : ACTIONS ! Revi r's Note: Ce ain LC0 Comp 1 ion Times a ' based on appro ed topical re rts. In ordel for a licens to use 2 Bl the ines, the lic see must justPy the Comple ion Times as reg red b the staff Safety .rvaluation Re rt (SER) for ( _ th topica repo . _J A Note has been provided to modify the ACTIONS related to RCIC System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been / (-.__ 7 entered. subsequent AEnit, subsystems, components, or verlanies expressee in the Condition discovered to be f inoperable or not within limits will not result in separate l OC# entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each aceitional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RCIC System 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 RCIC System instrumentation channel.
l (continued) l 1 , BWR/6 STS B 3.3 134 Rev. O, 09/28/92 l l l , - 1 G/oul Q Q aat/ 08 %-/V* ID e' Primary Containment Isolation Instrumentation . B 3.3.6.1 B 3.3 INSTRUMENTATION M 9/yse 8 3.3.6.1 Primary Containment Isolation Instrumentation BASES N mA efrywe// % <// n o / J o a v d e h BACKGROUND The primary containeen so ation instrumentati - ~ ~ - automatically initiates closure of a riat primary ~ containment isolation valves (PCIVs . The unction of the ! PCIVs, in combination with other accident mitigation , systems, is to limit fission product release during and : following postulated Design Basis Accidents (DSAs). Primary i containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the gangy environment will be consistent with the assumptions used in theanalysesforaDBA.y The isolation instrumentation includes the sensors, relays, ~ t o ;, p and switches that are necessary to cause initiation of t 3 #f rimar p(RCPS)y containment and reactor coolant pressure boundary isolation. Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint ; is exceeded, the channel output relay actuates, which then outputs a primary containment isolation signal to the isolation logic. Functional diversity is provided by [9,p _ monitoring a wide range of independent parameters. The ( # ~ input parameters to the isolation logic are (a) reactor ' i vessel water level, (b) ambient neiff*=niaD a-temperatures, (c) main steam line (M5L) flow measurement, d Standby Liquid Control (SLC) System initiation, - e condenser vacuum loss, (f) main steam line pressure, , g reactor core isolation cooling (RCIC) and RCIC/ residual heat removal (RHR) steam line flow, (h) ventilation exhaust radfation, (1) RCIC steam line exhaust diaphragm pressure, reactor (k) pressure, water cleanup (j) RCIC (RWCU) turbine : differential flow, (1) reactor steam dome pressure, and ; (m)drywellpressure. Redundant sensor input signals are i provided from each such isolation initiation parameter. The : only exception is SLC System initiation. In addition, ,/ manual isola ogics is provided. O The primary containment' isolation instrumentation has inputs to the trip logic from the isolation Functions listed below. l (continued) BWR/6 STS B 3.3-141 Rev. O, 09/28/92 l .-- ~.. .. - .- . . G A X 9J-/t' D 0 INSERT B141A The isolation of the drywell isolation valves, in combination I with other accident mitigation systems, functions to ensure . that steam and water releases to the drywell are channeled to ' / the suppression pool to maintain the pressure suppression i function of the the primary containment. \ s 1, 5.6./ i hl i i 4 INSERT N/ RIVER BEND B 3.3 10/1/92 7 i (.V/ l M AR' 9^ -/W L) Primary Containment Isolation Instrumentation B 3.3.6.1 i BASES BACKGROUND 1. Main Steam Line Isolation . (continued) l Most Main Steam Line Isolation Functions receive inputs from four channels. The outputs from these channels are combined ' in one-out-of-two taken twice logic to initiate isolation of l all main steam isolation valves (MSIVs). The outputs from the same channels are arranged into two two-out-of-two logic trip systems to isolate all MSL drain valves. Each MSL drain line has two isolation valves with one two-out-of-two logic systes associated with each valve. The exception to this arrangement is the Main Steam Line Flow-High Function. This Function uses 16 flow channels, four for each steam line. One channel free each steam line inputs to one of four trip strings. Two trip strings make up each trip system, and both tr< p systems must trip to cause an MSL isolation. Each trip string has four inputs 3 (one per MSL), any one of which will trip the trip string.. The trip strings within a trip system are arranged in a one-out-of-two taken twice logic. Therefore, this is ; effectively a one-out-of-eight taken twice logic arrangement ' to ihitiate isolation of the MSIVs. Steilarly, the 16 flow channels are connected into two two-out-of-two logic trip systems (effectively, two one-out-of-four twice logic), with each trip system isolating one of the two MSL drain valves. gy h./ 2. Primary Containment tX Each Primary Containment (Isolation Function receives inputs I free four channels. The outputs from these channels are odg jf/"# f arranged into two two-out-of-two logic trip systems. One nolded trip the o_Wer trip system initiates isolation of all outboardsytten init vu4e.s__f PCIVs." Each trip system logic closes one of the two valves on each penetration so that operation of either trip system isolates the penetration.
- 3. Reactor Core Isolation Coolina System Isolation r
M Most Functions receive input from two channels, with each ^ ws ' o stem using one-out-of-one logic. Mob ' -Y channel Functio i I (RHR Equipment Room Temperature) @ ' one channel' each trip system in each room for a total of , four ch:ennels fy- Tw,@gR, but the logic is the same 1 - I l , (continued) BWR/6 STS B 3.3-142 Rev. O, 09/28/92 l SM n-twy Primary Centainment isolation Instrumentation B 3.3.6.1 BASES BACKGROUND 3. Reactor Core Isolation Coolino System Isolation (continued) (one-out-of-one). Each of the two trip systems is connected to one of the two valves on each RCIC penetration so that operation of either trip system isolates the penetration. The exct.ption to this arrangement is the RCIC Turbine Exhaust Diaphragm Pressure-High Function. This Function' receives input from four turbine exhaust diaphragm pressure channels. The outputs from the turbine exhaust diaphragm pressure channels are connected into two two-out-of-two tri; systems, each trigen isolatino one of_ the two RCIC m valv mu is ne .i isetwm suihk wLick ce hetde wl 3 Ops' , e A,4 a.cics + m s ww % 3m. i j/
- 4. Reactor Water Cleanun Sys'tes Isolation Most Functions receive input from two channels with each n, s c,,I , channel in ogg tri sstesusingone-out-of-onelogic.vpD
("m -J Functions?4.pff (RWCU Pump Room Temperature) t f M channel in each trip system in each room for a total of four j channels $cr F=ctusftut the logic is the same (one-out-of-one). Each of the two trip systems is connected to one of the two valves on each RWCU penetration so that operation of either trip system isolates the penetration. The exception to this arrangement is the Reactor Vessel Water Level-Low Low, Level 2 Function. This Function receives input from four reactor vessel water level channels. The outputs from the reactor vessel water level channels are connected into two two-out-of-two trip systems, each trip system isolating one of the two RWCU valves. 5. -Mg.;tik 0;;clir@ System Isolation SuarbulDW-lml- ,(?yk yG Q- L La , Luel I; pg The,U,;tdr x c 2 Tin Isolation FunctionIr(ec.- eives input L. signals from instrumentation for the Reactor) Vessel Water Level-Low, Level 3; Drywell Pressure-High;f Dome Pressure-High; and RHR Equipment Room Ambient 41HP2 % ,1 ~ h eI u M u h&garant4sT Tescerature-High Functions. The Reactor q R d Vessel Water Level-Low,Meactor Steam Dome Pressure-High, Ogg d"d 'k [b and Drywell Pressure-High Functions each have four channels. The outputs from the reactor vessel water level and drywell pressure channels are connected into two two-out-of-two trip systems. The reactor steam done pressure is arranged into two one-out-of-two trip systems. (continued) BWR/6 STS B 3.3-143 Rev. O, 09/28/92 hi 93 -/<dQ Primary Containment Isolation Instrumentation B 3.3.6.1 l BASES-BACKGROUND
- 5. mqLD
;r . m M oc System Isolation (continued) / # ) The RHR Equipmenk Room Ambient
- Offfe m ti N eaperature
(*! J Function $'receivts) input from four channels with each channel in one trip system in one room using one-out-of-one logic. l l Each of the two trip systems is connected to one of the two I valves on each shutdown cooling penetration so that j operation of either trip system isolates the penetration. 1. W 1 - #pt l - / \ a d b y~ e// / APPLICABLE The isolation signals generated by the pr mary u ntainment 2 SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the LCO, and safety analyses of References 1 and 2 to initiate closure of APPLICA8ILITY valves to limit offsite doses: Refer to LC0 3.6.1.3, " Primary Containment Isolation Valves (PCIVs)," Applicable SafetyAnalysesBases,formoredetail.f ~ ' zNu /T Primary containment d isclation instrumentation satisfies < dfwa Criterion 3 of the NRC Policy Statement. Certain I L instrumentation Functions are retained for other reasons and I are described below in the individual Functions discussion. ad hyd The OPERASILITY of the primary containmentlinstrumentation \ g< is dependent on the OPERASILITY of the individual A' instrumentation channel Functions specified in / Table 3.3.6.1-1. Each Function must have a required number M 3,/ of OPERA 8LE channels, with their setpoints within the 4g specified Allowable Values, where appropriate. A channel is inoperable if its actual trip setpoint is not within its ' required Allowable Value. The actual setpoint is calibrated e consistent with applicable setpoint methodology assumptions. ' Each channel must also respond within its assumed response time where appropriate. l J Allowable V. lues are specified for each Primary Containment l j9t[ sIsolation Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less ! conservative than the nominal trip setpoint, but within its ' Allowable Value, is acceptable. Trip setpoints are those f predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and I (continued) BWR/6 STS B 3.3-144 Rev. O, 09/28/92 aL@ Aa n K- + - - 4-+- 3 ^--.A - & R - ' (AAR93tynI) , INSERT B144A The isolation of the drywell isolation valves, in combination with other accident mitigation systems, functions to ensure ~ that steam and water releases to-the drywell are channeled to / the suppression pool to maintain the pressure suppression - function of the the primary containment. Refer to LCO \ 3.6.5.3, "Drywell Isolation Valves," Applicable Safety Analysis Bases, for more detail. , \ \ g,3.b' 42S 1 f INSERT . RIVER BEND B 3.3- 14W /Y'l (IM 93-neD Primary Containment Isolation Instrumentation B 3.3.6.1 BASES , APPLICABLE when the measured output value of the process parameter SAFETY ANALYSES, exceeds the setpoint, the associated device (e.g., trip LCO, and unit) changes state. The aralytic limits are derived from APPLICABILITY the limiting values of the process parameters obtained from (continued) the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then detemined accounting for the remaining instrument errors (e.g., drift) . The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for . channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for. Certain Emergency Core Cooling Systems (ECCS) and RCIC valves (e.g., minimum flow) also serve the dual function of automatic PCIVs. The signals that isolate these valves are also associated with the automatic initiation of the ECCS , and RCIC. The instrumentation and ACTIONS associated with l these signals are. addressed in LC0 3.3.5.1, "ECCS i Instrumentation," and LC0 3.3.5.2, "RCIC Instrumentation," and are not included in this LCO. In general, the individual Functions are required to OPERA 8LE in MODES 1, 2, and 3 consistent with the rem %, , % Applicability for LC0 3.6.1.1, " Primary Containeen - Functions that have different Applicabilities are discussed below in the individual Functions discussion. The specific Applicable Safety Analyses, LCO, and I [r[r )77a// LCO c,3 2.k,r/ Applicability discussions are listed below on a Function by Function basis, a,y/hA cl . s 1. Main Steam Line Isolation s %2/ 1.a. Reactor Vessel Water Level-Low Low Low. Level 1 423 . Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease tg far, fuel damage could result. Therefore, isolation of tNF MSIVs and other interfaces with the reactor vessel occurs to prevent offsite dose limits _ from being exceeded. The Reactor Vessel Water Level-Low Low Low, Level 1 Function is one of the many Functions (continued) BWR/6 STS B 3.3-145 Rev. O, 09/28/92 . (U/ #N$ Primary Containment Isolation Instrumentation B 3.3.6.1 l BASES 1 APPLICA8LE 1.b. Main Steam Line Pressure-Low (continued) SAFETY ANALYSES, LC0, and separated from each other, each transmitter is able to APPLICABILITY detect low MSL pressure. Four channels of Main Steam Line Pressure-Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value was selected to be high enough to prevent excessive RPV depressurization. The Main Steam Line Pressure-Low Function is only required to be OPERA 8LE in MODE 1 since this is when the assumed transient can occur (Ref. 2). This function isolates the Group 1.c. Main Steam Line Flow-Hiah ralves. h Main Steam Line Flow-High is provided to detect a break ~of the MSL and to initiate closure of the M51Vs. If the i, team were allowed to continue flowing out of the break, the reactor would depressurire and the core could uncover. If the RPV water level decreases too far, fuel damage could occur. Therefore, the isolation is initiated on high flow to prevent or minimize core damage. The Main . Steam Line oel Flow-High Function is directly assumed in the anal -f the main stglas line break (NSLS) accident (Ref.1)ysis ~ i <p . The of ~/ isolation action, along with the scram function of the RPS, ! ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR 100 limits. The M5L flow signals are initiated from 16 transmitters that ' are connected to the four M5Ls. The transmitters are arranged such that, even though physically separated from i each other, all four connected to one steam line would be e able to detect the high flow. Four channels of Main Steam Liae Flow-High Function for each unisolated M5L (two channels per trip system) are available and are required to be OPERABLE so that no single instrument failure will preclude detecting a break in any individual MSL. The Allowable Value is chosen to ensure that offsite dose i
- limits are not exceeded due to the break.
(continued) J BWR/6 STS 8 3.3 147 Rev. O, 09/28/92 h 92 d@ Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICA8LE 1.c. Main Steam Line Flow-Hiah (continued) SAFETY ANALYSES, i LCO, and This function isolates the Group alves. P4 l APPLICABILITY ' 1.d. Condenser Vacuum-Low The Condenser Vacuum-Low Function is provided to prevent, overpressurization of the main condenser in the event of a loss of the main condenser vacuum. Since the integrity of the condenser is an assumption in offsite dose calculations, the Condenser Vacuum-Low Function is assumed to be OPERABLE i and capable of initiating closure of the MSIVs. The closure of the MSIVs is initiated to prevent the addition of steam that would lead to additional condenser pressurization and possible rupture of the diaphragm installed to protect the turbine exhaust hood, thereby preventing a potential radiation leakage path following an accident. Condenser vacuum pressure signals are derived from four pressure transmitters that sense the pressure in the condenser. Four channels of Condenser Vacuum-Low Function are available and are required to be OPERABLE to ensure no single instrument failure can preclude the isolation function. The Allowable Value is chosen to prevent damage to the condenser due to pressurization, thereby ensuring its - integrity for offsite dose analysis. As noted (footnote (a) , to Table 3.3.6.1-1), the channels are not required to be ' OPERA 8LE in MODES 2 and 3, when all turbine stop valves (TSVs) are closed, since the potential for condenser overpressurization is minimized. Switches are provided to. manually bypass the channels when all TSVs are closed. This Function isolatos tha G ufvalves. 4 2'" / / 64 Si l h3 .Ad .S l M ~ ( 1.a. 1.f4 Main steam Tunnel Ambient a n O!'fe m + W Tennerature-Hieh ~ Q%./ Ambient emperature-High is provided to detect a leak in the RCPB, and provides diversity to the high flow instrumentation. The isolation occurs when a very small leak has occurred. If the small leak is allowed to continue without isolation, offsite dose limits may be reached. However, credit for these instruse is not taken in any transient or accident analysis in th AR, since (continued) BWR/6 STS B 3.3-148 Rev. O, 09/28/92 [M @ Mf0 Primary Containment Isolation Instrumentation B 3.3.6.1
- 2. : . 01 BASES 7
so APPLICA8LE rbl*9 3 !3 h Ii _i d
- 1. e . 1. f4 main neem iunnei ambient aM nH 6 - t @
\ SAFETY ANALYSES, Tannerature-Hiah (continued) i LCO, and ) APPLICABILITY bounding analyses are perfonned for large breaks such as MSL8s. % a a s irJ % ,.e r e3 /- Aantent temperature signals are initiated from thennocouples ( located in the area being monitored. Four channels of(HiiB p ! $t--- T o :! ' - := ten = S Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. Each Function has one temperature element. l Eight thermocov$les p(ovide input to the Main Jteam Tunnel J ,- /Differe ial Temperature-High Funct n. The/ output I fo,c.o/ -ythese is use to date ine the/differen,6f tial ermocouplety'hannel consistof a different I 1 - jg 's jt ature? Eactr c rature instyhment t t recei es inpu 4 from f , permoc pies tMat are ocated the in t and tiet f ; for a i tal of our ava able j ig s cooling syst \\ \ p, the a annels. f/ N The ambient C ciff:n..;iaNeaperature monitoring \ \ Allowable Value is chosen to detect a leak equivalent to 1 s 25 pm. Y Functiorg solate)stheGroup alves. P4 Y g 1. . Manual Initiation P9 The Manual Initiation push button channels introduce signals into the MSL isolation logic that are redundant to the automatic protective instrumentation and p e manual
- 3,3 J isolation capability. There is no specift analysis that takes credit for this Functi safety
_Itjs h
- 48 -/
retained for at: =11 n '- '-- : P di m .i n op the isolation function as required by the MRC in the plant licensing basis. There are four push buttons for the logic, two manual initiation push buttons per trip system. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons. (continued) B 3.3-149 Rev. O, 09/28/92 BWR/6 STS = (IM 9:-/u Q Primary Containment Isolation Instrumentation B 3.3.6.1 . c . :.o./ BASES pQ e/2 APPLICA8LE SAFETY ANALYSES, LCO, and ~ nual Initiation (continued)_ h' Four channels of Manual Initial n Function are available APPLICA81LITY and are reautred to be OPERA 8tj n MUDE5 1, 2, an 3, sinc ( fth'esefrethe ES in whiph tl MSL Is lation a tomatic CI (Functjlonsarer utred to Ae OPERA 8LE. _ ~ s u A 'D ry w e/I g.3. s. /
- 2. Primary Containment solation syg 2.a_Y Reactor Vessel Water Level-Low Low. Level 2 Low RPV water level indicates the capability to cool the fuel may be threatened. The. valves whose penetrations communicate with the primary containment are isolated to limit the release of fission products. Tt:e isolation of the ,
primary containment on Level 2 supports actions to ensure i that offsite dose limits of 10 CFR 100 are not exceeded. ' O.3 6'f The Reactor Vessel Water Level-Low Low, Level 2 Func d s / associated with isolation is implicitly assumed in t analysis as these leakage paths are assumed to be iso d Rh ggg ' postLOCA.f 1 Gis@ Reactor vessel Water Level-Low Low, Level' 2 signals are initiated from 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 (variab e leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low, Level 2 Function are available and are required to be OPERA 8LE to ensure no single instrument failure can preclude the isolation function. The iteactor Vessel Water Level-Low Low, Level 2 Allowable ! Value was chosen to be the same as the ECCS Reactor Vessel l Ester Level-Low Low, Level 2 Allowable Value (LC0 3.3.5.1), , since isolation of these valves is not critical to orderly ' plant shutdown. l 1, '7 , 8 , 9 , l S;ed f(, g / h This Function isolates the Group C". 55. M Uvalves. 2.b4,4.4.2l, . Drywell Pressure-Hiah INS #M/,l High drywell pressure can indi ate a break in the RCPB. The l isolation of some of the on high drywell pressure ) i n/de, d ve /" supports actions to ensure that offsite dose limits of (continued) l l BWR/6 STS B 3.3-150 Rev. O, 09/28/92 QAR 93-MRI} INSERT B150A The isolation of valve Group 9 also includes the actuation of the Standby Gas Treatment System, the Control Room Fresh Air System, and the containment hydrogen analyzers. INSER.T B150B In addition, Function 2.a provides an isolation signal to certain drywell isolation valves. The isolation of the drywell isolation valves, in combination with other accident ' mitigation systems, functions to ensure that steam and water / releases to the drywell are channeled to the suppression pool g to maintain the pressure suppression function of the the primary containment. \ \ \ \ .y, L &e l di?S INSERT 9l/W RIVER BEND B3.3. 8 a n /1/o 2G 150 Q 93-k h _ l Primary Containment Isolation Instrumentation l 8 3.3.6.1 l aASEs I APPLICABLE 2.bdd. 2.'Y~Drywell Pressure-Hiah SAFETY ANALYSES, ~ (continued) PM l LCO, and i 10 CFR 100 are not exceeded. The Drywell Pressure-High APPLICABILITY Function associated with isolation of t rimary containment is implicitly assumed in th AR accident D analysis,as these leakage paths are ass j to be isolated 1 pstit7 postLOCAg I o rf t d High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High per Function are i available and are required to be OPERA 8LE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value was selected to be the same as the ECCS ! Drywell Pressure-High Allowable Value (LC0 3.3.5.1), since i this may be indicative of a LOCA inside primary containsect. Functionh'isolat6 the Group alveC)4-IN# #^ QM y :::t m 2 3l. m ::::::: :: n;; trui. = :: 2.:;. ='A. M ; 5" v h;; 'T;7: tie 7. 2.') .N '2.c. Reactor Yessel Water Level-log Low Low. Level 1 Low RPV we r level indicates t capability to cool the ! fuel may threatened. Shou far, fu damage could resul . RPV water level decr6ase too Therefore, isolation of the prima containment occurs o prevent offsite dose limits i from ing exceeded. Th Reactor Vessel Water,< Level-Low i ! Low ow, Level 1 Functhm is one of the many functions ! a umed to be OPERA 8tE and capable of proviAlng isolation I gnals. The Reactpf Vessel Water Level--Cow Low Low, ; l Level 1 Function associated with isola is implicitly OfR l assumed in the f5'AR analysis as theseakage paths are r , assumed to be solated post LOCA. Reactor ve el water level signal transmit rs that sense the diffe/are initia rence between the pressure due to constant column of water (reference leg) anf the I pressure due to the actual water level (variable lpq) in the vesp41. Four channels of Rnactor Vessel Water Leve'-Low Lod Low, Level 1 Function'are available and arvrequired to 1 b6 OPERABLE to ensure l 1 preclude the isolation'th'at no single instrume,r(t failure can w ' function. / j (continued) BWR/6 STS B 3.3-151 Rev. O, 09/28/92 ' [LAl. 93-ivQ INSERT B151A The i?$lation of valve Group 8 also includes the actuation of the Standby Gas Treatment System, the Control Room Fresh Air System, and the containment hydrogen analyzers. INSERT B151B In addition, Function 2.b provides an isolation signal to certain drywell isolation valves. The isolation of the 3,0.G drywell isolation valves, in combination with other accident 4pr mitigation systems, functions to ensure that steam and water s,- releases to the drywell are channeled to the suppression pool to maintain the pressure suppression function of the the primary containment. , INSERT RIVER BEND B 3.3-152 16 (i M 93 5I) I Primary Containment Isolation Instrumentation ! 8 3.3.6.1 i 8AsES t P12. d Containmentr - _. C(hbw] APPLICA8LE - - - - - SAFETY ANALYSE 5, kaniation-Hioh (continued, LCO, and APPLICA8ILITY i susy6eprovipedtoensur/offsite/selimitfarenot] I texteeded. 7 ! h These Functions isolate the Group [ valves. J. 3. 6, Manual Initiation gg The Manual Initiation push butto channels introduce signals into the primary containment isolation logic that are , redundant to the automatic protective instrumentation and i vide manual isolation capability. There is no specific i 091y safety analysis that takes credit for this Functi_og,._ t is retained forcc :r:M r:- - x ::: n.-1;_.;a siihe isolation function as required by the NRC in the plant -; d licensing basis. There are four push buttons for the logic, two manual initiation push buttons per trip system. There is no l Allowable Value for this Function since the channels are l mechanically actuated based solely on the position of the I push buttons. ' t Four channels of the Manual Initiation F tiottare available and are raau w +a M ~ n MODE 5 1, 2, ) and since t e are the in Primary Con neent Is ation aut ic Functions required a be O LE. _1 - j l
- 3. Reactor Core Isolation Coolina System Isolation 3.a. ' RCIC Steam Line Flow-Hiah i RCIC Steam Line Flow-High Function is provided to detect a I break of the RCIC steam lines and initiates closure of the j steam line isolation valves. If the steam is allowed to '
continue flowing out of the break, the reactor will l depressurize and core uncovery can occur. Therefore, the- 1 isolation is initiated on high flow to prevent or minimize core damage. The isolation action, along with the scram l function of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the ! limits of 10 CFR 50.46. Specific credit for this Function, (continued) ! BWR/6 STS 8 3.3-153 Rev. O, 09/28/92 ) l b P 9:-/ Q ) Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 3.c. RCIC Steam Sunolv Line Pressure-Low (continued) SAFETY ANALYSES, LCO, and (TS) because of the potential for risk due to possible APPLICA8ILITY failure of the instruments preventing RCIC initiations. 1 1 The RCIC Steam Supply Line Pressure-Low signals are initiated from two transmitters that are connected to the system steam line. Two channels of RCIC Steam Supply Line Pressure-Low Functions are available and are required to be OPERA 8LE to ensure that no single instrument failure can - preclude the isolation function. /2' The Allowable Value , selecte k . 1 damage to the system turbine @dtobehighenoughtoprevent l This Function isolates the Group valves. I \ 3.d. RCIC Turbine Exhaust Dianhraan Pressure-Hiah ! \ l s High turbine exhaust diaphragm pressure indicates that the : s , pressure may be high to continue operation of the l associated syst urbine. That is, one of two exhaust i diaphrages has ruptured and pressure is reaching turbine casing pressure limits. This isolation is for equipment i t assumed in any transient or accident ' @ protectionanalysis in thand . These instruments are included in the TS because of t potential for risk due to possible failure ; of the instruments preventing RCIC initiations (Ref. 3). ! The RCIC Turbine Exhaust Diaphragm Pressure-High signals are initiated from four transmitters that are connected to the area between the rupture diaphrages on each system's turbine exhaust line. Four channels of RCIC Turbine Exhaust Diaphragm Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument ,'-. failure can preclude the isolation function. ( 3 ' ' 3,I M i # #3 ( - The Allowable Values are i the systen g turbine @ g nough to prevent damage to h This Function isolates the Group [ valves. (continued) BWR/6 STS B 3.3-155 Rev. O, 09/28/92 {MK 9L +RJ) , Primary Containment Isolation Instrumentation 8 3.3.6.1 BASES <p Ambientegg n m ....H e \ i APPLICABLE , 3.e. [3.A l SAFETY ANALYSES,/ "eeneraturT-Hioh LCO, and / APPLICABILITY (continued)a /,'f leakAmbientedDifferrtIemperatures from the associated system steam piping. Theare provide / isolation occurs when a very small leak has occurred and is l diverse to the high flow instrumentation. If the small leak / is allowed to continue without isolation, offsite dose ' f li may be reached. These Functions are not assumed in ! transient or accident analysis, since bounding / h an ana ses are performed for large breaks such as recirculation or MSL breaks. / c2 d Ambient d_DifferenWTemperature-High signals are ' initiated from themocouples that are appropriately located # to protect the system that is being monitored. Two f instruments monitor each area. Six channels for RHR and 40-]'[ RCIC Ambient Temperature-High Function are available and ** are required to be OPERA 8LE to ensure that no single instrument failure can preclude the isolation function. \ There are two for the RCIC room and four for the RHR area. s ' There ars 12 themocoup1Rd (four for the RCIO room and Deight I 1 \ forrentia Dif tM RHRTemperatu 4rea) that -High pfovide Function 3 T einpdt outputto of thehgre / 5 t se the ouples i used to determine the differential \ emperat . Each annel co 'ists of e' differential ( l \ tempera re inst nt that ceives inputs f l s thermocouples th are loc ed in th inlet an outlet of l ( N the a coolin system f a total f six , area) av lable c%(t,wo nnels.for the RCIC Q andfour/orthe - l \ . The Allowable Values are set low enough to detect a leak l \ equivalent to 25 gpe. ; l Th s Function isolates the Group valves. 3 aJMMain Steam Line Tunnel Ambient Gnd-{Rf4eraritTE2P ~ Tamerature-Hiah ' i Ambiente Diff:mtiaNeaperature-High is provided to detect a Feak in the RCP9 and provides diversity to the high flow instrumentation. The isolation occurs when a very small leak has occurred. If the small leak is allowed to continue without isolation,. offsite limits may be reached. However, credit for these instruments is not taken in any (continued) l B 3.3-156 Rev. O, 09/28/92 BWR/6 STS =, ga 9: wd ) Frimary Containment Isolation Instrumentation 8 3.3.6.1 BASES l APPLICA8LE 3. SAFETY ANALYSES,[Teamerature-Hioh6.NMain Steam Line Tun LCO, and (continued) APPLICABILITY transient or accident analysis in th 1 f R, since bounding l , analyses are performed for large brea such as MSL8s. l / Ambient temperature signals are initiated from thersocouples / / located in the area being monitored. Two channels of Main Steam Tunnel Temperature-High Function are available and / are required to be OPERABLE to ensure that no single f instrument failure can preclude the isolation function. j Each Function has one temperature element. f[ pl _ _ _ _/ Four ereocouples providw in ut tw the Main Steam / Tunne / 'g Diff j -Hig Function. fThe outp'ut of the ntiel the ' Temperature oupljsisu /ed to determine the differential / i l t rat . E h chan el cons 14ts of differential ra ure in rument that Ives i uts f g l the ouples hat a locat in the niet a outle of i ; the pfea coo ing sy en for total f two allable j ) ( channels. f The Allowable Values are chosen to detect a leak equivalent i to 25 gps. \h' This Function isolates the Group alves. '] 3. Main Steam Line Tur.r.;l T- ~:rature Tiggt The Main Steam Line Tunnel Temperature Timer is provided to allow all the other systems that may be leaking in the main steam tunnel (as indicated by the high temperature) to be isolated before RCIC is automatically isolated. This ensures maximum RCIC System o wration by preventing iso 1a'tions due to ks in otter systems. This Function is hnotassumedinan R transient or accident analysis; however, maximizi CIC availability is an important function. Two channels for RCIC Main Steam Line Tunnel Timer Function are available and are required to be OPERA 8LE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are based on maximizing the ! availability of the RCIC System; that is, providing (continued) BWR/6 STS B 3.3-157 Rev. O, 09/28/92 fg < !?D Primary Containmeit Isolation Instrumentation B 3.3.6.1 BASES APPLICA8LE SAFETYANALYSES.h3, , V Main Steam Line Tunnel Temperature Timer (continued) LCO, and APPLICA8ILITY,,, sufficient time to isolate all other potential leakage sources in the main steam tunnel before RCIC is isolated. / This Function isolates the Group valves. / 3/ RCIC/RHR Steam Line Flow-Hiah / RCIC/RHR high steam line flow is provided to detect a break p.'d of the coennon steam line of RCIC and RHR (steam condensing eG mode) systems.and initiates closure of the isolation valves for both If the steam were allowed to continue flowing out of the break, the reactor would depressurize and the core I could uncover. Therefore, the isolation is initiated at I high flow to prevent or minimize core damage. ific ! credit for this Function is not assumed in an accident Pl . I or transient analysis since the bounding anal is perfonned for large breaks such as recirculation and MSL I breaks. However, these instruments prevent the RCIC/RHR steam line break from becoming bounding. \ The RCIC/RHR steam line flow signals are initiated from two \ transmitters that are connected to the steam line. Two channels with one channel in each trip system are available g and required to be OPERA 8LE to ensure that no single instrument failure can preclude the isolation function. The \ Allowable Value is selected to ensure that the trip occurs to prevent fuel damage and maintains the MSL8 as the y boundary event. x Thi Function actuates the Group alves. ,/ Drvwell Pressure-Hiah '}3 High drywell pressure can indicate a break in the RCPB. The RCLC isolation of the turbine exhaust is provided to prevent communication with the drywell when high drywell pressure exists. A potential leakage path exists via the turbine exhaust. The isolation is delayed until the system becomes unavailable for injection (i.e., low steam line pressure). The isolation of the RCIC turbine exhaust rywell Pressure-High is indirectly assumed in th R accident analysis because the turbine exhaust leaka ath is not
- assumed to contribute to offsite doses.
(continued) BWR/6 STS B 3.3-158 Rev. O, 09/28/92 7 , (4R 9' !YctQ Primary Containment Isolation Instrumentation B 3.3.6.1 BASES 1 APPLICABLE SAFETY ANALYSES,t h3. Drywell Pressure-y,{gh (continued) LCO, and i APPLICABILITY f High drywell pressure signals are initiated from pressure f transmitters that sense the pressure in the drywell. Two l ' channels of RCIC Drywell Pressure-High Function are available and are required to be OPERA 8LE to ensure that no l 4Ml eM single instrument failure can preclude the isolation function. ~ds s \ The Allowable Value was selected to be the same as the ECCS \ Drywell Pressure-High Allowable Value (LC0 3.3.5.1), since this is indicative of a LOCA inside primary containment. This Function isciates the Grpup valves. i 3. Manual Initf ah g ~ gg TheManualinitiationpushbuttonchannehintrodu ignad f- 1 into the RCIC System isolation logic that d p redun t to /W6 the automatic protect we instrumentation a e4 rovi nual isolation capaaility. s s retained forex :M r'-"- re is no speciff analysis that takes credit for this Functio sa ty It is y ...d d;.e iir oFThe h ]licensingbasis. ~ ~ isolation function as required by the NRC in the plant There pushbuttogforRC!r__ manual initiation e trip system. There 1s no Allowable Value forthisFunct'onsincethechannelf mechanically I actuated based solely on the position fthepushbuttonE c e s.wc ggr-channels of RCIC Manual Initiation are available and are required to be OPERA 8tE.
- 4. Reactor Water Cleanuo System Isolation t
4.a. Differential Flow-Hioh The high differential flow signal is provided to detect a break in the RWCU System. Th's will detect leaks in the RWCU System when area or differential temperature would not provide detection (i.e., a cold leg break). Should the reactor coolant continue to flow out of the break, offsite dose limits may be exceeded. Therefore, isolation of the RWCU System is initiated when high differential flow is (continued) BWR/6 STS B 3.3-159 Rev. O, 09/28/92 - . _ _ . - _ __ - _ _ ~ _ _ . _ . . . . _ _ . _ _ pt KW] Primary Containment Isolation Instrumentation B 3.3.6.1
- 1. M ,
BASES _@ 's APPLICA8LE L4.V 4.) 4.k d%P SAFETY ANALYSES, 4.c.- 4.d. 4.e. 4.f. 4.o.<% A F Ambient a8 DLffa sn M T=--era ture-Hi ch ~ ~ LCO, and APPLICABILITY g ,) Ambient (TE0iffereouspiesperature-High is provided to detect a leak from the RWCU System. The isolation occurs (continued)/ even when very small leaks have occurred and is diverse to / the hiph differential flow instrumentation for the hot portions of the RWCU System. If the small leak continues / without isolation, offsite dose limits may be reached. Credit for these instrumen s not taken in any transient @ or accident analysis in th AR, since bounding analyses / are performed for large b s such as MSLBs. I , -f Ambient C ifferer.:i d emperature signals are initiated - from temperature elements that are located in the room that
- ,- is being monitored. There are eight thermocouples at r~ provide input to the Area Temperature-High Funct two
' c.*G*f per area). Eight channels are required to be CPE to ensure that no single instrument failure can preclud the h isolation function. ,/ g [s /Therea '16 the uplestdatprov to the s Diffe tial/T ture-Righ Funcyti The output of g thes therip6 coup is used to det4rti . he differential t rature. E channel consipts of a/ differential \ /j ampera the 're in5(rument'that receives inputs from ouples/thatarelocate(inthei lI the a cooTing sy fem for / total o/ nlet afd outlet of \ f eight ' g ch nels ( per rea). (ght ch nelsarp/availabTerequi6dtob/ f !,0PERABLE o ensu that n single nstrumeWt faildre can/ ] g (preclud the is ation action. f \ h The Ambient 6 01 "a ba* M4esperature-High Allowable i Values are set low enough to detect a leak equivalent to as om. Ci m TiG ) These Functions isolate the Group pvalves. h ( F rra'Ylire-Hich Main Steam Line Tunnel Ambient 6 f Eiff m -+ D - ~ ! N Ambient Cf "iffemtidemperature-High is provided to detect a leak in the RCP8 and provides diversity to the high flow instrumentation. The isolation occurs when a very small leak has occurred. If the small leak is allowed to continue without isolation, offsite dose limits may be (continued) l l BWR/6 STS B 3.3-161 Rev. O, 09/28/92 l i @ 90 /'/*h Primary Containment Isolation Instrumentation B 3.3.6.1 ) BASES d Q"41 -2 APPLICABLE 4 Main Steas Line Tunnel Ambientr&.d Oifferei,U4D' SAFETY ANALYSES, TenneraTure-Hich (continued) LCO, and i APPLICA81LITY reached. However, credit for these instruments is not taken in any transient or accident analysis, since bounding analyses are performed for large breaks such as MSLAs. Ambient temperature signals are initiated from thermocouples located in the area bring monitored. Two channels of Main Steam Tunnel Temperature-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. Each Function has one temperature element. Differeptial revfour ,thersocoupIis Temperature-High that provide Function. The output input to%' of these4hermocouples H used to detemine the differential , ) tempe,ratur# Each channel co ists of Vdifferen'tial l / 'paperaturs' instrynent that eives i uts f i / thersocp6ples th,at are 1 in th inlet a outle of l / the ar4a cooli,og system a tota of two ailable , Janpels.f l / The Allowable Values are chosen to detect a leak equivalent to 25 g m / if, ..J N) !h Thi. Function isolates the Group ghalves. I 4 Reactor Vessel Water Level-Low Law. Level 2 1/ 9,M l Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too N far, fuel damage could result. Therefore, isolation of some reactor vessel interfaces occurs to isolate the potential sources of a break. The isolation of the RWCU System on Level 2 supports actions to ensure that fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The Reactor Vessel Water Level-Low Low, Level 2 Function associated with RWCU isolation is not directly assumed in any transient or accident analysis, since bounding analyses. are perforised for large breaks such as MSLBs. Reactor Vessel Water Level-Low Low, Level 2 signals are initiated from 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 (continued) [ l BWR/6 STS 8 3.3-162 Rev. O, 09/28/92 ;l Af m * /4ff) Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 4 actor Vessel Water Level-Low Low. Level 2 SAFETY ANALYSES, (ctnftinued) LCO, and APPLICABILITY level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low, Level 2 Function are available and are required to be OPERA 8LE to ensure that no single instrument failure can preclude the isolation function. The Reactor Vessel Water Level-Low Low, Level 2 Allowable Value was chosen to be the same as the ECCS Reactor Vessel Water Level-Low Low, Level 2 Allowable Value (LCO 3.3.5.1), since the capability to cool the fuel may be threatened. 7if 7 Jlf.) Th Function isolates the G(ro,up valves. y,9 P 4 LC System Initiation v / The isolation of the RWCU System is required when the SLC / System has been initiated to prevent dilution and removal of ,e the boron solution by the RWCU System (Ref. 4). SLC System 4 initiation signals are initiated from the two SLC pump start
- 1. od signals.
x ' There is no Allowable Value associated with this Function s since the channels are mechanically actuated based solely on s the position of the SLC System initiation switch. Two channels (one from each pump) of SLC System Initiation g Function are available and are required to be OPERA 8LE only in MODES 1 and 2, since these are the only MODES where the l' g reactor can be critical, and these MODES are consistent with Anolicability far the SLC System (LCO 3.1.7). M \ - hfw.mx be4Manual isal 4el 4he dae[v7 oJ l o whe3.] Initiation Pl+ s] The Manual Initiation push button channels introduce signals into the RWCU System isolation logic that are redundant to 1 the automatic protective instrumentation a rovide manual i isolation capability. There is no specifi safety l f-g analysis It is retainedthat takes credit r;:r-for y this rrr Function._. me--ity -tFThe ! - Z,; - for@eren l l s isolation function as required by the NRC in plant licensing ; basis. i \ l (continued) ! BWR/6 STS B 3.3-163 Rev. O, 09/28/92 l l G M i n-Q > Primary Containment Isolation Instrumentation B 3.3.6.1
- 4. . <r /
BASES Q<1 APPLICABLE SAFETY ANALYSE:,,/' anual Initiation (continued) LCO, and ! APPLICA81LITY . / There are four push buttons for the logic, two manual / initiation push buttons per trip s There is no Allowable Value for this Function,ystem. since the channels are / mechanically actuated based solely on the position of the / push buttons. I Four channels of the Manual Initiation F / available and are required to be OPERAn8thclion MOD g 1 re , f h ((ana Isol ince snez are the nupy5 in un [a onautomsbcFunctiorWarereq[icred Ineto RWC7 System Ip OPERA 8LE. l / h..-{RHtl 5. L* t a t::1f d System Isolation { 5.a.CM Ambient 6 d Diff:reidN -arature-Hich e ' Ambient 60!!f:r;r iP$emperature-High is provided to detect a leak from the associated system steam piping. The i -1c i s s isolation occurs when a very small leak has occurred and is j diverse to the high flow instrumentation. If the small leak g g\ s is allowed to continue without isolation, offsite dose \ /li may be reached.yeinsFFiinanun5sHPffet assumea inA g @ an transient or accident analysis, since bounding yp s ana yses are performed for large breaks such as MSLBs. tgj'., \ ' d Ambient rni77 m siademperature-High si nels are ' s initiated from thermocouples that are appropr ately located to protect the system that is being monitored. Two instruments monitar ea Four channels for RHR N AmbientCifferent4e(h Area. t:rTemperature-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. , m Eight hermocouples provide input to t'he rea Diff/re[ntial Temperature-High Function. output of [The /en . I th a the suples is nsed to det,e'rmine the pffferential, i rature Each channel consi s of a differential temperatu- instrument that re ives input from thersoc ples that,4re locat in the in t and out t of l . the arp'a coolingfystem for 'a total of our avail e chan 1 (continued) BWR/6 STS B 3.3-164 Rev. O, 09/28/92 $M C-Nal] Primary Containment Isolation Instrumentation 8 3.3.6.1 lcSr BASES ,- Q C , / I j APPLICABLE . J Ambient dindEf4m min a o Temperature-Hiah SAFETY ANALYSES, coninued) LCO, and APPLICABILITY The Allowable Values are set low enough to detect a leak equivalent to 25 gpe. M .ZMJERT GIM'sfk This Function isolates the Group
- 5. Reactor Vessel Water Level-Low. Level 3
~ ;.3 I , Low RPV water level indicates the capability to cool the sf; fuel may be threatened. Should RPV water level decrease too . \ far, fuel damage could result. Therefore, isolation of some reactor vessel interfaces occurs to begia isolating the potential sources of a break. The Reactor Vessel Water Level-Low, Level 3 Function ameisted with RHR Shutdown - Cooling System isolation is rut drectly assumed in any transient or accident analys m since bounding analyses are (/ [-f'ro'"'/ Ig # 005 ad *} 6 r; perfomed for large breaks such as MSLBs. The RHR Shutdown Cooling System isolation on Level 3 supports actions to ensure that the RPV water level does not drop below the top \/Efd#F##qf),a of the active fuei auring a vessel drainaown event + caused by l t a leak (e.g., pipe break or inadvertent valve opening) in I , the RHR Shutdown Cooling System. l /9 Reactor Vessel Water Level-Low, Level 3 signals are initiated from 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 l level (variable leg) in the vessel. Four channels (two - channels per trip system) of the Reactor Vessel Water ^ Y jy [/,/,'o,'f "3,gra r to [ OPERALevel Level-Low, be 8LE to3 ensure Functionthatare no available and are single instrument required. 2 s failure O j 'g can preclude the isolation function. As noted (footnote ( to Table 3.3.6.1-1), only two channels of the Reactor Vessel 3 3,t,d &f CMg Water Level-Low, L*c'. 3 Function are required to be gf ;; We g ,., ,g ,L OPERA 8LE in MODES 4 and 5 (both channels must input into the ' d" ' f,,. same trio system)fprovided the RHR Shutdown Cooling System integrity is maintained. System integrity is maintained 4h' ve "lrse"d f provided the piping is intact and no maintenance is being perfonned that has the potential for draining the reactor i vessel through the system. The Reactor Vessel Water Level-Low, Level 3 Allowable Value was chosen to be the same as the RPS Reactor Vessel Water (continued) BWR/6 STS B 3.3-165 Rev. O, 09/28/92 l G Ad 92 -N @l) INSERT B165A <i e g 9.l The RHR Equipse t Room Ambient ({nd Differer.t2 amperature - jg High Functio . - only required to be OPERABLE in PIODES 1, 2, and 3. In MODES 4 and 5, insufficient pressure and temperature are available to develop a significant steam leak in this piping and significant water leakage is protected by the Reactor Vessel Water Level - Low, Level 3 Function. l l l 1 l i i INSERT RIVER BEND B 3.3-165 10/1/93 CAdd 93 -/v a/ ) INSERT B165A ./ fh.<,./ _ The RHR Equipme t Room Ambient (dM Dif ferent4Memperature - (s jg High Functio , only required to be OPERABLE in MODES 1, 2, and 3. In MODES 4 and 5, insufficient pressure and temperature are available to develop a significant steam leak in this piping and significant water leakage is protected by the Reactor vessel Water Level - Low, Level 3 Function. l l l 1 l i RT RIVER BEND . - 65 10/1/93 (IAt 90-ivPQ I Primary Containment Isolation Instrumentation l 8 3.3.6.1 l l BASES l ~ ' APPLICABLE 5.r. Reactor Vessel Water Level-Low. Level 3 (continued) S Lfe l 4 ',22 SAFETY ANALYSES, LCO, and P Level-Low, Level 3 Allowable Value (LC0 3.3.1.1 since the h2 i, ' APPLICA81LITY capability to cool the fuel may esbedu/threateneM) _ _ . _ - . . - - _ - - - . . . ~ . y fa cs.1 l The Reactor Vessel Water Level-Low, Le g unction 1 t 'fAL /+dmm44, /,o,o only required to be OPERABLE in MODES 3 4, and 5 to pr%s'"^lll7f,, eventt# l ' 4 rep,,,B y, ce this ootential flow nath from lowerino reactor vessel level , O P E u e r ,; N a d E 5 io the top of the fuel.+/Tn ES an z, o er soia ion " j f(e.g. /Reac r 5t am P ssu -H h) da inis rat e ; fcf,J2,g*C M D'3-d conybls sure hat is low th i i ate to r eo ,f,e 3 /,,,, ,he,, e. Lprsvent nex ted ss f in nto vi thi flo na /ca /u M d4/ p/efsv/r hisFunctionisolatestheGroup(valves. or epe Yfo +le WHf ~ - -brNd/7 0 /6M 1 cs/-4> p e m $v;, 5.d. Reactor Steam Dome Pressure-Hiah //d." Y# #"/jv "! The Shutdown Cooling System Reactor Steam Dome f ac /<ms fo ewsure Pressure-High Function is provided to isolate the shutdown df ,,/3,ye /f,,c cooling portion of the RHR System. This interlock is i t provided only for equipment protection to prevent an /m. .de 0+a /0&R/oo intersystem LOCA scenario and credit for the interlock is ,fe ,,, / g ygg not assumed in the accident or transient analysis in the (4FSARD The Reactor Steam Dome-High pressure signals are initiated ' from four transmitters. Four channels of Reactor Steam Dome Pressure-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value was chosen to be Icw enough to protect the system equipment from overpressurization. This function isolates the Group valves. ', g O 5.e. Drywell Pressure-Hiah ' }/ High drywell pressure can indicate a break in the RCPB. The isolation of some of the PCIVs on high drywell pressure ' supports actions to ensure that offsite dose limits of 10 CFR 100 are not exceeded. The Drywell Pressure-High Function associated with isolation RHR Shutdown Cooling System is not modeled in an accident or transient analysis because other lea age paths (e.g., MSIVs) are more limiting. (continued) BWR/6 STS B 3.3-166 Rev. O, 09/28/92 I ?;p 9:- lWI) , l INSERT B166A
- 5. Reactor Vessel Water Level - Low Low Low. Level 1
/ I' ' Low RPV water level indicates that the capability to cool the o e,ro / fuel may be threatened. Should RPV water level decrease too l Therefore, isolation of the -- hp/ far, f'lel damage could result. shutdown cooling portion of the RHR System occurs to prevent i offsite. dose limits from being exceeded. The Reactor Vessel . Water Level-Low Low Low, Level 1 Function is one of the many Functions assumed to be OPERABLE and capable of providing ; isolation signals. The Reactor Vessel Water Level-Low Low l Low, Level 1 Function associated with isolation-is implicitly ; assumed in the USAR analysis since these leakage paths are. ! assumed to be isolated for a DBA. Reactor vessel water level signals are~ initiated from level transmitters that sense the difforence between the pressure < due to a constant column of water (reference leg) and the i pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Inw Low Low, Level 1 Function are available and are required to be l OPERABLE to ensure that no single instrument failure can i preclude the isolation function. ; The Reactor Vessel Water Level- Low Low Iow, Level 1 Allowable value is chosen to be the to same ensure asthat the ECCS the Level 1 Allowable shutdown cooling value (LCO 3.3.5.1) portion of the RHR System isolates on a potential doses from loss of coolant accident (LOCA) to prevent offsite exceeding 10 CFR 100 limits. This Function isolates the Group 10 valves. f INSERT RIVER BEND B 3.3-166 10/1/93 i 141 % - WI) Primary Containment Isolation Instrumentation s B 3.3.6.1 , 13. s. s. BASES gMG 7[ #' # l APPLICA8LE fe, Drywell Pressure-Hioh (continued) SAFETY ANALYSES, LCO, and High drywell pressure signals are initiated from pressure APPLICA81LITY transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High Function are available and are required to be OPERA 8LE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value was selected to be the same as the ECCS Drywell Pressure-High Allowable Value (LC0 3.3.5.1), since 1 this may be indicative of a LOCAinside primary containment. ' (Q U Co.e-a 84] ThisFunctionisolatestheGroupJ/ valves. 4-{ wssacr- BI(. 7 A @ ACTIONS Revi 's Note: Certain C0CompletionTimsIsarebasedon l appro topical reports In order for a 1 consee to use the ines, the licens must justify the lation Ti As i @ ired by the staff afety Evaluation eport (SER) f the ical report. ,f @f A Note has been provided to modify the ACTIONS related to primary containment 41 solation instrumentation channels l [ Section 1.3, Completion Times, specifies that once a d. , , ,ostotS) Condition has been entered, subsequent , suosystems, N adp )/"3// components, or variables expressed in t ondition C4 Jiscovered to be inoperable or not within limits 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 Condition. , However, the Required Actions for inoperable primary I containment' isolation instrumentation channels provide appropriate compensstory measures for separate inoperable channels. As such, a Note has been provided that allows l separate Condition entry for each inoperable primary ) containment j isolation instrumentation channel. { M \ Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours or 24 hours, depending on the Function, has been shown to be (continued) BWR/6 STS B 3.3-167 Rev. O, 09/28/92 h292-/<///} INSERT B167A , , -~l 0 , 'i - 5.W. Manual Initiation 'W ip'H . The Manual Initiation push button channels introduce signals into the RHR Shutdown Cooling System isolation logic that are redundant to the automatic protective instrumentation and provide manual isolation capability. There is no specific .0 f' It USAR safetyfor analysis Mer;11 that r e takes f r .dcredit e.. efor anuthis Function. 0F h e ~ e@ 'f is retained alversiL1 isolation function as required by the NRC in the plant licensing basis. There are four push buttons for the logic, two . manual initiation push buttons per trip system. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons. i Four channels of the Manual Initiation Function are available and are required to be OPERABLE. INSERT . RIVER BEND B 3.3-167 10/1/93 '54 e WO) Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS M (continued) [-ffdfELT BlbnAQ acceptable (Refs. 5 and 6) to permithstoration of any inoperable channel to OPERA 8LE status.& This out of service i time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action 8.1 Bases). If the inoperable channel cannot be restored to OPERA 8LE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable l channel in trip would conservatively compensate for the , inoperability, restore capability to accommodate a single l failure, and allow operation to continue with no further l restrictions. Alternately, if it is not desired to place I the channel in trip in(e.g., as. in result the case in anwhere placing)the inoperable channel trip would isolation , I Condition C must be entered and its Required Action taken. l 9,3,&.1 - 42.s y 1 1 Required Action 8.1 is intended to ensure that appropriate 'N actions are taken if multiple, inoperable, untripped ado.ma h l channels within the same Function result in redundant " # /, /# automatic isolation ca penetration flow paths). (pability being lost for the associated-The MSL isolation Functions are I va/ve f considered to be maintaining isolation capability when sufficient channels are OPEAABLE or in trip such that both trip systems will generate a trip signal from the given i Function on a valid signal. The other isolation Functions are considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip such that one trip ystem will generate a trip signal from the given Funct a val'd signal. This ensures that one of the n the asscc' ated penetration flow path can '53'l.h a mee ve an isolation sional from the given Function. For Functions 1.a.1.b 1.d.1.e, @ l.fAthis would require 0 h r l ' p.1 3 I both trip systems to have one channel 0PERA8tE or in trip For Function 1.c. this would require both trip systems g ('~ 3.W - sg ,PS have one channel, associated with each MSL, OPERABLE tria, For F ctions 2.a. 2.b, N 3.d 4 . N Q [g g4 3 5.gi@5.g, this would requ' ire one trip system to_ enannels, each OPERAS E or in tr . ForFunctionsk3.a.3.b,tg> twi 3.c 3.e, 3.f. 3.g. . , '3.1,13 fr .a. 4.b, 4.c, wou d require one tri@p [ . % . 4.h = *. =.,;.__.i :. s thisystem to have one[ For channel O @fs u3 Rj) W A
- N &&,,,,,,,0 BWR/6 STS B 3.3-168 Rev. O, 09/28/92
Q cm) Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS M (continu ) , fFunctfons3. ~ 5 4 .. 5.a. each Function / 1 consists locations.of hannels t at monitor several different \ Therefore, this would re ; location to be OPERA 8LE or in tripthe (quire one channel channels are not per i s required to be in the same trip system). Th condition d es inclyde the Manual Initiation Functions (Functions 1 ' I 22 r si 'l 2. acc1 3 4 ) y - M , since they are not assumed in any cent or transient analysis. Thus, a total loss of J h I manual initiation ca p . Q ed 5.h Required Action A.1)pability is allowed. for 24 hours (as allowed by g" The Completion Time is intended to allow the operator time to evaluate and repair any dtscovered inoperabilities. The ! Completion Time is acceptable because it minimizes risk I while allowing time for restoration or tripping of channels. 1 M 1 Required Action C.1 directs entry into the appropriate Condition referenced in Table 3.3.6.1-1. The applicable Condition specified in Table 3.3.6.1-1 is Function and M00E or other specified condition dependent and may change as the i l Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A or 8 and the associated Completion Time has expired, Condition C will be entered for ttat channel and provides for transfer to the appropriate subsequent Condition. 0.1 D.2.1. and 0.2.2 If the channel is not restored to 0PERA8LE status or placed in trip within the allowed Completion Time, the plant must be placed in a M00E or other specified condition in which l the LC0 does not apply. This is done by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours (Required Actions D.2.1 and D.2.2). Alternately, the associated MSLs ma and if allowed (i.e., y plant be isolated safety (Required analysis allows Action D.1), operation i i with an MSL isolated), plant operation with the MSL isolated may continue. Isolating the affected MSL accomplishes the i safety function of the inoperable channel. The Completion l (continued) BWR/6 STS B 3.3-169 Rev. O, 09/28/92 i % /wD Primary Containment Isolation Instrumentation 8 3.3.6.1 8ASES ACTIONS D.I. D.2.1. and 0.2.2 (continued) 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.
