ML20058M021
| ML20058M021 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 12/20/1993 |
| From: | CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | |
| References | |
| NUDOCS 9312200116 | |
| Download: ML20058M021 (424) | |
Text
i PERRY -
UNIT i 7
i CHAPTER 1 :
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TTACHMENT 1 ,
l CTS - PSTS :
r COMPARISON DOCUMENT '
1A: MARKUP OF CTS 18: DISCUSSION OF CHANGES ,
1C: NO SIGNIFICANT HAZARDS .
CONSIDERATIONS ,
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ATTACHMENT 1 A t t
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CTS - PSTS i t
COMPARISON DOCUMENT i:
6 MARKUP OF CTS r
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IDEFINITIONS i 5 $' -
<[~N$T h T N ing te m are-defined : that unWem 'nterpr-station of-these ten-Watiens- ey be achiausd. The defined terms appear in capitalized type and r applicable throughout these Technical Specifications ACTIONg k -,- - , a- - gr- -
Ad 4. b ,. e
{3 ACTIO@ hall be that part of a Specification which, rescribes r+
meawres renuired ' under designated conditions - - '
wd6 s pd.rJ C.
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. "/Er'It Mf"9 EX?O';' JR b c AF' ^@~ ~Q OM 1.2 TM "!ERAO: PLANAR EJCME soli Ls erri LLl; L ., d '? JPN 1 height and q ual~i.v dic~ a m.. ef-the-expc are of oild hc fusi . uJa -iu ths O specified.. bundle >Lthe specified heightNivi&J- L., the ~i.h ,T f =1.-vJ. o
.: f u d t M1a. .. .., - ~~ v-~~~ -
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, AVERAGE PLANAR LINEAR HEAT GENERATION -RATE .
[ML.uGGU' a l O The AVERAC4-PtAMAFLINEAR--iiEAT GENERAT-IGH " ATE -(APLHGR)Yh
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applicable to a specific planar height and is equal to the sum of the L-fMEAR A LHGR d iiEAT GENERATION RATES for all the fuel rods in the specified bundle at the ~
specified height divided by the number of fuel rods in the fuel bundle dN g39 ;
l CHANNEL CALIBRATION h h -
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g of the c annel A CHANNEL output such thatCALIBRATION it responds; theshall necessary be the rangg'an adjustment,oas necess!
known values of the parameter wWtr the channel monitors. /The CHANNEL QA5 ;+ad alarm qW CkA a IBRATION shall trip functions, and encompass shall include the theCRANNELentire channe44ncluding FUNCTIONAL TE57.' thensor:
The
' CHANNEL CALIBRATION may be performed by n
{ or total channel steps ack that the entir series an of sequential, alibrated. verlapping - ,
S c, m% J CHANNEL CHECK ggp g i g a A CHANNEL CHEfLshall be_the qualitative assessment $ of channel behavior during operationfby A e,r_vation, This determinati shall include, where pos-x
^ sible, comparisofy.o,f the channel indication a atus other indications %dTor status derived from independ instrument channels measuring the same parameter.
g CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall bep .
r o cL D C-i- % methern w $-he injection. of a simulated ignal into the inAchannel dis cioso m 4
%} ige dio alarm.,an m sentor as practicable 7 to verify OPERABILITgh
' functio r5annel failure trips 4 4
=.m= ue n cr tc m'fy CPEPJ2!LITi .uMir.; m
-m gg=
fm e;=1 = m Ord/cr t-ip f =th l The CHANNEL FUNCTIONAL TESTo"rsed may b'y be anyserTes of seq overlappinggr total channel steps ._A)thattheentirechannelistested PEERY - UNIT 1 1-1
- --. - - -- - -.._,._,--.-.--.._,__-_.-,,-.._..,.-.n _
t INSERT 1A i
Calibration of instrument channels with resistance temperature +
detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. :
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1 INSERT PERRY - UNIT 1 1-1 2njijg3
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DEFINITIONS CORE ALTERATION ~
~M 1.7 CORE ALTERATION shall be the (add +t4%- 1 E 1. , eivdt4Dn~'nB ovement ofroo uel, sources, Sc(TesWients or reactivity controlpwithin7 tis reactor g
~
Mv:e=mentofrcRM vessel)dth
. lEJis, LPRMs, the vassal TIPQ or fientremoved special movable and fuel in noNoma..-Q detectorsnis the vessel considered a CORE ALTERATION. eSuspension of CORE ALTERATIONS shall not risclude
,)
>CORE completion ofGLIM Ge SDMwvem_ent_nLa_Somponent W
h to a safe cen:cr" OPERATING TTREPORT;y @
,(co q mmi
)
- _..a c o cR. . . '_
- t. ag )-al.l .u.1',
1.8 The g.UPERATING LIMIT 5 REPORT,is the Perry Unit 1-specific documentW ' " "
that provides th.g.zere Oper: tin h 'phese cycleMpecific core per:g cycle in accordance with Specification 6.0.
tinglimits limitsfor thebecurrent shall 0.
eperati";
detemined reload for each cycle, reload Plant operation within these
~
ep;r t'ng limits is addresse4 in individual),5pecifications.
< c-3 .h w u. .E . p r., e '>
(CRITTCADOWER 10ffRB . {'7 / g' 2,. *.Ed
- 4 b , Q.9) assembly The W CbTICAL eb is calculated PCW:R Eli ha of =
PlTIG)CP by applicati~on
?that power in the 'A g prww correlation 2rfer:! El etric"&itic:1 o cause some point in the assembly to experience bQiling transition, divided by the actual assembly operating power. -
DOSE EQUIVALENT I-131 (S)
% ~
.4
- d. DOS J VALENT I-131 shall be that concentration of I-131dnicrocuries/per gram mixb). alone would produce the same thyroid dose as the quantity and isotopic
[q re ~of I-131. I-132. I-133, I-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, ' Calculation of Distance Factors for Power and Test Reactor Sites."___
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DRYWELL INTEGRITY '
. _ . . i 1.11 DRYWELL INTEGRITY shall exist when: '
- a. All drywell penetrations required to be closed during accident conditions are either: g
- 1. Capable of being closed by an OPERABLE automatic isolation system, or
- 2. Closed by at least one matual valve, blind flange, or deactivated automatic valve secured in its closed position.
- b. The drywell equipment hatch is closed and sealed.
- c. The drywell head is installed and sealed.
- d. The drywell air lock is in compliance with the requirements of I
Specification 3.6.2.3.
- e. The drywell leakage rates are within the limits of Specification 3.64 PERRY - UNIT 1 1-2 Amendment No. 20,33
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i INSERT 2A I
In addition, control rod movement with other than the normal j control rod drive is not considered a CORE ALTERATION provided j there are no fuel assemblies in the associated core cell. ,
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i PERRY - UNIT 1 1-2 10/1/93
DEFfNTT10NS
[DRYWELLINTEGRITY(continued)
- f. The suppression pool is in compliance with the requirements of S Specification 3.6.3.1. ,
g.
The sealing mechanism associated with each drywell penetration; e.g.,
welds, bellows or 0 rings, is OPERABLE.
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' E-AVERAGE DISINTEGRATION ENERGY I
1.
eat E shall be the average)(wighted in proportion to the conce% ration of b radionuclide in the reactor coolant at the time ampligg)pfthesumof the average beta and gamma energies per disintegratic' in MeV )
witLhalflivesgr::ttr nonf rodine activity in theP nolant. 15 minutes, making up at least of the 9W)ifor total isotop EMERGENCY CORE COOLING SYST JQ (ECCS) RESPONSE TIME r$. b '^
e O'::: O;:
YC0C:000LDO;Y;TOMECCSfRESPONSETIMEshallbethattime ititerval from when the monitored parameter exceeds its ECCS :.;t.;h setpoint at the channel sensor until the ECCS equipment is capable of performing its 4o f safety functio,ngi.[e.
r
, the valves travel to their required p charge pressu'esseach their required values, etc).', Times shall include diesel generatorstartingandsequenceleadingdelayffyhersapplicable. The response tj may be measured by 1 that the entire respons eries of sequent m , overlapping or total steps cy:: ed.
' EN CLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME
'1NThe
~ d L: : CIRCULATION P'."? TRI? SYSTEM RESPONSEJIME_shall be th_guime intervaljto ully open contacts complete of the suppression recirculation of the pump circuit electric breakep fromaregity initia en the
~
movement 1 T~the associare Q
, <urbine stop valvej, end or S
- p. Turbine control valvef. 6 ,
y j 4 m a m _. O The total response time may be measured byany series of sequential, overlapping or
[ steps that the entire response time is seasured, t
PERRY - UNIT 1 1-3 Amendment No. 42
DEFINITIONS FREQUENCY NOTATION k _'___-N---_--
1.17 The FREQUENCY NOTATION specMietMor the._p.erformance of Surveillance _. . . . . sj '
qquirements shall correspond to the intervals deTibetf-4n la
- .__z_ .
' FUEL HANDLING BUILDING INTEGRITY 's .._ - ..-- ..
1.18 FUEL HANDLING N BUILDING (FHB) INTEGRITY shall exist when: x s :
, a. The doors in'ba \ '
closed, except f access to the 620 foot elevation of the FHB n qrmal entry and exit.
@) b. The FHB railroad trackN door closed.
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- c. The fuel haxling area floor hatche 'ere in place.
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- d. .
The 3.7.7.1. FHS ventilation system is in compliance i y pecification !
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\ e. The shield blocks _are_ installed adjacent to the .:hield Buil ,
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}G EOUS ~RADWASTE TREATMENT x.,------.- (0FFGAS) SYSTEM ~,.,
11 h The GASE0US RADQASTE-TREATjiENT (0FFGAS SYSTEM is the system designed an.19d installed to reduce radioactive gaseous. e)ffluents by coolant system offgasses from the main condenser -evacuation system and provid-N
& collectin i
ing for delay or holdup for the purpose of reducing the total n dioactivity '
prior to release to the environment. ,_ ~ ---
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...IDEF!r:E 7 LEAKAGE' ~ '
/~b'# 1" 5bb"de3 L 20 IDENTIFIE0 LEAKAGE shall be:
, o.Id.,J 5,,s LF_rw 4 E y y bom .
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.h Q Leak ge into 5 such as[ pump seaQpr valve packing .
- e hhat or is captured and'E6nducted to a sump or collecting tang i t
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h Ltakage into the drywell atmosphere from sfJrces that are both specifically located and known either not to interfere with the i 4
operation of 1Wieakage detection systems or not to be PRESSURE
\' B0UNDARY LEAKAG "
dhJD EJr." ,J i
OLATIONSYSTEMIE540NSETIME r, u , J.T,idL A 8'(Jd D Q.tLukey(fy))ii<ob"E y gyg Lb yWer ,m 3
, ..J . a .1 J 1 r. Mc% t ', d
' The ISOLATION SYSTEM RESPONSE TIME shal e that tims'intmaTTro when M'**A-the monitored parameter exceeds its isolation rmt%n setpoint at the #
' channel sensor until the isolation valves travel o their required positions.
Times shall include diesel generator starting and eq'uence loading delay where applicable. The response time ma be measure by n series of sequential, overlappin r total steps that th entire response time is
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g( e.* As of .
PERRY - UNIT 1 1-4 Amendment No. 6
DEFINITIONS LIM $TINGCONTROLR0DPATTERN' '
hp 1.22 A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value
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_ LINEAR HEAT GENER'. TION F. ATE (L%d O L 7W- .-
.2 tigAyjEAyESERfJI0f' ",AIEjtHGR)Mhall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.
p . ,..~--. .._... .... ,
i LTQUIDJtADWASTE TREATMENT SYSTEM N g ~~.-...._..... -
.gf 1.24 The LIQUID RADWASTE4REATMENT SYSTEM is any process or control equipm used to reduce the amount or concentration of liquid radioactive materials i prior to their discharge to UNRESTRICTED AREASmIt involves all the installed ~
and available liquid radwaste management system equii>inent,-as, sell as their j
controls, power instrumentation, and services that make the systein7anctional/
LOGIC SYSTEM FUNCTIONAL TEST m g
.25 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of allI logic component '
i.e., all Q of a logic @ circuit, from:s_egar thragh d facluding the actuatestoand ~
f, contacts i
verify OPERABILITY. The LOGIC YS M FUNCTIONAL TEST may be performed by%ny i series of sequential, verlapping_r otal system steps wrA that the entire logic system is tested, re
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(S) 0F'THE PUBLIC ggg( -
i 1.26 MEMBER (S) 0F THE PUBLIC shall include 'all persons who are not occupa-
' tionally associated with the plant. This category does not include employees A :
, of the utility, its contractors, or vendors. Also excluded from this category l l are persons who enter the site to service equipment or to make deliveries.
iThis category does include persons who use portions of the site for ;
(recreational,o:cupational,orotherpurposesnotassociatedwiththeplant.
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MINIMUM CRITICAL POWER RATIO (vWLPb a _-
rh-QThe "IM!"U" ERITICAL TOWR RATIO (MCPRYshall be the smallestkT^, dich.h ')
exists in the corg q .79 g 3 g g
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[0FFSITEDOSElAL20LATIONMANUAL'TODCE)
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1.28 The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and b !
OM. parameters used in the calculation of offsite doses due to radioactive gaseous I
and liquid effluents, in the calculation of gaseous and liquid effluent moni-toring alarm / trip setpoints, and in the conduct of the radiological environmental ;
\monitoringprogram. _ . . :
N __.-
t 600 \19det 58I O PERRY - UNIT 1 1-5 t
e
i INSERT SA for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate !
correlation (s) to cause some point in the assembly to ;
experience boiling transition, divided by the actual assembly operating power.
INSERT SB MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor !
vessel.
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INSERT PERRY - UNIT 1 1-5 10/1/93 2
DEFINITIONS M- !
OPFRABLE - OPERABILITY T x er y
's ft91Asystem, subsystem, , component or device shall be OPERABLE or have OPERABILITY when it is capable of performing i specified function (s) and wh ,
\ all necessary attenda rumentation, controls,41ectrical power, cooling iseal water, lubricatio ther auxiliary equipment that are required for t a system, subsystes, tee 44 caponent or device to perform its unctio ) are also capable of performing their related support function (s).
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PERATIONAL CONDITION - CONDITION , , - -
%C J '
. . - . . . ~
1 h co.30 An OPERATIONAL CONDITION, i.e., CONDITION,~M shall
) be a Qpecified in Table 1.2. - .
fFHYSICS TESTS D '
f&
x I.31 PHYS
/'
TESTS s a TW~those tests performed to measure the fundamental nuclea descr'i 'racteristics of the reactor core and related instrtmentation.and 1) d in Chapter 14 of the fSAR, 2) uthorized under_the provisions of 10 '
tCFR $0.59 or 3) otheRiise approved by the Commission /
PRESSURE BOUNDARY LEAKAGE Lc - - --
y No aNactorfoolantphstesc;," 9eng::"g solable fault inthrough a
- } nt body,;ogr pipeRY LEAKAGE wall or vessel wall. :M11 h 1:W:p M RY CONTAI NENT INTEGRITV 1.33 PRIMARY CONTAINMENT INTEGRITY shall exist wtan
- a. All containment penetrations required to be closed during accident conditions are either: ,
g
) 1. Capable of being closed by an OPERABLE containment automatic
/
- isolation systes, or i
- i. ,
g 2. Closed by at least one manual valve, blind flange, or deacti-vated automatic valve secured in its closed position except for valves tnat clay be opened as pennitted by Specificat on 3.6.4. { i
- b. The containment equipment hatch is closed and sealed.
- c. Each containment air lock is in compliance with the requirements of Specification 3.6.1.3. ;
- d. The containment leakage rates are in compliance with the requirements ,
of Specification 3.6.1.2. !
( e. The suppression pool is in compliance with the requirements of I (x Specification 3.6.3.1.
/ _ -
[3fgR( C\ Q Q - k EO L d 004T Ik /
PERRY TUNIT'~ 1 ~ N '
/1-6 /c.endment No. 44
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INSERT 6A PRESSURE AND The PTLR is the unit specific document TEMPERATURE that provides the reactor vessel pressure LIMITS REPORT and temperature limits, including heatup j (PTLR) and cooldown rates, for the current !
reactor vessel fluence period. These. i pressure and temperature limits shall be ,
determined for each fluence period .in accordance with Specification 5.8.1.7.
Plant operation within these operating >
limits is adttressed in LCO 3.4.11, "RCS Pressure and Temperature (P/T) Limits." -
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I INSERT PERRY - UNIT 1 1-6 10/1/93 ,
DEFINITIONS
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The sealing mechanism associated w'ith each primary containmenDp enetration; e.g., welds, bellows or 0-rings, is OPERABLE. ,-
.__ -.,-~ _ ..
t PROCESS CONTROL PROGRAM (PCP N N~.- _
s 1.34 The PROCESS CONTROL PROGRAM shall contain the current formulas, sampling,'N analyses, tests, and determinations to be made to ensure that the processing h._ .
and packaging of solid radioactive wastes based on demonstrated processing of / '
actual or simulated wet solid wastes will be accomplished in such a way as to f assure compliance with 10 CFR Part 20,10 CFR Part 61,10 CFR Part 71 and ,/
Federal and State regulations, burial ground requirements and other '
, Qequirementsgoverningthedisposaloftheradioactivewaste '-
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[ PURGE - PURGING 'N i
. . . ~ . . - . . . - . . - .
01.35 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or !
kequiredtopurifytheconfineoent.f, w
other operating condition,
~ ~~
~ ~ ~ 'in such a mann RATED THERMAL POWER; (, R t P)
- R t P N..
.3 I RATED Tl'E"Ji"L P0ZO'sha11 be a total reactor core heat transfer rate to the reactor coolant of 3579'eQrriwt-]
3 REACTOR PROTECTION SYSTEM \ RESPONSE TIME
% / TWC~PJ iN gg Lt W6EAETORfROTECTION SY(ThESPONSE TIME shall be time interval from when the monitored parameter exceeds its rip setpoint at the channel sensor 4
until de energization of the scram pilot valve solenoids. The response time may be measured by n series of sequential, overlappi r total steps that the entire respons time _is measured. So w., y Q REPORTABLE EVENf_ ( % ~
OM 1.38 A REPORTABLE EVENT shai1 be a y of those conditions specified
~~ ' ~ ' '
50.73.,r ' ~ ~ ^ '"'
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l R00'DENSRY N f n. . .
y ,1.39 R00 DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted i 3 is equivalent to 100% R00 DENSITY. -
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f ~<: :.= = : . .:' :,.~
( SECONDARY CONTAINMENT INTEGRITY - . . .
1.40 SECONDARY CONTAINMENT INTEGRITY shall exist when: ~.
O e
N' a. All penetrations terminating in the annulus and required to be closed '
during accident conditions are either: -
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PERRY - UNIT 1 1-7 Amendment No.77 i
DEFINITIONS
- _ _ , - t
- 1. Capable of being closed by an OPERABLE containment automati h isolation system, or
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- 2. Closed by at least one manual valve, blind flange, or deactivated:.
automatic valve, as applicable secured in its closed position.
- b. The et,ntahment equipment hatch is closed and sealed and the shield blocks are installed adjacent to the Shield Building.
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- c. The door in each access to the annulus is closed, except for normal entry and exit. !
hH
- d. I The sealing mechanism associated with each Shield Building ,
penetration, e.g., welds, bellows or 0 rings, is OPERABLE. i t i
- e. The pressure within the secondary ccntainment is less than or equal '
{
to the value required by Specification 4.6.6.1.a. , except for normal .
entry and exit to the annulus.
)
- f. The Annulus Exhaust Gas Treatment System is in compliance with the /
sN re /
---..quirements of Specification 3.6.6.2. .
SHUTDOWN MARGIM(SOr4.
som%
QQ >
.41 (ML'TD0"" F/",CI, shall be the amount of reactivity by which the reactor is suberitTc~al or would be subcritical assuming all c=trcl ret Or: fully O4 inserted-except_.for +ha dogle. control-cod 1f-highest reactivity-eth-wMch-.is assumed to-be-ful-ly,rithdrawn-end-the reactor 4e in-the'-shutdown condition; 4 = 68'F anet vennn fear cc 7g ,
ISITE BOUNDARY
g 1.42 The SITE BOUNDARY shall be that line beyond which the land is neither y d, nor leased, nor otherwise controlled. . _by
. _ _the
_ licensee,. _
'[ SOLIDIFICATION ~
- . . . . . _ . . . ~ . . . . . . . ~ . . . . . . . . - - g 1.43 SOLIDIFICATION shall be the conversion of wet wastes into a form that .
Nmeets shipping and burial ground requirements. .- -- --- - -
l SOURCE CHEC I ' - ,
4 0 1.44 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.
STAGGERED TEST BASIS MIWIExt S B 3
< M3 -
45 STAGGEDST BASIS shair consist vi' y
&~
i a. A test schedule for n systems, subsystems, trains or other designated M -
components obtained by dividing the specified test interval into n equal subintervals. _ _ _ .__ _._ ._ _ .~.
PERRY - UNIT 1 1-8
.- , , , - . ~
i INSERT 8A r
- 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 f' control rod of highest reactivity worth, which is assumed to be fully withdrawn.
Witn control rods not capable of being fully inserted, !
the reactivity worth of these control rods must be I accounted for in the determination of SDM. j INSERT 8B .
STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist .;
of the testing of one of the systems, subsystems, channels, or.
other designated components during !
the interval specified by the i Surveillance Frequency, so that all ,
systems, subsystems, channels, or i other designated components are ;
tested during n Surveillance !
Frequency intervals, where n is the i total number of systems, subsystems, j channels, or other designated !
components in the associated !
function. l t
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INSERT PERRY - UNIT 1 1-8 10/1/93
DEFINITIONS 1
~. ~. .- -
M4 b. The testing of one system, subsystem, train or other designated A w component at the beginning of each subinterval. , -- --
THERMAL POWER '
h reactorTHERMAL coolant. POWER shall be the total reactor core heat tranrfer rate to the TURBINE BYPASS SYSTEM. RESPONSE TIME
(%d dipM sh The T }
..cr> !c.gRBINE BYPASS SYSTEM RESPONSE TIME consists of tw cepare;; M c aptimefrominitialmovementofthemainturbine: sto alve o control valve until 80% of turbine bypass capacity is establishe nd b k(, time from initial movement of the main turbine stop valve or cont i valvh. e -
until initial movement of the turbine bypass valve.9 E4therwesponse time may be measured by
-}
series of sequential, overlapping, or total steps we that the entire response time is measured so
..H- Qt re , ( Ifi10ENTIFIED LEAKAGE kgl-*& 43NMENTU!En gaECE-&" M [11 1eakace Q A ( Jr b dl A E ~
'Is7 0Y 5EUTITIED) LEAKAGE)*
I M RICTED A'REA w
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1.49 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY V7 ! access to which is not controlled by the licensee for purposes of protection of JMEMBERSOFTHEPUBLICfromexposuretoradiationandradioactivematerials,or i any area within the SITE BOUNDARY used for residential quarters or for ',-
Qndustrial, commercial, institutional, and/or recreati_onalgrposes. , . _
pVENT
_ILATION EXHAUST TREATMENT SYSTEMS- . . ,_' _- _ . . - - . ~ . ~ . - - - -
~~~.--.-.~-- _
1.50 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and N installed to reduce gaseous radiciodine or radioactive material in particulate
,M form in effluents by passing ventilation or vent exhaust gases through charcoal 1 I adsorbers and/or HEPA filters for the purpose of removing fodines or l l particulates from the gaseous exhaust stream prior to the release to the !
environment (such a system is not considered to have any effect on noble gas i effluents). Engineered Safety F>ature (ESF) atmospheric cleanup systems are not considered to be VENTILA MON EXHAUST TREATMENT SYSTEM components provided '
the-.-ESF system is not utilized to treat normal releases. .,
TE D s~s . .-
1.51 VENTING is the controlled process of discharging air or gas from a ~- ,' M '
Al confinement to maintain temperature, pressure, humidity, concentration or other i
{ operating condition, in such a manner that replacement air or gas is not ,
I provided or required during VENTING. Vent, used in system names, does not ly a VENTING process. -- - -- - ~ ~
PERRY - UNIT 1 1-9 l
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TABLE 1.1 SURVEILLANCE FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. :
D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. i W At least once per 7 days.
H At least once per 31 days.
Q At least once per 92 days.
SA At least once per 184 days. l k A At least once per 366 days. !
j R At least once per 18 months (550 days). l S/U Prior to each reactor startup. !
. P Completed prior to each release N .A. Not applicable.
x -
k- X 1
PERRY - UNIT 1 1-10
TABLE b l.l- ,-l-( ' '3
?49' M9 ,
OPERATIONAL CGNDITidN54/moOES __
j
_QE>Atof (9_s1C yw {FODESWIT)t CH AVERAGE REACTOR (
wNO: TION- POSITION i
r-e COOLANT TEMPERATURE ( F ) ,
1.r> Run P{w'ER OPERATI0ff Any tcrpcratur4- th l
i 29 S Startup/ Hot Standb[ Any-teeperettree t%
3F 4F (T,SHUTD0 Shutdow Shutdowl,',,****
Q > 200'*1UL-CfLD SHUTD0% *3 < 2004J.-- !
- 5. O *'#
R FUELIf" Shutdown or Refue 3140*fJ h --h '
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_ . . _ . _ ~ . _ _ . _ . . . - . . - - . -. . . -
- ~ ~ ..-... -
- ~ ~ ~ ~
.___.__ ~~~ . . . , _ ~ .
g"9 li fThe reactor mode switch may be placed in the Run, Startup/ Hot Standby,%
or Refuel posit 19n to test the switch interlock functions and related '
b instrumentation provided that the control rods are verified to rerain .
t c. o ., '
'I t o fully inserted by a second licensed operator or other technically '
s1%
qualified member of the unit technical staff. . - -
~- -
~ - . - . . .
m v,o D ##The reactor mode swli.ch fayEf~;ilaced in the Refuel position while a single b, !
o, control rod drive is being removed from the reactor pressure vessel per ,.
Lc o #
Specification 3.9.10.1. Ic:: x t -nc.yp .3 - .
3p9 --
Ud*yTwei in Use 6 9
.eectvr.'c::1witt.tiUTessel.headclosureboltslessthan fully tensioned."^ . N. N'N
..-.-.__...-Z..^._,.*. --
Aqq
@ "See SpeMTcst Scepticn: 3.10.1 and 3.10.0. ,
' ~ ~ - - - - - -- ..--
"U ___
- - .. m !
ho L'o ***The reactor mode switch may be placed in the Refuel position while a singl e control rod is being recoupled or withdrawn provided that the one-rod-out ' 2%
'3.j o. 3 interlock is OPERABLE. , "
% o.N
- ~ . .
h h )) _ SEC.'
~ - _ _
il A \ kdow -
W. .s pay > !
I PERRY-UNITi 1-11 l
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INSERT 12A '
i Insert new Sections 1.2, " Logical _ Connectors," 1.3, '
" Completion Times" and 1.4, " Frequency" as shown in the markup of the Improved Technical Specifications, NUREG-1434. l 1
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ATTACHMENT 1B i S
CTS - PSTS 5
COMPARISON DOCUMENT DISCUSSION OF CHANGES 6
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l DISCUSSION OF CHANCES i CTS: 1 - DEFINITIONS
- r ADMINISTRATIVE A,1 Reformatting and renumbering requirements is in accordance with l the BWR Standard Technical Specifications, NUREG-1434. As a l result, the Technical Specifications should be more readily l readable, and therefore understandable, by plant operators as well as other users. During this reformatting and renumbering process, no technical changes (either actual or interpretation- i al) to the Technical Specifications were made unless they were identified and justified. In the specific case of the
- Definitions Section, no individual numbering of each definition i is made. !
A.2 The sentence is deleted. It serves solely as background or !
basis material and is not incorporated in the BWR Standard !
Technical Specification, NUREG-1434.
A.3 The format of the Actions in the BWR Standard Technical ;
Specification, NUREG-1434, contains specific fields which are' l more accurately presented in this revised wording of the definition. !
A.4 As a requirement for OPERABILITY of a Technical Specification channel, not all channels will have a required sensor or alarm function. Conversely, some channels may have a required ;
displaf function. This is the intent of the existing wording,
[
and therefore the revised wording is proposed to more i accurately reflect this intent; consistent with the BWR !
Standard Technical Specification, NUREG-1434. Since the list i of equipment functions is intended to provide examples of !
attributes which must potentially be OPERABLE, dependent on l whether it is " required" or not, the list can be applied to .!
both analog and bistable channels, and the separate listings ,
can be combined.
s A.5 Usage of the terms "and/or" has been changed to "and" or to ;
"or". The BWR Standard Technical Specification, NUREG-1434, '
l Writer's Guide recommends the use of "and/or" be avoided. The !
intent of the definitions is not changed. j A.6 Editorial rewording / punctuation changes are made consistent with the BWR Standard Technical Specification, NUREG-1434.
During its development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. ;
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4 i PERRY - UNIT 1 1 10/1/93 !
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DISCUSSION OF CHANGES '
CTS: 1 - DEFINITIONS ADMINISTRATIVE '
(continued)
A.7 The words " addition, removal, relocation" have been deleted since " movement" encompasses these words, and " reactivity i controls" are now described as " reactivity control components." '
No change in intent or interpretation is made with this ;
4 proposal. The change is consistent with the BWR Standard
- Technical Specification, NUREG-1434. .
A.8 When CORE ALTERATIONS are required to be suspended, it is !
acknowledged that a specific movement may have to be completed. '
Completing the movement that was in progress at the time of the ;
requirement to suspend is required to establish a " safe" i' configuration (e.g., no fuel bundle suspended from the fuel mast). The requirement to establish a " safe" position is '
deemed proper and sufficient, in accordance with the BWR Standard Technical Specification, NUREG-1434. Eliminating the existing requirement to also be a " conservative" position avoid potential confusion and perhaps overly restrictive s interpretation. Since there is no reference to base the !
conservative evaluation on (i.e. , conservative with respect to '
what?), it is assumed that " conservative" is intended to ,
reflect the same context as " safe." That is, if it is " safe" it is also " conservative." Given this understanding, the ;
wording change is simply editorial. This is acceptable since
" safe" adequately controls the allowance to complete the move. !
A.9 The definition of CRITICAL POWER RATIO has been incorporated ;
into the definition of MINIMUM CRITICAL POWER RATIO. !
A.10 This comment number is not used for this station.
A.11 The definitions of DRYWELL INTEGRITY, FUEL HANDLING BUILDING I INTEGRITY, PRIMARY CONTAINMENT INTEGRITY and SECONDARY :'
CONTAINMENT INTEGRITY have been deleted from the proposed PNPP Technical Specifications. This was done because of the !
confusion associated with these definitions compared to their use in their respective LCOs. The change is editorial in that all the requirements are specifically addressed in the LCOs for ,
the Drywell, Fuel Handling, Primary Containment and Secondary l Containment; along with the remainder of the LCOs in the ;
Containment Systems chapter. Therefore the change is an l administrative presentation preference adopted by the BWR !
Standard Technical Specification, NUREG-1434. [
t A.12 This comment number is not used for this station. '
A.13 The definition of FREQUENCY NOTATION has been deleted since the l abbreviations in the existing Table 1.1 are no longer used.
All Surveillance Requirement Frequencies in the proposed -
, Technical Specifications are directly specified. l f
PERRY - UNIT 1 2 10/1/93 .
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I DISCUSSION OF CHANGES ,
CTS: 1 - DEFINITIONS !
r ADMINISTRATIVE (continued) '
I A.14 The definitions for IDENTIFIED LEAKAGE, PRESSURE BOUNDARY !
LEAKAGE and UNIDENTIFIED LEAKAGE have been combined into one term; LEAKAGE. The definitions of each of the categories of LEAKAGE are consistent with the existing definitions.
?
The definition of Total LEAKAGE has been added for clarity and completeness. The existing use of the undefined term " total lea.kage" is consistent with this proposed definition. !
A.15 As currently specified in the second portion of this definition, the intended leakage is that into the drywell space. The " collection systems" are intended to be those for :
i collection of leakages into the drywell space. This proposed change is simply clarifying the term, and therefore the revised wording is proposed to more accurately reflect this intent;
- consistent with the BWR Standard Technical Specification, NUREG-1434.
A.16 The definition of LOGIC SYSTEM FUNCTIONAL TEST (LSFT) has been modified to not include the actuated device. The actuated i device is (or will be) tested as part of the system functional test. Deleting the actuated device from the definition of LSFT eliminates the confusion as to whether a previously performed j LSFT is rendered invalid if the final actuated device is '
discovered to be inoperable as a consequence of another
- Surveillance (e.g., valve cycling). In instances where the ;
existing Technical Specifications do not contain a corresponding " system functional test" which would test the actuated device, one is being proposed for addition. !
4 Therefore, this change is seen as presenting the same technical ,
requirements, however, part of the existing requirements will be moved to other Specifications.
A.17 These definitions are deleted since the proposed revision to the specific Specifications referring to them no longer contain their use. Discussion of the technical aspects of this change !
are addressed in each Specification where the phrase is removed. The removal of a definition is considered purely j administrative, with no impact of its own, i
l 2
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PERRY - UNIT 1 3 10/1/93 1 i
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DISCUSSION OF CHANGES CTS: 1 - DEFINITIONS I ADMINISTRATIVE !^
(continued) !
A.18 The definitions of OFFSITE DOSE CALCULATION MANUAL and PROCESS i CONTROL PROGRAM have been incorporated into the Administrative Controls Section. Editorial wording changes = are consistent with the BWR Standard Technical Specification, NUREG-1434. ;
1 A.19 OPERATIONAL CONDITION has been deleted and a definition of MODE !
is added to be consistent with terminology sised in the BWR !
Standard Technical Specification, NUREG-1434. This is purely editorial. An additional clarifying statement is added-to indicate that defined MODES in proposed Table 1.1-1 apply only when fuel is in the reactor vessel. This intent was previously
- conveyed by the existing footnote "*" to Table 1.2 (refer also ,
to A. .
A.20 These additions add clarification of the existing requirement without any modification of intent.
A.21 The definition of PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) i has been added consistent with the Improved BWR Standard l Technical Specifications, NUREG-1434. Discussion of the ;
technical aspects of this change are addressed in the '
Specification where limits are replaced with a reference to the PTLR. The inclusion of a definition is deemed purely ,
administrative, with no impact of its own.
A.22 The definition of SHUTDOWN MARGIN has been modified to address stuck control rods. This is consistent with the existing '
requirement found in Surveillance 4.1.1.c to account.for the !
worth of a stuck control rod. The relocation of this requirement is purely editorial, i
A.23 The definition of STAGGERED TEST BASIS has been modified to be consistent with its usage throughout the proposed PNPP Technical Specifications. The intent of the' frequency of testing components on a STAGGERED TEST BASIS is not changed. 3 The revised definition allows the minimum Surveillance interval to be specified in the surveillance Requirements' Frequency ;
column of the applicable LCOs independent of the number of subsystems. This represents a human factored improvement to the current approach, which requires a determination of the i Surveillance sub-interval from the test schedule based on the number of subsystems.
i A.24 These footnotes are covered by the exceptions allowed to LCO -
requirements in the proposed Special Operations Section (currently titled "Special Test Exceptions"). .
Refer to !
proposed LCO 3.10.2, LCO 3.10.3 and LCO 3.10.4. i PERRY - UNIT 1 r
4 10/1/93 i l
DISCUSSION OF CHANGES CTS: 1 - DEFINITIONS i ADMINISTRATIVE (continued)
A.25 The intent of applying the MODE definition only when fuel is in e vessel is relocated to the definition of MODE (refer also k( to cobqt A.19) . Since the vessel head can only be removed if the head closure bolts are less than fully tensioned, there is no purpose in including "or with the head removed." These changes are purely editorial.
A.26 The footnote referencing Special Test Exceptions 3.10.1 and 3.10.3 has been deleted. This footnote only serves as a cross ,
reference and is not needed - consistent with the BWR Standard t Technical Specification, NUREG-1434. !
A.27 The following sections are being added to the Technical These additions aid in the understanding and Specifications. .
use of the new standard Technical Specification format and l style of presentation. Some conventions in applying the Technical Specifications to unique situations have previously been the subject of debate and interpretation by the licensee and the NRC Staff. Because the guidance in these proposed ;
sections is presented in the BWR Standard Technical Specification, NUREG-1434, as approved by the NRC Staff, and :
the guidance is not a specific deviation from anything in the ,
existing Technical Specifications, these additions are treated i as administrative in nature only. The added sections are as follows:
SECTION 1.2 - LOGIC CONNECTORS f Proposed Section 1.2 provides specific examples of the logical connectors "AND" and "OR" and the numbering sequence associated with their use. This revision is being proposed consistent with the BWR Standard Technical ;
Specification, NUREG-1434.
SECTION 1.3 - COMPLETION TIMES ,
l proposed Section 1.3 provides proper use of and I interpretation of Completion Times. The prcposed section also provides specific examples that aid the user in understanding Completion Times. The proposed Completion Times Section is consistent with the BWR Standard Technical Specification, NUREG-1434.
\
l PERRY - UNIT 1 5 10/1/93 L
I
DISCUSSION OF CHANGES CTS: 1 - DEFINITIONS :
i ADMINISTRATIVE (continued)'
SECTION 1.4 - FREQUENCY Proposed section 1.4 provides proper use and interpretation of the Surveillance Frequency. The '
proposed section also provides specific examples that aid the user in understanding Surveillance Frequency. The proposed Frequency Section is consistent with the BWR '
Standard Technical Specification, NUREG-1434.
A.28 The technical content of this requirement is being moved to another chapter of the proposed Technical Specifications. Any technical changes to this requirement will be addressed with the content of the proposed chapter location.
A.29 Specific CHANNEL CALIBRATION requirements for RTDs or thermocouples has been added. The intent of a CHANNEL .
CALIBRATION is to adjust the channel output so that the channel responds with known range and accuracy. Most instrument channels contain an adjustable transmitter (sensor) which is also subject to drift. Thus, for most channels, a CHANNEL ,
CALIBRATION includes adjustments to the transmitter (sensor) to '
re-establish proper input / output relationships. Certain types of sensing elements, by their design, construction and application have an inherent resistance to drift. They are designed such that they have a fixed input / output response which cannot be adjusted or changed once installed. When a credib2e mechanism which can cause change or drift in this fixed response does not exist, it is unnecessary to test them in the same manner as the other remaining devices in the -
channel to demonstrate proper operation. RTDs and thermocouples are sensing elements that fall into such a category. Thus, for these types of sensors, the appropriate calibration at the Frequencies specified in the Technical
~
Specifications would consist of a verification of OPERABILITY of the sensing element and a calibration of the remaining adjustable devices in the channel. Calibration of the adjustable devices in the channel is performed by applying the sensing elements' (RTDs or thermocouples) fixed input /octput relationships to the remainder of the channels and making the necessary adjustments to ensure range and accuracy. '
This proposed " verification of OPERABILITY" of the sensihig element (RTDs or thermocouples) is considered to be documentation of the currently accepted method for calibration of these instruments. As such, this change is considered to be administrative.
A L M u t u d p a l ).
PERRY - UNIT 1 6 10/1/93
- - ..~ _ . .
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DISCUSSION OF CHANGES CTS: 1 - DEFINITIONS j i
ADMINISTRATIVE !
(continued)
A.30 Phrase "at the height" added to the definition to clarify the existing way APLHGR is calculated. Since this does not change the methods currently used for the APLHGR calculations the ,
change is considered administrative. !
RELOCATED SPECIFICATIONS l R.1 This comment number is not used for this station.
TECHNICAL CHANGE - MORE RESTRICTIVE i
h.1 The intent of these changes is to provide clarity and completeness, and as such could be considered administrative.
However, technically, the changes eliminate the potential to interpret certain plant conditions such that no MODE, or a less restrictive MODE, would exist. Therefore, in proposing this change (consistent with the BWR Standard Technical ;
Specification, NUREG-1434), it will be discussed and justified ;
as a "more restrictive" change.
STARTUP MODE will now include the mode switch position of :
" Refuel" when the head bolts are fully tensioned (footnote
"(a)"). This is currently a plant condition which has no corresponding MODE and could therefore be incorrectly interpreted as not requiring the application of the majority of Technical Specifications. By defining this plant condition as '
STARTUP MODE, suf ficiently conservative restrictions will be applied by the applicable LCOs. ,
Clarifying the shutdown MODES with a new footnote stating ,
"all reactor vessel head bolts fully tensioned" eliminates the existing overlap in defined MODES when the mode switch is in l
" Shutdown" position: with the vessel head detensioned, both the definition of REFUEL as well as COLD SHUTDOWN could apply.
It is not the intent of the Technical Specification to allow an option of whether to apply REFUEL applicable LCOs or to apply l COLD SHUTDOWN applicable LCOs. This proposed change precludes an unacceptable interpretation.
The existing definition of REFUEL would cease to be' '
applicable when average coolant temperature exceeded 140 F.
With the mode switch in " Refuel" a plant condition which has no corresponding MODE exists. This could therefore be incorrectly interpreted as not requiring the application of the majority of Technical Specifications. By defining the REFUEL MODE as including plant conditions with no specific coolant temperature !
range, sufficiently conservative restrictions will be applied l by the applicable LCOs during all fueled conditions with the >
vessel head bolts detensioned.
PERRY - UNIT 1 7 10/1/93
l DISCUSSIOl; OF CllANGES i CTS: 1 - DEFINITIONS TECHNICAL CHANGE - LESS RESTRICTIVE i
"Genaric
None in this section.
" Specific" L.1 The phrase "or actual," in reference to the injected signal, has been added to the definition of CHANNEL FUNCTIONAL TEST.
Some CHANNEL FUNCTIONAL TESTS are performed by insertion of the actual signal into the logic (e.g. , rod block interlocks) . For -
others, there is no reason why an actual signal would preclude satisfactory performance of the test. Use of an actual signal instead of the existing requirement which limits use to a simulated signal, will not affect the performance of the channel. OPERABILITY can be adequately demcnstrated in either case since the channel itself can not discriminate between ,
" actual" or " simulated."