L1 If the channel is not restored to OPERA 8LE status or placed in trip within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LC0 does not apply. This is done by placing the plant in at least MODE 2 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 2 from full power < conditions ' n an orderly manner and without challenging plant systems. f.d If the channel is not restored to OPERABLE status or placed i in trip within the allowed Completion Time, plant operation may continue if the affected penetration flow path (s) is isolated. Isolating the affected penetration flow path (s) accomplishes the safety function of the inoperable channels. /f Wl _( r~)~ l '- -f For some of the Ambient (S Oiffue.,tilpTemperature t Functions, the affected penetration flow path (s) may be considered isolated by isolating only that portion of the system in the associated room monitored by the inoperable channel. That is, if the RWCU pump room A ambient channel is inoperable, the A pump room area can be isolated while ' allowing continued RWCU operation utilizing the B RWCU pump. ; Alternatively, if it is not desired to isolate the affected penetration flow path (s) (e.g., as in the case where isolating the reactor scram) penetration , Condition flowbe H must path (s) could entered result and its in a Required Actions taken. (continued) BWR/6 STS J 3.3-170 Rev. O, 09/28/92 [LAR 92-MN O Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS I.1 and I.2 (continued) F are provided by declaring the associated SLC subsystem M """ 0""g"f inoperable or isolating the RWCU System. 7 3 0'# by.s fem .s n bo,a St o *** .d 37 ' r(c re a d o r vesse / The Completion Time of 1 hour is acceptable because it ,' , minimizes risk while allowing sufficient time for erson_nel , to isolate the RWCU System. i ] 4 ' cibet /EG*Foof 7 3./ f h,\e.ked ,or " 3 " 9) If the channel is not restored to OP RABLE status or placed in trip within the allowed Complet on Time the <sg55im5iDe_ However, if the prod, e an Ornd]e shutdown 'c:::: :::= cooling
- = ;=:;,function shouldisbe tissedto provide core cooling, needed these Required Actions allow the penetration flow path to 4 C'S y ,y /
remain unisolated provided action is immediatel initiated c9%\ny ad to restore the channel to OPERA 8LE status or t isolate the RHR Shutdown Cool.ing SystenfD .e., = n e: :::: m::: ::::n su6 4veAO S5-i. - _ -::: -----ss m -s se +k 2:::t : tie- <! r : n-n il s d Jr6 OPE BL s t Sy em M iso ted. j LV.t ER T l - K.I. K.2.1. - t .- ~7' 2 ' * / d" If the channel is not restored to OPERA 8LE status or placed in trip within the allowed Completion Time, the associated penetration flow path (s) should be isolated (Required Action O ). Isolating the affected penetration flow 0Cf ,.. ll.I path (s) accomplishes the safety function of the inoperable instrumentation. Alternately, the plant must be placed in a condition in which the LCO does not acolv.1 If applic le, i ~ [ CORE ALTETtA IONS and movement irradiated fuel ass lies must be i diately suspended. Suspension of these 4 activitie shall not preclude completion of movene of a componen to a safe conditio . Also, if applicab , action must be usediately initiat d to suspend OPDRVs minimize the pro ability of a vesse draindown and subse ent i potent al for fission pro etion release. ACT e until OPORVs ar suspended. S must j / tconti " IA) L Seer B n 2. A - (continued) BWR/6 STS B 3.3-172 Rev. O, 09/28/92 (LAK 9.1-MAD' .o INSERT B172A This is done by placing the plant 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 plant conditions from full power conditions in an- ' orderly manner and without challenging plant systems. INSERT B172B or to provide means for control of potential. radioactive releases. This is accomplished by ensuring primary containment is OPERABLE. This may be performed as an. administrative check, by examining logs or other information, to determine. if the components are out of service for maintenance or other reasons. It is not necessary to perform t g,5 61 the Surveillances needed to demonstrate the OPERABILITY of the jgt components. If, however, any required component is . inoperable, then it must be restored to OPERABLE status. In this case, the Surveillances may need to be performed to ! restore the component to OPERABLE status. In addition, at least one-door in each primary containment air lock must be closed. The closed air lock door completes the' boundary for control of potential radioactive. releases. With the . appropriate administrative controls however, the closed air- ! lock door can be opened intermittently for entry and' exit. , This allowance is acceptable due to the need for containment ' access and due to the slow progression of events which may i result from a reactor vessel draindown event. Reactor vessel ! draindown events would not be . expected to result in the- ! immediate release of appreciable fission products to the j containment atmosphere. Actions must continue until all ' requirements of this Condition are satisfied. ., I -i O f f INSERT 0 RIVER BEND B 3.3-172 10/1/ W (MK 93- Nh.) l Primary Containment fsolation Instrumentation B 3.3.6.1 BASES (continued) l SURVEILLANCE Reviewer's Note: Certain Frequencies are based on approved
- REQUIREMENTS topft:a1 reports. In o Frpquencies, the licen/ see der must forjustif a licenheetousethese/
Frequenciep as A '. L- rfquired by the staff SER for the to al report. / _ As noted at the beginning of the SRs, the SRs for each l g // ~ Frimary Containment. Isolation Instrumentation Function are found in the SRs column of Table 3.3.6.1-1. I [ The Surveillances are also modified by a Note to indicate i ,9,3,3,I that When a channel is placed in an inoperable status solely l for perfomance of required Surveillances, entry into O associated Conditions and Required Actions may be delayed , for up to 6 hours provided the associated Function maintains ' trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be l returned to OPERABLE status or the applicable Condition l entered and Required Actions taken. This Note is based on, ' the reliability analysis (Refs. 5 and 6) assumption EEP nq.- J ' p=rs-tw the average time required to perfom channel r Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the ill isolate the penetration flow t path (s) when necessary. nola bo.y valv -~ 7.s.64 Os SR 3.3.6.1.1 Perfomance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A f 7 Sc rMd CHANNEL CHECK is44 comparison of the parameter indicated on one channe' to a similar parameter on other channels. It is MJ P based on the assumption that instrument channels monitoring rq ' the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excmive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly , between each CHANNEL CALIBRATION. ; Agreement criteria are detemined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is OCS outside the m criteria, it may be an ir.dication that the instrument has drifted outside its limit. (continued) BWR/6 STS B 3.3-173 Rev. O, 09/28/92 QAK93-/W$ Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.3 (continued) l REQUIREMENTS The Frequency of 92 days is based on the reliability analysis of References 5 and 6. SR 3.3.6.1.4 and SR 3.3.6.1.5 ! CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary I range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrationg -5 .. ;; ;;; n; 7. ; . . --- - ' n -o A ig. _ : -m ..- + k- ---<- --+ consistent with the plant ' specific setpoint methodolo rMr- : :::i ~. .,. ; w n I ' pl i;rnu .... . . .... . -- - gy. ___ Pa ___ --- 22: cf tr.; uL ht) - g m =5 p CS , ~ ' -if the as found point is not wi n its d - Allowable V , the plant spa e setpoi thodology be revis , as appropriate the hist and all oth pertin infomation i cate a n or the revis . Th iset nt shall be le set consis t with.the p sumption lo the current n1 , specific sefocint methodology.f The Frequency of SR 3.3.6.1.4 Js ased on ~ ass on of L !+ n+ yein eFrac$ ; 4 SR 3.3.6.1.5 is based on the a a: -- . m 1 CIO (ce!m - '--~~ = -- . ssumption - ---ofj r- theJmagnitude of equipment drift in the ;,etpoint analysis.
- 3. 3. 6./
##f SR 3'.3.6.1.6 i s The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the p[ u GPERASILITY of the required isolation logic for a specific Ay~d/ M g' _ channel. The system functional testing performed on PCIVs in LC0 3.6.1.3jverlaps this Surveillance to provide g/u, y, complete testing of the assumed safety function. The 18 month Frequency is based on the need to perfom this 4CO ?. 6 5. 3 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) BWR/6 STS B 3.3-175 Rev. O, 09/28/92 , L .h e 9;7/.MI) l Primary Containment Isolation Instrumentation l ' B 3.3.6.1 ' BASES SURVEILLANCE SR 3.3.6.1.6 (continued) REQUIRENENTS Operating experience has shown these components usually pass ! pg the Surveillance when perfonned at the 18 month Frequency. l ~} C '7&A I l SR 3.3.6.1.7 i This SR ensures that the individual channel response times ' / ' , are less than_ o ual to the maximum values assumed in the I , ident anati The instrument response times must be added to the' losure times to obtain the ISOLATION I I SYSTEN RESPONSE TINE. ISOLATION SYSTEN RESPONSE TINE \ t acceptance criteria are included in Reference 7. b t /the Surveilla e statessthat the radiation (h.:h! s @ * - detector's may be exclu from ISOLATION SYSTEN RESP 0NSEs TINE,tfesting./ This Note is necessary because of the - difficulty of generating an appropriate detector,fnput / dinaland,6ecause f frirtually/ ensure an,the princjplesresof detector instantshoous se ti . ration / Respopse time for radia 4n detectida channel hall measured from detector out or the in ectronic Qomp6nentin, he channeY.put f of thee irst g 43 !$0LATION SYSTEM RESPONSE TINE tests are conducted on an ! 18 month STAGGERED TEST BASIS. c : x- a test Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience that shows that random failures of instrumentation components causing ; serious response time degradation, but not channel failure, are infrequent. REFERENCES h 1. h SectionN 6.3f.' h h 2. h ChapterJ156 Q
- 3. Nf00-31466, " Technical Specification Screening Criteria Application and Risk Assessment,"
November 1987. h 4. h , Section 9.3.5[ h (continued) BWR/6 STS B 3.3-176 Rev. O, 09/28/92 1 gat 9:-t'MQ INS 2RT B176A , i Testing is performed only on channels where the assumed response time does not correspond to the diesel generator (DG) ) 7 S'O'I start time. For channels assumed to respond within the DG l # Si ,_ start time, sufficient margin exists in the 10 second start time when compared to the typical channel response time ' (milliseconds) so as to assure adequate response without a j specific measurement test. I O h i INSERT RIVER BEND B 3.3-176 40/1/ME2._ fjAf wmD Secondary Containment Isolation Instrumentation , 8 3.3.6.2 BASES APPLICA8LE 1. Reactor Vessel Water Level-Low Low. Level 2 SAFETY ANALYSES, (continued) LCO, and - APPLICA81LITY level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low, Level 2 Function are , available and are required to be OPERA 8LE to ensure that no single instrument failure can preclude the isolation function. The Reactor vessel Water Level-Low Low, Level 2 Allowable Value was chosen to be the same as the High Pressure Core Spray (HPCS)/ Reactor Core Isolation Cooling (RCIC) Reactor Vessel Water Level-Low Low, Level 2 Allowable Value (LCD 3.3.5.1, " Emergency Core Coolin System (ECCS) h I Instrumentation," and LC0 3.3.5.2 Cooling (RCIC) System
- eactor Core Isolation
), since this could indicate the capability to cool the fuel is being threatened. The Reactor Vessel Water Level-Low Low, Level 2 Functionf is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting i in significant releases of radioactive steam and gas. In ' MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitationsJ f these MODES: thus, thin Function is not requireak /In daftlon, ther Function 's also; required to be), iOPERABLE duri g CORE ALTERATIONS and operations with a ipotential f drainingthereactorvesselJ0PDRVs)because A the capabi ty of ise'ating potential sources of leakage OP3 m st be p vided to enipfre that offsite Aose limits are no [exc_eeded f core damag6 occurs. f
- 2. Drvuell Pressure-High o.
High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the.SGT System are' initiated in order to minimize the potential of an offsite dose release. The isolation of high drywell pressure ,c supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis. , ,, , , 3 ,2 However, the Drywell Pressure h Function associated with ~,%- fl_-fanalysis. isolation is not assumed in an cident or transient It is retained for t P n f r f: n.1 esM (continued) BWR/6 STS B 3.3-181 Rev. 0, 09/28/92 =_ . &i C: .vtI} Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES 1 APPLICABLE 2. Drywell Pressure-Hich (continued) SAFETY ANALYSES, p- l LCO, and Cdf = :ity of'fheJsecondary containment isolation APPLICABILITY / instrumentation as required by the NRC approved licensing %sses $ ,7*, 7; High drywell pressure signals are initiated from pressure transmitters that Qense the pressure in the drywell. Four A7 channels of Drywell*-High Function are available and are required to be OPERABLE to ensure that no single instrument j failure can preclude the isolation function. l The Allowable Value was chosen to be the same as the ECCS Orywell Pressure-High Function Allowable Value ; (LCO 3.3.5.1) since this is indicative of a loss of coolant j accident. ! The Drywell Pressure-High Function is required to be OPERA 8LE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature , limitations of these MODES. l Salad U F i h .- Rad' ation-Hioh o fc 3 bed % High secondary containment exhaust radiation is a indication of possible gross failure of the fuel cladding. The release may have originated from the primaryocontainment due to a break in the RCPS or the sfuelgW due to a fuel handling accident. When Exhaust Rad' ation-High SjE) ; **\g[ is detected secondary containment isolation and actuation of theM@f,ystem are initiated to it the release of veAh' fission products as assumed in the 5 (Ref.1). u safety analyses g The Exhaust Radiation-High@ signals are initiated from radiation detectors that are located on the ventilation exhaust piping coming from the fuel.t !P :: :=2 rrd = k*36$ . :: rrr : n.; ., a -. ' .-_ 7:::xe::S The signal veMM from each detecter is input to an individual monitor whose trip outputs are assigned to an isolation channel. @ (continued) BWR/6 STS B 3.3-182 Rev. O, 09/28/92 SR93ndj Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE h5 3. 4. Fuel F-(MMM -'- - --m Ventilation ~ r- ' " - '-T Exhaust SAFETY ANALYSES, Radiation-Hiah % (continued) LCO, and _ APPLICABILITY ' channels of Fuel andling Area Ven ation Exhaus , Radiation-Hi High Function an four channels of g Handling Ar Pool Sweep Exha Radiation-High gh Function e available and required to be RA8LE to ensure at no single in nt failure can clude the < iso ion function.J The Allowable Values are chosen to promptly detect gross failure of the fuel cladding. ' The Exhaust Radiation-High 6 FunctionEMD required to be.J0 e pen.m.1 in a m 1 Z, a .1 wnere nn uwreo .... ryf , exist is a p ability @,. pipe b ' thus, t s res ting in nificant leases of adioacti steam a . In M00 4 and 5, e probabi ty and c sequence of , hese eve are low e to the pressu and t ratu -limitat s of these ES; th , these F tions a not Lauf d- in ade i an the F *tian< a r;;2ir d to b l OPERA 8LE during . -- _ - -- - -~- movement of irradiated fuel assemblies in the g( n ; = z ;: because the capability o etecting radiation releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded. S. Manual Initiation : The Manual Initiation push button channels introduce signals into the secondary containment isolation logic that are redundant to the automatic protective instrumentation channels, and provide manual isolation capability. There is no specificu/SAR safety analysis that takes credit for this f _ hg @ i Function. It is retained for the diFer et=d- ry r% h 1 *d = m t, or i.ie secondary containment isolation [',Ss9 nstrumentation as required by the NRC approved licensing . N basis. There are four push buttons for the logic, two manual initiation push buttons per trip system. There is no Allowable Value for this Function since the channels are mechatiically actuated based solely on the position of the push buttons. (continued) , BWR/6 STS B 3.3-183 Rev. O, 09/28/92 @i %-/YW) Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2.4 (continued) REQUIREMENTS ___ ppropr if the ni ana all e 'D--(berevised, " " "' *' * " ' ' *" '- "- '"- shall t set con ent with ssumptions /_setpoe current-1[hnt suscif e.+nat + m. gata;y_f ,,, The Frequency is based upon the assumption of diAikanime caHhrat the magnitude of equipment drift in the setpoint analysis. SR 3.3.6.2.5 e-TheLOGICSYSTEMFUNCTIONALTISTdemonstratesthe M ' OPERABILITYoftherequiredisolationlogicforaspecific)P "4 g) channel. The system functional testing, performed on SCI)fs ,,j'v"" e* W h * \ f and the r ..i-L in LC0 3.6.4.2 M LC0 3.6.4.3, ' ( kyys+eg respective'y overlaps this Surveillance to provide complete ,_; testing of the assumed safety function. 'E co 2. 6.U 4 3 The 18 month Frequency is based on the need to perform this \ tco 3.c.u.6, ad Surveillance under the conditions that apply during a plant \ outage and the potential for an unplanned transient if the l L k c.o 23.d.1 3 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. SR A.3.6.2.6' [ is SR e es that tfie indi 'dualchann$1responset are less than or equ 1 to th maximum v ves assumed in the, ; d acciden analysi The ins nt res nse times must be l /I ' P added the SC losure times to ob in the ISOLATION ! [ SYST RESPONSE - acc tance criteria are . I. TION SYS RESPONSE TINE , l [ ?. 2.7 cluded in eference 5. t er [ 3 C Note to th Surveill ce states hat the radi tion detectors y be exc1 ded free ! TION SYST RESPONSE < \ TIME test g. This te is nec sary becaus of the \- , difficul of gener ting an ap priate det tor input (- ; l , signal nd because the princi es of detec r operati \ l \ virtu ly ensure n instanta eous respons time. Re onse j t time or radiati n detector channels shal be measu d from / ~ _ / (continued) , BWR/6 STS B 3.3-189 Rev. O, 09/28/92 1 hi 92 -/Q Secondary Containment Isolation Instrumentation 8 3.3.6.2 BASES SURVEILLANCE ISR 3.3.6.2.6 , REQUIREMENTS ( (con inued)~[ detector output o the input of the first/electronit j compon t in th channel. ., ( .- .- ISO TION SY EM RESPONS TIME testsA re condu ed on an ("# ' 1 month ST ERED TEST BASIS. g.;/L,5 w. W F equency it g onsisten with the t ical indus y refueli g cycle and s based up plant op sting exper' nee, whi shows t at random ailures of nstrumenta on compon ts causi / . serio response ime degrada on, but n t chann i e, yarej nfrequent currences. _ REFERENCES h 1. AR,SectionN6.3[h h2.hAR,ChapteN157l'@
- 3. MED0-31677-P-A, " Technical Specification Improvement Analysis for OWR Isolation Actuation Instrumentation,'
July 1990. G 4. NEDC-30851-P-A Supplement 2. " Technical Specifications Improvement Analysis for SWR ! solation 4 ,3,0 2 / Instrumentations Common to RPS and ECCS / ,> c 's. Instrumentation," March 1989. - -~ - --rr""S - _ OEc~/.5. - kS REksiud Rglicemah &c nu 'N_ y_- J q;? BWR/6 STS B 3.3 190 Rev. O, 09/28/92 [Adf 9:-/Y& Relief and LLS Instrumentatio B 3.3.6 l BASES SURVEILLANCE 3.3.6. h SR " (continued) REQUIRENENTS aerpont-snati :,. ; it ei s,..;.ie,4 .iin F...wn v C ', lof th6"eFrrent slant enacifte < meant _,1 a-l-Ay # j The Frequency is based upon the assumption of dC3'EiiHP C6 ]mnr-- r:::- r: -::- -- -- - mi the magnitude of equipment drift in the setpoint analysis. SR 3.3.6 #' ' The LOGIC SYSTEN FUNCTIONAL TEST demonstrates the >9 OPERASILITY of the required actuation logic for a specific The system functi ormed for S/RVs d -/ h channel.in LC0 3.4.4f_yra 2.--uC -J _ :reu' g
- r. 1_ testing g;. ,and LC0 overlaps this Surveillance to provide complete testing of the assumed safety function.
The 18 month Frequency is based on the need to perfom this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with tw reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency. ; REFERENCES I 1. i 2. , Sectio 5.2.2[ . Appendix SA. h
- 3. GENE-770-06-1, " Bases for Changes to' Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications,"
February 1991. l BWR/6 STS B 3.3-221 Rev. O, 09/28/92 4 [M " /& - CRFA System Instrumentation B 3.3.7.1 8 3.3 INSTRUMENTATION l B 3.3.7.1 Control Room Fresh Air (CRFA) System Instrumentation BASES \ BACKGROUND The CRFA System is designed to provide a radiologically controlled environment to ensure the habitability of the control room for the safety of control room operators under all plant conditions. Two independent CRFA subsystems tre each capable of fulfilling the stated safety functiori. The instrumentation and controls for the CRFA System automatically initiate action to isolate or pressurize the main control room (MCR) to minimize the consequences of radioactive material in the control room environment. In the event of a l Reactor Vessel Water Level-Low Low, Level 2. Drywell Pressure-High, or Control Room Ventilation Radiation - Monitor signal, the CRFA System is automatically started in the isolation mode. The MCR air is then recirculated , through the charcoal filter, and sufficient outside air is /-3, 2. 7, I drawn in through the normal _ intake to_ keep the MCR slightly i ( sp 3 pressurized with respect tot'r rr-irr --i- ".,. l ( ad v ee,a orcos ( ; yd,- The CRFAlinstrumentation has two trip systems: -one trip i system initiates one CRFA subsystem, while the second trip system initiates the other CRFA subsystem (Ref. 1). Each trip system receives input from the Functions listed above. The Functions are arranged as follows for each trip system. The Reactor Vessel Water Level-Low Low, Level 2 and Drywell Pressure-High are arranged together in a one-out-of-two i taken twice logic. The Contro' Room Ventilation P.adiation Monitors are arranged in a two-out-of-two logic. The channels include electronic equipment (e.g., trip units) , that compares measured input signals with pre-established ! setpoints. When the setpoint is exceeded, the channel ' output relay actuates, which then outputs a CRFA ,iaitiation l signal to the initiation logic. l APPLICA8LE The ability of the CRFA System to maintain the habitability l SAFETY ANALYSES, of the MCR is er itly assumed for certain accidents as l l LCO, and d hcussed in the R safety analyses (Refs. 2 and 3). Pi j l APPLICA8ILITY CRF3bperation e res that the radiation exposure of ; ontrol room personnel, through the duration of any one of
- b. 7./
kY (continued) 1 \ BWR/6 STS B 3.3-222 Rev. O, 09/28/92 I {MK 93-/YPb CRFA System Instrumentation B 3.3.7.1 BASES APPLICABLE the postulated accidents, does not exceed the limits set by SAFETY ANALYSES, GDC 19 of 10 CFR 50, Appendix A. LCO, and APPLICABILITY CRFA, instrumentation satisfies Criterion 3 of the NRC Policy (continued) Statement. - o ,f The OPERA 8ILITY of the CRFA inst ntation is dependent upon the OPERABILITY of the individual instrumentation 1 channel Functions specified in Table 3.3.7.1-1. Each i Function must have a required number of OPERA 8LE channels, l with their setpoints within the specified Allowable Values, l where appropriate. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. The actual setpoint is calibrated consistent with a plicable setpoint methodology assumpti6ns. 49 _. Allowable Values are specified for each CRF Function ' D' specified in the Table. Nominal trip setpoints are - specified in the setpoint calculations. These nominal 1 setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL l CALIBRATIONS. Operation with a trip setpoint that is less : conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetemined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then detemined, accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection : because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for. The specific Applicable Safety Analyses. LCO, and Applicability discussions are listed below on a Function by function basis. (continued) BWR/6 STS B 3.3-223 Rev. O, 09/28/92 4 l4 M' c-Nxt) CAFA System Instrumentation B 3.3.7.1 BASES (continued) ACTIONS ; Reviewer's Note: Certai[LCO Complation Times are based on R approved topical reports. In order for a licensee to use l thesedimes,thelicenseemustjustifftheCompletion. Times 4 LI f as required by the, staff Safety Evaluation Report (SER) for l u the topical repor$. / - _ ) A Note has been provided to modify the ACTIONS related to CRFA3 instrumentation channels. Section 1.3, Completion C, Times, specifies that once a Condition has been entered, l subseauent M, subsystems, components, or variables 1 (c!W,m expressed in the Condition discovered to be inoperable or not within limits 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 Condition. However, the Required Actions for inoperable ^ CRFA, instrumentation channels provide appropriate compensatory measures for separate inoperable char:nels. As such, a Note has been provided that allows separate 1 f , ' Condition entry for each inoperable CRFA instrumentation l aye Te" channel. N es / m..7 L,1 st?/ Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.7.1-1. The applicable Condition specified in the Table is Function dependent. Each time an inoperable channel is discovered, Condition A is entered for that channel and provides for transfer to the- l appropriate subsequent Condition. "~~~
- 3. 3 7, /
B.1 and B.2 #4 # Because of the diversity of sensors available to p*rovide ! initiation signals and the redundancy of the CRFA design, an allowable out of service time of 24 hours has been shown to be acceptable (Refs. 4 and 5) to permit' restoration of any - inoperable channel to OPERA 8LE status. Howe.er,thisoutof) i service time is only acceptable provided the associated / Function is still maintaining CRFAfinitiation capability. A i Function is considered to be maintaining CRFAfinitiation l capability when sufficient channels are OPERA 8LE or in trip, such that one trip system will generate an initiation signal from the given Function on a valid signal. This would (continued) BWR/6 STS B 3.3-226 Rev. O, 09/28/92 1 I -- _. =- -- -. - [AM s?-w& CRFA System Instrumentation B 3.3.7.1 BASES f ACTIONS B.1 and B.2 (continued) /'rIrysicM hfg require one trip systes to have two channels, each OPERABLE or in trip. In this situation (loss of CRFA initiation capability), the 24 hour allowance of Required Action B.2 is not appropriate. If_the Function i et maintaining CRFA ! g 'v[ R initiation capability,D CRFA must be declared inoperable within 1 hour of discovery of loss of CRFA ; initiation capability in both trip systems.- If the inoperable channel cannot be restored to OPERABLE status f within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure,'and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation); Condition E must be entered and its Required Actions taken. 3 3.7,1 -, C.1 and C.2 ' 91 .$yd.,.1 Because of the diversity of sensors available to provide . , initiation signals and the redundancy of the CRFArdesign, an allowable out of service time of 12 hours has been shown to be acceptable (Refs. 4 and 6) to permit restoration of any inoperable channel to OPERABLE status. However, this out of f service time is only acceptable provided the associated / - Function is still maintaining CRFA/ initiation capability. A / Function is considered to be maintaining CRFAfinitiation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate an initiation signal from the given Function on a valid signal. This would require one trip system to have two channels,_each OPERABLE orintrip). In this situation (loss of CRFA* initiation capability , the 12 hour allowance of Required Action C.2 is _ not _ appropriate. If th Function is not mainttining CRFA d( fattfatton capability, D CRFA q a sse_sunt be_ declared _ OC, L, inoperable within I hour of discovery of Loss of CRFA,@_&sysf_us e initiation capability in both trip systems. If the. " inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must j be placed in the tripped condition, per Required Action C.2. 1 Placing the inoperable channel in trip would conservatively compensate for the,inoperability, restore capability to i (continued) BWR/6 STS B 3.3-227 Rev. O, 09/28/92 [M K 93-NPQ CRFA System Instrumentation B 3.3.7.1 BASES ACTIONS C.1 an W (continued) accomodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition E must e entered and its Required Actions taken. D.1 and D.2 h/ Because of the diversity of sensors available to provide initiation signals and the redundancy of the CRFA Tesign, an allowable out of service time of 6 hours is provided to permit restoration of any inciperable channel to OPERA 8t Y status. However, this out of service time is only acceptable provided the associated Function is stil maintaining CRFAfinitiation capability. A Function is - considered to be maintaining CRFAfinitiation capability when ' sufficient channels are OPEAASLE or in trip, such that one trip system will generate an initiation signal from the given Function on a valid signal. This would require one ' trip system to have two channels, each OPERABLE or in tri). ,e In this situation (loss of CRFArinitiation capability), t ie , M. W 6 hour allowance of Required Action D.2 is not appropriate. - _r M If the Function is not_maintjining CRFA' initiation I-- d4 capability?@CR vithin 1 ho6r of> hFaustofbe discovery declared loss of CRFA6 inoperable nitiation i capability in both trip systems. If the inoperable channel
- naumu cannot be restored to OPERABLE status within the allowable -
out of service time, the channel m st be placed in the tripped condition, per Required Action D.2. Placing the inoperable channel in trip perfoms the intended function of the channel (starts the associated CRFA subsystem in the isolationmode). Altemately, if it is not desired to place the channel in trip (e.g., as in the case where it is not desired to start the subsystem), Condition E must be entered , l and its Required Actions taken. , , The 6 hour Completion Time is based on the consideration that this Function provides the primary signal to start the CRFA System, thus ensuring that the design basis of the CRFA L System is set. l (continued) BWR/6 STS B 3.3-228 Rev. O, 09/28/92 !L AA 93.mpi} CRFA System Instrumentation 8 3.3.7.1 BASES t ACTIONS E.1 and E.2 (continued) With any Required Action and associated Completion Time not met, the associated CRFA subsystem must be placed in the isolation mode of operation (Required Action D.1) to ensure that control room personnel will be protected in the event of a Desiga Basis Accident. The method used to place tu , CRFA subsystem in operation must provide for autood cally ' reinitiating the subsystem upon restoration of po:er _* following a loss of oower to the CRFA subsystem (s).j 45 Moled, if the toxic gas protection instrumensation is I concurrent y inoperable, the the CRFA subs stem shall N instead of th isolation mode. ' lplacedinthetoxicgasmod This p ides proper prot ion of the co rol room person 1 if both toxic j Techn a1 Specification *j s instrumentat and radiatio instr on (not, +=+4an requiredmby/i , i 'qconc ntly inoperab1v.fAlternately, if it is not desired l Ws iWtne subsystem', the CRFA subsystem associated with ! inoperable, untripped channels must be declared inoperable ! witiin 1 hour. i The 1 hour Completion Time is intended to allow the operator I time to place the CRFA subsystem in operation. The 1 hour l Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels, 1 or for placing the associated CRFA subsystem in operation. I SURVEILLANCE ' Reviewer's Note: Certain Frequencies are based on approved ) REQUIREMENTS topical reports. In order for a licensee to use these 1 Frequencies, the licensee must justify the Frequencies as piredbythestaf[SERforthetopicalreport. As noted at the beginning of the SRs, the SRs for each CRFA Instrumentation Function are located in the SRs column of Table 3.3.7.1-1. / The Surveillances are also modified by a Notu .. indicate (Iymb. that when a channel is placed in an inoperable status solely v for performance of required Surveillances, entry into Ls N M r upassociated Conditions and Required Actions may be delayed to 6 3 urs, provided the associated Function (2 n/ maintains CRFAlinitiation capability. Upon completion of ( ep/ the Surveillance, or expiration of the 6 hour allowance, the J channel must be returned to OPERA 8LE statu's or the i (continued) i , BWR/6 $TS B 3.3-229 Rev. O, 09/28/92 _ _ , , . _- - - - - -- , , , + - ,--g ,9,.- my -m, .-.m , Q Af 93-Mx& CRFA System Instrumentation B 3.3.7.1 BASES SURVEILLANCE applicable Condition entered and Required Actions taken. REQUIREMENTS This Note is based on the reliability analysis (Refs. 4, 5 (continued) and 6) assumption @ rt 5 t a n '~D the average time required to perfom channel surveillance. That analysis demonstrated that the 6 hour test 99 allowance does not significantly , reduce the probability that the CRFA will initiate when necessary. - 1 q. 7./ -- .Sys C/ SR 3.3.7.1.1 I Perfomance of the CHANNEL CHECK once every 12 hours ensures l that a gross failure of instrumentation has not occurred. A CHAletEL CHECK isa comparison of the indicated parameter for one instrument channel to a s,imilar parameter on other channels. It is based on the assumption that instrument h'e v channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of* excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifyirg the <nstrumentation continues to operate properly between each ; CHANNEL CALIBRATION. l 1 Agreement criteria are detemined by the plant staff based i on a combination of the channel instrument uncertainties, ) including indication and readability. If a channel is i h outside the ager criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that of demonstrates channel dii"ZiDUplEL-r . n=n;failure On is - rare. C -- :::x,m --> ,.. im-nce]e1; (3 f ::_. _ 9 14=4+ " '- 12 hrrn-J The CHAfstEL CHICK <==... supplements less fomal, but more frequent, checks of l channel status during nomal operational use of the displays associated with channels required by the LCO. ) SR 3.3.7.1.2 A CHANNEL FUNCTIONAL TEST is perfomed on each required channel to ensure that the entire channel will perfom the intended function. ~ A q se4pok+a ad,u w d u 7 ) g cn.shMb wl4k k assu& s d k cure C ~,4+ cpui A setpM m+Ldelen . ((continued) BWR/6 STS B 3.3-230 Rev. O, 09/28/92 - m -- ---- - --w w rum- ATTACHMENT 2B ITS - PSTS COMPARISON DOCUMENT REVISION 1 DISCUSSION OF CHANGES e G4s 93*N& DISCUSSION OF CHANGES TO NUREG-1434 TS: 3.3.3.1 - POST ACCIDENT MONITORING INSTRUMENTATION . l BRACKETED ADMINISTRATIVE CHOICE I B.1 Brackets removed and optional wording preferences revised to reflect appropriate plant _speci_fic r_equirements. g,0 _ . _.%e 4a4me I rs3e.eul.n3 9eiesl reparks we.D #eb' mued do 4% suun c.snssMed ,J,4 revms ; PLANT SPECIFIC DIFFERENCE I f* f
- 3' d 8 b cN* a S. O. ___
, { P.1 This comment number i not used for this station. P.2 *The j!:orrect/ plant jpecific eferengd is pr/vided for Leontistency/with revYsions pro sed in Aection 5je. : P.3 The plant specific Type A and Category 1 PAM instruments are identified in accordance with the NUREG Reviewer's Note. P.4 The current required allowed out-of-service time and CHANNEL CALIBRATION frequency for the Hydrogen Analyzers is retained. P.5 The current plant specific number of channels of suppression pool monitoring required is retained. - t CHANGE / IMPROVEMENT TO NUREG STS , C.1 This comment number is not used for this station. i C.2 The Bases are corrected. The reference Specification in the l Administrative Controls Section does not require the Special i Reports to be approved by any specific individual or committee. C.3 Condition G is not applied to reactor vessel ' water level in the I LCO. Therefore, the Bases are revised to match the LCO. ; C.4 These changes are proposed to revise specific terminology to that which is generically preferred for application to the BWR/6 plants. The BWR LCOs do not use the term " train", ; however, " division" is used in several places. , l C.5 This cossent number is not used for this station. RIVER BEND 11 Af , 3 10,5.of^% : l ([AK % tvf f) a RIVER BEND , SECTION 3.4 1 ATTACHMENT 1: CTS-PSTS COMPARISON DOCUMENT l ATTACHMENT 2:ITS-PSTS COMPARISON DOCUMENT (Tan 9.97I-3 A P ATTACHMENT 1 ITS - PSTS COMPARISON DOCUMENT : 1 REVISION 1 SECTION 3.4 REVISED PAGES l 1 A: MARKUP OF CTS 1B: DISCUSSION OF CHANGES 10: NO SIGNIFICANT HAZARDS CONSIDERATIONS 1 ATTACHMENT 1 A CTS - PSTS COMPARISON DOCUMENT REVISION 1 MARKUP OF CTS 1 QM 9:-Ngt) ; I 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM i RECIRCULATION Loops Lco 3.4.l ! /C0 3.4.2.' LCO 3,+.ll LINITING CON 0! TION FOR OPERATION L:3 3A.t TECH The reactor coolant syntam recirculation loons shall be in operation and in .egion I as speciften m figure 3.4.1.1-1 with either,: I b ; a. Two recirevlation loops operating /with limits and setpoints pe?) (Specif" cations z.1.Z. z.z.1, 3.z.1, 3.2.2. 3.3.6. or; ) i
- b. A single loop operating with:
/T. Volumetric 33,000 recirculation loop flow rate less than or equal to gpe, and ' f2. The recirculation loop flow control system in the loop Manual (Position Control) Mode, and '1 - - - 3. dt? THERMAL POWER less than or equal to 705 of RATED THERMAL POWE and (
- 9. Limits and setpoints for si ner snecifications 2.1.2. 2 le recirculation 1.'3.2.1, 3.2.2, and tion loop (3.3.6 Ai APPLICABILITY:
OPERATIONAL CONDITIONS 18 and 2* ' ACTION a. During single loop operation, with volumetric recirculation 1 D j action to reduce flow to less than or equal toLAl33, 1 hour. J During single loop operation, with the recirculation flow control Ab'h f b. #7 systes not in the Loop Manual mode, immediately initiate corrective / action to place the recirculation flow control system in the Loop * ( Manual mode within 1 hour. ' ~
- c. During si i M C[- RATEDTHEkleloopoperation,withTHERMALPOWER!veact POWER immediately initiate correct rester than 705 of reduce THEMAL POWEE to less than or equal to 75 of RATED THERMAL ,
POWER within 1 hour,. L R o 3.4.1 d. Within entry into sinole 'oss operation, verify that Nars - the operating limits (in spectricat' on 3.2.Dhave been appropriately adjusted for single loop operation. g RECElVED ' (aseeSpecialException3.10.4 NOV 301988 ' RIVER BEND - UNIT 1 3/4 4-1 " SOC y + v&O REACTOR COOLANT SYSTEM M LIMITING C0feIT!0N FOR~0PERATION 1 IN.sE/LT COND f ll J/7 - s U N g. Within 12 hours unen antew into sina' a. r-h the setpoints in Soecifications 2.2.: aan amaration. verify that rl3TL_. appropriate limits. 3.2.2 and 3.3.I)are within f. During single loop operation with either THERMAL POWER < 3 of /cc J 4. li " 4 RATED THEltut POWER or recire.ulation loop flo exceeding the limits in Surveillance W irement 4.4.1.1 THUMAL POWER 6r recirculatten loop riew increases." , & suspenaj 4'q ,gr 'With mera one or tuo reactor coolant system recirculation loops in # 200.3*4I en tiaMand rated core total em damaur r - ans mso snan e oO.3 THERMAL POWER CWO @2 h in Reg"on II o Figure 3.4.1.1-1:(greater than the ' 's' t specif ed 7 ' ', 1. fig f.co SM.I Determine tne APRM and LPIDf"* noise levels (Surveill lg-}CogpC, ~ a) At least once per 4 hours, and Ii b) ~ Within 30 minutes after completion of a THERMAL POWER lg increase of at least 55 of RATED THERMAL POWER. ;\ c,,,',l E s . With the APRM or LPRM** neutron flun noise levels greater than ' \ 'CO 3 4'i thme times their established baseline. noise 1 mis, \ '] cmJa O immediately initiate corrective action ta restore tha noise / 1 levels within the required limits witnin 2 hours byl<nereasing ; I icore flow to grosser snan or equal to enz or rated core flow or lbyreduci THERMAL POWER to less than or equal to the limit ,specified n Region II of Figure 3.4.1.1-1.; i g g h. With one or two reacter coolant system recirculation loops in_ \ g '; 1 operationandtotalcoreflowDessthan355ofrate THERMAL POWER (erester than the rigure init --- < f1eeJ1n Region III of 3.4.1.1-1,.tenediately _wdthin 15 e<nutes init< ata carrective i ~M $ acnon to intfgese core flow (ta n =ta, +" me % ta ME of7 m , (raun spec < f<core riswMr reduce.THEDIAL POWER te less than the limit/ ed in IFogien III of Figure'3.4.1.M within 4 neurs. ' 1 "With one restreulation loop not in operation and isolated tial temperature requirements of Surveillance R91 resent 4.4.I.the 1.46 diff and c are - m with respect to Surveillant:e Reevirementnotapplicable,andthepro 4.4.1.1.46 and e. LA **beDetector monitored. levels A and C of one LPIDI string inthecenterofthecores RECElYED I NOV 301988 RIVER BEND - UNIT 1 3/4 4-la Ano doent No. 31 SDC h REACTOR COOLANT $YSTEM $URVEILLANCE REQUIREMENTS = E~MSER.T m) Leo 3 4.2. N Each reactor coolant system rectrevTatio & shall be demonstrated OPERABLE at least once per 18 months by:n loop # #9. Verifying that the control valve fail SR 14 2 2. pressure at the hydraulic control unit, ands "as is" on loss of hy,6raulic
- p. 1 Verifying that the average rate of control valve movement is:
1. Less than or equal to 115 of stroke per second opening, and 2. Less than or equal to 115 of stroke per second closing Lp 4.4.1.3.2 fEstablish a baseline APRM and LPRM" neutron flux he regions for which monitoring is required (Specification 3.4.1.1 ACTION c within 2 hours of entering the region for which monitoring is required unless Q baselining () qfueling outage. has previously been performed in the region since the last t 1 f$ ,a [4.4.1.1.3 'M f once per 12 hours thereafter, verify that: Initially,within1houruponent a. ' THERMAL P WER is less than or equal to 70E of RATED THERMAL h . I b. The recirculation flow control system is in the Loop Manual j. (Position Control) mode, and
- c. ,
The 11.000 valuestric ene. recirculation flow rate is less than or equal to ' SR 34.n.1 U F34.u T I I G 1.13JWith onireastor cooient system recircuistion loop not in go3;. l m POWER oqual to 50E af_r or the recirculation loop flow in the operatine loop increase in THE recirculation loop flow,Iw1 thin la minutes prnF T5 49 following differential temperature requirements are set:HriiER or rectrcula ** 3 # "[. N T 100*Nbetsioen reactor vessel steam space coolant and bottom afiMine coolant, and " Detector levels A and C of one LPRM string per core octant plus detectors A (and C of one LPRM string in the center of core should be monitored. pgGE YLD l RIVER SEND - UNIT 1 3/4 4-2 Amendment No.31 SDC _ .. _ __ . _ . . .. .~ .._ .__ _ _ . _. . _ _ _ _- _ {AR 92-NPD , s./ J7 / 70 60 __ _ Rtoon y til _ 50 ;
- y r
'~ E ; E . ngGON 30 i q .,
- s ...
' m W 5 - 20 : 10 O 15 25 35 45 55 65 . 75 85 CORE FIDY (% RATED) FIGURE 3.4.1.1-1 THERMAL POWER VERSUS CORE FIDW R E C ElV E - NOV 3019?3 3.C i d RIVER BEND - UNIT 1 3/4 4-3 Amendnent.No. 31 REAcf04 00CLANT SYST[M R CrRCutarron t00P rt0w LIMITING CON 0! TION FOR OPERATION (tco3.4.i 140 3.4.1 das:s a Recirculatten loop flow mismatch shall be maintained within: 3'4*h SE of rated recirculation flow with core flow greater than of equal to 705 of rated core flow.
- y. .
105 ofcore rated rated recirculation flow with core flow less than 705 of flow. APPLIF, ABILITY: opera",1on. OPERATIONAL COSITIONS 18 and 28 during two recirculation l ACTION: f With recirculation loop flows different by more than the specified limits, either: Co#D A a. Eastore the recirculation im flows to within the specified lia_f_t5 W th'n 1. hours, of S'l Jt7 'fb. Shutdone one of the recirculation loons fand taka the ACTION required] (Dy *cirication 3.4 1.1. - 7 $URVEILLANCE REQUIRENENTS SA 3.4.1. I N Recirculation loop flow mismatch shall be verified to be within the limits at least once per 24 hours. .LVSERT SR 3.4.1. ) /00TE. basSpecto"TestE-Mien 3.10.47 **"he provisiens of specification 3.0.4 are not Applicable}@ l RECElyEO 1 NOV 301988 I ( SDC i A1 W B M - UNIT 1 3/4 4-5 Amendment No. 31 i n . ___..e _ - _ . _ ._ _ _.- _ - - _ _ [ ' /r AK 9:-NrD 1EACTOR COOLANT SYSTEM ' /CO 2 9 Tr \' ' 3/4.4.5 3PECIFIC ACTIVI'Y LIMITING CONDITION FOR OPERATION 4co M.9 > 332ff>Thespecificactivityoftheprimarycoolantshallcelimitedto:
- a. Less than or equal to 0.2 microcuries per gram DOSE EQUIVALENT l S. 'l I-131, and
-hr-{b. Less ther, oc equel ts 100/E mic a:urie: cer gr% ' APPLICABILITY: F- 1,$ ' OPERATIONAL CONDITIONS ; 4AL, do,d 3 4 ey n, A w ne not u ola k S w
- a. In'0PERATIONAL CONDITIONS 1, 2 or 3 with the specific activit of I
the primary coolant; , I [ gg.r m A as) Greater than 0.2 microcuries per gram 00SE EQUIVALENT I-131 but h i g [cg.At/.d less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT ! I-131 for more than 48 hours during one continuous time interval A or greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, ; f s'f 00^'8 6 be in at least HOT SHUTDOWN, with the main steam line isolation valves closed, within 12 hours. l \ _ l f 's] f f Greater t in 1 /E m'crocupies per/ gram, in lea . H J l / SHUJdOWN wit the ainsfeamlip6isol ion v ves los cJ (4 3' l l wiJhin 2h rs. f y g g hf In GPERATIONAL CONDITIONS 1, 2, 3 with tne spec ~ activity of the primary coolant greater than 0.2 microcuries per gram DOSE EQUIVA- \ b. Adl./ LENT I-131Agr::te- "=a in0/5 mer::m;; 4 c a.r,jserform tne 1 samoling and' analyst s requirements or item 4a or TaDie 4.4.5-1 unti! s h.t ' 4' g'7 jthe specific activity of tne primary coolant is restored to wi nin ' s : D g In OPERATIONAL CONDITION 1 or 2, with: l
- 1. THERMAL POWER changed by more than 15% of RATED THERMAL POWER in one hour", or
,7 2. The off gas level, at tne SJAE, increased by more than 10,000 9 4 ' microcuries oer second in one hour during steady stata coeration at release rates less taan 75,000 microcuries per secono, or i % 4 d./
- 3. The off gas level, at the SJAE, increased by more tnan 15% in one hcur curing steacy state operation at release rates grea:er ,
tnan 75,000 microcuries per second, perform the sampling and analysis requirements of Item 4b of Table 4.4.5-1 until the specific activity of the primary coolant is restore to within its limit. "Jo t sopiicaole during the startup test program. ] p RIVER BEND - UNIT 1 3/4 4-18 ! 1 ffAi' 93-l'di) t REACTOR COOLANT SYSTEM . 3URVEILLANCE RE0VIRE.uENTS 4-4 The soecific ac.ivity of the reactor coolant shall be demonstrated to
- e .itnin.tne limits Oy :erformanca of :ne sampling and analysis orogram of
%Taole 4.4.5-1. I Lf 7. V. f, / 4, 4 ~ Reg. 4J 4/ - _,;9 . R . As at 7 r ? RIVER BEND - UNIT 1 3/4 4-19 IA8tl 4.4.5-1 - x em g PRIMANY C00lANI SPIClflC ACllVITY SAMPLE AND ANALYSIS PROGRAM x m OPERAll0NAL CON 0ll10NS 9 IYPE Of MEASURLMINI SAMPit AND ANALYSIS IN WHICil SAMPLE AND ANALYSIS __ IREQUENCY AND ANALYSIS REQUINti) 8 $ ; ?rmiti i 0,;' Y " '"' V ^' '" Y ' '* " ' ' l ' ' &3 bx w w 2. n.t ' 7 m, L.,' ,2. Isotopic Analysis for DOSE At least once per ays 1 i l EQUIVALENT l-131 Concentration ' (3 iiadischenderl iur i Determinaison At l ea> i. unte pe, G ,,,,, , t ,',3 * - [ Isotopic Analysis for Iodine a) At least once'per 4 hours, If,(2#,3#,47) I wiever the specif ic 8 Reg. At A/ / activity exceeds a iinit, Af ,a R as required by ACTION b. SY ** g$ , g f 6,/ - duo i b) At least one sample, between- 1, 2 8 2 cod 6 hours following the - change in THERMAL POWER or ,4 / . off gas level, as required I by ACTION c. I 1) ~
- 5. Isotopic Analysis of an Off- At least once per 31 days /4/
gas Saanple including Quantitative e ' Measureisesits for at least Xe-133, A3 8 Xe-135 and Kr-88 f .---- _-_ f j 'i D ' /* Sample-to-be-taken ai ter a mistimum-et 2 Er"0 end 20 days ui TGWER OPERAiiGii inave c.eised since reErtd'e __- j k g,t..A,.o: > <--;3 gag 3 g j o, 7 #Until the specific activity of the priseary coolasit system is restored to within its limits. ;O ? E I 2 4Kc7 Act Al i c.t 6I ._ ._.- ~ . _ (EAR 93 '& REACTOR CCOLANT SYSTEM 3/4.4.5 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LCO 3,'/./l 'l 4,W J f LIMITING CONDITION FOR OPERATION L co 3.'t.// Cr. 't.'/D . y ' G The reactor coolant system temperature and pressure)shall be limited in accordance with the liett lines snown on Figure <E' 1 7 (1) curves A and A' for hydrostatic or leak testing; (2) curves 8 and 8' for heatup by non-M go iInuclear means, cooldown following a nuclear shutdown and low power PHYSICS I TESTS; and (3) curves C and C' for operations with a critical core other than low power PHYSICS TESTS, with: .sR 3.'gfI.I a. A maximum heatup of 100*F in any one hour period, I A b. A maximum cooldown of 100*F in any one hour period,
- c. A maximum temperature change of 10*F in any one hour period during
@ inservice hydrostatic and leak testing operations above the heatup .L and cooldown limit curves, and # M 'l p d. The reactor vessel flange and head flange temperature greater than or equal to 70*F when reactor vessel head bolting studs are under 3 A '3 'l'#f' 7 tension. APPLICABILITY: At all times. _IA/JER T ~Gwo .4/c . ACTION: )I (Co 2 A wore & With any of the above limits exceeded, restore the temperature and/or cressure to dthin the limits (Tthin 30 minuteh perform an engineering evaluation to f,p
- etermine tne effects of the out of-limit condition on the structural integrity of the reactor coolant system; and determine that the reactor c:olant sys a remains acceptacle for continued operations. Otherwise, be in at least HOT CMD 0 //5eUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours.
3URVEILLANCE REOUIREMENTS 1 sR 3A.!/./ GM. .O Curi g system n heatuo, :coicown sna inservice leak anc hycrostatic/I tR 2.v./^ / i ' testing operations, the reactor coolant system temoerature and pressure snall # 78 l M j :e determinea, at ! east onca car 30 minutes, to be within the above recuirec I r9 heatuo and coolcown limits and to the ignt of the ilmit iines af , 1 Figure 3.4.6.1-1 curves A and A', B anc B', or C and C', as applicable. RIVER 3END - UNIT 1 3/4 4-21 l _ . _. , _ _. __ -- .l SM 938 REACTOR COOLANT SYSTEM 1 7,4 SURVEILLANCE REQUIREMENTS (Continued) ) .c 2 .7. V.4 2 \ '54.E a The reactor coolant system temperature and pressure sha11de deter-S f mined to be to the right of the criticality limit line of Figure 6 g g 7,'ul / curves C and C' within 15 minutes orier to the withdrawal of control rods to ring the reactor to criticality and at least once per 30 minutes during system estup. r4.4.6.1.3 The reactor vessel material surveillance specimens /shall be removed ' and examined, to determine changes in reactor pressure vessel l material proper-OA2 ' qatri ties as' 'requiredy " ' ~ 10 CFR 50, Appendix H 6n accordance with the schedu the curves of / Figure 3.4.6.1-1.fThe results of these examinations shall be ~ .s R 3 4.!!. s yu '3.N -f ('Averifled A,S. RtoThe actor vessel flange and head flange temperature shall be 39 be greater than or equal to 70*F: a. In OPERATIONAL CONDITION 4 when reactor coolant system temperature is: .5 2 3. Y, //. 7 5 100*F, at least once per 12 hours'. gg g, y,g, y / 5 80*F, at least once per 30 minutes. # c'%ithin 30 minutes orior to andlat east once per 30 minutes during ['b. tensioning of,the reactor vessel head bolting studs. ,y zusar a uiu n,a \ j .S 2 '3.y.//.7 A/OTE e RIVER BEND - UNIT 1 3/4 4-22 - 1 ~ Ge? *?3-MRI) *l . 1600 '_ 3,1' j an AA'S 8'C C* 1sCO . 7 .r . e s '; $'2;; ll ,'j l E \ ! E IL } ll l' l @1C:0 l I -l I! I ll.' a y , m / , . 4.8.C - ce=E eELTuNE : # f',/ 8 l AFTER ASSUMED 111 *F ,5_ SCO (,, , l - , SHIFT FROM AN INITIAL wEto =T,.,OF -so r y a ! : 3 6CO , [ A - SYSTEM MYOROTEST UMIT g wiTH Fu(L IN VESSEL i 8 - NON-NUCLEAR MEATING M 3 UwT ? 7 P C - NUCLEAR (CORE CRITICAL) p g g g uu? - ~ ~ 3 400 m - VESSEL DISCONTINUITr $ st! 854 ,- UMITS A --- dORE SELTUNE WITH 1117 SHIFT 8 200 , N [g CURVES A*.e' C' ARE VAuo 70Y FOR 8 EFW OF OPERATION CURVES A.B.C ARE VAu0 .. FOR 2 EFW 0F QRtRATt0N O - 100 200 300 400 500 600 RECEIVED MNWUM REACTOR VESSEL WETAL TEMPERATURE (V) AUG 0 91990 2,y,//-/ r1 cure .s - - SDC MINIMUM TEMPERATURE REQUIRED vs PRESSURE RIVER BENO - UNIT 1 3/4 4-23 AMEN 0 MENT NO. 45 e a J wn.a m -.i _m. - an ._ - ATTACHMENT 1B CTS - PSTS COMPARISON DOCUMENT REVISION 1 DISCUSSION OF CHANGES 1 9 i (Lai< W'& DISCUSSION OF CHANGES CTS: 3.4.1.1 - RECIRCULATION LOOPS ADMINISTRATIVE A.1 The format of the proposed Technical Specifications does not include providing " cross references." Each LCO stands alone and adequately prescribes the use of its requirements without such references. Therefore the existing reference to the other specifications serves no functional purpose, and its removal is e 9,J purely an administrative difference in presentation. The existing / Actiods to k restpre. . .within ,(2 or 4) hours" is sI7 A.2 be f , ,/ f y proposed to / revised to/ " initiate action to . . . Imme'diate'ly" fbr operatiori in restricted power / flow Region ' restore III. The xist4hg requirement would appear to provide a time per du ing which power and flow requirements could exceed / th imi , e dn if capable' of being returned to within limit;s. Alsf o , i th param' ters j'are ip'capabl,5 of b'eing y'estored/ to e within the imits/withiri 4 hodrs, th'e existing (ction uld /ppea to 7esult in the/requi ment or an LER. he int to /the cti is b lieved to b more ppro iatel prese ed in( / pro ose Requ ed AcIion C 1. } is terpr ation of the j in,ent is upport4d b the/ BWR Stan ard chni' cal % l Spec icatio , NUREgl-1434 AsaI/enhacedpr entat n of the f (exi ing aditiinistr tent, /the , roposed ive. f ~ change i de d tb be" A.3 The format of the proposed Technical Specifications does not include providing " cross references." LCO 3.0.7 adequately prescribes the use of the Special Operations LCOs without such references. Therefore the existing reference to the Special Test Exception (s) serves no functional purpose, and its removal is purely an administrative difference in presentation. A.4 A new LCO, including appropriate ACTIONS, is proposed to clarify the inte.nt for OPERABILITY of the recirculation system flow control valves. These valves are currently included in the Recirculation Loops specification by required surveillances. The proposed change would provide only ' additional clarification of the current requirements, and is therefore considered an administrative change. A.5 Thermal stresses on vessel components are dependent on the i temperature difference between the idle loop coolant and the RPV coolant. Proposed SR 3.4.11.9 ensures the temperature \ difference between the idle loop and the RPV coolant is acceptable. A requirement to monitor the temperature difference between an idle loop and an operating loop is unnecessary and can be deleted as it is redundant to the loop-N to-coolant requirement of SR 3.4.11.9. However, the loop-to-coolant temperature check may use the operating loop temperature as representative of " coolant temperature." A.6 1his comment number is not used for this station RIVER BEND 1 10/1/93 [ EAR 93-/WI) DISCUSSION OF CHANGES CTS: 3.4.1.3 - RECIRCULATION LOOP FLOW ADMINISTRATIVE A.1 The format of the proposed Technical Specifications does not include providing " cross references." LCO 3.4.1 adequately prescribes the necessary conditions for compliance without such references. Therefore the existing reference to "take- the ACTION required by Specification 3.4.1.1**" serves no functional purpose, and its removal is purely an administrative difference in presentation. A.2 The format of the proposed Technical Specifications does not include providing " cross references." LCO 3.0.7 adequately prescribes the use of the Special Operations LCOs without such references. Therefore the existing reference to the Special Test Exception (s) serves no functional purpose, and its removal is purely an administrative difference in presentation. A.3 The revised presentation of actions (based on the BWR Standard Technical Specifications, NUREG-1434) does not propose to explicitly detail options to " restore. . .to OPERABLE status. " This action is always an option, and is implied in all Conditions. Omitting this action is purely editorial. A.4 This comment number is not used for this station. RELOCATED SPECIFICATIONS None in this section. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 This comment number is not used for this station. TECHNICAL CHANGE - LESS RESTRICTIVE " Gene g,,,,,,f ,,,,, g,, ,, ,,,f k h, 4kl, .f ' LA.1 TN Re red Action td "shutdgwn" one/ recir latio loop.is' reloca ed t procedu es. Sh down f the op is ot al)/ays ; lnece ary, but be p ferred '9 under some ondi ons./ Rel ati the m od to et the requir ments o pro edurps @i p vide addit} nal fle bility when ondit ns a' ow. YndJ .mainta ns the'turrent ction a an op on LA.2 This comment number is not used for this station. RIVER BEND 6 10/1/93 i (M R 93-NkD DISCUSSION OF CHANGES CTS: 3.4.5 - SPECIFIC ACTIVITY ADMINISTRATIVE A.1 Increased sampling is already required when the LCO limit is exceeded (Required Actions A.1 and B.1). Therefore~, no additional surveillance requirements are necessary as an action. 3 A.2 The one time exception for the startup test program is no longer needed and can be omitted. A.3 This comment deleted from LAR 93-14 based on deletion of E-bar requirements per BWR-12 C1. i RELOCATED SPECIFICATIONS None in this section. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 This proposed change modifies Item 2 of Table 4.4.5-1 to change the frequency for isotopic analysis for dose equivalent I-131 ,,,a concentration from at least once per 31 days to at least once *
- g. per seven days.
Modification of Table 4 . 4 . 5 - 1, Item 2, to require isotopic analysis for dose equivalent I-131 concentration at least once , per seven days, as opposed to the current requirement of at least once per 31 days, provides a compensatory tasasure for ensuring that even with deletion of the requirement that gross specific activity remain less thail or equs1 to 100/E-bar pCi/ gram, offsite doses will remain within a small fraction of the limits of 10 CFR 100. TECHNICAL CHANGE - LESS RESTRICTIVE " Generic" LA.1 The offgas isotopic analysis for xenon and krypton are not , direct measurements related to the LCO limits. They are used to routinely monitor and trend coolant activity and applicable ; to plant specific controls and administrative limits only. ; Therefore, this surveillance has been relocated to plant , specific administrative controls. l i RIVER BEND 18 6/16/94 , (LM 93 l'MQ DISCUSSION OF CHANGES CTS: 3.4.5 - SPECIFIC ACTIVITY IECHNICAL CHANGE - LESS RESTRICTIVE (continued) " Specific" L.1 The Applicability is limited to those conditions which represent a potential for release of significant quantities of radioactive coolant to the environment. Mode 4 is omitted since the reactor is not pressurized and the potential for leakage is significantly reduced. In Modes 2 and 3, with the main steam lines isolated, no escape path exists for significant releases and requirements for limiting the specific activity are not required. The Required Actions are also modified to reflect the new Applicability, and an option for exiting the applicable Modes is provided for cases where isolation is not desired. L.2 The requirement to conduct isotopic analysis for iodine when the gross specific activity limits are exceeded is deleted. Unless conditions exist that indicate that iodine is beyond limits, the increased frequency for analysis of iodine is unnecessary in that it provides no useful information with regard to the noncompliance. L.3 This proposed change deletes TS LCO 3.4.5.b, associated Actions and Surveillance Requirements which requires gross specific activity for non-iodines in the reactor coolant to be limited to less than or equal to 100/E-bar pCi/ gram. This proposed change also deletes Item 1 of Table 4.4.5-1 requiring gross beta and gamma activity detennination at least once per 72 hours. The Bases for TS 3.4.5 state that the intent of the requirement to limit the specific activity of the reactor coolant is to ensure that whole body and thyroid doses at che site boundary would not exceed a small fraction of the limits stated in 10 CFR 100 (i.e., 10% of 25 rem and 300 rem, respectively) in the event of a main steam line failure outside containment. To ensure that offsite thyroid doses do not exceed 30 rem, reactor coolant dose equivalent iodine-131 (DEI) is limited to less than or equal to 0.2 pCi/ gram. Likewise, reactor coolant gross specific activity is limited to less than or equal to 100/E-bar pCi/ gram to ensure that whole body doses to not exceed 2.5 rem. LCO 3.11.2.7 (ITS LCO 3.7.4) associated with radioactive effluents requires that the gross gamma radioactivity rate of the noble gases Xe-133, Xe-135, Xe-138, Kr-85m, Kr-87, and Kr-88 measured prior to the holdup pipe be limited to less than or equal to 290 millicuries /second. The Bases for LCO 3.11.2.7 state that restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total-body exposure to an individual at the exclusion RIVER BEND 19 6/16/94 9:-M/1) DISCUSSION OF CHANGES CTS: 3.4.5 - SPECIFIC ACTIVITY TECHNICAL CHANGE - LESS RESTRICTIVE (continued) area boundary will not exceed a small fraction of the limits of 10 CFR 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. The offgas treatment system, as required by LCO 3.11.2.7 (ITS LCO 3.7.4), provides reasonable assurance the reactor coolant ,p' a gross specific activity is maintained at a sufficiently low level to preclude offsite doses from exceeding a small fraction of the limits of 10 CFR 100 in the event of a main steam line failure. Therefore, LCO 3.4. 5.b is redundant and. places an unnecessary burden on the licensee without a commensurate increase in the margin of safety. Elimination of LCO 3.4.5.b will allow plant personnel to focus attention on efficient, safe operation of the plant without the unnecessary distraction of the redundant surveillance requirement. Additional assurance that the offsite doses will not exceed a small fraction of the 10 CFR 100 limits is provided by increasing the frequency of sampling and analysis of the reactor coolant for DEI from at least once per 31 days to at least once per seven days. i Since (1) the reactor coolant limit on DEI adequately assures that offsite doses will not exceed small fractions of the limits of 10 CFR 100 in the event of a main steam line failure outside containment and (2) gross gamma radioactivity rate of the noble gases measured prior to the holdup pipe is limited by LCO 3.11.2.7 (ITS LCO 3.4.7) to a value that provides reasonable assurance the reactor coolant gross specific activity is maintained at a sufficiently low level to preclude offsite doses from exceeding a small fraction of the limits of i 10 CFR 100, the requirements associated with LCO 3.4.5.b are unnecessary. L.4 A Note is added to the Required Actions for Condition A to indicate that LCO 3.0.4 is not applicable. Entry into the Applicable Modes should not be restricted since the most likely response to the condition is restoration of compliance within the allowed 48 hours. Further, since the LCO limits assure the dose due to a LOCA would be a small fraction of the 10 CFR 100 limit, operation during the allowed time frame would not represent a significant impact to the health and safety of the public. RIVER BEND 20 6/16/94 DA 92-niG l DISCUSSION OF CHANGES j CTS: 3.4.6.1 - RCS PRESSURE / TEMPERATURE LIMITS ADMINISTRATIVE (continued) A.5 A Note is provided to clarify the current intent of allowing entry into the applicable MODES without having performed this surveillance requirement. Since this requirement is only performed during the specified conditions, this change is consistent with current application and is considered administrative. A.6 Changes to this sechicn are being concurrently proposed in a Technical Specification amendment for relocation of component lists. The No Significant Hazards Consideration statement made in the referenced letter is still applicable to these changes. As such this change is considered administrative in this submittal. 1 RELOCATED SPECIFICATIONS None irt this section. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 A specific completion time for the engineering evaluation is proposed. The proposed time of 72 hours is considered reasonable for operation in Modes 1, 2 and 3 because the limits represent controls on long term vessel fatigue and usage factors. In Modes 4 and 5, the proposed frequency would prevent entry in the operating modes which is consistent with the current LCO 3.0.4. Additionally, exceeding the P/T limits for less than 30 minutes is not expected to present an immediate threat to the RCS integrity. TECHNICAL CHANGE - LESS RESTRICTIVE "Generi , s' LA.1 specific limit b or reactor coolant system pressure and ,. emperaturehavebeenrelocatedtoaplantspecificcontrolleg gf documentE^ g - rr---"* Tomnerature Limits Rennve %g . The design features and system operational limits are also h7 described in the USAR. ng to th PTLR ill be ontro)J.e - 9 the ov ^ lous e ro sed PT con ols in hapttpf 5 f he T ni 1S ific io . RIVER BEND 21 10/1/93 W - . . - .m r A_a A. ma _u ATTACHMENT 1C CTS - PSTS COMPARISON DOCUMENT REVISION 1 NO SIGNIFICANT HAZARDS CONSIDERATIONS P (AM 93-WID NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.4.5 - SPECIFIC ACTIVITY _L3" CHANGE Entergy Operations, Inc. (EOI) has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards.
- 1. Does the change involve a significant increase in the
'probabilicy or consequences of an accident previously , evaluated? The proposed change deletes Technical Specification (TS) g Limiting Condition of Operation (LCO) 3.4.5.b which requires 'd that the reactor coolant gross specific activity remain less y@ than or equal to 100/E-bar pCi/ gram and the requirement to determine E-bar at least once per six months, Table 4.4.5-1, Item 3. The proposed change also deletes Action a.2 associated with LCO 3.4.5.b that requires the plant to be in hot shutdown with the main steam isolation valves closed within 12 hours after the reactor coolant gross specific activity exceeds , 100/E-bar pCi/ gram. Additionally, the proposed change involves some administrative changes to support the deletion of LCO 3.4.5.b. BWR operatig experience has shown that as fuel leakage increases, dose equivalent iodine-131 (DEI) approaches the TS limit much more rapidly than does the gross spacific activity. The BWR design utilizes main condenser air ejectors to remove non-condensable gases from the reactor coolant. The non- : condensable gases are then sampled, monitored, and processed by the of fgas system prior to release to the environment. The l offgas pretreatment sample provides a more representative i sample of the noble gases that would be released in the event 1 of a main steam line failure outside containment than does the reactor coolant sample currently being taken from the reactor l recirculation system. The of fgas pretreatment monitor includes a setpoint which responds to release rates above a specified level which is established to ensure that untreated releases would not result in a whole body dose that exceeds a small fraction of the limits of 10 CFR 100. The sample point on the reactor recirculation system currently being used to collect information regarding gross specific activity will continue to be available for use in the event of a main steam line failure upstream of the offgas treatment system. I I The Bases for TS 3.4.5 state that the intent of the requirement to limit specific activity in the reactor coolant is to ensure that the whole body and thyroid doses at the site boundary will , not exceed a small fraction of the limits specified in 10 CFR 100 (i.e., 10 percent of 25 rem and 300 rem, respectively) in RIVER BEND 13 6/16/94 e 43 -> vx t) NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.4.5 - SPECIFIC ACTIVITY the event of a main steam line failure outside containment. To ensure that offsite thyroid doses do not exceed 30 rem, reactor coolant DEI is limited to less than or equal to 0.2 pCi/ gram. Likewise, reactor coolant gross specific activity is limited to less than or equal to 100/E-bar pCi/ gram to ensure that of fsite whole body doses do not exceed 2.5 rem. Reactor coolant gross specific activity is not an initiator of any accident evaluated in the USAR and therefore, deletion of LCO 3.4.5.b which limits reactor coolant gross specific activity to a value less than or T equal to 100/E-bar pCi/ gram will not result in an increase in #5 the probability of an accident previously evaluated in the USAR. LCO 3.11.2.7 (ITS LCO 3.7.4) associated with radioactive effluents requires that the gross gamma radioactivity rate of the noble gases Xe-133, Xe-135, Xe-138, Kr-85m, Kr-87, and Kr-88 measured by the main condenser offgas system pretreatment monitor be limited to less than or equal to 290 millicuries /second. The Bases for LCO 3.11.2.7 state that restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of the 10 CFR 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. The offgas treatment system, as required by LCO 3.11.2.7 (ITS LCO 3.7.4), provides reasonable assurance the reactor coolant gross specific activity is maintained at a sufficiently low level to preclude offsite doses from exceeding a small fraction of the limits of 10 CFR 100 in the event of a main steam line failure. Additional assurance that the offsite doses will not exceed a small fraction of the 10 CFR 100 limits is provided by increasing the frequency of sampling and analysis of the reactor coolant for DEI from at least once per 31 days to at least once per seven days. Since the proposed change will ensure that the offsite doses resulting from a main steam line failure will continue to be limited to a small fraction of the 10 CFR limits the proposed change will not invcive a significant increase in the consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical modification to the plant or to plant operation. The reactor coolant gross specific activity is a parameter that is monitored to prevent offsite doses from exceeding a small fraction of the 10 CFR 100 limits and support calculation of offsite doses in the event of a main steam line failure outside containment. As such, the reactor coolant specific activity is utilized to mitigate the RIVER BEND 14 6/16/94 h 93-/yS] NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.4.5 - SPECIFIC ACTIVITY radiological consequences of a main steam line failure and is not considered to be an initiator for any accident. Additionally, the offgas treatment system will provide an equal or better means for monitoring the reactor coolant gross specific activity than would the reactor recirculation system currently being used for this purpose. In the event of a main /' %Y G' steam line break upstream of the condenser that would prevent use of the offgas treatment system to monitor reactor coolant w gross specific activity, the existing sample point on the reactor recirculation system would continue to be available. Accordingly, deletion of the requirement to limit reactor coolant gross specific activity will not create the possibility of a new or dif ferent kind of accident from any previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The Bases for TS 3.4.5 state that the intent of the requirement to limit specific activity in the reactor coolant is to ensure that the whole body and thyroid doses at the site boundary will not exceed a small fraction of the limits specified in 10 CFR 100 (i.e., 10 percent of 25 rem and 300 rem, respectively) in the event of a main steam line failure outside containment. As stated above, LCO 3.11.2.7 associated with radioactive effluents requires that the gross gamma radioactivity rate of the noble gases Xe-133, Xe-135, Xe-138, Kr-85m, Kr-87, and Kr-88 measured by the main condenser offgas system pretreatment monitor be limited to less than or equal to 290 millicuries /second. The Bases for LCO 3.11.2.7 state that restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total-body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of the 10 CFR 100 in the event this effluent is inadvertently discharged without treatment directly to the environment. The offgas treatment system, as required by LCO 3.11.2.7 (ITS LCO 3.7.4), provides reasonable assurance the reactor coolant gross specific activity is maintained at a level sufficiently low level to preclude offsite doses from exceeding a small fraction of the limits of 10 CFR 100 in the event of a main steam line failure. Therefore, LCO 3.4.5.b is redundant and places an unnecessary burden on the licensee without a commensurate increase in the margin of safety. Elimination of LCO 3.4.5.b will allow plant personnel to focus attention on efficient, safe operation of the plant without the distraction of an unnecessary surveillance requirement. Accordingly, the proposed change enhances operation of the plant without reducing the margin of safety associated with a main steam line RIVER BEND 15 6/16/94 [4As' 9 3-NKI] NO SIGNIFICAN" H4ZARDS CONSIDERATIONS CTS: 3.4.5 - JPECIFIC ACTIVITY , t 7-- failure outside of containment (i.e., offsite doses remain a [ c,4 small fraction of the 10 CFR 100 limits). ' Jh Additional assurance that the offsite doses will not exceed a small fraction of the 10 CFR 100 limits is provided by increasing the frequency of sampling and analysis of the reactor coolant for DEI from at least once per 31 days to at least once per seven days. Therefore, the proposed change does not result in a significant reduction in a margin of safety. , t c 3 l s 9 RIVER BEND 16 6/16/94 ~ @sa-ns ATTACHMENT 2 ITS - PSTS COMPARISON DOCUMENT REVISION 1 , SECTION 3.4 REVISED PAGES 2A: MARKUP OF ITS 2B: DISCUSSION OF CHANGES 1 Q M 93 nap l 1 l l l l l i ATTACHMENT 2A l 2 ITS - PSTS l l COMPARISON DOCUMENT l REVISION 1 MARKUP OF ITS 9 0 t c _, , - - . , . . . - - - - - (Istr' @'4]) R; circulation Lorps Operating {q [ 41 i 3.4 REACTOR COOLANT SYSTEM (RCS) 7gepa/M 3* 2 e' ,1; n er smpe 3.4.1 Recirculation Loops Operating [, 4../gg*[, y LC0 3.4.1 CA.operatio Two recirculation rdepe loons T,N M --t:r.:d d 3ha11 [ l'h be -(~ M ,, , A r e - ' 7 __ f, k MS2/0~@~83 __QE 6 6 / c,n ' fshall --h U ' (3 s m . , s, /ca;9J (rfone recirculatica looKgip be in operatio # rrr m-F , co cit' ~ Tollowin limits are App .ies when Ahe associatec L.w is %M/o G,h. 7 app 11 cab e: ' /
- a. LC0 3.2.1 AVERAGE P LINEAR HEA GENERATION E
,, -fygus+ :ow:A / (APLHGR), single loop peration li 'its [speciff e in the COL 3; [ ! 70 'la f7f'l b. LC0 .2.2, 'NINI CRIYICAL ER RATIO (MCP ." single 9, % o,cae 'b~ c loop operation limits [specified in the COLR];fand j \ oa tnt /<*se . LC0 3.3.1.1, " Reactor Protection System (RPS) \ poee ,< ,,,, vl, ,, Instrumentation,' Function 2.b (Average Power Range \ g^ 3 ', Monitors Flow Blased Simulated Thermal Power-High), 1 AllowableValueofTable3.3.1.1-1presetforsingle loop operation. M rN3hgT IA @ APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A* Requi nts of the A.1 Sa sfy the 24 hour LCO met. virements o the 0. { pend % metutten of/ tant 11ty issue./g (continued) @ 2Nse.Rr IB - i. EltlCk BOJb 77 jg 373 b\\ IY 3.4-1 Rev. O, 09/28/92 &f 9:-MlQ i INEERT 1A ------------------------NOTE--------------------------- i Required limit and setpoint modifications for single recirculation loop operation may be delayed for up to 12 hours after transition from two recirculation loop operation to eingle recirculation loop operation. ,l ' i INSERT 1B s i A. fiiIO>.sl} ",'t Recirculation loop A.1 Aggtcre t.T 2 hours jet pump flow wr.ecirculation loop mismatch not within c witheeJ limits. fler: jei=ity j y f ([BN Total core flow as / .1 Determine APRM and once per ' a function of 4' LPRM neutron flux 8 hours , I THERMAL POWER noise levels. I g within Region I4 E ; \ ,- n J ? -- - x)' 47, c.qg 30 minutes f (f , f 7 - after an ,' increase of 2 5% RTP ! '[C Total core flow as klCnitimi.. I ::tir-- 4 1 j f I / a function of THERMAL POWER 'g) festore APRM and LPRM neutron flux (-Isurt iJl within Region IL noise level to s 3 -! [s {L / ) \ #'? ~ times established - [ E of 6pn M j baseline noise /'b , c/ APRM or LPRM levels. ( 2n I neutron flux noise I level > 3 times ; g established baseline noisa \ level. \ V t3 . 9tnvar ~ wen .>20% f'2[we 1//EM'N, ' ( ,1 f j ,[ are n,,~, c..y ie 9,,; ,, g y y, grp - ) te oa v,la l<'3 or /1;of y~~.-ykta!An, ~ _ .Y INSERT RIVER BEND 3.4-1 6) 10/1/93 hdi? v:-IV INSERT 1B (continued) /I 'D Total core flow as C$iti L- acuan_pl (4:d!"=1f f k a function of l festore total core !' ' / THERMAL POWER flow as a function \ ' within Region II . of THERMAL POWER to within -q 4 q'j o 7 _./'G 'O ,*d'L , y' f. f ~' Region I or II,. _ } ,,7 'i fob (Av<# Q$$ s I No recirculation @ itirt artic # )g',1 feduce THERMAL POWER TE91ste@ \ I ' N. loops in operation. -~ ~ to within Region I. h /'mh gg [ F,pn S Y./-r i \ ' . Be in MODE 2. 6 hours ? AliD .3 Be in MODE 3. 12 hours !? 8- %s,;s? /in;/ ad up l G I 'Deba ae r ~?0 ( .r; -ef,A p melia,G sd li-H&) &A sdp440 1 pe {a-e A, nd me \ +/ l se I i ~ l INSERT s RIVER BEND 3.4-1 O,I 10/1/93 1 1 (EAK 92 /YRI} INSERT 2A SR 3.4.1.2 ~erify total / core ow as fune- ion of/THE L POWE to e 24 hours wi in Rajafion I l I
- 3. /
.All [(e o ,, ntf v a -%w 2 wG red:S c.m -% ) or i, , 7//ebm powe.: ad }o kl euc. }/on dAih h Yf 7, /, /- / , l l l l l INSERT RIVER BEND 3.4-2 10/1/93 a - - . . - _ , - -- - _ + - -- a .- , _._n a y ) l 1 l 1 l 1 I 1 70 l l I I I I I 60 Rtoon k-N [ E I ) I I I 50 l I I M { I M I aF I I 1 g E 40 .' . I I I M 1 I M 1 I I I I I T I . I , 1 1 g- 30, 8 , l . I I a . I I . , 20 , 1 1 Y I I J l , I I . . 1 1 I : , 10 I : :: , : I : I I I I I I I [ I [ I I 1 I g I g I y y I I I I 1 1 E I ] I I I 1 1 0 15 25 35 45 55 65 . 75 85 CORE RDW (% RATED) ,d ' N Figure 3.4.1-1 (page i of 1) / Mal Power / Core Flow Stability Regions - t d, $? % ~ b % s RIVER BEND / ! 3.4-g3' 10/01/93 QAt? 9:-itsD S/RVs 3.4.4 $URVEILUUICE REQUIREMENTS SURVEILUUICE FREQUENCY SR 3.4.4.1 Verify the safet T!naccordance of the)requiredy function S/RVs lift setpoints are as follows: with the @ Number of Setpoint Inservice Testing Program QBi S/RVs (nsia) QI fu x--- r 7.59 P; * //'ff 7 a*A L //6 5 . 5 'g%' ; a //r6,'/ 4 4 //to .7.y p 4 y& .s ; E //66 2 a 2 s //90 ) *2 l Ove/5....!.s..._s.3-c..;.me=tti-..,,a-g u_ .., .j SR 3.4.4.2 ..--- -----.-------NOTE-------------------. . Valve actuation may be excluded. h Verify eachNrequi actuates on an actual or simulated relief function S/RV k18[ months h automatic initiation signal. SR 3.4.4.3 ---------------.---NOTE-------------------- Not afterrequired reactor steam to be p @erformed pressure @until 12 hours C l. d !.* i ........................... Verif eachkrequiredb/R'l opens when @ @ manuaflyactuated. k1 months on a S AGGERED TEST BASIS for _(yord beva a.re a2gia.h. do peden.4h Ost, each valve CL - solenoid x BWR/6 STS 3.4-8 Rev. O, 09/28/92 l l 1 fAC 9:-/ Ydl) l RCS Operational LEAKAGE 3.4.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME w . .. B. (continued) 'B *4rify source of 4 hours unidentified LEAKAGE increase is not service sensitive T' Y type 304 type 316 / . ' ,' # g2 austenit stainless I ___ , steelp ,, ,we, p ru u /, r ' < $$'A!$:$"%$'ch /;fle. ~o /etaI. ) C. Required Action and C.1 Be in MODE 3. 12 hours associated Concletion Time of Condition A .45 or B not met. C.2 8e in MODE 4. 36 hours
- 9.B -
Pressure boundary i LEAKAGE exists. l l SURVEILLANCE REQUIREMENTS l i SURVEILLANCE m,Q. FREQUENCY I SR 3.4.5.1 LEAKAG6,) , QQp Verify RCS unidentified a total LEAKAGh,' j/touYI and unidentified LEAKAGE increase are v within limits. l l l BWR/6 STS 3.4-10 Rev. 0, 09/28/92 (KM w!vy RCS PIV Leakago 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Pressure Isolation Valve (PIV) Leakage LCO 3.4.6 The leakage from each RCS PIV shall be within limit. APPLICABILITY: MODES 1 and 2, MODE 3, except valves in residual heat rencval (RHR) shutdown cooling fl $ _h (re net rece:re: x ;;;; :n:] o CI ,tpi; -- " - --" .@when in the shutdown cooling mode of operation. ,o r afs<in de +< - x] + ,, L ' ~' ACTIONS -------.-----------------.-----------NOTES-----------------------_------------
- 1. Separate Condition entry is allowed for each flow path.
- 2. Enter applicable Conditions and Required Actions for systems made inoperable by PIVs.
CONDITION REQUIRED ACTION COMPLETION TIME c Dr eakage from one or ----- ------NOTE------------- A.[fmoreRCSPIVsnot Each valve used to satisfy within limit. Required Action A.1 and t , Required Action A.2 shall s, y g,fe , , ,.,, , c have been verified to meet __., .2 70 ~ ' A-SR 3.4.6.11and ei thl 03 N'*' f * *E' (reactnr c olan pr. su jbou ary [or he gh jpr surlo_ ion of h /s stemJ (continued) BWR/6 STS 3.4-11 Rev. O, 09/28/92 42 % nit ) RCS PIV Leakage 3.4.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.1 Isolate the high 4 hours pressure portion of the affected system from the low pressure 1 portion by use of one closed manual, ' ,, deactivated , g , automatic, or check $Nf g ,' valve. 6,d 'usE AND L3q\ pviL 9i g \ A.2 Isolate the high I'(,;ach 72 hours pressure portion of s the affected system s from the low pressure s t portion by use of a second closed manual, deactivated automatic, or check / valve. B. Required Action and 8.1 Be in MODE 3. 12 hours associated Completion Time not met. N AN,Q B.2 Be in MODE 4. 36 hours I e BWR/6 STS 3.4-12 Rev. O, 09/28/92 l . _ _ l (J4R 93 -l'tal) RCS Leakage DetectiCn' Instrumentation 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Leakage Detection Instrumentation LCO 3.4.7 The following RCS leakage detection instrumentation shall OPERA 8LE: g ~ ...--j'g/
- a. Drywell floor drain sump monitoring system;
- b. One channel of either drywell atmospheric particulate. or h atmospheric gaseous monitoring system; Mand
- c. Drywell air cooler condensate flow rate monitoring systes P APPLICA8!LITY: MODES I, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A@. Drywel& floor drafn/or podeNQ ------------NOTE------------- sep monitoring system LC0 3.0.4 is not applicable. fnoperable. A.1 Restore drywell+ floor 30 days drain sup monitoring system to OPERABLE status. s s (continued) 31 t '/k s RNt/6 STS 3.4-14 Rev. O, 09/28/92 4 ; i [J AR 93 -/Yk 0 RCS Sp;cific Activity ! 3.4.8 3.4 REACTOR COOLANT SYSTEM (RCS) l l 3.4.8 RCS Specific Activity l l LCO 3.4.8 / Th
- q.e ,specific m v m ,2 activity of the reactor coolant shall be,.
{ 'l a. _ E EQUIVALENT I-51)6esWc ectiQ 1 2}$ 1:' _b. J ross/ specific activi W s 106/G Cf/g Mm APPLICABILITY: MODE 1 MODES 2 and 3 with any main _ steam 1in_e not isolated. -.---n g--- - --.y ACTIONS eco ,7 o . 4- i- s- ,44 ' app b ca b l e.- - - - - Jp CONDITION RE@ IRED ACTION COMPLETION TIME 4 A. Reactor coolant A.1 Determine DOSE Once per 4 hours specific activity E@IVALENT I-131. >,{0.2FyC1/gmand OSI . s 4.0 yC1/gm DOSE E EQUIVALENT I-131. A.2 Restore DOSE 48 hours EWIVALENT I-131 to within limits.
- 8. Required Action and B.1 Determine DOSE Once per 4 hours associated Comple. tion E@ ! VALENT I-131.
Time of Condition A not met. M 3 B.2.1 Isolate all main 12 hours steam lines. Reactor coolant >A4.0fyCi/gmDOSE specific activity B l h E@! VALENT I-131. (continued) l l BWR/6 STS 3.4-17 Rev. O, 09/28/92 ~)3 -/wh RCS Sp3cific Activity 3.4.8 ACTIONS C0WITION REQUIRED ACTION COMPLETION TINE 2 B. (continued) 8.2.2.1 Be in MODE 3. 12 hours E 8.2.2.2 Be in MODE 4. 36 hours 1 C. Reactor oolant C.1 Isolat'e all 'in 12 hods specif activi lines st / ~Q >I fyCi/ . C.2 Be i E 3. 2 hours q . 4 ~' i AfD M . C.2.2 Be in E 4. 36 ours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY v ty 1 y / * ' SR 3.4.8 -------------------NOTE-------------------- r.P Only required to be performed in M00E 1. s er, g,4 ,- ...-....................................... - ~ d# Verify reactor coolant DOSE . day EQUIVALENT !.131 specific activity is - O61 0.2 yCi/ge. (continued) BWR/6 STS 3.4-18 Rev. 0, 09/28/92 f 93-/</5D - ' RCS Spe M ActGii T , , 3.4.k ; , e ,, . . . ' SURVEILLANCE / REQUIREMENTS - (con,tinued) ' / t [ [ SURVf!LLANCE j' / ,/ f' FREQUENCY l /. '< .4.8.3 -......-. NOTE.-. / SR ........ ' f Not req red to bt'perfonned/ until 31 dd after minimum of 2 effective full p ' ys r day nd 20 days' of MODE Voperation ave ela ed since.the reacto was last i s ritical f4r a 48 ho . i .L......... ' . i ! l cR 3.0. 2.4 po. pel re6 +or coc(h................... t. 84 day 5 i \ DeteminegE from a 3am@e takenginK7 "^" t i5 \ m i mi . an n 41som o r + e . . , _ ; :, :: : ;;ur once wi ,iw hy: 2,d 20 ty: Of S 1: ( 7~ 0g th;:d s!=ce th:'- ::t:r ::(;;r:tur h:t, h::d 91 da op e '-d u= ? ==- h e '=_:.1 eak e p \e. ,! l o.c/c ed a a e Ol +k S R }i4.1' 3 - / / j' / / i / / / f' ' \ f Q .- j I \ i / # fl0 i BWR/6 STS 3.4 19 Rev. O, 09/28/92 6At 92 R RCS P/T Licits 3.4.11 3.4 REACTOR C00UWT SYS1(A (RCS) 3.4.11 ACS Pressure and Temperature (P/T) Limits LC0 3.4.11 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculationf- e'e- 7 temperature s </ requiremen shall be maintained! within q!p limigspes**4sub e72 APPLICASILITY: At all times. ACTIONS C0EITION REQUIRED ACTION COMPLETION TIME A. .-.-.----NOTE--------- A.1 Restore parameter (s) 30 minutes i Required Action A.2 to within limits. 1 shall be completed if l this Condition is B entered. , ...................... A.2 Deterisine RCS is 72 hours ! acceptable for i Requirements of the continued operation. l LCD not met in ! MDOES 1, 2, and 3.
- 8. Required Action and 8.1 Se in MODE 3. 12 hours associated Completion Time of Condition A not met.