L.2 As provided for with analog channels, the signa 1 used to test ;
bistable channels is proposed to be allowed to be injected "as '
close to the sensor as practicable." Injecting a signal at the ;
sensor would in some cases involve significantly increased probabilities of initiating undesired circuits during the test since several logic channels are often associated with a particular sensor. Performing the test by injection of a signal at the sensor requires jumpering of the other logic channels to prevent their initiation during the test, or increases the scope of the test to include multiple tests of other logic channels. Either method significantly increases the
- difficulty of performing the surveillance. Allowing initiation of the signal close to the F.ensor provides a complete test of the logic channel while significantly reducing this probability of undesired initiation.
f PERRY - UNIT 1 8 10/1/93 >
DISCUSSION OF CHANGES L CTS: 1 - DEFINITIONS i
TECHNICAL CHANGE - LESS RESTRICTIVE (continued)
L.3 " Normal" movement of SRMs, IRMs, LPRMs, TIPS or special movable ,
detectors (i.e., incore instruments) is not considered a CORE '
ALTERATION by the existing definition. In this definition, no delineation of what is and is not considered " normal" movement is given. This has lead to some confusion and perhaps overly restrictive interpretatien. The proposed change focuses the definition on activities that can affect the core reactivity. -
Since incore instruments have negligible .(if any) affect on core reactivity, any movement of incore instruments has essentially no impact on core reactivity. Therefore, the proposed change places no restrictions on incore instrument movement.
Maintaining CORE ALTERATIONS as movement of only that which can affect core reactivity is consistent with the BWR Standard Technical Specification, NUREG-1434. The basis for this is I evident in that the Specifications that are applicable during CORE ALTERATIONS are those that protect from or mitigate a ;
reactivity excursion event. '
L.4 Consistent with the rationale presented in item L.3 above ,
(i.e. , the proposed change focuses the definition on activities i that can affect the core reactivity), an additional change is ,
proposed to allow the physical removal of a control rod to not be considered a CORE ALTERATION. In this activity the control cell must first have all the fuel bundles removed prior to this i control rod movement. In this configuration, the negative reactivity inserted by removing the adjacent four fuel assemblies is significantly more than any minimal positive reactivity inserted durir.g the removal of the control rod.
Appropriate Technical Specification controls are applied during the fuel movements preceding the control rod removal to protect from or mitigate a rer , ty excursion event. After such i time, sufficient narg t am uesign features (the design of a ~
control rod precludes :r tsmoval without all fuel assemblies in the cell removed) -
- in place to allow removing the Technical Specification controls during the control rod s removal. This proposed change is consistent with the BWR !
Standard Technical Specification, NUREG-1434.
r f
l PERRY - UNIT 1 9 10/1/93
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CTS - PSTS ;
COMPARISON DOCUMENT :
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1 i
NO SIGNIFICANT HAZARDS 't CONSIDERATIONS i r
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s NO SIGNIFICANT HAZARDS CONSIDERATIONS {
CTS: 1 - DEFINITIONS. ;
"L1" CHANGE PNPP has evalusted 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?
The phrase "or actual," in reference to the injected signal, has been added to the definition of CHANNEL FUNCTIONAL TEST.
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 revisions to them, dictate the accept-ability of generating this signal. The proposed change does not affect the procedures governing plant operations and the a acceptability of creating these signals; it simply would allow such a signal to be utilized in evaluating the acceptance criteria for OPERABILITY of an instrument channel. Therefore, the change does not involve a significant increase in the probability of an accident previously evaluated.
Since the function of the channel remains unaffected, and no changes result to any setpoints, the change does not involve 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 possibility of a new or different 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.
- 3. Does this change involve a significant reduction in a margin of safety?
Use of an actual signal instead of the existing requirement which limits use to a simulated signal will not affect the performance of the channel. OPERABILITY is adequately demonstrated in either case since the channel itself can not discriminate between " actual" or " simulated." Therefore, the change does not involve a significant reduction in a margin of safety.
PERRY - UNIT 1 1 10/1/93
l l
NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 1 - DEFINITIONS "L2" CHANGE i
PNPP 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 l 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?
Testing of bistable instrument channels such that the test signal does not include the " sensor" will significantly reduce ,
the complications associated with performance of a surveillance .
on a sensor that provides input to multiple logic channels. i This change will not affect the failure probability of the ,
equipment. But this potential extension of the surveillance interval of the sensors will slightly increase the probability of the sensors being failed upon demand to operate. However, this slight increase is offset by the reduction in complication which reduces the probability of personnel error during the '
surveillance. 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 possibility of a new or different 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.
- 3. Does this change involve a significant reduction in a margin of safety?
This change does not involve a change to the limits or limiting conditions of operation, but only to the method for performing a surveillance. This change will not affect the failure probability of the equipment. But this potential extension of the surveillance interval of the sensors will slightly increase the probability of the sensors being failed upon demand to operate. However, this slight increase is offset by the [
reduction in complication which reduces the probability of "
personnel error during the surveillance. Therefore this change does not involve a significant reduction in a margin of safety.
PERRY - UNIT 1 2 10/1/93
i NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 1 - DEFINITIONS "L3" CHANGE j i
PNPP 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?
The SRMs, LPRMs, IRMs, TIPS and special movable detectors are not assumed to cause the initiation of any analyzed event. .
Their movement will not cause reactivity in the core to be .
discernibly changed. Therefore, this proposed change will not ,
involve a significant increase in the probability of an !
accident previously evaluated.
The OPERABILITY of the incore detectors to function in mitigation of analyzed events is unaffected by this change.
The proposed change involves allowing movement of these incore ;
detectors while not enforcing requirements necessary for CORE ALTERATIONS. Since there would be no concurrent CORE ALTERATIONS at this time (if there were, then the requirements i for CORE ALTERATIONS would be independently applied), no i analyzed event is assumed. Therefore, this proposed change will not involve a significant increase in the consequences of an accident previously evaluated. }
t
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? l The proposed change does not involve physical modification to t the plant. OPERABILITY requirements for the incore instruments remains required when necessary for monitoring and event '
mitigation. Movement of the detectors involves:
1
- normal retraction to below the core plate via an installed i drive system (this movement is excluded from the existing ,
definition of CORE ALTERATIONS); :
- manual retraction to the under-vessel cavity for replacement i and subsequent reinsertion of a new detectors. This activity i causes no potential of changing core geometry or core '
reactivity; !
- in-vessel movement of " dummy" assemblies containing special moveable detectors (this movement is excluded from the existing t
definition of CORE ALTERATIONS).
PERRY - UNIT 1 3 10/1/93 i
NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 1 - DEFINITIONS "L3" CHANGE (continued)
Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.
- 3. Does this change involve a significant reduction in a margin of safety?
The SRMs, LPRMs, IRMs, TIPS and special movable detectors are not assumed to cause the initiation of any analyzed event.
Their movement will not cause reactivity in the core to be discernibly changed. The OPERABILITY of the incore detectors to function in mitigation of analyzed events is unaffected by this change. The proposed change involves allowing movement of these incore detectors while not enforcing requirements necessary for CORE ALTERATIONS. Since there would be no concurrent CORE ALTERATIONS at this time (if there were, then the requirements for CORE ALTERATIONS would be independently applied), no analyzed event is assumed and therefore, this proposed change will not involve a significant reduction in a margin of safety.
PERRY - UNIT 1 4 .
10/1/93 Il o
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l NO SIGNIFICANT HAZARDS CONSIDERATIONS i CTS: 1 - DEFINITIONS l l
"L4" CHANGE l PNPP has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards cons.ideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following j evaluation is provided for the three categories of the significant l hazards consideration standards:
- 1. Does the change involve a significant increase in the ;
probability or consequences of an accident previously ;
~
evaluated?
The proposed change, to allow the physical removal '(i.e. ,
movement other than with the normal control rod drive) of a_ ,
control rod to not be considered a CORE nLTERATION, involves I first removing all the fuel bundles in a cell prior to this j control rod movement. In this configuration, the negative reactivity inserted by removing the adjacent four fuel assemblies is significantly more than any minimal positive ,
reactivity inserted during any movement of the control rod. ,
Therefore the probability of an unexpected positive reactivity !
insertion event is not significantly increased. !
Since no reactivity insertion event is expected as a result of i the control rod movement, and since there vould be no l concurrent CORE ALTERATIONS at this time (if there were, then >
the require.ments for CORE ALTERATIONS would be independently l applied), no analyzed event unique to CORE ALTERATIONS (note: l other requirements, such as those for handling loads over irradiated fuel), will remain applicable. Therefore, this F proposed change will not involve a significant increase in the 3 consequences of an accident previously evaluated.
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PERRY - UNIT 1 5 10/1/93 j I
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1 NO SIGNIFICANT HAZARDS CONSIDERATIONS I CTS: 1 - DEFINITIONS "L4" CHANGE l (continued) l
- 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 physical modification to i the plant. Movement of a control rod with other than with the >
normal control rod drive involves unlatching and ,
withdrawal / insertion from over-vessel handling equipment.
These activities necessitate, by design, the removal.of the adjacent four fuel assemblies. With this configuration (no ;.
fuel in the cell; handling the associated control rod), the proposed change will allow movement of a " reactivity control i' component" while not imposing requirements unique to CORE ALTERATIONS (note: other requirements, such as those for i handling loads over irradiated fuel, will remain applicable). 1 Since the reactivity affects of this control rod movement are more than compensated for by the initial removal of the fuel !
assemblies, this new activity-does not create the possibility. I 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?
Since the negative reactivity inserted by removing the adjacent i four fuel assemblies is significantly more than any minimal '
positive reactivity inserted during any movement of the control rod, not considering the proposed activity (movement of a i control rod, other than with the normal control rod drive) to i be a CORE ALTERATION does not involve a significant reduction _i in a margin of safety. !
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PERRY - UNIT 1 6 10/1/93
ATTACHMENT 2 :
ITS - PSTS 1 COMPARISON DOCUMENT i
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Definitions 1.2 1.0 USE AND APPLICATION 1.1 Definitions
-.--------------------.----.--NOTE-------------------------------------
The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.
Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
~
AVERAGE BuliDILE'TXF03ttRE T E BUNDLE EXPOSURE shall beJguaL4e-the sum of rage &Trposure of the fuel specifi "ded by the number of fuel rods in the fuel bundle.
D ;tnm " AR EXPOSURE The AVERAGE PLANAR EXPOSURE shall be applicahlm.ttf O a s ecific planar height and M o the sum
\b] -
o . -
,ce2= & ii tne fuel rods in the specified bundle at the 7 eciB ed height divided g by the number of fuel rods in the fuel bundle. %,
AVERAGE PLANAR LINEAR The APLHGR shall be applicable to a specific HEAT GENERATION RATE planar height and is equal to the sum of the (APLHGR) 9 LHGRs}# h at ganerM un rate por grit -lengt4H4 s ful red; for all the fuel rods in the specified Q bundle at the specified height divided b numberoffuelrodsinthefuelbundle$ythe at the height)%
CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass 4 the entire channel, including the required sensor, alam, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors consist of an inplace cr;;; cclitr;ticf =W 4 a s
{---+ renig-elcrds and nomal_c.alikELtion of the -
Q. l . h% c c,4 5 u s~ ,-4
' ot 6 m bb Q h (continued) !
PERRV - umT 1 o EWR/fr-55 "cv. O, 00/20/9e 1.[-1 ed
% yD -
v
-4+ w
Definitions 1.1 1.1 Definitions CHANNEL CALIBRATION remaining adjustable devices in the channel.
(continued) Whencier ; sen;ing c h ment i; repieced, the next required- irglect eres;--;clibrati:n :;n;ists of O :::p; ring thc ether -sca;ing ch::nt: -ith the
' r+cco;1j instslied ,c sing : h : r,t. The CHANNEL CALIBRATION may be performed by 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 behavi<r during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be
- 4.- -
G. '=le; 9-re1Qhe injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarm, interlock, display, and trip functions, and channel failure trips.4 ct
~
- t. "i;tek 6nnch '; .3. , g casui c . it & c ad nitch wwntectd -- the injcc;.vu vf a sim i ted er h iusi signoi iniv ihc channel es wisse t th: ,caser as practicatric te verify OProen" !"', inuiuding s epired 9 e nd + rip fUnuivna.
The CHANNEL FUNCTIONAL TEST may be perfomed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is ,j tested. ~'
CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, e sources,+ reactivity control components
'! cc7"eatr ef fe: ting r:as .iy withinYa. t eorreactor 6EFia/ P--
vessel with the vessel head removed and fuel in the vessel. Movement of source range monitors, local power range monitors, intermediate range (continued)
% R/G STS P( n y V r.'-t 1.1-2 Rev. O, 0948M2-
Definitions !
1.1 .
1.1 Definitions '
I CORE ALTERATION monitors, traversing incore probes, or special i (continued) movable detectors (including undervessel replacement) is not considered a CORE ALTERATION.
In addition, control rod movement with other than ;
the nonnal control rod drive is not considered a -
CORE ALTERATION provided there are no fuel i assemblies in the associated core cell. 5 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. These cycle specific limits !'
shall be determined for each reload cycle in accordance with Specification 3.".1. 0 Plant operation within these limits is/ addressed in individual Specifications. h,g,(,g...h DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration i of I-131 (microcuries/ gram) that alone would produce the same thyroid dose as the quantity and 6 isotopic mixture of I-131 I-132 I-133, I-134, ;
and I-135 actua esent. The thyroid dose conversion fac r ed for this calculation shall be those listed n able III of TID-14844 AEC, 1962, "Calcu ation of Distance Factors for Power and Test Reactor Sites," or these listed irr h ' gleg ei Rcguhtery Cuide 1.109, Re.
nne, u , , ; .-
1,-
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$-AVERAGE $ shall be the average (weighted in proportion
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DISINTEGRATION ENERGY to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the J sum of the avera -
Q disintegration (ge beta and ansna in MeV s6t' s, energies other than per- .j iodines, with half live utes, making up I at least 9554 of the total noni ne activity in 1 the coolant.
EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval ,
SYSTEM (ECCS) RESPONSE from when the monitored parameter exceeds its ECCS T1HE 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)
-0WR/6 ~;i."c Pu r , \.L. A i 1.1-3 Ren 0, 09/2Sf92
Definitions 1.1 1.1 Definitions EMERGENCY CORE COOLING required positions, pump discharge pressures reach SYSTEM (ECCS) RESPONSE theirrequiredvalues,etc.). Times shall include 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.
END OF CYCLE The E0C-RPT SYSTEM RESPONSE TIME shall be that )
RECIRCULATION PUMP TRIP time interval from initial ngnal gcneration by p ?_
(EOC-RPT) SYSTEM RESPONSEWthe associated turbine stop valve limit : witch or TIME frca h the turbine control valve hydraulic ci-1 ccntrol cil prc:sure drept b 10.; the presme switch retpcint] to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. 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,
[e m pt +er the t,reekei m u aspieTf'ci, time, h which is not ::::ured but i: -;;1? dated t: 0 0 M0 '
to the c;nefacturer's d::ign c:!uc]
ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that RESPONSE TIME time interval from when the monitored parameter exceeds its isoiatT6Msetpoint at the channel O M bh g sensor until the isolation velves travel to their required positions. Times snall include diesel generator starting and sequence loading 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.
LEAKAGE LEAKAGE shall be:
- a. Identified LEAKAGE I. LEAKAGE into the drywell such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or (continued)
Esi/6 5T-5 9w, . V J t 1.1-4 S e 0, 0g/28/g2
Definitions 2.1 1.1 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 drywell that is not <
identified LEAKAGE; l
- c. Total LEAKAGE Sum of he identified and unidentified LEAKAGE,
- d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a P.eactor Coolant System (RCS) component body, i pipe wall, or vessel wall.
'i k ~ LINEAR HEAT GENEUTION RATE (LHGR)
The LHGR shall be the heat generation rate per unit length of fuel rod. It is the integral of
~f the heat flux over 1e heat transfer area
~ associated with_the_ unit length. ~
LOGIC SYSTEM FUNCTIONAL
v?$ r A LOGIC SYSTEM FUNCTIONAL TEST hall be a test C1 f alla ogic components (i.e., al elays and TEST '
contacts, trip units, solid state logic elements, ,
etc oo ep.p.3 sens.) or of a logic circuit, as practicable from up to, but as close to the not including,
~ the actuated *dcvicc, to verify OPERABILITY. The '
LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps 50 that the entire logic system is tested.
MAXI h 0N ~ ... The MFLPD shall be the largest value of the .
OF LIMITING fraction of. limiting power density in the core. :
POWER DENSITY (HFLPD) The fraction oY Timiting power density shall be the LHGR existing at a given 16'c ation-divided by
_ the specified LHGR limit for that bundle type.' - _
(continued) s R/6 STS b 5 -C ' 1.1-5 Rev. O, Og/20/g2-
Definitions 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 core (for each class of fuel 39 The CPR is that power in the assembly that is calculated by application of the appropriate 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 reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in ble 1 1-I with fuel in the react.or_yess.el u '
OPERABLE-OPERABILITY C S 'd wmW e %ve O PEM G.u W;' C synem, suosystem, 'tniiG'Tomp0n nt For ' device shall be OPERABLE en it is capable of performing its specified safety function (s) and when all necessary attendant instrumentation, controls, nomal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary
. h equipment that are required for the system, ,
subsystem,4trem, component, or device to perfom its specified safety function (s) are also capable of performing their related support function (s).
I PHYSICS TESTS PHYSICS)EJTS'shall be those tests performed to m measureAhe fundamental nguar characteristics of .
N s the actor core and r tfted instrumentatio3,/
se tests are: f
- ./ a. Describe Chapter [14, Init dl Test '
eg - Program) f the fSAR;
\ b. Authorized under the.pr'ovisions of 'f
-10 CFR 50.59; ory' '
\s c. Otherwise. approved by the Nuclear Regulatory Commission.
PRESSURE AND The PTLR is the unit specific document that TEMPERATURE LIMITS provides the reactor vessel pressure and REPORT (PTLR) temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits (continued)
E'm/C 53 h r t , V < A i 1.1-6 h. G, 09/28/92
Definitions 1.1 1.1 Definitions p g,l,Q PRESSURE AND shall be detemined for each fluence period in TEMPERATURE LIMITS accordance with Specification 79.! M Plant REPORT (PTLR) operation within these operating limits is (continued) addressed in LC0 3.4.11. "RCS Pressure and ,
Temperature (P/T) Limits."
RATED THERMAL POWER RTP shall be a total reactor core stjtransf rate to the reactor coolant of
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(RTP) -
REACTOR PROTECTION n9)MWt.
The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until l de-energization of the scram pilot valve solenoids. 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.
SHUTOOWN MARGIN (SDM) SDM shall be the arount of reactivity by which the reactor is subtritical or would be subtritical assuming that:
- a. The reactor is xenon free;
- b. The moderator temperature is 68'F; and
- c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
Nhcontrolrodsnotcapableofbeingfully inserted, the reactivity worth of these control C rods must be accounted for in the detemination of SDM N.
- STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, -
channels, or other designated components during i the interval specified by the Surveillance '
Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function. '
(continued)
--BWR/6-SM 9c (<, V A ' 1.1-7 Rev. O, 09/2S/93 '
f
Definitions 1.1 1.1 Definitions (continued)
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. ,
TURBINE BYPASS SYSTEM The TURBINE BYPASS SYSTEM RESPONSE TIME consists RESPONSE TIME of two compo s: ,
\' cc E} -
- a. The ime for nitial movement of the main 31.,
turbine stop valve or control valve until 80%
[2 g of the turbine bypass capacity is established; and C4 Fe s -,
- b. The time nitial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve.
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.
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04 /5 STL Is.frg \)n.\ \ 1.1-8 Rr. . O, 09/25/92 !
Definitions 1.1 Table 1.1-1 (page 1 of 1)
H0 DES REACTOR H0DE AVERAGE REACTOR MODE TITLE SWITCH POSITION COOLANT TEMPERATURE
(*F) 1 Power Operation Run NA 2 Startup Refuel (a) or Startup/i ot NA Standby 3 Hot Shutdown (a) Shutdown > 200f 4 ColdShutdown(a) 4 5 Refueling (b) Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned.
(b) One or more reactor vessel head closure bolts less than fully tensioned.
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IFJR/S STL b "1 U ^^- 1.1-9 Rr. . O, 00/ 3/02
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Legical Connectors -
1.2 1.0 USE AND APPLICATION 1.2 Logical Connectors
._=
PURPOSE The purpose of this secticn is to explain the meaning of logical connectors.
Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and frequencies. The only logical connectors that appear in TS are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings.
BACKGROUND Several levels of logic may be used to state Required Actions. These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the number of the Required Action). The successive levels of logic are identified by additional digits of the Required Action number and by successive indentiotis of the logical connectors. i
~ When logical connectors are used to state a Condition, only' W OCM the first level of logic is used, and the logical connector W eft justified with the Condition statemen p -
l When logical connectors are used to state aOCompletion Timed Surveillance, or frequency, only the first level of logic is !
used, andofthe statement the logical connector Completion is left justified Time, Surveillance, or with the ,/
Frequency. W RPonc'l.h &
EXAMPLES The following examples illustrate the use of logical '
connectors.
l (continued) l 1
Ee/0 STS Eu5 M \ 1. -10 (#6 h.. O,00/20/92-
Logical Connectors 1.2 1.2 Lorical Connectors 1
EXAMPLES EXAMPLE 1.2-1 (continued) '
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i
A. LCO not met. A.1 Ve ri fy . . . !
AND A.2 Restore . . .
In this example, the logical connector AND is used to indicate that, when in Condition A, both Required Actions A.1 and A.2 must be completed.
I P
{ continued) i
%'o/6 STS- 9e.tt3 Va' 4 \ 1.2-11 Ocv. O, 0;/20/02 i
i
t Logical Ccnnectors 1.2 l 1.2 Logical Connectors '
EXAMPLES EXAMPLE I.2-2 (continued)
ACTIONS
~
CONDITION REQUIRED ACTION COMPLETION TIMs ,
A. LCO not met. A.1 Trip . . .
l
_OE A.2.1 Veri fy . . . ,
i AND A.2.2.1 Reduce . . . !
A.2.2.2 Perform . . .
E3 !
A.3 Align . . .
i This example represents a more complicated use of logical connectors. Required Actions A.1, A.2, and A.3 are alternative choices, only one of which must be performed as ;
indicated by the use of the logical connector QR and the left justified placement. Any one of these three Actions ;
may be chosen. If A.2 is chosen, then both A.2.1 and A.2.2
- must be performed as indicated by the logical connector AND. !
Required Action A.2.2 is met by performing A.2.2.1 or A.2.2.2. The indented position of the logical connector 1
QR indicates that A.2.2.1 and A.2.2.2 are alternative !
choices, only one of which must be performed.
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?>5J6 ST& 9stq Um A ' 1.2-12 Acw. O, 00/24 53-t
Completion Times 1.3 f
1.0 USE AND APPLICATION I.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.
6.k$ C.JA... C.,Opu h 2 @
BACKGROUND (LCOs));pecify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which '
the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action (s) and CompletionTime(s).
i DESCRIPTION The Completion Time is the amount of time allowed for i completing a Required Action. It is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Ccndition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO Applicability.
If situations are discovered that require entry into more than one Condition at a time within a single LCO (multiple Conditions), the Required Actions for each Condition must be performed within the associated Completion Time. When in multiple Conditions, separate Completion Times are . tracked for each Condition starting from the time of discovery of the situation that required entry into the Condition. _
Once a Condition has been entered, subsequent N n 4 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 unless ;
specifically stated. The Required Actions of the Condition l continue to apply to each additional failure, with Completion Times based on initial entry into the Condition. ;
(continued) mR/c ST& Pur v U d.1 \ (0 0 p g pe e. O, 09/23/02-
Completion Times 1.3 1.3 Completion Times DESCRIPTION However, when a subsecuent tra SODS, ubsystem, component, or (continued) variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time (s) may be extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability:
- a. Must exist concurrent with the first inoperability; and
- b. Must remain inoperable or not within limits after the first inoperability is resolved.
The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either:
- a. The stated Completion Time, as measured from the initial entry into the Condition, plm an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; or
- b. The stated Completion Time as measured from discovery of the subsequent inoperability doe '4 The above Completion Time extensio i tie ot apply to those Specifications that have exceptions at allow completely separate re-entry into the Condition (for eachurain,yMsg Ocs subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual '
Specifications.
The above Completion Time extension does not apply to a Completion Time with a modified " time zero." This modified
" time zero" may be ex)r'essed as a repetitive time (i.e.,
"once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />," w1ere the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery . . ." Example 1.3-3 illustrates one use of this type of Completion Time. The 10 day Completion Time specified for Conditions A and B in Example 1.3-3 may not be extended.
(continued) e=/; sis 9ct y V 4 d i 1.3-14 Fev. O, 00/2S/02
Completion Times 1.3 1.3 Completion Times (continued)
EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions.
EXAMPLE 1.3-1 '
i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action and -
associated AND Completion Time not B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> met.
Condition B has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition B is entered.
The Required Actions of Condition B are to be in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. A total of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is allowed for reaching MODE 3 and a total of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (not 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) is allowed for reaching MODE 4 from the time that Condition B was entered. If MODE 3 is reached within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, the time allowed for reaching MODE 4 is the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> because the total time allowed for reaching MODE 4 is 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
If Condition B is entered while in MODE 3, the time allowed for reaching MODE 4 is the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
t (continued)
C'Jn/5 STS ic t r 1 \3 ' k i 1.3-15 Fev 0, 09/?S/92-
Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump A.1 Restore pump to 7 days '
inoperable. OPERABLE status.
B. Required B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> met. !
When a pump is GI SQQ clared inoperable, Condition A is entered.
- If the pump is not estored to OPERABLE status within 7 days, Condition P is entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable pump is restored to OPERABLE status after Condition B is entered, he Required Actions of Condition B may be terminated. % ,, IN ,,,.) G ol J.NJ, c[-
When a second pump is declared ino~jieTa'ble while the first pump is still inoperable, Condition A is not re-entered for the second pump. LCO 3.0.3 is entered, since the ACTIONS do not include a Condition for more than one inoperable pump.
The Completion Time clock for Condition A does not stop after LCO 3.0.3 is entered, but continues to be tracked from the time Condition A was initially entered.
While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has not expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition A.
(continued) m:n/S STS- 9 rr 3 Vel.k \ 1.3-16 h.. O, 00/20/92
Completien Times I
1.3 1.3 Completion Tim:s !
EXAMPLES EXAMPLE 1.3-2 (continued)
While in L'C0 3.0.3, if one of the inoperable pumps is !
restored to OPERABLE status and the Completion Time for Condition A has expired, LCO 3.0.3 may be exited and i operation continued in accordance with Condition B. The . i Completion Time for. Condition B is tracked from the time the Condition A Completion Time expired. .l On restoring one of the pumps to OPERABLE status, the i Condition A Completion Time is not reset, but continues from the time the first pump was declared inoperable. This a Completion Time may be extended if the pump restored to '
OPERABLE status was the first inoperable pump. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ,
extension to the stated 7 days is allowed, provided this l' does not result in the second pump being inoperable for
> 7 days. :
L s
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(continued)- -
i
.)
CWP/05T5 kt.fr3 -\Jc:A\ 1.3-17 hv.- 0, 09/?8/92 1
l
Completion Times 1.3 1.3 Completion Times '
EXAMPLES EXAMPLE 1.5-3 i (continued)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ;
A. One A.1 Restore 7 days i Function X function X '
subsystem subsystem to AND '
inoperable. OPERABLE status. I 10 days from '
discovery of failure to meet the LC0 B. One B.1 Restore 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> !
Function Y Function Y subsystem subsystem to AND inoperable. OPERABLE status.
10 days from discovery of failure to meet the LCO i t
C. One C.1 Restore 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> function X Function X subsystem subsystem to :
inoperable. OPERABLE status.
AND OR One C.2 Restore 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> function Y Function Y subsystem subsystem to inoperable. OPERABLE status.
i f
I (continued)'
Own/5STS9er, A A\ 1.3 18 3 ,, 3,- cgj;;f r i
Ccmpletion Times ,
1.3 1 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued) sggjh@
When one Function X tres an e function Y tren are _\ !
inoperable, Condition A and Condition _ B are concurrently i
applicable. The Completion Times for Condition A and - !
Condition B are tracked separately for each tsa.i.nfstarting i from the time each WIwas declared inoperable and the Condition was entered. A separate Completion Time is .
established for Condi.t_ionlanttrackeltrom the time the _. J second tsabbwas dic~lared inoperable ti.e., the time the situation described in Condition C was discovered). ;
If Required Action C.2 is completed within the specified ;
Completion Time, Conditions B and C are exited. If the Completion Time for Required Action A.1 has not expired, '
operation may continue in accordance with Condition A. The remaining Completion Time in Condition A is measured from the time the affected M was declared inoperable (i.e.,
initial entry into ConditionTQAs,sg Q -
r The Completion Times of Conditions A and B are modified by a logical connector, with a separate 10 day Completion Time '
measured from the time it was discovered the LCO was not '
met. In this example, without the separate Completion Time, it would be possible to alternate between Conditions A, B, ;
and C in such a manner that operation could continue
- indefinitely without ever restoring systems to meet the LCO.
The separate Completion Time modified by the phrase "from .
discovery of failure to meet the LC0" is designed to prevent t indefinite continued operation while not meeting the LCO. '
This Completion Time allows for an exception to the nonnal -
- time zero" for beginning the Completion Time " clock". In !
this instance, the Completion Time " time zero" is specified as commencing at the time the LCO was initially not met, ;
instead of at the time the associated Condition was entered. -
i (continued) 0"R/0 STS kt HM -U k i 1.3-19 En. G,65/20/92 i
f Completion Times 1.3 1.3 Completion Times ,
EXAMPLES EXAMPLE 1.3-4 (continued)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restorevalve(s) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> valves to OPERABLE inoperable. status.
B. Required B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> met.
1 A single Completion Time is used for any number of valves inoperable at the same time. The Completion Time associated with Condition A is based on the initial entry into Condition A and is not tracked on a per valve basis.
Declaring subsequent valves inoperable, while Condition A is still in effect, does not trigger the tracking of separate Completion Times.
Once one of the valves has been restored to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first valve was declared inoperable. The Completion Time may be extended if the valve restored to OPERABLE status was the first inoperable valve. The Condition A Completion Time may be extended for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provided this does not result in any subsequent valve being inoperable for > 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. - --
If the Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ( f3 extensio[
expires while one or more valves are still inoperable, Condition B is entered.
(continued)
-0W;/5-;;; 9tr i g -\)el.\ \ 1.3-20 Aer 0, 00/20/92
Completion Times I 1.3 r
1.3 Completion Times t i
EXAMPLES EXAMPLE 1.3-5 (continued)
- ACTIONS
NOTE---------------------------- l Separate Condition entry is allowed for each inoperable '
valve.
i CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.I Restore valve to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> valves OPERABLE status.
B. Required B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> '
Action and associated AND Completion i Time not B.2 Be in NODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> met.
The Note above the ACTIONS table is a method of modifying how the Completion Time is tracked. If.this method of modifying how the Corr letion ' e is tracked wa, applicable q4 only to Condition the Not mayhippear in tha*' Condition 3 n ~ h D 3{ir.Z.T,,*E) ou\ A + h.t k.. P l -
The Note allows Condition A to be entered a parately for
.%rf%
rt. e *W'-rw each inoperable valve, and Completion Times tracked on a per valve basis. When a valve is declared inoperable, Condition A is entered and its Completion Time starts. If subsequent valves are declared inoperable, Condition A is entered for each valve and separate Completion Times start and are tracked for each valve.
(continued) ;
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l' Cl,/5 STS 9c t t 3 V K.4 1 1.3-21 D=" 0, 07/24/G2-
Completion Times !
1.3 ,!
1.3 Completion Times !
l i
EXAMPLES EXAMPLE 1.3-5 (continued) -
If the Completion Time associated with a valve in- -f Condit'on A expires, Condition B is entered for that valve. !
If the Completion Times associated with subsequent valves in !
Condition A expire, Condition B is entered separately for :
each valve and separate Completion Times start and are {
tracked for each valve. If a valve that caused entry into Condition B is restored to OPERABLE status, Condition B is exited for that valve. ,
i Since the Note in this example allows multiple Condition. ,
entry and tracking of separate Completion' Times, Completion f Time extensions do not apply. j EXAMPLE 1.3-6 -
ACTIONS I CONDITION REQUIRED ACTION COMPLETION TIME i i
A. One channel A.1 Perform Once per inoperable. SR 3.x.x.x. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> !
0R l
A.2 Reduce THERMAL 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> :
POWER to s 50% RTP. l!
B. Required B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 3
Action and associated - "
Completion !
Time not i met. i 1
(continued) !
I i
SEn/5 STS 9e.tf t-() '.h 4 1.3-22 P,v n, ng/23/93_ ;
n
Completion Times i 1.3 .
1.3 Completion Tires EXAMPLES EXAMPLE 1.3-6 (continued) i Entry into Condition A offers a choice between Required {
j Action A.1 or A.2. . Required Action A.1 has a "once per" !
Completion Time,.which qualifies for the 25% extension, per 1 SR 3.0.2, to each performance after the initial performance. '
If Required Action A.1 is followed and the Required Action' !
, C.4 is not met within the Completion Time (tec!ecism the G57- las i extension allowed by SR 3.0.2), Condition B is entered. If j Required Action A.2 is followed and the Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is not met, Condition B is entered, t If after entry into Condition B, Required Action A.1 or A.2 is met, Condition B is exited and operation may then !
continue in Condition A. i i
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{
BWR/; ;;; 9ttr$ .\)p.\ \ 1.3-23 Rev 0, 09/2C/02
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- -- . . . -. -. - _-___-.___ ____ __ ._.____ __ _ _ _ l
Completion Times I 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Verify affected I hour .
subsystem subsystem inoperable. isolated. AND Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter AND A.2 Restore subsystem 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to OPERABLE ,
status.
B. Required B.1 Be in H0DE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 1 Action and associated AND .
Completion !
Time not B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> met.
Required Action A.1 has two Completion Times. The I hour Completion Time begins at the time the Condition is entered r and each "Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter" interval begins upon performance of Required Action A.1.
If after Condition A is entered, Required Action A.1 is not met within either the initial I hour or any subsequent plus
_! 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval from the previous > e j G33) extension allowed by SR 3.0.2) performance
, Condition (6hc:uct B is entered.
The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time i
Condition A was initially entered. If Required Action A.1 .
(continued)
Em/CSTSkr.-(fM 'Un 4 5 1.3 24 nc<. c. c:/: ssa i
e
Completion Times 1.3 1.3 Completion Times RAMPLES EXAMPLE I.3-7 (continued) is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided 1 Na far Dequired artian has not Jex fred.S')Jt'eetheCc7 atian s and Compi tTon Time o equired as a . ified "ti zero" (i.* , after t h L(Action A).
initiaV1 hour ot from t of Condi ~ n entry) he lp 4nce fp a Completi Time ext ion does t apply.
IMMEDIATE When "Imediately" is used as a Completion Time, the COMPLETION TIME Required Action should be pursued without delay and in a controlled manner.
^'
0=/5 STS itiN 1.3-25 Rc. .- 0, 00/2G/92-
Frequency ,
1.4 1.0 USE AND APPLICATION ,
1.4 Frequency '
PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.
DESCRIPTION Each Surveillance Requirement (SR) has a specified frequency ,
ld. - in which the Surveillance must be met in order to meet the associated *(LCO') An understanding of the correct application G g,9 g, of the specified Frequency is necessary for compliance with qq ,
the SR.
Q D p' " The "specified Frequency" is referred to throughout this isection and each of the Specifications of Section 3.0,
- " Surveillance Requirement (SR) Applicability. The "specified Frequency" consists of the requirements of the frequency column of each SR, as well as certain Notes in the Surveillance column that modify performance requirements.
Sometimes special situations dictate when the requirements !
of a Surveillance are to be met. They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Surveillance, or both. Example 1.4-4 discusses these -
special situations.
Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such ,
that it is only " required" when it can be and should be perfomed. With an SR satisfied, SR 3.0.4 imposes no restriction.
h The use of(" met"or"perfonrhin these instances conveys specified meaffings. A Surveillance is " met" only when the acceptance criteria'are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being " performed," constitutes a Surveillance not " met." "Perfomance" refers only to the requirement to specifically determine the ability to meet the acceptance (continued) nam xw o h s6 s
,.i.3S s =. c. ==::=iS ,
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Frequency 1.4 '
1.4 Frequency DESCRIPTION criteria. SR 3.0.4 restrictions would not apply if both the (continued) following conditions are satisfied:
- a. The Surveillanr.e is not required to be performed; and t
- b. The Surveillance is not required to be met or, even if required to be met, is not known to be failed.
EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these exam Applicability of the LCO (LC0 not shown)isples,MODESthe1, 2, and 3.
EXAMPLE 1,4-1 SURVEILLANCE REQUIREMENTS -
SURVEILLANCE FREQUENCY Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications TS). The Frequency specifies an ir.terval (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)(during which the associated
, Surveillance must be performed at least one time.
Performance of the Surveillance initiates the subsequent infer v*l interval. Although the Frequency is stated as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, an >
5f fCibl extension of the time interval to 1.25 times the (YITill
_4. Frequency is allowed by SR 3.0.2 for operational in the flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the unit is outside the Applicability of the LCO). If the interval specified by ;
SR 3.0.2 is exceeded while the unit is in a H0DE or other specified condition in the Applicability of the LCO, and the performance of the Surveillance is not otherwise modified s
(continued)
OW2/0-ST-L EcfM -V sb I 1.4-27 6. O, 09/23/92-
Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-1 (continued)
(refer to Examples 1.4-3 and 1.4-4), then SR 3.0.3 becomes applicable.
If the interval as specified by SR 3.0.2 is exceeded while the unit is not in a H0DE or other specified condition in the Applicability of the LCO for which performance of the SR is required, the Surveillance must be perfomed within the Frequency requirements of SR 3.0.2 prior to entry into the MODE or other specified condition. Failure to do so would result in a violation of SR 3.0.4.
EXAMPLE 1.4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENC)
Verify flow is within limits. Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after a 25% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter Example 1.4-2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level < 25% RTP to a 25% RTP, the Surveillance must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The use of "once" indicates a single performance will satisfy the specified frequency (assuming no other Frequencies are connected b "AND"). This type of Frequency does not qualify for the extension allowed by SR 3.0.2.
(continued)
I P's/5 m 9edr3 V n.k 1.4-28 Sv 0 7 09/20/92-
Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-2 (continued) j "Thereafter" indicates future perfomances must be established per SR 3.0.2 but only after a specified !
condition is first met (i.e., the "once" performance ~ in this-
~
l example). If reactor power decreases to < 25% RTP, the i measurement of both intervals stops. New intervals start l upon reactor power reaching 25% RTP. l l
EXAMPLE 1.4-3 !
I SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY } ,
i
NOTE-------------- ---
Not required to be perfomed until .'
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after E 25% RTP.
Perfom channel adjustment. 7 days I
The interval continues whether or not the unit operation is ;
< 25% RTP between perfomances. !
1 As the Note modifies the required performance of- the Surveillance, it is construed to be part of the "specified :
' Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />'after i power reaches a 25% RTP to perform the Surveillance. The !
Surveillance is still considered to be within the "specified $
Frequency." Therefore, if the Surveillance wereJot Col performed within the 7 day (-p:n .:2. i,ei- 2 0.0.2X interval, ;
. but operation was < 25% RTP, it would not constitute a failure of the'SR or failure to meet the LCO. Also, no !
violation of SR 3.0.4 occurs when changing MODES, even with - l the 7 day Frequency not met, provided operation does not j exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with power a 25% RTP. - - - .
) !
(plus de extension J ,
med by SR 3.0.2h
~
, l (continued) -
Am/ ST: 9e U 3 4 4.11 1.4-29 k.. G, Gii/Z8fW i
I
Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued,'
Once the unit reaches 25% RTP, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> would be allowed for completing the Surveillance. If the Surveillance were not perfomed within this 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval, there would then be a failure to perform a Surveillance within the specified Q @frequencezi ther w'u!d be rc:tricted V crden : ritFchange SP -3.0 d liind the provisions of SR 3.0.3 would apply. !
EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
.-----------------NOTE------------------ 1 Only required to be met in MODE 1.
Verify leakage rates are within limits. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance.
Therefore, if the Surveillance were not performed within the
- f'y s 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ('Ifht Ncmm the gextension allowed by 'SR 3.0.2)
(A inteTvsT; but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, ne violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency exceeded, provided the MODE change was not made~into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency were not met),
SR 3.0.4 would require satisfying the SR.
E Z 'S STS hc.ff3 -b \ \ 1.4-30 Rn. O, OViW92
e l
b t
ATTACHMENT 2B
'i ITS - PSTS !
h i
COMPARISON DOCUMENT l t
f DISCUSSION OF CHANGES 9
?
I O
i r
i P
1 l
l I
1 1
DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 1 - USE AND APPLICATION
]LRACKETED ADMINISTRATIVE CHOICE B.1 Brackets and optional wording preferences removed to reflect appropriate plant specific requirements.
PLANT SPECIFIC DIFFERENCE P.1 The defined term PHYSICS TESTS was used only in Specifications which were not adopted for this station (LCO 3.10.9,
" Recirculation Loops -
Testing") and the definition is not required. ,
P.2 Consistent with Perry Nuclear Power Plant (PNPP) safety analyses, the appropriate intervals assumed are reflected in i the definition of EOC-RPT SYSTEM RESPONSE TIME.
P.3 References to other Technical Specifications are revised in accordance with plant specific proposed renumbering.
CHANGE / IMPROVEMENT TO NUREG STS C.1 For consistency with other Response Time definitions and the existing TS, the clarification " initiation" is added.
C.2 For consistency with other definitions, e.g. CHANNEL CALIBRATIONS and CHANNEL FUNCTIONAL TEST, the clarification
" required" is added. This is consistent with the OPERABILITY requirement intent for necessary functions.
C.3 The phrase "or have OPERABILITY" is added to this NUREG STS definition to provide a specific relationship between this . term and the definition. The preferred wording remains consistent with the existing Technical Specification (TS) definition.