E 8.2 Se in M00E 4. 36 hours l l (continued) i BWR/6 STS 3.4-25 Rev. 0, 09/28/92 (LM 93 N& RCS P/T Litits 3.4.11 ACTIONS (conthiwd) COMITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE--------- C.1 Initiate action to Ismediately Required Action C.2 restore parameter (s) shall be completed if to within limits. this Condition is entered. AMI C.2 Deteneine RCS is Prior to Requirements of the acceptable for entering N00E 2 LC0 not met in other operation. or 3 than N00E5 1, 2 and 3. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 -------------------NOTE-------------------- Only required to be performed during RCS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing. ify @C$ ;;- zrn = " x----- - r' ~ 30 minutes '3A -
- b. heatus and coo' down rates are644444)
J79 Q _ ,,,,,, ___, . _ ,,, __ ,, A f (6 to0*F m og ode Asur Aar'ob SR 3.4.11.2 -Verify RCS pressure and RCS temperature are once within within the criticality limits specified in 15 minutes p - r. ~ ({dgre. 7,y.u.TD prior to ,.4 - control rod j - . pore . - - , ----- ~ . 3 withdrawal for n 0 1l retfra a 9 ae ed 4Ar, y co oho I y the purpose of roR wi+4 J<ana l or 1 e pv<pne of achieving oc4eeveh ced 4,//$. criticality ; g- -. -- -- $o ($ jj VIC QN f gp {$ontfnued) we ~ 6 +4e 1,a , Ye(r /UlC , c. 3,4.//4, uk BWR/6 STS 3.4 26 Rev. O, 09/28/92 [LAR92-NKD RCS P/T Licits 3.4.11 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY l ) SR 3.4.11.3 ............-....--N0TE-...--....-.-....-.- ' l T ' Only required to be met in MODES 1, 2, 3 3 { g and44fwit reactor steam dama pressure eg n 25 psi 9 j ..........f..--- <.h.._non y,q p ., is 6-4 i^- I :_; .; ;__. . fid i;; --- _ . ;w; startup of a j j h g.g recirculation pump ; SR 3.4.11.4 --.....-...........N0TE.----........-.-.... ' Only required to be met in MODE $ l 2. 3 J sad L.....GL . Q ruhu wae,y J,,i Verify the difference between the reactor Once within coolant temperature in the recirculation 15 minutes loop to be started and the RPV coolant prior to each [g ,.) t reture is 6 startup of a / .: t.2 .--- f recirculation N [f50'E) SR 3.4.11.5 ..-.....-....--....N0TE..........-...-.-.-. 3
- Only required to be performed when ,
tensioning the reactor vessel head bolting 1 stads. l > ........................................... l , S.1 Verify reactor vessel flange and head 30 minutes , l ej ,/ flange temperatures are m. .. . _ . . _ ...._ l l ,/ c__......._._ . _ _ _ t i [k90'F:] ! i (continued) SWR /6 STS 3.4 27 Rev. 0, 09/28/92 I (LAR 93-/YND RCS P/T Licits 3.4.11 i SURVEILLANCE REQUIREMENTS (continued) ! SURVEILLANCE FREQUENCY SR 3.4.11.6 ........--..-.-.---N0TE....-............... Not required to be performed until 30 minutes after RCS temperature s 80*F in i MODE 4. ........................................... t 1 Verify reactor vessel flange and head 30 minutes ,jp.f ../ flange.tteperatures are "e -- : r M: tc ~ i /m n _ ::: -' LS e- ,1 70_* F. } SR 3.4.11.7 -.........--..-.--.N0TE--..-.-.-.........-. Not required to be perfomed until 12 hours after RCS temperature s 100*F in MODE 4. 's, i Verify reactor vessel flange and head 12 hours I # 72 flange temperatures are d A h :== h 5 ""j ) g n . ... .. e r w ) { @ 1 0
- Y) , '
l , i INSERT .26h h ' l l l - t OWR /6 STS 3.4-28 Rev. O,09/28/92 1 (LM 93-tyd1 ) INSERT 28A 4 SR 3.4.11.8 --------------NOTE------------- ' [47f Only required to be met in
- g. ,
single loop operation +with the q cima 9 -ac o u recirculation loop flow in the /4 WEtmst operating loop s 50% of rated poggg or recirculation loop flow or THERMAL POWER < 30% of RTP. r e <!< a /d<.k /cc/ ____________________..__________ , fj v Verify the difference between once within the bottom head coolant 15 minutes temperature and the RPV coolant prior to an 7'q _,, temperature isjeit?.ir 07.: li;it-) increase in A12 @p :ified i.; tr.; TTL.i. T THERMAL POWER or an
- to0'F. ) _
increase in loop flow SR 3.4.11.9 --------------NOTE------------- Only required to be met in s ngle loop operation 7with the h,T~ macoce. recirculation loop flow in the /o TN, E
- r"A 4 operating loop s 50% of rated pyy7g fr recirculation loop flow, or THERMAL POWER < 30% of RTP, and
'4 ' ' " " '/" "t"" the idle recirculation loop not . isolated from the RPV. (m/r9#, " /c" ________________..______________ t g'y Verify the difference between once within 15 minutes the reactor coolant temperature #7 # in the recirculation loop not in prior to an operation and the RPV coolant increase in ^ temperature is)Vithia - _ _ _ _ _ _J THERMAL '7 Y - p;;;i . . . _. .. . _ . . _ . .) POWER or an >72 increase in L 450'FJ s loop flow \ \ 1NSERT fi&vRe 3. YJ/-/ i ) INSERT fol' 9Y RIVER BEND 3.4-286) -10/1/0:23 1500 ,- I AA* $ S'C C' 1400 - f , r .. e # 1 0 l 'S ' I .a *] .,
- c. 1200 -
i l lll l ? $ 100C I m: J I: 1! a[':. - 1 m , W / s
- A'.5'.C' - CCRE BELTUNE
/ e,' ' ,[ AFTER ASSUMED 111'F i g 800 SHIFT FROM AN INITIAL ,0 f,' ,sl , e WELD RT gr OF -SQ'F # . s ! x ' l A - SYSTEM MYOROTEST UMIT E L, g SCO WITH FUEL IN VESSEL f 8 - NON-NUCtEAR HEATING *! E uuT ? ? 7 y g a g C - NUCLEAR (CCRE CRITICAL) uuiT S 400 A - VE5SEL 015 CONTINUITY 3:2 ase ,,,, umis A --- CORK BELTUNE WITH 1117 SHIFT 8 200 , aow [C CUM 5 A'.B'.C' ARE VAUD ' 70Y / FOR 5 EFPV 0F OPERATION CUM 5 A s.C ARE vAu0 .- FOR 2 EFPY OF OPERAfl0N , 0 i i i ..:G 100 200 300 400 500 600 i RECElVED MINIMUM REACTOR VE55EL METAL TEMPERATURE (V) ; l AUG 0 91990 g y, a. - g l FIGUREo - " * ? M, J' # SDC MINIMUM TEMPERATURE REQUIRED v5 MPRE55URE RIVER BENO " h
- h@ C6'" "C.,"
i , zuar - 7. '/- 27 [h @' 9: /6/) Recirculatien Locps Operating B 3.4.1 BASES BACKGROUND The subcooled water enters the bottom of the fuel channels (continued) and contacts the fuel cladding, where heat is transferred to the coolant. As it rises, the coolant begins to boil, creating steam voids within the fuel channel that continue until the coolant exits the core. Because of reduced moderation, the steam voiding introduces negative reactivity that must be compensated for to maintain or to increase reactor power. The recirculation flow control allows operators to increase recirculation flow and sweep some of the voids from the fuel channel, overcoming the negative reactivity void effect. Thus, the reason for having variable recirculation flow is tu compensate for reactivity effects of boiling over a wide range of power generation (f.e., 55 to 100% RTP) without having to move control rods and disturb desirable flux patterns. Each recirculation loop is manually started from the control room. The recirculation flow control valves provide regulation of individual recirculation loop drive flows.- The flow in each loop can be manually or automatically- , controlled. ' 0,47 3 APPLICABLE The operation of the Reactor Coolant Recirculation System is f SAFETY ANALYSES an initial condition assumed in the design basis loss of coolant accident (LOCA) (Ref. 1). During a LOCA caused by a t recirculation loop pipe break, the intact loop is assumed to s p/z howeW,ve e 'X# ./ , provide coolant flow during the first few seconds of the s accident. The initial core flow decrease is rapid because pi,r,/y>,3 -a; res,e.u[ the recirculation pump in the broken loop ceases to pump (i }f' / Cdic [ reactor coolant to the vessel almost immediately. The pump in the intact loop coasts down relatively slowly. This pump g, coastdown governs the core flow response for the next a,,n se, Me, u several seconds until the jet pump suction is uncovered . o v c, (Ref.1). The analyses assume that both loop]s are operating cro5 7. . . at the same flow prior to the accident. Q4 M av with a flow mismatch between the two loops, <CE -"-='e ~ (c/ tv ,/e. Me *Ieeerrvatively assg: er ef;e =:t irffn the loop with the itgher flow.TISEflow coastdown and core response are ( CMy M'tentially more severe in tnisicase, ince the intact loop 7 - cuch e :t:-tm at a lower flow rate and t e core response is '# "// '" 'N the same as if both loops were operating at the lower tiow rate) AThe recirculation system is also assumed to have [b eec , c[efe,,g[ sufficient flow coastdown characteristics to maintain fuel l 40 os peuy :, e' 'c Ya:o.5c<t wy neurkn, / alJ ,,,,,,) -- -A- - '/ (continued) BWR/6 STS ' B 3.4-2 Rev. O, 09/28/92 (f4 C -m LC - Recirculation Lo:ps Op3 rating 8 3.4.1 ~ BASES APPLICA8LE themal margins during abnonaal operational tran s , SAFETY ANALYSES (Ref. 2), which are analyzed in Chapter 15 of th R. Pl (continued) i A plant specific LOCA analysis has been perfonned assuming only one operating recirculation loop. This analysis has demonstrated that, in the event of a LOCA caused by a pipe < break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Ref. 3). The transient analyses of Chapter 15 of th been performed for single recirculation loo peration (Ref. 3) and demonstrate sufficient flow coastdown R have also h characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation. loop ; operation, modification to the Reactor Protection System l ~ average power range monitor (APRM) instrument setpoints it I also required to account for the different relationships ' ' between recirculation drive flow and reactor core flow. The gM457 APLH6R and MCPR seseemes-)for si le loop operation are specified in the COLA. The APRM low biased simulated thermal power setpoint is in LC0 3.3.1.1, " Reactor Protection System (RPS) Instrumentation." Recirculation loops operating satisfies Criterion 2 of the l NRC Policy Statement. l n- a _ l LC0 Two recirculation loopsregliired are+&@L) to be in Q4) operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loan the assumptions of the LOCA analysis are satisfied.f Wit the imits ect rea ing ^ MM oop wi the / ower/ [h .9 \ U g s A[W(SR 3/4.1.1 [n t met./the rycircul tion flod must b considered rot in perat on. W hfonly one recirculation toon in ooeration,mdifications to the required APLH6R limits (LC0 3.El, ' AVERAGE PLANAR LINEAR YNEa* FeR HEAT GENERATION RATE (APLH6R)'). MCPR limits (LCD 3.2.2, uf ne 4704Rr3 " MINIMUM CRITICAL POWER RATIO (MCPR)"), and APRM Flow 81ased , fh 4,fg / Simulated Thermal Power-High setpoint (LC0 3.3.1.1) be applied to allow continued operation conP stent with t e j * ' . ' ,he j' " f'*" assumptions of Reference 3.
- we,,1,0.0 a mt M8E
'"'" * " f h .nisEltT 8,3 B (continued) BWR/6 STS B 3.4-3 Rev. O, 09/28/92 i w I f 7 . - m {1B 9 -/WI) prusi c ? %S $ oA wlch cu e. dl;>v or ~/o hl core. Non> INSERT B3A ; gq,i In addition, the total core flow Foxpressed as a functhon of THERMAL POWER must be in Region I' as identified inMfigure 03 4 1-1, " THERMAL POWER / Core Flow Stability Regions." .p7 (4 - Alternatively, with INSERT B3B The LCO is modified by a Note which allows up to 12 hours before having to put in effect the required modifications to required limits and setpoints after a change in the reactor operating conditions from two recirculatien loops operating to single recirculation loop operation. If the required limits and setpoints are not in compliance with the applicable requirements at the end of this period, the associated equipment must be declared inoperable or the limits "not satisfied," and the ACTIONS required by nonconformance with the applicable specifications implemented. This time is provided due to the need to stabilize operation with one recirculation loop, including the procedural steps necessary to limit flow and flow control mode in the operating loop, f * '-' ==" eh : monitor for excessive APRM and local power range monitor (LPRM) neutron flux noise levels; and the complexity and detail / required to fully implement and confirm the required limit and, setpoint modifications, e r 7s l INSERT , RIVER BEND B 3.4-3 10/1/93. ]At G:-//r/] l 1 Recirculation Le<ps Op3 rating B 3.4.1 BASES (continued) l APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is , considerable energy in the reactor core and the limiting i design basis transients and accidents are assumed to occur. In MODES , 4, and 5, the consequences of an accident are reduce- od the coastdown characteristics of the recir- ation loops are not important. g, g, ACTIONS Ad g ,h .g,, g y y g i'l I With c : 7::rM - . : 21 tt ;.;; c.d 5dkhe recirculation ~ ! loops must be restored to operation w th matched flows I s wit siit*S hours.fA ci u tio 1 pi c si re no i i i i id o wh ,oper ion when.the in ha 1 e t b n1 ta j p f f a wo oo. 4 ,' , , , ' '"f migp6at. an autre li t I _. th.th lower f ~ ra on hbuld a LOCA occ ""1 (unusune o circulation oog zonsidered obt no in peration' m o , the so ' low - v st own and res its s se na 6on d bf g he an ys o o a 'I i i t al tof st ab p r ig ta s.) Alternatively, if the single loop requirements of the LC0 are applied to operating imits and RPS setpoints, operation pl with only one recirculation loop would satisfy the requirements of the LC0 and the initial conditions of the accident sequence. ~ The our Completion Time is based on the low probability ! of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected. 1 F G.p .4 .[/y This Req Action does not requ tweircul ionpumpinth lowest ow loog et.?n the mismateb trippjing W : betweera otal j pump ows of e tu) bhp is greater , I' "" #'4 '"" ff redn el ?> -s *'e , than t requi limi . Howe r, ik G5=i, sMye large f ) flow ismatch occur low fl or reve/se flow can occur //,.;6 ML2 the ow flow oopje pumps, ausing v bration of the jed'in / lL) , A ms , .me. ,e nn 4 /Jo p. s. If ero or verse f ow is d ected, e condi ~ l should be 11evia d by ch ging f1 contro valve po, tion sition / j#Y f# to re-e ablish orward f or b trippin the pump ( i XI '["" )n' , J (continued) BWR/6 STS B 3.4-4 Rev. O, 09/28/92 i _ Q 9 -iv,1) f - ~ ~ MU (J t C Recirculation Loops Operating j ,' 3 3,4,3 .A E, %r l N uem - r _ ACTIONS C b C .'! d d D d ' (continued) ' - v? 1/ - / 1,With no irculati loops in o ration, Action d associat Completio Time of theApuired > met. t unit is itio A not 4 utred to brought a in which 0 does not pply. To hieve thi stauus the niant; MM theybebrought gust; o MODE 3 w hin 12 h Ln his 9 j BSA q f con tion, tne fo rating bec se of the circulati loops a not re ired to be uced sev rity of s and nimal depe ence on th recircul ion loo coastdown haracteri ics. The a owed C etion Ti of 12 ho is reasonab1 , based on rating a rience, o reach E3 from ful power cond ions in ap orderly nner and g11e ing plant s tems. / thout) SURVEILLANCE SR 3.4.1.1 REQUIREMENTS
- This SA ensures the recirculation loop flows are within the allowable limits for sisaatch. At low core flow (i.e.,
h <J70Ref rated core flow), the MCPR requirements provide Targer margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch 001 can %arefore core flew. The be allowed when core flow is </Jo]Tof rated recirculation loop jet pump f ow, as used in f this Surveillance, is the summation of the flows from all of e the jet pumps associated with a single recirculation loon. Q w q=gs, g F w'm / / 5$N-The mis ch is measured in terms of narcant o" rated cag y ! d.b s k , Et f I*'P/ d5R is not required when both loops are not in operation .pt.,4,fLh: ' i since the mismathh limits are meaningless during single loop / or natural circulation operation. The Surveillance must be I /N WM N 1 / performed within 24 hours after both loops are in operation. / # The 24 hour Frequency is consistent with the Frequency for b _L M ' ,/ jet pump OPERASILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner. { J/7 e NKW MB -@ (continued) BWR/6 STS B 3.4-5 Rev. O, 09/28/92 ~ /T42 52-10 t ) INSERT B5B SR 3.4.1.2 This SR ensures the reactor THERMAL POWER and core flows are within appropriate parameter limits to prevent uncontrolled - power oscillations. At low recirculation flows and high reactor power, the reactor exhibits increased susceptibility to thermal hydraulic instability. This SR identifies when the conditions requiring interim actions are necessary. The Frequency is based on operating experience and the operators' inherent knowledge of reactor status, including significant changes in THERMAL POWER and core flow. n . 'l 'SR 3[4.1 M , h ; / 0' T is S ens res the reactor THERMAL POWER and core flows are ' ithi app opriat'e parameter limits to prevent uncontrolled - powe osc latipns by verifying the APRM'and LPRM noise Jevels. i At ow ows and high power,/ the reactor exhibits increased sdcap bility to thermal hydraulig instability. / This SR ent les when thp conditions requ' iring in'terim adtions are nece ary. The Frequency gs based /on oper ting expierience and the opera ors' in'herent knowledge including si nific nt changes in THERMAL Pp/of WER and reac core or stat,us, flpw. The,IR is j mddifiagf by a, Note requiring performanc of thef'SR when j in the; / / j appli, cable region. ,/ , j/ ,/ p - / J /) L E6 7 [ !' C ! ~- ---..- -.- ',"] \ Z bovY a. /OCA oc e v r' wik 77/E/'*w Qpe.>Eit' 7 7 0 Vo f7?y I' ! s^ kco<e ,csso~ u mai dn= c o ~M 4r tie A 'A '* a., a yl s e; . Thaie Are , en /y a ll~ des +sa e is a/loE h \ i \ aface rdemn PoweK k A 70 6 crP, l l' T$e .i hur Oy A/O 7%< a a<sc0 ou he low \ p,, A,is, ,) n s a :Je 4 occo<,G Li,~, du n i p <aE, en a eea o~6/c h i., e h coy e/e l te l hj,c? ! A /,;n, n2 o r heg,,aJ cm ,,,ov;/s,,~j hy ycu,/o'u si/~,y , c %es ,,, rn an pwu c , N,,L 4 1 \ fC guteh ch!ee!ef , / N _._ INSERT ) RIVER BEND B 3.4-5 10/1/93 . - - _=. - __ . .. . . . . (2AR 9:-10 I} c , '/ - - ./ h .t r 7 Pdhm 4 hen: INSERT B5A - Due to thermal hydraulic stability concerns, operation of the plant is divided into three regions based on THERMAL POWER and core flows. Region III is a power / flow ratio with core . flow < 39% of the rated core flow. Region II is a power / flow ratio f with core flow a 39% and < 45% of the rated core flow. Deliberate entry into Region III is not permitted, and if'it
- ,.d occurs,9- "i=w action is required to exit the region y f reducing THERMAL , POWER through control rod insertion or by increasing recirculation loop flow by opening the flow control d . 'l valva. Operation in Region II is also more susceptible to fffy instability than normal operating parameters. However, !
s operation in this region is allowed with the exception that if l' ;'N evidence of instability occurs (i.e., APRM or LPRM neutron flux level is three times the established baseline) then (=- m ar P l \ action is required to exit this regionj F ; l C - l \ A determination of APRM and LPRM neutron flux noise levels every ! I 8 hours provides frequent periodic information relative to l gfestablished baseline noise levels (see that l l Condition @ ion of indicate stable steady state operation. A determinat ' I these noise Imvels within 30 minutes after an increase of a 5% i RTP provides a more frequent indication of the stability of , operation following any significant potential for change of the /M,od thermal hydraulic properties of the system. These Frequencies fy provide early detection of neutron flux oscillations due to core , /g,,lc/y b;hV thermal hydraulic instabilities. 4 ' J: m-r -- r = -3 restore the plant to a more stable ' ' power / flow ratio if such indications of limit cycle neutron flux oscillations are. detected. ?f) ,D > . E.1 L2, and I'/3 p- vp With no recirculation loops in operation, the unit is required to be brought to a MODE in which the Lc0 does not apply. Action d must be initiated &% Wto reduce THERMAL POWER to be / within the limits to assure thermal hydraulic stability concerns / are addressed. The plant is then required to be placed in MODE ; 2 in 6 hours and MODE 3 in 12 hours. In this condition, the i recirculation loops are not required to be operating because of the. reduced severity of DBAs and minimal dependence on the ; I recirculation loop coastdown characteristics. The allowed l l Completion. Times are reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner ,. and without challenging plant systems. c,1 ,~ - u42 ::!wk, Coy ff,-n7, - _%.a,m na m Q ,ru 6 : ,' ' y , E,- , * <,: 7 c , -p % ~-> t 1 ,c /. -i, .e ,-: u ,1 -lo nono in cm c a 4,ly , wn,- 4 udad cbs'/hyy p /W ., b y - . INSERT RIVER BEND B3.4-56) "10/1/ N r (2AK 93-/Yf t) INSERT BSA (continued) Gd e 9,4 ' If the required limit or setpoint modifications are not j jV performed within 12 hou;s after transition from two s recirculatrion loop operation to single loop operation, the required limits and setpoints which have not been modified must be immediately declared not met. The Required Actions for the associated limits and instrument channels must then be taken. INSERT g RIVER BEND B 3. 4 - 5 (b) 10/1/9 6 (EAR 9.:-15 R circulatica Le:ps Operating ! B 3.4.1 BASES (continued) m REFERDtCES 1. & 2. R, Section %.:.: .i t & . Section d!FNM99if. MS_,4. l.4] i~ . x3. m..; #.; ;; . .;7. .. ... d :.. -,. .,h2;....rp g USA R , S ec.4 15.0 4
- r
^ dll SNSGXT / kEA.) , o -) oW 483 "fV) . e BWR/6 STS B 3.4-6 Rev. O, 09/28/92 ~ jg 3.el. .1 }s f pore. 2n$ h ') 3 ': ' ' D l l I INSERT B13A ince refueling activities (fuel assembly replacement or *d/d'*[ shuffle, as well as any modifications to fuel support orifice fD size or core plate bypass flow) can affect the relationship f etween core flow, det pump flow, and recirculation loop flow, / these relationships
- g be re-established eac (cycle. During
/ vn/er s w/t th nitial weeks of operationf li a -- cycl.a wnile aselining coe8ty,,b new " established patterns", engineering judgement of the daily q f # surveillance results is used to detect significant abnormalities ' which could indicate a jet pump failure. s s / G ot l 'I dW 0 C' ll f,,,, , u,;nj ptk,s//1/by nnyLo < <r~~ a a,,aa a v. o L - c,s a u a / ) \ 4 77 INSERT RIVER BEND B'3.4-13 10/1/93 (J# & - 5/Y's 8 3.4.4 8ASES (continued) APPLICABLE The overpressure protection system must accommodate the SAFETY ANALYSE 3 most severe pressult transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves high neutron flux (i.e.,(MSIVs) failure followed by reactor of the direct scram scram on h p* "N associated with MSIV position) (Ref. 2). For the purpose of Ine analyse W of the S/RVs are assumed to operate in @ b. the relief mode. an4gg in the safety mode. The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 11% of vessel design pressure (110% x 1250 psig = 1375psig). This LC0 helps to ensure that the acceptance Ilmit of 1375 psig is met during the design basis event. Reference 3 discusses additional events that are expected to From an ove ressure sta W : actuate the S/RVs.' "%:d ad" "' '"' 5/RVs satisfy Criterion 3 of the NRC Policy Statement. 4 a L. LC0 The safety function of 5/RVs is required to be OPERABLE in the safety , and an additional S/RVs - OAe (otaer than eneseven S/RVs that satisfy the safety function) must be OPERABLE in the relief mode. The '4 Q requirements of this LCO are a plicable only to the !Aq. capability of the S/RVs to mec cally open to relieve '- excess pressure. In Reference n evaluation was I.- performed to establish the parametric relationship between the peak vessel pressure and the number of OPERABLE S/RVs. f g g " Y The results show that with a minimum of egges($/RVs in the sarety mese a ASME Code lim @it of 1375 psig is not exceeded.S/RVs in th The S/RV set mints are established to ensure the ASME Code , limit on peat reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve be set at or l below vessel design pressure (1250 psig) and the highest safety valve be set so the total accumulated pressure does not exceed 110% of the design pressure for conditions. The transient evaluations in Reference 3 are based on these set oints, but also include the additional uncertainties of i P3 .2 % of the nominal setpoint to account for potential l setpoint drift to provide an added degree of conservatism. (continued) i BWR/6 STS B 3.4-16 Rev. 0, 09/28/92 l (t4K c?3 /+'!D RCS Operaticnal LEAKAGE S 3.4.5 BASES (continued) ACTIONS M & LEAKAGE With RCS unidentified or total LEAKAGE greater t e limits, actions must be taken to reduce the . cause the LEAKAGE limits are conservatively below the LEAKAGE that would constitute a critical crack size, 4 hours is allowed to reduce the LEAKAGE rates before the reactor must be shut down. If an unidentified LEAKAGE has been identified and quantified, it any be reclassified and considered as identified LEAKAGE. However, the total LEAKAGE limit would remain unchanged. i An unidentified LEAKAGE inc /$CPet ase o >2spswithina7 hour h period is an indication of a tantial flaw in the RCPS and (Ido gggyf fQ must be quickly evaluated. A hough the increase does not lealaye Mk necessarily violate the ab olute unidentified LEAKAGE limit, sucA TA2f de certain susceptible c nts must be detemined not to be Currewf pde the source of the LEAKAG within the required Completion #5 g',85 g,g , h. N an W M W M E N masegnaw - -- E rueaired l' aits, an alternative to reducing L E o Mc f ye w' t sin imtsJistoevaluateRCStyMe304 a ypeI6/. 'c g ,gere4sc /4 #e austenitic stainless steel pipingettet is,,sub.iect to hiah ,, /f u,on 2Wf>* stress or that contains relative y~ stagnant or intermittent , source of the increased M m w fluids and - - - - it i-erf/ce 47 de Souted on LEAKAGE. h' . twa ' 4 *,,g,6te The 4 hour Completion Time _to jiroperInverify the ' / source before the reactor must be shut down. mehd.s) ,,,9 g ,,,,,q,. L /,,. Jan l C.1 nad C.2 " " ' ' ' ' " * ' d " '/ ' A"' d" " If any Required Action and associated Completion Time of , l Agauge ,w e u* Condition A or B is not met or if pressure boundary LEAKAGE I # ,, exists, 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 MODE 3 within 12 hours and to MODE 4 within [1 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant j.79 conditions from full power conditions in an orderly manner and without challenging plant systems. (continued) BWR/6 STS 8,3,4-2a Rev. 0, 09/28/92 (IM 92 MI] RCS Operaticnal LEAKAGE 8 3.4.5 8ASES (continued) SURVEILLANCE SR 3.4.5.1 v. I~ REQUIREMENTS io f The RCS LEAKAGE is monitored by a variety of instruments ' designed to< @ id c h. - 22.. LE"."J. : : - ; . . m., .no io,f-quantify the various types of LEAKAGE. Leakage detection instrumentation is discussed in more detail in the Bases for LC0 3.4.7, "RCS Leakage Detection Instrumentation." Sump OI leverc... _, .. .... _ , typically monitored to determine actual LEAKAGE rates. However, any method may be used to quantify LEAKAGE within the guidelines of Reference 7. In conjunction with alares and other administrative controls,
- 12. ^ hcur Frequency for this Surveillance is appropriate for
- Mtifying changes in LEAKAGE and for tracking required trends (Ref. 8).
REFERENCES 1. 10 CFR 50.2. Fel e b W.e s'< AST m \ 2, 10 CFR 50.55a(c). 3. I , AAtot. B Wres 4 Gig L _x a l h k1- W all z f % >' -. C_
- 4. 10CFR50,AppendixA,GDC55 GEAP-5620dApril1968.
,+ ca. w ee ,+. 5 @. IRlREG 067,Mtober 1975. A mlest S b<t P3W J kl.- _ s,
- 6. M Wder Reudor %%"
OEl .Sectiop5.2.5.5.3T. _ DI - - - "
- 7. Regulatory Guide 1.45(Ms[1973. __
- 8. Generic Letter 88-01, Supplement i 3
RC Pos%'on s,. "I& SCL 8nIR a ve;.w m.m moi w,,y Fe 6$ m2' -) BWR/6 STS B.3.4-25 Rev. O, 09/28/92 CMX 9J-A1E/ ) , RCS PIV Leakage B 3.4.6. 8ASES (continued) APPLICA81LITY In MODES 1, 2, and 3, this LCO applies because the P!V leakage potential is greatest when the RCS is pressurized. or
- In MODE 3, valves in the RHR flowpath are not required to 3
fra,,sMd}k meet the requirements of this LCD when inithe RHR mode of cg operation. g , c ## p In M00ES 4 and 5,' leakage limits are nofprovided because the lower reactor coolant pressure results in a reduced \ -y, potential for leakage and for a LOCA outside the og containment. Accordingly, the potential for the consequences of reactor coolant leakage is far lower during these MODES. ddms ACTIONS The ACTIONS are modified by two Notes. Note I has bee - -M provided to modify the ACTIONS related to RCS P!V flow paths. Section 1.3, Completion Times, specifies once a l Condition has been entered subsequent tee 4a C subs ! components or variables exp,ressed in the Condition,ystems, ! discovered to be inoperable or not within limits, will not l result in separate entry into the Condition. Section 1.3 : also specifies Required Actions of the Condition continue to l apply for each additional failure, with Com based on initial entry into the Condition. pletion Times However, the Required Actions for the Condition of RCS P!V leakage limits l exceeded provide appropriate compensatory measures for ; separate affected RCS PIV flow paths. As such, a Note has been provided that allows separate Condition entry for each i affected RCS PIV flow path. Note 2 requires an evaluation l r of affected systems if a PIV is inoperable. The leakage may j have affected systes OpDIABILITY, or isohtion of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety function. As a result, the applicable Conditions and Required Actions for systems made inoperable by PIVs must be entered. This ensures appropriate remedial actions are taken, if necessary, for the affected systems. A.1 and A.2 If leakage free one or more RCS P!Vs is not within limit, g the flow path must be isolated by at least one closed _g 0C4 manual,dep-'ivatedeutoestic;orcheckvalvewithin 4 hours. - e " --" N - ~ ' -- "" -- - ' : -M MSESTl - b 2. 8 A j ' (continued) IWR/6 STS B 3.4 28 Rev. 0, 09/28/92 l (1M v3-Hk INSERT B28A check valve may be used for this purpose if leakage past the /checkvalvedidnotexceedtheallowableleakagelimit at the last reOteling outage, or after the last time the valve was known to have opened, whichever is more recent.
- a. 4 .-
fcguNe tl'oh d.l e h Wetarreh kslNa A.2 aie uohis y fdh 270 , ,y, g , .j, j, , ,f INSERT RIVER BEND B'3.4-28 10/1/93 [AM 93-19lI) RCS Leakage Detection Instrumentation ; 8 3.4.7 ; B 3.4 REACTOR C0OLANT SYSTDI (RCS) l 8 3.4.7 RCS Leakage Detection Instrumentation BASES l BACKGROUND GDC 30 of 10 CFR 50, Appendix A (Ref.1), requires means for detecting and, to the extent practical, idcntifying the location of the source of RCS LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems. Limits on LEAKAGE from the reactor coolant pressure boundary (RCPS) are required so that appropriate action can be taken before the integrity of the RCPB is impaired (Ref. 2). Leakage detection systems fot the RCS are provided to alert i the operators when leakage rates above normal background levels are detected and also to supply quantitative measurement of rates. The Bases for LC0 3.4.5, 'RCS - Operational LEAKAGE." discuss the limits on RCS LEAKAGE rates. Systems for separating the LEAKAGE of an identified source from an unidentified source are necessary to provide prompt , and quantitative information to the operators to permit them ' to take immediate corrective action. LEAKAGE from the RCPB inside the drywell is detected by at h least one of gip three independently monitored variables, ' such as sump leve changes and dr particulate radioactivity levels.ywell gaseous The 3rimary and means of , quantifying LEAKAGE in the drywell is dr ' drain sump no na em Q .ywell ,,, f1a ped The drywel floor ersTirsump monitoring syste2monitorFIhe P4 LEAKAGE collected in the floor drain sump. This unidentified LEAKAGE consists of LEAKAGE from control rod I se.J e ,LY' drives, valve flan _,,, e..ges.,,orand packings, I #3 5*h wdc e ;-2:li drywell floor drains,unit air cooling athcGasan d.M e condensate drains, anst any LEAKAGE not collected in the ' . _ . drywell equipment drain sump. The drywelllfloor drain su transmitters that supply level indications in the main ', i control room. 's I -r $314 % - II The or drain sump 1 indicators have tches that 4 Ml start an the sump pump utred. r starts ) **' ) I % ach time the s s-pumped down to e low level se oint.j (continued) BWR/6 STS B 3.4-31 Rev. O, 09/28/92 (1 A R 93 /& RCS Leakage Detection Instrumentamn l B 3.4.7 1 8ACKGROUND ff the sump fills to the high level setpoint before the ' (continued) timer ends, an al sounds in the cont I room, indicating ' a LEAKAGE rat to the sump in exce of a preset limit. A l v second ti starts when the sum s start on high level. ! / Should s timer run out before'p-tae sump level reaches the low el setpoint, an alamis sounded in the control room /c/, ' i cating a LEAKAGE rate into the sump in excess of a ' preset limit. A flowAndicator in the discharge line of the drywell floor draiVsump pumps provides flow indication in i P , the control roce. ,._3[1- The d drywel 11D monitoring systems continuously monitor the l atmosphere for airborne particulate and gaseous radioactivity. A sudden increase of radioactivity, which any be attributed to RCPS steam or reactor water LEAKAGE i 2 annunciated in the control roce. The drywell atmospherhsh particulate and gaseous radioactivity monitoring systems are not capable of quantifying leakage rates, but are sensitive enough to indicate increased LEAKAGE rates of 1 gpa within 1 hour. Larger changes in LEAKAGE rates are detected in proportional y shorter times (Ref. 3). 1 l Condensate from 6 the 49 drywell coolers is routed to l the drywell floor drain sump and is monitored by a flow i Ogl transmitter that provides indication and slams in the j - control room. This drywell air cooler condensate flow rate - f monitoring system serves as an added indicator, but not _ quantifier, of R$$ unidentified LEAKAGE. _ APPLICA8LE A threat of significant compromise to the RCPS exists if the SAFETY ANALYSES barrier contains a crack that is large enough to propagate rapidly. LEAKAGE rate limits are set low enough to detect sh g the LEAKAGE emitted from a single cragk in the RCPS (Refs. 4 and5). Each of the leakage detection systems inside the ) #s drywell is designed with the capability of detectino LEA _KAGE a -r- p;M /4 " le 1 \j gss than the:!= 7,riete established ef - rees LEAKAGE LUX 5 " teelett rate limitgird
- l - 3 ;; nidy -
o.( d, e # 6## 4 :-:=: c::: e:: c allows the operators to evaluate the ? significance of the indicated LEAKAGE and, if necessary, shut down the reactor for further investigation and i corrective action. The allowed LEAKAGE rates are well below the rates predicted for critical crack sizes (Ref. 6). - (continued) IWR/6 STS B 3.4-32 Rev. O, 09/28/92 hr 9cW@ ' RCS Leakage Detection Instrumentation ' B 3.4.7 i l BASES ' l APPLICABLE Therefore, these actions provide adequate response before a SAFETY ANALYSES significant break in the RCPB can occur. (continued) RCS leakage detection instrumentation satisfies Criterion 1 I of the NRC Policy Statement. -- m n \ QM P*dt!WQ4) s - -1% i 2 l d' LCO The drywell/floo} drain suur monitorin sys Mequired # to quantify the unidentified LEAKAGE from the RCS. Thus, ! @ for the system 3t o be considered OPERABLE, either the flow monttoring or the sump level monitoring portion of the l k a stem,must be 0PERAS4E. The other monitoring systems P" g 4;, privTdep; ch_..- to the operators so closer examination in[lo /e, t of other detection systems will be made to determine the extent of any corrective action that may be required. With the leakage detection systems inoperable, monitoring for LEAKAGE in the RCPB is degraded. . 1 APPLICABILITY In MODES 1, 2 and 3, leakage detection systems are required to be OPERABLE to support LC0 3.4.5. This Applicability is consistent with that for LC0 3.4.5. ) ) l ACTIONS &J With the drywellvfloor drain sump mnitoring system i inoperable, no other form of sampling can provide the equivalent information to quantify leakage. However, the drywell atmospheric activity monitoryand the drywell air-cooler condensate flow rate monitorLwill provide indications of cha=r h la g \ (top fec uL With the drywe11MT'our vrain s - nitoring system PG in<eerable. but with RCS unidentified and total LEAKAGE 11 be'nqdeterminedeverTy hours (SR 3.4.5.1), operation may cont <nue for 30 days. The 30 day Completion Time of , Required Action A.1 is acceptable, based on operating experience, considering the multiple forms of leakage detection that are still available. Required Action A.1 is modified by a Note that states that the provisions of LC0 3.0.4 are not applicable. As a result, a MODE change is allowed when the drywell floor drain sump monitoring system (continued) BWR/6 STS B 3.4 33 Rev. O, 09/28/92 (IAR93-14'Ph ~ RCS Leakage Detection Instrumentat'on 8 3.4.7 BASES SURVEILI.ANCE jL 3.4.7.1 (continued) REQUIREMENTS properly. The Frequency of 12 hours is based on instrument reliability and is reasonable for detecting off nonnal conditions. SR 3.4.7.2 g, 0 This SR mquires the perfomance of a CHANNEL FUNCTIONAL e9 TEST of the required RCS leakage detection instrumentation. N \ The test ensures that the monitors can perfom their functio The test also verifie e \ mem <=n in the eaai-t desiredaccuracy ido relative manner. of the instrumen * ,oa h8 ;** Meg. The Frequency of 31 days considers instrument reliability, and operating ekperience has shown it proper f3 for detecting degradation. SR 3.4.7.3 This SR requires the perfomance of a CHAf91EL CALIBRATION of therequiredRCSleakagedetectioninstrumentationchannys. C) The calibration verifies the accuracy of the instrumentation . ' etsdal,includingtheinstruments1ptedinside .i %d r t r :_:.^. .. TheFrequencyofM18}monthsisatypical j refueling cycle and considers channel reifability. gM Operating experience has proven this Frequency is acceptable. REFERENCES 1. 10 CFR 50, Appendix A, GDC 30. _.
- 2. Regulatory Guide 1.45, May 1973. d' b * " i" h 3. h , Section) $.2.5.2 [ h ASTM g ,A 100 ,., g6Pm i
- 4. GEAP-5620,[ April 1968.
I%Q A _- 411 FL,
- 5. NUREG-75/067,4 0 ctober 1975.
- 6. ,Section/5.2.5.5.3[ h I hhg44a_ __ ud fuc. loa 6 oI O db d d )I
- 7. 05% Mk U?l p3 k,;,us ut.n sw: P:q J >
f4 {iilg (A{e keubr hufs BWR/6 STS B 3.4-36 Rev. 0, 09/28/92 l l - .= . - - . . . . -- -- _- - -- - . _ . .- - _ - . [L A K k -l i f ') RCS Sp;cific Activity B 3.4.8 8 3.4 REACTOR C00LANT SYSTEM (RCS) B 3.4.8 RCS Specific Activity 8ASES BACKGROUND During circulation, the reactor coolant acquires radioactive materials due to release of fission products from fuel leaks into the coolant and activation of corrosion products in the reactor coolant. These radioactive materials in the coolant can plate cut in the RCS, and, at times, an accumulation will break away to spike the nomal level of radioactivity. The release of coolant during a Design Basis Accident (DBA) could send radioactive materials into the environment. Limits on the maximum allowable level of radioactivity in the reactor coolant are established to ensure, in the event of a release of any radioactive material to the environment during a D8A, radiation doses are maintained within the limits of 10 CFR 100 (Ref. 1). - - This LC0 contains odine "O pecific activity g '/ limits. The iodine isotopic activ' ties per gram of. reactor conbat are expressed in_ teras of a DOSE EQUIVALENT I-131. j '!a b ae bb[a ry e dhecodlant.fTheallowable'evelsareintendedtolimit the 2 hour radiation dose to an individual at the site boundary to a small fraction of the 10 CFR 100 limit. APPLICABLE Analytical methods and assumptions involving radios e , SAFETY ANALYSES material in the primary coolant are presented in th AR Pt ' (Ref.2). The specific activity in the reactor coo (the source tem) is an initial condition for evaluation of the conseq)uences of an accident (MSL8 outside containment. Noduefuel to a mainissteam damage line break postulated in the MSLB accident, and the release of radioactive material to the environment' is assumed to end when the main steam isolation valves (MSIVs) close completely. This MSLB release foms the basis for detemining offsite doses (Ref. 2). The limits on the specific activity of the primary coolant ensure that the 2 hour thyroid and whole body doses at the site boundary, resulting from an MSLB .i 4 (continued) BWR/6 STS B 3.4-37 Rev. O, 09/28/92 , _ __ _- _ __ - - - - - --- - a /In 9:47) RCS Specific Activity B 3.4.8 BASES APPLICA8LE outside containment during steady state operation, will not SAFETY ANALYSES exceed 10% of the dose guidelines of 10 CFR 100. (continued) The limits on specific activity are values from a parametric evaluation of typical site locations. These limits are conservative because the evaluation considered more restrictive parameters than for a specific site, such as the location of the site boundary and the meteorological conditions of the site. RCS specific activity satisfies Criterion 2 of the NRC Policy Statement. 1 LCO The specific iodine activity is limited to k0.2[yCi/gm h DOSE E(UIVALENT I-13Jf and th: ; ::: :;:: fic . i.; w w - 59 ~. M;ite i.e 100/E 4:/CThese !ieits :: -dttle source M l ' tem assumed in the safety analysis for the MSL8 is noQ exceeded, so any release of radioactivity to the environment , during an MSLB is less than a small fraction of the 10 CFR 100 limits. ( . ( 7 th limf answes s APPLICABILITY In MODE 1, and MODES 2 and 3 with any main steam line not isolated, limits on the primary coolant radioactivity are ! applicable since there is an escape path for release of radioactive material from the primary coolant to the i l environment containment. in the4event _of angge MSLB outside y of p[rim In MODES 2 and 3 with theM N A W , such limits do not applysinceanescapepatEdoesnotexist. In MODES 4 ' i and 5, no limits are required since the reactor is not pressurized and the potential for leakage is reduced. i ACTIONS A.1 and A.2 When the reactor coolant specific activity exceeds the LCO DOSE must beEQUIVALENT analyzed for DOSEI-131 limit, but isI-131 EQUIVALENT s 4.0atyCi/ p , t oncesamples eas every 4 hours. In addition, the specific activity must be restored to the LC0 limit within 48 hours. The Completion (continued) BWR/6 STS B 3.4-38 Rev. O, 09/28/92 2An' 9:twI] RCS Specific Activity 8 3.4.8 BASES ACTIONS A.1 and A.2 (continued) Time of once every 4 hours is based on the time needed to take and analyze a sample. The 48 hour Completion Time to Cl restore the actiirity level provides a reasonable time for temporar bursts) ytocoolant be cleaned activity up withincreases (iodine the normal spikes systems. processing or crud hGW pA f ' B.I. B.2.1. B.2.2.1. and B.2.2.2 If the DOSE EQUIVALENT I-131 cannot be restored to sk0.2 ! yCf /gm within 48 hours, or if at any time it is >'T4.0F Cf/ge, it must be determined at least every 4 hours and all the main steam lines must be isolated within 12 hours. fTsTTating tne NATirTYtan lines precludes the possibility o releasing radioactive material to the environment in an amount that is more than a small fraction of the ^ requirements of 10 CFR 100 during a postulated MSLB . accident. !. Alternately, the plant can be brought to M00E 3 within / i 12 hours and to MODE 4 within 36 hours. This option is / l provided for those instances when isolation of main steam y / lines is not desired (e.g., due to the decay heat loads). , ', ' g In MODE 4, the requirements of the LCO are no Ion er h I A applicable. - gg \ /) i The Completion Time of once every 4 hours is the time needed s to take and analyze a sample. The 12 hour Completion Time ! is reasonable, based on operating experience, to isolate the ,4 ' main steam lines in an orderly manner and without challenging plant systems. Also, the allowed Completion / d / D /l l Times for Required Actions B.2.2.1 and 8.2.2.2 for bringing the plant to MODES 3 and 4 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. l ,.-- - ,/ C.I. .2.1. and C.2.2
- f
/ ' / \ ,/ When the c'eactor cools t specifi/ activity is > 100/$' ' E l / yCi/gs ,Ill main steam lines aftgisolded withiry 12 hou s. w . ..y u m nm-..-....--.-----.-J l ~ ~- / f l (continued) l BWR/6 STS B 3.4-39 Rev. O, 09/28/92 (ZAK 93-ivE!) .a gi INSERT B39A feqv,*ee[/c-/r$,u o (04 /$4 A Note to the @ xcludes the MODE change restriction of 4 LCO 3.0.4. This exception allows entry into the applicable l MODE (S) while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of a limiting event while exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to, power operation. 1 1 F II O . INSERT RIVER BEND B 3.4-39 10/1/93 W) 04 93-/& RCS Spe.cific Activity B 3.4.8 BASES I ACTIONS .. ' k.1. C .1. and C.2.2 (contin Uf"!% .I. vernTilm wIvironment auMriran W l j . l ternate y, the pl$nt can b brought t N00E3within' 12 hour and to MODE 4 withi/i 36 hours. This option is / provid d for those instances when isol ion of main steam Q, ./ lines is not des' ired (e.g ~ due to th decay heat loads). In E 4, the requirene s of the L are no longer ap icable. , e 12 hour Completto Time is re onable, base. on ' / operating experience to isolate he main steaul lines / / without gfiallenging/ plant syst . Also, th allowed ' Comple n Times for Required tions C.2.1 nd C.2.2 are reaso le, based /on operatin experience, o. reach requ d plant 96nditions fr full power onditiorp,the in an or(e ly manner /and withou hallenging nt syst9ms. SURVEILLANCE SR bb.1 ,/ ad , REQUIREMENTS / This requires perf 9rmin g isotop analysis as a 3 measdre of the grossjtpeci i ctivity of he reactor co ant at least once per 7 ys. While asicallyfa antitative measue' of radi nuclides wi h half 1,fves longer '\ han 15 utes, cluding odines, th measur nt is the \ (~ G sus of t degastetgasma ctivities d the g ous3 W .' C 2 activit es in pe sample aken. Thi Surveil 14nce provides / od an in cation f any in ase in g s specif c activity ' : Tre ing t results this Sury 111ance a lows prope . dial action to b taken bef reachi the LCO 1 mit , / 's der nontal operat conditi s. The 7 day Frequ cy M consid the unli el hood of a gross f I failure uring g this s rt time f an y ~-- .SR 3.4.8 7 This Surveillance is performed to ensure odine remains within limit during nonnal operation. The @ day Freque cy is adequate to trend changes in the iodine activity leve , (continued) BWR/6 STS B 3.4-40 Rev. O, 09/28/92 {AM @HRI) RCS Sp2cific Activity I B 3.4.8 BASES SURVEILLANCE , ' SR 3.4.8 / (continued) REQUIREMENTS - d en:idering ; ::: specific activity is w,iitored eveD ' , ' hWJ - This SR is modified by a Note that requires this - /' Surveillance to be performed only in M00E 1 because the q, ,/ level of fission products generated in other MODES is much i jgt, less. ( , . . . - S 3.4.8.3 ond FA 3.4.9 4 J A radiochemic analysis for determination is quired { , g ;,j;c hv 9 + with the pla' operating in M00 1 with equi libri:== d;' conditions. T e G detemination frectly ' relates o the LCO and i required to veri plant operatio within he gross speci c activity LC0 1 it. The analys for is a measur of the average ergies per 1 di ntegration for sotopes with half ives longer th minutes, exc1 ing todines. ting experiene has i _, shown that G d s not change rap y and the Fre ncy of j 184 days rec nizes this. t \ // 4 g; t @ SR s been modified a Note that st es that j samplin,g'is required to performed with 31 days after a s minimum of 2 effective 11 power days d 20 days of MOD oper(tion have elaps since the reac r was last s C.E l sub' critical for at I ast 48 hours _.f his ensures the I radioactive materials are at equi,Mbrium so the ana) is for ! 7gggr ,/ G is representative and not skewed by a crud burst'or other / \ similar abnomal event. / / j %_m BD Q r / ' /- REFERENCES 1. 10 CFR 100.11o @ h 2.hAR,SectionC5.1.t'i - - BWR/6 STS B 3.4-41 Rev. O, 09/28/92 [M 92-/NQ [ , INSERT B41A- - - ~ - The a lysis Fra ncy is base n a start ti coinciden ith Ns obt ning a r ctor coola sample, an providing me t p pare the s ple, conduq/the measurem process, d ana e as neces ary'to properly termine E. / / >$9 1 INSERT l RIVER BEND B 3.4-41' 10/1/93 l L , _ ._. @ 9:wrQ RHR Snutdown Cooling Systen-Hat Shutdown B 3.4.9 BASES LCO associated piping and valves. Each shutdown coolinq (continued) subsystem is considered OPERA 8LE if it can be manua' ly , aligned (remote or local) in thu shutdown cooling mode for removal of decay heat. In MODE 3. one RNR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. ' Operation of one subsystem can maintain or ; reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly ; continuous operation is required. Note 1 permits both RNR shutdown cooling subsystems and l recirculation p be shut down for a seriod of 2 hours in an 8 hour peri a;. ;;;; ;.. 4;,;;;;; M . _-fnGL Note 2 allows one shuteown coo <ng sunsystem to ne C.) laoperable for uo to 2 hours ace of surveillance test 'm. _ _~ L ese s may ystem or on some other lant system or component that necessitates placing the stem in an inoperable status during the ; performance. In is pemitted because the core heat I generation can be low enough and the heatup rate slow enough l i to allow some changes to the RNR subsystems or other operations requiring RNR flow interruption and loss of redundancy. APPLICA81LITY In MOD 2, and in MODE 3 with reactor steam does r ifressu he mm cut in permissive pressure, this LC0 is not a able. Operation of the RHR System in the fu4'Yf"" y e8u / .4 shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Oecay heat removal at reactor ~ pressureT%EEPtheRNRcutinpermissivepressureis g,4 typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 below this pressure, the 4 j7q OPERASILITY requirements for the Emergency Core Cooling Systems (ECCS) (LCD 3.5.1, 'ECCS-Operating") do not allow OC6 placing the 4ew erena m e RHR shutdown cooling subsystem into operation. In MODE 3 with reactor steam done pressure below the RNR cut in permissive pressure (i.e., the actual pressure at which the interlock resets) the RHR System may be operated in the (continued) BWR/6 STS 8 3.4 43 Rev. 0, 09/28/92 ! (i4R 93 anD RHR Shutdown Cooling System-Het Shutdown 8 3.4.9 , BASES APPLICA8!LITY shutdown cooling mode to remove decay heat to reduce or (continued) eJsiain coolant temperature A er~oe ,a reo,,.,/, M p ; fu regge# 1o cc i,1 opera lim . ,4 The requirements for decay heat removal in H00ES 4 and 5 are j 79 Shutdown Cooling System-Cold Shutdown *; LC0 3.9 d ! OG4 " Residual Heat Removal (RNR)-High Water Level"; and l LC0 3.9g. , ' Residual Heat Removal (RHR)-Low Water Level." J W h v- , l g C)(c efY Q $ fer~wo10cd ACTIONS A.I. A.2. and A.3 b34,c o go7g 2, -iWith one SFWWElh required RHR shutdown cooling subsystem ( _TA)S E RT inoperable for decay heat removal, inoperable 0W'lsubsystem@mustberestoredto ERA 8LE status without r SWA delay. In this condition, the remaining OPERA 8LE subsystem / can provide the necessa decay heat removal. The overall reliability is reduced, ver, because a single failur^e in Oc7 the OPERA 8LE subsystem could result in reduced RHR shutdown cooling capability. 7 ubsystems C5 bWith iaarer of the the requi the =two %iaias NR a = shutdown is cae*61a o co Mo heatremova.u=. a r_ + h overL11 uias L reliability is reduced s inerefore an alternate mettod o ',,(Tecap heat remova' must be provided.fifith both RHR shutdown Sd - coor ng subsystems Inoperanie, an alternate method of decay gd heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO. The 1 hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. (DEf The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) 4,f its capability to maintain or reduce temperature. Dec 'q _ - "-*m val h_v ambient losses can be considered a # _/ the alternate method capability. A rnate met o< can be used include (but are not limited to) the8 System or the Reactor Water Clean $ up System. Q Aal God DW pp (continued) BWR/6 STS g 3.4 44 Rev. 0, 09/28/92 - - - - - - - - - - - - - - - _ - - - - - - - - - - - - - ~ . . - - - - - - - , - - - , - , , . - - - - - - - (2AR 93-HPTb RHR Shutdown Cooling System-Hot Shutdown l 8 3.4.9 8ASES ACTIONS A.1. A.2. and A.3 (continued) However, due to the potentially reduced reliability of the alternate methods of decay heat removal, it is also required to reduce the reactor coolant temperature to the point where MODE 4 is entered. . B.1.B.2.andB.} y With no RHR shutdown cooling subsystem and no recirculation pump in operation, except as is Mrsitted by @ LC0 Note, reactor coolant circulation by t1e RHR shutdown cooling subsystem or one recirculati,on pump must be restored without delay. l Until RHR or recirculation pump operation is re-established, ! an alternate method of reactor coolant circulation must be placed into service. This will provide the necessary l circulation for monitoring coolant temperaturg The 1 hour Completion Time is based on the coolant circulation function and is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation. Furthermore, verification of the functioning of the alternate method must be reconfinned every 12 hours thereafter. This will provide assurance of continued temperature monitoring capability. / or During the period when the reactor coolant is being recircu b ---- circulated by an alternato method (other than by the ' requiredRMRJhutdown star), the reactor coolant h, pq temperatureandpressu$6o'ing re must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is appropriate. i ub i SURVEILLANCE 1R 3.4.9.1 : REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours is sufficient in view of other visual and audible indications i (continued) BWR/6 STS 8 3.4-45 Rev. O, 09/28/92 i , . _ . . - - ~- - " ' ' (L4K 93-/YD ' ' RHR Shutdown Cooling Systes-Cold shutdown 8 3.4.10 BASES LC0 piping and valves. Each shutdown cooling subsystem is (continued) considered OPERA 8LE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. In H00E 4 one RNR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. Operation of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. Note 1 pemits both RHR shutdown cooling subsystems and recirculation pumps _ to be shut down for a period of 2 hours in an 8 hour periogr_r --- :-: -; = r :: -.fr.g ". Note 2 allows one RRR snut h coo ing suosvassa so se . inoperable for upA t 2 hours for-der - perfomance of l surveillance tes(C--+ - m - r -- a r % i 6 ThH e tests any be on the affected M system or on some other plant system or component that necessitates ; placing the RHR System in an i rable status during the performance. This is pemitted ause the core heat ! generation can be low enough and the heatup rate slow enough ! to allow some changes to the RHR subsystems or other l operations requiring RNR flow interruption and loss of redundancy. APPLICABILITY In ICDES 1 and 2, and in MODE 3 with reactor steam done pressuNthe RHR cut in pemissive pressure, this LCO is not applicable. Operation of the RNR System in the (/"d g" g"d ~ shutdown cooling mode is not allowed above this pressure i or egul -/o because the RCS ressure may exceed the design pressure of the shutdown ing piping. Decay heat removal at reactor , pressures he RNR Cut in pemissive pressure is l'l typically accomplished by condensing the steam in the main a 79 condenser. Additionally, in MODE 2 below this pressure, the OptRASILITY requirements for the Emergency Core Cooling Systems (ECCS) (LC0 3.5.1, 'ECC5-Operating") do not aliow placing the :p ==- :RNR shutdown cooling subsystem into operat:on. In MODE 4, the RHR System any be operated in the shutdown Ml ~' cooling mode to remove decay,.hapit to maintain coolant temperature below 200*F. f o dec , a e,,,,,,f /,4v # 7 'l y ; 9,48 do be A cyedok ' (continued) SWR /6 STS B 3.4 48 Rev. 0, 09/28/92 0/I97-S ) RHR Shutdown Cosling System-Cold Shutdown 8 3.4.10 BASES (m 74 9 i APPLICA8ILITY The requirements for decay heat removal in MODE 3 below the (continued) cut in permissive pressure and in MODE 5 are discussed in ~ LCO 3.4.), " Residual Heat Removal System-Hot Shutdown'; LC0 3.9.& '(Residual Heat RemovalRHR (RHR)-Hi Water Level"; and LCO 3.9.$, " Residual Heat Removal (g R)-Low Water Level.' g ACTIONS j# L1 ef k3 W M*h h jggg With one of he two required RHR shutdown cooling subsystems inoperabl , the remaining subsystem is capable of providing !$49A L the requi d decay heat removal. However, the overall reliability is reduced. Therefore, an alternate method of decay heat removal must be provided. With both RHR shutdown 6 cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LC0. The I hour Completion Time is based on the decay heat removal function and the probability of a loss of the
- 7. 4 available decay heat removal capabilities. Furthermore, g/ 9 verification of the functional availability of these alternate method (s) must be reconfirmed every 24 hours i
thereafter. This will provide assurance of continued heat removal capability. ' f16 T The required coolin capacity of the alternate method should be ensured by verif ng (by calculation or demonstration) its capability to ma ntain or reduce temperature. Dec heat - val tent losses can be considered a oe c---'- m'-- - he alternate method capability. e e ! methods tnat c be used include (but are not limited to) the System or the Reactor Water Cleanup System. Cg 1 ~ B.1 and B.2 , With no RHR shutdown cooling subsystem and no recirculation h pump in operation, except as is pemitted by 49 LCO Not and until RHR or recirculation pump operation is re-established, an alternate method of reactor coolant circulation must be placed into service. This will provide (continued) BWR/6 STS B 3.4-49 Rev. O, 09/28/92 - . , , , - n - - , v-- ,- +-, 6 AR 93
- fD RCS P/T Licits
, B 3.4.11 B 3.4 REACTOR C00UWT SYSTEM (RCS) B 3.4.11 ACS Pressure and Temperature (P/T) Limits BASES i BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to systes pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and < - reactor trips. This LC0 limits the pressure and temperature f'f ## changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation. l l q GEs Mi3%ontains P/T limit curves for heatus, cooldown, and
- f -
f!t N!f f !b e heatup curve provides limits for both heatup and i criticality. Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to detersins that operation is within the allowable region. The LCO establishes opersting limits that provide a margin to brittle failure of the reactor vessel and pipir.g of the reactor coolant pressure boundary (RCPS). The vessel is the , component most subject to brittle failure. Therefore, the LC0 limits apply mainly to the vessel. , 10 CFR 50, Appendix G (Ref. 1), requires the establishment-of P/T limits for material fracture toughness rsquirements of the RCPS asterials. Reference 1 requires an adequate margin to brittle failure during normal operation, ant <cipated operational occurrences, and system hydrostatic tests. It aandates the use of the American Society of Mechanical Engineers (ASME) Code, Section !!I, Appendix G (Ref.2). The actual shift in the RT of the vessel material will be established periodically b y removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and 10 CFR 50, Appendix H
- (Ref.4). The operating P/T limit curves will be adjusted, (continued)
BWR/6 STS B 3.4-51 Rev. 0,09/28/92 , , , _ , . . , , ,.....r +-,----m,- 4 (L f 93-/YRI ) RCS P/T Licits 8 3.4.11 , BASES APPLICABLE are acceptance limits themselves since they preclude SAFETY AMALYSES operation in an unanalyzed condition. (continued) RCS P/T limits satisfy Criterion 2 of the NRC Policy ' Statement. r LC0 The elements of this LC0 are: q'y a. RCS pressure, temperature, and heatus or coo 1N rate c3 . a+hin eh. 11 i+= --=== a '- *' . = "'durme WC5 l 12 7 Qes tup , coel de.ve ,a- e wseau- m am *Q~ste s'c m tmy. ' \s , b. The tem prature difference between the reactor vessel - bottom wad coolant ant #M reactor pressure vessel (RPV) coolant is within as limit af - "n"Murina pThem w e.se w.gqgyy m d c temperature difference M n sne reacto coo ant , ' in the respective recirculation _ loop and in the ' reactor vessel meets the limit -- --- ---- during pump j l
- d. RCS pressure and temperature a C3 within the_
Q reR z-l g pv limits c: x'?"- -
- 2: n"nphee/oaclee*[l i
- e. The reactor vessel flange and the head flange [C3 '
temperatures are within @Pfinits d = - G whW ' Mk,,q reactor vessel head bolting studs .t !"-sirrE "- i These limits define allowable operating regions and pemit a large number of operating cycles while also providing a wide l margin to nonductile fai' ure. j The rate of change of temperature limits control the thermal gradient through the vessel well and are used as inputs for calculating the heatup, cooldown, and inservice leak and hydrostatic testing P/T limit curves. Thus, the LC0 for the
- rate of change of temperature restricts stresses caused by j thermal gradients and also ensures the validity of the P/T l Ilmit curves.
, Violation of the limits places the reactor vessel outside of , ! the bounds of the stress analyses and can increase stresses ! I in other RCS components. The consequences depend on several factors, as follows: i (continued) BWR/6 STS , B 3.4-53 Rev. 0, 09/28/92 QM 93 IYEQ RCS P/T Liaits B 3.4.11 BASES ACTIONS C.1 and C.2 - (continued) Operation outside the P/T limits in other than M00ES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPS is returned to a condition that has been verified by stress analyses. The Required Action must be ; initiated without delay and continued until the limits are ! restored. Besides restoring the P/T limit parameters to within limits, an evaluation is required to detemine if RCS operation is allowed. This evaluation must verffy that the RCPS integrity is acceptable and must be completed before ! approaching criticality or heating up to > 200*F. Several methods may be used, including comparison with pre-analyzed ; transients, new analyses, or. inspection of the components. ; ASME Section XI, Appendix E (Ref. 6), may be used to support I the evaluation; however, its use is restricted to evaluation of the beltline. . 1 SURVEILLANCE SR 3.4.11.1 , REQUIREMENTS ' fl Verification that operation is within limits is f required every 30 minutes when RCS pressure and temperature f conditions are undergoing planned changes. This Frequency # is considered reasonable in view of the control room ' indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments 30 minutes pemits assessment and correction of 1 g minor deviations. D , Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria pff f given in the relevant plant procedure for ending the g activity are satisfied. ggs This SR has been modified by a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and inservice leakage and hydrostatic testing. (continued) 8WR/6 STS 8 3.4-56 Rev. 0, 09/28/92 Q4 93-/v// ] RCS P/T Licits B 3.4.11 BASES SURVEILLANCE E 3.4.11.2 - REQUIREMENTS (continued) A separate limit is used when the reactor is approaching -f4/4 R e, a; c et
- criticality. Consequently, the RCS pressure and temperature r".M';8 must be verified within the appro riate limits before A4% withdrawing control rods that wil aske the reactor d / r y u,< o4 f *M critical .
' Js<re. //ne s ,,s,, to b c/ yd '"I m/ / ' Performing the Surveillance within 15 minutes before control "" h'"v,,7f ,4, , /,, rod withdrawal for the purpose of achieving criticality i provides adequate assurance that the limits will not be ; />"'?"5' ? # '" 4 " ")' exceeded between the time of the Surveillance and the time ot'6 < /d;/. of the control rod withdrawal. t l N 3R 3.4.11.3 and SR 3.4.11.4 l '/ Differential temperatures within the applicable GEEdimits i ensure that thermal stresses resulting from the startup of i an idle recirculation pump will not exceed design 1 allowances. In addition, compliance with these limits ensures that the assigtions of the analysis for the startup of an idle recirculation loop (Ref. 8) are satisfied. -3. 'I # 7f Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate - 'e assurance that the Ifmits will not be exceeded between the ' f time of the Surveillance and the time of the idle pump , . a ,,.. <ec,,e -,4on /- start. 3 /"d I" I pay,f.G al. , R ? An /Macceptabledifferential temperature means of demonstrating compliance with the requirement in SR 3.4.11.4 is to \ g ,,,4 , ,, /
- compare the temperatures of the operating recirculation loop
\ \ a,,, ,.7s<e[/,o, and the idle loop. su~ s s SR 3.4.11.3,Wheen modified by a Note that requires the ] Surveillance to be met only in MODES 1 2, 3, and 4$5E9- \ reacter steam done pressure a 25 psig . In MODE 5, the , overaII stress on limiting components is lower; therefore, g,,/c2c,y.//.y' AT limits are not required. {hm SR 3.4.11.5. SR 3.4.11.6. and SR 3.4.11.7 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits (continued) r BWR/6 STS B 3.4-57 Rev. O, 09/28/92 (LM 93-/YRT) RCS P/T Limits B 3.4.11 SASES SURVEILLANCE SA 3.4.11.5. SR 3.4.11.6. and SR 3.4.11.7 REQUIREMENTS (continued) during. system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in M00E 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LC0 limits. The flange temperatures must be verified to be above the limits 30 minutes before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature s 80*F, 30 air.ute checks of the flange temperatures are required because of the reduced margin to the limits. When in Mc0E 4 with RCS temperature s 100*F, .h, ./ monitoring of the flange temperature is required every -U 12 hours to ensure the temperatures are within limits (fpss***sFHANFFFEip, . p2 The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour - - - -t l _TASEflT Frequency is reasonable based on the rate of temperature gggg change possible at these temperatures. REFERENCES 1. 10 CFR 50, Appendix 6.
- 2. ASME, Boiler and pressure Vessel Code, Section !!I, stocN[bM(aJseN
- 3. ASTME185-82,IJuly1982.
Smeili ace Year & Lih4-id "
- 4. 10 CFR 50, Appendix H. c i.a u , , a + y g es.
- 5. Regulatory Guide 1.99, Revision 2. May 1988.
- 6. r s_ure Ytsnel Code. Section 11. -
ASME, Boile' andy"bki,d f+euvre Riset A4 Appendix E. - - f % h u s Re w er w h 4 e 8 $ s,
- 7. NED0-21778-A,fDecember Isis.
- 8. . Section NEEEED.
BWR/6 STS 8 3.4-58 Rev. O, 09/28/92 vwr , , - - - - - (TAC 93-HRD i INSERT B58A SR 3.4.11.8 and SR 3.4.11.9 Gd ./ 72 Differential temperatures within the applicable @ limits ensure that thermal stresses resulting from increases in THERMAL POWER or recirculation loop flow during single recirculation loop operation will not exceed design allowances. Performing j the Surveillance within 15 minutes before beginning such an ) increase in power or flow rate provides adequate assurance that i the limits will not be exceeded between the time of the Surveillance and the time of the change in operation. An acceptable means of demonstrating compliance with the l temperature differential requirement in SR 3.4.11.9 is- to l compare the temperatures of the operating recirculation loop and l the idle loop. { d e ,, . Plant specific test data has determined that the bottom head is l i pga not subject to temperature stratification with natural powet er circulation at power levels as low as 36% of RTP or with any hep //#" single loop flow rate when the recirculation pump is on high , ' speed operation. Therefore, SR 3.4.11.8 and SR 3.4.11.9 have ; g g ), ' j, % < been modified by a Note that requires the Surveillance to be met "f## 4 only' h conditions The Note for SR 3.4.11.9 further : limits the requirement for this Surveillance to exclude comparison of the idle loop temperature if the idle loop is x isolated from the RPV since the water in the loop can not be 3.4 introduced into the remainder f the reactor coolant system. ' ~- - - - - . , , , , , #c. nd ed) } ^ / n _. - i sn 3. 4.11.1V ,,pl6,[lAp}.,#e) / Verifi ion t operation withi[ PTLR h is essure nd tempdrature c quired ditions j 6 eve 30 minu a when RCS ar undergo g planned c nges. is Freqy'ency is naidered dn e control oom in$. cation av ilable to t asonable n view of of chang i - nitor CS status. .lso , sin tempez;4ture rat limits are specifi in hourly neremeryt's , 30 min tes permi s \ asse nt and co action of nor dev stions. rveillance f inservice akage a d hydrost ic testin may / l I l 'be discontin d when the iteria van in t relevant lant f ' procedure r ending the etivity re satief d. / This S has been ified b a Note t requi s thi's I Surve ance to be formed o y during neervice 1 ge and j hydr tatic testin . , - L f i INSERT RIVER BEND B 3.4-58 10/1/93 ._, - u ~ -nA - - ei e 4 - A. (AR 93-/YMi] ATTACHMENT 2B !TS - PSTS COMPARISON DOCUMENT . REVISION 1 DISCUSSION OF CHANGES I J i y ,. - .., - , - - , --- ,, . , , . . , _ . _____ _ _.___.__ _ __ _ _ _______ _ _ _____ ___ ___ _________ _ (dAf 99 N)[]D DISCUSSION OF CHANGES TO NUREG-1434 TS 3.4.8 - RCS SPECIFIC ACTIVITY BRACKETED ADMINISTRATIVE CHOICE B.1 Brackets removed and optional wording preferences revised to reflect appropriate plant specific requirements. PLANT SPECIFIC DIFFERENCE P.1 The safety analysis report for this station is identified as the Updated Safety Analysis Report and is correctly referred to as the USAR. P.2 The date on this reference 1:o 10 CFR 100 is deleted since the limits are based on the current regulations. P.3 This comment number is not umV. for this station. CHANGE / IMPROVEMENT TO NUREG STS ; C.1 A Note is added to the Required Actions for Condition A to indicate that LCO 3.0.4 is not applicable. Entry into the Applicable Modes should not be restricted since the most likely i' response to the condition is restoration of compliance within the allowed 48 hours. Further, since the LCO limits assure the . dose due to a IDCA would be a small fraction of the 10 CFR 100 l limit, operation during the allowed time frame would not j represent a significant impact to the health and safety of the r p ic. g , ., f g g g. ,g , ! C.2 CE proposeo to provias clasislatie . Q:t it'l h*= m'WC - & _=uv. . ... ..,. .. im m . . m . m eoo.. - ; C.3 Change proposed to provide consistent wording with the 140. 7 [ C.4 % orae contro to provice ditio al cl ifica on re rding the of ra'dioac ve a erial duri a ulate LB g ,'l facci nt, the seco sen ce o ara ph is place ith sp'p --i la sentence c led rom a f st p agrap of Bj. ~ I scussi6n y Tbh com.,ed nundec is nofviek h b ' ' RIVER BEND 11 10/1/93 a . . . _ -- .__ _ _. . ___ _ _ _ _ _ _ [4N( 93-/W/) l DISCUSSION OF CHANGES TO NUREG-1434 TS 3.4.8 - RCS SPECIFIC ACTIVITY CHANGE / IMPROVEMENT TO NUREG STS h, o,.mwf roue be h A!Of V3 IJ YoE C.5 /The analysi time f comple ng a determination o Ei currently interpre d to 184 day from last ion. Ho ever, s may easi be exces d if an ldeterni extelid outage w a to oce . The cur t Note
- lies that
.- Y E de ruinatio are no required til after sample is /s@ ta n. Sinc E date ination a empts dur non-MODE 1-o ration pr ide inde arminate r uits, the ont is clearl ,not a simp calenda type Frequp6cy. There re, the propo d sampling equency a ave g 18Vdays with Note that al sa delay r sampl g then cent oper ion has n been approp ate to ovide e r inble scep . sinca sampling i ' the i itial'p nt from w ch an anal ~ is can be ne, ao tin / Freque y for sa .:9mple i also pro sed. s Frgquency it based the nues ry time perfo the gnalysis orisite or a an offsite Yaboratory. / C.6 This change provides consistency with the wording of other Bases for similar type Required Actions. t RIVER BEND 12 10/1/93 &!' 93-/YRI ) DISCUSSION OF CHANGES TO NUREG-1434 TS 3.4.11 - RCS PRESSURE AND TEMPERATURE LIMITS BRACKETED AI3tINISTRATIVE CHOICE B.1 Brackets removed and optional wording preferences revised to reflect appropriate plant specific requirements. 1 PIANT SPECIFIC DIFFERENCE P.1 The safety analysis report for this station is identified as the Updated Safety Analysis Report and is correctly referred to as the USAR. P.2 Additional Surveillance Requirements are included to incorporate current plant specific verifications of single loop operation limits. P.3 Additional information is included in the references for ease of identification. This additional information may include the title, revision number and/or date. P.4 This comment number is not used for this station. CHANGE / IMPROVEMENT TO NUREG STS . C.1 These changes made to the LCO elements to include when each . element applies. C.2 Editorial information is deleted since it is not required information and it is not consistent with the format and ' Q content of other proposed Bases. en -- c.3 C.,- s e m.J e. A b com ae,,4 d a A.,g., y,.opos,J ! in Sechea 5.o . ) l 4 /2o /14 RIVER BEND 17 10/1/01 6 e 95 w ri) ; i l l l . i RIVER BEND SECTION 3.5 l l ll l i ATTACHMENT 1: CTS-PSTS COMPARISON DOCUMENT l ATTACHMENT 2:ITS-PSTS COMPARISON DOCUMENT l 6AR 93wQ , i ATTACHMENT 1 ITS - PSTS . l COMPARISON DOCUMENT i l REVISION 1 SECTION 3.5 REVISED PAGES l 1 A: MARKUP OF CTS 1B: DISCUSSION OF CHANGES
- 1C
- NO SIGNIFICANT HAZARDS CONSIDERATIONS
L L E e 3 7 Fi r) l i i I ATTACHMENT 1 A 1 CTS - PSTS i COMPARISON DOCUMENT REVISION 1 MARKUP OF CTS l l l l , i h 93 /dh EMERGENCY CORE OCOLING SYSTIMS 3/4.5.2 ECCS - IMUTOOWN - T'S 1 E.7 LINITING CON 0!T:0N FOR OPERATION G.S __---. 3.5.2 At leas of the following sna11 be OPERA 8LE: [4. l of taking suction from the suppression pool a water through the spray sparger to the reactor vessel. l q b .' Low-pressure coolant injection (LPCI) subsystem "A" of the RHR ! sion pool and transferring the water to the reactor v c. gg Low pressure coolant injection (LPCI) subsystem "B" of the RHR sion pool and transferring the water to the reactor v d. Low pressure coolant injection (LPCI) subsystem "C" of the RHR , sion pool and transferring the water to the reactor v .e. The high pressure core spray (HPCS) system with a flow path capabl ' of taking suction from the condensate storage tank or suppression pool, as applicable, when these sources of water are OPERA 8LE per ( Specification sparger to the reactor 3.5.3.b,vessel. and transferring the water through the spra j App (ICA8ILITY: OPERATIONAL CONDITION 4 and 5*. ACTION: t ! 4. cwa a"With one of the above required subsystems / systems inoceraole, r at least two subsystems / systems to OPERA 8LE status within 4 hours or Cbu G suspend all operations that have a potential for draining the reactor Ivessel. b. 'With U -~ both of the above required subsystems / systems inoperable, surccnd "1**.",".*' :: cwoc the reactor vassel. 4 ja11 operations that have a potential for draining Restore at least one suesystan/ sveta= ta opfitARLE eso O statue within 4 the hours BANDLINujwithin nextor3 establish aours ) (PRIMARY. CONTAINMENT INT < e 4C0JR.'*The ECCS is not required to be OPERABLE provided that the reactor vess # g is removed, the cavity is flooded, the upper containment fuel pool gate is opened, and water level is maintained within the limits of Specifications 3.9.P nd 3.5.Y3 c-RIVER BENO - UNIT 1 3/4 5-6 - m (ur 93&D EMERGENCY CORE COOLING SYSTEMS 9 : :.: SUPPRESSION P0OL /TS352 . .. pINITINGCONDITIONFOROPERATION \ b 3 b 2-,1 ( 4EEb The suppression pool shall be OPERA 8LE: Novrolo a. J.c . z.2 In OPERATIONAL CONDITION 1, 2 and 3 with a contained water volume of g at least 137,571 ft3, equivalent to a level of 19'6". , S as.2J b. ) 50.5.2/2 fin OPERATIONAL CONDITION 4 and 5" with a contained water vo J_t least 94,000 fts, equivalent to a level of 13'3".rexcept tnat /co a.s.2 the suppression drained providedpool that:level may De less than the Ifmit or may be Owe Bf 1. cwoc No operations are performed that have a potential for draining the reactor vessel, w f,p; 2. The reactor mode switch is "::MCin the Shutdown or Refuel position,
- 3. The condensate storage tank contain at least 125,000 avail-g able gallons of water,(equivalent to a level of 11'1"3 and !
il; 3R J.r.2 2 4. The HPCS system is OPERA 8LE per Specification 3.5.2fwith an ! (OPERA 8LE flow path capable of taking suction from the conden-] LA2 i sate storage tank and transferring the water through the spray; i tsparger to the reactor vessel. 1 (d ( c. Witn wo OPERA LE suppres on pool p ack syst eac consisti g of #: arrW suo stems g9 I --
- 1. At lea one OPE .E crescen area sump and u
An OP RABLE flow cath to th suppressio cool. a APPLICABILITY: OPER4TIONAL CONDITIONS (1, 2, 3, 5" god % ACTION: .2.Q
- a. In OPERATIONAL CONDITION 1, 2 or 3 with the suppression pool water MD2 J. '.. 2. 2 level less than the above limit, restore the water level to within g the limit within 1 hour or be in at least HOT SHUTOOWN within the next 12 hours and in COLD SHUT within the following 24 hours.
- b. In OPERATIONAL CONDITION 4 o th the suppression pool water level im M less than the above limit or drained and the above required conditions not satisfied, suspend O^*: u ! OT!T: u, Gell operatiaan that have g.)cwo g ,
a potential for drainina the reactor vessei and C gg ock tm reactor =edsh switch INTEGRnY <ntheStutdown>osition2EstablishrRmRYCDEAIMMENTg - FUE. HANDLI Mift1 thin 8 hours. 4co "The suppression pool is not required to be OPERA 8LE in OPERATIONAL CONDITION 5 M.2 provided that the reactor vessel head is removed, the cavity is flooded, the t , Appg, upper containment fuel pool gate is open, and the water level is maintained within the limits cf Specifications 3.9.8 snd 3.9. 2 F#ThoAPP5 A not/requiped tWDe OPERABLEAhen pne suppressuur pony isjno@ J _reA01 redt/o bg(OPERAM.E./' ' RECEIVEo RIVER BEND - UNIT 1 3/4 5-8 Amendment No. 38 AL 191989, . 7?" 2-.- - . ,. _ _ . , _ , - . , - , , _. .~,.-m f (L M 93- NA).) #11 EMERGENCY CORE COOLING SYSTEMS , ACTION (Continued) j , 'c . With o SPPS subsystem noperable, restore he SPPS subsystem to OPE E status withi 31 dcys or demonst to the OPERA 8ILITY of he rem ning SPPS subs tem at least once r 31 days by: t 1 A functional at of the crescen area sump pump, and
- 2. Demonstr ng that the associ ed flow path can be a gned to the sup assion pool.
- d. I With both PPS subsystems ino rable, restore one SPP subsystem to OPERA 8L status within 7 de or:
- 1. n OPERATIONAL COND ION 1, 2, or 3 be in least HOT SHUTDOWN..J2 within the next ours and COLD SHUT within the following 24 hours.
- 2. In OPERATIO CONDITION 4 or 5* pr ide at least one altern REC __ : , , E D pumpback me od and demonstrate th OPERA 8ILITY of an alter te method wi in 24 hours and at le t once per 24 hours thereaft , otherwise suspend E ALTERATIONS and all Sgp l in0 operat ns that have a potent I for draining the re r vesse and lock the reactor de switch in the shu S OC pos ion. Establish PRI CONTAINMENT INTEGRI - FUEL u H LING within 8 hours.
SURVEILLANCE REQUIREMENTS 4.5.3.1 The suppression pool shall be determined OPERABLE by verifying the water level to be greater than or equal to, as applicable: .h 19'6", at least once per 24 hours, in OPERATIONAL CONDITION 1, 2 and 3. 3R J.6.2 : b. 13'3", at least once per 12 hours, in OPERATIONAL CONDITION 4 and 5. 3FS5.22 4.5.3.2 With the suppression pool level less than the above limit or drained in OPERATIONAL CONDITION 4 or 58, at least once per 12 hours: L@ 3,f JI2 a Verify the required conditions of Specification 3.5.3.b to be /// i h . satisfied, or
- b. Veri footnote conditions
- to be satisfied.
1.5.3.3 At least once er 92 days, t SPPS shall b emonstrated CPE E by
- a. Verifyin och crescent sa sump pump velops 50gpe, a
- s. >
b.' Verif ng the flow p can be align to the suppress n pool. A *The suppre on pool is no required to b PERABLE in OPE IIDHAL CONDIT 5 provided at the reacto vessel head is emoved, the cav sy is flooded the upper c ainment fuel ol gate is op , and the water evel is sai ined within he limits of ecifications 3 .8 and 3.9.9.. RIVER bet:0 - UNIT 1 3/4 5-9 Amendment No. 3W, 47 WR43-tyk ATTACHMENT 1B CTS - PSTS COMPARISON DOCUMENT REVISION 1 DISCUSSION OF CHANGES ~ Q R 93-/ W D ; DISCUSSION OF CHANGES CTS: 3.5.2 - ECCS SHUTDOWN TECHNICAL CHANGE - LESS RESTRICTIVE (continued) L.3 A Note. clarifying the alignment requirements of a LPCI subsystem has been included for proposed SR 3.5.2.4. The note is similar to the existing footnote ## to LCO 3.5.1 except that the proposed note allows operation of one of the RHR subsystems in the shutdown cooling mode during MODE 4 and 5, if necessary, g.I and clarifies that the subsystem is still considered OPERABLE jf for the LPCI mode. Because manual valve positioning removes ( ' the capability of the subsystems to respond automatically, the subsystem would be considered inoperable without this note. Although no specific analysis of this condition has been performed, the allowance provided by the note is acceptable because the return to OPERABILITY entails only the repositioning of valves, either remote or locally, and the energy. requiring dissipation in MODE 4 and 5, is considerably less than that at 100% power with normal operating temperature and pressure. Further, because of the low probability of an event requiring an ECCS and the certain need ' for shutdown cooling, it is considered appropriate to have a subsystem aligned for the decay heat removal. 4 I l 4 mlh/9y RIVER BEND # 67 Q. 6 j l (MR ' 93 h'f 0 DISCUSSION OF CHANGES CTS: 3.5.3 - SUPPRESSION POOL 1$ J// RELOC&TEll_J2ECIELCATIQHS k' M The suppression pool pumpback system provides control of post- 5 LOCA leakage from the ECCS in the crescent area of- the j dA- auxiliary building by returning the leakage to the suppression f pool. However, the evaluation (No. . 377) summarized in NEDO-31466, Supplement 1, determined that this system is not required to mitigate the consequences of any transient or design basis accident. Therefore, the requirements specified for this function did not satisfy the NRC Interim Policy Statement technical specification screening criteria as documented in the Application of Selection Criteria to the RBS TS and the requirements have been relocated to plant documents controlled in accordance with 10 CFR 50.59. ~gp ue, ik +/th sedib" o TECHNICAL CHANGE - MORE RESTRICTIVE I None in this section. TECHNICAL CHANGE - LESS RESTRICTIVE " Generic" LA.1 Movement of the reactor mode switch from the shutdown or refuel position is adequately controlled by the MODES definition table (proposed Table 1.1-1), and the requirement to " lock" the mode switch is adequately controlled by plant procedures. 4 LA.2 The details relating to system design and purpose have been relocated to the Bases. The design features and system operation are also described in the USAR. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. l LA.3 This comment number is not used for this station. 1 LC.1 This comment number is not used for this station. i " Specific" l l L.1 The requirement for suspension of CORE ALTERATIONS with no ECCS l available due to a degraded water source is deleted. Refueling ; LCOs provide requirements to ensure safe operation during CORE , ALTERATIONS including required water level above the RPV l flange. The ECCS function provides additional protection for loss of vessel inventory events. However, these events are not f initiated by, nor is the response of ECCS hampered by, CORE ALTERATION operations. RIVER BEND 8 rNu <1"71/ T l g 4 _ -. 4-.-- B 1- J_ . 2- - p a. u aa., 9 ( .1x es-mi) 4 ATTACHMENT 1C CTS - PSTS COMPARISON DOCUMENT REVISION 1 NO SIGNIFICANT HAZARDS CONSIDERATIONS l l l (w swiD 's /3.b jf NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.5.2 - ECCS SHUTDOWN "L3" CHANGE \ ' ~ Entergy Operations Inc., has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
Previous analysis indicates that a LPCI line break may initiate an accident. However, since LPCI is currently analyzed for the requested mode of operation, this change will not affect the probability of such an event. This equipment's role is in mitigating and thereby limiting consequences of analyzed events. The LPCI equipment required for OPERABILITY is only interrupted by valve alignment and is still capable of being manually realigned if the LPCI subsystem is needed to mitigate the consequences of design basis accidents. In addition, the proposed Note is applicable when the reactor is shutdown in MODE 4 or 5. Thus, the reactor heat load is much less than in MODE 1 (the MODE assumed in the accident analysis) , and another subsystem of the ECCS are still required to be OPERABLE. These changes are consistent with the philosophy stated in the proposed BWR Standard Technical Specifications. Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change introduces no new mode of plant operation and it does not involve physical modification to the plant. Therefore it does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed change does not reduce a margin of safety because the change has no impact on any safety analysis assumption. The clarifying Note recognizes the conflict in the dual purpose system and allows the decay heat removal function to proceed in MODE 4 and 5. Further, it recognizes that the amount of time to realign the LPCI system f rom the decay heat removal function has no significant impact on the margin of safety because heat loads under these conditions are far below that assumed in the safety analysis. Because the Note allows decay heat removal to /0//(s/N RIVER BEND M YA - 10 /1/9-3c"" 6 $N 93-dMy q.5 NO SIGNIFICANT HAZARDS CONSIDERATIONS /t CTS: 3.5.2 - ECCS SHUTDOWN \ "L3" CHANGE (continued) continue, the movement of the plant towards increased safety conditions and reduced energy levels is unimpeded. 4 l to shr RIVER BEND I f$ ' n '9 3 - I 1 (les g.3-nao ; l l l ) l l l i ATTACHMENT 2 , I l l l ITS - PSTS ; COMPARISON DOCUMENT REVISION 1 1 l SECTION 3.5 REVISED PAGES 2A: MARKUP OF ITS 2B: DISCUSSION OF CHANGES . I' ,l 4 . _ - - - - "3 - * - l I ATTACHMENT 2A ITS - PSTS COMPARISON DOCUMENT REVISION 1 MARKUP OF ITS I 4 a m , - - - . - - , - ,. I i $ 93-NfD ECCS-Operating 3.5.1 ACTIONS , C0EITION REQUIRED ACTION COMPLETZON TIME G. (continued) G.2 Reduce reactor steam done pres ure to 36 hours Required Action and s associated Completion sig. joo Time of Condition E or F not met. 3 H A I . HPCS andforefhray(L[PCS)9 ressure N.1 Enter LC0 3.0.3. Immediately inoperable. ($ysh
- g m '
Three or more ECCS ' injection / spray g ' subsystems inoperable. 4?/ E On or ao ECCS j fon pray su st an two or val s no able i HPCS Systes and one or more ADS valves inoperable. . 8 l Two or more ECCS injection / spray subsystems and one or more AD$' valves inoperable. BWR/6 STS 3.5-3 Rev. O, 09/28/92 4 4CO $LA 593- IVRI } IHjiERT 8A AHQ D.3 Initiate action to Immediately close one door in each primary containment air lock, exc p dr g ntyy a Xi de m ni tr ti e - co tr 1. .....n;...h . , say a,Je<:t a jjf -f $d f !'"il ' C UAhe r 1 a nssc~e Cpts!ryl, INSERT RIVER BEND 3.5-8 10/1/93 . . -. - - . . . = . . (lag 9pMQ.) ECCS-Operating 8 3.5.1 BASES SURVEILLANCE SR 3.5.1.2 REQUIREMENTS (continued) operation in the RHR shutdown cooling mode during MODE 3 if necessary. . l SR 3.5.1.3 h ,, h *N sVerification -~ 1:teg[p psi every 31 days that ADS air &eee4verf ressure is i operation. g assures adequate air pressure for re iable ADS ! The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The designed ; t pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at leas.t two valve actuations can occur h with the drywell at 70% of design pressure (Ref. Qf. ThQ i ECCS safety analysis assumes only one actuation to achieve 9 '.J O03 the depressurization required for operation of the low pressure ECCS. This minimum required pressure of tm bsdA rAdd _ a,ir , is orovided by the f inn. -- sig . . - . . , J. , :;r =:) The l 3 31 day Frequency takes into consideration administrative I rappi h g sp+% (SW) control over operation of the (;;;. 1-. A., ddgcohhd and alarus for low air pressure. ... 1 .,nno kckup fem M ~ygg psyg,pg ~ pensf rAhem Valve-SR 3.5.1.4 lebettt Cn+'oI j j #* g
- The performance requirements of the ECCS pumps are determined through application of the 10 CFR 50, Appendix K, criteria (Ref. 8). This periodic Surveillance is performed (in accordance with the ASME Code, Section XI, requirements for the ECCS pumps to. verify that the ECCS pumps will develop the flow ra)tes required by the respective M a The ECCS pump flow rates ensure that adequate core cooling .4p Pf : ,
is provided to satisfy the acceptance criteri f 10 Crn 50.4s (R f. 10J. gg$(; gut _ s , y f; ;,w; /g,, .l The pump flow rates are verifiedM?kt! e 5 _ =ai[Yhat ,' l fo overce,as IsT:;;5 emet tDthe RPV pressure expected during a LOCA. -^ - TheD.m :;;;w pump outlet pressure is adequate to k, overcome the e:evation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during LOCAs. These values may be established during pre-operational testing. t " 27 - O27 h Frequency for this Surveillance is in accordance with the Insevice Testing Program requirements. (continued) BWR/6 STS B 3.5-10 Rev. O, 09/28/92 1 l l5AR 93
- NR )
ECCS-Shutdmen B 3.5.2 B 3.5 COOLING (RCIC) SYSTEMEMERGENCY CORE COOL B 3.5.2 ECCS-Shutdown t i - BASES BACKGROUND } A description of the High Pressure Core Spray HPCS System, Low Pressure Core S i, coolant injection ; pray)(LPCS) System, and low (p LKI mode of the Residual Heat Removal (RHR) System is provided in the Bases for LC0 3.5.1, 'ECCS-Operating. " -3.S ;
- g. g APPLICA8LE SAFETY ANALYSES ECCS performance is evaluated for the entire spectrum of break sizes for a postulated'
' g (LOCA). The lon loss of coolant accident s $ /.s n aso a h /e,- -roaugme'4,,,pg,, basis LOCA (Ref.g 1 ) demonstrates that only one ECCSt i . - injection / spray subsystem is required, post LOCA, to ' e'f'de"'U y " "'"/f maintain the peak cladding temperature below the allowable d / *M//l t e- limit.4 To provide redundancy, a minimum of two ECCS e/ y,,p g ge CEJ Ecc5 'I subsystems are required to be OPERABLE in MODES 4 and . i Two de ate naeif8M 'p rM,3 1 " U On pray z syst - 'asso en ure entoty <g ed/o d /'/'#7 keu the ctor ssure \R ) inethe evi of a nadve at we 1drai===;yl s p[ ~[ The ECCS satisfy Criterion 3 of the NRC Policy Statement. LCO stems are required to be Two ECCS OPERABLE. The ECCSinjection injection/ spray'subsy/ spray subsyste are defined as the three LKI subsystems, System. tie LKS System, and the HPCS Tite LPCS Syste one motor driven pump, m and each LPCI subsystem consist of from the suppression pool to the RPV. piping, and valves to trans The HMS System consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or condensate storage tank (CST) to the RPV. One LPCI subsystem (A or 8) may be aligned for decay heat removal in function, if itMODE can be4 manually or 5 and considered realigned remoteOPERA 8LE for the ECC to the LPCI mode and is not otherwise inope(rable. or local) s Because of low press *e and low temperature conditions in MODES 4 Q and 5, s'afficient time will be available to manually align (continued) BWR/6 STS B 3.5-14 Rev. O, 09/28/92 GAR 93-M*h RCIC Systes B 3.5.3 EASES l BACKGR0 LNG The RCIC pump is provided with a minimum flow bypes ine, ) (continued) ;which discharges to the suppression pool. The valv in ' 1this line automatically open{to prevent pump damage us to ! Cl overheating when other discharge line valves are closed. To ! ensure rap'd delivery of water to the RPV and to sinimize water hammer effects, the RCIC System discharge line " keep fill" system is designed to maintain the pump discharge line ! filled with water. APPLICA8LE The function of the RCIC System is to respond to transient ! , SAFETY ANALYSES events by arevidins ==%e coolant to the reactor.>The RCI7 M (3yst is an inee saist estu ystem nd no .m - ~/ Jc t is aken i the s ety yses f RCIC ystem N# / : rati . Bas on it contri tion the action f I erall plant sk, ver, sys 16R04 / Techn al 5 ificat s as i by t is i luded Po cy the i 1 { 5ta ~
- t. 7 .
LCO The OPERASILITY of the RCIC System provides adequate core cooling such that actuation of any of the ECCS subsystems is h not loss7feeduster flow.tred in the event of RPV isolation accompanie The RCI has sufficient capacity to maintain RPV inventory during an i lation event. Sydeu : APPLICA8ILITY The RCIC System is required to be OptRASLE in MODE 1, and i MODES 2 and 3 with reactor steam done pressure > 150 psig ' since RCIC is the primary non-ECCS water source for core cooling when the reactor is isolated and pressurized. In l~ M WES 2 and 3 with reactor steam done pressure s 150 psig, and in fWDES 4 and 5 ACIC is not required to be OPERABLE since the ECCS injection / spray subsystems can provide sufficient flow to the vessel. ACTIONS A.1 and A.2 If the RCIC System is inoperable during MODE 1, or MODES 2 or 3 with reactor steam done pressure = 150 si , and the l @ NPCS System is (tmE5H53Bverified to be LE, the RCIC (continued) BWR/6 STS S 3.5-20 Rev. O, 09/28/92 i (44K 93-MKD. INSERT B20A .g ,( Should a design basis control rod drop accident occur, the , pp/ RCIC System can be used in conjunction with the HPCS System to meet the single failure criteria in mitigating the - consequences of the event (Ref. 4). The RCIC System is an , Engineered Safety Feature for this event and satisfies Criterion 3 of the NRC Policy Statement. l h + 2 1 8 i INSERT RIVER BEND B 3.5-20 5/13/94 4 -,w+- 4 . .s . . m - . . _ . ... . - . . . . _ . _ . . _ , _ ,_ (444 93 -H KD INSERT B20A , g,( /Shouldadesignbasis control rod drop accident occur, the .pyf { RCIC System can be used in conjunction with the HPCS System to ' meet the single failure criteria in mitigating the (' - consequences of the event (Ref. 4). The RCIC System is an Engineered Safety Feature for this event and satisfies Criterion 3 of the NRC Policy Statement. INSERT RIVER BEND B 3.5-20 5/13/94 3-NR$ RCIC System R 3.5.3 BASES SURVEILLANCE SR 3.5.3.5 (continued) REQUIREMENTS ensures that the RCIC System will automatically restart on S ,i an RPV low water level (Level 2) signal received subsequent ~ to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the 6 suppression pool. The LOGIC SYSTEM Ft30CTIONAL TEST l ) keur,or (ou r " '" ) performed in LCO 3.3.5.h overlaps this Surveillance to /. gs ,..g i /.'qc g Syste, provide complete testinglof the assumed safety function. ' ^ , [i % d m"#* N. ^' 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 perfomed with the reactor at power. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptaale from a reliability standpoint. . 4 This SR is modified by a Note that escludes vessel injection during the Surveillance. Stace all active camponents are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance. REFERENCES 1. 10 CFR 50, Appendix A. GOC 33. h 2. M, Section)5.4.6.2[ h
- 3. Memorandum from R.L. Beer (lutC) to V. Stello, Jr.
{HRC), " Recommended Interie Revisions to LCa's for ECCS Camponents," _ December 1, 1975. $'~~(4. ~ v.s4R , .5 e e soa L '/. (s} b* h Fwn/6 STS s 3.5-24 Rev. O. 09/28/92 ,- . ,, . ~ . - . - - - - - - ,n_, - . . , . , , , - . - - - Cu8 n- d ATTACHMENT 2B ITS - PSTS COMPARISON DOCUMENT . REVISION 1 l I DISCUSSION OF CHANGES i l {Af, 93-/d/D DISCUSSION OF CHANGES TO NUREG-1434 SECTION 3.3 - ECCS and RCIC BRACKETED ADMINISTRATIVE CHOICE B.1 Brackets removed and optional wording :evised to reflect appropriate plant specific requirements. PLANT SPECIFIC DIFFERENCE P.1 This change provides plant specific equipment terminology. P.2 This change provides plant specific equipment or analysis description. P.3 The safety analysis report for this station is identified as the Updated Safety Analysis Report and is correctly referred to as the USAR. P.4 The Condition and Bases are revised to reflect the use of primary containment rather, than secondary containment, for compensatory protection in this situation. Included with the primary containment boundal.y is the need for penetration isolation (or isolation capability) and closure of the , associated air locks. One closed door provides a boundary which is consistent with the normally required closure of secondary containment access doors. P.5 Changed reterence to provide appropriate plant specific ref<ronce. Since this reference has previously been noted. the original reference 13 is deleted, and the original reference 14 is being changed to 13. P.6 This comment number is not used for this station. P.7 This change is made to provide consistency with the choice of - plant specific bracketed information in t.he Specification. 11 //f_ P.8 The RCIC System is considered an Engineering Safety Feature for the Control Rod Drop Accident for this plant. P.9 This comment number is not used for this station. P.10 This comment number is not used for this station. P.11 This comment number is not used for this station. RIVER BEND 1 5/13/94 G4R 93-/y>f/) i RIVER BEND SECTION 3.6 : 1
- ATTACHMENT 1: CTS-PSTS COMPARISON DOCUMENT ATTACHMENT 2:ITS-PSTS COMPARISON DOCUMENT 1
. . _ _ - ._ . -. - . = . CU( es-m3 ATTACHMENT 1 ITS - PSTS COMPARISON DOCUMENT REVISION 1 SECTION 3.6 REVISED PAGES 1 A: MARKUP OF CTS 1B: DISCUSSION OF CHANGES 1C: NO SIGNIFICANT HAZARDS CONSIDERATIONS .4 " ATTACHMENT 1 A CTS - PSTS COMPARISON DOCUMENT REVISION 1 MARKUP OF CTS I l I l l li)K n-lvnh RECEIVED CONTAINNENT SYSTEMS , sm %_ . MAR 14 W PRIMARY CONTA!W4ENT(IETEG Iri - T;;CL ::'::^aI;j SDC LIMITING CON 0! TION FOR OPERATION a
- 3. to . l. c o ope.sra t,s
, EE:3:RB PRIMARY CONTAIMIENT shall be h APPLICA8ILITY: Operational Condition ACTION: Without PRIMARY CONTAINIENT @ T 0^r;7 cz .""."* ' M uspend handling of h @ CesA irradiated fuel in the primary containment, COREERATI0ftS and operations with a potential for draining the reactor vessel. 3, s .l. m SURVEILLANCE REQUIREMENTS 7 I r
- 4. 6.1. 2 PRIMARY CONTAIW4NT """0"I"! Tm ^ "- ^M shall be demonstrated:
- a. Githin 24 hours prior to enterinD at least once per @ u auring operan onal conettion= ey verifying that all primary con-3 14'110.1 tainment penetrations required to be closed during accident condi-tions are closed by hatches, valves, blind flanges, or deactivated automatic valves secured in position.** - -
l , (b, By verifying each containment air lock is in compliance with the ( requirements of Specification 3.6.1.4. ' 1 Axsr t. - ] <r a n . I r o I f f.co 3.c..t. so Aen. . "When handling irradiated fuel in the primary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel ! Up to twelve vent and drain line pathways any be opened under aministrat v l __ control for the punoses of surveillance testing provided the total calculate . Iow rate through the open vent and drain line pcthways is less than or al 70.2 cfa., j g- - m m r . . . 3 ,a n g 1 RIVER BEND - UNIT 1 3/4 6-2 Ameneent No. ] ~ _ -- . - _ . - _ - .-._ _ (LA/ 93*/YN l') CONTAIP98ENT SYSTEftS PRIMARY CONTAINMENT PURGE $YSTEx 4 oO 3. Io.l .3) h LIMITING CONDITION FOR OPERATION / LCO 3. (, A 3) sto3kI3 m The primary containment purge 36-inch supply and exhaust isolation valvss shall be OPERABLE and closed exce - J.M a. Each 36-inch purge valve may be open to purge system _ooeration, with 45 , p2% such operation limited to (1000 hour a per 365 days._for reducing af reorng sa. 3.t..g.3.1 ll G_ctivity and for neeneura enntrol . and a g C- -- .' c )/ 4 . If the SGTS is in the purge flow path, both trains of the SGTS must e Ad _i l 7 be OPERA 8LE, but only one train of SGTS may be operating in the purge po k 3 & cflow path.] a s to.t. =, , 1. M 3 APPLICA8!LITY: OPERATIONALCONDITIONS1,2,and3.})N'fjj ^ / ACTION: *-[fMENT 4 c c> Z (e / 3 i y f,4 mea.o port / LCO 34 3 3 a. With a 36-inch primar Qonlaihe. dt urge s'upply or exhaust isolation como A valve open for more thanU000 hour _ per 365 days, close and/or seal the 36-inch valve or otherwise isofacte' the penetration within 4 hours or be in at least NOT SHUTDOWN wit.hin the next 12 hours and cwvo E in COLD SHUTDOWN within the following 24'hourt e ~ 4 With both both SGTS SGTS OPERAtrains in one 8LE with operation SGTS in in theth[ purge
- urge f1'ow flow path or witho2 path, discontinue.