C.4 These changes are proposed primarily for consistency and to improve the understanding of the explanations. These improvements are also being considered generically by the appropriate vendor Technical Specification owners Groups.
C.S 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, a
ct 's a s ysw C.6 These changes are editorial corrections of typographical or grammatical errors.
PERRY - UNIT 1 1 10/1/93
DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 1 - USE AND APPLICATION l
CHANGE / IMPROVEMENT TO NUREG STS (continued)
C.7 The phrase " cross calibration of the sensing elements" implies '
activities which are not possible on RTDs or thermocouples.
Calibrations typically require adjustments of devices to cause ,
them to conform to a desired output. In this sense, RTDs and thermocouples can not be " calibrated." The appropriate activity to require for an RTD or thermocouple is a comparison of RTD or thermocouple output indications from sensors measuring the same temperature. This activity is precisely that activity described by a CHANNEL CHECK for an. individual ,
sensor. Therefore, this proposed change is intended to provide '
a more appropriate presentation of the intended requirement.
The sentence beginning "Whenever a sensing element- is replaced. . ." also describes an activity that is included in the revised discussion and is, therefore, repetitive and-can be -
deleted. Also, "shall" is replaced with "may" to allow other appropriate mechanisms for " calibration" of these devices, if they are developed.
C.8 Including the terms " interlock" and " display" in the list of attributes to verify OPERABILITY for an analog channel introduces confusion for bistable channels which have an interlock function. Since the list is intended to provide examples of attributes which mu.,t potentially be OPERABLE, dependent on whether it is " required" or not, the list can be l applied to both analog and bistable channels. This revision i will remove any perception of an intended difference. '
C.9 "Other components affecting reactivity" could easily be misinterpreted to include the manipulation of components snich i cause parameter changes that affect reactivity. For example, any activity G at causes a moderator temperature change would then be considered a CORE ALTERATION. The current BWR/6 licenses do not include "other components affecting reactivity." The presumed intent is to incorporate other than the normal reactivity control components, such as control rods.
' However, such additional control components are not addressed in any procedures and their introduction would involve !
procedure revisions in accordance with 10 CFR 50.59.
Therefore, this phrase represents an unnecessary, and ;
potentially confusing, addition to the currently sufficient '
definition of CORE ALTERATIONS, and is deleted.
C.10 The term " equipment" is used in place of " device" for '
consisten;y with the terminology from IEEE-308.
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PERRY - UNIT 1 2 10/1/93
% _. i DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 1 - USE AND APPLICATION i
CHANGE / IMPROVEMENT TO NUREG STS (continued)
C.11 In the example being discussed, the unit is in MODE 1, (e.g.,
2 25% RTP) and no MODE changes would be necessary which could be made or restricted. Therefore, potential confusion is eliminated by deleting this phrase and there is no loss of understanding. !
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- 2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS
'i THERHAL POWER, Low Pressure or Low Flow wi 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow l 1ess than 10% of rated flow, g { APPLICABILITY: OPERATIONAL CONDITIONS I andD ACTION:
'1. 'l With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 735-psig e* e e flow less than 10%
of rated flow,Qat least HOT S__HlJTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1. g --
m f
p34 be ? I.o'1 w +ao rec'nddes jTHERMALPOWER,HighPressureandHighFlow loep opereAlem or h 1.08 Ao'
- sigle retirculat'onloff 4t'AD 11I[ 2.1.2 TheMINIMUMCRITICALPOWERRATIO(MCPR)shallh): le 233
.@with the reactor vessel steam dome pressure (gveater th@785 psig l
'and core flow (fr~ eater than110% of rated flow. (
h hPPLICABILITY: OPERATIONAL CONDITIONS I aM gg" ACTION:
713 VithMCPRhessthan1.07andthereactorvesselsteamdomepressure l gfreater thaib785 osia and core flow.fgteater tha310% of rated flow,@
1 at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)ana comply with the requirements o 5pecincation 6.7.1.
g REACTOR COOLANT SYSTEM PRESSURE S .I 7 I 2.1 Ivess.3elThe reactor coolant system pressure, as measured in the reactor steam dome, shall not exceed 1325 psig.
@ BILITY: OPERATIONAL CONDITIONS 1, 2, 3 and h JACTION:
g ) 7., y -
- 1. 2. With the reacter coolant system preJs yn.)_as measured i ' ctor_
vessel steam dome, above 1325 psig,%e.jn at least H DOWN th fiiactor coo 1~anCsys_ tem pressiire 'lesT than or_ equal _ t'o~1326 ps withi]n>
423/and comply with the requirements of Specification b./.1. \ '
D 1.1.1 >
PERRY - UNIT 1 2-1 Amendment No. 20
- 2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS !
SAFETY LIMITS (Continued)
REACTOR VESSEL WATER LEVEL 1113 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.
h [ APPLICABILITY: OPERATIONALCONDIb dp
' ACTION:
~2. t
[
With the radiated reactor _yg_ illy initiate thssel water level at or below the top of fuelJmanu ,
~ e ECCNrestore the water leveTO i 1.2.1}-Dep'ressurize
,Ithe requirements othe reactor vessel, fSpeEiiication 6.7.1. as necessary for ECCS C operation.j o
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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS
- REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 ff The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1. !'
APPLICABILITY: As shown in Table 3.3.1-1.
- ACTION: MI L S$ 5 hdvdlq %N 12.1-l
[
With a reactor protection system instrumentation setpoint less conservative t than the value shown in the Allowable Values column of Table 2.2.1-1, declare ,
i the channel inoperable and apply the applicable ACTION statement requirement ;
of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value. !
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.s-. I PERRY - UNIT 1 3 nr onk 1- 5
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l ACTIVITIES (Continued) !
- f. The Plant Security Plan and Emergency D' .ng ;
instructions, shall be reviewed at ' snths.
Recomended changes to the Plans .istructions shall be reviewed pursuant tn J. .. Specification 6.5.1.6 and approved by th (# ' rerry Nuclear Power Plant Department
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appropriate. ktf . be obtained as j cQf U .,
t 6.6 REPORTABLE EVENT
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6.6.1 The foll' gfL ce taken for REPORTABLE EVENTS:
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be notified and a report submitted pursuant l
r .nts of Section 50.73 to 10 CFR Part 50, and erABLE EVENT shall be reviewed by the PORC and the results
.: review submitted to the NSRC and the Vice President - Nuclear. :
- 6. 7 SAFETY LIMIT VIOLATION !
t 13 ,
6.7.1 The following actions shall be taken in the event a Safety Limit is Iviolated: h S
't.2.l y. The NRC Operations Center shall be notifi .by telephone. n as possible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> !
u3 --4@Wancflh'elSRC'sNalThe notified within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />shyThe PM
(
- b. A(Safety Limit Violation RefoEsha11'be ' prepared. /The-cerertwT) em <
13- 4-by-the-? ORC.)'T3ir reporAtall doccribe (1 il Qe-reviews ci~rc . tances pyeceding thf viol (n,(2) fects o he viol on pplicab' i
' . - upo d' nit pcponents stems struc es, and correc ve i
j tion 1aren to p ent r rence. p
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- g. The(_SqLety fl the NSRC,,and the Limitlica Violation Pr**Wnt _- Nuclea3RepW411 within 30belsubmitted days of the to the C{
violati
, E 9%g. In. .se,) g g't g il d. Critical operation of the unit shall not be resumed until authorized '
ti by the Comission.
6.8 --~ ~~~**" "wemiemE AND PROGRW4 6.f tec, and ma
[ Md K3 sed w g { h , ,
- a. ine appi. cow.o ....... :
tory Guide 1.33, Revision 2, February 1978. l PERRY- UNIT 1 5 15 4:endment No. 22,;6, 42 l ItJS EE T 2-3 i
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DISCUSSION OF CHANGES CTS: 2 - SAFETY LIMITS ADMINISTRATIVE A.1 Where possible, plant specific management position titles in the proposed Technical Specifications are replaced wit.h generic titles as provided in ANSI /ANS 3.1. Personnel who fulfill these positions are required to meet specific qualifications as detailed in proposed Specification 5.3, and compliance details relating to the plant specific management position titles are identified in licensee controlled documents (such as the USAR) .
The two major specific replacements are the generic " Plant Manager" for the manager level individual responsible for the overall safe operation of the plant and the generic descriptive use of "the corporate executive responsible for overall plant nuclear safety" in place of the Vice President position. The plant specific titles fulfilling the duties of these generic positions will continue to be defined, established, documented and updated in a plant controllec' document with specific regulatory review requirements for changes, such as the Bases,.
USAR or QA Manual. This approach is consistent with the intent of Generic Letter 88-06 which recommended, as a line item improvement, relocation of the corporate and unit organization charts to licensee controlled documents. The intent of the Generic Letter, and of this proposed change, is to reduce the unnecessary burden on NRC and licensee resources being used to process changes due solely to personnel titles changes during reorganizations. Since this change does not eliminate any of the qualifications, responsibilities or requirements for these '
personnel or the positions, the change is considered to be a '
change in presentation only and is therefore administrative.
The use of generic titles will decrease the administrative burden on both the utility and the NRC associated with Technical Specification changes due to reorganizations and title changes which do not affect the functions of these positions. Specific titles are provided in the Bases.
A.2 The technical content of this requirement is being moved to another chapter of the proposed Technical Specifications in accordance with the format of the BWR Standard Technical Specification, NUREG-1434. Any technical changes to this requirement will be addressed with the content of the proposed )
chapter location.
t A.3 The reporting requirements for a special report are replaced with the specific requirements from the regulations which have been promulgated to address the reporting to the NRC of this ;
type of situation. Since the regulations now address this report, the conflicting details of the specification can be revised and/or deleted.
PERRY - UNIT 1 1 10/1/93
1 1
I DISCUSSION OF CHANGES !
CTS: 2 - SAFETY LIMITS ADMINISTRATIVE (continued) '
A.4 Single loop operations are being incorporated into this change request. The amendment submittal for single loop operations for Perry was submitted to the NRC in a letter, PY-CEI/NRR-1353 L dated 6/28/91. The NSHC submitted as part of that change- request letter is still valid and will not be restated here. Since a complete submittal has been sent to the NRC staff on this issue, the change for purposes of this +
revision is considered administrative. ;
RELOCATED SPECIFICATIONS None in this section. j TECHNICAL CHANGE - MORE RESTRICTIVE '
M.1 The Applicability of each of the SLs is extended to all MODES of operation. Although it is physically impossible to violate .
some SLs in some MODES, any SL violation shccid receive the !
same attention and response.
M.2 An additional reporting requirement of notification of the ,
highest level management specifically responsible for the l operation of the plant is added.
M.3 Limits on steam dome pressure and core flow are to be specified !
as 1. The current SLs do not address a pressure or flow which !
is equal to the limit. This proposed change will resolve an i incontinuity between current SL 2.1.1 and SL 2.1.2. l TECHNICAL CHANGE - LESS RESTRICTIVE I
" Generic"
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LA.1 This comment number is not used for this station. l LA.2 The Required Action has been made less specific to allow operator flexibility in determining the best method to restore .
the water level. Directions for the methods to be used for compliance are included in the appropriate response procedures. i The time frame for completion of the action is made consistent ;
with the allowed time to restore other Safety Limit violations.
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PERRY.- UNIT 1 2 10/1/93
DISCUSSION OF CHANGES CTS: 2 - SAFETY LIMITS TECHNICAL CHANGE - LESS RESTRICTIVE (continued) '
LA.3 Details of the content of the required report are relocated to '
the Bases and procedures. The general requirements are !
dictated by 10 CFR 50.73 for content of an LER. Additionally, l changes to the Bases will be controlled by the provisions of ;
the proposed Bases Control Process in Chapter 5 of the Technical Specifications.
LA.4 Details of the time frame for providing this report to utility management are relocated to the Bases and procedures. Chr.nges to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications.
" Specific" L.1 This review is duplicated by the offsite review and audit function. Deleting this review from the onsite review group responsibilities provides additional review time for the remaining functions. Since these are after-the-fact reviews,.
the offsite function provides sufficient, adequate and timely review. This change is consistent with proposed changes in i current Technical Specification Section 6.5. '
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ATTACHMENT 1C :
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COMPARISON DOCUMENT NO SIGNIFICANT HAZARDS CONSIDERATIONS F
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i NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 2 - SAFETY LIMITS t
"L1" CHANGE '
PNPP has evaluated this proposed Technical Specification change and i has determined that it involves no significant hazards '
consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.02. The following evaluation is provided for the three categories of the significant '
hazards consideration standards: 4
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? j i
The proposed change would remove a long term review function :
from the onsite committee that is duplicated by the functions (
of the offsite committee. This will allow additional time for i review of short term plant conditions. These proposed reviews l are not considered as initiators for any previously evaluated i accident and are not required for . the mitigation of any [
evaluated accident. Therefore, the proposed change will not l increase 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.
Therefore it does not create the possibility of a new or i different kind of accident from any accident previously j 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 to provide for adequate and timely review and audit of LERs.
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ATTACHMENT 2 .
ITS - PSTS COMPARISON DOCUMENT i
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ATTACHMENT 2A ,
ITS - PSTS COMPARISON DOCUMENT MARKUP OF ITS i
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SLs 2.0 i
2.0 SAFETY LIMITS (SLs) i 2.1 SLs .
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2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome _ 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 dome pressure e 785 psig and core flow e 10% rated core flow:
MCPR operation shall or be $or1.07 1.08 forootwo oo recirculationrh single operation. recirculation 5 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall befiha4*tekretf)s 1325 psig.
2.2 SL Violations With any SL violation, the following actions shall be completed:
2.2.1 Within I hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72.
2.2.2 Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: !
2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 Insert all insertable control rods. 1 2.2.3 Within 24_ hours, notify the 44ifiHR9dManager.5=:rgr
,.-+ywe-vrenee n: {
_:xie r essa-Lii;Iiind the [offsite reviewees {
OPS , f --sp;d ficd ... pecificetica ';.0.2, "[0ff2.ic]g{c end A dit"3im 1 ,4ke corystke specuCve responikke, }q q,;, Q,4 for overcM phd tmclear safet S3 (continued) )
PGMV-ut]lT1
-BWR/6-$TS- AII dia 2.0-1 M ev. O, 09/28/92-i l
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SLs 2.0 6
2.0 SLs 2.2 SL Violations (continued) 2.2.4 Within 30 days, a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73.
h(%$qi7" theMotM4 *McA2wer4-4pm +TdThe LERa-4F".1Ticcur shall be submitted to the NRC,
~
> 7F and)the
[R=gI@anagerCgTmfPB,and thre-Freracent-Nuc4efe QHmum --
v l'ile corporeAe eyew+lve resfusob e
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@ r ove r,Il plod n ucle<t r cate ht . _
2.2.5 Operation of the unit shall not be resumed until authorized by the NRC.
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l BWR/6 STS 2.0-2 Rev. O, 09/28/92 1
Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs) !
B 2.1.1 Reactor Core SLs ,
BASES
^
BACKGROUND GDC 10 (Ref.1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, nomal operational transients, and anticipated operational occurrence 00s). ,-
Q g b g_
The fuel cladding integrity SL is set suchthatno7 fuel damage is calculated to occur if the T[imit is not violat Because fuel damage is not directly observable, stepback -
approach _ .is_used. to establish an SL, such that the MCPR is not_less than the limit specified in Specification 2.1.1.2.
[foMboth-GenerabE4 ectric-tGEj ~ Cer 4 Nuclear Fuel-(ANF)-Corporation-fuel +poration-and-Adveeed3 rj MCPR greater tnan the r specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from themal stresses, which occur from reactor operation significantly above design conditions.
While fission product migration frosn cladding perforation,is just as measurable as that from use related cracking, the themally caused cladding perforations signal a threshold beyond which still greater themal stresses may cause gross, i rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the i conditions that would produce onset of transition boiling l (i .e. , HCPR = 1.00) . These conditions represent a ugnificant departure from the condition intended by design l for planned operation. The MCPR fuel cladding integrity SL l ensures that during nomal operation and during A00s, at i least 99.9% of the fuel rods in the core do not experience transition boiling.
(continued)
PE R R4 - LU)\T L .
-BWR/fr-SM pd g B 2.0-1
~ Revc--0,-09/28/92-
Reactor Core SLs B 2.1.1 BASES BACKGROUND Operation above the boundary of the nucleate bciling regime (continued) could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat trah %r coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. 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 normal operation and A00s. The reactor core SLs are established to preclude violation of the fuel design O(,9 criterion that an MCPR is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.
.5L The Reactor Protection System setpoints (LCO 3.3'.1.1,
" Reactor Protection System (RPS) Instrumentation"), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERHAL POWER h leve hat would result in reaching the MCP 2.1.1.1 Fuel Claddino Inteority 4Genemi-E4tetMc S'
-Gorocration (CE) ruell g- -
GE critical power correlations are applicable for all critical power calculations at pressures a 785 psig or core flows a 10% of rated flow. For operation at low pressures '
and low 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 (Ref. 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 a 4.53 psi driving head will be
> 28 x 10 lb/hr. Full scale ATLAS test data. i taken at pressures from 14.7 psia to 800 psia (continued)
BWR/6 STS B 2.0-2 Rev. O, 09/28/92
Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.15 Fuel Claddino Intecrity 4 General-E4ec-tMcD -
SAFETY ANALYSES groorat4en (GO ruel) - (continued) indicate that the fuel assembly critical power at this flow is approximately 3.35 My,t. With the c lo design peaking factors, this corresponds to a THERMAL POWER > 50% RTP. Thus, a THERMAL POWER limit of 25% RTP for reactor pressure < 785 psig is conservative.
hl.1.lb Fuel Claddino Intecrity (Advanced Nuclear Fuel CorDoration (ANF) Fuel)
, The u of the XN-3 correlation is valid for critical power calcula 'ons at pressures > 580 psig and bundle mass fluxes
> 0.25 x O' lb/hr-ft2 (Ref. 3) . For operation at low pressures low flows, the fuel cladding inte.grity SL is established a limiting condition on core MiERMAL POWER, with the folio ing basis:
Provided tha the water level in the vessel downcomer is intained above the top of the OM active fuel, na ral circulation is sufficient to ensure a minimum ndle flow for all fuel assemblies that hav a relatively high power and potentially can appr ch a critical heat flux condition. For the AN 9x9 fuel 3design, the minimum bundle flow is > 30 x 10 lb/hr. For the ANF 8x8 fuel design, the m{nimum bundle flow is
> 28 x 10' lb/hr. For all dqsigns, the coolant minimum bundle flow and maxim flow area are such that the mass flux is al s
> 0.25 x 10' lb/hr-ft . Full se e critical power 2
tests taken at pressures down to .7 psia indicate that the fuel assembly cri cal power at 0.25 x 10' lb/hr-ft2 is a: 3.35 Hwt. 25% RTP, a bundle power of 3.35 Mwt corresponds t a bundle ,
radial peaking factor of > 3.0, which is significantly higher than the expected pea ing factor. Thus, a THERMAL POWER limit of 25% TP for reactor pressures < 785 ps'g is conserva'tive. ,
(continued)
BWR/6 STS B 2.0-3 Rev. O, 09/28/92
Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) 2.1.1.2a HCPR -ME-fueML h\
The fuel fuel significant cladding damageintegrity SL issuch is calculated setthat
[to no occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, .
the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur.
Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in '
the procedures used to calculate the critical power result in an uncertainty in the value of tne critical power.
Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.
The MCPR SL is detemined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transitjon is detemined using the approved General Electric Critical C1 gowercorrelations. Details of the fuel claddi'ng integrity SL calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties used in the detemination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.
7;% 1.2b HCPR (ANF Fuel)
The MCPR SL es sufficient conservatism in the operating MCPR limit that, in event of an A00 from the limiting condition of operation, east 99.9% of the fuel rods in the core would be expected to 'd boiling transition. The Ot3) margin between calculated boiling ' tion (i .e. ,
MCPR = 1.00) and the MCPR SL is based etailed statistical procedure that considers the un inties in monitoring the core operating state. One speci uncertainty included in the SL is the uncertainty 1 he.. rent 1
(continued)
BWR/6 STS B 2.0-4 Rev. O, 09/28/92 i
i
Reactor Core SLs B 2.1.1 BASES APPLICABLE 7.sl.l.2b HCPR (ANF Fuel) (continued)
SAFETY ANALYSES N in the'XN-3 critical power correlation. Reference 3 describes e methodology used in determining the MCPR SL.
The XN-3 critica ower correlation is' based on a significant body ractical test data, providing a high degree of assurance at the critical power, as evaluated by thecorrelation,iswitKnasmallpercentageoftheactual Og critical power being estimated. As long as the core pressure and flow are withi he range of validity of the XN-3 correlation, the assumed actor conditions used in defining the SL introduce conse atism into the limit because bounding high radial powe factors and bounding flat local peaking distributions are use to estimate the number of rods in boiling transition. Still urther conservatism is induced by the tendency of the XN-3 rrelation to overpredict the number of rods in boiling transition. These conservatisms and the inherent accuracy of he XN-3 correlation provide a reasonable degree of a rance that there would be no transition boiling in the co during sustained operation at the MCPR SL. If boiling ransition were to occur, there is reason to believe that th integrity of the fuel would not be compromised. Significant st data accumulated by the NRC and private organizations indi ate that the use of a boiling transition limitation to pro ct against cladding failure is a very conservative approach.
Much of the data indicate that BWR fuel can survive for ang extended period of time in an environment of boiling transition.
2.1.1.3 Reactor Vessel Water level During MODES I and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad erforation in the event that the OCll water level becomes of ti,e core height. The reactor vessel water level SL has been established at the top of the less bem 4uo br)5 (continued)
BWR/6 STS B 2.0-5 Rev. O, 09/28/92
Reactor Core SLs i 8 2.8.1 BASES !
APPLICABLE 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 l action. ~
i SAFETY LIMITS The reactor core SLs are established to protect the :
integrity of the fuel clad barrier to the release of !
radioactive materials to the environs. .SL 2.1.1.1 and l SL 2.1.1.2 ensure that the core operates within the fuel '
design criteria. SL 2.1.1.3 ensures that the reactor vessel ;
water level is greater than the top of the active irradiated '
fuel, thus maintaining a coolable geometry. !
l APPLICABILITY SLs 2.1.1.1, 2.1.1.2,and 2.1.1.3 are applicable in all-d t i
SAFETY LIMIT R.2 1
' VIOLATIONS -
If any SL is violated, the NRC Operations Center must be :
, notifi within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in accordance. with 10 CFR 50.72 ,
P4 (Ref.
2.2.2 . .
{
Exceeding an SL may cause fuel damage and create a potential -
for radioactive releases i ess of.10 CFR 100, " Reactor l OPM Site Criteria," limits (Re . Therefore, it is required' ;
to insert all insertable cont I rods and restore compliance' ;
with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time '
ensures that the operators take prompt remedial action.and l also ensures that the probability of an accident occurring - :
during this period is minimal. ;
1 (continued)
. BWR/6 STS B 2.0-6 Rev. O, 09/28/92
.m.. - -
,, , .wc--- -t = +-a+-- - - * + - - - - - - - - - - - - - - - - - - - - -- --
Reactor Core SLs B 2.1.1 BASES SAFETY LIMIT 2.2.3 Af:ela S$eb 2e vQo; C- yyddee [tJSgc ]
VIOLATIONS -
(continued) If any SL is violated, the7ppe~ opriate :erier managamant nO '
4[theauclear-olant-and-the -utilitfshall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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.
Oe Crentw\ Mavgtg Pettg Odckrher Plant Dcprfr.eetkad k Yse hrei.Jenb A)ucleatf 2.2.4 If any SL is violated, a Licensee Event Report sba_ll be prepared and submitted within 30 days to the NR A the senic$ o=
wanagement of the- nuc4e*@nh-anothe ut '1 ty vice-
,, t fres4<fent-fbclear Operat4 ens. This re; 9 eme
- i in -
accordance with 10 CFR_50.73 (Ref. Aq J 4u r Q@io kejW<R635Mgf P4 2.2.5 If any SL is violated, restart of the unit shall not comence 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 10.
- 2. NEDE-24011-P-A,j(latest approved revision).
i XN-NF524{Ah-R ' ion-h-November 4983 -
m 94 O3 r. 10 CFR 50.72. ,G, . ><e ra i Ele e1e;c cCl,,w,l.
G [ 10 CFR 100. / y ;< a l -? R e u t e 10 CFR 50.73. ( F ei 6 E STAR T-[
G l
i BWR/6 STS B 2.0-7 Rev. O, 09/28/92
B 2.0 SAFETY LIMITS (SLs)
B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND hhe SL on reactor steam dome pressure protects th RCS against overpressurization. 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 atmosphere. 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 (RCPB) shall be designed with sufficient margin to ensure that the design ,
conditions are not exceeded anticipated operational during occurrences normal (A00s). op]
During erat-ion-and-.
MODES 1 3 and 2, th
~
eactor vessel wat~ EFT OO~ abov top of thi active ,fyei,evel to provide is requirfd'to e cooling be apa ility. - _ - -
During normal operation and A00s, RCS pressure is limited from exceeding the design pressure by more than 10%, in i 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 core. Any further
- hydrostatic testing with fuel in the core @ done under rm OCS LCO 3.10.1, " Inservice Leak and Hydrostatic Testing M Operation." Following inception of unit operation, RCS ca componentsshallbepressuretestedginaccordancewiththe i requirements of ASME Code,Section XI (Ref. 3).
Overpressurization of the RCS could result in a breach of the RCPB. If this occurred in conjunction with a fuel :
cladding failure, fission p_roducts could enter the :
containment atmosphere, fusing concerm rem +ive iglimits f O(,13 -
- n-raAtoA nivc rcirasen specified in 10 CFR 100,'"kcactor '
Site
.--Criteria" (Ref. 4).(
rejalq k Mbtr c& prokeche \>scriers dHiyed *
. Yo reve ttMear&e telc45tf b B/ccedhuj 08 (continued)
BWR/6 STS B 2.0-8 Rev. O, 09/28/92 i
RCS Pressure SL B 2.1.2 BASES (continued)
APPLICABLE The RCS safety / relief valves and the Reactor Protection SAFETY ANALYSES System Reactor Vessel Steam Dome Pressure-High Function have settings established to ensure that the RCS pressure SL will not be exceeded.
The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to ASME, Boiler and Pressure Vessel Code, h SectionIIIN191EditionfincludingAddendathroughthe Nwinterof1972 Ref. 5), which permits a maximum pressure transient of 110%, 1375 psig, of oesign pressure 1250 psig The SL of 1325_psig, as measured @)'the reactor steam domi 1993 (pressure-mo4catVFe is equivalent to 1375 psig ats'tthe lowe g elevation of the RCS. The RCS istdesigned to ASME Co See+ ion III. I@D Edition.,(Ref. 6), for the reactor ,
I g M J ,, D -
recirculation piping, which permits a maximum pressure b d d W' a n'4 transient of 110% of design pressures of 1250 psig for
' 4c wh+er of IW4 -jsuction piping)(SP 1650 psig for discharge pipinge The RCS -
P pressure SL is selected to be the lowest transient overpressure allowed by the applicable codesu __ P1
&aen 5e pwp a J % d;ulwy va\ve,ad tssa isii hvadhcLaqt va\ve.
SAFETY LIMITS The maximum transient pressure allowable RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 110% of design pressures of 1250Thepsig,for-suction psig for discharge most limitingpipingb@ese
? pipi@ of t allowances is the '
110% of esign pressure; therefore, the SL on maximum allowabl R ressure is established at si 4LesucGyyggvD 1325 -
APPLICABILITY SL2.1.2appliesinallH00ESJ;Awevar. in MODE 5. hacxus the-re a c to r-ve ss eM e a d-clo s u re-b o lt s-a re-n o t-ful l y -
CI (tightened,-it-is-unMkely-the-RCS would-be-pr+ssur4cL4 {
_._ . . _ _t n - ~_ . _ __
SAFETY LIMIT VIOLATIONS 2.2.1 h M rafa m re d d d e tea c fo r N M / w If any SL is violated, the NRC Operations Center must be notified within I hour, in accordance with 10 CFR 50.72 l (Ref. 7).
l (continued) .
BWR/6 STS B 2.0-9 Rev. O, 09/28/92 l
RCS Pressure SL B 2.1.2 BASES SAFETY LIMIT 2.2.2 VIOLATIONS ,
(continued) Exceeding the RCS pressure SL may cause imediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits (Ref.4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time e_nsuretthatJhe operators take prompt remedial / action c ' od aho ewsures 4ka RT^
\ i (b D 2 Y 2 ! " & y " "'$ { 9 ".'L N F"EOb i5 W"'"[*_
223 (den wp~pk mszeQ@
If any SL is violated,' the ffppropriate-senior-management-of-J2
)ftne-nuclear pient-ener thewt444tyfsFall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period provides time for plant -
operators and staff to take the appropriate imediate action and assess the condition of the unit before reporting to the
.'- senior management.
m mm j he Gemc3 tk qu%?ev&srVowes Ya yortney.4pd .}& }lee Predjed u) ear)
~. - __ - /'7,'73 If any SL is violated, a Licensee Event Report shall be _
i prepared and submitted within 30 days to the NRCf.the-senior 9t>- i Iman a g ement-of;thF-enir;14 a r@4 nt , Bd the ett 'ity vicef / ,
bide n t ---- Nuc4 ear-Ope ra Hnn e--4h4 s=cemri rement=k n
'accordance with 10 CFR 50.73 (R_ef. 8)_./(\c. p 8h /
i 6- P s d M a b , bc s ubli&e d L d-w_ e A/SEC, A, ,
?.2.5 ~ ..-
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 normal operation.
i REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.
pc] 2. JSHE, Boiler and Pressure Vessel Code, Section IIg 61: = - /ygg l
(continued)
BWR/6 STS B 2.0-10 Rev. O, 09/28/92
RCS Pressure SL B 2.1.2 BASES REFERENCES 3. ASME, Boiler and Pressure Vessel Code,Section XI, (continued) Article IW 0,
- 4. 10 CFR 100.
- 5. ASME, ler and Pressure essel Code 1971 Edition Addend Twinterof1972 6.
C03 aa,a w u seo+i9 Edition
- 7. SLBoiler vese,c and Pressurkssel Code
- 8. 10 CFR 50.73.
BWR/6 STS B 2.0-11 Rev. O, 09/28/92
ATTACHMENT 2B ITS - PSTS COMPARISON DOCUMENT DISCUSSION OF CHANGES f
i l
i l
t DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 2 - SAFETY LIMITS r BRACKETED ADMINISTRATIVE CHOICES B.1 Brackets removed and optional wording revised to reflect [
appropriate plant specific requirements.
PLANT SPECIFIC DIFFERENCE P.1 As indicated in current Technical Specification Bases, the recirculation system discharge piping design pressure is 1650 psig from the recirculation pump to the discharge valve and 1550 psig befond the discharge valve. Additionally, the piping with the most limiting design pressure, the suction piping, is ;
identified.
i P.2 This comment number is not used for this station.
P.3 The correct addenda for the piping design are providedr' The' currently approved Technical Specifications referenEthis same addenda so no change is being requested in this amendment. .
P.4 Nonapplicable references are deleted. In addition, the sections and references were appropriately renumbered to :
accommodate removal of the non-applicable information and reference (s).
P.5 The fuel vendor used at the plant is only iraportant to choose .
the appropriate discussions. Once tnis is complete, i unnecessary references to the vendor can be deleted to reduce
- administrative burden associated with vendor corporate name .
changes. !
i P.6 Where possible, plant specific management position titles in :
the proposed Technical Specifications are replaced with generic '
titles as provided in ANSI /ANS 3.1. Personnel who fulfill ,
these positions are required to meet specific qualifications as !
detailed in proposed Specification 5.3, and compliance details relating to the plant specific management position titles are identified in licensee controlled documents (such as the USAR). ;
The two major specific replacements are the generic " Plant Manager" for the manager level individual responsible for the '
overall safe operation of the plant and the generic descriptive ,
use of "the corporate executive responsible for overall plant i nuclear safety" in place of the Vice President position. The . ,
plant specific titles fulfilling the duties of these generic i positions will continue to be defined, established, documented and updated in a plant controlled document with specific regulatory review requirements for changes, such as the Bases, !
USAR or QA Manual. This approach is consistent with the intent of Generic Letter 88-06 which recommended, as a line item ;
improvement, relocation of the corporate ad unit organization j PERRY - UNIT 1 1 10/1/93 4
i DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 2 - SAFETY LIMITS BRACKETED ADMINISTRATIVE CHOICES B.1 Brackets removed and optional wording revised to reflect appropriate plant specific requirements.
PLANT SPECIFIC DIFFERENCE :
P.1 As indicated in current Technical Specification Bases, the recirculation system discharge piping design prer te is 1650 psig from the recirculation pump to the dischar> valve and 1550 psig beyond the discharge valve. Additionally, the piping with the most limiting design pressure, the suction piping, is
, identified.
P.2 This comment number is not used for this station.
P.3 The correct addenda for the piping design are provided. The version and addenda listed are the currently correct references for.this facility, and are updated from those in the Bases of the current Specification.
P.4 Nonapplicable references are deleted. In addition, the i sections and references were appropriately renumbered to accommodate removal of the non-applicable information and i reference (s).
P.5 The fuel vendor used at the plant is only important to choose the appropriate discussions. Once this is complete, !'
unnecessary references to the vendor can be deleted to reduce administrative burden associated with vendor corporate name changes. ,
P.6 Where possible,-plant specific management position titles in ,
the proposed Technical Specifications are replaced with generic titles as provided in ANSI /ANS 3.1. Personnel who fulfill !
3 these positions are required to meet specific qualifications as -
detailed in proposed Specification 5.3, and compliance details e relating to the plant specific management position titles are .
identified in licensee controlled documents (such as the USAR). I The two major specific replacements are the generic " Plant i Manager" for the manager level individual responsible for the ;
overall safe operation of the plant and the generic descriptive ;
use of "the corporate executive responsible for overall plant nuclear safety" in place of the Vice President position. The plant specific titles fulfilling the duties of these generic ,
positions will continue to be defined, established, documented !
and updated in a plant controlled document with specific regulatory review requirements for changes, such as the Bases, USAR or QA Manual. This approach is consistent with the intent i of Generic Letter 88-06 which recommended, as a line item i PERRY - UNIT 1 1 10/1/93
DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 2 - SAFETY LIMITS PLANT SPECIFIC DIFFERENCE (continued) improvement, relocation of the corporate and unit organization charts to licensee controlled documents. The intent of the !
Generic Letter, and of this proposed change, is to reduce the unnecessary burden on NRC and licensee resources being used to process changes due solely to personnel titles changes during reorganizations. Since this change does not eliminate any of ,
the qualifications, responsibilities or requirements for these personnel or the positions, the change is considered to be a cnange in presentation only and is therefore administrative.
The use of generic titles will decrease the administrative 3 burden on both the utility and the NRC associated with Technical Specification changes due to reorganizations and title changes which do not affect the funqtions of these positions. Specific titles are provided in the s Bases.
P.7 Titles are included for fuel vendor specific references. !
P P.8 This comment number is not used for this station. ,
P.9 (The reference is made more generic to include other applicable krticles of ASME Section III. nd d -
P.10 This comment number is not used for this station.
y i
P.11 Change includes NSRC in the functions to be notified of any SL P violation. This change was included as a bracketed change in h; l specification, but no brackets exist in Bases. -[,
CHANGE / IMPROVEMENT TO NUREG STS C.1 A typographical correction is identified. The parentheses or !
comma is not needed in the context of the sentence, or !
capitalization is made consistent with other uses.
C.2 The Bases paragraph is reworded to show that only the submittal to the NRC is required by 10 CFR 50.73 and identify the i specific plant and utility management. l C.3 The misplaced Bases statement on water level requirement applicability (a reactor core SL) is removed from the RCS }
Pressure SL discussion. Deleting the statement has no impact ;
since the statement is appropriately included in Bases 2.1.1.3. !
C.4 The Bases discussion of the Safety Limit is revised to match i the SL and clarified. !
i C.5 The verb is revised to reflect that the use of the LCO is I optional. I f
t PERRY - UNIT 1 2 10/1/93 ;
- g. y s-- --
DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 2 - SAFETY LIMITS I CHANGE / IMPROVEMENT TO NUREG STS ;
(continued) '
C.7 An editorial correction is made to provide for presentation that is consistent with the other requirements. +
C.8 The phrasing on Pages B 2.0-1 and B 2.0-4 are editorially !
revised for consistency, and correct grammar.
C.9 The correct reference in this context is to the " limit." I C.10 "Mwt" is revised to "MWt" per the use in the Definitions Section for RATED THERMAL POWER.
C.11 The use of the reduced size font is unreadable. The symbol and f number are editorially revised to use full size text.
C.12 Applicability Bases content is not appropriate for discussion of probabilities of a limit violation. ,
C.13 Preferred wording is incorporated. The proposed text is more f descriptive of the bases for the " concerns" regarding potential
" radioactive releases."
C.14 Enhancement for consistency with 2.2.2 Bases on Page B 2.0-6 I (same Bases for same Specification section).
C.15 A correction is provided for the identified reference.
C.16 The pressure indicator is not the only method of determining the reactor steam dome pressure and is secondary to the actual pressure. The Bases are revised to reflect that the pressure '
is the important parameter and prevent any potential confusion that may occur if the pressure indicator were not available.
[
h T
4
- 4 PERRY - UNIT 1 3 10/1/93 ,
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PERRY -
UNIT 1 :
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SECTION 3.0 :
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ATTACHMENT 1 :
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CTS - PSTS !
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COMPARISON DOCUMENT !
1 1
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1 I
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i 1A: MARKUP OF CTS b
1B: DISCUSSION OF CHANGES i t
1C: NO SIGNIFICANT HAZARDS l
CONSIDERATIONS '
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ATTACHMENT 1 A i t
CTS - PSTS i.
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.L COMPARISON DOCUMENT 1 1
i MARKUP OF CTS i I
If P
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[0/4. 0 AP Pi.I C A<i++
LIMITING CONDITION FOR OPERATION
~
MMW
' -Q [3.0.1 Compliance with the Limiting Conditions for Operation contained in the T.NEM succeeding Specifications is required during the OPERATIOKAL CONDITIONS or other conditions specified therein; except that upon failure to steet the :
i A f Limiting g f Conditions for Operation, the associated ACTION requirements shall be s
i 4
-% __ i M oncompliance with a Specification sha u exist when the requirements of_
L st , not the Limiting Condition for Operation and associated ACTION requirements are met within the specified time intervals. If the Limiting Condition for gg Operation ompletionisofrestored prior the Action to expirationis of requirements notthe specified required P -time intervals,j ;
G. O. 3 VEn~a'Tiiiiiting Condition for Operation is not met, except as providedx in the associated ACTION requirements, within one hour action shall be initiated -A T.ns te to place the unit in an OPERATIONAL CONDITION in which the Specification does ;
not apply by placing it, e applicable, in:
- 1. At least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, 2.
- 3. At least HOT 3HUTD0kN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and At least COLD SHUTD0hH within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Where corrective measures are completed that permit operation under the ACTION L requirements, the ACTION may be taken in accordance with the specified time limits , )
Exceptions to these requirements are stated in the individual Spec I G his Specification is not ap g- in OPERATIONAL CONDITIONS 4 or 5. _
rro!4 E~ntry into an OPERATIONAL CONDITION or oOieTspecified con'dition shall Q not be made when the conditions for the Limiting Conditions for Operation are
\not met and the associated ACTION requires a shutdown if they are not met IrdEce ithin a specified time interval. Entry into an OPERATIONAL CONDITION or -
-other specified condition may be made in accordance with the ACTION require- '
[ments when conformance to them pemits continued operation of the facility for an unlimited period of time.
t This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements. -
Exceptions to these requirements are stated in the individual Specifications.
~
d kOD L c. o log) ,
- <AOD L C- O 306'/
y Q - 4 Aoo Lc o 32 7 > -
J PERRY - UNIT 1 3/4 0-1 Amendment No. 30 !
INSERT 1A LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except '
as provided in LCO 3.0.2 and LCO 3.o.7.
INSERT 1B i LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.
If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time (s), completion of the Required Action (s) is not required, unless otherwise stated.
INSERT 1C 1
LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in: ,
- a. MODE 2 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />;
- b. MODE 3 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and ,
- c. MODE 4 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.
Exceptions to 'this Specification are stated in '
the individual Specifications.
Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by :
LCO 3.0.3 is not required.
l LCO 3.0.3 is applicable in MODES 1, 2, and 3.
A OA I
I INSERT {
PERRY - UNIT 1 3/4 0-1 (1) 10/1/93 i I
i 6
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 associa*:9d ACTIONS to be entered permit continued operatien in the MODE or other specified condition in ti.a ,
Applicability for an unlimited period of time. '
This Specification shall not prevent changes in i MODES or other specified conditions in the Applicability that are required to comply with ACTIONS, or that are part of a shutdown of the r unit.
Exceptions to this Specification are stated in ,
the individual Specifications. These exceptions i allow entry into MODES or other sl?ccified ,
conditions in the Applicability when the j associated . ACTIONS to be entered allt.w unit -
operation in the MODE or other specified condition in the Applicability only for a limited period of time, c
i i
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b i
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l INSERT PERRY - UNIT 1 3/4 0-1 (2) 10/1/93 i
i INSERT 1E [
LCO 3.0.5 Equipment ren.aved from service or declared l inoperable to comply with ACTIONS may be returned to service under administrative control solely to ,
perform testing required .to demonstrate its i OPERABILITY or the OPERABILITY 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 OPERABILITY ,
e LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO 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 accordance with Specification 5.7.16, " Safety function i Determination Program (SFDP)." If a loss of [
safety function is determined to exist by this !
program, the appropriate Conditions and Required ;
Actions of the LCO 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 i Conditions and Required Actions shall be entered '
in accordance with LCO 3.0.2. .