200 3.(,,A 3 36-inch purge system operation and close the open 36-inch valve (s) or l i g A d A.I e otherwise isolate the penetration (s) within 4 hours or be in at / g, g least HOT SHUTDOWN within the r.axt 12 nours and in COLO 3HUTDOWN j Qithin the following 24 hours. f t,to g ,g,} c. With a primary containment purge sup31y : 4xhaust isoistion valve (s; ~ with resilient material seals having a measureo 'eakace sea a v-==M o (eto thJt limit of surveillanca Recui rement 4. f 1. 3. 3 Jeester* -
- inceeraoTo l Eave (s)toOPERA8LEstatua>within24hoursorseinatleastHOT SHUwuwM within the next 12 hours and in COLD SHUTDOWN within the i cu o E following 24 hours.
SURVEILLANCE REQUIREMENTS T R.3 6 1.3.1 * ' '
- 1 Each 36 inch primary containment purge suooly and exhaust isolation valve shall be verified to be closed at least once ser 31 days.
l 6.1.9.2 The cumulative time that the 36-inch primary :entainment purge supply l and/or exhaust isolation valves have been open during the past 365 days shall p determined at least once per 7 days. mit of 2000 hours per 365 days from initial fue loaainguntil3eenths e first refueling outage. f l RIVER BEMO - UNIT 1 3/4 6-14 - I l (SM 92-N/d CONTAIMENT' SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 9 3. t. .U Z. 3 g At least once per 18 mont b ,(durina COLD SHUTD0W or REFUELIN7Gby , verifyingthat,onacontainmentisolationtestsignal,eacqisolation damper actuates to its isolation position. , 3 ,
- f. By verifying the isolation time to be within its limit when tested 5- Edm M ecification 4.0.5.)
o%per 92 $ s \ M9 - ,7 %, y.2 ~ , _ ,/ * /3 l 1 l . c, c ,_ x _ The specified 18 month interval during the first operating cycle may be f extended to coincide with completion of the first refueling outage, scheduled 1 to begin 9-15-87. . RIVER BEND - UNIT 1 3/4 6-52 Amendment No. 9, l l v W93-Nil) CONTAllOGENT$1ggg, MoxD To . D 'I O SURVEILLME REQUIREMEltTS (Continued) . I - /c. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position , C.5.b of Regulatory Guide 1.52. Revisi a 2. March 1978, meets the laboratory testina criteria of Regulatory Potition C.6.a of Regulatory Guide 1.52. Revi= ton 2, March 1974, for a methyl iodide penetration f of less than 0.1755. ge, 3 ,, 4 3 j d At least once per 18 months by;
- 1. Perfoming a system functional test which incl ~'-= rauTated (73") automatic actuation of the avs* = 4hroughout its emergency ~
Y y (operat'ng sequence for the: D UI M l
- (*a) LOCA, and gs ut (b) Annulus ventilation exhaust high radiation signa
% Verifying that the pressure drop across the combined EPA HO' ', , filters and charcoal adsorber banks is less than 8 inches water [ ) 3 61 ga while the filter train is operating at a flow rate ofJ 91G ,J , 12 cfm t 105. f**2~. #,. .3 erifying that tha filter train starts and isolation dampers open on each of the following test signals: k9,,ou 8. /,. M/ a ', Manual initiation from the control room, and , 7 kO) Simulated automatic initiation si =al. w'ff M "' N 4 . Verifying that the filter cooling bypass dampers can be manually I opened and the fan can be manual y started. tsono to ) r =- - . Verifying that the heators dissipate > 51 kw when tested in l TS 3'7'13 J _ accordance with ANSI N510-1900 at the design supply voltage. w-.-(e. , bank, Verifying, after each complete or partial replacement of a HEPA filter that the E PA filter bank sat,isfies the inplace penetration
- and bypass leakage testing acceptance critation of less than 0.055 in
@accordancewithANSIM510-1900whileoperatingthesys l te of 12,500 cfm t IGE.y l
- l l r l "The specified 18 month interval during the fint operating cycle may be extended to coincide with completion of the first refueling outage, scheduledj (to begin 9-15-87.
RECElvED AUG 191987 Amendment No. 9 ! RIVER BEND - UNIT 1 3/4 6-56 i SDC l ^ ^ - - - ' - - - - - - - - - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ .s..- a e a n a - m - <m.M . a 3 _ , _u. _._ N 0 ATTACHMENT 1B . CTS - PSTS COMPARISON DOCUMENT - REVISION 1 DISCUSSION OF CHANGES p,w w . . _ . ._.-__-_---__.__ ------- - , - - 1 1 1 Qi45' 93-/YAf } DISCUSSION OF CHANGES CTS: 3.6.1.2 - PRIMARY CONTAINMENT INTEGRITY.- FUEL HANDLING 1 h l TECHNICAL CHANGE - T RstS ' RESTRICTIVE ! " Generic" LA.1 The details'of this footnote have been removed'to procedures. The commitments in this footnote are not requirements set forth I in the analyses, and are commitments which will be contained in i plant procedures. " Specific" L.1 The frequency of the surveillance has been changed from 24 hours to every 31 days. The new surveillance interval is consistent with that denoted in:
- s. kMO SR 3.6.1.3.1, and SR 3.6.1.3.2 of LCO 3.6.1.3, "#CIVs" N SR - 3. 6. 4.1. 2 and SR 3. 6.4.1. 3 of LCO 3. 6. 4.1 " Secondary Containment- Operating" SR 3.6.4.2.1 of LCO 3.6.4.2, "SCIDs"
- SR 3.6.4.5.2 and SR 3. 6.4. 5. 3 ' of - LCO 3.6.4.5, "Puel Building". ,
Moreover, administrative controls ensure that open vent and , drain patnways 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- ; auxilicry building differential pressure every 2 hours; and (4) assure at least one person is assigned to each open : penetration. , k a F i. RIVER BEND 4 10/1/93 , e; (44895NXI) DISCUSSION OF CHANGES ; CTS: 3.6.1.9 - PRIMARY CONTAINMENT PURGE SYSTEM TECHNICAL CHANGE - LESS RESTRICTIVE ' (continued) L.2 The time limitations currently applied to opening the containment purge supply and exhaust valves are proposed to be replaced with specific criteria for opening. The time limits , were based c' engineering judgement and/or early plant operating experience, and not based on any legal requirement. 4 The new limits on when the valves are permitted to be open as : defined in Note 2 to SR 3.6.1.3.1 will ensure appropriate controls, and use of the system will be minimized and limited to safety-related reasons. The operating history indicates that these lines are opened only for the proposed specified reasons and for cumulative periods which are generally significantly less than the allowed cumulative times. L.3 This comment number is not used for this station. L.4 A new action (ACTION is proposed to be added which would permit continued operation with a containment purge supply / and/or exhaust valves with resilient seals having a measured leakage _ rate in excess of the limits as long as the affected g, penetration is isolated within 24 hours, this isolation is #y verified every 31 days, and the leak rate test is performed oa the purge valves used to perform the isolation every 92 days. i These actions assure that the penetration will not leak in , g excess of limits should an accident occur while operating, and thus alleviate the need to shutdown the facility. If any o ' the conditions of ACTION h can not be performed, ACTION N requires the plant shutdown that is presently required. 6 1 L.5 This comment number is not used for this station. i L.6 A note has been added to CTS LCO 3.6.1.9 to allow inoperable ' 49,g 3 purge valves to be reopened under administrative controls. Mq This change is consistent with the CTS allowances for PCIVs (including these valves) under LCO 3.6.4. However, since these \' valves are also addressed under LCO 3.6.1.9 which does not contain this note, reopening these valves is not currently allowed if they are inoperable under LCO 3.6.1.9. This change is consistent with NUREG-1434 and is acceptable based on the limitation that the valves be under administrative control while open. This ensures that the valves will be reclosed promptly in the event containment isolation is required. Y /0 N RIVER BEND 23 iC/1/f (IAR 93*MEQ DISCUSSION OF CHANGES CTS: 3.6.5.3 - SECONDARY CONTAINMENT AUTOMATIC ISOLATION DAMPERS ADMINISTRATIVE (continued) A.8 A Required Action is included to periodically verify that the isolated penetration remains isolated. This verification will assure that if the penetration were inadvertently re-opened, it would eventually be identified. Since this requirement is currently addressed in CTS 4.6.5.1.b this change is considered administrative. g(p.H'g I jg A.9 Specification 4.0.5 refers to the inservice inspection and testing of ASME components. Surveillance requirement 4. 6. 5. 3. c infers that the SCID's being tested are ASME components. A However, for RBS, the surveillance requirement applies to non-ASME SCID's. During development of the RBS Technical Specifications, the non-ASME SCID's were associated with the inservice inspection and testing program to ensure that they were being tested under a structured program. There was no intent during the development of the Technical Specifications to require additional ASME testing requirements for these non-ASME SCID's EELOCATED SPECIFICATIONS None in this section. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 An additional surveillance requirement is included to periodically verify that each secondary containment isolation manual valve and blind flange that is required to be closed is closed. These passive isolation devices have not previously been included in the verification of closure except through the ability of the standby gas treatment system to develop and maintain a vacuum. Therefore, this periodic verification constitutes a more restrictive change. M.2 The proposed specification will now apply to all types of secondary containment isolation devises not just automatic isolation dampers. Since this is an added scope the change is considered more restrictive. M.3 This comment number is not used for this station. TECHNICAL CHANGE - LESS RESTRICTIVE " Generic" LA.1 This comment number is not used for this station. j /0 WW RIVER BEND 54 10 /' 1/ Q, Q0( }}.D evve0 fo IJ fay t- be pgcy.ptpQ f, , 4/,2/4 d DCC d. '? om fags SV - -DISCUSSION OF CHANGES P CTS: 3.6.5.3 - SECONDARY CONTAINMENT AUTOMATIC ISOLATION DAMPERS s- TECHNICAL CHANGE - LESS RESTRICTIVE ,(continued) LA.2 Any time the OPERABILITY of a system or component.has been af fected by repair, maintenance or replacement of a component, post maintenance testing is required to demonstrate OPERABILITY of' the system or component. Explicit post maintenance Surveillance Requirements have therefore been deleted from the specifications. " Specific" L.1 An allowance is- proposed for intermittently opening closed secondary containment isolation dampers under administrative control as is allowed in the existing primary containment Technical Specifications. The. allowance is presented in proposed Actions Note 1 and in SR 3.6.4.2.1 Note 2. Opening of secondary containment penetrations on a intermittent basis is required for many of the same reasons as primary containment penetrations and.the potential impact on consequences is less significant. L.2 In the event both dampers in a penetration are inoperable, the existing Specification, which requires maintaining one isolation damper OPERABLE, would not be met and an immediate shutdown is required. The proposed actions for the secondary containment penetrations provide 4 hours prior to commencing a required-shutdown. This proposed 4 hour period is consistent with the existing time allowed for conditions when the , secondary containment is inoperable. The proposed change will provide consistency in actions for these various secondary containment degradations. L.3 The proposed surveillance for _a functional test of each secondary containment isolation damper does not include the restriction on plant conditions that-requires the surveillance to be performed during Cold Shutdown or Refueling. Some isolations could be adequately tested in other than Cold Shutdown or Refueling, without jeopardizing safe plant operations. The control of the plant conditions appropriate to t perform the test is an issue for procedures and scheduling, and has been determined by the NRC Staff to be unnecessary as a Technical Specification restriction. As indicated in Generic Letter 91-04, allowing this control is consistent with the vast , majority of other Technical Specification surveil ~ ances that do not dictate plant conditions for the surveillance. -s /0 Y RIVER BEND 55 10/1/ L^._ (LAK 93-Hit DISCUSSION OF CHANGES CTS: 3.6.5.4 , STANDBY GAS TREA'INENT SYSTEM i TECHNICAL CHANGE - LESS RESTRICTIVE " Generic" LA.1 Details of = the ' methods for performing this surveillance are , relocated to the Bases and procedures. The design features and system operation which dictate the methods are described in the USAR. Additionally, changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. g , (,,4, 3 LA.2 Details related to the design of the system iin this case the , i~ standby gas treatment system is designed as two " independent" jj systems) are--discused in the bases.of the proposed Technical ' - Specifications. These details do not provide any useful or necessary information to the' operator. Additionally, changes l to the system design and function are adequately controlled by , the requirements of 10 CFR 50.59. " Specific" L.1 This comment number is not used for this station. , L.2 The phrase " actual or," in reference to. the automatic initiation signal, has been added to the surveillance requirement for verifying that each subsystem actuates on an automatic initiation signal. This allows satisfactory automatic system initiations for other than surveillance ; purposes to be used to fulfill the surveillance requirements. OPERABILITY is adequately demonstrated in either case since the subsystem itself can not discriminate between " actual" or " simulated." i } i RIVER BEND 57 5/5/94 (}v4L 93 l'lkh DISCUSSION OF CHANGES CTS: 3.6.6.2 - PRIMARY CONTAINMENT /DRYWELL HYDROGEN MIXING TECHNICAL CHANGE - LESS RESTRICTIVE (continued) " Specific" L.1 A statement is inserted to indicate that Specification 3.0.4 is not applicable for the condition of one primary containment /drywell hydrogen mixing subsystem inoperable. An OPERABLE primary containment /drywell hydrogen mixing subsystem remains available in this condition, and igniters are also available to backup the system. In addition, the purge system does not impact normal operation of the plant in any way, and hence, would not provide any additional initiators for plant transients during startup or MODE changes. Since probabilities have been determined to be acceptable for a 30 day allowed out-of-service time for one division of hydrogen control function equipment, redundant equipment in this system and the other systems is available to perform the function, and there is no impact on normal plant operations from the unavailability of this specific equipment, the exception is considered to provide no significant impact on safety. L.2 An additional Action is proposed for the Condition of both primary containment /drywell hydrogen mixing 2bsystems inoperable. The igniters are also designed to control hydrogen in a post-LOCA environment. However, redundancy for the hydrogen contvol function would be reduced. Therefore, a period of 7 days is proposed to allow attempts to restore at least one division of the purge system to OPERABLE status before requiring a shutdown. This Action would possibly prevent an unnecessary shutdown and the increased potential for transients associated with each shutdown. L.3 This comment number is not used for this station. L.4 The time limitations currently applied to opening the hydrogen g ,(p,3M mixing system inlet or outlet valves are proposed to be j.3 replaced with specific criteria for opening. The time limits were based on engineering judgement and/or early plant s operating experience, and not based on any legal requirement. The new limite on when the valves are permitted to be open as defined in the Note to SR 3.6.5.3.2 will ensure appropriate controls, a2d use of the system will be minimized and limited to safety-related reasons. The operating history indicates that these lines are opened only for the proposed specified reasons and for cumulative periods which are generally significantly less than the allowed cumulative times. RIVER BEND 64 5/4/94 es - A A A nb ,.,Mm 4m--, A6- m 2- s- -eL+a h 6 k ATTACHMENT 1C - CTS - PSTS COMPARISON DOCUMENT REVISION 1 NO SIGNIFICANT HAZARDS CONSIDERATIONS 1 l k i _ _ _ . _ _. _ , _ . - __ _ . . . _ ~ _ _ _ _ _ - -- i )]Af 93-NRI , 1 NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS:'3.6.1.2 - PRIMARY CONTAINMENT INTEGRITY.- FUEL HANDLING 1 "L1* CHANGE .Entergy Operations Inc., has evaluated this proposed Technical Specification change and has determined that it involves no - significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92'. The following evaluation is provided for the three categories of the significant hazards consideration standards: * 'g 1. Does the change involve a significant increase in the 9 d probability _or consequences of an accident previously f(p evaluated? This change would increase the survei;'ance frequency of the verification of the primary containment penetration flow paths. This increase in frequency is consistent with the frequencies during operation of the plant. The proposed change is offset by administrative controls that ensure open vent and drain . pathways will: (1) only be opened to support leakage rate testing; (2) not exceed 12 valves;- (3) require monitoring 5 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 is assigned to each open penetration. This provides additional assurance that the conditions required by this surveillances are being maintained. Therefore, this proposed change does not involve an increase in i~ the probability of an accident previously evaluated. Further, i since the change impacts only the frequency of verification and , . does not result in any change in the response of the equipment to an accident, the change does not increase the consequences of any previously analyzed accident.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? ,
This change does not result in any changes to the equipment design or capab!1 ties or to the operation of the plant. Further, since the change impacts only the frequency of verification and does rot result in any change in the response of the equipment to an accident, the change does not create the possibility of a new or different kind of accident from any previously analyzed accident.
- 3. Does this change involve a significant reduction in a margin of safety?
The change impacts only the frequency of verification of-primary containment penetration flow paths. This change does , not involve a significant reduction in the' margin of safety because of the additional administrative controls required. This essentially confirms the capability to maintain the building leak tightness. RIVER BEND 3 5/4/94 fMd 92-H& NO SIGNIFICANT HAZARDS CONSIDERATIONS ,3, (p .I' CTS: 3.6.1.9 - PRIMARY CONTAINMENT PURGE SYSTEM
- t.
- l "L6" CHANGE Entergy Operations Inc., has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92.
The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
This change will add a note to allow inoperable purge valves to be reopened under aaministrative controls. Failure of purge valves to close is not assumed to initiate any accidents. Therefore, this change does not increase the probability of a previously analyzed accident. Because the valves are required to under administrative controls when opened under the proposed note, adequate assurance is provided that the valves will be reclosed promptly in the event containment isolation is required. Therefore, this change does not significantly increase the consequences of a previously analyzed accident.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change introduces no new mode of plant operation and it does not involve physical modification to the plant. Therefore it does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
This change does not involve a significant reduction in a margin of safety since the proposed change will continue require administrative controls to be in place while the valve (s) are open. This provides adequate assurance that the valves will be reclosed promptly in the event containment isolation is required. NYY RIVER BEND 16 10/1/03 % QR 93-Nah NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.6.6.2 - PRII@.RY CONTAINMENT /DRYWELL HYDROGEN MIXING "L4" CHANGE f4(p .3',,) Entergy Operations Inc., has evaluated this proposed Technical f ~ Specification change and has determined that it involves no ~ \ jg significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. g The following evaluation is provided for the three categories of the g significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
This change would replace quantitative restrictions for opening of the hydrogen mixing system inlet or outlet valves with qualitative restrictions. The hydrogen mixing system is not considered as the initiator for any previously evaluated accidents and, therefore, revising the opening criteria will not significantly increase the probability of any previously evaluated accident. Further, the change maintains the, current method of operation and response of the hydrogen mixing valves to an accident. Therefore, the change does not increase the consequences of any previously analyzed accident.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
This change does not result in any changes to the equipment design or capabilities or to the operation of the plant. Further, since the change impacts only the opening criteria for the system and does not result in any change in the response of the equipment to an accident, the change does not create the possibility of a new or different kind of accident from any previously analyzed accident.
- 3. Does this change involve a significant reduction in a margin of safety?
This change impacts only the opening criteria for the hydrogen mixing system. The methodology and limits of the accident analysis are not affected, nor is the system response. Therefore, the change does not involve a significant reduction in the margin of safety. RIVER BEND 51 5/4/94 @a goeID ATTACHMENT 2 ITS - PSTS COMPARISON DOCUMENT REVISION 1 . SECTION 3.6 REVISED PAGES A 2A: MARKUP OF ITS 28: DISCUSSION OF CHANGES k 4 ATTACHMENT 2A . ITS - PSTS COMPARISON DOCUMENT REVISION 1 MARKUP OF ITS - . . _ . - ~. . . hat' 9L:-/YEI) Primary Containment 3.6.1.1 " rahl SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.1.1 Perform required visual examinations and -----NOTE------ leakage rate testing except for primary SR 3.0.2 is not containment air lock testing, in 3/jpf t accordance with 10 CFR 50, Appendix J. applicable ' as modified by approved exemptions. / In accordance [OI /the ximum lowaVeles ge r te, with 10 CFR 50, 3 s % of rima (cont noen air I, Appendix J, as t x p day a the e cula dp modified by 1 ntai nt p sure P,. approved exemptions i I ' i FSR 3.6.1.2 Verify primary containment inte ity in acco ance with t. i stru ural I accordance -3 1 Primary th the - Con inment Tendo Surveillan Program. erinary / < jContainment (Q V Tendon A : Surveillanc Program ,o ' ~ . ?.- UWG O (Q!C. f (f(CCplfgr/Cd Cftfth$ h .1 /. Aa. L wee, A<q h ),ls any n'a,fp fo//o~,y s6 per-fu-Q in acco/A-ce M ) [ toire so, Ap,sessf, x as nohird ay ayoftue.% egep/m u,1Ac , /caxa,c eale a<c . h < a m h<yfauce th e,dc'na. 6c0 %cc fy)oa C feds us k a trh 4< i tkc hp:. A l=7'. ~ .. BWR/6 STS 3.6-2 Rev. O, 09/28/92 (1,AR 93-l'fN ) Primary Containment Air Locks 3.6.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.2.1 ------.--------..-NOTES------...-...-----
- 1. An inoperable air lock door does not ,
invalidate the previous successful perfomance of the overall air lock leakaae test. - dDu ao Mo 0Es 1,2. , a43] @ tes
- 2. y/acce!a u snali ptance be evo usten against criteria of SR 3.6.1.1.1 in accordance with 10 CFR 50 Appendix J, as modified by approved exemptions.
Perfom required primary containment air .----NOTE.----- lock leakage rate testing in accordance SR 3.0.2 is not G.6,/,2 , with 10 CFR 50, Appendix J, as modified applicable . #7 ' 'Ph,ckY k"[ [ [ N bC' ~ ~ ~ ~ ~ ~ ~ h e der.ce ci;::-': f:7 ;;r :e Y. In accordance ) ^ ters:x; ' with 10 CFR 50, 's Appendix J, as
- a. ~0v rail a lock eskagefateishI ' modified by p7 [2scf when ested t a P, . '
, approved 3 exemptions 'b. For ea door, eakag rate i l s[2 fh) wh n the p het n the 's door eals i press ized t '^ 7, 6. /,2 m [ .0 P,] .- ' .-49/9 s s \ g y-SR 3.6.1.'2.2 Verif primary containment air lock seal j 7 days y; i air f ask pressure is a)90}*'psig. - t .a. ._ [V mbde cara u ! I s a sw ce 4, he cal <e acQ a~As ! Q k a"a ' ereu s A~~d he~ la u 1 l l BWR/6 STS 3.6-7 Rev. O, 09/28/92 l l 04 9.1-Mit) \ PCIVs 3.6.1.3 3.6 CONTAIMIENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs) LCO 3.6.1.3 Each PCIV shall be OPERA 8LE. g M. co. /> - ~ ' l ,- .d t/2 - l APPLICA81LITY:- MODES 1, 2, and 3 '# # !* k' s ! ,,,' ghenassociated6nstrumentationisrequiredtobeOPERA8LE ' ##Yf .te m a s d '* 4 . ($57TC"3.a.o.i,
- 1. . . - .~ ~. " a' - --rr .-unsry G.. =. r.: ::= e+4ee2.
c c, !- sAs w, ?> n -- A fer L CO 3. 2. 6. /, " 916,,, ' -?to at '*I e tea e *pr *UeI [o,,,j,,o,.,,,,9 pj Q, l A5lDNSX"' L'? " La~se~Mde, &"*f'd" p,/f fj,/,4'o db. l I .....................................N0TES.--....-..-...-...-....--........... . i 1. Penetration flow paths fhe ;t ffTT*."h L.rx -"- ----*-" h - ;@ may be unisolated interuittently under administrative controls.
- 2. Separate Condition estry is allowed for each penetration flow path. I
- 3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.
- 4. Enter applicable Conditions and Required Actions of LC0.3.6.1.1, " Primary l Containment." when PCIV leakage results in exceeding _gy.grall containment OPi leakage rat 7e acceptance criteriap poOEJ t, t,.pd u -
CONDITION REQUIRED ACTION COMPLETION TIME A. ... ..N0TE.--.T -- A.1 Isolate the affected 4 hours except nly a icable.to na==+*mHan flow path for main steam c.% i penetrat with two flow s. ths k / 07 vec. of at least one closedad line w ...................., ae-activate 2 u ta,- valv m One or more penetration flow paths [ 6/e'a.d,.idrawa c/03c f J /a + cr e. *e c , f'vMc 8 hours steam line for main with one PCIV check valve udh f/0* h ; inoperable (excent des I throug4.M e valve- l $ ;---- 9 x e- @ e, t D l r ec u rek. m = c -, . ... . . . - .. s < Q* +Ppesa, leakage not M \ s withiti 1initf t' 'J lel '3 #31 (continued) BWR/6 STS 3.6 9 Rev. O, 09/28/92 su (A M 93-t Q PClis I ACTIONS (continued) C0WITION s REQUIRED ACTION CONPl.ETION TIME
- 8. ------- NOTE------- 8.1 ' Isolate the affected I hour Sq1y app able to naastration flow path h pe with ration low paths PCIV t
g by v;e o< at /c4;+ one. c/vsed a c/ l q.---------.----.---J_ r h- acio'vde.2 adon one or more dose k eavalvl' wa n/n, penetration flow paths of h/<d //mc.e.. l with two PCIVs 1 -' I g d e *!'d 'P*e '" notwithinlimith A" LED l C ---------NOTE--------- C.1 Isolate the affected [4] hours Only appifcable to penetration flow path penetration flow paths by use of at liest with only one PCIV. one closed and \ . de-activated automatic valve, N One or more penetration flow paths .'-h closed manual valve,' or blind flange. ; with one PCIV g CS inoperable. -s < l ! C.2 -------NOTE--------- .s ves and blind fla s in high ' ,N ' radia on areas may ' be veri ed by use of N administr ive means. f Verify the aff ed once per 31 days , penetration flow ath is isolated. .._.- C., - , < ,e a ,2 ,2 Afl.~ 9Aks w W ' _ l
- r-2 7 :: =~T Restore leaka e rate 4 hours 1
( C '-- -'Testage rate to within lie t. - F M thin 1 ~ a l ~~~ - ' % c.e.g\ E.e p ur g. v wh y . (enntinued) BWR/6 STS 3.6-11 Rev. O, 09/28/92 I ) QA 93 N{I} PC2Vs 7 3.6.1.3 (S ,(p.I ,3 631 ACTIONS (continued) / C0WITION f REQUIRED ACTION COMPLETION TIME ~ ' - t l )p I One or more penetration flow paths 1 I Isolate the affected penetration flow path 24 hours d with one or more t cv vsc v4 at /cas t i vh [containmentpurge valves not within 1 s dee-ed c/ofe4 ve d i purge valve leakage W /"82 \fais-ado]W c Im e f ino~ vel e2 wlve, o r L /i o 2 J/s u g e-. ! ? .. f f' 8 2 ,-- ,77 g -- 7 --- f QIMGgn'~ ' a y\,w} ,,, b ) radiation areas may be verified by use of s administrative means. ' I Verify the affected once per 31 days ; y g/ penetration flow path is isolated. for isolation devices outside - g / l gontainment l ~ r Prior to
- l entering MODE Y v3 5 '
from M0CE f 8 l not perfomed C ' within the ' l previous 92 days ,' { for isolation y,p j devices inside
- r ontainment
@) ' 3 Perfom SR 3.6.1.3 Once er for the resilient f92 ay: seal purge valves' closed to compi ith Required Action . l. (continued) 1 BWR/6 STS 3.6-12 Rev. O, 09/28/92 Aj n /v47) PCIVs 3.6.1.3 , SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 3 ,___ , SR 3.6.1.3 ----------- ------NOTE------- ------... Only requi ed to be met in ES 1, 2, j an( 3. / -...--. ----------.---.. ..-....----.- / Verif each [ ] inch inary contain t 31 days ' pu valve is sealed losed except r on purge valve in penetration fl f p h while in Cond ion E of this CO. ~1 1 -7. (e. l. C SR 3.6.1. f ------------------NOTES------ ;------..-- ~ g
- 1. 'Alaj w'
/ Only required to be met in MODES 1, p 2, and 3. fe p[de&<l f
- fe
).. O l' t
- 2. Not required to be met when the inch primary containment purge
=- r valves are open for pressure control, Al. ARA or air qualit considerations / gor Ep6dWig V for personn ent , ern urvellia se , that requi he va ves to be op , "'g <aL?-m::: ;;; ; fa;; . ;; h:.:r;;:)r-- a .u...-+, n... .-:. s a. p3 ..-------- -- ~ " .-.. 7-'-'..~~- ----.- -- : Verify each inch primary containment 31 days purge valve i losed. \ i = ~ - - I ~ ontinued)
- 3. h c4c. SYa"Sby G as 'fEedmed (%7 Jub .syjfen is i kthepr%'y ;
l i$_ - coulaw befb 567* nesubcy d []atye. skMs snus .f/o l > pa1k,be ' & l(p 0 FEM 6AE, -["a/AM O 'oulf one ScrT .s uf ryde ny be ofou t'i~ i N tbe oc~ jo aiy pan . cod ln~ed w;e ; \, -- , I l i l BWR/6 STS 3.6-14 Rev. O, 09/28/92 l (i Ai 9:-nth PCIVs 3.6.1.3 . SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.1.3 ' Verify the isolation time of each power In accordance - / / h MSIVs,(is within limits.operatedandeachautomaticPCly,ex with the /I nservice 4 G 0 d' .2 ~ Testing dd'1 Procram ge & dy:Je @ - SR 3.6.1.3 , ------------------NOTE ------------------ h# 7 ,< Only 2, andrequired
- 3. to be met in MODES '
t 1, ,[/ 2. sesItsghallbpevalu ed agai'n e'6 " i,; a epta ce cri ria of SR 3.6 '1.1. 67 ac rdance ith 1 CFR 50 ' BW / Appe fx J, s modi ed by ppr ed 49 @f / / exe tions l Perfom leakage rate testing for each 184 days primary containment purge valve with resilient seals. ANA Once within 92 days after opening the valve (p - - SR 3.6.1.3 Verify the isolation time of each MSIV is In accordance SGd 3 m with the e @ 3(secondsands 5(seconds. d" n;;" < . ~ _ Program g ( ' u w.u;=- . o (continued) i BWR/6 STS 3.6-16 Rev. O, 09/28/92 (MR 93-NRO PCIVs > 3.6.1.3 l ,, g SURVEILLANCE REQUIREMENTS (continued) : #9 SURVEILLANCE ! FREQUENCY ' SR 3.6.1. Verify each automatic PCIV actuates to l ithe isolation position on an actual or J18[ months l simulated isolation signal. -} 7 - i Q SR 3.6.1.3 [ b n g-- b --------N0 -- ----.--- ,-- - - - NOTF - " ~ esults shalT be e W uatedfagains G' S R 3 n1 h a * + G acceptance criteria of SR,3.6.1.1.1 in! - *PPl*US. T l ' -- . C l f;j)'l s n' acdords'nce kit 10 CFR 50, Appendix J 4mddiffed td ao roved exemptions./ _f, as \ g Mo ho h Areptaed p-- ify the combinea ieauage ress h y ror / m moorrj Ver secondar containment bypass leakage & ---=Ts3 ,a Q,1,a eA '5 ._ ** !'.( pm. M , 9 [""*"yo,ooo P""'I**d toera so, %dia 4J ch opmed ene- % ph _ Q a$aAW h SR m , s,; 3.6.1.3 < t ---------------.-NOT -------------- suJts s, hall se evaluar,ed TgtfMW - , 'n y _hO " )pscceptancec terie of R 3 6.1.141 i , ' j accordance th30CF 50 ppe f - usediffed b ap1 roved x ions"ndixJ/as 2 pig, ---..-----------.------------------------ \ r Verify leakage rate through :' ~ rm ----NOTE----- : is s M scfh ,when tested SR 3.0.2 is g at t . $;. 150 / f, , not s p, Q[:ec et, v o w4 yapplicable m y _ ihe vo\<es ser<ed k eod dUuion oj. i In accordance ths. Pt.c.S s - & ~~. - FR 50, " Appendix J, ' [ , As. modified s ' by approved exemptions (continued) L BWR/6 STS 3.6 17 Rev. O, 09/28/92 1 'k/M c?:-H & INSERT 17A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY [T SR 3.6.1.3 7/ Verify in-leakage rate of 5 340 18 months scfh for each of the following ,' valve groups when tested at 11.5 , __ psid for MS-PLCS valves and 33 (c,G./S}- psid for PVLCS sealed valves. i \ =9 - a. Division I MS-PLCS valves , and Division I PVLCS valves.
- b. Division II MS-PLCS valves and Division II PVLCS valves.
- c. Division I MS-PLCS valves and all first outboard PVLCS valves.
4 e O O INSERT 7/r/t/ RIVER BEND 3.6-17 (),As 9: /51} " PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY-f SR 3.6.1.3.11 ------------------NOTh--------------- [- j...- Only required to be met in MODES 1, p ## hd/_ 2, and 3. i_'["/ff#f[ s _ 'S'- y'g'g 3 ~ /'2. 'sultg shaly be evaluated 'ag2ny ] d7,, / I Ad acceptance,<:riteri4 of SR' 3.6.1.1. / T - i in Accor nce with 10 CfR 50s/ / s Appendi J, as,Modifie,d by 9' pro ed N xempilo n - ojj Ver f c n r e f ( ~h times the total number of PCIVs through hydrostatically tested lines that fJa ac"'g'* d TB i -hg,7 penetrate the primary containment is not exceeded when these isolation valves are ' /o cre . o, sp<~/4 C /fA tested at a 1.1 P,. ! ) a, , d/.4;S hy yg,uf , ,,,,g,1, w _ _ . _ . . ~ / J ~ ~{.6.1.3.12/- SR ------------------ OTE------ --------- / i / Only/equired to e met in DES 1, ; 2, nd 3. d ! , erify each ] inch mary cont noen[ [1 months - purge val is block to restri the ' valve fr opening [50]4. ) , \ - -) 1 l u BWR/6 STS 3.6-18 Rev. O,09/28/92 . . l [A.4$ 93-/YIlh l INSERT 18A SURVEILLANCE REQUIREMENTS ~~ , _ SURVEILLANCE FREQUENCY l2 SR 3.6.1.3. fR ult f 4NOT ^ ----+--- sh il b eva ate J--- ----NOTE---- SR 3.0,2 is g ,6 ,/,.2 ,r,, , /, ; / ai ta ept ce iter a of. not st/9 ~ ac rda e wigh} ,SR .6.1 .1 applicable gf @/j10 f FR 0, pend J, s / ------------ Im6dif db appr ed emp ').fr .l. 3 - (' o[s)-- ,g -- .- - ~ ;rcf'""p Verifykthecombinedleakagerate In for alf<anrulus bypant leakage accordance paths gs s 13,500 cc/hr)when with 10 CFR pressurized to a Pa. 50, Appendix J, as modified by approved exemptions l i 4 e k INSERT RIVER BEND 3.6-18 10/1/93 [MS. 93-l'/fl} ff s- PLC.t 6 3.6.1.9 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY rd su st er e ment 'qircuitry, r SR 3.6.1.3.3 system functional test of each ,[18%nths MS-PU.S W ) __ I i 'f' N . ~ - 3, 4, /,1 s tg 2 6.f. c) - #1 s ! BWR/6 ST3 3.5-28 O, 09/28/92 Rev. Q 97-/fd/) Primary Containment-Shutdown 3.6.1.10 -SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.10.1 ----------------NOT ---------------- d()p Not drainrequired line pathways to beprovidedmet for$ he vent and total calculated flow rate through open vent and drain pathways is s 70.2 cfm M t.': :- ^r- '-- '^r . f/ g+_v" e ' <-- - ' d : : : .a
- 2. r 1 red be me for pat $ays L apab of ing cl ed by OP M8L prir ry c ai f t iss(ati valves.
automa 'c g Verify each penetration flow path, f.1do,dDr/ m required to be closed during accident ,o t " , o conditio,ns, ism..c,lo, mn sed.~J^- ----"'^ 4 - as . , . 2 ,- A/ 65i6mmt. Q , t 2.6.t.m #6 ' u l l BWR/6 SUPPLEMENTAL 3.6-28 11 10/1/93 (fAR 93-IkI] Primary Containment Hydrog:n Recombiners 3.6.3.1 3.6 CONTAlletENT SYSTEMS 3.6.3.1 Primary Containment Hydrogen Recombiners (fi ::r:::::P; in:::1!:C' h LCO 3.6.3.1 Two primary containment hydrogen recombiners shall be OPERA 8LE. APPLICA8ILITY: MODES 1 and 2. ACTIONS .C0fGITION REQUIRED ACTION COMPLEi!0N TINE A. One primary A.1 ------.-NOTE--------- containment hydrogen LCO 3.0.4 is not recombiner inoperable. applicable. , Restore primary 30 days containment hydrogen recombiner to OPERA 8LE status. B. Two primary 8.1 Verify by 1 hour containment hydrogen administrative means recombiners that the hydrogen 8lE inoperable. gg ci ; control function is -- = maintained. ,62pdurs/ [the afte s /
- 3. h,3. /
B.2 Restore one primary 7 days * (o containment hydrogen recombiner to OPERA 8LE status. s C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. BWR/6 STS 3.6 37 Rev. O, 09/28/92 gcc c,6, g, p { Ad 93 /vBI) INSERT 39A B. Two primary B.1 Verify by 1' hour containment and administrative drywell hydrogen means that the igniter divisiona hydrogen control inoperable. function is e pe 1 I maintained. o s jA (t r ft j s
- 3. (p . 3./
B.2 Restore one 7 days d(p primary containment and drywell hydrogen igniter division to OPERABLE status. ' eh e ) INSERT RIVER BEND 3.6-39 10/1/93 Primary Cantainment and Drywell 14 D/WI) Hy(rogen Ign 3.6.3.2 ACTIONS (continued) ColeITION REQUIRED ACTI0ff COMPLETION TINE' / f w p3 8. -.. .-..-NOTE-.---- .- B.1 Restore djacent 30 day { S arate Conditio ignito s) to try is allowe for OPERA E status, each area. One or no open areas with ad cent ignitors inoper le. / h C. ..-.-.N0TE-..----.- Separate Condition 1 Restore one i nitor in the enc 1 ed area 7 days / entry is allowed for to OPERA 8L status. each area. One or more enc 1 ed areas with no 0 LE / ignitors. C Q Required Action and C[ associated Completion p.1 8e in MODE 3. 12 hours Time not met. i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.2.1 f?:ry :::t f tit:r i: ;;; qi::d " :: : .* 184. days assee4:t:diditrdicii;;i: ::ti':ltid.' ' I Ee. eve.e- e.o.tk prl mar 1 cbA4o.nme44 and doywe.ll bydken 'qnikt- divise (continued) C Q"5 P"-4.cm c.orred versos volge_ .,, rneosuremeo45,4o <- wr'ify rectuired .. 9 'syi4ers ,- q- &f VI&e ." ; . BWR/6 STS ( Y; aw iG ~' '/ p Rev. O, 09/28/92 (kk93-14II) Primary Containment and Drywell Hydrogen 19nitors 3.6.3.2 .o SURVEILLANCE REQUIREMENTS (continued) 1 SURVEILLANCE FREQUENCY { SR 3.6.3.2.2 ..................N0TE..-....--.--....-.. Not required to be performed until 92 days.after discovery of four or more ignitors in the division inoperable. so e i i n c t i , R3 .3.2.3 Visually xamineeac!accessibl ignitor 8] months to veri y cleanlinps. C f SR 3.6.3.2/) sufficient current draw 3/i e417001/F surface temperatu . /18[ months h M"- uired ignito y inD i r ---- p rr er Tinacceysible areasp credr.pg r _ '~ viced igaer m o<dssibleT f SR 3.6.3. Q Verify A _______ ac0 $ [f Y uru of @ J18[ months h h e) 1700KF. g ku y te e. u b Q/ % e.r3a c d a e a. d ) tb yenk,tr.ys v N cl iv i i u m uJ ge ib - c.o r r ch vi.r3, 3 'v u o De meau.-e h ^ (, 'L h ya rt gred g. e Aars s ~ Se N u e. - ' ,. 4.m ._. 3 '7 l *2, G, L 2 42 / 1 BWR/6 STS 3.6-41 Rev. 0, 09/28/92 i i _ _ GM 93 -/YIc[] Y "!.'.__. $yft b ~ JA 3.6.3.3 P~r di0i4weJilhfrop Ti@a4Q 3.6.3.3 (IEiiiiiETE*UjpqD ystem] j fnNw cordd,Asd[drhell hydro 9e Ilx j LC0 3.6.3.3 wo if =:r"-" p su5 systems shall be OPERA 8LE. l APPLICABILITY: MODES 1 ar.d 2. ACTIONS l CONDITION REQUIRED ACTION COMPLETION TINE h44_eL6,WM /kmi! LA.wsa ,n3, A ~ ----f K.1 ....-.--NOTE--------. Q . One n r! subsystes inoperacle. f LC0 3.0.4 1s not applicable. i Restore W 30 days subsystem to 081 OPE LE status. [r ery c.a Fe.3,14/g,$_IIId Tairf. 7
- 8. Two W ir:: r;;) 8.1 Verify by 91 1 hour subsystems inoperable. administrative means that the hydrogen g --
control function is fg__Er N maintained. 12/ours after bE ~ 362./ 8.2 Restore one GEiiiiiG4 7 days #(o (M subsystem to OPERA 8LE status. C. Required Action and C.1 Be in MODE 3. s 12 hours associated Completion Time not met. 1 BWR/6 STS 3.6-42 Rev. O, 09/28/92 l
- l. ._ _ ._
i h lA AR e3- mj) en h ) Secondary Contai ACTIONS (continued) COMITION REQUIRED ACTION COMPLETION TIME i r '- C. [Se ndary C.I --------N0 --------- c tainment) LCO 3.0. is not ' operable duri applica e. movement of ir diated ----- -------------- fuel assembli s in the / [ primary or econdary Su end movement of amediately containeen , during i adiated fuel CORE ALT TIONS, or ssemblies in the during RVs. [primaryand secondary g containment]. C.2 Suspend /. , RE Immediately ' ALTERA ONS. ' i M C.3 nitiate action to Imme ately suspend OPDRVs. e-- -) SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY n - shjela builde4aa utus ,ameliaq Luldau o.J Eel yeldig.) l Q w . SR 3.6.4.1.1 Veri f,.M : f . c ..:= - r Wyacuum is ' 24 hours m ~ -M l h,_M; inchad 10.0I of[0bvacuum water (8 gaugg"?'8'b - SR 3.6.4.1.2 Verify a114 secondary containmentF -g' 31 days g equipment hatches are close {,, x k L ,- e y i (continued) cwh ze gh,
- 3. 6. 4. /
'7 BWR/6 STS 3.6-45 Rev. O, 09/28/92 }/ L .rp D S s 3.6.4 3.6 CONTAllMENT SYSTEMS l Oompers 3.6.4.2 Secondary Containment Isolation h SC l LCO 3.6.4.2 Each 50 hall be OPERABLE. - l &al boulda b_Ee_i beiid!-- , ' Sol @ W - _ _ - ) APPLICABILITY: MODES 1, 2, and 3, i During movement of irradiated fuel assemblies in the ! y i.- y ; ;;;;; t ; ci-t:f.. .nt S Suring ORE Alit TIONS, 7C Durin operatio with a pot ntial for raining t reactor u essel (OP Vs). M i ACTIONS 1. ------------------------NOTES------------------------------------ Penetration flow paths may be unisolated intermittently under , administrative controls. ' t
- 2. Separate Condition entry is allowed for each penetration flow path.
3 itions e and by 5Required Actions for systems made h . Enter ap licabiinoperab CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected 8 hours penetration low paths penetration flow patit. OP2 with one S Oy d e d at lea r ' inoperable. O one closce and ~~~ .- de-udi vof=Z ' lET avicera hc. 2amper,. ' TM* !4 c!vsed waruol\s%er cr blind llwroc. 8G (continued) BWR/6 STS 3.6-47 Rev. O, 09/28/92 O,:r < tvil s @ 3.a. . ACTIONS ' CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 --------NOTF-- 2 .... f:h;; ;;d M :sO--- Maks's r ar may I be verified by use of administrative means. Verify the affected Once per 31 days l penetration flow path is isolated. B.1 Isolate the affected 4 hours 8.f--(y-app-NOTE.-.... On able to nenetration flow cath. , penetration ow paths l hy use of of least with h o isola n f one closedand e ~F valves.\............ q...... f de-acflyefg> j yynie et \ c135e<(qviyft I do,yer One or more \ t da -~ 2"sC' penetration paths N with two SC ' inoperable. H i*, 5: C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A 6tlQ or B not met in . MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours g (continued) 1 i BWR/6 STS 3.6 48 Rev. O, 09/28/92 l l 1 'p p m:- tv <I ) 3. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY ~_ mptr1 SR 3.6.4.2.1 ------------ ----NOTES------------------
- 1. "%h:: and blind flanges in high radiation areas may be verified b -
" A[df 3GA# j use of administrative ' ca7 /
- 2. Not required to be met for SC s that ,
h -----.$.$'$" $" --- Verify each secon ary contamment 0"J*' 31 days isolation manual and bliM flange that is required to be closed duttnq g, y ',
- s. f.j accident conditions is closede ,
/ 9 SR 3.6.4.2.2 Verify the isolation time of e power operated and each automatic 5 s cordance P within limits. ifA thv .- Its Jce ~ / e in , , 8 *.*_g a ' ; -- c _ 3.6.4.2.3 SR Verify each automatic SC the isolation position oh-in actual or ctuates to ' ,[18[ months h simulated automatic isolation signal. en.4 4 9 e I BWR/6 STS 3.6-50 Rev. O, 09/28/92 t @A4 93-/ vel) SGT System 3.6.4.3 3.6 CONTA! MENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE. , APPLICA8ILITY: MODES 1, 2, and @ 'Ourin fuel as h w ,Duri vesent f irradiate imary a secondary CORE A ERATIONS, ntainee ], Dur' g opera ons with a otential or drainin thereactor) lies in th ]A i vessel OPORVs). j ACTIONS . CONDITION REQUIRED ACTION COMPLETION TIME 1 I A. One SGT subsystem inoperable. ./ Restore SGT subsystem to OPERA 8LE status. 7 days 1
- 8. Required Action and 8.1 Be in M00E 3. 12 haurs associated Completion Time M 8.2 Be in M00E 4. 36 hours j l C. Require Action and / -NOTE ------------
associ ted Completion LC0 3.0.3 is no applicable. Time Condition A -------------- -------------- 4' not t during mov t of irradia C.1 Place OPERABLE SGT Immediately fue assemblies in he subs stem in (03 t [pr mary or second ry cotainment],durng ope ation. C E ALTERATIONS or 98 , g,6, o ring OPORVs. 4/(f (mt h=t-- 1 L _. T ,; cggAg i c, & = 4, n a ,.a v u , y g,, @ sen %3.6 s. _ _ . , J BWR/6 STS /wo Rev. O, 09/28/92 ^ A'N$L , z. e.c '/b (A M K - tv *) y Orywell 3.6.5.1 3.6 CONTAINMENT SYSTEMS 3.6.5.1 Orywell LCO 3.6.5.1 The drywell shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell inoperable. A.1 Restore drywell to I hour OPERABLE status. B. Required Action and B.1 Be in MODE 3. associated Completion 12 hours Time not met. 6lg1 B.2 Be in MODE 4. 36 hours r c 2)J, SERT 54 A hh . 3 ' SURVEILLANCE REQUIREMENTS SURVEILLANCE , FREQUENCY SR 3.6.5.1 k : ",(y or ey#t Verify bypass leakage is bypass-leakage limit. BI p{ r the /18Mnths h a Jr.s-w . *9 SR 3.6.5.1.2' Visually inspect the exposed accessible -n" " ' l interior and exterior surfaces of the ac e. p drywell . pec or~j,,or az ce m4 - - ~ 1y,oe A se n ,<.r e , ~ y y; 7 6 1. r. i JIOso ov: r , c v r p v, b fc N1rY J*t Y sl'fbf
- C 'fC'~' /
i byfft!H t e L .'. , 21ela1fteocc l l00Vtry6 Arvc: u :' u M.hl:a, e i ic a,4L% ,-,,.i s ee ,< / ~c :. , p . a /; c ,p'&a n ,/c BWR/6 STS 3.6 54 Rev. O, 09/28/92 e (2 A2 9 0-/* l) Orywell Air Lock 3.6.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.2.1 ---- /--- >----- TE----------.----.M Only equ red to e pe ormed nce a of eac cl ing, ~~~__'#' 0d' /A ,....._______________-_-----,g _ . , Verify seal leakage rate is s scfh 72 hours a Nee I when the gap between_the door sea is pressurized to a 2;;.:: ;H euf air ,./rf~f.// /,a ur 3.0 pssh }l?h] SR 3.6.5.2.2 Verify drywel lock seal air flask 7 days pressure is a psig. l SR 3.6.5.2.3 -------.---- -...-NOTE------.--.-.------- : Only required to be performed upon entry l into drywell. .........----.....--..-.-.-.........-.... l Verify only one door in the drywell air 18 months lock can be opened at a time 1 SR 3.6.5.2.4 --------------NOTE-----------.------- 'h---inoperable air lock door does not invalidate the previous successful performance of the overall air lock leaka e .;t...ge test. / I1.85' Verify overall drywe , air lock leakage 18 months rate is s Q @ overall mir ' ock leakage test atscfh by performing an am 1. : ::n. gswl , i (continued) i . 4, PrTo,. 5 p , fomance. d %e en a ll N+ at 1 3. o psial , 4ke wl.ck- sht u L e P4 prestv/i teJ -4e 19.2.p(.cl. _ c +- BWR/6 STS 3.6,-58 Rev. O, 09/28/92 , <GF <C - 161 ) j GJOCAL S PEH OM T C \\ [$o& 4t1 N D/Vv) -> Styvell sol ho" ' ~ 3.6 CONTAINMENT SYSTEMS 3.6.5.3 Drywell Isolation Valves - 1115ETCT I)ila FIOTL 4 l LCO 3.6.5.3 Each DIV shall be OPERABLE. 4. Edee a pplict.ble. Co,d, b s u d ikqv,,,,l AeLm J 4C0 26.5.1, APPLICABILITY: MODES I, 2, and 3. " Drpe ll . wk e4 OW leakage ' res u l-f s !, exceed g over.all i ACTIONS h, . i drywell b yp is lea kage r.c.4e cr ele n.u. .....................................N0TES.---.- ----- -........-.....-.......