LCO 3.0.7 Special Operations LCOs in Section 3.10 allow specified Technical Specifications (TS) requirements to be changed to permit performance
- of special tests and operations. Unless 7 otherwise specified, all other TS requirements remain unchanged. Compliance with Special ,
Operations LCOs .is optional. When a Special Operations LCO is desired to be met but is not met, the ACTIONS of the Special Operations LCO i shall be met. When a Special Operations LCO is not desired to be met, entry into a MODE or other ,
specified condition in the Applicability shall ;
only be made in accordance with the other :
applicable Specifications.
INSERT PERRY - UNIT.1 3/4 0-1 (3) 10/1/93
h h PLICABI URVEILLANCE REQUIREMENT SCb ga ca /sh - weu o p ooe m
'C. . . L9ettf:=Gdrger4ihshall)be met during theXd;fMIONEQ:;DIII "59 L or othergconditions h(eciTiedMor individuaT%imi&g-Genditiens Joe Operat4g g mirss3otherwise stated inCndi/iduiFSTNdllance Requi ggg t emcat, WeQ
~
k.D~2~E.ach durvmTance Requirement shalf"b'e pirformed within the sp surveillance interval with a maximum allowable ext _ension not to exceed ) 25 Mercent of the_specified surveillance intervalgQ N sea 3 a g[
4.0.3 '
Failure to perform a Surveillance Requirement within the allowed surveil-lance interval, defined by Specification 4.0.2, shall constitute noncompliance ith the OPERABILITLr.equiregrtts. for.._a, Limiting Condition for Operation./ ihe w
/ time' limits 6f the ACTION requirements are applicable at the time it is identified
{ that a Surveillance Requirement has not been performed. The ACTION requirement
. l yg g may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance gc, ; when the allowable outage time limits. of the ACTION re_quirements i are less than b6Q4 hoursJurveillance Requirements do not have to be performed on inope
~
'4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable condi N tion shall not be made unless the Surveillance Sequirement(s) associated with A i the Limiting Condition for Operation have been performed within the applicable surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as requi nd to comply with ACTION e
Jequirements.j6-{ 19 sRv gg
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l O .5 S M 11ance Requirements for inservice inspection and testing of ASME
,j Code lass 1, 2, and 3 components shall be applicable as follows:
To a.
innservice inspection of ASME Code Class 1, 2, and 3 components and i
st% be pe vice testing of ASME Code Class 1, 2, and 3 pumps and- valves'shall g.7' g Pressure rmed in accordance with Section XI of the ASME Boiler and qsel Code and applicable Addenda as required by 10 CFR 50, d Section 50.55h granted by the Cof, except where specific written relief has been .
Ch Il ssion pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).'
- b. Surveillance intervals Pressure Vessel Code and ecified in Section XI of the ASME Boiler and licable Addenda for the inservice inspection and testing activities requi d by the ASME Boiler and Pressure Vessel Code and applicable Addenda sha be applicable as follows in these OA10 Technical Specifications:
ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda terminology for inservice performing inservice inspection and testing activities ins -tion and testing activi s 1
Weekly Monthly At least h per 7 days Quarterly or every 3 months At least once r 31 days -f Semiannually or every 6 months At least once pe 02 days f Every 9 months At least once per 1 days Atleastonceper27(6das Yearly er annually At least once per 366 days V' l PERRY - UNIT 1 3/4 0-2 Amendment No. M, 39 ft f
h INSERT 2A Failure to meet a Surveillance, whether such failure is I experienced during the performance of the Surveillance or ;
between performances of the Surveillance, shall be failure to i meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as !
provided in SR 3.0.3. Surveillances do not have to be ,
performed on inoperable equipment or variables outside
[
specified limits.
INSERT 2B SR 3.0.2 The specified Frequency for each SR is met if the !
Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured Al) from the previous performance or as measured from ;
the time a specified condition of the Frequency '
is met.
For Frequencies specified as "once," the above 8
interval extension does not apply.
If a Completion Time requires periodic performance on a "once per . . .
" basis, the Oil above Frequency extension applies performance after the initial performance.
to each Exceptions to this Specification are stated-in i OOI the individual Specifications.
i INSERT 2C 1
SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency,_ then compliance with the requirement to declare the j LCO not met may be delayed, from the time of I discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of !
the specified Frequency, whichever is lers. This' l delay period is permitted to allow performance of the Surveillance.
If the surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition (s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition (s) must be entered.
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i INSERT PERRY - UNIT 1 3/4 0-2 (1) 10/1/93 I
l INSERT 2D f
SR 3.0.4 Entry into a MODE or other specified condition in I the Applicability of an LCO shall not be made l 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 l are required to comply with ACTIONS or that are part of a shutdown of the unit.
)
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INSERT PERRY - UNIT 1 3/4 0-2 (2) 10/1/93
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APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) I i
% l visions of Specification 4.0.2 are applicable to the above . N(
- c. 7he-pr(sd requir Trequcies for performing inservice inspection and testing activities.
- d. Performance of the above inservic % ection and testing activities ff\.ve shall be in addition to other specified1 eillance Requirements. !
^= e. Nothing in the ASME Boiler and Pressure Vessel Code be construed 5,7 10 to supersede the requirements of any Technical Specificatio . '
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a n .l Sn. II , .
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PERRY - UNIT 1 3/4 0-3 i
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CTS - PSTS i COMPARISON DOCUMENT i
DISCUSSION OF CHANGES l i
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h DISCUSSION OF CHANGES '
CTS: 3/4.0 APPLICABILITY ADMINISTRATIVE A.1 Editorial rewording is made consistent with the BWR Standard Technical Specification, NUREG-1434. During-its development '
certain wording preferences or English language conventions '
were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. '
i The lead-in phrase " Compliance with the...is required" was replaced with "LCOs shall be met during".
OPERATIONAL CONDITIONS was changed to MODES to be consistent with the BWR Standard Technical Specification, NUREG-1434, terminology.
The phrase " ...that upon failure to meet the Limiting '
Conditions for Operation, the associated ACTION requirements shall be met" was changed to "...as provided in LCO 3.0.2 and LCO 3.0.7." LCO 3.0.2 addresses the requirement of meeting the associated A'CTIONS when not meeting a Limiting Condition for Operation. LCO 3. 0.7 addresses another situation when an LCO requirement is allowed not to be met. The requirements remain, albeit in a combination of proposed LCO 3.0.1 and LCO 3. 0. 2. (The addition of the exception to LCO 3.0.7, is discussed below in comment A.6.)
A.2 Editorial rewording is made consistent with the BWR Standard Technical Specification, NUREG-1434. During its development certain wording preferences or English language conventions ,
were adopted which resulted in no technical changes (either '
actual or interpretational) to the Technical Specifications.
The lead-in sentence (" Noncompliance with a Specification ;
shall exist when...") is replaced with "Upon discovery of >
a failure to meet an LCO..." This elimination of the definition of " noncompliance" is administrative in that the Technical Specifications make no use of it. The proposed ,
new first sentence is conceptually relocated from existing LCO 3.0.1 (see comment A.1 above). The addition of the exceptions to LCO 3.0.5 and LCO 3.0.6 are due to their proposed inclusion. Refer to the associated discussion below in comment L.1 and A.5 respectively. ,
15 The phrase " restored" is changed to " met or no longer applicable," "specified time intervals" is changed to
" Completion Time (s " " ACTION requirements" is changed to
" Required Action (),)
s ."
l The phrase "unless otherwise stated" is added consistent with existing exceptions found in a few LCOs. This clarity avoids potential mis-application of requirements.
PERRY - UNIT 1 1 10/1/93 ,
l DISCUSSION OF CHANGES CTS: 3/4.0 APPLICABILITY ADMINISTRATIVE (continued) ,
A.3 Editorial rewording is made consistent with the BWR Standard Technical Specification, NUREG-1434. During its development-certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications. ,
The phrase "except as provided in the associated ACTION requirements" is replaced with "and the associated ACTIONS are not met, an associated ACTION is not provided, or if '
directed by the associated ACTIONS."
o, '
The ACTIO to be taken is changed to "the unit shall be placed in MODE or other specified condition in which the LCO is not applicable" for clarification. The previous wording (" place the unit in an OPERATIONAL CONDITION in '
which the Specification does not apply by placing it") left some confusion as to what to do when the Applicability is an "other specified condition."
The terms STARTUP, HOT SHUTDOWN and COLD SHUTDOWN were replaced with their respective MODE designations in ;
accordance with the MODES Table (proposed Table 1.1-1).
The times to reach each MODE are proposed to include the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> currently allowed by the existing LCO 3.0.3 for '
initiating the shutdown. Additionally, the time represents the total time allowed from the entry into LCO 3.0.3, replacing the existing presentation where each time is referenced as "the next," or "the following," or "the subsequent."
The phrase "under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation" is changed to "in accordance with the LCO or ACTIONS, completionofthe[Cg)(srequiredby LCO 3.0.3 1s not required."
The sentence "This Specification is not applicable in :
OPERATIONAL CONDITION 4 or 5" is changed to "LCO 3.0.3 is :
only applicable in MODES 1, 2, and 3. " This administrative i change is made in conjunction with relocating all existing exceptions to LCO 3.0.3 for Specifications whose .
Applicability is other than MODES 1, 2, or 3, to be encompassed by the proposed LCO 3.0.3.
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. Me)Y - UNIT 1 2 10/1/93
t I
L DISCUSSION OF CHANGES CTS: 3/4.0 APPLICABILITY ADMINISTRATIVE (continued) l A.4 Editorial rewording is made consistent with the BWR Standard ;
Technical Specification, NUREG-1434. During its development I certain wording preferences or English language conventions ,
were adopted which resulted in no technical changes (either i actual or interpretational) to the Technical Specifications. !
The first two sentences of the existing LCO 3.0.4 ,
(" Entry. . .shall not be made. . . . Entry. . . may be made. . . ") I may lead to contradictory interpretations. These two sentences are replaced with a single sentence ("When an LCO is not met, entry into a MODE..."). '
- The existing statement "
... prevent passage through or to ,
OPERATIONAL CONDITIONS as..." is reworded to " ... prevent I changes in MODES or other specified conditions in the Applicability that...". !
An additional sentence is proposed to clarify the intent of the individual exceptions to LCO 3.0.4 ("These exceptions allow entry into MODES or other specified conditions in the i 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."). b A.5 LCO 3.0.6 is added to provide guidance regarding the appropriate ACTIONS to be taken when a single inoperability l (e.g., a support system) also results in the inoperability of one or more related systems (e.g., supported system (s)). In the existing Technical Specifications, alor.g with their intent i and interpretation provided by the NRC over the years, there is I not an unambiguous approach to the combined support / supported inoperability.
Guidance provided in the June 13, 1979 NRC memorandum from Brian K. Grimes (Assistant Director for Engineering and Projects) to Samuel E. Bryan (Assistant Director for Field Coordination) would indicate an intent / interpretation consistent with the proposed- LCO 3.0.6 -
without the necessity of also requiring the additional actions of a Safety Function Determination Program. That is, only the inoperable support system ACTIONS need be taken.
PERRY - UNIT 1 3 10/1/93-
DISCUSSION OF CHANGES CTS: 3/4.0 APPLICABILITY ADMINISTRATIVE (continued)
Guidance provided by the NRC in their April 10, 1980 letter to all Licensees, regarding the definition of OPERABILITY and its impact as a support system on the remainder of the Technical Specifications, would indicate a similar philosophy of not taking ACTIONS for the inoperable supported equipment. However, in this case, additional ACTIONS similar to the proposed Safety Function Determination Program ACTIONS, were addressed and required.
Generic Letter 91-18 and a plain-English reading of the existing STS provide an interpretation that failure to perform a required function, even as a result of a Technical Specification support system inoperability, requires all associated ACTIONS to be taken.
- Certain existing Specifications contain ACTIONS such as
" Declare <the supported system > inoperable and take the ACTIONS of <its Specification >." In many cases the supported system would likely already be considered inoperable. The implication of this presentation is that the ACTIONS of the inoperable supported system would not have been taken without the specific direction to do so.
Considering the history of disagreement and misunderstandings in this area, the BWR Standard Technical Specification, NUREG-1434, was developed with the, Industry input and approval of the
~
NRC to include LCO 3.0.6. Since its function is to clarify existing ambiguities and to maintain actions within the realm of previous interpretations, this new provision is deemed to be administrative in nature.
A.6 LCO 3.0.7 is added to provide guidance regarding the meeting of Special Operations LCOs in proposed Section 3.10. These Special Operations LCOs allow specified Technical Specification requirements to be changed (made applicable in part or whole, or suspended) to permit the performance of special tests or operations which otherwise could not be performed. If the Special Operations LCOs did not exist, many of the special tests and operations necessary to demonstrate select plant performance characteristics, special maintenance activities and special evolutions could not be performed. This Specification eliminates the confusion which would otherwise exist as to which LCOs apply during the performance of a special test or ,
operation. This is consistent with the intent of existing j Special Test Exceptions, however, without this specific j allowance to change the requirements of another LCO, a conflict ;
of requirements could be incorrectly interpreted to exist.
l Therefore, this change provides only administrative clarity. '
l PERRY - UNIT 1 4 10/1/93 l l
i
l DISCUSSION OF CHANGES CTS: 3/4.0 APPLICABILITY ADMINISTRATIVE (continued) j A.7 Reformatting and renumbering requirements is in accordance with the BWR Standard Technical Specifications, NUREG-1434. As a result, the Technical Specifications should be more readily readable, and therefore more understandable, by plant operators as well as other users. During this reformatting and renumbering process, no technical changes (either actual or interpretational) to the Technical Specifications were made unless they were identified and justified.
A.8 Editorial rewording is made consistent with the BWR Standard Technical Specification, NUREG-1434. During its development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational, to the Technical Specifications.
A.9 Editorial rewording is made consistent with the BWR Standard Technical Specification, NUREG-1434. During its development .
certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications.
The first sentence of the insert (" Failure to meet a Surveillance, whether such failure is experienced during the performance of the surveillance or between performances of the Surveillance, shall be failure to meet the LCO.") is proposed to clarify an existing intent that is not explicitly stated.
The concept (editorially rewritten) found in the first sentence of existing Specification 4.0.3, has been moved to the second sentence of the insert (" Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO, except as provided inM 3.0.3. ") -
STL The sentence " Surveillance Requirements do not have to be performed on inoperable equipment" is moved from the last sentence of Specification 4.0.3, to proposed SR 3.0.1 (last sentence of the insert). Since all LCOs do not deal exclusively with equipment OPERABILITY, a clarifying phrase is also added "...or variables outside limits."
A sM. d This presentation of proposed SR 3.0.1 combines the basis issues relating Surveillances to the associated LCO. ,
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PERRY - UNIT 1 5 10/1/93 ,
DISCUSSION OF CHANGES ,
CTS: 3/4.0 APPLICABILITY ADMINISTRATIVE (continued)
A.10 The technical content of this requi.rement is being moved to another chapter of the proposed Technical Specifications. Any technical changes to this requirement will be addressed with the content of the proposed chapter location.
A.11 Editorial rewording is made consistent with the BWR Standard Technical Specification, NUREG-1434, which resulted in no technical changes (either actual or interpretational) to the basic application of the 25% extension to routine surveillances. Also, the sentence " Exceptions to this Specification are stated in the individual Specifications" is added to acknowledge the explicit use of exceptions in various surveillances. These changes provide consistency of wording ,
and clarity for understanding. No technical changes (either actual or interpretational) to the Technical Specifications are intended by these changes.
RELOCATED SPECIFICATIONS None in this section.
TECHNICAL CHANGE - MORE RESTRICTIVE M.1 The sentence "For Frequencies specified as "once," the above interval extension does not apply" is proposed to be added.
This is because the interval extension concept is based on scheduling flexibility for repetitive performances, and these Surveillances are not repetitive in nature and essentially have '
no interval as measured from the previous performance. This precludes the ability to extend these performances, and is therefore an additional restriction. The existing Specification can be seen to allow the extension to apply to all Surveillances.
TECHNICAL CHANGE - LESS RESTRICTIVE ;
i
" Generic" None in this section.
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PERRY - UNIT 1 6 10/1/93
DISCUSSION OF CHANGES ,
CTS: 3/4.0 APPLICABILITY '
TECHNICAL CHANGE - LESS RESTRICTIVE (continued) l
" Specific" L.1 LCO 3.0.5 is added to provide an exception to LCO 3.0.2 for instances where restoration of inoperable equipment to an OPERABLE status could not be performed while continuing to comply with Required Actions. Many Technical Specification ACTIONS require an inoperable component to be removed from service, such as: maintaining an isolation valve closed, disarming a control rod, or tripping an inoperable instrument channel. To allow the performance of Surveillance Requirements to demonstrate the OPERABILITY of the equipment being returned i to service, or to demonstrate the OPERABILITY of other equipment which otherwise could not be performed without returning the equipment to service, an exception to these Required Actions is necessary.
l LCO 3.0.5 is necessary to establish an allowance that, althcogh informally utilized in restoration of inoperable equipment, is ,
not formally recognized in the present Technical Specifications. Without this allowance certain components could not be restored to OPERABLE status and a plant shutdown would ensue. Clearly, it is not the intent or desire that the Technical Specifications preclude the return to service of a .
i suspected OPERABLE component to confirm its OPERABILITY. This allowance is deemed to represent a more stable, safe operation than requiring a plant shutdown to complete the restoration and confirmatory testing.
t L.2 The sentence "If a Completion Time requires periodic performance on a "once per..." basis, the above Frequency extension ahmr4 applies to each performance after the initial performance" is proposed to be added. This provides the ;
consistency in scheduling flexibility for all performances of I periodic requirements, whether they are Surveillances or i Required Actions. The intent remains to perform the activity, '
on the average, once during each specified interval. i l
l PERRY - UNIT 1 7 10/1/93
I DISCUSSION OF CHANGES CTS: 3/4.0 APPLICABILITY l
[
TECHNICAL CHANGE - LESS RESTRICTIVE (continued) l L.3 The sentences in the existing Specification 4.0.3 which are replaced by the insert have two major differences: I
- 1) One change eliminates the requirement to declare the equipment inoperable upon discovery of a Surveillance that has not been performed on time. It is proposed to permit the declaration of the LCO-not-met, to be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The net effect of this portion of :
the change is negligible, since in either case the :
delaying of the ACTIONS for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is the outcome. The rationale for the change is to eliminate -
confusion in applying the correct ACTION time limits at the end of this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, which is also consistent !
with the wording of the BWR Standard Technical -;
Specification, NUREG-1434. ~
- 2) The second change applies this allowance, for the delay in declaring the LCO-not-met, to all ACTIONS instead of ,
only to those ACTIONS whose " allowable outage time ;
limits" (hereafter referred to as Completion Times, ,
consistent with the proposed BWR Standard Technical Specification, NUREG-1434) are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. As discussed in NRC Generic Letter 87-09, it is overly t conservative to assume that systems or components are '
inoperable when a surveillance has not been performed. l The opposite is in fact the case, the vast majority of surveillances demonstrate that systems or components in f act are OPERABLE. When a Surveillance is missed, it is primarily a question of OPERABILITY that has not been verified by the performance of the required )
surveillance. Based on consideration of plant conditions, adequate planning, availability of personnel, the time required to perform the Surveillance and the safety significance of the delay in completing the Surveillance, the NRC has concluded that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is an acceptable time limit for completing a missed Surveillance when the allowable outage times of the ACTIONS are less than the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limit or a shutdown is required to comply with ACTIONS.
PERRY - UNIT 1 8 10/1/93
- .1 l
i DISCUSSION OF CHANGES l CTS: 3/4.0 APPLICABILITY l l
l TECHNICAL CHANGE - LESS RESTRICTIVE l (continued) I Since shorter Completion Times are generally provided for more safety significant Required Actions, if a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay can be safely applied to a Required Action with less than a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time, there should be less of a safety impact when a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay is applied to a Required Action with a greater than 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time. Furthermore, consistent application of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay regardless of Completion Time is .
critical to eliminating potential confusion and !
misapplication. Therefore, this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> deferral should apply to all systems or components, regardless of whether or not their ACTION Completion Time is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less.
Additional clarifying statements are added. The sentences "If the Surveillance is not performed within the delay period. . .",
and "When the Surveillance is performed within the delay period..." are included from the Generic Letter 87-09, LCO 4.0.3 Bases. This clarification will help avoid confusion as to when the Completion Time (s) of the Required Action (s) begin in various situations.
L.4 This comment number is not used for this station.
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PERRY - UNIT 1 9 10/1/93
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ATTACHMENT 1C ;
i CTS - PSTS :
COMPARISON DOCUMENT !
I NO SIGNIFICANT HAZARDS 4 CONSIDERATIONS
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NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3/4.0 - APPLICABILITY "L1" CHANGE PNPP 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 ,i evaluated?
t The addition of Specification LCO 3.0.5 allows restoration of equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ;
ACTIONS. Temporarily returning inoperable equipment to service ~
may in some cases increase the probability of a previously '
evaluated accident. However, the potential impact of temporarily returning the equipment to service is considered to be insignificant since the equipment will have been restored to conditions which are expected to provid'e the required safety :
function. As indicated in Generic Letter 87-09, the vast majority of surveillances do in fact demonstrate that systems or components are OPERABLE. Also, returning the equipment to service will promote timely restoration-of the OPERABILITY of the equipment and reduce the probability of any events that may have been prevented by such OPERABLE equipment. Therefore, the '
change does not involve a significant increase in the probability of an accident previously evaluated.
Since the equipment to be restored is already out of service, the availability of the equipment has been previously considered in the evaluation of consequences of an accident.
Temporarily returning the equipment to service in a state which ;
is expected to function as required to mitigate the i consequences of a previously analyzed accident will promote timely restoration of the OPERABILITY of the equiptnent and restore the capabilitieb of the equipment to mitigate the ;
consequences of any events as previously analyzed. _ Therefore, i the change does not involve a significant increase in the consequences of an accident previously evaluated. '
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l PERRY - UNIT 1 1 10/1/93 1 1
NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3/4.0 - APPLICABILITY "L1" CHANGE (continued)
- 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.
Operation with the inoperable equipment temporarily restored to service is not considered a new mode of operation since existing procedures and administrative controls prevent the restoration of equipment to service until it is considered capable of providing the required safety functions.
Performance of the surveillance is considered to be a confirmatory check of that capability which demonstrates that the equipment is indeed OPERABLE in the majority of the cases.
Short restoration in a condition which is not ultimately confirmed is comparable to such equipment being determined to be inoperable during operation and operation continued in such a condition during an allowed time to complete Required Actions. Since this condition has been previously evaluated in the development of the current technical specifications, the-possibility of a new or different kind of accident from any accident previously evaluated is not created.
- 3. Does this change involve a significant reduction in a margin of safety?
Temporarily returning inoperable equipment to service for the purpose of confirming OPERABILITY places the plant in a condition which has been previously evaluated and determined to be acceptable for short periods. Additionally, the equipment has been determined to be in a condition which provides the previously determined margin of safety. The performance of the surveillance simply confirms the expected result and capability of the equipment. Therefore, the change does not involve a significant reduction in a margin of safety.
PERRY - UNIT 1 2 10/1/93
i NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3/4.0 - APPLICABILITY ;
"L2" CFANGE PNPP 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?
l, 1
The application of the 25% extension to Required Action ;
Completion Times which have a specified frequency on a periodic '
"once per" basis has been determined to not significantly i degrade the reliability that results from performing the i surveillance at a specified frequency. As indicated in Generic Letter 87-09, "the vast majority of surveillances do in fact demonstrate that systems or components are OPERABLE."
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident ;
previously evaluated.
i
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? '
The possibility of a new or different 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.
- 3. Does this change involve a significant reduction in a margin of i safety?
The application of the 25% extension to Required Action Completion Times which have a specified frequency on a periodic l "once per" basis has been determined to not significantly degrade the reliability that results from performing the surveillance at a specified frequency. As indicated in Generic Letter 87-09, "the vast majority of surveillances do in fact demonstrate that systems or components are OPERABLE."
a Therefore, the proposed change does not involve a significant reduction in the margin of safety. -
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PERRY - UNIT 1 3 10/1/93 1
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t NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3/4.0 - APPLICABILITY '
"L3" CHANGE !
PNPP has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards I 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 probabil-ity or consequences of an accident previously evaluated? ,
The assumption of inoperability of equipment upon the t identification that a surveillance has been missed has been determined to be overly conservative because the vast majority '
of surveillances do in fact demonstrate' that systems or components are OPERABLE. Twenty-four hours has been 4
determined as an acceptable time limit for completing a missed surveillance. This time limit prevents unnecessary shutdowns and other changes in plant conditions which may in themselves increase the probability of an operational event. Therefore, this proposed change will not involve a significant increase in the probability of an accident previously evaluated.
The delay in declaring- equipment inoperable upon the identification that a surveillance has been missed does not change the system design or operation and therefore does not directly affect the capability of the equipment to mitigate the consequences of a previously analyzed accident. Therefore, this proposed change will not involve a significant increase in the consequences of an accident previously evaluated, r
- 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 physical modification to r the plant or a change in the mode of operation. Therefore, the possibility of a new or different hind of accident from any accident previously evaluated is not created.
- 3. Does this change involve a significant reduction in a margin of !
safety?
i The allowance of a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time limit to conduct missed surveillances is provided to avoid unnecessary shutdowns and other changes to - plant conditions. Since the majority of r surveillances confirm the OPERABILITY of the equipment, the delay does not reduce the reliability of the equipment, and therefore, the proposed change will not involve a significant '
reduction in a margin of safety.
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l PERRY - UNIT 1 4 10/1/93 !
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i ATTACHMENT 2 :
i ITS - PSTS ;
COMPARISON DOCUMENT i 1
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2A: MARKUP OF ITS 2B: DISCUSSION OF CHANGES :
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ATTACHMENT 2A :1 6
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COMPARISON DOCUMENT i I
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LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LCO 3.0.7.
LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required '
Actions providedofinthe LCOassociated 3.0.6. Conditions shall be met, except as h FJ-(.o 3.oM If the Lc0 is met or is no longer applicable prior to expiration of the specified Completion Time (s), completion of the Required Action (s) is not required, unless otherwise
. or 8 3. rec el b b an,6.ketAcnohn
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LCO 3.0.3 When ee anan LCO isACTION associated not met and is not the associated provided, the unit shall4CTION
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' be ced in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within I hour to place the unit, as applicable, in:
- a. MODE 2 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />;
, b. MODE 3 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and
- c. MODE 4 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.
- Exceptions to this Specification are stated in the individual Specifications. -
Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion '
of the actions required by LCO 3.0.3 is not required. ,
LCO 3.0.3 is lica in MODES 1, 2, and 3.
hLCO 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 (continued)
Ban /6 CERM-lA0171
( 3.0-I WG, 09/28/42-
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LCO Applicability I 3.0 3.0 LCO APPLICABILITY '
hLC0h.4 specified conditions in the Ann 14eebmty that aic .c- '
(continued) to comply with ION oroE6he.i art
%.5 .9. d ,C c sk A J %
Exceptions to t is Specification afE 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.
LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under '
administrative control solely to perform testing required to demonstrate its OPERABILITY.or the OPERABILITY of other >
equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perfom the testing required to demonstrate OPERABILITY.
LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO i ACTIONS are required to be entered. This is an exception to
'- LCO 3.0.2 for the supported system. In this event, 7
SV additional evaluation _s and limitations accordance with Specification Q "
may be required in ,
Determination Program (SFDP)."If g, Safety a loss offunction safety function g
is determined to exist by this program,' the' appropriate .
Conditions and Required Actions of the LCO 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.
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I (continued) ;
t
-BWR/fr5 % 9 o 3 A Ai 3.0-2 Rev. G, 09/28/92 !
~ . .. . = .~
LCO Applicability 3.0 ,
3.0 LCO APPLICABILITY (continued)
LCO 3.0.7 Special Operations LCOs in Section 3.10 allow specified' Technical Specifications (TS) requirements to be changed to permit performance of special tests and operations. Unless h M unchanged.otherwise specified, Compliance all other TS requirements with Special-Operations LCOs is rem optional. When a Special Operations LCO is desired to be met but is not met, the ACTIONS of the Special Operations LCO shall be met. When a Special Operations LCO is not desired to be met, entry into a MODE or other specified i condition in the Applicability shall only be made in accordance with the other applicable Specifications.
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3,0 3 nc.. ?. 69 /?A /92 t
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SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR.
Failure to meet a Surveillance whether such failure is experienced during the performance ,
of the Surveillance or between perfomances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3.
Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1 specified in the Frequency, as meas.25 times the interval ured from the previous performance or as measured from the time a specified condition of the Frequency is met.
For Frequencies specified as "once," the above interval extension does not apply.
If a Completion Time re "once per . . ." basis, quires periodic the above performance Frequency on a extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, u the specified Frequency,p to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> whichever or up to the limit of is less.
period is permitted to allow performance of theThis delay Surveillance.
If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition (s) must be entered. L Cumpletion Thu J the Requ; icd Aet4cns4ecia f = diately upon O(_3 expi=t!Mf ihewishy-ye4d.
(continued) l W R/5 STS 9o U 3- M i 3.0-4 Pr.. 0,4 #28492 l
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i SR Applicability 3.0 3.0 SR APPLICABILITY SR 3.0.3-When the Surveillance is performed within the delay period (continued) and the Surveillance is not met, the LCO must irmnediately be declared not met, and the applicable Condition (s) must be Og entered. The-Ces letion " ::
-imedistely upr.r. @failurs tc ;ast th; Sur/ci"--* aef the R :;;ir;d Actier.5 ,
SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall not be made unless the LCO's Surveillances have been met within their specified ub 3 >wequency.
Fr This provision shall not prevent p:::::;
mg te MODES or other specifji d.4on 07 thr$Ith-";h
- o nequired Ae g s % gg, I w
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i = /5 STS kefIq M ' 3.0-5 new. u, 09/28/92 !
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LCO Applicability B 3.0 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY BASES LCOs LCO 3.0.1 through LCO 3.0.7 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.
LCO 3.0.1 LCO 3.0.1 est' ablishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the unit is in the MODES or other specified conditions of the Applicability statement of each Specification).
LC0 3.0.2 LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that:
- a. Completion of the Required Actions within the specified Completion Times constitutes compliance with a Specification; and
- b. Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.
There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the '
Specification is not applicable. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering (continued) swr /5-;;; PEERV-ltulT 1 B 3.0-1 Rev 0A9R6M2
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LCD Applicability B 3.0 DASES LCO 3.0.2 ACTIONS.) l (continued) '
remedial measures that permit continued operation unit that is not further restricted by the Complet an acceptable level of safety for continued operation.
Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.
The nature of some Required Actions of some Conditions l nece.ssitates Required Actionsthat, musonce the Condition is entered, the e completed eve though the i associated Conditio no longer eris ACTIONS specify the S The individual LCO's equired Actions where this is the case.
An example of this is in LCOf.0.1. "E h m-Gueruind Pressee #j The Completion Times of the Required Actions are also
-pupe rA re applicable when a system or component is removed from service intentionally.
g wh." relying on the ACTIONS includeThe reasons for intentionally but are not limited to, perfomance of Surveillances, p,reventive maintenance, CN problems. maintenance, or investigation of operational corrective Entering ACTIONS for these reasons must be done in a manner that does not compromise safety. Intentional entry into A.CTIONS should not be made for operational convenience.
A "d""d'"* Alternatives that would not result in
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- b'i"' ' " " b ' ' 'h "3 d
Ocs @"D ~0ofng so lWi~trthe time both'"5*d- d subsystems function are inoperable and limits the time other conditions exist which result in LCO 3.0.3 being entered. Individual Specifications may specify a time limit for performing an SR I when testing. equipment is removed from service or bypassed for In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.
When a change in H0DE or other specified condition is required to comply with Required Actions, the unit may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition (s) are entered.
(continued)
BWR/6-STS h%-%\ t B 3.0-2 Rev e Or09RG/W
LCO Applicanility B 3.0 BASES (continued)
LCO 3.0.3 LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and:
a.
An associated Required Action and Completion Time is not met and no other Condition applies; or b.
The condition of the unit is not specifically addressed by the associated ACTIONS.
.This means that no combination of Cc,nditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit. Sometimes possible combinations of Conditions are suc,h that entering LCO 3.0.3 is warranted; in such cases the ACTIONS specifically state a Condition corresp,onding to such combinations finnediately. and also that LCO 3.0.3 be entered This Specification delineates the time limits for placing the unit in a safe MODE or other t,,ecified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its ACTIONS. It is not intended to be used as an operational convenience that pemits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable.
Upon entering LCO 3.0.3, I hour is allowed to prepare for an orderly operation.
shutdown before initiating a change in unit This includes time to pemit the operator to _
coordinate the reduction in electrical generation with the loadelectrical the dispatcher to ensure the stability and availability of grid. The time limits specified to reach lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities of the unit, assuming that only the minimum required equipment is OPERABLE.
This reduces thermal stresses on components of the Rea: tor Loolant System and the potential
, for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LCO 3.0.3 Times.
Completion are consistent with the discussion of Section 1.3, 3
(continued)
BwR1fr-STS- h unM\ B 3.0-3 Re d 09/23792
LCO Applicability B 3.0 t
BASES LCO 3.0.3 (continued) A unit shutdown required in accordance with LCO 3.0.3 may terminated and LCO 3.0.3 exited if any of the following occurs:
- a. The LCO is now met.
- b. i A Condition now exists for which the Required Actions have been perfonned.
c.
ACTIONS Times. exist that do not have expired Completion These Completion Times are applicable from the point in time that the Condition is initially entered and not from the time LCO 3.0.3 is exited.
The time limits of Specification 3.0.3 allow 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> for MODE 1 operation.the unit to be in MODE 4 when a shutdown If the unit is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE applies.
If a lower MODE is reached in less time than allowed, however, the total allowable not time to reach MODE 4, or other applicable MODE, is reduced.
then the time allowed for reaching MODE 3 is the n 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />, because the total time for reaching MODE 3 is not reduced from the allowable limit of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Therefore, if remedial measures are completed that would permit a return to MODE 1 a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.
In MODES 1, 2, and 3, LCO 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 4 and 5 because the unit is already in the most restrictive Condition required by LCO 3.0.3. The requirements of LCO 4
3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, or 3) because the !
ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.
Exceptions to LCO 3.0.3 are provided in instances where i requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the associated condition of the unit.
LCO 3.7.7, " Fuel Pool Water Level."An example of this is in LCO 3.7.7 has an Applicability of "During movement of irradiated fuel :
(continued) awn /C sis- 9 u h D A i B 3.0-4 Rev - 0. 00/IC/9T I
LCO ApplicaDility B 3.0 s BASES ,
0.3
@LC0(tinued)
(con this LCO can be applicable in any or all M[
Therefore, If the LCO :
MODE 1, 2, or 3, there is no safety placing the unit in a shutdown condition. y bi The Required '
assemblies in the associated fuel sto !
appropriate actions of LCORequired 3.0.3. Action to complete in Ifeu o)f the i individual Specifications.These exceptions are addressed in the i
LC_0_3. 0_. 4 g QQ tcosl LCO 3.0.4 establishes limitations on changes in MODE Appliabil;h l'4 $ 1 other specified conditions in the Applicability when ani' is not met.
It precludes placing the unit in a4EfRMe L Agl;c,&lig desired ~
MODE or other specified conditiongwhen the following; i b.h y ,deregg a.
L requirements of an LC0f" +ha "* er JQ ;
~-
Pka e dit W are (w..a eth met i Ae'AM,6~ii4r4 pee;iiCcMU4+0to be. en deM~
b.
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9
%ch 44 ike (2[q gp p ttabW4_ y would result in the unit being reContinued no!
3 @Jin a m er eGei 4eM" ~* quired ^"'"tofh pha i
i MP5 *"h"#g "~" 1 E GC dee; rat oppisto comply with the Required Actions.
[ed 6NeoliabilN3 !
cduireA
' ~
b be <dtred Compliance with Required Actions that pemit continued ,
~~ ~- ~
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MODE or other specified condition provid level of safety for continued operation.
i
, regard change. to the status of the unit beforeThis is without or, after the MODE !
Therefore, in such cases, entry into a MODE or i other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions' The provisions of this Specification should not be- .
practice of restoring systems or compon status before unit startup.
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MODES or other specified conditions in that are required to comply with ACTIONS.
i provisions of LCO 3.0.4 shall not prevent changes in MODES; (continued) twRffrSYS 9%di B 3.0-5 Rew 0, 09f2Sf93 '
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LCO Applicability B 3.0 BASES CO 0.4 (c tinued) or r other specified m
conditions in the Applicability that shutdown.
Exceptions to Lc0 3.0.4 are stated in the individual Specifications.
Exceptions may apply to all the ACTIONS or to a specific Required Action of a Specification.
Surveillances do not have to be perfomed on the associated :
inoperable equipment (or on variables outside the specified limits), as pemitted by SR 3.0.1. Therefore, changing MODES or other specified conditions while in an ACTIONS Condition, either in compliance with LCO 3.0.4, or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be perfomed due to the associated inoperable equipment.
However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.
LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been 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 perfomance of SRs to demonstrate:
- a. The OPERABILITY of the equipment being returned to '
service; or
- b. The OPERAPILITY of other equipment /
~
I 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 perfom any other preventive or corrective maintenance.
An example of demonstrating the OPERABILITY of the equipment being returned to service is reopening a containment 4
(continued)
-BWRM-STS 9e.n , 'EA T B 3.0-6 Rev, -4,--09ftef9f
LCO Appl'icability B 3.0 BASES LC0h0.5 (continued)
Actions, and must be reopened to perfom .
An h ot sys ple of demonstrating the OPERABILITY of uipment is taking an inoperable channel or trip out of the tripped condition to prevent the trip )
function from occurring during the perfomance of an another channel in the other trip system.
of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to permit the logic to function and indicate the appropriate response during the performan an SR on another channel in the same trip system.
LCO 3.0.6 systems that have an LCO specified in the Specifications (TS).
LCO 3.0.2 would require that the Conditions an h-l Actions of the associated inoperable supporte entered system. solely due to the inoperability of the support, Q that are required to ensure the plant is main safe condition Required Actions.are specified in the support systeLCO's These Required Actions may inci Actions or may specify other Required Actions When a support system is inoperable and there is an LCO required to be declared inoperable if detemin However, it is not necessary to enter into .
systems' Conditions and Required Actions unless directed to do so by the support system's Required Actions.The c2 jI potential related toconfusion and inconsistency of requirements systems' L entry into multiple support and supported !
onditions and Required Actions are eliminated by providing all the actions that are necessary to ensure the plant is maintained in a safe condition in the support system's Required Actions.
However, there are instances where a support system's Required Action may either direct a supported system to be
+
)
(continued)
SWR /fr-STS-- C 'q -M i B 3.0-7 Rev. O, 09f28M2 :
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LCO Applicability B 3.0 l BASES h LCO 0.6 _
(cofftinued)
Required Actions for the supported system other Required Action.immediately or after some spe Regardless of whether it is immediate or after some delay, when a support system Required Action directs a supported system to be dec inoperable or directs entry into Conditions and Requir Actions for a supported system, the applicable Conditio and 3.0.2.
LCO Required Actions shall be entered in accordance with Specification ga g g (SFDP), ensures loss of safety function is det appropriate actions are taken.
determine if loss of safety function existsm other limitations, remedial actions, or comp. Additionally, ensatory actions may be identified as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions.
implements the requirements of LCO 3.0.6. The SFDP Cross division checks to identify a loss of safety function for those required. support systems that support safety systems ar The cross division check verifies that the supported systems of the redundant OPERABLE support sy are OPERABLE, thereby ensuring safety
. exists, the appropriate Conditions and Required Actions of the LCOtoinbewhich required the loss of safety function exists are entered.
LCO 3.0.7 be performed at various times over the lif These special tests and operations are necessary to :
perfom special maintenance activities, an special evolutions.
allow specified TS reSpecial Operations LCOs in Section 3.10 perfomances of these'quirements to be changed to pemit special tests and operations, which the requirements of these TS.otherwise could not be pe all the other TS requirements remain unchanged.Unless This will j
j (continued)
M ki S V s'. A ) B 3.0-8 88 W 9/38pJa 1
LCO Applicability B 3.0
. BASES
{
LCO J.7 ensure all appropriate requirements of the MODE or other (co(tinued) specified condition not directly associated with or required to be changed to perform the special test or operation will remain in effect.
The Applicability of a Special Operations LCO represents a condition not necessarily in compliance with the normal requirements of the TS. Compliance with Special Operations LCOs is optional. A special operation may be performed either under the provisions of the appropriate Special Operations LCO or under the other applicable TS requirements. If it is desired to perfom the special operation under the provisions of the Special Operations LCO, the requirements of the Special Operations LCO shall be followed. When a Special Operations LCO requires another LCO to be met, only the requirements of the LCO statement are required to be met regardless of that LCO's Applicability (i.e., should the requirements of this other LCO not be met, the ACTIONS of the Special Operations LCO apply, not the ACTIONS of the other LCO). H r, there are instances where Special Operations L TIONS may C direct the other LC TIONS be met. The S eillances of the other LCO are not required to be met, unless specified in the Special Operations LCO. If conditions exist such that the Applicability of any other LCO is met, all the other LCO's requirements (ACTIONS and SRs) are required to
- be met concurrent with the requirements of the Special Operations LCO.
5 h ,
B 3.0-9 p 9g
SR Applicability ,
B 3.0 B 3.0 l SURVE!LLANCE REQUIREMENT (SR) APPLICABILITY I i
BASES ;
=
i SRs _
SR 3.0.1 through SR 3.0.4 establish the rements- general' i applicable to all unless otherwise stated.Specifications and apply at, all time SR 3.0.1 .-
during the MODES or other specified i i
Applicability for which the requirements of the LCO unless otherwise specified in the individual SRsp y, Specification is to ensure that Surveillancesperfomed This are.
to verify the OPERABILITY of systems and components that variables are within specified limits. F and i
with SR 3.0.2, constitutes a infailure accordanceto m 4
1 associated SRs have been met. Systems and co!
Specification Nothing in this systems or com,ponents are OPERABLE when:ho!j a.
although still meeting the SRs; orThe syste ,
b.
The requirements of the Surveillance (s ,
are known to be not met between required Surveillanc)e perfoman .