- 1. Penetration flow paths ma be unisolated intermittently under administrative control .
2. ,e vee m ...p 4. .m_, m p ( a, p ,p v.l , O;; Separate Condition entry is allowed foF each penetrationllow path.
- 3. Enter applicable Conditions and Required Actions for systems made inoperable by DIVs. '
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected [peners for 4 penetration flow paths penetration flow path , ion flow with one DIV by v52 of nr /ca2r paths t inoperable. ] cne clo5e4 and has in de-adivafe0 d ter r 8 ', - f, 3.b auf anoM close.C volve,lue eanvai va ,g L ; SQ blu ct 4lcap, or L. G dect value whh flou) ~ 8_ hours) for 'l \ +hrevf +bevu M net tion f 1
- c urec0. pa s :} ;
inc ; diame r11 ( M (continued) l l l BWR/6 STS 3.6 60 Rev. O, 09/28/92 1 /,4 Ch// ) O!Vs 3.6.5.3 .e ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 -- ----. NOTE-- --..-. {p:n r: :::.-y .. 1.5dg o[ar may EV) LE-T be verified by use of administrative means. Verify the affected Prior to penetration flow path entering MODE 2 is isolated. or 3 from MODE 4, if not performed within the previous 92 days 8. hlyappli TE--------- B.1 Isolate the affected Ihourfor On le to . penetration flow path penetration flow ,enetr.ation p paths l hy v5c o' a+ Ic ort paths't ] with t isola n , s o.s.: cne clo JuC ad jnches N[ valves. s , i sg de- oc'Ea /e8 ' s A
- e. . . . . . . . . . . . . . . . . .g g,
'f ' a vfa na M vo/cte, c! scd P.nucel JAbc, % i' - i One or more ' , 4 hourDfor penetration flow paths h /.ug
- c. //h{ J/m
,. gw',j,$[J, (penetratlon flow ' with two DIVs e vg/ge paths < [ '] inoperable. ,4f g/g 4 ,# g e c f.c , (nchesy ; , i C. Required Action and C.1 Be in MODE 3. 12 hours i associated Completion --- l Time not met. M + ! C.2 Be in MODE 4. 36 hours , BWR/6 STS 3.6 61 Rev. O, 09/28/92 kM 93-n'21) OIVs 3.6.5.3 { SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Og SR 3.6.5.3.1 Verify each inch drywell purge 31 days isolation va v is sealed closed. y f- _ SR 3.6.5.3.2 QNw*QT - - - -- . - . - . - - . - - -- - NO T E - - - - - - - - - -/- - - - - - - - - r-~-- Not required to be met when the %rywell N h'"F %=g_gskyr;- 5-; . =----e for pressu;r,e control, ALARA or airvalves are openj 8) 5 quality co sidcrations for personnel !ufy;(( g L\ entry, be urveillances6that require the pr s,fw,[/e - C o,, open .cr - _:_----- , -f,, /,,y /,e J jw,,~4 ) valves ___~_-_.e___-....., W , e .= =. u h i 1 1 2 .' ' -- ,,,,,o ;,s - M'y^/,yA = :... : . # 1:M eaEm.a e-A E ~ This Specification ensures that the performance of the primary containment, in the event of a DBA, meets the assumptions used in the safety analyses of References 1 and 2. SR 3.6.1.1.1 leakage rate requirements are in (continued) RIVER Br2 --BWR4 STS - @ t'. B 3.6-1 Rev. O, 09/28/92 1. (LM 9:'SI) Primary Containment-B 3.5.1.1 BASES l BACKGROUIS (continued) conformance with 10 CFR 53, Appendix J (Ref. 3), as modified by approved exemptions. APPLICABLE SAFETY ANALYSES The safet9 design basis for the primary containment is that it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate. The DBA that postulates the maximum release of radioactive I material within primary. containment is a LOCA. In the i analysis of this accident, it is assumed that primary j containment is OPERA 8LE such that release of fission products to the environment is controlled by the rate of primary containment leakage. , Analytical methods and assumptions involving the primary containment are presented in References 1 and 2. The safet y analyses assume a nonsechanistic fission product release following a DBA, whtch forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERASILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded. ,r - /5.62 3e maximum allowable, leakage rate for the primary containment (L ) is g&% by weight of the containment ; e ,7 and drywell ai,r er_24 hours at the maximum peak containment JI 61 pressure (P,) of w psig)L(Ref. 4). r 7G ' Primary containeen s isfies Criterion 3 of the NRC Policy t Statement. LCO Primary containment OPERA 8!LITY is maintained by limiting leaka f w m .w., ? p gr.ge= top'Tsir.- 3 (" i 3} f =: Comp':nem;lance .s crwith ters this LCO will ensure a W1 mary containment configuration, including equipment ff'g ,i ' ._ 1atches, that is structurally sound and that will limit 1 - *N leakage to those leaka e rates assumed in the safety analysis. Individual eskage ' rates specified for the s ' primary ~ containment air lock.s are addressed in LCO 3.6.1.2. ~ _Q = ./,P lo., enepf k n vat i., m rary a-w x p:- a w ,ep,re fpieci/ lwe$cn n?, A,'y e 4 7, \( e~' a A+ fib /,Je . de /w %se 2 c<n.Q C /cen e % / i:e ; .,m+ < o. 6 h , ~A i tc omt/ T,re y ,4 /ea,y eo'ntinuen) ,:e x o. , r A BWR/6 STS - ' 8 T 6-2 ev. O, 09/28/92 -.-_------_--__---_u---- e. --mm- - ' ' - * ' M' )M 9$5f/[' Primary Containment B 3.6.1.1 BASES (continued) (\e.d. g ( \ 3 ~__ , - APPLICA8ILITY In MODES 1, 2, and 3, a D8A could cause a release of radioactive material to primary containeen In MODES 4 and 5, the probability and consequences of reduced due to the pressure and temperature hose events are , these MODES. Therefore, primary containment imitations of to be h in MODES 4 and 5 to prevent lea age ofnot required , toacti e material from primary containmentet- 1 Fever 4, L t.o R I.10 ~ 3 "Piim*3 C A=G ACTIONS L.1 '
- u. m.g In the event that primary containment is inoperable, primary containment 1 hour. must be restored to OPERABLE status within The 1 hour Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining primary containment OPERA 8ILITY during MODES 1, 2, and 3. This time that the probability of an accident (period alsoprimary requiring ensures .
containment OPERABILITY) occurring during periods where primary containment is inoperable is minimal. 8.1 and 8.2 If primary containment cannot be restored to OPERA 8LE status within the associated Completion Time, the plant must be l brought to a M00E in which the LCD does not apply. To achieve this status, the plant must be brought to at least MODE allowed Com 3 within 12 hours and to MODE 4 within 36 hours. The experience,pletion Times are reasonable, based on operating to reach the required plant conditions from fu'I power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE -SR-3.6.1.id REQUIREMENTS Maintaining the primary containment OPERA 8LE requires e ! compliance with the visual examinations and leakage rate * ,- test requirements of 10 CFR 5 'A i modified by approved exemptio0, ns. Appendix J (Ref. 3), as Failure to meet air lock h leakagetesting(SR3.6.1.2.1andSR3.6.1.2.4 e--2 = t L;= help (5:: 2 ' > M ),P.res ~( rimary containment _ purge valve leakaoe testinge J _m-('5R '3 6.l.3.f) , seawk ~ ~"y cord..ded " bpss leakaje (SR 3.lc.l.3% ~ (continued BWR/6 STS B 3.6-3 Rev. O, 09/28/92 I w -, [ LAP @H ot } g Primary Containment p _ 8 3.6.1.1 BASES ( krau (akacge [% h).3.th Om - ~ , /W i.s f , . SURVEILLANCE SR 3.6.1.1.1 ' o i /&), ed 'Y QlI REQUIREMENTS y kaW/ 318 ' y , main steam d:M -- - ]-- leakage posah6 s , a: e -n ~ (SR 3.6.1.3. 9 , % hydrostatically tested valve leakage e/ W,,j j-((SR3.6.1.3. sdoes not necessarily result in a failure of Inis sa. f "'7"* '" me Tupact of the faflure to meet these SRs must be evaluated against the Type A, 8, and C acceptance criteria of 10 CFR 50 Appendix 1.(Ref. 3).f The Frequency x=T 4,0 ' ' is required by 10 CFR 50, Appendix J, s modified by approved exemptions. Thus, SR 3.0. j/l extensions) does not apply. which allows Frequency s SR 3.6. 1.2 The st etural integrity .the primary ntaincent is ensuri by the successful completion of he Primary Cont # neent Tendon Surve 11ance Program and by associate visi41 inspections of t e steel liner d penetrations r O81 evi 'ence of deteriorat or breach o integrity. Thi ures that the stru ural integrit of the primary c tainment will be intained in a ordance with the revisions of the P mary Containee t Tendon Surveil ance regram. Testing d Frequency a consistent with the recommendations of Regulatory Gui 1.35 (Ref. 5). REFERENCES h 1. R.Section).27 h 2. h . SectionJ 15.6.57'
- 3. 10 CFR 50, Appendix J.
h
- 4. . Section '2. fo t
@ e.=memeus.==.) =m n h. kl Y tttKdes ff/O f YO Y # ' h ' npse0 WcfR to, Apessu f 7, / camp e +ed a reyw"O i y he +o ;ad a b Aa 0< c&eaf fpe 13 xd C j.a, ~ N 0,7 C k for overall 1}pe e, di M \ l o8er h es ba n en rep,leS l' axe-eAtan /eam% h,d'= laceerlaue cibirk is case 0 ou aY overa// fype d A'arsy ' \ i \ u&llcw l4h} O ja4. .wawp v s /, o / ^ 14 : ods w ' n ~, 4 & o/ rAe .:A'Y' A MQ aaa lse,v ' y - _] BWR/6 STS ~8 3.6-4 ^ nev. O, 09/28/92 , f: n utvly Primary Containment Air Locks 8 3.6.1.2 83.6 CONTAINMENT SYSTEMS B 3.6.1.2 Primary Containment Air Locks BASES BACKGROUND Two double door primary containment air locks have been , built into the primary containment to provide personnel access to t% primary containment and to provide primary i containment and exit. The isolation during the process of personnel entry air locks are designed to withstand the same loads, temperatures, and peak design internal and external pressures as the primary containment f the primary containment, the air lock (Ref.1). As part of Mhh 2 limits the release of 471 radioactive material to the environment during normal unit operation and through a range of transients and accidents up \ to and including postulated. Design Basis Accidents (08As). x Each air lock door has been designed and tested to certify s its ability to withstand pressure in excess of the maxinam s, expected pressure following a DBA in primary containment. ~ Each of the by 460]Cpsig doors has inflatable the seal air flask andseals that are maintained neumatic system, @ @Wich is maintained at a pressure a) }"psig. Each door has two seals to ensure they are sing a failure proof in maintaining the leak tight boundary of primary containment. Each air lock is nominally a right circular cylinder,10 ft 2 inches in diameter, with doors at each end that are interlocked to prevent simultaneous opening. The air locks-are provided with limit switches on both doors in each air ck that provide control room indication of door nasition. O\ the e ee k r f Ldefea .; rlgiri puring periods when' primary containment is not i be OPERABLE, the air lock interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent primary containment entry is necessary. Under some conditions, as 'al'1 owed by this LCO, the primary containment may be accessed through the air lock when the door interlock mechanism has . failed, by manually perfoming the interlock function. The primary containment air 16cks form part of the primary containment pressure boundary. As such, air lock integrity and leak tightness are essential for maintaining primary containment leakage rate to within limits in the event of a (continued) BWR/6 STS B 3.6-5 Rev. O, 09/20/92 torny o o w c d k ir fows (g)g qg, y Q 9, b. f 2 & /;w f,'/, Vy i INSERf B7A g , [. Therefore, maintaining OPERABLF orimary containment air locks in i MODE 4 or 5 for to ensure whicha control volume [leasess onlyofrequired during 3 17 situations significant re radioactive material can be postulated; such as during operations with a potential for draining the reactor vessel (OPDRVs) , during CORE ALTERATIONS, or during fuel movement of irradiated fuel assemblies in the primary containment. INSERT B7B \ 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 i lock is governed by subsequent Condition entry and application of associated Required Actions. 4 l 1 6 8 [ ' l 4 f INSERT RIVER BEND B 3.6-7 10/1/93 (MR 93-/YN ) Primary 5.nc.4. k bbovet.\\ qr * *r ce L ~ tsb can ainment Air 1.ocks . \ e. . hge r e e. u sea 8 3.6.1.2 i 3a V bk k M*CC b L A J BASES (continued)> m *) " W *b A i nq he re m mo E i T SURVEILLANCE ( a N 3.6.1.2.1 i SR p REQUIREMENTS Naintaining primary containment air locks OPERABLE requi gE^ ' ' compliance with the leakage rate test requirements of 10 CFR 50, Appendix J (Ref. 2), as modified by approved exemptionst N o E.S % This SR reflects the leakage rate testin requirements with regard to air lock leakage (Type 8 eskage , 3',g 1 r 5 4 T#9 The period <c testing requirements verify that the pyg7 air locTc leakage does not exceed the allowed fraction of the overall primary containment leakage rate. G/M required by exemptions. 10 CFR 50, Appendix J as modified by app extensions) does not apply.Thus, 45 SR 3.0.2 (which allows 1 The SR has been modified by two Notes. Note 1 atates that an inoperabl.e 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 d capable d onng~opea b. of a otA. of providing a fission product barrier in the event ! sults to be evaluatedNote 2 has been added to this SR, req u s _ mo0E5 k1 4 33. i _ SR 3.6.1.1.1F This esu%ainst the acceptance criter' a of = res that air lock leaka e is ) properly accounted far in determining the overa 1 primary containment leakage rate. 4 _ SR 3.6.1.2.2 ~ The seal air flask pressure is verified to be at a490fpsi every 7 days to ensure that the seal system remains viable.g It must be checked because it could bleed down during or following access through the air lock, which occurs ~/ / regularly. 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. SR 3.6.1.2.3 , The air lock interlock sechanism 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 (continued) BWR/6 STS B 3.6-11 Rev. 0, 09/28/92 h A$fo /f g'/ ) INSERT BilA ... (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 O'E'g,g isolated by use of (for this Specification) one closed and locked air lock door. The leakage rate of the isolated bypass 4%li leakage path is assumed to be the actual pathway 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 a of quantifying maximum p'thway leakage is only to be used for this SR (i.e., Appendix J maximum pathway leakage limits used to evaluate Type A, B and C limits are to be quantified in accordance with Appendix.J). During the operational conditions of moving irradiated fuel assemblies in the primary containment, CORE ALTERATIONS, 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 applied to the air locks. In these operational conditions the reactor coolant system is not pressurized and specific primary containment leakage limits are not imposed. 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. INSERT RIVER BEND B 3.6-11 5/3/94 (MR 92IVI) '^ PC8Vs W .6 .1.3 ^ i olenN! No~pdk hyke R J L /Ao~" (o* & BASES (continued)y,,y / a;f.n/ p -f, 4,- A ic yys ,/ < l e%ge/,4~ ~y n~ ~s/e -r APPLICA8ILITY / In MODE 5 I, 2, and 3, a DBA could cause a release of / radioactive material to primary containment. In MODES 4 f and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of ' these MOD Therefore, most PCI h are nne renuired en h.
- d it: ;& =; aft 51.; ;;,;r;; .i;_; r f, ,f, ; /
PS :; 5: ::J; -- ::j g- valves are requ_.._.: ; .:.... - " ' ' ce rtai n .ire m 5 to De opt RAsLE, hyever .. g ~ ;- H g ;revent _ r - ___g r:0314;. : J. ' ' ,_ p 95 y : ;"...:.. p nesrWivfr1 ;;; ; e1 ... .M.. iliE....%t; ' .f. M.. n.s-b wEoie assoctaten 4/ insnumTntation is required to be OPERA 8LE acco \G ) LC0 3.3.6.1, " Primary Containment 4 Isolation mad #25 d>~e# - <N"'/* O This does not include the valves that isolatetheass8ciaedinstrumentation.)Instr . ACTIONS The ACTIONS are modified.by a Note allowing penetration flow pa<;h(s) b re;t 7;r 7.; i . ..e y , i . , ces;.e..x..; e.rs) 6 : n ' M O d S ): Jto be unisolated intermittently under~ administrative controls, f T prima containee pu Valve xception applies to imary tainment rge va es re not querified to lose un r acciden conditi ns.] ' These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicsted. et the size o theT Fcontainee purge line penedi , penetrati ns exhaust rectly from the inary cont neent n and th fact that hose atmosphe Po to the en ronment, the pen ration flo path contain ng these va es may not be o ed under #-- admini rative con is. A single p rge valve i a penet tion flow th may be opene to effect it_no able valve .,a s allowed by t e Note to S pairstoanj 3.6.3.1. f , A second Note has been added to provide clarification that, , for the purpose of this LCO, separate Condition entry is , allowed for each penetration flow pa g g D gg 31Ae ACTIONS are modified by ^ :t'. Notel whseh ensure / ! appropriate remedial actions are aken, if necessary, if the affected systan(s) are rendered inoperable by an inoperable (continued) BWR/6 STS B 3.6-16 Rev. O, 09/28/92 - - - - - __._._m - _ _ _ . _ _ . _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ - - _ _ _ - _ - _ _ _ _ _ . _ . _ _ _ _ _ _ (T M 92-MA h PClys B 3.6.1.3 8ASES ACTIONS A.1 and A.2 (continued) drtwe h d lit v emeO Q i m h ontainment ano capesse or oeing stspositioned are in t correct position. The Completion Time for this verification of "once per 31 days for isolation devices outside primary containment, drywell, and steam tunnel " is appropriate because tne v are ..... operated under administrative CD dewas Fo trols and tTie probability of their misaliennent is low. S 4 period a " rior to entering MODE 2 or,"A from M performed within the previous 92 daystJis based on engineering judgment and is considered reasonable in view of the inaccessibi' ity of the 4"'-~ 7 and administrative controls ensuring that existence of other Oct misalig it M, an unlikely possibility. Ns do.o Con tionAismodifiedbyAnot indicating that t ied wA\h Cg ' Condit is only applicable to th Vs. For penetration flow penetration flow path hwithtwo dition C vides appropriate, Require ths etions. with one PCI.V. I % obh ., Co Bgquired Action A.2 is modified by a Note that applies to d "t sJ ~,= ... ..a en ; C . located in high radiation areas and allows tM'td pe~ verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically ~ astricted. Therefore, the probability of misalignmen ' ani.;.Cd proper posit on.they once have been verified to be in the is law. 4.xc.eg\ due.he \eahge o n h MN NO - RJ With one or more penetration flow paths with two PCIVs inoperabl either the inoperable PCIVs must be restored to p /, : OPERABLE status or the affected penetration flow path must gn be isolated within I hour. The method of isolation must i g include the use of at least one isolation barrier that ' cannot be adversely affected by a single active failure. [g gf~., f. : Isolation barriers that meet this criterion are a e de-activated automatic valve, a closed manual valve d i i blind flange / 4The 1 hour Completion Time is consistent w . theACTIONSof}LCO3.6.1.1. \ ^ Q , g pondt n B is modified by a e indicati g this Condit gsonly QeJbletopenetratio flow paths 9fth two PCIVs. (continued) BWR/6 STS 8 3.6-18 Rev. O, 09/28/92
- -Nat )
PCfVs 8 3.6.1.3 , BASES ACTIONS (C 1 and C.2 (continueo) - b restricted. herefore, the probabili of misalignment o " t'hese valves, ce they have been verifi i l_Droper position, s low - to be in the ( Sn%b h qist biage idte, %d roska _, Magt, ($ cr/ @ $h/ bby tw\ e i With the secondary containment bypass leakage rate %ot N within limit, the assumptions of the safety analysis met. Therefore, the leakage must be restored to within not limit within 4 hours. Restoration can be accomplished by 1 isolating the penetration that caused the limit to be exceeded by use of or.e closed and de-activated automatic valve, closed manual valve,.or blind flange. When a penetration is isolated, the leakage rate for the isolation penetration is assumed to be the actual pathway leakage s J'/!' , through the isolation device. f;/ If two isolation devices are used to isolate the penetration, the leakage rate is assumed f to be the lesser actual pathway leakage of the two devices. The 4 hour Completion Time is reasonable considering the 7' g,o time required to restore the leakage by isolatin M penetration and the relatiye immrtance que:::=g_the ^? -]f ' ,, c2aeir ruaron. rA ;;y;;;;: Me:ipto tie overall contaTrWUfn't _ i -:7. 9./, 3 h .2. and 3 ( ~ In the event one or more tontainment purge valves are not within the purge valve leakage limits, purge valve leakage must be restored to within limits or the affected penetration must be isolated. The method of isolation must be by the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isol ion O$1 - barriers that meet this criterion are afclosed and (5 - f/ V de-activated automatic valve, closed manual valve d i J flange}.9Fpurge valv th resilient seals zed to. st san sry Required Actio 1,%ust have Deen strated to I o meet the leakage requir nts of SR 3.6.1.3 The specified Comp'etion Time is reasonable, co ring that 'oneicontainment purge valve r'emains closed & f;r : in ;,7 WC ;A .[ , does not exist. m y , so that a gross breach of containment fon tf - - ~ fz e,,/. ' y (continued) BWR/6 STS B 3.6-20 Rev. O, 09/28/92 l (fA 92- e WI) PC8ys B 3.6.1.3 8ASES ACTIONS .1. .2. an .3 (continued) -hde - In accordance with Required Action 2, this penetration i flow path must be verified to be i olated on a periodic g r e-a ff basis. The periodic verification is necessary to ensure , _I thatJcontainment penetrations required to be isolated following an accident, which areJo longer caoable of being i fgg,3 automatically isolated, will be G -- rz: n:: ::n:n jg should an event occur. This Required Action does not gg require any testing or valve manipulation. Rather, it ; involves verification.Cz- ;G S:M that those l isolation devices outsi % contai._..t and potentially capable of being mispositioned are in the correct posi1 1 For the isolation devices insides:entainment he time ) period specified as " prior to entering MODE from MODE ! if not performed within the previous 92 day is based o i engineeriC ' red reasonable in view of ve contro s t w T' ensure ~that isoiat devica misaligammat is an unlikely possibility. on y N isolated or ontainment purge valve with resilipnt seal t is cordance with Required Actiorf.1, % SR 3.6.1. t be performed at least once eve assurance that degradation of the resF) days. This prov 'ent seal is detected and confi ms that the leakage rate of the wontainment purge valve does not increase during the time the penet is isola 1 . The nomal Frequency for SR 3.6.1 54 days. C W .. ; r e r T i-itu"w Since more reliance is placed o , on a single valve whf in this condition, it is prudent to perform the SR more 'o .'Therefore, a Frequency of once pe & days was chosen and has been shown. acceptable based on ting experience. __. @ If any Required Action and associated Completion Time cannot be met in MODE 1, 2. or 3, 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 withini36 hours. The allowed Completion Times are reasonable, based on operating experience, to rwach the rwquired plant conditions from full (continued) BWR/6 STS B 3.6-21 Rsv. O, 09/28/92 ,~ ,, . , , , . . . _ _ _ I yg c, 6, /.1 / { ( 9;'-h'd / ) . J INSERT B23A If a purge valve is open in violation of this SR, the valve is -considered inoperablQs . @ndit.ica A-- suvMesh -ano--u_niessM --ootherwise known to have excessive leakage when close M Q t_ / ( dered to have leakage outside limi g fgendit w -- \ i . % j() u/r}e int ~atle nk u* en f - t _. s (p } , (9,l O ' ( 3 4T l I e INSERT RIVER BEM B 3.6-23 10/1/93 - - . . _ , - . ,, ,. % - , , - .c- .,r- y . . . , [w a: iGI ) [.%) \ Ccw% mch hdeM,w r gA M.g PCIVs a 3.6.1.3 C t I. s ' w.,A A be reptJ h %. ~1M.3. 9, u/ BASES ' ~ m _ _b'N1 _ x l SURVEILLANCE SR 3.6.1.] .h (continuedh" I' E - REQUIREMENTS @::::L_::J-required to be capable of At the en the purge valves are i ,b,p - / irradiated fuel assemblies) ization ing pressur(e.g., concerns are during not / mo s p ~ present and the purge valves are allowed to be opent ~ g +ct The SR is modified by a Note (Note 2 stating that the SR is : not required to be met when the purge) valves are open fo the stated reasons. The Note states that these valves may h.;V-- , ;p#d. v ,.' . - / be opened for pressure control, Al,/LRA, or air quality consicerations for personnel ent ji " ~"'f t purp re;' w that require the valves to _be -- ..nSurveillances q * ,
- g
- : :-::::q mo_:: _._ e m : g::;
_ ; = fd:: '- 83 erg. - s y,u//ce.;), hq L- drimary contiTriment pi#g'e'v'alves a ' re capable of closing in ' the environment following a.LOCA. Therefore, these valves S N' are allowed to be open for limited periods of time. The 2tp 7 pc#y Q _..:n 1 day Frequency is consistent with other q* P_ _ t 7 _. ::G requirements, =^-~:raa- r: +c I p ,~n, ~ ^ ' ' ~ s' - ~ 3 or sydal 4eshc3 h h SR 3.s.1. - "N M5 9% Nd 3 4 4 a c"'" ' This SR verifies that each rimary containment isolation duf uf t, .scrf e,6e manual valve and blind flange that is located outside primary containment, drywell, steam tunnel, and is a,e y,/ ,p,, ,7 ,J2 /, /E.,y hmquired > to be closed during accident conditions, is closed. a # " " " / " /" ~.'"j The SR helps to ensure that post accident leakage of ig reu.a.,, 4 e d # #" '. radioactive fluids or gases outside of the primary containment boundary is within design limits. This SR does /rLf ,~c h s / ,,p , not require any testing or valve manipulation. Rather, it 4 ,,/,A 4, //i 5" :.. . Q that those Q,f,,,1 fy,,ga,' V h@M involves mespe+ves)outside verificationp.. primary conta' ,_. 59 anc capable mispositioned, are in the correct / position. _; : of being , 1 ~ Since kd verification of 625p position for @ outside primary containment is relatively eas/,\the 31 da Frequency was dem 6el chosentoprovideaddedassuranceithatth)e,are in the Positions. @ demees es are adde <1 to this SRMf The fled 55Lte applies to sce. aid ' ' pisaur vavesandblindf'angeslocatedinhighradiationareasan allows them to be verified by'use of administrative f 7 62# A controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA (continued) BWR/6 STS B 3.6-24 Rev. O, 09/28/92 (, A M a/r/) PCIVs B 3.6.1.3 ! BASES i SURVEILLANCE REQUIREMENTShSR 3.6.1. (continued) gi N i p N reasons. - Therefore, the probabil ty of misalignment of (llrgb thes C - , cace they have bee verified to be in the ' i proper posit Kis low. A Note is included to ' not required to eQme;h he SR during the istime are thecla! are ws - M l SR 3.6.1. .h ' b M U This SR verifies that each primary containment manua N bnet ' i isolation valve and blind flange located inside primary containment, drywell, or steam tunnel, and required to be closed during accident conditions, is closed. The SR helps ! to ensure that post accident leakage of radioactive fluids oresign gases outside the primary containment boundary is within limit (t chwen Frequency o rar"q!!gs inside primary containeen for fa entering MODE 2 or free MODE 4, if' ! not oerformed withir. the previous 92 day is appropriate since thehCR~TIIEps are operated under 1 % g / administrative misalignment is 100. controls and the probability of thei 1 ^ _ _ - N M - % Notes are added to this D The e 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. dm5 probability of misalignment of these ]There,ore, the once tne been verified to be in their proper m_sition, is low.y have A -}Md Note is included to clarify taat e that are open under administrative controls are not require to meet the SR during the time that the are open. PCIVs h SR 3. 6.1. . S' ' Verifying the isolation time of each power operated and each automatic PCIV is within limits is required to.. demonstrate ! ,n cnex tLiry, t / - i. L - I msivs coy k excIvM hm %& SK slave l":'V r - f hil closure. isolofior tmc is de<&shold by SR 36.12.6, Q-w l A ,3cl,hw 6 s td com m i t af the volvc w il/ w , ~ in a t ran pe, va le:s +4an or epal -lo +!,af a ss onC + 'Sc (continued) i i BWR/6 STS B 3.6-25 Rev. O, 09/28/92 l i l fM c.^. -N$ ) PCIVs B 3.6.1.3 BASES s SURVEILLANCEhSR 3.6.1. . (continued) REQUIREMENTS safety analysis. The isolation time and Frequency of this l =SRxxareMin" _ accordance with the Inservice Testing Program (D ~ , SR 3.6.1. ~ [ ' For primary containment purge valves with resilient seals, additional leakage rate. testin be of 10 CFR 50, Appendix J (Ref.g'(h. yond is required the test require to ensure OPERASILITY. Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than do other seal types. Based on this observation, and the importance of maintaining this penetration leak tight (due'to the direct path between , primary containment and the environment), a Frequency of i ,, _ J j 184 days was established.;; ;r er r.; I % .;; M = T i p e rte is r :-Z; Gei. O .T Additionally, this SR austehe performed within 92 days after opening the valve. The 92 day Frequency was chosen recognizing that cycling the valve could introduce additional seal degradation beyond that which occurs to a valve that has not been open(ed). / Thus, decreasing the interval measure after a valve has been(opened.from 184 days) is i / I # The SR is modified by a Note stating that the primary containment purge valves are only required to meet leakage rate testing requirements in MODES 1, 2, and 3. If a LOCA inside primary containment occurs in these MODES, purge g( valve leakage must be minimized to ensure offsite radiological release is within limits. At other times when the purge valves are required to be capable of closing ts h tt re) 4. (e.g., during handling of irradiated fuel), pressurization concerns a not present and the A econFnoteeis ac=;to purge valves requiarecF~~*% NE
- M this 5R 4
resus s to ng ug ,SR ev(luate(agai st th acce ance c iteria of N .6.1 .1. is sure that rima contai nt y ve 3 akag sp perly ccou ed f ge bby cn\ugy overaf pri ry c tai nt I age t indeferwini th (~s > , h,/r 3 4W (continued) BWR/6 STS 8 3.6-26 Rev. O, 09/28/92 _+w y . . .- -w- + - == [M %NtPCIVs /) B 3.6.1.3 BASES 16 SURVEILLANCE f SR 3.6.1.3 # REQUIREMENTS (continued) Verifying that the full closure isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY, The full closure isolation time test ensures n 6 /> D ' ' that the MSIV will isolate in a time period that does not N9 exceed the times assumed in the DBA analyses. The Frequency of this SR is din accordance with the Inservice Testing Program (r 10 h@. "- l'7 SR 3 . 6 .1. 3 .'B' Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.1.6 overlaps this SR to provide (/ complete test. of the safety function. Theef187-month Frequency is c i on the need to perfonn this Surveillance under the condit ms that apply during a plant outage and the potential for an unplanned transient if the Surveillance were perfonned with the reactor at power. Operating experience has shown that these components usually pass this (dp 2 Surveillance when performed at the 418K month Frequency. < Therefore, the Frequency was concluded to be acceptable from ' a reliability standpoint. , JN1.5 > i i M SR 3.6.1.3.9 ' y , This SR ensures that the leakage rate of. secondary containment bypass leakage paths is less than the specified leakage rate. This provides assurance that the assumptions in the radiological evaluations of Reference re met. The leakage rate of each bypass leakage path is assumed to be the maximum pathway leakage (leakage through the worse of the two isolation valves) unless the penetration is isolated g h/ v by use of one closed and de-activated automatic valve, closed manual valve, or blind flange. In this case, the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two valves. This method of quantifying maximum j (continued) , BWR/6 STS B 3.6-27 Rev. O, 09/28/92 ()bff /Y/[) '3,(pl.3 INSERT B27A bN SR 3. 6.1. 3 JT i The use of MS-PLCS as a positive leakage barrier results in in-leakage and gradual pressure buildup within the containment. The total allowable MSIV in-leakage rate does not have radiological consequences. This surveillance ensures that the total allowable air in-leakage rate from the MSIVs and valves served by the PVLC3 is limited such that containment pressurization does not exceed 50 percent of the design value in a 30 day period due to these sources. 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 that these components usually pass this Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. 4 l i l 1 i INSERT RIVER BEND B 3.6-27 d$9953EP , l I ~~ (b/9:-!HI) FCivs q, h- B 3.6.1.3 , 8ASES i SURVEILLANCE ~ SR 3.6. 3 ( tinued) - REQUIREMENTS N pathway leakage is only to be used for this SR (i.e., ! Appendix J maximum pathway leakage limits are to be Mitgived h 10CFR50 3 quantified in accordance with Appendix J). The 61s = d_ ydt, s (M.4h os Frequency is_Jiased on snw n == iv ywr io snu 5 riinante M J'yWU*J v %nder ne conditi s that a ply during plant o tage and 1 the po ential foi an unpla ed transi ue pby Sa , 3R,3.o,t(hchalu Mre erfomd th the actor at p r.t if the Op urveilla ce ating jexpe ence has hown tha these c < 5 . '"*" y# Su illance en perfo d at the 18]nents mon u ally pas thi Frequene . 7 Th fore, t 0*[* Pf l 1' Frequen y was conc uded to / a liabili stando nt. r acceptab e fro j/ ,Lqty g{ /4dev uated ed to the of'sults ' ains /this SR/requiri the ace ptancegriteri SR to/be 6.1.1 . _ . _ . , s ens es th secon ry con inment ypass 1 roper ka s 7- accoipt ed for n dete ining e over 1 pr tr , conta nt 14akage r e.f ry _ \ ~__._. .- g ob b wg y I' We \lbues Sea \ vI e (fe l H eack f.viieo , SR 3.6.1.3 cm%s med \ I *i
- pg3 ;
The analyses in Referencesl2 and 3 are based on leakage ,is less_than the specifie , leakage trate. Leakage through hat S/'I I l i qil;_fra = m must be P. 2 ~ ~n _ij psigf'. The scfh When tested at / I leakage rate must be verified to / be in accordance with t eakage test requi n i f RefernnNe as~midff ed by appWWid exemption ' 2 - 1 6 d ~ Pfote is i kGKSS i tni 5R re irinfteria oftpe resu s to be evaluated g O gg J W ruin t he the ce e cr eskage is pro R 3.6 verallp[piry ntai t lea ger rde .1.1. ly a unted Jhis in / h q ,- s s The Frequency 1s required by 10 CFR 50, Appendix J (Ref. - fj ' as modified by approved exemptions; thus, SR 3.0.2 (which allows Frequency extensions) does not apply. ' ~~ s [ q .t.3 5R 36 3 '! J99 Surveillance of hydrostatically tested lines provides ; assurance that the calculation assumptions of References 2 \ and 3 aru met. fY6tEthat ds function valves st pass all 3 j applicah e SRs, incid ing the . f 1*=kaa* rat latt d ! SR 3.6.1M 11 M ap . riate The combined leakage rates t l must be demonstrateA)e in accargance wttbyleakage - 1 (04 Se krq a q d b ) (continued) x --- BWR/6 STS B 3.6-28 Rev. O, 09/28/92 ![/4/ 9:- I'!E l ) INSERT B28A. (2 places) 1/ / is added to this SR which states that these valves are / ' #oniy required to meet this leakage limjt in MOPES 1, 2, and 3. ( .l In the other conditions, the dactor poolant Aystem is not \ s pressuri::ed and specific primary containment leakage limits s are not required. \ s N - 9 g j, 'j % j *3 , & -l* A I J% y tltl t l INSERT RIVER BEND B 3.6-28 10/1/93 L -l'!!I) 4; , tj ,1, f PCIVs B 3.6.1.3 '.W 1 BASES sk l SURVEILLANCE SR 3.6.1.3 e tinued) REQUIREMENTS ' 4 - test requ (fA, ,w4; h exemption nts of Reference _ as modified b u b SR 3.02. NWI ch 6 my., $ a hnu, $ proved ( 3 og yptp/ ( ;, h w ' m,, Ol g t_,.__ nis 5 d eo h b u M: tee. A te 9 states that these v valves are only required to meet the combined leakage rate ) - f t in MODES 1, 2, and 3 since this is when the Reactor Coolant : System is pressurized and primary containment is required. \ In some instances, the. valves are required to be capable of i j automatically closing during MODES other than MODES 1 and 3. JE_;.u. senditions--i: th;indr "Pht:bs'~2$:nMM=p, 2 Q conditions.R ::t m2;gnihe~se-other~M00E$7 g g,,g ,\,,,'Ay g gre _J , Hote 2 is added to ,tnis a quirin the res its to be sNevaluafedetainsttheacce ance er eria o SR 3. 1.1.1 - Thi/ensu'stha account these v ve lea ges are prope y for i deterni ing the verall rima conta g sfeaka , rate. f > t b SR b l.3(12 / 9 i j / , . / Reviewe s Note: This SR is'only required for those plants e valves with re with p [ MODE 1, 2, ornd3)having duri plientblockin seals allowed.to be open } ._ val s that are not pe nently installed. /g devices on the - t rifying that each ] inch primary containment purge valve , . .s blocked to res ict opening to s (504] is required to i ensure that the the time limit lves can close under' D8A conditions within / and 3. essumed in the anal,yses of References 2 ( Ogg _ , l 1 The SR is ified by a Note st ting that this SR is only' required o be met in MODES (2,and3. If a LOCA inside prima containment occurs these MODES, the purge.v'lves a must ose to maintain c ainment leakage within the values ass y intheaccidentgnalysis. At other time ves are required to4e capable of closing (e (s., <during shen purge vesent of irradiat fuel assemblies), press rization t concerns are not pr sent, thus the purge val s can be fully / open. The[18) th Frequency is appropr te because the / blocking device, are typically removed o y during a refueling outage. L'~~ _ C (continued) BWR/6 STS B 3.6-29 Rev. O, 09/28/92
- a. . . . .- - . - _- -
(EAK 93 WIj o
- INSERT B29A
( I? : /q SR 3. 6.1. 3 1F i This SR ensures that the coJchined leakage rate of annulus bypass leakage paths is less than the specified leakage rate. G '0 'I' 4~ This provides assurance that the assumptions in the #97 radiological evaluations of Reference 4 are met. The leakage rate of each bypass leakage path is assumed to be the maximum pathway leakage (leakage through the worse of the two isolation valves) unless the penetration is isolated by use of one closed and de-activated automatic valve, closed manual valve, or blind flange. In th'is case, the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both c , G' 3 isolation valves in the penetration are closed, the actual /WI leakage rate is the lesser leakage rate of the two valves. This method of quantifying maximum pathway leakage is only to be used for this SR (i.e., Appendix J maximum pathway leakage I limits are to be quantified in accordance with Appendix J). l The Frequency is required by 10 CFR 50, Appendix J (Ref. 4), ( as modified by approved exemptions; thus, SR 3.0.2 (which , allows Frequency extensions) does not apply. i y s adcled to t sSRre[ iring 'he resul/s to _ s .ANote[da evaluate inst t accept nce cr' eria of R 3.6. '.1.k. This nsur that conda contai ent by ss. leak ge fss . ' pr rly ccountep for in determ' ing the overall r i M;ry c tain .nt leakdge rat -fe . w i i l i l l INSERT 1 RIVER BEND B 3.6-29 10/1/93 - l S' R 3 . 6, /, (p , / G M 9 3 I'! di ) INSERT B39A ... Also, adequate steam flow must be passing through the main , turbine or turbine bypass valves to continue to control reactor pressure when the valves divert steam flow upon opening. 444 ,__ 1 (, ,/. fo INSERT B3JJ The Frequency of 18 months on a STAGGERED TEST BASIS ensures that each solenoid for each S/RV is alternately tested. ~ w $
- INSERT RIVER BEND B 3.6-39 10/1/93
y __ .._ . _._ __ _ __ _ .