Surveillances do not have to be perfomed when the u!
in a MODE or other specified condition for which the l requirements of the unless otherwise specified.associated LCO are not applicable i
The SRs associated with a I
' Operations LCO is used as an allowabl requirements of a Specification. !
i Surveillances, Actions including Surveillances invoked by Req because,the ACTIONS define the reme Surveillances have to be met and performed in accordan .
4 with SR 3.0.2, prior. to returning equipment to OPERABLE status. ^
-h
- 4 I
(continued)
-BWR/6-STS 9tH M ^ \
B 3.0-10 Rev. C, 09/20/92-e v v4- - - - , - - - - _
SR Applicability B 3.0 BASES SR 3.0.1 (continued) Upon completion of maintenance, appropriate post maintena'nce testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2.
Post maintenance testing may not be possible in the current H0DE or other specified conditions in the Applicability due to the necessary unit parameters not having been established.
may be considered OPERABLE provided testing has be satisfactorily completed to the extent possible and the perfoming is equipment itsnot otherwise believed to be incapable of function. This will allow operation to proceed to a MODE or other specified condition where other recessary post maintenance tests can be completed. Some exemples of this process are:
o Cor. trol rod drive maintenance /during refueling that h a. requires scram testing at f 000 B . However, if other appropriate testing is satisfac fly completed Ocg and the scram time testing of SR 3.1'. .
is satisfied. YJ the control rod can be considered OPE BLE. This allows startup to proceed to reach ch a to perform other necessary testingm
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b.- -Gem % (ene 5sM& Cse\% (Rc& #f Wressurs=rar*=NAy-ftfPtty maintenanCg during shutdown that requires system functional tests at a specified pressure. Provided other appropriate testing is satisfactorily completed, startup can (Q
b g' vh>
proceed operation considered OPERABLE. This allows with_H@h the specified to reac pressure to c the necessary post maintenance testing.
SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic perfonr.ance of the Required Action on a "once per..."
interval .
SR 3.0.2 pemits a 25% extension of the interval specified in the Frequency.
This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (continued) bMK/O aw iS- bi B 3.0-11 Rev A , 09/E0f92
SR Applicability B 3.0 BASES SR 3.0.2 (continued) (e.g., transient conditions or other ongoing Surveillance or maintenance activities).
The 25% extension does not significantly degrade the reliability that results from perfonning the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of confonnance with the SRs.
The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply. These exceptions are stated in the individual Specifications. An example of where SR 3.0.2 does not apply is a Surveillance with a Frequency of "in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions." The requirements of regulations take precedence over the TS. The TS cannot in and of themselves extend a test interval specified in the regulations.
Therefore, there is a Note in the Frequency stating, "SR 3.0.2 is not applicable."
As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic requires perfonnance on a "once per..Consletion
." dasis. TheTime 25%
that extension ap perfo mance. plies to each performance after the initial The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25%
extension to this Completion Time is that such an action ;
usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.
The provisions of SR 3.0.2 are not intended to be used !
repeatedly merely as an operational convenience to extend !
Surveillance intervals r periodic _Compittion Time intervals beyond those specified.
M gg g g ggg w Ak etL % whava s 3 /
~ _. _-w _
_s SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring :
affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not !
l been completed within the specified frequency. A delay '
s (continued)
-BWR/6-ST& h a f yD I B 3.0-12 Rev. O, 09/28f92
SR Applicability B 3.0 j BASES C.~3 or up k. N I.d A of Ot. Af odi t. '
p g g , aig e.vu bM w -s 3 j SR 3.0.3 (continued) period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> applies from the point in time that it is discovered that the Surveillance has not been perfomed in accordance with SR 3.0.2, and not at the time that the specified frequency was not met. ;
provides been missed.adequate time to complete Surveillances that h ;
This delay period pemits the completion of a Surveillance before complying with Required Actions or other !
remedial measures that might preclude completion of the Surveillance. l.
i The basis for this delay period includes consideration of unit conditions, adequate planning, availability of the safety significance of the delay in comple 1 i
required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.
When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions or operational situations, is discovered not to have been perfonned when specified, to perfom theSR Surveillance.
3.0.3 allows the full delay period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i
SR 3.0.3 also provides a time limit for completion of Surveillances that become applicable as a consequence of ,
MODE changes imposed by Required Actions.
Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence.
Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals.
period, then the equipment is considered ino variable then is considered outside the specified limits and the Completion Times of the Required Actions for the applicable of the delayLCO Conditions begin imediately upon expiration period.
delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCO (continued)
-BWR/6-STS u n , D E.\ l B 3.0-13 4ev. Or09/28/92
SR Applicabili%y B 3.0 BASES SR 3.0.3 Conditions be (continued) Surveillance. gin immediately upon the failure of the i
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.
SR 3.0.4 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 OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the Applicability for which these s stems and components ensure Ogg
- f (changes'in App (icabi glit MODES or otheLr specified conditi
'TJJSERX (i- y_4tartup. J -
as'sociated'with unit shutdoMs we'11sas1
. s DMh
- The provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. } xmEET The precise requirements for performance of SRs are specified such that exceptions ta 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 performanc,e of Surveillances when the prerequisite condition (s) s;;ecified 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" until the specific conditions needed are met. Alternately, the Surveillance may be stated in the fom of a Note as not required condition,(to be met or performed) until a particular event, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, frequency.
ewR # - W S- c. f y N A I B 3.0-14 Rev_ 0,494W92 l
INSERT B14A l
However, in certain circumstances failing to meet an SR will not result in SR 3.0.4 restricting a MODE change or other specified l condition change. When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed (per SR 3.0.1 "Surveillances do not have to be performed on inoperable equipment"). 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 i 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.
INSERT B14B 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 unit shutdown.
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PERRY - UNIT 1 B 3.0-14 j
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ATTACHMENT 2B .
F ITS - PSTS !
COMPARISON DOCUMENT ;
DISCUSSION OF CHANGES I
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i DISCUSSION OF CHANGES TO NUREG-1434 SECTION 3.0 - APPLICABILITY BRACKETED ADMINISTRATIVE CHOICE None in this section.
P_LANT SPECIFIC DIFFERENCE P.1 References to other Technical Specifications are revised in accordance with plant specific proposed renumbering.
t CHANGE / IMPROVEMENT TO NUREG STS C.1 LCO 3.0.5 also contains an exception to LCO 3.0.2. For ,
completeness it should also be referenced here.
C.2 Typographical correction is identified. Missing spaces are added, words out of place (apparently inadvertently included during word processing) are removed, and grammatical corrections are m~ade as necessary. .
C.3 This change avoids the error introduced if the particular Required Action has a stated alternate time of beginning.
0.4 Editorial rewording to make consistent with the same application found in LCO 3.0.4.
C.5 These changes are proposed to revise specific terminology to .
that which is generically preferred for. application to the BWR/6 plants. ,
C.6 "Or if directed by the associated ACTIONS," is provided to include this particular entry condition for LCO 3.0.3. The I two original entry conditions for the LCO are not the only ;
times when it is appropriate. '
, C.7 Since no other 3.0 Specification includes direction on MODE applicability, this statement could be read to imply something unique to LCO 3.0.3. As worded, it could inappropriately imply MODE 1, 2, and 3 applicability is unique to LCO 3.0.3. As proposed to be modified-(add {
"only"), the implication would be that LCO 3.0.3 is unique in 1 that other MODES are not included - which is the appropriate l interpretation. 3 i PERRY - UNIT 1 1 10/1/93 ;
i
DISCUSSION OF CHANGES TO NUREG-1434 !
SECTION 3.0 - APPLICABILITY l CHANGE / IMPROVEMENT TO NUREG STS (continued) :
C.8 The NUREG STS Bases state, as fact, that LCO 3.0.4 does not !
impose restrictions on normal or forced shutdowns. Generic ,
Letter 87-09 proposed Bases state that the LCO 3.0.4 restrictions are only to preclude entry into " higher" HODES of operation. This understanding and intent is not clearly ,
interpreted from the Specification wording. The proposed j change clarifies the intent, consistent with the Bases and past NRC guidance. Additionally, for consistency of application this phrase is also incorporated in SR 3.0.4.
This will preclude any potential conflict in restricting MODE changes during a plant shutdown. !
C.9 The standard use of HPCS as an example is not applicable for a BWR/6 design. However, the RCIC System is appropriate for discussion in the identified condition.
C.10 Some ACTIONS would preclude " continued operation in the MODE" but may not require completely exiting the applicability of the LCO (especially in instances where the LCO applicability includes all MODES). It is the intent of LCO 3.0.4 to ,
preclude entry into the MODE when the ACTIONS will not allow continued operation in that MODE.
C.11 The referenced pressure is not appropriate for all BWRs. The appropriate normal operating pressure for BWR/6s is proposed.
C.12 Consistent with the guidance of Generic Letter 89-14 which '
deleted the 3.25 limit and Generic Letter 91-04, the Bases are clarified to indicate the intent for application to refueling surveillances.
C.13 The relation between the SR 3.0.1 statement that 3 "Surveillances do not have to be performed on inoperable kEd' equipment," and the SR 3.0.4 restriction on MODE changes Jd",ME Ms are ot met, is clarified.
C.14 No exception to LCO 3.0.2 is found in LCO 3.8.1. An appropriate reference is now presented.
.a% o Osah 'C k E Y*
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i PERRY - UNIT 1 2 10/1/93 ,
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PERR -
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~ r CHAPTER 4 .
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i ATTACHMENT 1 CTS - PSTS COMPARISON DOCUMENT u
1 A: MARKUP OF CTS ;
i 1B: DISCUSSION OF CHANGES i 1C: NO' SIGNIFICANT HAZARDS CONSIDERATIONS j i
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i ATTACHMENT 1 A !
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P CTS - PSTS '
COMPARISON DOCUMENT t
MARKUP OF CTS i i
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M.o M DESIGN FEATURES
~
@ '5 -1iT 5 od (das M N, o Sw dr[e
- w. V h 11 9EEEN 2Eh325SPIEI,53.,.f[^
Buvnumu ..mvm- m._
- LIGIO MUCNIC
~
Figuy 5.1.1-1 shows the PNPP site area, includ'ing the meteorological tower.
The excl(sion area boundary is 2900 feet from the center line of the reactor.
All land wit.hin the exclusion area is jointly owned by the CAPCO Group
- Companies.
for oil, gas,;and salt.'CEI controls the exclusion area; controls include mineral p righ In addition, the U.S. Coast Guard provides control over the Lake Erh-portion of the exclusion area. A railroad spur serves the plant, heading in an'Nask-north easterly direction from the railroad company k right-of-way to the plant site &CEI owns the tracks and only railroad cars consigned to the PNPP are brought ont6-the site over this spur. h i
M Figure 5.1.1-1 also shows the liquid and gaselbs,. effluent discharge locations !
as well as the plant SITE BOUNDARY for gaseous releas_es and the UNRESTRICTED '
AREA for liquid effluent releases.
The dose rate andNo.ses due to radioactive ,
materials released in gaseous effluents from the site to areas at and beyond s the SITE B0UNDARY v3.11.2.2, and 3.11.2.3.shall be limited pursuant to Specificationhql.2.1, All gaseous effluent releases at PNPP are<onsidered to be ground-level releases.
i The concentrations of radioactive materials released in liquid effluents to UNRESTRICTED AREAS shall be limited purs'uant Q 'o Specificationl 11.1.1. y i
yh/ pdm 6MfdA 264E([pQ
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- p 0 M :n : rphcli k c: :h:r '-"gure5. D
/x5.2NONTAINMENT
~~ - 'N I
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CONFIGURATION 's. . . _
~ _ . . -
s 5.2.1 The primar i cylinder and an el(lt soidal dome. containment is a steel structure composed of a vertical rig Inside and at the bottom of the primary con- i
' tainment is a reinforc concrete drywell composed of a vertical right cylinder i
and a steel head which pression pool connected to c tains an approximately 18'3" deep water filled sup-e drywell through a series of horizontal vents. <
The primary containment has a inimum net free air volume of 1,160,000 cubic feet. The drywell has a minim et free air volume of 276,500 cubic feet. I DESIGNTEMPERATUREANDPRESSURE 5.2.2 The containment and drywell are designe nd shall be maintained for:
\
- a. Maximum internal pressure:
( 1. Drywell 30 psig.
N 2. Containment 15 psig.
K__. ~
- - + ...
PERRY - UNIT 1 5-1
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INSERT 1A I 4.1.1 Site and Exclusion Area Boundaries- '
The site boundary shall include the area shown in-Figure 4.1-1 and the exclusion area boundary ~shall '
have a radius of 2900 feet from the centerline of I the reactor. lt t
4.1.2 Low Population Zone (LPZ) ;
The LPZ shall have a radius of 2.5 miles from the l centerline of the reactor, l 5
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'tJePr SITE BOUNDARY TOR GACEOUS EFFL'.'E"T.W
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.J.. A E PERRY - UNIT 1 5-2
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POWER PLAN ,
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_ DESIGN FEATURES i
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DESIGN TEMPERATURE AND PRESSURE (Continued 5 N
~
N
- b. MaxYmom-internaltemperature: \
.I 1. Drywell'330"F.
2.
N Suppression'poql 185 F.
f
\ t
/ c. Maximum external to internal differential pressure:\
- 1. Drywell 21 psid. N
/
/
- 2. Containment 0.8 psid. -
N f ,' N SECONDARY CONTAINMENT ,,N N s
\
5.2.3 The secondary containment consists of the annulus b$ tween the shield k building and the primary containment and has a minimum free voluge,of 392,548 ,
3 cubic feet. - . .
.. _ . _ _ ~ . . . . . . .
g}~-~kACTS coqrb 4 2 Ps u d or Co h- edi,\
. . - - .. f --.J A M z , rt L, ..
blNThe reactor shall contain 48 fuel assemblies. Each assembly shall con of>yist of aenriched slightly matrix ofuranium Zircaloy? clad fuel dioxide (UOJ rodsas with an initial compos 4 tion' fuel materi Fuel assemoii miteo to those fuel ~ designs approved by the NR taff for use in >
~
7D45FJLt 4M h O hbh $h. . $5Y ,
LY
~~~ fEsch h.UE.O.2Nhe
~ d_ i~ embl;c.lball consist reactor core of e- cruci form-aFrcyst The control material sh~5T1 be boron carbide powdcc (5' C shall nca contain j nium 177 contr metalpnd he,c c raie:L2hsc+ber 'cgh of M2.7 'ade) *_
- 5. 4 REACTOR COOLANT S
~~~~
DESIGN PRESSURE AND TEMPERATURE ~~ '
5.4.1 The reactor coolant system is designed and shall be maintained:
m N
' ) a.
ib In accordance with the code requirements specified in Section 5.2 of the FSAR. with allowance for normal degra' dation pursuant to the applicable Surveillance Requirements, ,
\ b. For a pressure of:
- 1. 1250 psig on the suction side of the recirculation pump. /
L~___ - -
- ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ '
~ ~ ~ ~ ~
PERRY - UNIT 1 5-4 Amendment No. 33,40
f i
1 i
l INSERT 4A Limited substitutions of zirconium alloy clad or stainless steel filler rods for fuel rods, in accordance with approved :
applications of fuel rod configurations, may be-used. Fuel l assemblies shall be limited to those fuel designs that have :
been analyzed with applicable NRC staff' approved codes and- I methods and shown by_ tents or analyses to . comply with all i safety design _ bases._f A limited number of lead test assemblies !
that have not completed representative testing may be placed -'
Ll in nonlimiting core regions.
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INSERT '!
PERRY - UNIT 1 5-4 10/1/93 ;
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1 1
l DESIGN FEATURES
~~
N
~
DESIGN PRESSURE AND TEMPERATURE (Continue'd)~
- 2. 1650 psig from the recirculation pump clischarge to the outlet side.of the discharge block valve.
-s }
! )
- 3. 1550 psig'from the discharge block valve to the jet pumps.
Lu - c. For a temperature of 575?(.
VOLUME 5.4.2 The total water and steam volume of t6s' reactor vessel and recirculation system is approximately 19,000 cubic feet at a niiminal steam dome saturation f temperature of 549 F.
x /
' x
- 5. h 8ETEOROLOGICAL TOWER LOCATION N }j g5.1 eorological tower shal1 be tedas_shownonfigh .1-1/
[5-6-) Fdd SYORAGd_ - . . -
r~
Qil,\)C/ITICALIT/
~u o
@li 54 The spent fuel storage racks are designed and shal
-- maintained with:
b tw / ip Oh
- a. A k,ff(agupdent to. 'ers thger cquci ts'0.95g ooded with A3 unborated water, Mc-kdir al' cc?cPti=c! uncertainties and-biases- !
as.. described in Section .
9.l. t of the us ML k a ucL W ill%g Q\,g114.c25 A nomin -wc4 center-to-center dis,tance-ba%an m%i& a s'ics
' h 'plac^i Sthe Qv s4.ra y sem$=JJ.aS Q ,*AK.~.m _ storage racksjILthe 1 Handlin Fue1_Mbo.g
-= San.1 G .r6,espatcy Ah
<c . b nomicili1Ti9 ccatcHTnter-distance-fromvark Lu .ock estria 4Le % t
. minimm.4ue4-+tecage<e!! cpaci9 c' ' inches-can+er-te-center un hin oM y i>
Qrack-in.-the-upper.-containment-pool. @N1 Erd 5 A] __
, _ . _ . . _ ~ . . . . .
T QLh(, xduring ~7,he storageCONDITIONS OPERATIONAL of spent fuel I and 2. in the upper containment pool is prohibited gg[yQ ImErlT 5]h b Sr6,#'The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevatio 39t'f/.
fPACIT/
% % d- -
g The spent fuel storage pool is designed and shall be maintained with a torage__ capacity ifm" d to no more than 4020 fuel assemblies.
-%0f)OMPONENT
/5. 7C CYCLICW CTF CJ,.M8 _
R TRANSIENT LIH N JL, G g 5. 7.1 The components identified in Table 5.7.1-1 are designed and shall be )
t maintained within the cyclic or transient limits of Table 5.7.1-1.
(x,f,9 3
M PERRY - UNIT 1 5-5 l
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.l INSERT SA '
- b. A nominal fuel assembly center to center ,
storage spacing of 7 inches within rows and '
12 inches between rows in the storage racks ;
in the upper containment pool; and i INSERT SB 4.3.1.2 The new fuel storage racks are designed and shall be maintained with: i
- a. k eff s 0.95 if fully flooded with unborated water, which includes an allowance for i uncertainties as described in Section 9.1.1 i of the USARy and
- b. A nominal 7 inch center to center distance I between fuel assemblies placed 1.: etorage '
racks. i INSERT SC I
4.3.3.2 No more than 190 fuel assemblies may be stored in 4 the upper containment pool. l
-t 0 $
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i 4
INSERT PERRY - UNIT 1 5-5 10/1/93 {
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_, . . - :P h
, - ~
TABLE 5.7.1-1
' 'y ~N E / 'N x COMPONENT CYCLIC OR TRANSIENT LIMITS n
y COMPONENT
~~
LIMIT DESIGN CYCLE
(
Reactor 120 heatup and cooldown cyci b s N.N 70'F to 560*F to 70'F 80 step change cycles ~
dMwdO Loss of feedwater heaters ,
180 reactor trip cycles b ' 100% to 0%'of4ATED THERMAL POWER 3 6Cn 3 e 40 hydrostatic pressure or leak tests Pressurized to > pTig.Jind /
g < 1250 psig
) _ - _ - - -
- - - ~ ~ - - - - ~ ~~
1/
a
. __mm_
_ . _ _ . _ . _ _ _____-_-.__m..um._mma___2m_.-_w_.mi_m.u_m_m_u-____m _._2_ ._ m_ _ _c _ . _ - - -
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ATTACHMENT 1B !
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.i CTS - PSTS i i
COMPARISON DOCUMENT l i
DISCUSSION OF CHANGES- -l i
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DISCUSSION OF CHANGES CTS: 5 - DESIGN FEATURES
{
bDMINISTRATIVL i A.1 Reformatting and renumbering requirements is in accordance with
- the BWR Standard Tecnnical Specifications, NUREG-1434. As a i result, the Technical Specifications should be more readily readable, and therefore understandable, by plant operators as !
well as other users. During this reformatting and renumbering process, no technical changes (either actual or interpretation-al) to the Technical Specifications were made unless they were identified and justified.
A.2 The figures showing the exclusion area boundary and the low !
population zone are replaced with specific descriptions of i these boundaries. Since these descriptions and revised figures continue to provide the information pertinent to 10 CFR 100 requirements, these revisions are presentation preferences only '
and are considered to be admir.istrative changes. i A.3 Editorial rewording is made consistent with the BWR Standard Technical Specification, NUREG-1434. During its development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the Technical Specifications.
A.4 This comment number is not used for this station.
A.5 This comment number is not used for this station. ,
A.6 Usage of the terms "and/or" has been changed to "and". The BWR Standard Technical Specification, NUREG-1434, Writer's Guide 4 recommends the use of "and/or" be avoided. The intent of the 'I definitions is not changed. i A.7 The technical content of this requirement is being moved to i another chapter of the proposed Technical Specifications in accordance with the format of the BWR Standard Technical Specification, NUREG-1434. Any technical changes to this ;
requirement will be addressed with the content of the proposed ;
chapter location, j l
1 PELOCATED SPECIFICATIONS None in this Lection.
PERRY - UNIT 1 1 10/1/93 ]
'i DISCUSSION OF CHANGES l CTS: 5 - DESIGN-FEATURES '
TECHNICAL CIIANGE - MORE RESTRICTIVE !
M.1 The existing Tecnnical Specifications do not contain these
['
limitations on fuel storage in the fuel storage racks. The addition of this Specification imposes restrictions which will require a formal licente amendment request and approval to !
modify. >
TECHNICAL CIIANGE - LESS RESTRICTIVE .
" Generic" t
LA.1 The specific boundary for the Unrestricted Area remains detailed in the USAR Section 2.1.1.3. The requirements for and q restrictions on locating the Unrestricted Area must conform to t regulations found in 10 CFR Part 20. If this feature of the facility were altered in accordance with 10 CFR Part 50.59 and 10 CFR Part 20, there would not be a significant impact on safety (which is the criteria of 10 CFR 50. 3 6 (c) (4 ) for- ,
including as a Design Feature). Therefore, allowing the removal of the figure, showing the location of this boundary from the Technical Specifications l will not impact safe ;
operation of the facility. !
l
.LA.2 Configurations, design temperatures and pressures, and volumes >
of Primary Containment and Dryuell, Secondary Containment, and the Reactor Coolant System remain detailed in the USAR Sections ;
6.2.1, 6.2.3, and 5.1, respectively. Any changes ~ to these l design parameters must conform to the requirements of 10 CFR !
50.59. Furthermore, sufficient detail relating to these features exists in LCOs to ensure any changes which may impact safety would require prior NRC review and approval. Since the r features with a potential to impact safcty are s tfficiently addressed by LCOs, and other features, if altered in accordance with 10 CFR 50.59, would not result in a significant impact on r safety, the criteria of 10 CFR 50.36(c) (4) for including as a i Design Feature are not met. Therefore, allowing the removal of :
these details from Technical Specifications, with their ;
discussion'in the USAR, will not impact safe operation of the )
facility. 1 LA.3 These figures are revised to eliminate details which are not required by the Specification, and are not necessary to show :
features that, if altered, would have a significant impact on i safety (which is the criteria of 10 CFR 50. 3 6 (c) (4 ) for including as a Design Feature). .
PERRY - UNIT 1 2 10/1/93
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DISCUSSION OF CHANGES CTS: 5 - DESIGN FEATURES TECHNICAL CHANGE - LESS RESTRICTIVE '
" Generic"(continued)
LA.4 This comment number is not used for this station.
t LA.5 The construction of the tubes containing the boron carbide and ;
the nominal length containing the boron carbide remain detailed ,
in USAR Section 4.2.2 and Figure 4.2-1. Any changes to these design parameters must conform to the requirements of 10 CFR ,
50.59. Furthermore, the LCOs which address safety parameters '
which would be impacted by such changes ensure any changes 1 would require prior NRC review and approval. Since the features {
with a potential to impact safety are sufficiently addressed by }
LCOs, and other features, if altered in accordance with 10 CFR i 50.59, would not result in a significant impact on safety, the {
criteria of 10 CFR 50. 36 (c) (4 ) for including as a Design j Feature are not met. Therefore, allowing the removal of these ,
details from Technical Specifications will not impact safe operation of the facility.
LA.6 Administrative limitations are relocated to the USAR or plant I procedures. This limitation is not related to materials of '
construction or geometric arrangements, that if altered or ,
modified, would have a significant impact on. safety.
Therefore, the requirement is removed from the Design Features section.
LA.7 This comment number is not used for this station.
LA.8 This comment number is not used for this station. ,
t LA.9 This comment number is not used for this station.
I
- " Specific" L.1 This allowance provides a recognition of a specific kind of special test that may be performed. This is intended to avoid potential confusion regarding whether a Technical Specification l change is required to conduct this test. The requirements of 10 CFR 50.59 regarding conducting special tests remain applicable, and are sufficient to ensure that a limited number of lead test assemblies placed in non-limiting core regions, l will not have a significant impact on safety (which is the criteria of 10 CFR 50. 3 6 (c) (4 ) for including as a Design Feature). This change is in conformance with Supplement 1 of ,
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PERRY - UNIT 1 3 10/1/93
ATTACHMENT 1C CTS - PSTS COMPARISON DOCUMENT T
NO SIGNIFICANT HAZARDS CONSIDERATIONS i
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t NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 5 - DESIGN FEATURES "11" CHANGE PHPP 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?
Fuel assemblics are not considered as accident initiators for ,
any previously analyzed accident and therefore, the addition of ,
a design feature requirement that allows a limited number of i lead test assemblies to be place in nonlimiting core regions does not involve a significant increase in the probability of an accident previously evaluated.
Since the revised requirement will only allow the lead test l assemblics to be placed in nonlimiting core regions and i previously analyzed accidents do not result in fuel failures in i these regions, the change does not involve a significant ,
increase in the consequences of- an accident- previously evaluated.
i i
- 2. Does the change create the possibility of a new or different i kind of accident from any accident previously evaluated?
.i The proposed change introduces no new mode of plant operation and it does not involve physical modification to the plant.
Operation with fuel assemblies that have not -completed representative testing does not create the possibility of a new or different kind of accident from any accident previously evaluated since individual minor -fuel failures have been considered.
- 3. Does this change involve a significant reduction in a margin of safety?
Allowing limited use of lead test assemblies that have not completed representative testing does not involve a significant reduction in a margin of safety since the assemblies will '.,
restricted to nonlimiting core regions.
1
-l PERRY - UNIT 1 1 10/1/93 l
ATTACHMENT 2 ITS - PSTS COMPARISON DOCUMENT l
2A: MARKUP OF ITS 28: DISCUSSION OF CHANGES
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ATTACHMENT 2A l i
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ITS - PSTS
^
COMPARISON DOCUMENT -
MARKUP OF ITS !
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l Design Features 4.0 :
4.0 DESIGN FEATURES '
}
4.1 Site 4.1.1 Site and Exclusion Area Boundaries ' -
gg 4V
- [shall be ashcWe 4hc er area I
l The site (and%xclusion are "-^- uew .L,d qs> shown in Figure 4.1-1]s- gog J,3 36u we N __ l o..tclie oR M on Ve d St - N l' 4.1.2 Low Population Zone (LPZ) C' * \"* 45 N fL
- d
- f -
The
( hvea LPZ [Nf3b5 n described ead N s of L s ~-
cr__;;Mu thre ir,_G u.x 4.1-2 II N i Q% cv & ,r y kh c.J e sbat) 4.2 Reactor Core 4.2.1 Fuel Assemblies
_" - E 9 _ _"
> fue14 assemblies. Each assembly
\ The reactor shall consist of shall contain a matrix o ~CM =i= :nci) fuel rods with an j
initial composition of natura or slightly enriched uranium 1 O@\ dioxide {002 ) as fuel material [, erd .te. .vdi]. Limited- '
substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to >
comply with all safety design bases. A limited number of lead '
test assemblies that have not completed representative testing may be placed in nonlimiting core regions. <
1 1
4.2.2 Control Rod Assemblies ,
\^17 h} ;
The reactor core shall contain 3 crucifonn shaped control rod assemblies. The control materia shallbPfboroncarbide, hafnium metal)tr.,appr;;d4ythe':n 8 ,
i b
Gde 1 (continued) 1 l
CR/E STs 9E11'LUurr 1. 4.0 1 pe.g . c, c:/2s/a2 l
4 Design Features 4.0 4.0 DESIGN FEATURES (continued) !
4.3 fuel Storage >
4.3.1 Criticality !
4.3.1.1 The spent fuel storage racks are designed and shall be maintained with: ;
~
-(a. Fuy fassembli having ximum inity o .31] '
configur [kon' at c
~
eactor c condit
[the[ nonnaaver U-235 e 'chment of_.{ .5] weig percent}yJ b'. k, in$ludesanallowanc5 0.95 if fully flooded with unborated waterI I
WSectio f the uncertainties as described in -
i, sA l't i
A nominal g!
1 assembly center to center storage spacing oM7}~Mnches within rows and [1:25] inches between rows i' in the [ic., d:::ity storage racks}Hn the upper containment pgol;_and]R.- '
G,.
C
. A nominal fuil@ assembly center to center storag of [5.25] inches, within a neutron poison material i
between. storage spaces, in theMhigh density storage racks]"in th 4g 1
ees44=::t p;. .,.es e..i. i el- Mer:;:--;;;si end -inme m pag;y glke) uuuci 4.3.1.2 The new fuel storage racks are designed and shall be maintained with: !
i
-(a. F assemblie ving a max' [k-infinit he nonnal actor core figuration a
[1.31] [ f d condit j
[averaj
-235 enri ntof-[4.5] ght percen.t ,f -
)f. k,,,ludess 0.95 if fully flooded with unborated water, which p" a inc allowanc r uncertainties as described in NSection 9,4 of the /SA4}t' ,
_ _ 9.l.) *R
- c. ,,,s 0.98 it A_erateu ay 3queous foam ich incluj
/ an allowan#for uncerta dies as d ibed in /
w[Sectiond.1 of the ;J'and A nominal g hinch center to center distance between fuel assemb ies placed in storage racks.
(continued)
BC/5STObrt1 Wd 1 4.0-2 Rev. % --09ffBf9P l
Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) 4.3.2 Drainace The spent fuel storage pool is designed and shall be maintained to prevent inadvertent _ draining of the;eol below elevation {tofM4
-- 5. 2 5 i r. h Q, ,g (, , 4 A 4.3.3 Capacity 4.3.3.1 The spent fuel storage pool is designed and shall be maintained..with a storage capacity limited to no more than ssemblies.
c232"P{f H0h9 ' 50 4.3.3.2 No more than [000] [,uel ass
-- ] may be stored in the emblies upper containment pool.
eWR/& STS 9a v 3 V 4.h 1 4.0-3 Rev. -0, 09/20/92
Design features 4.0 fiJ6Ek.T HAJ
'Y This figure shall c'nsist o of [a map of] the s\ite area and provide, as a minimum, the infomation described in Section
[2.1.2] of the. FSAR relating to [the map] .
N 4
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I Design Features 4.0 N
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(( ( [t q,y-igheshallcons) tof[amapo the sit area showing t and recreat LPZ boundary. Features suc as towns roads, al areas sha be indicate in suffi o s detail to al identificati of signific t shift ent N '
N population di in ribution within the LPZ.
O Figure 4.1-2 (page 1 of 1) J7 '
- Low Population Zone I
-BWR/6-STS Ec", V e'. \ I 4.0-5 ReV- 0. 09/26/si
ATTACHMENT 2B' i
ITS - PSTS ;
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COMPARISON DOCUMENT i
DISCUSSION OF CHANGES ~ I l
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DISCUSSION OF CHANGES TO NUREG-1434 l CHAPTER 4 - DESIGN FEATURES !
ERACKETED ADMINISTRATIVE CHOICE I i
B.1 Brackets removed and optional wording revised to reflect t appropriate plant specific requirements. (
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l PLANT SPECIFIC DIFFERENCE i i
P.1 USAR Section 9.1.1 indicates that administrative controls will !
be used to prevent optimum moderation conditions from !
occurring. This approach was accepted in section 9.1.1 of the PNPP Safety Evaluation Report, NUREG-0887, dated May 1982.- 't PNPP has-also determined that other administrative controls, '
such as alternate spacing of fuel bundles in the new fuel :
storage vault are adequate to prevent the geometry necessary l for the optimum moderation conditions from being a concern even #
under conditions of water spray / aqueous foam (such as from fire '
protection system actuation). These controls are discussed in i two letters from PNPP to the NRC dealing with deletion of the criticality monitors for the new fuel storage vaults. The {
first, PY-CEI/NRR-1387 L dated February 28, 1992 requested an j Exemption from 10 CFR 70.24, while the second letter, j PY-CEI/NRR-1388 L dated February 28,_1992 requested removal of. !
the criticality monitors from the present Technical l" Specifications. These letters discuss the bases for determining that appropriate fuel spacing in the new fuel vault {
precludes the chances of inadvertant criticality, even under !
optimum moderation conditions. '
P.2 This comment number is not used for this station. ,
P.3 This comment number is not used for this station. t P.4 Title changed to use plant specific nomenclature (Fuel Handling Building). .
P.5 This comment number is not used for this station.
CHANGE / IMPROVEMENT TO NUREG STS '
C.1 Since 10 CFR 50.46 and 10 CFR 50, Appendix K require the use of j "Zircaloy or ZIRLO," the TS are changed to be consistent. ,
C.2 This appears to be reviewer's note material (probably should !
have been inside the brackets). If the control material is
- specified in the Technical Specifications, it is by default "as approved by the NRC. " Any change likewise would have to be "as )
approved by. the NRC. " Since all other TS requirements are also "as approved by the NRC" this phrase is removed for j consistency.
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PERRY - UNIT 1 1 10/1/93 1
l DISCUSSION OF CHANGES TO NUREG-1434 l CHAPTER 4 - DESIGN FEATURES ;
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CHANGE / IMPROVEMENT TO NUREG STS l (continued)
C.3 This additional limitation.is unnecessary to provide adequate I margin of safety for criticality. In order to demonstrate compliance with the requirements for k err, calculations must be '
performed, as described in the safety analysis report, to 1 determine the maximum ek er of the rack. This calculation is ;
dependent on the actual k. (or the enrichment) of the fuel to be stored in the rack. For case of demonstrating compliance with the ktr limit for a particular rack design, a bounding compliance criterion on the k. (or the enrichment) of the fuel .
to be stored in the rack may be established such that the rack's k.rr limit is met. Because the plant is required to-maintain the k ort of the storage rack within the specified limit, each new fuel assembly to be loaded into the reactor i must be compared to the storage rack's bounding compliance criterion (either k. or enrichment, as applicable) . This check :
is required to ensure that the k e rt limit for the storage rack is not exceeded as required by the current (and proposed) -
Technical Specification. In addition, the proposed k. l limitation represents a fuel design feature, rather than a limit on the storage rack design, and the fuel design. !
requirements and limits are already provided in proposed .
Specification 4.2.1. Design reviews for reloads also verify ;
continued compliance with the bounding k. (or enrichment) compliance criterion prior to utilizing the new fuel. These !
current controls provide suf ficient requirements for prevention !
of criticality and make the proposed additional k. or i enrichment limitation unnecessary. !
\
Furthermore, k. is not currently presented in any approved l BWR/6 license. Consistent with all domestic BWR/6's current licensed basis, this limit is not presented within the l Technical Specifications. ,
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I PERRY - UNIT 1 2 10/1/93
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PERRY -
UNIT 1 ;
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SECTION 3.1
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ATTACHMENT 1 CTS - PSTS COMPARISON DOCUMENT 1A: MARKUP OF CTS 1B: DISCUSSION OF CHANGES 1C: NO SIGNIFICANT HAZARDS CONSIDERATIONS i
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ATTACHMENT 1A CTS - PSTS COMPARISON DOCUMENT
- MARKUP OF CTS
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN b ( O 3. k d LIMITING CONDITION FOR OPERATION 3.1.1 The SHUTDOWN MARGIN shall be equal to or greater than:
- a. 0.38% delta k/k with the highest worth rod analytically detemined, or
- b. 0.28% delta k/k with the highest worth rod determined by test.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5.
ACTION:
With the SHUTOOWN MARGIN less than specified:
ConD h a.
In OPERATIONAL CONDITION 1 or 2, reestablish the required SHUTDOWN QB MARGIN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT SHUTD0 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Fu Cm>0 C,0 b. In OPERATIONAL CONDITION 3 or 4 immediately all insertable cnntrol_ rods _to_be_lgserted_andsuspend all activities that couTJ) tGVrgu@g_SHUIDOW x
MARGIN./Ili'DPERMRjNAL CONDITION 47estabj ERIMARY CQfilA1NMENI]$((Gh%hin 8_ _ hours.i - - - -
w ._ q
.c. In OPERATIONAL CONDITION 5, suspend CORE ALTERATIONS .uthlFF Cw0E ~
U ~ 7 C ' "M i e s W t c o u l d re d :" "* """" " "A""Mbin_tTail" g in Q' EstablishfRIMAR'r CONIAlp Q @ iifttable control rodi %
T INTEGRITJiWithTThourti -
SURVEILLANCE REQUIREMENTS (si '3.\.\ \
4-1--E, The SHUTDOWN MARGIN shall be determined to be equal to or greater than specified at any time during the fuel cycle:
a.
By measurement @rior -to or during the first startuplaf ter each GelWl% o 7 '
b.
By measurement, within 500 MWD /T prior to the core average exposure [
at which the predicted SHUTDOWN MARGIN, including uncertainties an_d -
Ncalcidation es_, is equal to Jhe_specified,,1.imit., F 3 g3 c. Withtf1(I TIourraf ter detection of a withdrawn control rod that is immovabre, as a result of excessive friction or mechanical interference, U eG LC-9 nu % k 'l or is untrippable, except that the above required SHUTDOWN MARGIN g withdrawnworthoftheimmovableoruntrippable e 50m w I Or.c t w mo M d ( D O O I47 w O1 G s.a 0 4W S 's
- ~~
PERRY - UNIT 1 3/4 1-1
I REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIES LC O 3. l {
LIMITING CONDITION FOR OPERATION The reactivity equi d land the predicted t
lenceofee'ifferencebetweentheactualf"0D' D DENSIT ha ot exceed 1% delta k/k. -
At.
APPLICABILITY: OPERATIONAL CONDITIONS 1 an 2.
ACTION:
With the reactiv t valence difference exceeding 1% delta k/k:
C6 c o ~ t- a. within nourpperrorm an anairsis to determine and explain the cause3 Cof the reactivi g ifference gperation may continue if the difference is explained and corrected.
Coac e a2 b. Otherwise, be in at least HOT SHlfTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS
.~ .
S Q. ~5.\ 'l t The reactivity equi alenced- the difference between the actual -
} nd the predicted Q){ijiral to 1% delta k/k: ~~
D.SENSI hall be verified to be less than or e
tg
- a. @ ring the first. startud)followingk ALTERATION) and
- b. At least once per 1000 MWD /T during POWER OPERATION.
A h (c. - The provisions of Specification 4.0.4 are not applicab i
PERRY - UNIT 1 3/4 1-2
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LLo 3.l.3 'l 3. ) 3 '
REACTIVITY CONTROL SYSTEMS
- 3 " 'I '^D 3/4.1.3 CONTROL RODS CONTROL R0D OPERABILITY '
N k00 Mt m3 [ 614. D LIMITING CONDITION FOR OPERATION (L15 3'1.3[ All control rods shall be OPERABLE.
APKICABILITY: OPERATIONAL CONDITIONS I and 2. f C*" D ACTION: ,h ,
- a. With lone! control rod inoperable due to beingfiernovable,'as a result of C- 4 4 D A .eTcessiveTrTction or mech' a nic'aTiiiteFference, or known to be untrippable:)
LO. 1.
Within @ hour:
DD a) Verify that the inoperable control rod, if withdrawn'; is
( separated from all other inoperable contro ~iids by~ at least >
g i two control cel1s in all directions.
b) Disarm thejassociated directional control valves ** eithD q.
6o p. 1) Electrically, or s s
Ns i
- 2) Hydraulically by closing the drive water and exhaust water n isolation valves. , ,
ME Otherwise, be in at least HOT SHUTDOWN within the next 12 urs.. i U Q D " 3 /.
~
Comply with Surveillance Requirement 4.1.1.c within hours, or '
Co e O E be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
estore e inoperable withdrawn control rod to OPERABLE status 3 h within 4 12 hou y or be in atdeast HOT SHUTDOWN within the next_ _ b
- b. With one or more control rods trippable but inoperable for causes otner ;
b r> C than addressed in ACTION a, above: g
- 1. If the inoperable control rod (s) is withdr within @ hour:
C O a) Verify that the inoperableGdthdraVrbcdtrol rod (s) is separated
- D D from all other inoperable withdrawn control rods by at least two control cells in all directions, and i b 94 / Demonstrate the insertion capability of the inoperable withdra h
[k control rod (s) by inserting the control rod (s) at least one notch by drive water pressure within the normal operating ran e*. g i
Q' the associated directional control valves"* bither:t-rwise, insert the inoperable a) Electrically, or +
) Hydraulically by closing the drive water and exhaust wat
's_olation valves.
g *The inoperable control rod may then be withdrawn to a position no further wwithdrawn than,its position when .found..to be, inoperable. -- i LA 0 Q'May be rearmed intermittently, under administrative control, to permit
~10 5,d testing associated with restoring the control rod to OPERABLE status.
PERRY - UNIT 1 3/4 1-3
F i
REACTIVITY CONTROL SYSTEMS CowoE - 4. CLER 1 Ts c i I L. ;- o '3 . \ .9 1 m td k V P S l LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Continued) W
- 2. ktheinoperabl_econtrolrod(s)isinserted,withinonehour. disarm bseO ( ~
the associated directional control vaTves"^ either(
~ '
x x, a) Ele'cteically, s
i or -s
() b HydraulicaQy by closing the drive water and exhaust water N isolation valves. -
C-w o E Otherwise, be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
~
Yegrovisions of Spechication 3.0.4 ar'egot appli p\
G a n E' c. With more than 8 control rods inoperable, be in at least HOT SHUTDOWN l within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
t C o ~3,\ ,% * -
d.
44( Or'a Ndl'>* b^
With one scram discharge volume vent valve and/or one scram discharge C o.40 h volume drain valve inoperable _and open, restore the inoperable valve (s) to OPERABLE status within a Lior be in at least HOT SHUTDOWN within C 040 C the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. deg3 -. Q
- e. With any scram discharge volume vent valve (s) and/or any scram dischar q q g volume drain valve (s) otherwise inoperable, restore pt-len t on; vent oc va be =d -e de valve to OPERABLE status within'8*"8 or be in at ;
Ot> 0 C least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. ( ~1 A 35 SURVEILLANCE REQUIREMENTS
' 1. 2. 2. The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by:
t
$0. 3,b%\ % At least once per 31 days verifying each valve to be open,* and 5 0 D ' g ',' b.
At least once per 92 days cycling each valve through at least one complete cycle of full travel.
c, R. 3, l.D 4.1. ^ .1. E Where above tha low power setpoint of the RPCS, all withdrawn _ control t;L N s.
'"*3 required to have their directional control valves disarmed electricall Q ;
r hydraulicallyghall be demonstrated OPERABLE byCmsvingleach control rod at Q@
least one notch: g
- a. At least once pery, days, and -
L2 . O k b. Qtlhper) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when any control rod is immovable as a g g , g A ,1, result of exc~edve friction or mechanical interference.
Q 3.l.t M
i *These valves may be closed intermittently for testing under administrative controls.
g g p]**May be rearmed intermittently, under administrative control, to permit i testing associated with restoring the control rod to OPERABLE status.
PERRY - UNIT 1 3/4 1-4
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
)4.1.3.1.3All ntrol rods shal demonstrath ERABLE by perform Surveillance Re meats 4.1. 3. 2, 4.1M. 3, 4.1. 3 4.1.3.5. , 7
~
(4.1.3.1.4 The scram discharge volume shall be determined OPERABLE by yemonstrating:
b a. The scram discharge volume drain and vent valves OPERABLE at least
. once per 18 months by verifying that the drain and vent valves:
- 1. Close within 30 seconds after receipt of.agsignal for control rods to scram,
\ orsiE 3 gh
- 2. Open when thetscram signal is racet.
N %
- b. roper level senso response by performah of a CHANNEL F TIONAL 4
[(.q T of the scram di ins {trynentation at harge volume scram a leasnce per 31 days.
ontrol rod bloc evel
- l t
PERRY - UNIT 1 3/4 1-5 !
k,l C. D 31. 4 )
RE4TJTVITi~CENTROL SYSTEMS L(g g(q -
CONTROL ROD MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION I
3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position, based on de energization of the scram pilot valve solenoids as time zero, shall not exceed the following limits: ;
Maximum Insertion Times Reactor Vessel Dome to Notch Position (Seconds)
Pressure (psig)* 43 29 13 950 0.31 0.81 1.44 '
1050 0.32 0.86 1.57 :
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
- a. :'
k With the maximum scram insertion time of one or more control rods exceeding the maximum scram insertion time limits :;i Specification 3.1.3.2 as determined by Specification 4.1.3.2.a or b, operation may continue g ded_.1 hat- -
'17 For all " slow" control rods, i.e. , those which exceed the limits of Specification 3.1.3.2, the individual scram insertion times do I
not exceed the following limits:
'l Maximum Insertion Times
( Reactor Vessel Dome to Notch Position (Seconds) d i Pressure (psig)* 43 29 13 950 0.38 1.09 2.09
\ 1050 0.39 1.14 J.22 I. 2.
' For " fast" control rods, i.e. , those which satisfy the limits of i
Specification 3.1.3.2, the average scram insertion times do not exceed the following limits:
V '-
Maximum Average Insertion Times LO to Notch Position (Seconds)
Reactor vessel Dome 3 1. 9 . A Pressure (psig)* 43 29 13 950 0.30 D 75 1.40 1050 0.31 0.84 1.53 3.
The sum of " fast" control rods with individual scram insertion times in excess of the 1 ts of ACTION a.2 and of " slow" control rods ld 0 does se a ::d-7. d.j
- 4. No " slow" control rod, " fast" control rod with individual scram 3Mb insertion time in excess of the limits of ACTION a.2, or other-wise inoperable control rod occupy adjacent locations in any direction, including the diagonal, to another such control rod.
C.W O 4 Otherwise, be in at least HOT SHllTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, pgg. ' *For intermediate reactor vessel dome pressure, the scram time criteria is B determined by linear interpolation at each notch position.-
31.4-i-
\ % (ch PERRY - UNIT 1 3/4 1-6
F ]
i l
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) l ACTION: (Continued)
TO> h '\b. With a " slow" control rod (s) not satisfying (ACTION a.h above:
E9d 1. Declare the " slow" control rod (s) inoperable, and 9.n i i sq,3g33 (. . Perform the Surveillance Requirements of fpecification 4.1.3.2.c at s
least once per 60 dayq when operation is continued with'three or ' ,/
5L Qnore " slow" control rods declared inoperablef
~
-il Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. I
- c. With the maximum scram insertion time of one or more control rods exceed-ing the maximum scram insertion time limits of Specification 3.1.3.2 as !
determined by Specification 4.1.3.2.c, operation may continue provided that:
1 " Slow" control rods, i.e. , those which exceed the limits of TDD1 ,
Specification 3.1.3.2, do not make up more than 20% of the 10% l sample of control rods tested. I h h 2. Each of these " slo " ntrolrodssatisfiesthelimitsof(A
- 3. ThehghtTjWnt trol rods surrounding each " slow" control rod are: !
3 L H 1. D ~ Demonitrated' thro 6gh meas'ur'ement within 124ours-to ntisfy 'the ~ A
( maximum scram insertion time.l.imits of_ Spec.if.i. cation _3A .3.2, and b) OPERABLE. '
t r o ~f.1.9 o 4. The total number of " slow" control rods, as determined by gj Specification 4.1.3.2.c, when added to the sum of ACTION a.3r as determined by Specification 4.1.3.2.a and b, does not excee 'W I
D k Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- d. ~fep' T rov ions of Specification 3.0.4 are not applicaMs.N A -
SURVEILLANCE REOUIREMENTS Oo! 4.1.3.2 The maximum scram insertion time of the control rods shall be gt demonstrated through measurement with reactor coolant pressure greater than or l equal to 950 psig and, during single control rod scram time tests, the contro11 M i rod drive pumps isolated from the accumulators: i N +if
- a. For all control rods pr' to THERMAL POWER exceeding 40% of RATED
$ Q. 3.\.4, g THERMAL POWER following ' f ALTERATICIiiS) r a reactor shutdown that is greater thah~120' Hays, 50.1l9 3 b. For specifically affected individual control rods
- following i g 3gqq {g maintenance on or could modification drive system which affect theto scthe control
'usertionrod timeorofcontrol those rod j specific control rods, and . r 5(L3,1,y,1 c. For at leas
( ggg once per 120(days ~of POWER OPERATION.~ " ~ -10FN"th N 3 D 3 khe provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 2 provided this surveillance is completed prior to k 1lA 4 :
g entry into OPERATIONAL CONDITION 1.
PERRY - UNIT 1 3/4 1-7 i
REACTIVITY CONTROL SYSTEMS lC03IT i CONTROL ROD SCRAM ACCUMULATORS Q d. G m r.1 h o q. e. tbs wh LIMITING CONDITION FOR OPERATION h 4 00 S C-1 \b e4 s N og y 3.1.3.3 All control rod scram accumulators shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and h -
In OPERATIONAL CONDITIONS 1 or 2: r sY soo
- 1. control rod scram accumulator inoperablee within
( g a)--Restore the inoper21: ::cM ate;- to OK"AP8:E-st t= ,a r b) Declare the control rod a iated with the inoperable accumulator QnoperabTe;' t._
4ther he, oe in ar, inst-10T-stt!T00VN sithin the next 12-Aoups.
- 2. With[more than ont control rod scram accumulator inopera le, Coso GA [
de-lare the aMciated control r<cem. > rods dnoperabTd)and: L. t O wi ---
a) If the contro1\ rod associated with any inoperable scram fKO A li Cl accumulator is Lwithdrawnl immediately verify that at least j_n_econtrol_roddrivepumpisoperatingJbyinsertin at lea @
2.EQ ) ti. G3 jg) (one withdrawn control rod at least one notch. f no control roTdrTve ~ pump is operating:
qq q gy qi t g (1) If reactor pressure is >(50d,g psig, restart g at least one control rod drive pump within 20 minutes or place the reactor mode switch in th utd position.
s ocp -
(I C o t. 6 (. tD (2) If reactor pressure is < 5~1g, p ace the reactor mode switch in the Shutdown position.
b) [.1nsert the inoperable control rods and disarm thehciated
/ 6irectional control valves either.
LC.0 ' -
/ '
- 1) Electrically, or g l2) Hydraulically by closing the drive water and exhaust g water isolation valves. /
(Otherwise,_ be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.-pr.
- 5. In OPERATIONAL CONDITION 5*:
DvMb 1. With one withdrawn control rod with its associated scram accumu-To lator inoperable, insert the affected control rod and disarm the tg associated directional control valves within one hour, either:
1 a) Electrically, or SE b) Hydraalically by closing the drive water and exhaust water p isolation valves.
' 'At least the accumulator assc:iated with each withdrawn control rod. Not plicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
PERRY - UNIT 1 3/4 1-8
l REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
'* V ' ^ \ 2.
12 t. With more than one withdrawn control rod with the associated s L,, t o accumulator inoperable and eith no control rod drive pump operating, i immediately place the reactor mode switch in the Shutdown position. '
provisions fSpecifb 3.0.4 are-not apphcable SURVEILLANCE REQUIREMENTS 4.1.3.3 ,
Each control rod scram accumulator shall be determined OPERABLE:
bb a.
b 3g5) At least once per 7 days by verifying that_ the_ptessure is creater than 9r-._equalto_1520 psigjuiiless the control rod is inserted aiiBy
{ rmed on s
'cr}amme p M g u y 9 sq -
- b. At least once per 18 months b' y performance of a:
Qj 1.
CHANNEL FUNCTIONAL TEST of the leak detectors, and -
R 2.
CHANNEL CALIBRATION of the press'ure detectors, and verifying n alarm setpoint of > 1520 psig cnNecreasino pressure.
J 1
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PERRY - UNIT 1 3/4 1-9 Amendment No. 32 ,
REACTIVITY CONTROL SYSTEMS dQ 1l}
CONTROL ROD DRIVE COUPLING 3 ).3 $
LIMITING CONDITION FOR OPERATION ND - 1. 3. 0 All control rod; shall be s.oupled to their drive mechanisms APPLICABILITY:
ACTION:
OPERATIONAL CONDITIONS 1, 2
[
~~
- a. [In OPERATIONAL CONDITION 1 and 2 with one control rod not coupled to its C o % r> q, associated drive mechanism, within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: N
\
[ 1. kpermittedbytheRPChinsertthecontrolroddrivemechanismto ,
N accomplish recoupling and$ verify recoupling by withdrawing the j control rod, and: ( '
a) Observing any indicated response of the nuclear instrumentation, LN -s and t fb) overtravel Demonstrating that the control rod drive will not go to the position.
y Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. L.7.-
- b. rIf recoupling is not acdomplished on the first attempt or.
\' Jermitted by the RPCS,[then until permitted by the RPCS, declare the ifno3 mgantrol rod inoperable,d nsert the_ronten sand disarmgthe ociaTid lIrectional contro'l valves ** either. ;
L4 1 .
a) Electrically, or
\ b) Hydraulically by closing the drive water and exhaust water N isolation valves.
C.0" o E Otherwise, be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
-. - s In OPERATIONAL CONDITION 5* with a withdrawn control rod not coupled to
its assoc,ii.ted drive mechanism, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, either: \
s
- 1. InserthecontrolrodtoaccomplishTecouplingandverifyrecoupling \ !
i by withdrawing the control rod and demonstrating that the control I rod will not go the overtravel position',sor
- 2. If recoupling is not accomplished, insert the c'oatrol rod and disarm a '
s the associated direc.'3nal control valves ** either a) Electrically, c- '
s l N. b) Hydraulically by closing the drive water and exhaust water w ._i, solation valves. -
k The'phvisions NpIifickios 3.0.4 are not applicablefk
~
/ *At least each withdrawn control rod. Not applicable to control rods removed Qer Specification 3.9.10.1 or 3.9.10.2. .p _
,g ;.""MayWreatmed16termittently, under administrative control, to permit 33.5 testing associated with restoring the control rod to OPERABLE status.
PERRY - UNIT 1 3/4 1-10
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS SCL 3. \ 3 I 4.1. 2. t Each af fected contrn] zod_shalLbe demonstr.ated_.toJe couf l ed to its drive __mecha M observing any indicated response of the nu' clear instrum jen-h Q t_i[on whileJ ithdr3MiDg_theJ ontrol rod to the f.ully withdrawn positioprand then verilying that the control rod drive does not go to the overtravel position:
m 1 a. Prior to reactor criticality after completing CORE ALTERATIONS that
'N s s__could have affected the control rod drive coupling integrity,
- b. Anytime the control rod is withdrawn to the " Full out" position in subsequent operation, and ;
- c. Following maintenance on or modification to the control rod or control rod drive system which could have affected the control rod drive coupling integrity.
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4 PERRY - UNIT 1 3/4 1-11 i
P REACTIVITY CONTROL SYSTEMS L(, Q 3, l 3
. CONTROL R0D POSITION INDICATION .$ 1 3 , j , 3 ,
LIMITING CONDITION FOR OPERATION M I 3 .1.3. 5 OPERABLE.***
At least one control rod position indication system shall be !
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2 k f* )
ACTION:
C.N D C. a.
In OPERATIONAL CONDITION 1 or 2 when the position of any OPERABLE control rod cannot within indicator, be determined 1 hour: by at least one OPERABLE control rod position
- 1. 1 Move indicator, the control or j rod to a position with an OPERABLE pos % .
- 2. When THERMAL POWER is:
e a) Less than or equal to the low power setpoint of the RPCS:
ll1) Declare the control rod inoperable, and
- 2) Verify the position and bypassing of control rods with 3
inoperable " Full-in" and/or " Full out" position indicators (SR33.p,q j by a second licensed operator or other technica11y qualified member of the unit technical staff. >
b) Greater than the low power setpoint of the RPCS, declare the control rod inoperable. insert the control rod nd disarugthe :
[associateddirectionalcontrolvalves** fther: / i
- 1) Electrically, or l 2) f
( N isolation valves. Hydraulically by closing the drive water and ex Cwo& Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
., b.
In OPERATIONAL CONDITION 5* with both position indicators of a withdrawn control rod inoperable, move the control rod to a position with an OPERABLE position indicator or insert the' control rod.
@ c[ ~Th y visic h tSpecificari d 3.0.4 ire not applic WQ s
- At least each withdrawn control rod.
- *A y per Specification 3.9.10.1 or 3.9.10.2.Notapplicabletocontrolrodsremoved).
f
{
l b !
3.g.f *May be rearmed intermittently, under administrative control, to persi
. st33 associated with restoring the control rod to OPERABLE _ status. .
~
- This reqMt-~iliaN WiaB5 TIED Tf tiie pobt[ioE of eac6'~0PERABIEN i
bI control rod can be determined by at least one OPERABLE control rod /- i j
(% position indicator. , _.
-~ '
PERRY - UNIT 1 3/4 1-12 r
l
RfACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 5 3,13 I) 4.1.3.5;]The above required control rod position indication system shall be determined OPERABLE by verifying:
- a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the position of each control rod is indicated, and That thelndicated controProd. position changes during the movement A g- {b.
s of the control'Tod drive when performing Surveillance Requirement s i M,J. 3.1.)w_
t 9
, PERRY - UNIT 1 3/4 1-13 !
i
r-1 9 '
REACTIVITY CONTROL SYSTEMS 4 N
CONTROL Q DRIVE HOUSING SUPPORT *
,\
LIMITING CONDITION FOR OPERATION 3.1.3.6 The control ~ rod drive housing support shall be in place.
APPLICABILITY: OPERATIONd'ts. CONDITIONS 1, 2 and 3.
ACTION: 'N N
With the control rod drive housing hopport not in place, be in at least HOT SHUTOOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
\s.
N i
'x SURVEILLANCE REQUIREMENTS
\ ,
\
g i 4.1.3.6 Thecontrolroddrivehousingsupportshallbeverifiedto\e,inplace by a visual inspection prior to startup any time it has been disassembled or when area. maintenance has been performed in the control rod drive housing supp'or,t l N '
9 f
6 b
1 i
PERRY - UNIT 1 3/4 1-14
r i
REACTIVITY CONTROL SYSTEMS 3/4.1.4 CONTROL R0D PROGRAM CONTROLS [
CONTROL R0D WITHDRAWAL ,
LIMITING CONDITION FOR OPERATION ,
i
] 3.1. 4.1 Control rods shall not be withdrawn. i f, 'l APPLICABILITY: OPERATIONAL CONDITIONS I and 2, when the main turbine bypass
~tvalves are not fully closed and THERMAL POWER is greater than the low power L Lo isetpoint of the rod pattern control system (RPCS).
- 21 ACTION:
(' j a any control rod withdrawal when the main turbine bypass valves are not
.itully closed and THERMAL POWER is greater than the low power setpoint of the t VRPCS, immediately return the control rod (s) to the position prior to hcontrolrodwithdrawal.
l
\
ii ,
i SURVEILLANCE REQUIREMENTS 4.1. 4.1 Control rod withdrawal shall be prevented, when the main turbine bypass valves are not fully closed and THERMAL POWER is greater than the ~10w t power setpoint of the RPCS, and verified by a second licensed operator or '
i other technically qualified member of the unit technical staff.
h PERRY - UNIT 1 3/4 1-15 ,
- ( bCru.t d k - o r $ <. M M A REACTIVITY CONTROL SYSTEMS U 3. l 3 O
ROD PATTERN CONTROL SYSTEM 3. I A LIMITING CONDITION FOR OPERATION t G . I . ? .? The rod pattern control system (RPCS) shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2@ SQ. 2 3.2 g 3 ACTION:
% a. With the RPCS inoperable or with the requirements of ACTION b, 4 ; below, not satisfied and with:
7o ttg 1. THERMAL POWER less than or equal to the RPCS low power setpoint, control rod movement shall not be permitted, except by a scram.
3311 2.
l THERMAL POWER greater than the RPCS low power setpoint, control rod withdrawal shall not be permitted.
~ h*k b. With an inoperable control rod (s),'DFERXBLE control rod movementW' M b 2 M k onti_nue by Dypassing the inoperable control rod (s) in the RPCS '
L C 0 3.1,3, Qrovideo tnawy . -- -
- 1. With one controT~ rod ~ inoperable due to being immovable, as a
[ C o a c) g result of excessive friction or mechanical interference, or g known to be untri a R thic J g M ia ennten1 rnd may be . o W (h -f bypassed in th d ang ^drin systee-RG05Pand/or thclo"d~'
action control system 1Tp~rovided~that the SHUTDOWN MARGIN' aTbeen deteiiidni' d to be equal to or greater than required by/
Spec i fica t io n 3.1.1.y' ' ~~~~~~~'~~ ~~
- 2. Withhi~topt1Lcpjltr rods iroperable for causes other than
( 0w D C'4 - addresse~d in ACTION b.1, above, these inoperable may be bypassedrin the RACScprov'ided thatD-trol rods i J - a) e control rod Qo be bypassed]is' '~ inserted and the ' dir~e~cTion1TToritrol valves are disarmed fiThe~r. L .)
- 1) LAt Electrically, or ;
- 2) Hydraulically by closing the drive water and exhaus '
water isolation valves. 1
\
[b) All inoperable control rods are separated from all other 340 0 , inoperable control rods by at least two control cells in l I ( all directions.
. [ I t
(~ c) Th8arenotmorethanQinoperabl'econtrolrodsinany] ( RPCS 'broup. ~s N N y
" ' * ' 3.i l The position and bypassing of an inoperable control rod (s) is SR h2 ' verified by a second licensed operator or other technically 3 3 f,(,q L qualified member of the unit technical staff.
e_Special Test Ex.geption 3.10.2 )
~
4 400 Lc. o 3 4 t, 5.. j
# Entry into OPERATIONAL'CdRDTT!5HT and withdrawal of selected control rods gg is pemitted for the purpose of determining the OPERABILITY of the RPCS prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.
x i 1 d =}^
;g_
o S . M_ .p*=e .s..M;M3 9 on %J . PERRY - UNIT 1 3/4 1-16 i
REACTIVITY CONTROL 1 YSTEMS SilRVEILLANCE REQUIREMENTS 4.1.4.2 The RPCS shall be demonstrated OPERABLE by verifying the OPERABILITY of the: MW6 a. Rod pattern controller when THERMAL POWER is less than the low power T S setpoint by selecting and attempting to move an inhibited control rod:
- 1. After withdrawal of the first insequence control rod or gang, O 2.
for each reactor startup. As soon as the rod inhibit mode is automatically initiated at the RPCS low power setpoint, during power reduction.
- 3. The first time only that a banked position, N1, N2, or N3, is 4
reached during startup or during power reduction below the RPCS low power setpoint. h
. b. Rod withdrawal limiter when THERMAL POWER is greater than or equal to the low power setpoint by selecting and attempting to move a restricted control rod in excess of the allowable distance:
- 1. As power is increased above the RPCS low power setpoint and RPCS high power setpoint, and as power is decreased below the RPCS high power setpoint.
- 2. At least once per 31 days while operation continues above the l
RPCS low power setpoint. 4 PERRY - UNIT 1 3/4 1-17 , i l
REACTIVITY CONTROL SYSTEMS I 3 /L0 1. 5 STANDBY LIQUID CONTROL SYSTEM C - LIMITING CONDITION FOR OPERATION Lfo't17 iHhdJ
'w andby liquid control system subsystems shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2(and 5*.} ACTION:
- a. In OPERATIONAL CONDITION 1 or 2:
C. 2 4 0 b 1. With one system subsystem inoperable, restore the inoperable c ,a g q subsystem to OPERABLE status within 7 days or be in at least HOT SHUTD0hH within the next 12 hours. c ogo @ 2. With both standby liquid control system subsystems inoperable,
, restore at least one subsystem to OPERABLE status within 8 hours
'340 C or be in at least HOT SHUTD0hH within the next 12 hours.
- b. In OPERATIONAL CONDITION Sh y --
- 1. With onegystem subsystem inoperable,. restore the inoperable subsystem h OPERABLE status within 30 Jays or insert all insertable cbatrol rods within the next hour. x[
- 2. With both stand iquid control system subsystems inoperable, insert all insertable control rods within one hour. _
SURVEILLANCE REQUIREMENTS 4.1.5 Each standby liquid control system subsystem shall be demonstrated OPERABLE:
- a. At least once per 24 hours by verifying that; S CL 3 .12).7 1. The temperature of the sodium pentaborate solution is greater than or equal to 70 F.
R3i71 2. The available volume of sodium pentaborate solution is within the limit of Figure 3.1.5-1. g r;.g. i,3 3. The heat tracing circuit is OPERABLE by determining the temperature of the pump suction piping to be greater than or equal to 70*F.
\ . ~ ~ ~ . . - - -
*With any tontrol rod withdrawn. ' Jot applical -
to control rods emoved per N.A. Specificatioh 4.9.10.1 or 3.0.10.2. . . , .
._ S.__. . . . . _ . . _ _ . . _ - - -
PERRY - UNIT 1- 3/4 1-18
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- b. At least once per 31 days by; 50.3.l. D 1. Verifying the continuity of the explosive charge.
~
~
- 2. Determining _that'ThTavaiia51Ew~elght of"sodikentaborate theYdium~pentatiorate g 3. ),-), q Qs1reiter solutioh'ciin~ than or equal centration is to 5236 within lbs and)its of Figure 3.1.5-1 thy'11m by chemical analysis."
- 3. Verifying that each valve, manual, power operated or automatic, S C. 3.s.1 L ,
in the flow path that is not locked, sealed, or otherwise f secured in position, is in its correct positianc or cu be .hyd ' QL 4. %e carnd p sg, ,
- c. Demonstrating that, when tested pursuant to Specification 4. . , ~u 503.\d.7 the minimum flow requirement of 41.2 gpm per pump at a pressure of greater than or equal to 1220 psig is met.
F LA 2
- d. At lea once per mgnths Qurina shutdoM y; -
L . om 1: Ini latingttrth, f the standby liquid control system subsystems, including an'esplosive valve, and verifying that a flow pa h g g g
- 3 ') O from.thepumpstothereactorpressurevesselisavailableyby jhumping demineralized ~waterjoto. the re Nreplacement charge ~for 'th~e explosiviValve shalTbi from the C>
Q 3~Jlwhich same manufactured batch as the one fired or from another batch has been certified by having one of that batch success-
' ~~
(fullyfired._ , 5 /2 3 ,J ,7 7 2. Demonstrating that all heat traced pipipg jetween_the . h storage tank and the reactor vessel is unblockedJby pumping from'the storage tank to the test tank and then draining and flushing ') LA l the g ing with demineralized water.** f- --
~ ~
9~ ; s d. Demons'tratirig that~the'st6fEge tank operating heater is OPERABLE l L7 by verifying the expected temperature rise of the sodium pentaborate ! solution in the storage tank after the operating heater is energized. J l g ny,g l"This test shall also be performed anytime water or boron is added to the I solution or when the solution. temperature drops below 70*F. st7)gq **This test shall also be performed whenever both heat _.trittiDg_ circuits 4 ave-been j
, found to be inoperable and)may be performed by any series of sequential, over- -l I
(~' Tapping or total flow path steps such that the entire -- - flow path is included.-
-) '
PERRY - UNIT 1 3/4 1-19
15 A 3 i c g LG'$ HIGH-3g g4 LEVEL ALARM LEVEL ALARM _ OVERFLOW-
, VOLUME - ms ! . .l.. , , ,,
y '% % 2 ~% 8>3 m w MARGIN m
}$
- REGION OF APPROVED VOLUME - CONCENTRATION _ \ s'
\ \
w OE \ 1 e n. ~ ' N / $0 12- NJ MINIMUM REQUIRED CONCENTRATION t.INE II 4260 4409 4647 5013 V - NET TANK VOLUME (GALLONS) SODIUM PENTABORATE SOLUTION CONCENTRATION / VOLUME REQUIREMENTS Figure -3rl:Srl- 3,/, V- /
p-ATTACHMENT 1B CTS - PSTS COMPARISON DOCUMENT DISCUSSION OF CHANGES
i i DISCUSSION OF CHANGES CTS: 3.1.1 - SHUTDOWN MARGIN 4 l ADMINISTRATIVE A.1 In MODES 3 and 4 a single control rod may have been withdrawn 'i under the provisions of the Special Operations LCO 3.10.3 and LCO 3.10.4, or some unanticipated event may have resulted in l uninserted control rods. Therefore, rather than an action to !
" verify... inserted," the proposed ACTION is more definitive;
" Initiate action to insert. . . . " This wording provides the same intent in the event all insertable control rods are found to be ;
inserted, but also clarifies that any uninserted control rods are to be inserted. ; A.2 In MODES 3 and 4 the vessel head is bolted in-place and the , only activity that could significantly reduce SHUTDOWN MARGIN (SDM) is control rod withdrawal. Since an ACTION is provided which ensures control rods remain inserted, any additional action to suspend activities that could reduce the SHUTDOWN MARGIN is redundant and unnecessary. Similarly, in MODE 5 the' , only activities that could affect SHUTDOWN MARGIN are CORE ALTERATIONS and control rod withdrawal . Since ACTIONS are provided to suspend CORE ALTERATIONS and ensure control rods remain inserted, any additional action to suspend other activities is redundant and unnecessary. , A.3 This change replaces the use of the defined term _ PRIMARY CONTAINMENT INTEGRITY with the essential elements of that definition. The change is editorial in that all the i requirements are specifically addressed by the proposed , Required Actions D.2, D.3, and D . 4 and E.3, E.4, and E.5. , Therefore, the change is purely a presentation preference adopted by the BWR Standard Technical Specification, NUREG-1434. Refer also to the comment in the Definitions Section which addresses deletion of the various CONTAINMENT INTEGRITY ! definitions. l A.4 The existing ACTION to " establish PRIMARY CONTAINMENT INTEGRITY : within 8 hours" would appear to provide a period of time (8 hours) in which integrity could be violated even if capable of being maintained. Additionally, if the plant status is such 3- that integrity is not capable of being established within 8- t hours, the existing ACTION results in "non-compliance with the l Technical Specifications" and a requirement for an-LER. The i intent of the ACTION is believed to be more appropriately ; presented in the proposed Required Actions D.2, D.3, D.4, and : E.3, E.4, E.5. With the proposed ACTIONS, a significantly more : conservative requirement to establish the primary containment boundary and maintain it is imposed. No longer would the l provision to violate the boundary for up to 8 hours appear to r exist. With this conservatism however, comes the understanding ! that if best ef forts to establish the boundary took longer than l' 8 hours, no LER would be required. PERRY - UNIT 1 1 10/1/93 . w
DISCUSSION OF CHANGES CTS: 3.1.1 - SilUfDOWN MARGIN ADMINISTRATIVE (continued) This interpretation of the intent. is supported by the BWR Standard Technical Specification, NUREG-1434. As an enhanced presentation of the existing intent, the proposed change'is deemed to be administrative. A.5 The existing ACTION to " insert...within 1 hour" (see comment L.2 for a change to what is required to be inserted) is' proposed to be revised to " initiate action to fully insert ... Immediately." This change is similar to that discussed in A.4 above. The existing requirement would appear to provide an hour in which control rods could be left withdrawn, even if able to be inserted. Also, if the control rod is incapable of being inserted in 1 hour, the existing ACTION would appear to result in the requirement for an LER. The intent of the ACTION is believed to be more appropriately presented in proposed Required Action E.2. With the proposed ACTION, a significantly . more conservative requirement to insert the_ control rod (s) and maintain them inserted is imposed. No longer would the provision to withdraw or leave withdrawn one or more control' rods for up to I hour appear to exist. With this conservatism however, comes the understanding that if best ef forts to insert the control rod (s) took longer than 1 hour, no LER would be required. This interpretation of the intent is supported by the BWR Standard Technical Specification, NUREG-1434. As an enhanced presentation of the existing intent, the proposed change is deemed to be administrative. A.6 A specific performance time for the SDM test is proposed to clarify when " prior to or during the first startup." Most SDM tests are performed as an in-sequence critical and therefore four hours after reaching criticality is provided as a reasonable time to perform the required calculations and have appropriate verification completed. Interpretations both more and less conservative could be made far the existing requirement, however this interpretatica of the . intent is supported by the BWR Standard Technical Specification, NUREG-1434. As an enhanced presentation of the existing intent, the proposed change is deemed to be administrative. A.7 More explicit wording is proposed to replace the activity referred to as " refueling." The intent is to perform the SDM test after in-vessel activities which could have altered SDM. These are more explicitly presented as " fuel movement within the reactor pressure vessel or control rod replacement." As an enhanced presentation of the existing intent, the proposed change is deemed to be administrative. PERRY - UNIT 1 2 10/1/93
DISCUSSION OF CHANGES CTS: 3.1.1 - SHUTDOWN MARGIN ADMINISTRATIVE (continued) A.8 The definition of SDM has been modified to require additional margin when a control rod is stuck (refer to Section 1.1). RELOCATED SPECIFICATIONS None iL this section. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 An additional Surveillance Frequency for SDM verification has been added to clarify the requirements necessary for assuring SDM during the refueling process. Because SDM is assumed.in several refueling mode analyses in the USAR, some measures must be taken to ensure the intermediate fuel loading patterns during refueling have adequate SDM. This change imposes a-requirement where none is explicitly provided in the existing Technical Specifications. This new requirement does not, however, require introducing tests or modes of operation of a new or different nature than currently exist. As presented in the Bases corresponding to this requirement, this is best accomplished by analysis (rather than in-sequence criticals) because of the many changes in the core loading during a typical refueling. Bounding analyses may be used to demonstrate adequate SDM for the most reactive configurations during refueling thereby showing acceptability of the entire fuel movement sequence. TECHNICAL CHANGE - LESS RESTRICTIVE " Generic" None in this section. PERRY - UNIT 1 3 10/1/93
r DISCUSSION OF CHANGES i CTS: 3.1.1 - SHUTDOWN MARGIN i TECHNICAL CHANGE - LESS RESTRICTIVE : (continued) !
" Specific" ,
L.1 The existing requirement to suspend all CORE ALTERATIONS precludes off-loading fuel and inserting con.rol rods'. The - proposed change is to modify the requirement to suspend CORE ALTERATIONS by inserting "except. for control rod insertion and ' fuel assembly removal." This exception allows for the ; continuation of activities that have a potential to correct the l problem, and restore the margin of safety. This additional operational flexibility does not require new or different actions, but allows corrective actions which would have otherwise been precluded (except under the provisions of 10 CFR -i 50.54 (x) ) . The corrective actions would only be pursued in ; accordance with approved procedures. ; L.2 The ACTION to insert all insertable control rods in MODE 5 has been modified to only require those control rods in core cells.- ' containing one or more fuel assemblies to be fully inserted. ' If all fuel _ assemblies are removed from a_ core cell, inserting. t the associated contrel rod has a negligible impact on core reactivity. Furthermore, during MODE 5 operation, refueling procedures could have cells emptied and the control rod j withdrawn, but " insertable." However,.due to a variety of : considerations (i.e., location of. blade guides, ongoing instrumentation maintenance, water chemistry, etc. ), it may not ; be desirable to insert these control rods. Since there is
- negligible impact on SDM should the control rod be inserted i with no fuel in the cell, it is acceptable to provide this ,
flexibility. _ L.3 The SDM limits adequately account for uncertainties and biases, and for fuel cycle changes. As long as the required margin is , met, as determined by the initial- startup test and as 1 corroborated by the periodic reactivity anomaly surveillance (current surveillance 4.1.2), there should be no need for l additional Surveillance Requirements. Furthermore, the ; requirement to perfonn an additional demonstration just prior , to the predicted SDM equaling the limit, would. require a ' shutdown of the plant. This shutdown to perform the test would i be in addition to the shutdown that is required a short time i later when the SDM requirement is no longer. met. The l acceptability of the proposed' surveillances, ' which do not l include an additional test just prior to the SDM limit not ' being met, is further supported in the BWR Standard Technical Specification, NUREG-1434 presentation of these requirements. PERRY - UNIT 1 4 10/1/93
b DISCUSSION OF CHANGES CTS: 3.1.1 - SHUTDOWN MARGIN TECHNICAL CHANGE - LESS RESTRICTIVE (continued) L.4 The requirement to verify SHUTDOWN MARGIN with a control rod stuck has been moved to LCO 3.1.3 as Required Action A.3 for discovery of the stuck control rod. With a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach COLD SHUTDOWN is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram. Even with this postulated additional single failure, sufficient reactivity control remains to reach and maintain HOT SHUTDOWN conditions. Also, a notch test is required by LCO 3.1.3 for each remaining withdrawn control rod to ensure that no ad.ditional control rods are stuck. Given these considerations, the extended time allowed to verify SHUTDOWN MARGIN provides a reasonable time to perform the analysis or test. PERRY - UNIT 1 5 10/1/93
I r DISCUSSION OF CHANGES CTS: 3.1.'2 - REACTIVITY ANOMALIES ADMINISTRATIVE A.1 A specific time for completing the reactivity anomaly : surveillance is proposed. This will clarify when "during the ! first startup" the test must be performed. This test is l performed by comparing the monitored rod density to the ! predicted rod density as a function of cycle exposure while at . steady state reactor power conditions. Therefore, 24 hours i af ter reaching these conditions is provided as a reasonable ; time to perforr the required calculations and have appropriate verification completed. Interpretations, both more and less conservative, could be made for the existing requirement; , however this interpretation of the intent is supported by the BWR Standard Technical Specification, NUREG-1434 As an . enhanced presentation of the existing intent, the proposed i change is deemed to be administrative.
~
A.2 The term rod density has been removed from the Definitions section (1.0) of these proposed Specifications. The term is not used repeatedly in the proposed Specifications and the definition of the term is no longer required. This is i consis, tent with NUREG-1433 which is written for the BWR-4 i product line, and uses the rod density term, but does not l require a definition. .! A.3 Surveillance Requirement 4.1.2.c dealing with the provisions of ; Specification 4.0.4 have been incorporated into the Frequency ; requirement of new Surveillance Requirement SR 3.1.2.1 by requiring the SR be performed within 24 hours after reaching ! equilibrium conditions following startup after fuel movement , within the reactor vessel or control rod replacement. I s RELOCATED SPECIFICATIONS None in this section.
]
TECHNICAL CHANGE - t40RE RESTRICTIVE None in this section. PERRY - UNIT 1 6 10/1/93 l l
m DISCUSSION OF CHANGES CTS: 3.1.2 - REACTIVITY ANOMALIES ! TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 This ACTION involves re-evaluating predicted core reactivity conditions in an effort to explain and correct the difference j such that, based on the new evaluation, the reactivity difference is returned to acceptable limits. It is proposed to have the specifics of this process be removed from the Technical Specifications in accordance with the BWR Standard Technical Specification, NUREG-1434. Generally, the details of ,
how a specific action is performed are not located in the I Improved Technical Specifications. This change does allow procedural revisions in accordance with 10 CFR 50.59 in lieu of a formal amendment request for any changes to this method of restoring the reactivity difference to within limits.
" Specific" l L.1 The time allowed to restore the core reactivity difference to within limits (i.e., to " perform an analysis to determine and explain the cause of the reactivity difference") is proposed to be increased from 12 hours to 72 hours. Typically, a reactivity anomaly would be indicative of incorrect analysis inputs or assumptions of fuel reactivity used in the analysis.
A determination and explanation of the cause of the anomaly would normally involve an offsite fuel analysis department and the fuel vendor. Contacting and obtaining the necessary input may require a time period much longer than one shift (particularly on weekends and holidays) . Since SHUTDOWN MARGIN has typically been demonstrated by test prior to reaching the conditions at. which this surveillance is performed, the safety impact of the extended time for evaluation is negligible. Given these considerations, the BWR Standard Technical ~ Specification, NUREG-1434 allows this time to be extended to 72 hours. L.2 " CORE ALTERATIONS" is proposed to be replaced with " fuel ' movement within the reactor pressure vessel or control rod replacement." The intent of this surveillance is to verify. the core reactivity after in-vessel operations - which could have significantly altered the core reactivity. .Certain CORE' ; ALTERATIONS have a known affect which is reversible and, in fact, are activities consistent . wi' a those assumed to occur. during routine operations. Normal C ntrol rod movement is such an activity. Since this activity does not require- I i PERRY - UNIT 1 7 10/1/93 1 1
.i
DISCUSSION OF CHANGES : CTS: 3.1.2 - REACTIVITY ANOMALIES ! i TECHNICAL CHANGE - LESS RESTRICTIVE (continued) re-verification of core reactivity during normal operations ! with the vessel head on (i.e., not defined as a CORE j ALTERATION), it should also be allowed with the reactor vessel head removed (i.e., defined as a CORE ALTERATION) without a ! requirement to re-verify core reactivity. The proposed wording provides a specific list of those CORE ALTERATIONS which would constitute a core reactivity change not expected to . occur during normal operations, specifically excluding normal control rod movement. i L.3 This comment number is not used for this station. f f i i i i i r
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PEOV - I"IIT 1 8 10/1/93
DISCUSSION OF CHANGES ! CTS: 3.1.3.1 - CONTROL ROD OPERABILITY 1 ADMINISTRATIVE A.1 The organization of the Control Rod OPERABILITY Specification is proposed to include all conditions that can affect the ability of the control rods to provide the necessary reactivity
- insertion and also to be simplified as follows:
- 1) a control rod is considered " inoperable" only when it is degraded to the point that it cannot provide its scram function and all inoperable control rods (except stuck .
rods) are required to be fully inserted and disarmed.
- 2) "Untrippable" is no longer treated as a " stuck" control rod provided it can be fully inserted. If only the "trippable" capability is lost, the control rod is considered inoperable (and must be fully inserted and disarmed) and is not considered " stuck." _
- 3) a control rod is considered " inoperable" and " stuck" if it is incapable of being inserted and requirements are retained to preserve SHUTDOWN MARGIN for this situation. [
- 4) a control rod is considered " slow" when it is capable of providing the scram function but may not be able to meet !
the assumed time limits.
- 5) special considerations are provided for conformance to the ,
banked position withdrawal sequence (BPWS) at less than + 20% of rated thermal power. The scram reactivity used in the safety analysis allows for a specified number of inoperable and slow scramming rods, and the control rod drop accident analysis provides additional ! considerations of the BPWS at low power levels. ! A.2 This proposed Note (" Separate Condition entry is allowed ...") l provides more explicit instructions for proper application of , the ACTIONS for Technical Specification compliance. In i conjunction with the proposed Specification 1.3 " Completion Times," this Note provides direction consistent with the intent i of the existing ACTIONS for inoperable control rods and scram discharge volume (SDV) vent and drain valves. In the first case, each inoperable control rod is intended to be allowed a l specified period of time to verify compliance with certain-limits and, when necessary, fully insert and disarm the control , rod. Additionally, each SDV line is intended to be allowed a : specified period of time to confirm it is isolated or capable of isolation, and to restore the complete function of the line. r PERRY - UNIT 1 9 10/1/93
E DISCUSSION OF CHANGES l CTS: 3.1.3.1 - CONTROL ROD OPERABILITY t ADMINISTRATIVE (continued) ! I A.3 " Immovable, as a result of excessive friction or mechanical ! interference, or known to be untrippable" has been replaced I with the term " stuck." The intent of the-existing wording is I consistent with the proposed simplification. Details of I potential mechanisms, mechanical or electrical, by which , control rods may be " stuck" are not necessary for inclusion ! within the ACTION. t i A.4 This comment number is not used for this station. , i A5 Due to the proposed changes to the requirements for inoperable , control rods, there are no withdrawn OPERABLE control rods ' required to have directional control valves disarmed. l Inoperable control rods are not required to meet this ! Surveillance (per existing 4.0.3, and proposed SR 3.0.1), and OPERABLE control rods will not be required to have their directional control valves disarmed. Therefore, this proposed t change reflects a deletion of a unusable allowance. A.6 These listed surveillances are required' by other l Specifications. Repeating a requirement to perform them is not ' necessary. Elimination of this " cross-reference" is therefore j administrative. l A.7 The phrase " actual or simulated" is added to clarify that ; either type of scram signal can fulfill the . - Surveillance -{ Requirement. This allows satisfactory automatic system j initiations for other than surveillance purposes to be used to ; fulfill the Surveillance Requirements. OPERABILITY is ; adequately demonstrated in either case'since the RPS can not { discriminate between an " actual" or " simulated" signal.