- . 1 (AfR93-Mf/}
Primary Containment Unit Coolers .B 3.6.1.7 Pb BASES (continued) l l L APPLICABLE- Reference 1 contains-the results of analyses that predict l SAFETY ANALYSES the primary containment pressure response for a LOCA with- l the maximum allowable bypass leakage area. The containment i unit coolers are not required for mitigating LOCA except in ' the case of steam bypass. ! i I The equivalent flow paph area for by) ass leakage has been ' specified to be 1.0 ft . The analysis demonstrates that with primary containment unit cooler operation the primary containment pressure and , temperature remains within design i limits. i c 4, /, y The 3rimary containment unit coolers satisfy Criterion 3 of : #y the RC Policy Statement. i i , LC0 In the event of a Design Basis Accident (DBA), a minimum of one primary containment unit cooler is required to mitigate 1 potential bypass leakage and maintain primary containment M an[c ,4d peak pressure below design limits. To ensure that these 4 Mu wM f9 bf i requirements are met, two primary containment unit cooler; must be OPERABLE Therefore, in the event of an accident. 3"'"." [d*, f/JkJ at least one unit cooler is OPERABLE assuming the worst case i single active failure. : l 1 APPLICABILITY In MODES 1. 2, and 3, a DBA could cause pressurization and i increased temperatures within the primary containment. In . MODES 4 and 5, the probability and consequences of these events are reduced due to the )ressure and tem erature ~ limitations in these MODES. T1erefore, maintaining primary. ' containment unit coolers OPERABLE is not required in MODE 4 or 5. , i I ACTIONS L1 With one primary containment unit cooler ino)erable, the inoperable subsystem must be restored to OPEMBLE status - wit 11n 7 days.' In this condition, the remaining OPERABLE ! L primary containment unit cooler is adequate to perform the - primary containment cooling function. However, the overall i (continued) l BWR/6 SlJPPLEMENTAL B 3.6-45 11 10/1/93 - , , - - - - - ,,,,.e---,,m ,,a , , -,n-,- ,- . , , . - . _ --.-c -. - _ _ _ . _ . _ _ _ _ _ _ _ _ . _ _ . _ - - - (ZAP 93-NRt} Primary Containment Unit Coolers B 3.6.1.7 9 BASES SURVEILLANCE ' f SR 3.6.1.7 (continued) REQUIREMENTS rather. it involves verification that those dampers capable 4 of being mispositioned are in the correct position. The 31 day Frequency of this SR is justified because the dampers are operated under procedural control and because improper positioning would affect only a single unit cooler. This Frequency has been shown to be acceptable based on operating experience. , h SR 3.6.1.7 Verifying each unit cooler develops a flow rate L.E50,000?-cfm ensures overall performance has not degraded during the cycle. Such inservice tests confirm component OPERABILITY. trend performance, and dete':t inc131ent failures by indicating abnormal performance. T1e Frequency of this SR is consister.t with that applied to pumps by the Inservice Testing Program. S'R 3.6.1.7 This SR verifies that each primary containment unit cooler actuates upon receipt of an actual or simulated automatic ' actuation signam The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.3.6 overlaps this SR to provide complete testing of h the safety function. The 18Nnonth Frecuency is based on the need to perform this urveillance uncer 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 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. SAR,Sectiond2.1.1.3.@ K Jrfteq saf A & eme<3 coy c,oers /,, ueme a,, Q ~~ d d the p aswie reZe q/' a&J2 M a Ac h p& aduh ' aaa k h #jh utred pa L el,ddo. BWR/6 SUPPLEMENTAL B 3.6-45 iv 10/1/93 1 L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ QAK 93 - /'/fl: l PVLCS B 3.6.1.8 BASES - - 4 ACTIONS C.1 and C.2 (continued) If the inoperable PVLCS subsystem cannot be restored to OPERA 8LE status within the required Completion Time, the plant must be brepght to a MODE in whien 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 con,ditions in an orderly manner and without challenging' plant systems. SURVEILLANCE SR 3.6.1.8.1 ' REQUIREMENTS
- The minimum air supply necessary for PVLCS OPERABILITY varies with the system being supplied with compressed air from the PVLCS accumulators. Due to the support s 4,f;g function of PVLCS for S/RV actuator air, however, ystemthe -
specified minimum pressure o 1 / "*j 4"2 g,,I : 01)*$sig is required, which proviites sufficient air for' -# 3/RV actuations M th " R g,,,,1 ; ;;;;,; ;; = ;;;ig. his minimum air pressure
- s
( /]caf-7- AOrAj, dione is sufficient for PVLCS to support the OP M * 'd these S/RV systems and is verified every 24 hours. The e /0 24 hour Frequency is considered adequate in view of other indications available in the control room, such as alanes, to alert the operator to an abnormal PVLCS air pressure condition. SR 3.6.1.8.2 ! h Asimulatedsystemoperationisperformedevery.4[18fInonths to ensure that the PVLCS will function throughout its operating sequence. This includes correct automatic , position <ng of valves once the system is initiated manually. ! Proper functioning of the compressor and valves is verified O3 by this Surveillance. The118FoonthFrequencywas ' developed considering it is prudent that many Surveillances be perfonned only during a plant outage. Operating experience has shown that these e nents usually pass the h Surveillance when performed at the 18foonthFrequency. Therefore, the Frequency was conc 1 to be acceptable from a reliability standpoint. I i (continued) BWR/6 STS B 3.6-48 Rev. O, 09/28/92 - - - - ~ r w E G AA 93*MYD Primary Containment-Shutdown B 3.6.1.10 ~ BASES BACKGROUND T (continued) C [1ministrative controls ensure that open vent and drain pathways will: (1) only be opened to support u,, I. /o , leakage rate testing: (2) not exceed 12 valves: (3) require /(_ #4 -/ monitoring opened vent and drain valves, as well as the / containment-to-auxiliary building differential pressure i every 2 hours; and (4) assure at least one person is . ghn. 19 _ signed to_each open penetration (Ref. 1). _ 'g e w pal ~ This Specification ensures that the performance of the 4 "384' N primary containment, in the event of a fuel handling o** accident. inadvertent criticality, or reactor vessel draindown, provides an acceptable leakage barrier to contain fission products, thereby minimizing offsite doses. APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it contain the fission products from a fuel handling accident inside the t'rimary containment (Ref.2), to limit doses at the lite boundary to within limits. The primary 9taugs containment (performs no active function in response to this event; however, itspC*EPf3ILI"' in cG6junctiOr, dt tt ;utc etic cl= urc . selected OP LE containment solatio /al s (LC0 3.6. 3 " Primary ntainment Isolat' Valves C/ 5 LC0 3.3.6.1 " " Primary Co ainment Isolatio I strumentati .- n;urc : '^? tinht ficci _ nervh u-t W ier. It leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage rates assumed in safety analyses. The fuel handling accident inside the primary containment is assumed to occur only after = 80 hours since the reactor was last critical. The fission product release is, in turn. I based on an assumed leakage rate from vent and drain valves / with a combined flow rate of 70.2 cfm (based on an assumed 1 g/ [ 0.367 inch water gauge differential 3ressure)]. This assumed pressure reflects the fact tlat the fuel handling dl accident 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 ad<iitional leakage of 0.26% of,the containment volume per day. , (continued) BWR/6 SUPPLEMENTAL B 3.6-53 ii 10/1/93 l l _ . 1 Ji4K 93 -!'// /) Primary Containment-Shutdown B 3.6.1.10 BASES SURVEILLANCE SR 3.6.1.10.1 (continued) REQUIREMENTS leakage of radioactive i " m di e i-gases outside of the Oct primary containment boundary is within design '.imits. The ; method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a i 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 that the required valves are in the correct position. The!M dov) Frequency was chosen to Op2 provide added assurance' that the valves remain in the ~ _ _ _ _ _ correct positions. -- NRCSafetyEvaluationReportforherBendTechnical f REFERENCES h 1. Specification Amendment #35 dated March 3.1989FA- ! -["EM9t.1cg@ciff g c;tf ca J.c.A..c A. dated - c.-; =- _ : . ....,. hh 2. USAR.Sectiond5.7.67 - ~ ~f g, $ f b pts e A / 5 9 / M rbe .sx' ii ,u./ reg a u e 0 40 ac -raf for *jj ~ veut a~.0 A ik /la pa/A~ sys p io 2 0 de 44/ ca/w/,/d S/n , ate + 4,,., 4 1 cyJe n ven / an.0 bar*n pufdwsyr il A 70.2 c h . f zAntsr pian p . D 3.&-s3 2 - i BWR/6 SUPPLEMENTAL B 3.6-53 y 10/1/93 f lL M 95iQ Primary Containment and Drywell Hydrogen Ignitors B 3.6.3.2 BASES BACKGROUlO (continued) Bl When the hydrogen ignitors are energized they heat up to a surfacetemperaturet'T1700}<*f. At this temperature, they ! ignite the hydrogen gas that ::. present in the airspace in the vicinity of the "gnitor. The hydrogen ignitors depend on the dispersed location of the ignitors so that local pockets of hydrogen at increased concentrations would burn befora reaching a hydrogen concentration sianificantly . higher than the lower f' ammability lisib/,tdi.;;- ir9 r e er: , , , , ,, , q . . _ t-t l- n! hydre-^ 'e='--t _^.tr. . .......-.--ae ter -
- --- x-- %
[6.oj ; _ er:: t C.5 t =: r:=t e ir. :"'.n f :.: t.,: -;= pre:e-tMew (co./.nsumedf ~~ ~ APPLICABLE The hydrogen ignitors cause hydrogen in containment to burn SAFETY ANALYSES in a controlled manner as it accumulates following a degraded core accident (Ref. 3). Burning occurs at the lower flammability concentration where the resulting - temperatures and pressures are re,latively benign. Without the system, hydrogen could butid up to higher concentrations that could result in a violent reaction if i random ignition source after such a buildup.gnited by a The hydrogen ignitors are not included for mitigation of a Design Basis Accident (08A) because an amount of hydrogen equivalent to that generated from the reaction of 75% of the fuel cladding with water is far in excess of the hydrogen calculated for the limiting 08A loss of coolant accident (LOCA). The hydrogen concentration resulting from a 08A can - be maintained less than the flammability limit using the hydrogen recombiners. However, the hydrogen ignitors have been shown by probacilistic risk analysis to be a significant contributor to limiting the severity of accident sequences that are commonly found to dominate risk for units with Mark 111 containment. The hydrogen ignitors are considered to be risk significant in accordance with the NRC Policy Statement. LCO % Two divisions of primary containment and dr hydrogen ignitors must be OPERA 8LE, each with /g N 00 0 @ % ss ONMBL . i en ' __ h~ #} (continued) BWR/6 STS B 3.6-80 Rev. O, 09/28/92 J)r - an) Q h .,ESecondary ContainmenQA-- 8 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1,{ Secondary Containment [ g BASES BACKGROUND The function dilute, and holdofupthefsecondary containment fission products that may leak from [is to contain, ~ primary containment following a Design Basis Accident (DBA). ' M &#nI_n conjunction with operation of the Standby Gas Treatment f4 9)(56T) System /and closure of certain valves whose lines 6W45 penetrate thejsecondary containment 7theAsecondary containment}-1s designed to reduce the activity level of the g fission products prior to release to the environment and to Q' h ' isolate and contain fission products that are released .# l$ during certain operations that take place inside primary containment, when primary containment is not required to be f n OPERABLE, or that take place outside primary containment. , g em iel4 bl. M .;. y- i Thelsecondary containment $- - H completely l g ,g ,,, D o l'l'a 9 _ g b ' g ".b ' . .,myabeGpostulated I ( encloses to contain the primary primary system fluid. containmen This , ' e 3 structure for1es a control volume that serves to hold up and m dilute the fission products. It is possible for the g' JN P P4 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 I allowingtheMsecondarycontainmentFtobedesignedasa n conventional structure, thet{ secondary containment}7equires Bt Q(Qfor^(9A i support systems to maintain the control volume pressure at less than the external pressure. Requirements for these f OP4 k systems are_specifie stely LC0 3.6.4.2, " Secondary Containment Isolatio (SC 6 )LCD 3.6.a M _- s andby Ga h t _T c o a. t, . u .64 , !b . e l d % . lel - 5 1" N'b M Y o^d1003M.5"LIELl'4WAINa _ i b s h e* . " ~ NAPPLICABLE There are principal accidents for which credit is h SAFETY ANALYSES e craatr taken for LOCA JPef. 1),fa econdary containmentV0PERA81LITY._ These are a i nandTing_ accident in~ s ide priiiiaW3 h nannt (Ref. ,f4 ^ TYuel handling ac~cTdent in the ; fu l siili4MP building' Ref.L . Thef{ secondary containment}Fh pe no active function in response to each of these limiting events; however, its leak tightness is required to ensure that the release of radioactive materf als from ihe primary containment is restricted to those leakage paths and , associated leakage rates assused in the accident analysis, (continued) BWR/6 STS B 3.5-90 Rev. O, 09/28/92 l TAR 93-tyR/) INSERT B90A The isolation devices for the penetrations in the secondary containment boundary are a part of the secondary containment barrier. To maintain this barrier: , g,q.l a. All Auxiliary Building penetrations, Fuel Building e penetrations and Shield Building annulus penetrations JII required to be closed during accident conditions are either: s
- 1. Capable of being closed by an OPERABLE secondary containment automatic isolation signal, or
- 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve or damper, as applicable, secured in its closed position, except as provided in LCO 3.6.4.2;
- b. All Auxiliary Building, Fuel Building and Shield Building Annulus equipment hatches are closed and sealed.
- c. The Standby Gas Treatment System is OPERABLE, except as provided in LCO 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 Luilding, Fuel Building and Shield Building Annulus is closed, except for routine entry and exit of personnel and equipment.
INSERT RIVER BEND B 3.6-90 10/11 4 [A AR 93-NR)) M SGeondary Containmen P l B 3.6.4.1 BASES ACTIONS C.I.I.2.andC.3 (continued) h h _ enent of irra isted fuel ass lies would n fficient reas n to require a, eactor shutdo . be a ,) SURVEILLANCE 3.6.4.1.1 REQUIREMENTS SR @p f g 3QQ w.w.c ,; Q _ Q- ' This SR ensures that the e eterrr r ::: :::_ Q/boundaryis sufficiently leak tight to preclude exfiltration under expected wihd conditions. The 24 hour Frequency of this SR was developed based on operating experience related to 8 :' g Jsecondary containment}Macuum variations during the applicable MODES and the low probability of a 08A occurring between surveillances. - [ Furthermore, the 24 hour Frequency is considered adequate in 3' /r' 1. / view of other indications available in the control room,. , #7 including alarms, to alert the operator to an abnormal ' _Nsecondary containmentP' vacuum condition. t SR 3.6.4.1.2 and SR 3.6.4.1.3 fjnha, , 6./y,, Verifying that'<[ secondary containmentfequipment hatches and e** ,, X* gJ,, access doors are closed ensures that the infiltration of / 1[. /, y3 og ccQoM, outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying ; l that all such openings are closed provides adequate io[/cax~/[d" L -- - assurancethatexfiltrationfromtheNsecondarycontainmentP will not occur. 4 Maintaining'1 secondary containment}^- i OPERAbsLITY requires verifying each door in the access . i opening is closed, except when the access opening is being * used for entry and exita ::: , e: :ee;; e..e 2 :. :.;; -- q t l ' (;;;;;;." The' 31 day Frequency for these 5Rs has been snown I to be adequate based on operating experience, and is considered adequate in view of the other i "'--- ' '--- gN ' Mtch st:t : 'M t :n evef 4 M ^ *Er :t r:tw. $ ccnkoIs onsecondy unhimfanos 0t'"HD ' h .45R 3.6.4.1.4andfSR 3.6.4. .I P4 ( y L d;,, m,L 2 l h The SGT System exhausts the atmosphere to the environment through appropriate treatment ! equipment. To ensure that all fission products are treated, (continued) l BWR/6 STS 8 3.6 93 Rev. O, 09/28/92 ; i @M95D o 8 3.6.4.2 l B 3.6 .CONTAI MENT SYSTEMS s 8 3.6.4.2 Secondary Containment Isolation SC BASES h h is L L o BACKGROUND @ The. function of the SC s in combination with other I accident release during mitigation sys Mas, is to Itait fission product and followin Accidents (08As) (Ref.1). g postulated Design Basi:; g Secondary containment isolation within the time limits specified for those isolation :hr--%<. designed to close automatically ensures that fission products that leak from primary containment following a that are released containment is notduring certain operations when primary requ' red to be OPERABLE, or that take lace ounside orinary containment. are main 1 Lined within C5 gEi: 'id ts. h se W .s @t%. 'A h.,J .,.g i @) The OPERASILITY requirements for 5 sdh c r adequate secondary containment c._ is .T w [<ef _ maintai . ,b7 ' 0 w ins and after an acc i N to tne environmentV t by minimizing potential /patuD** ocH)ersjoassive or active (automaisolation c). Manual device 6 ::,Fare de-acteithor iatedE / automatic *::h:; secured in their closed sitio U. ; 2 7 checkmatesFwith flow through the ureb and blind ij @b.nanees~ automatic'4vehes are considereddesigned passive devices. _=== Checkg & h , _ action following an accident are consideres acutve devices. rIsolauen ca.r ierts) for tne penetration are d< scusses uD to clos without operator e' Taference e GM ', 4/Y h Automatic-ei - 1 to v-close ask.khA ah*WNw @ on a seconeary containeFnt isolation Qw Ab Mseconear_y____CNitainment M--i;t untreated followingradioactive a OSA or other material Other penetrations are isolated b h a g erJ D closed position er blind flanges.y the use of valvefin the APPLICA8LE 'The must L o ensure *
- SAFETY ANALYSES contaidnin ece ry leek-tight barrier to fission product' gg releases The principal ageidents for whichrse)condary containment y:n tj C^r::5ris requires are a loss of coolant 1
accident (Ref. m r%snarsau . . m unsrae prints ; grainmetqRe". 3), fans a fuel handling accident in the (continued) BWR/6 STS 8 3.6-95 Rev. O, 09/28/92 ,-, , . - - . . , , . , - . - - - P rI,carg tenho,\werd is J j resseJ adapaheh in LCo eNM m@ - W.10 N sc g "& rdn arg kb emtak-3b"\ d
- w a i, Mada$
- B 3.6.
Wradi.h 9' *ed3*C he\ .33 e,,W,3 , M-he\ %e\ J;g Mbe achig BASES 6m A \ e A m% " req d *Nmee C" mh" h'"i *uUmbass o'bc.sf""\ah) u.hW C'^ N We tuei b\ Aiq % be OPEd%LE s '_ APPLICA8ILITY r x; ;,;.n,OTS, et during movement of irradiated fue (continued) P2. psemblies. Noving irr g j y g:pj :adiated -- .. .... s
- = nfuel assemblies nni.:; _i.n3Dthe ty a! n eccwr ACTIONS The ACTIONS are modified by three Notes. The first Note ;
allows 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, a,t the controls of t h . In this way the penetration can be Y rapidly isolated when the need for secondary containment}A isolation is indicated. 6( The second Note provides clarification that for the purpose of this LCD separate Condition entry is allowed for each Hj$EZ[ penetration flow pathpr - O1 i The third Note ensures appropriate remedial actions are taken, if necessary, if the a d system (s) are rendered inoperable by an inoperable S g< , A.1 and A.2 ,j /$ [jq.2 In the event that O \ 7 ~ paths with one SCI flow path (s) must are one 'or more penetration flow noperable, the affected penetration isolated. h t Y 'N includetheuseofatleastonelThemethodofisolationm isolation barrier that y cannot be adversely affected a single active failure k N Isolation barriers that _ Jhis criteria are a closed and ; de-activated automatic" liclosed manual ;;.h., ::t a gc#a,'/" l blind flange. For penetrations isolated in accordance i Required Actfon A.1, the d W used to isolate the ' l penetration should be the closest available eatme/to-~ gN'M i secondary containment. l This Required Action must be i completed within the 8 hour Completion Time. The specified . time period is reasonable considering the time required to < l isolate the penetrati nd the low probability of a 08A, h which requires the SC short time. to close, occurring during this < J j For affected penetrations that have been isolated in k accordance with Required Action A.1, the affected ' (continued) BWR/6 STS B 3.6-97 Rev. O, 09/28/92 (A M b /y a? h 83 42 BASES SURVEILLANCE SR 3.6.4.2.1 continued) REQUIREMENTS 5C/Ge <& Since these are readily accessible to personnel during nomal'udi operation and verification of their position is relatively easy, the 31 da requency was chosen to provide added assurance that the [vdy re in the ' , correct position,s (scio; , Two Notes have[b'8 ed to this SR. The first Note g s applies to @2Pand blind flanges located in high Qn. cam ' radiation areas and allows them to be verified by use of administrativeAontrRDC Allowing verification by ad_ministrative @ s considered acceptable, since # access to these' areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Theref re, the ' 7 M' probability of misalignment of these tr nce they have h */d ~ been verified to be in the proper posi ton (is low. - - - - - ~ _ . . - - , o A second Note has been included to clarif SC J that h are open under administrative controls _ are not required to , are open. meet the SR during the time the[coCJ 3.6.4.2.2 SR [()a Verifying the isolation time of each power operated and each ' automatic SCIV'is within limits is required to demonstrate y OPERABILITY. The isolation time test ensures that the OTii4:- will isolate in a time period less than or eaual to that t 4 0*# assumed in the safety analyses. The dTM eiaa + " " 491 --- Frequency of this SR C Dr ouoroao - -i tti tr e ; >iaer. waz # 4;a .w , . .~ -..- + 92 days] . ~ ~ / SR 3.6.4.2.3 D Verifying that each automatic SCI, closesona.fsecondary containment}?-isolationsignalisrequiredtopreventleakageg of radioactive following a DBAmaterial or,otherfromacci(dents.secondarycontainment% This SR ensures that each automatic SCD(will actuate to the isolation position [ @ on a (secondary containment?_. isolation signal. The LOGIC SYSTE'M FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this SR to provide complete testing of the safety function. The ([. f1BT-monthFrequencyisbasedontheneedtoperformthis Surveillance under the conditions that apply during a plant (continued) BWR/6 STS B 3.6-100 Rev. O, 09/28/92 b SY ^ (A AA 90-/YKl (3 3, 0, 'l 3 INSERT B104A A.1 and A.2 With one SGT subsystem inoperable, action must be taken i .._..;;;;; ele to verify that the OPERABLE SGT subsystem is not / operating in the purge flowpath. l ~ . ~ s ,_,,,c , ,e -. lft' w f conbl w h i j e 1 INSERT RIVER BEND B 3.6-104 10/1/93 ) {LAR 93-!YKl) SGT System B 3.6.4.3 BASES '. SURVEILLANCE SR 3.6.4.3.3 (continued) REQUIREMENTS : 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 Survelliance can be performed with the reactor at power, operating experience has shown these components usually pass the Surveillance when performed at the i f{18Thonth Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. 3.G,y 3 t /O % ,,,,, i SR 3.6._4.3.4 co, /*y This SR requires verification that the SGT filter @ bypass damper can be opened.and the fan started. This , ensures that the ventilation mode of SGT System operation is y available. While this Surveillance can be performed with - 00,1 ' the reactor at power, operating experience has shown these components usually pass the Surveillance when performed at thq,.418)*'uonth Frequency, which is based on the refueling cycle. Therefore, the drequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. 10 CFR 50, Appendix A, GDC 41. h 2. ,Sectiond.2.3}I . I t h 3. l., , Section J15.6.5 7 ' 68 4. Regulatory Guide 1.52, Rev y 2}( h l l l l t I i BWR/6 STS B 3.6-107 Rev. O, 09/28/92 l l @ 03-ivPI) Annulus Mixing System B 3.6.4.4 BASES (continued) LC0 Following a DBA a minimum of one shield building annulus mixino subsystem is reauired to1Nintr^ "'e aanu'es et 34 > MA ,6~:::= ::rc=r .2 2 -~-- tc th: :nyira:=nt a:. & 4 (p - adequately mix gaseous releases for processing by the Standby Gas Treatment System. Meeting the LC0 requirements for two operable subsystems ensures o>eration of at least one shield building annulus mixing su) system in the event of a single active failure. 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 6.l With one shield building annulus mixing subsystem inoperable, the ino)erable subsystem must be restored to OPERABLE status wit 11n 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 adeouately performed. The 7 day Com)letion Time is based on consideration of such factors as t1e availability of the OPERABLE redundant shield building annulus mixing subsystem and the low probability of a CBA occurring during this period. (continued) BWR/6 SUPPLEMENTAL B 3.6-107 ii 10/1/9'3 /l X 9,'/vR/) Fuel Building Ventilation System-Fuel Handling B 3.6.4.7 BASES BACKGROUND gheFuelBuildingVentilationSystemautomaticallystarts and operates in response to actuation signals indicative of (cortinued) conditions or-an accident that could require operation of __ the system j oll ng 1 t1ation. ot.n enclosyre buildi, fig l 08/ fecircul lon f 5 and th charc 1 filter t in fans ' - start.- Fuel B iding ntilatio System f1 are / 7, o,'/,7 pontr led by dulat' g inlet anes coal fil er trai exhaust ans and t instal ed on th positio vo'gme / g /,, W f ch trol da rs in alled 1 ranch duc to ind' idual / regions the s ondary c tainment. f APPLICABLE The design basis for the Fuel Building Ventilation System is SAFETY ANALYSES to mitigate the consequences of a fuel handling accident (Ref. 3). For all events analyzed, the' Fuel Building Ventilation System is shown to reduce, via filtration and adsorption, the radioactive material released to the environment. Since the system is assumed to filter all releases, with the analysis not accounting for any delay in system startup, at least one subsystem must be in operation while handling irradiated fuel. The Fuel Building Ventilation System satisfies Criterion 3 of the NRC Policy Statement. LC0 Following a FHA, a minimum of one Fuel Building Ventilation 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 o xrable subsystems ensures operation of at least one Fuel Building Ventilation subsystem in the event of a single active failure,. Requiring one subsystem to be in operation ensures no releases occur that are not filtered and adsorbed. (continued) BWR/6 SUPPLEMENTAL B 3.6-107 xiv 10/1/93 EM es-N#/) Drywell B 3.6.5.1 8 3.6- CONTAlm4ENT SYSTEMS 8 3.6.5.1 Drywell BASES BACKGROUND l The drywell houses the reactor pressure vessel (RPV), the reactor coolant recirculating loops, and branch connections Or oyDb %g of the Reactor Coolant System (RCS), which hve isolation valves at the primary containment boundary. The function of e the drywell is to maintain a pressure boundar dtfIESS[on Pook. steam from a loss of coolant accident (LOCA) to they that channel ----- ----tsuppress.jortpool,_ whg_te it h condensed __AirJorced from i [W, ~~_ the drywell is released into this primary containmentk The pressure suppression capability $ assures that peak LOCA S g re5Sion & cokiwithin design limits.;- Temptfathe anTpfessure in Ene seinary conta J The drywell also protects accessible areas of the containment from radiation originating in the reactor core and RCS. To ensure the drywell pressure suppression capability, tiie drywell bypass leakage must be minimized to prevent overpressurization of the primary containment during the drywell pressurization phase of a LOCA. This re periodic testing of the drywell bypass leakage, quires confirmation that the dry < ell air lock in leak. tight, OP.ER_A8ILITY. of the E7 ~I drywell isolation valves g W s f \ @li vacuuurTettif vapyss an: closed. Hri:~o\irmation that the ~ The isolation devices for the drywell penetrations are a part of the drywell barrier. To maintain this barrier: l
- 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 OPERABLE automatic DIV, or
- 2. closed b manual valve bl flange 9 or N deractivatedautomatic[valv secured'in ;
position [exceptasprovidedi LE0 1 A 5.3 3.(,.r,/ "Drywell Isolation Valves (DIV ; ae '4 -- _c 3 ; -) ~
- c. A// by-e// eQ,re d As 4 es an e/ mea ';
~ A ~/ca , >
(c.ontinued) i BWR/6 STS B 3.6-108 Rev. O, 09/28/92 1
~ _ _- _ _ _ -__ _ __ . - _ . _ --
l (J.Ad <?:-WPi) l Drywell !
8 3.6.5.1 i BASES l j
8ACKGR002 c. Th Drywell Vacuum R ief Sys is OP LE exce as p vided i LC0 3.6 .6, 'D 11 Vacu Relief (continued) Uystes.' (
This Specification is intended to ensure that the perfomance of the drywell in the event of a DBA meets the 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 is condensed. Maintaining the pressure suppression capability assures that safety analyses remain valid and that the peak LOCA temperature and pressure in the primary containment 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 drywe ] is 0PERA8LE if#t e _
. . - bypass leakage is withinl im 7, . ; a ne & to 4 v 24 !< i., J /
- r. yp n o a.. :,, :t ,
^ ~ - - -
BWR[6STS 8 3.6-109 Rev. O, 09/28/92 1
n
[M ::-i/It)l orywel B 3.6.5.1 BASES. ':
8.d- (continued)
ACTIONS commensurate with the importance of maintaining the drywell 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 ,
drywell is inoperable is minimal. A'so, the Completion Time-is the same as that applied to inoperability of the primary containment in LCO 3.6.1.1, " Primary Containmen ."
93 '
B.1 and B.2
(- Ope <4} -
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 h JAMD f3 //oA 7'
challenging plant systems.
4.u.9 w
'b SURVEILLANCE SR 3.6.5.1 [/e s, _ff ,,,,,-2,, 9,, /
~' '
REQUIREMENTS i -
',' /c"n c!'/ ~$7/8 The analyses in Reference,2'are based on a maximum drywell -
5 bypass leakage. This Surveillance ensures that the actual !
/,,ea44 / 4 de Gfdrywell bypass leaka f /,rdg,pr ,,4,rg,ed,,-Qdesign cr value of.fl.0he t
2 is CC = Mhe acceptable A/6 ass ,
u rufs,rif A The leakage test is performed every [18] raonths, consistent
/e,y 2
//,egfyy, en ,,;.c with the difficulty of performing the test, risk of high s/, ,7,,//c;p,,,
g ,/, radiation exposure, and the remote possibility that a
.6, f component failure that is not identified by some other
}f'/~'g h/ ""/ drywell
' or primary containment SR might occur. Operating
<-a- dv a/ #erO "J"Jexperience has shown that these components usually pass the
! fin: hekeen <efv,r'[ Surveillance when performed at the ,{18]e month Frequency.
/ </ry-f// /e,v,ge rate Therefore, the Frequency was concluded to be acceptable from h
ferf dr a reliability standpoint, c r&r, u< is ><c base, g[oe ,,
5: .A+ *Ic SR 3.6.5.1 despf //ri-A/M. de c - f ,, b ,d -t y ra /gre The exposed accessible drywell interior and exterior c,* -L vem seurs,co f
surfaces are inspected to ensure there are no apparent les, m := ,re 1:n 1<f ,
Ae assurp/ishs o-( yle z, telycina/ysis. (continued)
' ~~ -
.-)
BWR/6 STS B 3.6-110 Rev. O, 09/28/92 i
i w .en-.
[AR 93'lY V /D Drywell B 3.6.5.1 I
8ASES
.I i p4 P1) ~
(. a9 SURVEILLANCE SR 3.6.5.1.6(Yontinued)
REQUIREMENTS d !
I physical defects that would prevent the drywell from f [,, c, ' performing its intended function.
\fwJefhgufone
' ,eEedd;drywell structural integrity is maintained.
This SR ensures that
#7 7' 4 The D Frequency was chosen so that the interior and exterior ceda '""/^"#7 ' p surfaces of the drywelT ea'n ce inspectedhary oth ed /ff"'"
/O crR 50 r_.__,,,,,
6 . ,___. Due to the passive nature of the drywell
,,, f ; y g/ [.2), Bl structure,theXc'a m Frequency is sufficient to identify component'degrada y affect drywell y structural integrity.
,jf,g])
REFERENCES h 1. NR, Chapter)6Nnd Chapter)15[ h F
BWR/6 STS B 3.6-111 Rev. O, 09/28/92
g)g cp:3wij)
Drywell Air Lock B 3.6.5.2 RASES ACTIONS D.1 and D.2 (continued) conditions from full power conditions in an orderly manner and without challenging plant systems.
g 4.0f This SR requires a test be h of the drywell air lock doorsAt e- ~" :onned to: verify seal,, leakage
- .11 ::-- . A seal leakage rate lieit of 8 6 scfh has been established to ensure the integrity of the seals. The Surveillance is 4 [only mquired to be perforiiiWonce after each closing. The
'72 4"" uency of 72 hours is based on operating exper'ience. n r is e ns erefedequay in vips of theAher i tea gas av 1 le to fiant 4perations personriel th C
* 'I the sesl '
.1 SR 3.6.5.2.2 Every 7 days the 11 air lock seal air flask pressure is '
ODL verified to be a psig to ensure that the seal system remains viable.
I must be checked because it could bleed down during or following access through the air lock, which occurs regularly. The 7 day Frequency has been shown to be acceptable, based on operating experience, and is considered adequate in view of the other indications to the plant operations personnel that the seal air flask pressure is low.
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 I pressure, closure of either door will support drywell
~
, OPERASILITY. Thus, the door interlock feature supports
/ and '
drywell opfRASILITY while the air lock is being used for on I transit in and out of the drywell.
w 9 f Wedw.e, I- Periodic test f this interlock demonstrates that the interlock ;
ed andnot thatinadvertently simultaneous inner and f.~, ^gMM t , =
will outerfunction door openingas wi desi@'I .
occur Due to the (continued) {
i BWR/6 STS B 3.6-117 Rev. O, 09/28/92 l
I
[], AR '&/97))
ofvs B 3.6.5.3 BASES
"~ "
continued) N3 '
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 DIVs are not required to be OPERA 8LE in MODES 4 and 5.
ACTIONS The ACTIONS are modified b allows penetration flow pat to beNotes. The first Note unisolated
,0 CAPI bMeh Puy A Pe4<ab intermittentiy unoer administrative controls. These flow res, 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 provfdes clarification that for the purpose of this LC0 separate Condition entry is penetration flow path.g g g g_ allowed for each The third Note requires the OPERA 8!LITY of affected systems to be evaluated when a DIV is inoperable. This ensures appropriate remedial actions are taken if necessary if the affected system (s) are rendered inopera,ble by an inoperable DIV. -
ruscar 1.g A.1 an%Ind A.2 6f 7'
17 With one or more penetration flow paths with one DIV inoperable, the affected penetration flow path must be t isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely iMc- affected by a single active failure. Isolation barriers
. that_ meet th riterion are a closed and de-activated automatic ,
closed manual valve, a blind flange, and a check valve with flow through the valve secured. In this hKCondition, the remaining OPERABLE DIY is adequate to perform T6e isolation function. However, the overall reliability is '
(continued)
BWR/6 STS B 3.6-122 Rev. O, 09/28/92
)
i
(),AR 9"- H//) '
OfVs I B 3.6.5.3 1 8ASES i SURVEILLANCE SR 3.6.5.3.1 (continued) --'
i REQUIREMENTS V Close-l isolation valves 4534 are not qualified to @ under accident conditions. This SR is designed to ensure that a l
gross breach of drywell is not caused by an inadvertent or l
spuriousofdrywell these; purg]e Detailed analysis inch isolation drywell purgevalvevalvesopening. failed to i conclusively demonstrate their ability to close during a LOCA in time to support drywell 0PERASILITY. Therefore, k these valves are required to be in sealed closed position 7:
during MODES 1, 2 and 3.
Thes:M ] inch drywell purge '
l valves that are sealed closed must have motive power to the b d valve operator removed. This can be accomplished by
/
de-energizing the source of electric power or removing the ;
air supply to the valve operator. In this application, the '
ters 'Frequency The sealed" hasisnoa 1ennnotation H ult :f theof"*"leakane r:: ' within limits.
eth: Of S::MO
,,, ,:-Ire": M4 GM. " re'et;d te purge valve use during unit operations. j j ,,,,
~
3.6.5.3.2 f_ PrimaCenkinW SR ~
y This SR ensures that the G4] "- ^ brywell isolation valves are closed as requir f allowable reasoyg : ed :a or, ifpad:d open, tg open forf an y ..,_-y.,. ...... == -- --- --t - 1 v - - - m > == r -
& =de ---d ! ~ t -:::f '. t'. : ; . % - fer . e= :: .e 1..; c -
e :f 1: C_ for limited periods of time. This SR has g;my ,
been modified by a Note indicating the SR is not required to ,_
panwg
- De est when sne' drywel))=. ,- 5:;;?y = e-t:0 valves are
~ open for pressure control ALARA or air ualit ge , mn,q consideretions for personn,el entry, or rveil ancesfthat n g n Mtd i !
require the_._....__.......___.____...t,
.,_____._ m , valve to be open (=rz:J",- ' "= Q C ',.~.:'_, _
- i. ,
bi-t-il. The 31 day Frequency ~ is consistent with the '
/
valve requirements discussed under SR 3.6.5.3.1. ,
or 90<= / ,_
- 7 f SR 3.6.5.3.3
/
sv #, [ L/
This SR requires verification that each drywell isolation manual valve and blind flange that is required to be closed t
during accident conditions is closed. The SR helps to 3, g 5, a ensure that drywell bypass lea)Lage is maintained _to a _
j fp minimum. n~::n tin: rt"-* = G2" ;tri =:t:F :_t.' t Qhe he We \o cah.n oT Wese d evices )
~ 3 (continued)
BWR/6 STS 8 3.6-125 Rev. O, 09/28/92 1
8 ATTACHMENT 2B ITS PSTS COMPARISON DOCUMENT- -
REVISION 1 DISCUSSION OF CHANGES
[MR 93-/UI]
DISCUSSION OF CHANGES TO NUREG-1434 SECTION 3.6.1.3 - PRIMARY CONTAINMENT ISOLATION VALVES PLANT SPECIFIC DIFFERENCE (continued)
P.14 This comment number is not used for this station.
P.15 A surveillance for the annulus bypass leakage has been added consistent with existing requirements.
P.16 Since the drywell purge valves are required to be sealed closed in MODES 1, 2 and 3, no separate restriction is necessary.
P.17 This comment number is not used for this station.
3.(, .I, 3 p @ G) bov g gg p7 P.18 K gurveillanc6 and Bases description 7for the MS-PLCS been added consistent with existing requirements.
P.19 This comment number is not used for this station.
CHANGE / IMPROVEMENT TO NUREG STS C.1 This change corrects an obvious oversight.
C.2 The improved TS separate primary containment (LCO 3. 6.1.1) and PCIVs (LCO 3.6.1.3). The point of separation revolves around Appendix J leakage limits and testing, which are identified as primary containment OPERABILITY. PCIVs form a portion of this boundary (and it is this " boundary that is addressed in LCO 3.6.1.3), however Appendix J leakage limits not met is not a direct impact on PCIV OPERABILITY. Bases discussions of " leak tight" and/or " leakage" related to PCIVs is modified to reflect the more appropriate " boundary" relation.
C.3 This change is proposed to maintain consistency between the Specifications and Bases.
C.4 These changes provide a more detailed and editorially enhanced discussion of the PCIVs design and scope.
C.5 This change provides the appropriate MODE references (as they are also presented in Condition A and SR 3.6.1.3.3.)
C.6 This comment number is not used for this station.
C.7 An exclusion to MODES 1, 2 and 3 is added to Note 4 for clarity. Since LCO 3.6.1.1 is only applicable in MODES 1, 2 and 3, and the overall leakage criteria is not applied outside i of these MODES, the exclusion is appropriate. ,
)
Similarly, other leakage criteria (SR 3.6.1.3.8 and SR 3.6.1.3.9) have this same exclusion added. This is consistent with the same exclusion already provided in SR 3.6.1.3.10. l 1
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ATTACHMENT 1: CTS-PSTS COMPARISON DOCUMENT :
ATTACHMENT 2:ITS-PSTS COMPARISON DOCUMENT U l
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ATTACHMENT 1 ITS - PSTS COMPARISON DOCUMENT REVISION 1 SECTION 3.7 REVISED PAGES 1 A: MARKUP OF CTS 18: DISCUSSION OF CHANGES 1C: NO SIGNIFICANT HAZARDS CONSIDERATIONS
gh aw ATTACHMENT 1 A CTS - PSTS COMPARISON DOCUMENT REVISION 1 MARKUP OF CTS i
L
S M 9 & /YRf}
PLAWT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any vent'14 tion zone consuni-cating with the subsystes by:
l, McVED h 1. Verifying that the subsystes satisfies the in-place penetration T5 C .13J'r and bypass leakage testing acceptance criterion of less than
< O.055 and uses the test precedure guidance in Regulatory Posi-tiens C.5.a. C.5.c and C.S.d of Regulatory Guide 1.52, Revision
- A3 2. March 1978, and the system flew rate is 4000 efe 105.
l i
- 2. Verifying within 31 days after removal that a laboratory analysis l
- of a representative carben sample obtained in accordance with !
Regulatory position C.G.h of Regulatery Guide 1.52, Revision 2, i March 1978, meets the laboratory testing criteria of Regulatory !
Position C.6.a of Regulatory Guide 1.52. Revision 2. March 1978, i for a methyl iodide penetration of less.than 0.1755; and
- 3. Verifying a subsystem flew rate of 4000 cfe 2105 during sub-system operation when tested in accordance with ANSI N510-1980. ,
After every 720 hours of charcoal adsetter operation by verifying d.
within 31 days after removal that a laboratory analysis of a repre-sentative carhen sample obtained in accordance with Regulatory Post-ten C.S.h of Regulatory Guide 1.52, Sovisten 2. March 1978, meets the laboratory testing criteria of Segul Position C.E.a of Regulatory Guide 1.52. Esvisten 2 , for a methyl fodice I penetration of less than 0.1755. . March At least once per 18 months by:
(e. 1. Verifying that the pressure drop acrest the combined HEPA filters e g and charcoal adsarber banks is less than 8 inches water gauge j g,7 ( while operating the subsystem at a flev rate of.4000 cfm a 105._ /
s N 2. (Verifying that en each of the below emergency sede actuation l (test signals.jtne sumeystem ausensucany saidtenes se sne w '
, se n31 . emergency mese of eseration, Jtne iseianten valves close withD
'/j 't. , escenessans the centrol rees is.es' ntained at a positive
( pressure of > 1/8 inch water gauge relative to the outside j % 2 7'J 4 etasephere dring subsystem operation at a flev rate less than er aquel to 4,000 cfe:
7f L# _ **e) LOCA and h
M*D io b) Local air intake radiation monitor - Nish. .
Verifying that the heaters dissipate 23 e 2.3 lar when tested in
, J 3. accordance with ANSI N510-1900, at the design supply voltage.
1'S 5713; L
- Ine specirted la month interval during the first operating cycle may be extended to coincide with completion of the first refueling outage, sched
( to begin 9 15-87. _1 RECEIVED
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RIVERBOS-UNIT 1 3/4 7-6 Amendment No.1 AU.G 191987 sac u
(y t go - nh ATTACHMENT 1B CTS - PSTS COMPARISON DOCUMENT REVISION 1 DISCUSSION OF CHANGES l
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DISCUSSION OF CHANGES CTS: 3.7.2 - MAIN CONTROL ROOM AIR CONDITIONING SYSTEM TECHNICAL CHANGE - LESS RESTRICTIVE I (continued)
" Specific" L.1 The Applicability of this specification is revised to exclude MODE 4 & 5 if no activities are being conducted which may lead to a need for control room isolation. The probability and consequences of a design basis accident are significantly reduced due to pressure and temperature limitations in these MODES. However, some activities increase the probability of some accidents and these are retained in the applicability.
L.2 The requirement to perform the syscem testing on a staggered test basis is deleted. The current requirement delineates unnecessary detail for scheduling of the testing. Since the f requency was not affected, i.e., both current and proposed require monthly testing for each subsystem, and scheduling is not a safety concern as long as both subsystems are not tested simultaneously, this requirement can be deleted with no impact on safety.
L.3 This surveillance requirement is being deleted. In its place an entire new LCO on control room heating, ventilating, and air conditioning (HVAC) is being added in this proposal. Although this 12 hour surveillance requirement is being removed, the new spec 4fication better demonstrates the adequacy of the control room HVAC system to remove the necessary heat loads.
L.4 The phrase " actual or simulated" in reference to the automatic initiation signal, has been added to the surveillance requirement for verifying that each subsystem actuates on an automatic initiation signal. This allows satisfactory automatic system initiations for other than surveillance purposes to be used to fulfill the surveillance requirements.
OPERABILITY is adequately demonstrated in either case since the system can not discriminate between " actual" or " simulated."
L.5 The Frequency for demonstrating that the Control Room G'7 Ventilation System is capable of pressurizing the control room gll has been changed from each train every 18 months to one train
, each 18 months. Since this test is intended to demonstrate the leak tightness of the control room boundary, this test need only be performed on one train each 18 months. The TS will still require verification that each control room ventilation train is capable of automatic initiation every 18 months.
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ATTACHMENT 1C CTS - PSTS COMPARISON DOCUMENT REVISION 1 NO SIGNIFICANT HAZARDS CONSIDERATIONS
.-4
-, , . - . . _ . . ~ - - -- .. -.
L y .k.5& fro #0 /. ne e miN fjAl93-/VR}} ,
-fro -r org i ) u ,d*(. e /M 9241 4
NO SIGNIFICANT HAZARDS CONSIDERATIONS- ,
CTS: 3.7.2 - CONTROL ROOM EMERGENCY FILTRATION SYSTEM i i
"L4" CHANGE Entergy Operations Inc., has evaluated this proposed Technical Specification change and has determined that it involves no l significant hazards consideration. This determination has been i performed in accordance with the criteria set forth in 10 CFR 50.92. !
The following evaluation is provided for the three categories of the !
significant hazards consideration standards:
}
- 1. Does the change involve- a significant increase in the ;
probability or consequences of .an accident previously i evaluated? [
L The phrase " actual or simulated" in reference to the automatic initiation signal, has been added to the system functional test l 4 surveillance test description. This does not impose a requirement to create an " actual" signal, nor does it eliminate :
any restriction on producing an " actual" signal. While creating an " actual" signal could increase the probability of an event, existing procedures and 10 CFR 50.59 control of j revisions to them, dictate the acceptability of generating this signal. The proposed change doer not affect-the procedures governing plant operations and the acceptability of creating !
these signals; it simply would allow such a signal to be [
utilized in evaluating the acceptance criteria for the system i functional test requirements. Therefore, the change does not ;
involve a significant increase in the probability of an ;
accident previously evaluated. j 1 Since the function of the system functional test remains unaffected the change does not involve a significant increase j in the consequences of an accident previously evaluated. ;
- 2. Does the change create the possibility of a new or different l kind of accident from any accident previously evaluated?
- The possibility of a new or dif ferent kind of accident from any accident previously evaluated is not created because the proposed change introduces no new mode of plant operation and ,
it does not involve physical modification to the plant. l
- 3. Does this change involve a significant reduction in a margin of [
safety? >
f Use of an actual signal instead of the' existing requirement i which limits use to a simulated signal, will not affect the i performance of the surveillarme test. OPERABILITY is y adequately demonstrated in eith r.'ase since the system itself !
can not discriminate betweer " actual" or " simulated." -
Therefore, the change does not involve a significant reduction !'
in a margin of safety.
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NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3. 7.2 - CONTROL ROOM EMERGENCY FILTRATION SYSTEM Jgl
(
"L5" CHANGE s'
Entergy Operations Inc,. has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92.
The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
This change will decrease the Frequency of demonstrating that each Control Room Ventilation subsystem is capable of pressurizing the control room within the required time limit.
control room pressurization time is not assumed to initiate an accident sequences. Therefore, this change does not increase the probability of a previously analyzed accident. Because this test is really a test of the leakage integrity of the
- ontrol room boundary, it does not need to be performed on each train each refueling outage. This LCO will still require that the automatic initiation capability of each train of control room ventilation be demonstrated each 18 months. Therefore, this change does not significantly increase the consequences of a previously analyzed accident.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change introduces no new mode of plant operation and it does not involve physical modification to the plant.
Therefore it does not create the possibility of a new or i different kind of accident from any accident previously l evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
This change does not involve a significant reduction in a ;
margin of safety since this test is really a test of the :
1 leakage integrity of the control room boundary. Demonstration of this leakage integrity can be accomplished utilizing only one train of control room ventilation each refueling outage.
This LCO will still require that the automatic initiation capability of each train of control room ventilation be demonstrated each 18 months.
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G M go WiD ATTACHMENT 2 ITS - PSTS COMPARISON DOCUMENT
. REVISION 1 SECTION 3.7 REVISED PAGES 2A: MARKUP OF ITS 2B: DISCUSSION OF CHANGES l
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ATTACHMENT 2A !
ITS - PSTS l
COMPARISON DOCUMENT l REVISION 1 MARKUP OF ITS .
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' 3. 7.1 l SURVEILUUICE REQUIREMENTS (continued)
SURVEILUWICE FREQUENCY
\
f SR 3.7. . s Doerate each Mooling tower fa or 31 days /
/, -
=/ [15] minutes. '
l (c% SR 3.7.; .
WE----/-----J----4--
\'_~J 4 Iso tion o flow tg individual e ' n ts q do s not ndertre- , u = rji ra e.
27 !
}.. ------------y--- -- ---
--)
\ gi Verify each3(SSW} subsystem manuai, power 31 days 7
s operated, stid automatic valve in the flow 6 path servicing safety related systems or Qf'"" components, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3. 7.1, Verify eachj$$ ubsystem actuates on an nths actual er simulated initiation signal. J18 {
BWR/6 STS 3.7-4 Rev. 0, 09/28/92
(AR92tv21) h J CRFA yst g.h'D ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME h E. Two,ECRFA%ubsystems ino'perable during f--3.0 LC
--- ---N E--- ----- -
i not plic le
_- #27 37 movenent of irradiated - --- --- ---- ---- --- -
fuel assemblies in the h primary or secondary
) containment}f"during E.1 Suspend movement of irradiated fuel Isunediately CORE ALTERATIONS, or assemblies in the during OPORVs. Jrimary and secondary containment .
A.N,Q E.2 Suspend CORE Insnediately ALTERATIONS.
bhQ E.3 Initiate action to Issuediately suspend OPDRVs.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. Operat,e eaciv{CRFA bsystem for a 10 31 days continuous hours with the heater S 9 (fer ry-tn. -. u.vn .....-..)]perati(.
h __15 ri-"tes] M
< b.
SR 3.?'. 2 Perfor1srequiredjCRFA lter testing in accordance with theM Ventilation Filter Testing Program (VFTP)F withthjeInaccordan VFT (continued)
P BWR/6 STS 3.7-9 Rev. O, 09/28/92
Q 93-/YeO h ) Control Room AC h ste 3..Ao 3.7 PLANT SYSTEMS 3.7. Control Room Air Conditioning (AC) stem 3 -
LCO 3.7/3 Two) control room ACkubsystems shall be OPERABLE.
h APPLICABILITY: MODES 1, 2, and 3, Durin movement of irradiated fuel assemblies in the O81 rimary or secondary containmentW Du ng CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PORVs).
ACTIONS -
CONDITION REQUIRED ACTION l COMPLETION TIME A. One,[ control room AC[ A.I Restoredcontrolroom 30 days Obl subsystem inoperable. ACPsubsystemto OPERA 8LE status.
Required Action and Be in MODE 3. 12 hours Associated Completi l i
Time of Condition NE !
not met in MODE 1 ,
or 3. J' Be in MODE 4. 36 hours r
__ _n (continued) 4o B. *1No ' codrof Icom OI VerMy con frof rco" v Oncc j3ef -
Ac sunsvMs a"# *~p'a 4" if Inua in opora b/c, d to y *f, 76 I AND h v 13 . 2 Wedne one a j
'3.1 s
control room AC 7 #9' '
)W .sua sys & -fo I A4 OFEi464E s h -lw . '
\
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BWR/6 STS 3.7-11 Rev. O, 09/28/92
-e
(} M 9:-N MI) h J Control Roco AC Nys 3.
ACTIONS (continued)
CONDITION REQUIRED AC710N CONPLETION TINE Required Action and ------------NOTE-------------
associated Completion LCO 3.0.3 is not applicable.
Ti m of Condition A -----------------------------
not met during D movement of irradiated ( Place OPERA 8LE Innediately fuel assemblies in the jcontrolroomACf g
/ containment}7duringprimaryoperation. or secondary 6l subsystem in CORE ALTERATIONS, or during OPORVs. QB
.1 Suspend movement of Immediately irradiated fuel assemblies in the Jprimary and secondary containment [
- _M
.2 Suspend CORE Immediately ALTERATIONS.
ale !
.3 Initiate action to Ismediately suspend OPDRVs.
T {co rol ACF .1 E er LCO 3. 3. Isumedia ly OOI\ in b,sys in rable 1, 2 or'3.
6,
/ ,
(continued) 37
~ 34 BWR/6 STS 3.7-12 Rev. O, 09/28/92 l
[5M ?.1/Yfl)
MControl Room AC t ACTIONS (continued)
CONDITION REQUIRED ACTION CONPLETION TIME h [E te i during movement of n ab
/ C--
I LC 3.0 i
TE- -- --- --
not app icab e.
irradiated fuel assemblies in the E.1 Suspend movement of Immediately Mprimary or secondary irradiated fuel containmentWduring assemblies in the CORE ALTERATIONS, or Mprimary and during OPDRVs. secondary
% , containment [
6.4 fby aa:EAk an0 g t
i4 au: ,,,4Q Coy /s k- -
f),,, e e l Crf./ f,; E.2 Suspend CORE Immediately ALTERATIONS.
d nf mi .
g E.3 Initiate action to Ismediately suspend OPDRVs.
i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
- l SR 3. . . Verify eachNcontrol onACfs'ubsystemhas $18[ months ,
the capability to remove the assumed heat l load.
BWR/6 STS 3.7-13 Rev. O, 09/28/92
(LM 9:4 81) g Main Condenser Offga 3.71 4 l
i SURVEILLANCE REQUIREMENTS j
SURVEILLANCE FREQUENCY h SR 3.,.
4
-------------------NOTE--------------------
Not required to be performed until 31 days )
after any main steam line not isolated
- . jW
'3 A $ --SSS S---------------------
erify the gross amma activity rate of the 31 days CW y p noble gases is s decay of 30 minutes >mct/secondjafter refoi .
' - - y Once within 4 hours after a
; SR s.7. v. / t 50% increase j s in the nominal
'~~ g steady state fission gas
- I C l , ,, releasee after factoring out increases due to changes in THERMAL POWER level BWR/6 STS 3.7-15 Rev. O, 09/28/92
&K93-lynt)
[h M s3W b stem 3.7.1 and BASES 00 APPLICABLE SAFETY ANALYSES is the failure of one of the two stan y DGs, which would in (continued) turn affect one Q Q system TheM5WP1' low assumed in gpe) per pump to the g the analyses .Tablei6.2is 'b,Ref. 7) Reference 2atdiscusses ystes perfomance during these conditions, exchan er
$5 7 g g
The95SW}Psystem,togetheruiththeTtAtS[sattsfy Criterion 3 of the NRC Posf:y Statement.
LC0 The OPERASILITY of subs stem A (Division 1) and subsystes B (Division 2 }7ystem is required to ensure the e}}