'l A.8 ACTION b.3 discussing the applicability of Specification 3.0.4 j is being deleted. The new wording of LCO 3.0.4 along with the j new ACTIONS have negated the need for this statement. Since i Conditions A, C, and D do not require leaving the applicable MODES while in these ACTIONS, the wording of LCO 3.0.4'would- i permit MODE changes while in these Conditions. Therefore, the i need for an exception from the applicability of Specification 1 3.0.4 is not required. l r
-)
A.9 This comment number is not used for this station PERRY - UNIT 1 10 10/1/93 I
DISCUSSION OF CHANGES CTS: 3.1.3.1 - CONTROL ROD OPERABILITY RELOCATED SPECIFICATIONS
- ]
R.1 Testing of the SDV level scram instrumentation is duplicated in j existing surveillance 4.3.1.1. Surveillance 4.1.3.1.4.b for i the scram function is proposed to be moved to the RPS i specification (3.3.1.1). The Control Rod Block function of the f Scram Discharge Level Instrumentation is relocated to plant ; controlled procedures in accordance with the " Application of i Selection Criteria to the PNPP TS" and the NRC Interim Policy i Statement. Refer to the application document discussion of l TS 3/4.3.6.5 for additional information. : TECHNICAL CHANGE - MORE RESTRICTIVE _j M.1 The existing ACTIONS would appear to require LCO 3.0.3 entry if [ more than one control rod were stuck. The proposed Condition B ; Required Actions maintain the equivalent shutdown ACTION as [ LCO 3.0.3, but also contain an additional requirement to disarm the stuck control rod. This additional requirement provides a i level of protection to the control rod drive. should a scram ; signal occur. If mechanically bound, the stuck control rod I could cause further damage if not disarmed. Disarming normally i could preclude control rod insertion on a scram signal, ! however, since this control rod is stuck, this effect of ! disarming is moot. l 1 M.2 Condition D of the proposed LCO 3.1.3 applies to all inoperable 1 control rods whether inserted or withdrawn. {; M.3 The proposed changes to Actions for non-stuck inoperable ! control rods eliminates the check of insertion capability; -l replacing it with a requirement to fully insert and disarm all. .l inoperable control rods. The existing Action, requiring the ; insertion capability to be verified and then allowing .the ; control rod to remain withdrawn, is currently applicable to i conditions such as: 1) " slow" control rods; 2) one inoperable 1 CRD accumulator; 3) and loss of position indication while below l the low power setpoint. The first two of these conditions are i addressed later in comments for LCO 3.1.3.2 and LCO 3.1. 3. 3, -l respectively. .i The latter condition would no longer allow the affected control } rod to remain withdrawn and not disarmed. This added l restriction on control rod (s) with loss of position indication; ! is conservative with respect to scram' time and SDM. Actions a for inoperable control rods not complying.with BPWS (proposed j Condition D) assure that insertion of these control rods remain i appropriately controlled. ! PERRY - UNIT 1 Il 10/1/93 *
.i'
r-b f i DISCUSSION OF CHANGES CTS: 3.1.3.1 - CONTROL ROD OPERABILITY j k TECHNICAL CHANGE - MORE RESTRICTIVE (continued) M.4 Proposed surveillances SR 3.1.3.2 and SR 3.1.3.3 require control rods to " insert" in lieu of the existing requirement [ for " moving. " The existing requirement could be met by control ! rod withdrawal. It is conceivable that a mechanism causing j binding of the control rod that prevents insertion'could exist ! such that a withdrawal test would not detect the problem. l Since the purpose of the test is to assure scram insertion } capability, restricting the test to only allow- control rod . insertion provides an increased likelihood of this test : detecting a problem that impacts this capability. ; I TECHNICAL CHANGE - LESS RESTRICTIVE !
" Generic" l l
LA.1 Details of the methods of disarming control rod drive (s) (CRD) ; are relocated to the Bases and procedures. The requirement to i disarm the CRDs remains in the Specification. i i
" Specific" {
L.1 Proposed Condition D provides the requirements and actions for the local distribution of inoperable control rods. Three j distinct changes are addressed: ; i
- 1) Condition D is modified by a Note excluding its !
applicability-above 20% power. The existing separation ! requirements for a stuck control rod, in part, accounted ! for allowing withdrawn inoperable control rods (refer to ! comment M.3 above). To preserve scram reactivity, a stuck ! rod must be separated from other withdrawn inoperable l control rods which may also not scram. In the proposed ! Technical Specifications, all inoperable control rods ! which would not scram or could not be verified to scram ! (e.g., loss of position indication) are required to be j fully inserted and therefore can not impact scram ! reactivity. The local distribution of withdrawn " slow" control rods is addressed in proposed LCO 3.1.4.
- Therefore, scram reactivity remains preserved at all power
levels, and is unaffected by.this proposed change. :
.(
Separation requirements are required when below 20% power j because of Control Rod Drop Accident (CRDA) concerns j related to control rod reactivity worth. Above 20% power, l control rod worth that is of concern for the CRDA is not possible. Further, the ef fects of inoperable control rods j in close proximity are adequately controlled- by the j l PERRY - UNIT 1 12 10/1/93 ! i
C DISCUSSION OF CHANGES CTS: 3.1.3.1 - CONTROL ROD OPERABILITY TECHNICAL CHANGE - LESS RESTRICTIVE (continued) requirements f or monitoring the fuel thermal limits, e.g. , MCPR, APLHGR and LHGR, above 25% power. Between 20% and 25% power, sufficient margin exists for adequate protection. Therefore, adequate limits to control core reactivity and power distribution above 20% power remain with this proposed change.
- 2) Condition D also does not require ACTIONS for inoperable control rods whose position is in conformance with BPWS constraints, even if the inoperable control rods are within two cells of each other. As discussed above in the first item of this change, adequate limits to control core reactivity and power distribution above 20% power remain with this proposed change. Below 20% power the appropriate core reactivity and power distribution limits.
are controlled by maintaining control rod positions within the limits of BPWS and maintaining scram times within the limits of existing Specification 3.1.3.2 (as modified to reflect proposed LCO 3.1.4). If the two inoperable control rods were both " stuck," ACTIONS require an immediate shutdown - regardless of their proximity. Therefore, the limitation on the local distribution of inoperable control rods that comply with BPWS, is either overly restrictive, or adequately controlled by the distribution restrictions for " slow" control rods.
- 3) Finally, the Required Actions for Condition D allow 4 hours to correct the situation prior to commencing a required shutdown, while the existing ACTION a.1 allows 1 hour. This increase is proposed in recognition of the actual operational steps involved on discovery of inoperable control rod (s). Time is first required to attempt identification and correction of the problem.
Additional time is then necessary to fully insert (some operational considerations may be necessary to adjust control rod patterns and/or power levels), and then disarm the af fected control rod (s) . After these high priority steps are accomplished, then attention can be turned to correcting localized distributions of inoperable control rods that deviate f rom BPWS. Given the low probability of a CRDA durine this brief proposed time extension, and the desire not .o impose excessive time constraints on operator actions that could lead to hasty corrective actions, the proposed extension to this ACTION is deemed acceptable. PERRY - UNIT 1 13 10/1/93
o I DISCUSSION OF CHANGES CTS: 3.1.3.1 - CONTROL ROD OPERABILITY TECHNICAL CHANGE - LESS RESTRICTIVE i (continued) L.2 Disarming a control rod involves personnel actions by other than control room operating personnel, with the necessary activity being within the Primary Containment. These processes i require coordination of personnel and preparation of equipment, and potentially require anti-contamination " dress-out" and access into the Primary Containment, in additicn to the actual procedure of disarming the control rod. Currently, all these ; activities must be completed and the control room personnel must confirm completion within one hour. This is proposed to be extended to two hours (consistent with the BWR Standard Technical Specification, NUREG-1434) in recognition of the potential for excessive haste required to complete this task. The proposed two hour time does not represent a significant : safety concern as the control rod is already in an acceptable position (in accordance with other ACTIONS), and'the ACTION to disarm is solely a mechanism for precluding the potential for , damage to the CRD mechanism. . L.3 The requirement to demonstrate SHUTDOWN MARGIN with a control rod stuck is proposed as Required Action A.3. With a single control rod stuck in a withdrawn position, the remaining , OPERABLE control *ods are capable of providing the required l scram and shutdown reactivity. Failure to reach COLD SHUTDOWN ' 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 postulated additional single failure of ' an adjacent control rod to insert, sufficient reactivity control i remains to reach and maintain HOT SHUTDOWN conditions. Required Action A.2 of LCO 3.1.3 performs a notch test on each t remaining withdrawn control rod to ensure that no additional i control rods are stuck. Given these considerations, the time i allowed to demonstrate SHUTDOWN MARGIN has been extended from + 12 hours to 72 hours to allow a reasonable time to perform the { analysis or test. L.4 It is proposed to remove the upper limit on restoration time ! for one stuck control rod. By deleting existing ACTION a.3, . ; continued operation with a stuck control rod may be allowed. ' With a single withdrawn control rod stuck, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. The assumptions utilized in - establishing the proposed scram time limits account for a single stuck control rod in. addition to an assumed single failure during a transient. SHUTDOWN MARGIN remains required i to be met, accounting for the loss of negative reactivity due ; to the stuck control rod (refer to the proposed definition of SDM and proposed Required Action A.3 of LCO 3.1.3) . Given that PERRY - UNIT 1 14 10/1/93 ! i
1 DISCUSSION OF CHANGES l CTS: 3.1.3.1 - CONTROL ROD OPERABILITY ; TECHNICAL CHANGE - LESS RESTRICTIVE (continued) 1 operation remains within the bounds of analyzed events, all , remaining limitations continue to be required, and prompt j action is required to confirm no additional stuck control rods ; exist, continued operation is proposed to be allowed, as are i MODE changes in accordance with SR 3.0.4. ! I L.5 All inoperable non-stuck control rods are required to be fully ; inserted and disarmed (refer to comment M.3 above). The time - :i allowed to complete the insertion is proposed to be extended to I three hours for all cases. In the existing ACTIONS for an uncoupled control rod (LCO 3.1.3.4, ACTION a.2) two hours are : allowed before entering the LCO 3.1.3.1 ACTION b.1 which then l gives an additional hour to insert and disarm the control rod l (total of three hours to insert and disarm) . Uncoupled' control ! rod ACTIONS are proposed to be addressed by LCO 3.1.3 ! Condition C, as are other non-stuck inoperable control rod ACTIONS. This existing three hour allowance, before requiring , an inoperable (uncoupled) control rod to be inserted, is the ! time found in the proposed Required Action C.1 for control rod ' in;ertion. For consistency of presentation, this three hour + limitation is also proposed for all other instances of : inoperable control rods. These other instances (loss of i position indication, exceseive scram time, certain combinations ! of conditions with a low pressure on a control rod scram i accumulator) also warrant a minimal time to attempt restoration prior to inserting and disarming. It is for these other i instances that the extended time to insert are proposed. Given i that these instances do not represent loss of SDM,. and are ; limited to a total of no more than 8 inoperable control rods = (refer to proposed Condition E) , the extended, time does not represent a cianificant safety concern. t Disarming a control rod involves personnel actions by other i than control room operating personnel, with the necessary < activity being within the Primary Containment. These processes 4 require coordination of personnel and preparation of equipment. - and potentially require anti-contamination " dress-out" and , access into the Primary Containment, in addition to the actual j procedure of disarming the control rod. Currently, all these activities must be completed and the control room personnel , must confirm completion within the same one hour allowed to ' insert the control rod. The disarming is proposed to be ; extended to four hours; one hour beyond that allowed to insert ; (consistent with the BWR Standard Technical Specification, ! NUREG-1434) in recognition of the potential for excessive haste required to complete this task. The proposed four hour time : does not represent a significant safety concern as the control ; rod is already in its required position (in accordance with i f PERRY - UNIT 1 15 10/1/93 l t i r p
F DISCUSSION OF CHANGES CTS: 3.1.3.1 - CONTROL ROD OPERABILITY TECHNICAL CHANGE - LESS RESTRICTIVE (continued) does not represent a significant safety concern as the control rod is already in its required position (in accordance.with other ACTIONS), and the ACTION to disarm is solely a mechanism for precluding the potential for future mis-operation. L.6 The SDV vent and drain valve requirements are proposed to be moved to LCO 3.1.8. The SDV vent and drain valves primary safety functions are to-maintain the scram discharge volume with sufficient capacity to accept discharge water following a scram signal, and to isolate the SDV during a scram to contain the reactor coolant discharge. The isolation function can still be satisfied if at least one valve is OPERABLE in.each line or the line is isolated. However, this action could prevent proper venting and draining of the SDV under normal conditions. Therefore, a note has been added to periodically permit opening the affected line for draining and venting. The following ACTIONS are proposed to be modified:
- 1) Allow 7 days to restore inoperable SDV vent or drain valves provided at least one valve in each line is OPERABLE or the line is isolated.
- 2) The 8 hour limit with both valves in a line inoperable, is proposed to require isolation of that line in this time concurrent with the 7 day limit to restore both valves to OPERABLE status.
- 3) Recognizing that the SDV vent and drain valves are normally open to prevent accumulation of water in the SDV from leakage, a Note has been added to the ACTIONS that require isolation of the line, to allow periodic opening of the affected line for draining and venting of the SDV.
This will be necessary to avoid automatic reactor scrams on high level in the SDV. These extended times, and the option to administratively unisolate a SDV line isolated to comply with a Required Action, are consistent with the BWR Standard Technical Specification, NUREG-1434. These increased allowances are deemed not to substantially increase the risk of a scram with an additional failure that could allow the SDV to remain unisolated; nor to substantially increase the risk of the SDV failing to accept the control rod drive water displaced-during a scram. PERRY - UNIT 1 16 10/1/93
e , vf DISCUSSION OF CHANGES CTS: 3.1.3.1 - CONTROL ROD OPERABILITY l TECHNICAL CHANGE - LESS RESTRICTIVE , I (continued) L.7 The surveillance on control rods to verify them to be non-stuck i is proposed to be extended from 7 days to 31 days for control ' rods that are not fully withdrawn, consistent with the BWR-Standard Technical Specification, NUREG-1434. Partially withdrawn control rods have a significantly greater affect on core flux distribution than do fully withdrawn control rods, i Historically, power reductions are required each week to ' perform this test on these partially withdrawn control rods. This impact on plant capacity is deemed excessive given the , following considerations: .
- 1) At full power a large percentage of control rods ;
90%) are fully withdrawn and would i (typically 80 - continue to be exercised each week. This represents a ! significant sample size when looking for an unexpected ! random event.
- 2) Operating experience has shown " stuck" control rods to be i an extremely rare event while operating.
- 3) Should a stuck rod be discovered, 100% of the remaining control rods (even partially withdrawn) must be tested within 24 hours (proposed Required Action A.2).
L.8 This comment number is not used for this station. i L.9 After discovery of a stuck rod, all withdrawn control rods are required to be notch tested as per Required Action A.2'. This , provides adequate assurance that the cause of the stuck rod is -! not of generic concern. Continued testing of control rods per the normal Frequency is sufficient to ensure continued OPERABILITY of the remaining control rods and therefore the 24 hour Frequency is only required once. P PERRY - UNIT 1 17 10/1/93-
DISCUSSION OF CHANGES CTS: 3.1.3.2 - CONTROL ROD MAXIMUM SCRAM INSERTION TIMES ADMINISTRATIVE A.1 The organization of the Scram Insertion Time Specification is proposed to be modified as follows:
- 1) Specific conditions with " slow" control rods, which allow unrestricted operation, are relocated to the LCO.
statement. This administratively allows operation to continue without the label of also having an "LCO not met." No functional difference is created with this proposed change.
- 2) The category of " fast" control rods is eliminated, as is any requirement on average scram times, thus simplifying the presentation. (In the proposed change, the LCO limit becomes the existing " fast" average limit. Tnis additional restriction on the LCO scram time is offset by a change in the allowed number of " slow" control rods, and in elimination of any average scram time requirement.
These changes are specifically the subject of Less Restrictive Change "L.1")
- 3) The LCO scram time limits are presented in proposed Table 3.1.4-1.
A.2 Note 2 of the proposed Table 3.1.4-1 is equivalent to existing ACTION b.1 (" Declare the " slow" control rod (s) inoperable"). However, the proposed Specification does not contain the equivalent " default" ACTION ("Otherwise, be in at least HOT SHUTDOWN") for failures of the existing b.1 ACTION. There are no circumstances which would preclude the possibility of compliance with an ACTI')N to " Declare the " slow" control rod (s) . inoperable." Omission of this " default" ACTION is therefore inconsequential. A.3 The "or" in the identified SR Frequency is revised to "AND" to 4 clarify that this not a " choice" but is required following either condition. A.4 ACTION d discussing the applicability of Specification 3.0.4 is being deleted. The new wording of LCO 3.0.4 along with the new wording of LCO 3.1.4 have negated the need for this statement. Since the LCO ACTIONS permit continuous operation of the plant, the wording of LCO 3.0.4 would. permit MODE changes while in these Conditions. Therefore, the need for an exception from the applicability of Specification 3.0.4 is not required. RELOCATED SPECIFICATIONS None in this sectica. PERRY - UNIT 1 18 10/1/93
i y i DISCUSSION OF CHANGES CTS: 3.1.3.2 - CONTROL ROD MAXIMUM SCRAM INSERTION TIMES I TECHNICAL CHANGE - MORE RESTRICTIVE .! M.1 The Surveillance Requirement "for specifically affected" CRDs. ! is proposed to have the flexibility (provided- by current l footnote *) to delay post maintenance testing until " prior to I entry into OPERATIONAL CONDITION.1," deleted. This proposed- , modification is to ensure adequate testing is performed. prior ! to declaring the control rod OPERABLE, and entering MODE 2. In . support of this. proposed additional restriction, along with ! deleting the existing flexibility, an_ additional surveillance -l is proposed (SR 3.1.4.3). This new surveillance will. require
. j a scram time test, which may be done at any reactor pressure, :
prior to declaring the control rod OPERABLE-(and thus, enabling { its withdrawal during a startup). { To allow testing at less than normal operating pressures, additional scram time limits will be contained in plant ; procedures. These limits are reasonable for application as a .i s test of OPERABILITY at these conditions. -Since this test, and i therefore any limits, are not applied in the existing i Specification, any value could be construed as being more
- restrictive. Furthermore, the existing scram time test ~ ;
requirement (performed at normal operating reactor pressure) is: ; still required to be performed prior to exceeding 40% power. l It is noted that if the control rod remained inoperable (which f would require it to be inserted and disarmed) until ~ normal operating pressures, a single scram time test would~ satisfy both Surveillance Requirements. i TECHNICAL CHANGE - LESS RESTRICTIVE l
" Generic" LA.1 A " representative sample" of control rods is required to be ,
tested each 120 days of power operation (existing Surveillance l 4.1.3.2.c). Additionally, a limit 'on the number of . " slow - i 3 control rods in this sample is imposed (existing ACTION c.1). 1
; The existing Surveillance and ACTION specifically delineate the. 1 scope of this sample and the statistical . limit for " slow" !
control rods in the sample. The proposed change adopts the BWR ! Standard Technical Specification, NUREG-1434, position that t these details be located within plant procedures and summarized 1 it the Bases for the surveillance. ; LA.2 This comment number is not used for this station. . d e r i PERRY - UNIT 1 19 10/1/93 j i
DISCUSSION OF CHANGES [ CTS: 3.1.3.2 - CONTROL ROD MAXIMUM SCRAM INSERTION TIMES f TECHNICAL CHANGE - LESS RESTRICTIVE (continued) ,
" Specific" L.1 The Design Basis Accident (DBA) and transient analyses simplistically assume all of the control rods scram at a i specified insertion rate. The resulting negative scram I reactivity forms the basis for the determination of plant thermal limits (e.g., the MCPR). Other distributions of scram l times (e.g., several control rods scramming slower than the i' average time, with several control rods scramming faster than the average time) can also provide sufficient scram reactivity !
within the assumptions of the analyses. To account for single failure and to allow a certain number of ,
" slow" scramming control rods, the scram times proposed in l Table 3.1.4-1 are faster than those assumed in the design basis >
analysis. The scram times have sufficient margin to those assumed in the analyses to allow up to 7.5% of the control rods t (e.g., 177 x 7.5% = 13) to have scram. times that exceed the specified limits (i.e., assumed np1 to scram; however they are i assumed to insert to provide long' term decay heat removal - these insertion times would be on the order of minutes or hours i versus seconds). Scram times exceeding . the LCO limit are referred to as " slow" control rods. Furthermore, the proposed limits have additional margin to account for a single stuck i control rod allowed by LCO 3.1.3, Condition A, and an additional control rod failing to scram per the single failure ! i criterion (total of two stuck control rods). i Changes to the existing Actions follow the above discussion. Specifically: - 1) ACTION a.1 - maximum limits on " slow" control rods, is j relaxed to 7 seconds for notch position 13, with no i intermediate limits for other notch positions. This ! proposed limit is found in Table 3.1.4-1 Note 2, and in l SR 3.1.3.4 of LCO 3.1.3. As discussed above, these " slow" l control rods are not assumed to scram. The 7 second limit 1 is chosen based on historical use of this value (from the l BWR/4 STS), which adequately bounds assumptions for long j term reactivity control. , I
- 2) ACTION a.2 -
average scram times are proposed to be :1 eliminated. The proposed LCO ]imit' is consistent with the existing " average" values. This particular change is more restrictive in that the existing average limit is more restrictive than the existing LCO limit. PERRY - UNIT 1 20 10/1/93
DISCUSSION OF CHANGES CTS: 3.1.3.2 - CONTROL ROD MAXIMUM SCRAM INSERTION TIMES TECHNICAL CHANGE - LESS RESTRICTIVE (continued)
- 3) ACTIONS a.3 and c.4 - the number of control rods exceeding the LCO limit (which is the equivalent of the existing
" fast" and " slow" control rods) is increased to 13 (from 7). The overall core scram reactivity remains.
consistent with that assumed in analyses. The allowed number of slow control rods is plant specific and is 7.5% of the total number of control rods. This value is higher than currently allowed (4%) because withdrawn inoperable control rods are no longer allowed (refer to item 5 below) .
- 4) ACTIONS b and c.2 -
excessively " slow" control rods (proposed to be control rods with scram times in excess of 7 seconds) are declared inoperable. The proposed actions for inoperable control rods (refer to comment M.3 to , existing LCO 3.1.3.1) require these excessive " slow" control rods to be fully inserted and disarmed, while the existing actions would allow these control' rods to remain - withdrawn. While the criteria for excessively " slow" is relaxed (from values in existing ACTION a.1, to 7 seconds), the accident analysis assumptions are still met. L.2 The requirement for increasing scram time surveillance testing when more than three control rods are " slow" is proposed to be deleted. During normal power operating conditions, scram , testing is a significant perturbation to steady state operation; involving significant power reductions, abnormal control rod patterns and abnormal control rod drive hydraulic system configurations. Requiring more frequent scram time-surveillance tests is therefore not desirable. Because of the frequent testing of control rod insertion capability-(SR 3.1.3.2 and SR 3.1.3.3) and accumulator OPERABILITY (SR 3.1.5.1), and the operating history demonstrating a high degree of reliability, the more frequent scram time testing is not deemed necessary to assure safe plant operation. L.2 Scram insertion time data exists from previous testing for control rods surrounding a " slow" control rod. No additional testing should be required to determine if an adjacent " slow" drive exists provided the results of the current statistical sampling of scram times produces acceptable results _ (i.e. , ' criteria of existing ACTION c.1 met). The limitations on number and distribution of " slow" rods are deemed stringent enough to assure that severe degradation of scram performance does not exist. Because of the frequent testing of control rod insertion capability (SR 3.1.3.2 and SR 3.1.3.3) and accumulator OPERABILITY (SR 3.1.5.1) , and the operating history demonstrating a high degree of reliability, the more frequent scram time testing is not deemed necessary to assure safe plant operation. ,
^
PERRY - UNIT 1 21 10/1/93
II i, i i DISCUSSION OF CHANGES ; CTS: 3.1.3.2 - CONTROL ROD MAXIMUM SCRAM INSERTION TIMES TECHNICAL CHANGE - LESS RESTRICTIVE l (continued) , t L.4 The Surveillance Frequency has been modified to require testing after fuel movement within the reactor vessel. This is i equivalent to CORE ALTERATIONS excluding normal control rod i movement. Normal control rod movement with the vessel head detensioned or removed however, is not any different than j nonnal control rod movement during power operations. This ! activity would not be expected to result in any affect on scram i speed. Therefore, no increased frequency of scram time test
- performance is necessary simply due to normal control rod I movement while the vessel head is detensioned. ;
I i i i
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I 1 l l . PERRY - UNIT I 22 10/I/93 l
= DISCUSSION OF CHANGES CTS: 3.1.3.3 - CONTROL ROD SCRAM ACCUMULATORS ADMINISTRATIVE A.1 In general, the outline for the format and editorial changes proposed in this Specification consist of rewriting the existing two ACTIONS for operating modes (a.1 and a.2) as three Conditions (A, B and C): Existing ACTION a.1 for one inoperable accumulator at any reactor pressure, becomes Condition A for one inoperable accumulator at reactor pressures which will support control rod insertion, and Condition C for ' reactor pressures which will not support insertion. Existing ACTION a.2 for two or more inoperable accumulators is split into Condition B and C; which are dependent on reactor pressure. A.1 This proposed Note (" Separate Condition entry is allowed for each control rod scram accumulator") provides more explicit instructions for proper application of the ACTIONS for Technical Specification compliance. In conjunction with the-proposed Specification 1.3 - " Completion Times," this Note provides direction consistent with the intent of the existing ACTIONS for inoperable control rod accumulators. Upon discovery of each inoperable accumulator, it is intended that each specified ACTION be applied regardless of it having been applied previously for other inoperable accumulators. A.3 The revised presentation of ACTIONS (based on the BWR Standard Technical Specification, NUREG-1434) is proposed to not 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 The proposed Specification does not contain the equivalent "def ault" ACTION ("Otherwise, be in at least HOT SHUTDOWN") for failures of the existing ACTION a.1. There are no circumstances which would preclude the possibility of compliance with an ACTION to " Declare the control rod... inoperable." Omission of this " default" ACTION is therefore inconsequential. A.5 The existing ACTION a.2 requires the affected control rod to be declared inoperable. Once declared inoperable, the TlTIONS for an inoperable control rod are applied, which contain the requirement to insert and disarm, as well as a shutdown requirement if this ACTION is not performed. The proposed ACTIONS for inoperable accumulators do not repeat the ACTION to insert and disarm, or the shutdown requirement for failure to perform this ACTION. A.6 This comment number is not used for this station. PERRY - UNIT 1 23 10/1/93
n i DISCUSSION OF CHANGES l CTS: 3.1.3.3 - CONTROL ROD SCRAM ACCUMULATORS r ADMINISTRATIVE i (continued) A.7 St3 ting the conditions for an exception to performance of the accumulator surveillance which are equivalent to the actions , required if the accumulator is inoperable, is unnecessary. If . the accumulator is inoperable, existing Specification 4.0.3 (proposed SR 3. 0.1) states that surveillances are not required to be performed. . A.8 The technical content of this requirement is being moved to ; another chapter of the proposed Technical Specifications. Any technical changes to this requirement will be addressed with -l the content of the proposed chapter location. A.9 ACTION c discussing the applicability of Specification' 3.0.4 is i being deleted. The new wording of LCO 3.0.4 along with the new [ ACTIONS have negated the need for this statement. Since t Conditions A, B, and C permit continuous operation of the plant ! while in these ACTIONS, the wording of LCO 3.0.4 would permi: MODE changes while in these Conditions. Therefore, the need' ! for an exception from the applicability of Specification 3'.0.4 ! is not required. ; A.10 The requirement to verify that a control rod drive pump is [ operating has been maintained, but the method of verifying this j has changed from inserting one control rod one notch to ! verifying that charging water header pressure is at least 1520 [ psig. These methods both assure that sufficient control rod ' drive pressure exists to insert control rods. The proposed ! method of determining charging water header pressure provides ! added assurance that the charging header pressure is sufficient - to drive all rods, whereas the existing method only assures , that one rod is still capable of insertion. Since the change is ! merely exchanging one test method for another equivalent (or better) test method, this change is considered administrative. ! RELOCATED SPECIFICATIONS j None in this Section. I PERRY - UNIT 1 24 10/1/93 ___ _a
- . ~
f-; .1 DISCUSSION OF CHANGES CTS: 3.1.3.3 - CONTROL ROD SCRAM ACCUMULATORS l TECHNICAL CHANGE - MORE RESTRICTIVE M.1 The proposed ACTIONS for inoperable control rod accumulators ; provides an 8 hour allowance when a single accumulator is inoperable, but'onlv if the reactor pressure is sufficiently I high to support control rod insertion. Existing Action a.1 l would allow 8 hours for one accumulator at any reactor ' pressure. At reduced reactor pressures, control rods may not insert on a scram signal unless the associated accumulator is ; OPERABLE. Given the allowances in the proposed LCOs 3.1.3 and ; 3.1.4 for number and distribution of inoperable and " slow" l control rods, an additional control rod failing to scram (due ! to inoperable accumulator and low reactor pressure) for up to 8 hours with out compensatory action is not justified. l Therefore, existing ACTION a.1 is reflected in proposed i Condition A for one inoperable accumulator during sufficiently high reactor pressures, and existing ACTIONS a.1 and a.2 are reflected in Condition C for lower reactor pressures for one or , l more inoperable accumulators. _ TECHNICAL CHANGE - LESS RESTRICTIVE !
" Generic" LA.1 Details of the methods of disarming control rod drive (s) (CRD)~ ~
are relocated to the Bases and procedures. The requirement to' disann the CRDr remains in the Specification. LA.2 This comment number is not used for this station. : LC.1 The scram accumulator leak detectors, pressure detectors, and associated alarm do not necessarily. . relate directly to accumulator OPERABILITY. In general the BWR Standard Technical Specification, NUREG-1434, does not specify indication-only or ; test equipment to be OPERABLE to support OPERABILITY of a ' system or component. Control of the availability of , and necessary compensatory activities if not available, for - indications, monitoring instruments, alarms, and test equipment are addressed by plant operational procedures and policies. Therefore, the control rod accumulator leak detectors, precsure detectors, and alarm surveillances are removed from the : Technical Specification. ! PERRY - UNIT 1 25 10/1/93 5 i
DISCUSSION OF CHANGES CTS: 3.1.3.3 - CONTROL ROD SCRAM ACCUMULATORS TECHNICAL CHANGE - LESS RESTRICTIVE (continued) " Specific" L.1 An inoperable control rod accumulator affects the associated control rod scram time. However, at sufficiently high reactor pressure, the accumulators only provide a portion of the scram force. With this reactor pressure, the control rod will scram , even without the associated accumulator, although probably not within the required scram times. Therefore, the option to declare a control rod with an inoperable accumulator " slow" when reactor pressure is sufficient is proposed. Since the existing ACTION to declare the control rod inoperable would allow the control rod to remain withdrawn and not disarmed, the 6 proposed ACTION to declare the control rod " slow" is > essentially equivalent. The proposed limits and allowances for numbers and distribution of inoperable and " slow" control rods (found in proposed LCOs 3.1.3 and 3.1.4, respectively) are appropriately applied to control rods with inoperable accumulators whether declared inoperable or " slow. " The option for declaring the control rod with an inoperable accumulator .;
" slow" is restricted (by a Note to Required Action A.1 and B.2.1) to control rods that were not previously known to be
" slow." This restriction limits the flexibility to control rods not otherwise known to have an impaired scram capability. !
Additionally, the requirement for declaration of " slow" or ! inoperable (and the implied concurrent restoration allowed > time) is extended to 1 hour in proposed Required Actions B.2.1 and B.2.2. This provides a reasonable time to attempt investigation and restoration of the inoperable accumulator. ! Furthermore, the ACTIONS have been " human factored" to present the more critical ACTION, i.e., with a shorter allowed
- completion time, first. Proposed Required Action B.1 addresses the situation where additional accumulators may be rapidly becoming inoperable due to loss of charging pressure. Once verification of adequate charging pressure is made, and considering that reactor pressure is adequate to assure the '
scram function of the control rods with inoperable accumulators, the proposed 1 hour extension is not significant. L.2 This comment number is not used for this station. , l i l PERRY - UNIT 1 26 10/1/93 i l l
i DISCUSSION OF CHANGES i CTS: 3.1.3.3 - CONTROL ROD SCRAM ACCUMULATORS l TECHNICAL' CHANGE - LESS RESTRICTIVE (continued) L.3 As indicated in USAR Sections 4.6.1.1.2.5.3 and 4.6.2.3.2.3, 600 psig reactor pressure is sufficient to ensure that the ; control rods will insert without supporting accumulator ; pressure. However, as indicated, the insertion times may be ; slower and may not support the' required scram times used in the 1 safety analysis. Since this is the identified. basis for the bracketed pressure to be used in the proposed LCO 3.1.5, this ! proposed pressure is in accordance with the approved safety analysis. r s i i F i a 5 P PERRY - UNIT 1 27 10/1/93
DISCUSSION OF CHANGES CTS: 3.1.3.4 - CONTROL ROD DRIVE COUPLING ADMINISTRATIVE A.1 The requirement that control rods be coupled to their drive mechanism is presented in SR 3.1.3.5, making it a requirement for control rods to be considered OPERABLE. The ACTIONS for uncoupled control rods remain effective (as commented on below). Eliminating the separate LCO for control rod coupling, by moving the Surveillance and ACTIONS to another Specification, does not eliminate any requirements, or impose-a new or different treatment of the requirements (other than those separately proposed) . Therefore, this proposed change is considered administrative. A.2 The time allowed for this particular portion of the ACTION (" insert the control rod, and disarm") is not clearly presented in the existing Specification. It follows an ACTION. to
" declare the control rod inoperable" and has been interpreted to give direction as to the proper option for inoperable control rod ACTIONS (refer to existing LCO 3.1.3.1, ACTION b.1 for two options) . Those ACTIONS for inoperable control rods in-LCO 3.1.3.1 allow one hour from the time the control rod is declared inoperable, until it is required to be inserted and disarmed. Therefore, the intent of the existing requirement is deemed to allow a total of three hours (two hours in LCO 3.1.3.4 ACTION a., plus the one hour discussed here) . This three hours is consistent with the BWR Standard Technical Specification, NUREG-1434, and the proposed ACTIONS, for inoperable control rods (and therefore uncoupled control rods) .
As an enhanced presentation of the existing intent, the proposed change is deemed to be administrative. A.3 Existing Surveillance 4.1.3.4.c addresses the requirement to perform coupling checks af ter performing activities which could , have affected coupling integrity. This surveillance must be
- completed prior to allowing the control rod to be considered OPERABLE (the consideration of OPERABILITY- is more clearly presented in the proposed editorial rewrite of 4.1.3.4.c into :
the Frequency for proposed SR 3.1.3.5). Therefore, the ! existing Surveillance 4.1.3.4.a is redundant. -"CORE i ALTERATIONS that could have affected the control rod drive ; coupling integrity" is a subset of the 4.1.3.4.c requirement !
" maintenance...which could have affected the control rod drive !
coupling integrity. " Performance of the integrity verification .; prior to control rod OPERABILITY (which is the understanding of i 4.1.3.4 as presented in the proposed SR 3.1.3.5) bounds " prior to reactor criticality. " Elimination of 4.1.3.4.a is therefore administrative, and represents no actual change in i requirements. j i I I PERRY - UNIT 1 28 10/3/93 I
.q f
i i DISCUSSION OF CHANGES CTS: 3.1.3.4 - CONTROL ROD DRIVE COUPLING ' ADMINISTRATIVE l (continued) . A.4 ACTION c discussing the applicability of Specification 3.0.4 is : being deleted. The new wording of LCO 3.0.4 along with the new i ACTIONS have negated the need for this statement. Since . Conditien C of LCO 3.1.3 permits continuous operation of the i plant while in this ACTION, the wording of LCO 3.0.4 would' ! permit MODE changes while in this Condition. Therefore, the ; need for an exception from the applicability of Specification i 3.0.4 is not required. i i RELOCATED SPECIFICATIONS None in this section. t TECHNICAL CHANGE - MORE RESTRICTIVE ' P None in this section. TECHNICAL CHANGE - LESS RESTRICTIVE i
" Generic" ,
LA.1 (Refer also to comment A.1 for this LCO. ) Exista . IION a.1 contains detailed methods of restoring coupling intcy 2 y to an - uncoupled control rod. The revised presentation of ACTIONS (based on the BWR Standard Technical Specification, NUREG-1434) is proposed to not explicitly detail options to " restore...to j OPERABLE status." This ACTION is always an option, and is implied in all Conditions. Omitting this ACTION is purely ; editorial. The details contained within the existing actions are located in plant procedures. Furthermore, the specific detailed ACTION requiring the control , rod to be declared inoperable (existing ACTION a.2) is < addressed by the requirement of SR 3.0.1 on the proposed - SR 3.1.3.5. SR 3.0.1 requires that " failure to meet a . Surveillance ...shall be failure to meet the LCO." In the l proposed presentation, failure to meet the LCO ("LCO 3.1.3 ; Each control rod shall be OPERABLE") , results in the control
- rod being considered inoperable. i LA.2 Details of the methods of disarming control rod (s) are l relocated to the Bases and procedures. i LA.3 Details of the methods of verifying control rod coupling are relocated to plant procedures. l 6
PERRY - UNIT 1 29 10/1/93 .
i: DISCUSSION OF CHANGES CTS: 3.1.3.4 - CONTROL ROD DRIVE COUPLING i i TECIUJICAL CHANGE - LESS RESTRICTIVE (continued) ;
" Specific" L.1 Coupling requirements during refueling are not necessary since only one. control rod can be withdrawn from core cells '
containing fuel assemblies. The probability and consequences t of a single control rod dropping from its fully inserted position to the withdrawn position of the control rod drive are negligible (i.e., reactor _will remain subcritical). However, l these requirements are retained in proposed special operations of SDM testing in MODE 5 (Specification 3.10.8). 1. L.2 If an uncoupled control rod is not allowed by the applicable rod pattern control system to be inserted to accomplish recoupling, the current Technical Specifications require the ; control rod to be inserted. This may also require bypassing of : the applicable rod pattern control system and operation with an out-of-sequence control rod. Therefore, coupling attempts are allowed regardless of the condition cf the rod pattern control system because of the short time allowed. If coupling is not established in 3 hours, the control rod must be fully inserted { and then disarmed within the next hour (proposed Required Actions C.1 and C.2). Also, because of the limited time , i allowed to recouple, the number of attempts does not need to be restricted. The number of attempts to recouple a control rod may be restricted by plant procedures which consider the potential for equipment damage during successive recoupling attempts. i i' 0 r t PERRY - UNIT 1 30 10/1/93
DISCUSSION OF CHANGES . CTS: 3.1.3.5 - CONTROL ROD POSITION INDICATION l
,?
ADMINISTRATIVE A.1 The essence of the requirement that each control rod have at least one control rod position indication is presented in SR 3.1.3.1 of proposed LCO 3.1.3 " Control Rod OPERABILITY. " The effect of relocating the requirement for control rod position i to be indicated, is to make it a requirement for control rods : to be considered OPERABLE. Eliminating the separate LCO for control rod position indication (by moving the Surveillance and , ACTIONS to another Specification) does not eliminate . any requirements, or impose a new or different treatment. of the l requirements (other than those separately proposed). j A.2 The requirements for control rod position indication during refueling have been moved to LCO 3.9.4. The details of the , changes to the existing requirements will be discussed in comments for Chapter 3.9 " Refueling Operations." ! A.3 ACTION c discussing the applicability of Specification 3.0.4 is being deleted. The new wording of LCO 3.0.4 along with the new- ' ACTIONS have negated the need for this statement. Since Condition C of LCO 3.1.3 permits continuous operation of the - plant while in this ACTION th3 wording of LCO 3.0.4 would permit MODE changes while in this Condition. Therefore, the ! need for an exception from the applicability of Specification j 3.0.4 is not required. l RELOCATED SPECIFICATIONS None in this section. TECHNICAL CHANGE - MORE RESTRICTIVE { M.1 The existing ACTION for inoperable control rods (refer to LCO 3.1.3.1 ACTION b.), provides the option to verify the insertion capability, and then allows the control rod to remain withdrawn. The proposed changes to ACTIONS . for non-stuck inoperable control rods (refer to proposed LCO 3.1.3 Condition C) eliminates the check of insertion' capability; replacing it with a requirement to fully insert and disarm all inoperable control rods. These changes are discussed in the markup to existing LCO 3.1.3.1. The effect on the ACTIONS for control rods with position unknown, when below the low power setpoint of the RPCS, is to eliminate the option to leave'the ; control rod withdrawn and continue to operate. The control rod , will be required to be inserted and disarmed, regardless of the power level (this is currently the action if power is greater than the low power setpoint). 4 . PERRY - UNIT 1 31 10/1/93 l l 1
DISCUSSION OF CHANGES CTS: 3.1.3.5 - CONTROL ROD POSITION INDICATION ; TECHNICAL CHANGE - MORE RESTRICTIVE (continued) i M.2 Verification of the position and bypassing of control rods, j refers to an operation of the Rod Action Control System. 'The- ; bypassing activity relates to the function of this system to ; control rod movement and patterns via the Rod Withdrawal .; Limiter and the Rod Pattern Controller. These systems and the ! related control rod pattern control functions are addressed in- ' the proposed LCO 3.3.2.1. Specifically, proposed SR 3.3.2.1.9 contains the requirement of the existing ACTION a.2.a)2).
- However, this proposed surveillance requires the verification of position and bypassing be applied to any and all control rods bypassed; not just those control rods with inoperable '
" Full-in" and/or " Full-out" position indicators. The increased scope of this existing ACTION serves to provide the additional ;
level of safety obtained from the verification "by a second i licensed operator. . . " to all control rod bypassing operations. -! The relocation of the requirement to the Specification for the ' RPCS is administrative - allowing the requirements to be more closely associated'with the affected system. j b TECHNICAL CHANGE - LESS RESTRICTIV_g
" Generic" IA.1 The details of the methods for determining the position of. the control rod are proposed to be relocated to the Bases for the proposed Surveillance (SR 3.1. 3.1) and associated procedures.
The safety significant requirement remains within the Technical Specification - the position of the control rod must be known. LA.2 Details of the methods of disarming control rod (s) are relocated to the Bases and procedures. LA.3 To perform control rod movement tests (4.1.3.1.2) and coupling ! verification (4.1. 3. 4 .b) , position indication must be ! available. If position indication is not available, these tests can not be satisfied and appropriate ACTIONS will be > taken for inoperable control rods. Surveillance Requirement- , 4.1.3.5.b, therefore, provides details found within existing i procedures for the performance of the Surveillances referenced t therein. If at any time the position of a control rod-is t unknown, the control rod would be considered-inoperable, and. appropriate conservative ACTIONS taken. Therefore, relocating , the details of Surveillance 4.1.3.5.b does not significantly : affect safety. l 4 ! ~ PERRY - UNIT 1 32 10/1/93 I h
t t DISCUSSION OF CHANGES CTS: 3.1.3.5 - CONTROL ROD _ POSITION INDICATION , TECHNICAL CHANGE - LESS RESTRICTIVE , (continued) " Specific"
^
L.1 The time allowed for this particular portion of the ' ACTION (" insert the control rod, and disarm") is not clearly presented in the existing Specification. It follows an ACTION to
" declare the control rod inoperable" and has been interpreted r to give direction as to the proper option for inoperable control rod actions (refer to existing LCO. 3.1.3.1, ACTION b.1 for two options) . Those ACTIONS for inoperable control rods in LCO 3.1.3.1 allow one hour from the time the control rod is -
declared inoperable, until it is required to be inserted and disarmed. Therefore, the intent of the existing requirement is , deemed to allow a total of two hours (one hour in LCO 3.1.3.5
- ACTION a, plus the one hour discussed here).
Control rods whose position is unknown are proposed to be addressed by LCO 3.1.3 Condition C, as are other non-stuck . inoperable control rods. The existing 2 hour allowance, before requiring an inoperable (position unknown) control rod to be , inserted, is proposed to be extended to three hours (the time ' in the proposed LCO 3.1.3 Required Action- C.1). For ; consistency of presentation, this three hour limitation is also proposed (in other comments) for all other instances of ' inoperable control rods. These other instances (excessive scram speed, certain combinations of conditions with a low pressure on a control rod scram accumulator) also warrant a minimal time to attempt restoration prior to inserting and ! disarming. Given that these instances, including unknown control rod position, do not represent loss of SDM, and are limited to a total of no more than 8 inoperable control rods (refer to proposed LCO 3.1.3 Condition E) , the extended time does not represent a significant_ safety concern. Disarming a control rod involves personnel actions by other than control room operating personnel, with the necessary activity being within the Primary Containment. These processes require coordination of personnel, preparation of equipment, the potential to requiring anti-contamination " dress-out," and necess into the Primary Containment, in addition to the actual procedure of disarming the control rod. Currently, all these activities must be completed and the control room personnel must confirm completion within the same time allowed to insert the control rod. The disarming is proposed to be extended to four hours; one hour beyond that allowed to insert (consistent with the BWR Standard Technical Specification, NUREG-1434) in PERRY - UNIT 1 33 10/1/93 i I __ ._ J
i l DISCUSSION OF CHANGES CTS: 3.1.3.5 - CONTROL ROD POSITION INDICATION l ! ( ) TECHNICAL CHANGE - LESS RESTRICTIVE (continued) recognition of the potential for excessive haste required to , complete this task. The proposed four hour time does not- l represent a significant safety concern as the control rod-is already in its required position (in accordance with 'other ACTIONS), and the ACTION to disarm is solely a mechanism for precluding the potential for future mis-operation. i L.2 This comment number is not used for this station. f
'l I
( f i i 1 i PERRY - UNIT 1 34 10/1/93 i f-
I 1 l t DISCUSSION OF CHANGES i CTS: 3.1.3.6 - CONTROL ROD DRIVE HOUSING SUPPORT- l l ADMINISTRATIVE ' None in this section. f RELOCATED SPECIFICATIONS R.1 The requirements for the control rod drive housing support to ! be in place are included in the physical design requirements { for control rod OPERABILITY. Plant configuration management i provides adequate controls to assure the supports are in place i through post-maintenance checks and surveillances. Relocation ! of the controls to the plant configuration management program ; is in accordance with the " Application of Selection Criteria to ; PNPP TS" and the NRC Interim Policy Statement on Technical Specification Improvements. Refer to the application document l discussion of TS 3/4.1.3.6 for additional information. -l TECHNICAL CHANGE - MORE RESTRICTIVE l None in this section. ! [ s TECHNICAL CHANGE - LESS RESTRICTIVE .[ t None in this section. ! i I f I r 1 f I i F I i t k h f PERRY - UNIT 1 35 10/1/93 ! i
DISCUSSION OF CHANGES CTS: 3.1.4.1 - CONTROL ROD WITHDRAWAL ADMINISTRATIVE A.1 The technical content of this requirement is being moved to another chapter of the proposed Technical Specifications. Any technical changes to this requirement will be addressed with the content of the proposed chapter location. RELOCATED SPECIFICATIONS None in this section. I TECHNICAL CHANGE - MORE RESTRICTIVE None in this section. TECHNICAL CHANGE - LESS RESTRICTIVE None in this section. I i PER.RY - UNIT 1 36 10/1/93
DISCUSSION OF CHANGES CTS: 3.1.4.2 - ROD PATTERN CONTROL SYSTEM f ZLDMINISTRATIVE , A.1 The . existing Specification for Rod Pattern Control System - (RPCS) is being moved to the Instrumentation- Section (LCO 3.3.2.1). Comments related to proposed changes to the RPCS Specification are addressed within the markup to the existing Instrumentation Section, LCO 3.3.6. However, the existing Specification for the RPCS contains ; requirements / ACTIONS that are not directly associated with the function of the " system." These non-system issues (e.g., determination of SDM with a stuck control rod - b.1; positioning and disarming of inoperable control rods - b.2.a; inoperable control rod separation requirements - b.2.b;_ s number of inoperable control rods allowed in a " group" - b.2.c) ' are proposed to be addressed within the appropriate Section 3-.1 l LCOs, and as such, comments on these details are addressed i below. , i A.2 The technical content of this requirement is being moved to another chapter of the proposed Technical Specifications. Any technical changes to this requirement will be addressed with the content of the proposed chapter location. . A.3 The current design of the RPCS provides for only 8 control rods to be bypassed (bypassing done in the Rod Action Control System ; (RACS)). It is simply a statement of the design to include ;
"8 control rods may be bypassed," and therefore, removing this statement is of no consequence.
A.4 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. t RELOCATED SPECIFICATIONS None in this section. I i PERRY - UNIT 1 37 10/1/93
DISCUSSION OF CHANGES i CTS: 3.1.4.2 - ROD PATTERN CONTROL SYSTEM i TECHNICAL CHANGE - MORE RESTRICTIVE i M.1 Bypassing a control rod in the RGDS prevents insertion or , withdrawal of that rod. Bypassing control rods in the RACS , allows insertion or withdrawal of that rod irrespective of normal rod pattern controller constraints. For the purpose of , bypassing inoperable control rods, in accordance with the proposed SR 3.3.2.1.9, the RACS bypassing function is utilized. l The allowance for bypassing control rods in the RGDS can be utilized as a method of electrically disarming one control rod's directional control valves. TECHNICAL CHANGE - LESS RESTRICTIVE ,
" Generic a LA.1 Details of the methods of disarming control rod (s) are !
relocated to the Bases and procedures. l LA.2 This comment number is not used for this station. L
" Specific" L.1 The RPCS is designed to control rod patterns and movement to within the bound of analyzed conditions. If the system becomes l incapable of perf orming its function (i.e. , " inoperable") then ;
the actions presented in ACTION "a" appropriately apply. These actions actually will automatically occur by the fail-safe nature of the RPCS in the majority of conceivable RPCS inoperabilities. However, the requirements of ACTION "b" contain conditions that may not be directly related to the RPCS function, and therefore do not automatically warrant the same actions. When these , conditions result in operation beyond that allowed by the design of the RPCS, the RPCS will either preclude _ their i occurrence (by blocking control rod movement) or the system ! would be considered inoperable and the appropriate conservative
- ACTIONS applied (consistent with existing ACTION a.1 and a.2)
When the condition does not result in a situation which the RPCS is intended to preclude, other more appropriate Technical : Specification ACTIONS are proposed. These conditions not controlled by the RPCS:
- a. ACTION b.1 - determination of SDM with a stuck control rod; i r
- b. ACTION b . 2. a - positioning and disarming of inoperable control rods; and f
l PERRY - UNIT 1 38 10/1/93 i
i DISCUSSION OF CHANGES CTS: 3.1.4.2 - ROD PATTERN CONTROL SYSTEM , TECHNICAL CHANGE - LESS RESTRICTIVE (continued) ;
- c. ACTION b.2.b -
inoperable control rod separation requirements; and ACTION b. 2. c - number of inoperable control rods allowed in a " group"; ; are either concurrently addressed in other existing Chapter 3.1 LCOs/ ACTIONS, or are proposed to be addressed within Chapter 3.1 LCOs/ ACTIONS, as discussed below. (L.la) Existing ACTION b.1 addresses operation with a stuck ' control rod. This condition is also addressed in the existing ACTIONS f or LCO 3.1.1 and LCO 3.1.3.1. The RPCS is not intended to support maintaining SDM when a control rod is stuck, and should not require actions to suspend ' control rod movement until SDM has been confirmed. Appropriate ACTION is provided for this condition in other i Specifications. Separate ACTIONS are specified in the proposed LCO 3.1.3 allowing 72 hours for re-determining , the SDM. The time is justified by the assumptions in safety analyses which account for the effects of a single stuck control rod. No significant impact on safety is i introduced by allowing the control rod to be bypassed , during this 72 hour period, for continued operation of the - plant while determination of the SDM is being made. (L.lb) The existing restriction appears to be mis-worded. The provision allows bypassing and continued operations only after an inoperable control rod is inserted. In actual practice, to complete the insertion may require bypassing the control rod. In either case, proposed LCO - 3.1 3 provides the necessary ACTIONS for inserting and disarming an inoperable control rod. The proposed change here ' simply allows the bypassing of inoperable control rods, at any time, and allows other Specifications to control the insertion and disarming of that control rod. Since the more safety significant action would be to address the inoperable control rod, if bypassing it prior to insertion would facilitate that action, it is not a significant safety issue to allow that bypassing. (L.lc) The requirement for inoperable control rods to be separated is not necessary in all cases. The separation requirement is adequately controlled by proposed LCO 3.1.3 Condition D. Inoperable control rods are required to be - inserted, and may be adjacent if their respective BPWS groups are allowed to be inserted. If not in accordance with BPWS, proposed LCO 3.1.3 Condition E would require a ; shutdown. Existing LCO 3.1.4.2 ACTION a and ACTION b.2.b, require a shutdown by scram (once below the Low Power Setpoint) in all cases of inoperable control rods not ! PERRY - UNIT 1 39 10/1/93 I w, - c-- - - _ - - - ,
DISCUSSION OF CHANGES , CTS: 3.1.4.2 - ROD PA'l'I'ERN CONTROL SYSTEM TECHNICAL CHANGE - LESS RESTRICTIVE (continued) separated. However, a scram initiates an unnecessary challenge to the equipment which mitigates such a transient. Four hours are allowed to attempt to return the control rod (s) to OPERABLE status. This time frame is based on the requirement to fully insert the inoperable rods per Condition C of proposed LCO 3.1.3. This is based on the high probability of success and the low probability , of a control rod drop accident during this time frame. Again, a controlled shutdown prevents an unnecessary ! challenge to the equipment which mitigates a transient , cause by a scram. (L.1d) A specific requirement for control rods to be in i compliance with the BPWS during operation below the Low Power Setpoint is proposed as LCO 3.1.6. This LCO also contains an allowance (ACTIONS to LCO 3.1.6) for a limited ; number of out-of-sequence OPERABLE control rods, which is presented in the bht Standard Technical Specification, , NUREG-1434, and also proposed to be included in' the i revised Technical Specifications. The ACTIONS allow up to 8 out-of-sequence OPERABLE control rods (separate from any inoperable out-of-sequence control rods) to be returned to their correct position within 8 hours. This allowance for , correction is proposed in recognition of the occurrence of ; such events as " double-nctch" rod withdrawals, and minor ! misalignment of rod pattern during CRD hydraulic transients (control rod drif t due to excessive cooling . water pressure) or during a plant shutdown. These events can introduce out-of-sequence control rod patterns which the RPCS was unable to preclude, even though the RPCS was functioning as designed. Since the time allowed for correction is small, with each ! step bringing the pattern closer to compliance with BPWS, ! and the probability of a Control Rod Drop Accident is ; remote during this brief period, this allowance is deemed to not present a significant impact on safe operation, and would preclude the plant transient introduce by a required reactor scram. L.2 Generic analyses do not specifically restrict the number of inoperable control rods in a BPWS group. Up to eight j inoperable rods can be allowed in a BPWS group if the required i separation is maintained. The separation requirements force , the inoperable control rods to be evenly distributed which minimizes the consequences of the out-of-sequence rods. I PERRY - UNIT 1 40 10/1/93 l l
f DISCUSSION OF CHANGES CTS: 3.1.5 - STANDBY LIQUID CONTROa SYSTEM !
-j ADMINISTRATIVE l t
A.1 With the temperature below its limit the system is inoperable. ! The required verification is not productive during the time the temperature is low. The intent of this requirement (as clarified in the ITS) is to confirm, after restoring the heat i tracing to OPERABLE status, that the low temperature condition ' did not result in precipitation. Since the requirement has not l changed, and the Frequency is one interpretation of the current- i requirements, this change is considered administrative. ; I RELOCATED SPECIFICATIONS f None in this section. I
'I i
TECHNICAL CHANGE - MORE RESTRICTIVE ! 1 None in this section. TECHNICAL CHANGE - LESS RESTRICTIVE' f
" Generic" LA.1 The method of performing the surveillance test is relocated to plant procedures.
r LA.2 The requirement that this test be performed "during shutdown" { is relocated to procedural,- administrative controls in : accordance with the guidance of Generic Letter 91-04. ! r LA.3 Requirements on the replacement charges for explosive valves { have been relocated to the Bases and plant administrative - controls. i LA.4 This comment number is not used for this station.
" Specific" ;
L.1 The SLCS is not required during refueling since only a single
- control rod can 3 withdrawn and adequate SHUTDOWN MARGIN prevents criticalicy under these conditions. t L.2 This comment number is not used for this station.
L.3 This comment number is not used for this station. i l PERRY - UNIT 1 41 10/1/93 . l
, 7 y _> - - - - - _ _ _ _ _ _ _ _ _ _ _ -
d DISCUSSION OF CHANGES ; CTS: 3.1.5 - STANDBY LIQUID CONTROL SYSTEM TECHNICAL CilANGE - LESS RESTRICTIVE : (continued) f l L.4 An additional allowance for a valve in the flow path:to be !' capable of being aligned to the correct position is provided. ; This allowance may be generically added for any system which is 1 manually initiated. Operator action will provide system l initiation consistent with the safety analysis. ! v L.5 Daily surveillance testing of the solution temperature provides I an adequate check on the capability of the storage tank heaters ! to maintain solution temperature and therefore the 18 month test of the tank heaters is deleted. t i L.6 The requirement to determine the available weight of the sodium i pentaborate is being deleted. Determining the sodium i pentaborate concentration within the limits assures that an j acceptable amount of sodium pentaborate is available. This is ! consistent with the Surveillance Requirements contained in s NUREG-1434 for this system. l L.7 The testing Frequency is being changed to every 18 months on a. 7 STAGGERED TEST BASIS. This testing Frequency is sufficient to ; assure that the flow path and squib valve on each subsystem is ! tested within a sufficient period of time. This'will reduce i the number of squib valves required to be used/ replaced, but at ! the same time give proper assurance that the system would i operate if required. L.8 This comment number is not used for this station. i i f i r 6 4 F l t PERRY - UNIT 1 42 10/1/93 l t , _ . . -~ _ - -_ . . _ . ._, , . _.. - . - . . .
M ATTACHMENT IC ; 1 CTS - PSTS COMPARISON DOCUMENT NO SIGNIFICANT HAZARDS CONSIDERATIONS
7;n NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.1.1 - SHUTDOWN MARGIN "11" CHANGE PNPP has evaluated this proposed Technical Specification change and has determined that it involves no significant ha::ards 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?
The proposed change would allow ACTIONS (insertion of control rods and removal of fuel bundles) that are not considered as initiators of any accidents previously evaluated involving a , potential criticality of the core and therefore would not affect their probability. Additionally the proposed ACTIONS : would provide negative reactivity to control the event and reduce the consequences. Therefore, the change does 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 does not involve new equipment, design or , operations, but provides for compenshtory ACTIONS to reduce the ! consequences of a previously analyzed event. Therefore, the proposed change does not create the possibility of a new of different kind of accident from any accident previously' ' evaluated. ,
- 3. Doec this change involve a significant reduction in a margin of safety? ;
The proposed change would allow operations which would increase ; the SHUTDOWN MARGIN when it is below the expected levels and ' would result in a more expeditious return to the required j SHUTDOWN MARGIN. Therefore, the proposed change does not allow i operations which would involve a significant reduction in the i margin of safety. j
\
PERRY - UNIT 1 1 10/1/93 l
i NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.1.1 - SHUTDOWN MARGIN
L2 " CHANGE PNPP 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? ,
e The proposed change deletes the requirement for inserting control rods in core cells with no fuel bundles, but normal control rod movement is not considered as an initiator of a i previously evaluated accident. Therefore, this action or inaction will not significantly increase the probability of an accident previously evaluated. Further, since the reactivity effect of a control rott in a core cell with no fuel bundles is , negligible, the lack of this insertion requirement will not i involve 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 physical modification to ! the plant. Movement of a control rod with fewer than one fuel : assembly in the core cell does not significantly affect the , reactivity and therefore does not create the possibility of a I new or different kind of accident from any accident previously < evaluated.
- 3. Does this change involve a significant reduction in a margin of ,
safety? Considering the negative reactivity inserted by removing the adjacent fcur fuel assemblies is significantly more than any , minimal positive reactivity inserted during any movement of the i control rod, the proposed change does not involve a significant , reduction in a margin of safety. l t I J PERRY - UNIT 1 2 10/1/93 f'
~ . - - ~ , _ . . - . . - , , , ,
NO SIGNIFI3NT HAZARDS CONSIDERATIONS CTS: 3.1.1 - SHUTDOWN IRRGIN I
"L3" CHANGE ,
PNPP has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards-consideration. This determination has been performed in accordance j with the criteria set forth in 10 CFR 50.92. The following ! evaluation is provided for the three categories of the significant i hazards consideration standards-l
- 1. Does the change involve a significant increase in the i probability or consequences of an accident previously !
evaluated? ! The proposed change deletes a surveillance requirement which , provides only confirmatory information. The change would not affect equipment design or operation and involves only a ! surveillance of a specified parameter which is not considered ; as an accident init iator
. . Therefore, the deletion of this :
surveillance will not significantly increase the probability of an accident previously evaluated. Further, since the surveillance would provide only confirmation of the parameter ! value for which sufficient uncertainties and biases have been adequately considered in the limit development, the deletion ofl the surveillance will not involve a significant increase in the ; consequences of an accident previously evaluated. i
- 2. Does the change create the possibility of a new or different l kind of accident from any accident previously evaluated? !
The proposed change does not involve physical modification to ! the plant or a change in the operation. The surveillance ! provides only confinnation of an adequately known value of a. j parameter for which sufficient uncertainties and biases have j been adequately considered in the limit development. ' Therefore, the change does not create the possibility of a new > or different kind of accident from any accident previously ! evaluated. L
- 3. Does this change involve a significant reduction in a margin of !
safety? The SDM limits account for uncertainties and biases, and for ; fuel cycle changes. The required margin is determined by the ! initial startup test and as corroborated by the periodic , reactivity anomaly surveillance (current surveillance 4.1.2) . ] Therefore, the additional surveillance requirement provides no j additional useful information and the proposed change does not i involve a significant reduction in a margin of safety. l l PERRY - UNIT 1 3 10/1/93 ~ l 4 a ,
- - . _ ~ . - - .
i NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.1.1 - SHUTDOWN 1RRGIN I
"J.,4 " CHANGE PNPP has evaluated this proposed Technical Specification change and .
has determined that it involves no significant hazards ! consideration. This determination has been performed in accordance l with the criteria set forth in 10 CFR 50.92. The following l 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?
The proposed change provides an extended time to perform a l SHUTDOWN MARGIN surveillance after identifying a stuck rod.-A , single control rod stuck in a withdrawn position .does not af fect the capability of the remaining OPERABLE control rods to ; i provide the required scram and shutdown reactivity. Therefore, this extended time frame to perform the surveillance will not significantly increase the probability of an accident previously evaluated. Further, since the remaining OPERABLE control rods provide the required scram and shutdown reactivity, this change will not involve a significant increase . I 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 physical modification to ! the plant or a change in the operation. The occurrence of a single stuck control rod was included in the development of the ; limit. As a result, sufficient margin to achieve cold shutdown -j is assured, without an additional failure, until: adequate , margin to accommodate an additional failure is confirmed. , Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously ' evaluated. i
- 3. Does this change involve a significant reduction in a margin of i safety? .
i not ~h test is promptly performed for each of the remaining wiF ,tr wn control rods to ensure no additional control rods are sto s. The extension of the time allowed ~ to demonstrate i ShT2DOWN MARGIN provides a reasonable time to perform the i en41ysis or test to confirm an expected result due to prior ; malysis which includes sufficient uncertainties and biases to , i account for the stuck rod. Th~ cfore, the proposed change does 4 not involve a significant reduction in a margin of safety. ; f t PERRY - UNIT 1 4 10/1/93 ! I
v NO SIGNIFICANT HAZARDS CONSIDERATIONS ! CTS: 3.1.2 - REACTIVITY ANOMALIES : 1 1
"L1" CHANGE PNPP has evaluated this proposed Technical Specification change and l 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? j The proposed change would increase the ACTION time allowed to evaluate and determine the cause of any reactivity anomalies to 72 hours. Such a reactivity anomaly is not considered as an ! initiator of any accidents previously evaluated and therefore , would not affect their probability. Additionally, substantial margin exists in the analysis which predict core reactivity and I in those which analyze the accidents. Further, adequate SHUTDOWN MARGIN is demonstrated by test prior to determining the existence of a reactivity anomaly. with regard to the expected reactivity based on analysis. Based on experience, l any anomalies are expected to be small and slow developing, and insignificant with regard to the consequences. Therefore, the. ; change does not involve a significant increase in the l probability or consequences of an accident previously evaluated. i
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? 1 l
The proposed change does not involve new equipment, design or i operations, but provides for additional time to complete the I previously approved ACTIONS. Therefore, the proposed change l does not create the possibility of a new of different kind of- ' accident from any accident previously evaluated. + 3 -. Does this change involve a significant reduction in a margin of . safety? I The proposed change would allow additional time to detennine i the cause of any reactivity anomalies during which the core ; parameters may not be as analyzed. However, these conditions ' occur infrequently and any minor decrease in the margin during this additional time is offset by not hastily inducing core- , transients while in this condition. Therefore, the proposed ! 1 change does not allow operations which would involve a ; significant reduction in the margin of safety. l l PERRY - UNIT 1 5 10/1/93
i NO SIGNIFICANT HAZARDS CONSIDERATIONS' ! CTS: 3.1.2 - REACTIVITY ANOMALIES ; 5 "L2" CHANGE PNPP 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: i t
- 1. Does the change involve a significant increase in the ,
probability or consequences of an accident previously evaluated? The proposed change revises the activities which initiate the . surveillance to only those which could have significantly , altered the core - reactivity and are not readily. reversible. , Excluded are those activities which alter core reactivity on a ' frequent basis as part of the normal operation, such as control rod movement. The performance of this surveillance does=not 1 involve the operation of, or change to, any equipment which is- , assumed as an initiator for any analyzed accidents. Since the ' excluded operations are previously approved normal activities with reversible effects, the change does not impact the. ; consequences of any analyzed accidents. Therefore, this change i will not significantly increase the probability of an accident l previously evaluated, nor .w ill it involve a significant increase in the consequences of an accident previously ; eval uated. ! t
- 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 a new mode of operation and therefore 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? ,j Allowing this surveillance to be omitted- following outages which could not have significantly affected the core reactivity does not impact the ability of the equipment to maintain the plant within acceptable limits. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
4 PERRY - UNIT 1 6 10/1/93-a
_ ._ - _. m . L k NO SIGNIFICANT HAZARDS CONSIDEPJiTIONS i CTS: 3.1.3.1 - CONTROL ROD OPERABILITY - t i
"L3" CHANGE l PNPP has evaluated this proposed Technical Specification change and l has determined. that it involves no significant hazards consideration. This determination has been perfomed in accordan :e 1 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'- f probability or consequences of an accident previously ,
evaluated? ! The proposed change would revise the Required Actions for ! inoperable control rods below 20% of rated themal power to be ] applicable only when the rods are in noncompliance with the- , banked position withdrawal sequence and are not separated. ! Inoperable rods are not in themselves considered as initiators ' for any accidents previously evaluated and therefore cannot , increase the probability of such accidents. The current , analyses account for the excluded conditions and will therefore -j not contribute to an increase in the consequences of previously . evaluated accidents. Additionally, the extended time . f or - ACTION does not affect the ability of the systems to respond to such accidents and also do not contribute to an increase in the consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different i kind of accident from any accident previously evaluated? l The proposed change does not involve new equipment, design or fi operations changes, but provides additional time to complete the previously approved ACTIONS and eliminates some Required l e Actions for conditions which are accounted for in the current !
analysis. Therefore, the proposed change does not create the , possibility of a new of 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 would allow additional time to . correct . i control rod patterns which may not be as analyzed. 'However, .; these conditions occur infrequently and any' minor decrease'in i the margin during this additional time is of fset:by not hastily j inducing core transients while in this condition. Therefore, i the proposed change does not allow operations .which would : involve a significant. reduction in the margin of safety. ! PERRY - UNIT 1 7 10/1/93 l
NO SIGNIFICANT HAZARDS CONSIDERATIONS l CTS: -3.1.3.1 - CONTROL ROD OPERABILITY ; L2 " CHANGE l PNPP has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards ! consideration. This determination has been performed in accordance i with the criteria set forth in 10 CFR 50.92. The following i 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? The proposed change revises the time allowed to disarm an inoperable control rod. Inoperable rods are not in themselves ;
^
considered as initiators for any accidents previously evaluated and therefore cannot increase the probability of such accidents. Additionally, the extended time for ACTION does not affect the ability of the systems to respond to such accidents since the control rod is in its required position and therefore does not contribute to an increase 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?
i The proposed change does not involve a physical modification to the plant or a new mode of operation and therefore 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 l safety?
The proposed change would allow additional time to disarm inoperable control rods. However, the control rod is in its ! required safe position and disarming only deters future l mis-operation. Such mis-operation is of low probability during : the time immediately following the identification of such l inoperability and any minor decrease in the margin during this additional time is offset by not hastily inducing core
- transients while in this condition. Therefore, the proposed change does not allow operations which would involve a significant reduction in the margin of safety.
PERRY - UNIT 1 8 10/1/93
1 NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.1.3.1 - CONTROL ROD OPERABILITY "L3" CHANGE i PNPP 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?
l The proposed change allows additional time to confirm the SHUTDOWN MARGIN with one stuck control rod. Inoperable rods are not in themselves considered as initiators for any accidents previously evaluated and therefore cannot increase the probability of such accidents. Additionally, the extended , time for ACTION does not affect the ability of the systems to respond to such accidents since the one control rod is assumed . to be fully withdrawn in analyses and therefore does not ! contribute to an increase in the consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different i kind of accident from any accident previously evaluated?
The proposed change does not involve physical modification to !
~
the plant or a change in the operation. The occurrence of a , single stuck control rod was included in the development of the limit. As a result, sufficient margin to achieve cold shutdown , is assured, without an additional failure, until adequate- .; margin to accommodate an additional failure is confirmed. i Therefore, the change 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 _f safety? ;
The SDM limits account for uncertainties and biases,'for fuel ! cycle changes and for one stuck fully withdrawn control rod. The occurrence of a single stuck control rod was included in. ' the development of the limit. As a result, sufficient margin i to achieve cold shutdown is assured, without an additional : failure, until adequate margin to accommodate an additional , failure is confirmed. Therefore, the proposed change does not allow operations which would involve a significant reduction in the margin of safety. , i PERRY - UNIT 1 9 10/1/93 , i
-n-
NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.1.3.1 - CONTROL ROD OPERALILITY "L4" CH14LQS PNPP 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 evaluaticn 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?
The proposed change would allow unlimited continued operation with one stuck control rod. Inoperable rods are not in themselves considered as initiators for any accidents previously evaluated and therefore cannot increase the probability of such accidents. . Additionally, the current analysis accounts for one inoperable fully withdrawn rod and one additional single f ailure. Therefore, this change will not involve 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 physical modification to the plant. The change in the operation is consistent with current safety analysis assumptions and therefore the change 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 is consistent with the assumptions of the current safety analysis and since prompt and regular surveillance confirms no additional inoperable rods, the proposed change does not involve a significant reduction in a margin of safety. PERRY - UNIT 1 10 10/1/93
I i NO SIGNIFICANT HAZARDS CONSIDERATIONS # CTS: 3.1.3.1 - CONTROL ROD OPERABILITY "L5" CHANGE PNPP 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?
P The proposed change revises the time allowed to fully insert an inoperable control rod wh: 1 is not stuck. Inoperable rods are not in themselves considered as initiators for any accidents previously evaluated and therefore cannot increase the probability of such accidents. Additionally, the extended time , for ACTION does not affect the ability of the systems to respond to such accidents since a single control rod is assumed ' to be withdrawn in the accident analyses. Therefore, the proposed change does not contribute to an 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 a new mode of operation and therefore does not i 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 would allow additional time to insert I inoperable control rods. However, the control rod is assumed to be fully withdrawn in the accident analysis and any minor decrease in the margin during this additional time is of fset by not hastily inducing core transients while in this condition. ! Therefore, the proposed change does not allow operations which , would involve a significant reduction in the margin of safety. [ i PERRY - UNIT 1 11 10/1/93
NO SIGIJII'ICANT HAZARDS CONSIDERATIONS CTS: 3.1.3.1 - CONTROL ROD OPERABILITY "L6" CHANGE PNPP has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards consideration. This determination has been perfr med 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?
The proposed change modifies the Required Actions for inoperable scram discharge volume (SDV) vent and drain valve (s). These valves are not identified as initiators for any accidents previously analyzed. Therefore, the proposed change will not significantly increase the probability of an accident previously evaluated. Further, the proposed change continues to provide ACTIONS which assure the SDV will be available to perform its safety functions. Therefore, this change will not involve 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 physical modification to the plant. A minor change in the operation will allow ACTIONS that return the SDV to a capability to perform its safety functions. Therefore, the change does not create the possibility of a new or different kini of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
Since the proposed ACTIONS will continue to provide a SDV that can perform its safety functions, the proposed change does not involve a significant reduction in a margin of safety. PERRY - UNIT 1 12 10/1/93
-. . . . . . ~ . .--. _ - . = . , -
i NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.1.3.1 - CONTROL ROD OPERABILITY t
"L7" CHANGE T PNPP has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards consideration. This detemination has been performed in accordance r 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? The proposed change extends the surveillance frequency for partially withdrawn control rods. The change would not affect equipment design or operation and involves only a surveillance of a specified parameter which is not considered as an accident initiator. The control rod drive system has demonstrated high reliability at each of the BWR6 power facilities and at otherl similar f acilities. Additionally, the ccntrol rod drive system - is not identified as an initiator for any accidents previously analyzed. Therefore, the change of this surveillance frequency !; will not significantly increase the probability of an accident previously evaluated. Further, extension of the surveillance frepency would not impact the ability of the system to perfom . ! its function following an accident and therefore - the change will not involve 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? ;
m The extension of the surveillance frequency does not involve I physical modification to the plant or a change in the operation. Therefore, the change 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 change in the surveillance frequency does not remove any < requirements for control rod OPERABILITY and only provides a ' minor reduction in the probability of finding an inoperable control rod. Since most of the control rods will continue to- l be tested on the current frequency and if one stuck rod is identified, all rods must be checked promptly, the proposed ' change does not involve a significant reduction in a margin of .; safety. PERRY - UNIT 1 13 10/1/93 . i
I NO S1GNIFICANT HAZARDS-CONSIDERATIONS CTS: 3.1.3.1 - CONTROL ROD OPERABILITY "L9" CHANGE j PNPP 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 i probability or consequences of an accident previously 3 evaluated?
The proposed change would remove an unnecessary additional performance of a surveillance which has been performed within its normally required frequency. Not performing the surveillance would not affect any equipment which is assumed to be an ' initiator of any analyzed event. Further, since.the surveillance continues to be performed on its normal frequency, . there is no impact on the capability of the system to perform its required safety function. Therefore, the proposed change-does 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?
3 The proposed change does not create the possibility of a new or different kind of accident from any accident previously i evaluated because the proposed change introduces no new mode of plant operation nor does it require physical modification to ' the plant.
- 3. Does this change involve a significant reduction in a margin of safety?
The normal surveillance frequency has been shown,. based on operating experience, to be adequate for assuring the equipment ; is available and capable of performing its intended function. Additionally, the requirements of SR 3.0.4 (current i Specification 4.0.4) provide assurance the equipment- is : OPERABLE prior to beginning the functions for which it is ! required. Therefore, the proposed change does not involve a significant reduction in a margin of safety. i
+
PERRY --UNIT 1 14 10/1/93
NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.1.3.2 - CONTROL ROD MAXIMUM SCRAM INSERTION TIMES ; I "L1" CHANGE PNPP has evaluated this proposed Technical Specification change and i 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'?
The proposed change would revise the limits on scram insertion times for control rods. Inoperable rods are not in themselves considered as initiators for any accidents previously evaluated i and therefore cannot increase the probability of such - accidents. Additionally, the current analysis- provides ; sufficient margin to account for the proposed allowances of slow and inoperable control rods. Therefore, this_ change will not involve 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 physical modification to the plant. The change in the operation is consistent with current safety analysis assumptions and therefore the change i 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 is consistent with the assumptions of the ! current safety analysis and since regular surveillance confinns-conformance to the scram time assumptions, the proposed change , does not involve a significant reduction in a margin of safety. PERRY - UNIT 1 15 10/1/93
NO SIGNIFICANT HAZARDS CONSIDERATIONS-CTS: 3.1.3.2 - CONTROL ROD MAXIMUM SCRAM INSERTION TIMES "L2" CHANGE PNPP 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?
I The proposed change deletes the requirement for increased frequency of control rod testing when more than three rods are slow. Inoperable rods are not in themselves considered as-initiators for any accidents previously evaluated and therefore cannot increase the probability of such- accidents. Additionally, the current analysis provides sufficient margin [ to account for the proposed allowances of slow and inoperable , control rods. Therefore, this change will not involve a I significant increase in the consequences of an accident previously evaluated. !
- 2. Does the change create the possibility of a new or different 4 kind of accident-from any accident previously evaluated?
The proposed change does not involve physical modification to -j the plant or a change in the operation. 5herefore, the change . does not create the possibility of a new ir different kind of 1 accident from any accident previously es,ouated. ; i
- 3. Does this change involve a significant reduction in' a margin of -
safety? i The proposed change is consistent with the assumptions of the current safety analysis and therefore does not involve a i significant reduction in a margin of safety. , i n Y i 1 i i PERRY - UNIT 1 16 10/1/93 l i
.- . - - .. ._. ~- - . -
e NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.1.3.2 - CONTROL ROD MAXIMUM SCRAM INSERTION TIMES f
"L3" CHANGE l PNPP has evaluated this proposed Technical Specification change and l has determined that it involves no significant hazards !
consideration. This determination has been performed in accordance i 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?
The proposed change deletos a surveillance requirement which provides only confirmatory information. The change would not i affect equipment design or operation and involves only a surveillance of a specified parameter which is not considered as an accident initiator. Therefore, the deletion of this l surveillance will not significantly increase the probability of '; an accident previously evaluated. Further, since the surveillance would provide only confirmation of information already available, the deletion of the surveillance will not , involve 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 physical modification to ! the plant or a change in the operation. The surveillance provides only confirmation of known information. Therefore, l the change 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? i Since the scram time for each rod is available from prior testing and the insertion capability of each rod is determined , frequently, sufficient information exists without additional testing to provide high probability of conformance to . the safety analysis. Therefore, the proposed change does not involve a significant reduction in a margin of safety. PERRY - UNIT 1 17 10/1/93 r ., _
NO SIGNIFICANT HAZARDS CONSIDERATIONS l CTS: 3.1.3.2 - CONTROL ROD MAXIMUM SCRAM INSERTION TIMES ;
"L4" CHANGE PNPP has evaluated this proposed Technical Specification change and i 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 i evaluation is provided for the three categories of the significant. i hazards consideration standards:
- 1. Does the change involve a significant increase in the- +
_ probability or consequences of an accident previously evaluated? The proposed change deletes a surveillance requirenent which- , provides only confirmatory information. The change would not : affect equipment design or operation and involves only a l surveillance of a specified parameter which is not considered as an accident initiator. Therefore, the deletion of this ' surveillance will not significantly increase the probability of i an accident previously evaluated. Further, since the surveillance would provide only confirmation of information already available and not affected by the currently initiating ; event, the deletion of the surveillance will not involve a ' significant increase in the consequences of an accident i previously evaluated. ; L
- 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 physical modification to I the plant or a change in the operation. The surveillance i provides only confirmation of known information and is not 4 affected by the currently initiating event. Therefore, the r change does not create the possibility of a new or different i kind of accident from any accident previously evaluated. .
- 3. Does this change involve a significant reduction in a margin of f safety? (
i Since the scram time for each rod is available from prior } testing, the insertion capability of each rod is determined ' frequently, and these parameters are not affected by the currently initiating event of fuel movement, sufficient ,. information exists without additional testing to provide high ; probability of conformance to the safety analysis. Therefore, l the proposed change does not involve a significant reduction in a margin of safety. I f PERRY - UNIT 1 18 10/1/93
NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.1.3.3 - CONTROL ROD SCRAM ACCUMULATORS "L1" CHANGE PNPP 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?
The proposed change would revise the declared status of control rods with an inoperable accumulator and extend the time allowed to declare such status. Inoperable rods are not in themselves considered as initiators for any accidents previously evaluated and therefore cannot increase- the probability of such accidents. Additionally, the current analysis provides sufficient margin to account for the proposed allowances of slow and inoperable control rods. Therefore, this change will not involve 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 physical modification to the plant. The change in the operation is consistent with current _ safety analysis assumptions and-therefore the change 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 is consistent with the assumptions of the current safety analysis and since the reactor pressure is sufficient to provide the scram function of the control. rods, the proposed change does not involve a significant reduction in a margin of safety. PERRY - UNIT 1 19 10/1/93
t. t NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.1.3.3 - CONTROL ROD SCMJ1 ACCUMULATORS .
"L3" CHANGE PNPP 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 i evaluated?
The proposed change would reduce the reactor pressure at which a control rod can continue to be considered OPERABLE but slow. , Inoperable or slow rods are not in themselves considered as initiators for any accidents previously evaluated and therefore I cannot increase the probability of such accidents. Additionally, the current analysis provides sufficient margin to account for the proposed allowances of slow control rods. Therefore, this change will not involve 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 physical modification to the plant. The change in the operation is consistent with
- current safety analysis assumptions and therefore the change.
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 is consistent with the assumptions of the-current safety analysis and since the reactor pressure is - sufficient to provide the insertion function of the control rods, the proposed change does not involve a significant , reduction in a margin of safety. t i PERRY - UNIT 1 20 10/1/93 , t - - . , , ~ - - , .
i NO SIGNIFICANT HAZARDS CONSIDERATIONS f CTS: 3.1.3.4 - CONTROL ROD DRIVE COUPLING -
"L1" CHANGE PNPP has evaluated this proposed Technical Specification change and has detennined 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? The proposed change would delete the requirements for control rod coupling during the refueling mode. Since only one control rod is allowed to be withdrawn from core cells containing fuel i assemblies during refueling, the coupling requirements provide l no additional controls. Therefore, the elimination of such , controls does not increase the probability of a previously evaluated accident. Additionally, the remaining requirements provide controls consistent with the assumptions of the current analysis. Therefore, this change will not involve .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 physical modification to l the plant or change in the operations. Therefore, the change ' does not create the possibility of a new or different kind of .j accident from any accident previously evaluated. i 1
- 3. Does this change involve a significant reduction in a margin of i safety?
e The proposed change removes redundant controls and is consistent with the assumptions of the current safety analysis. Therefore, the proposed change does not involve-a significant , reduction in a margin of safety. l t 1 l l l PERRY - UNIT 1 21 10/1/93 !
t NO SIGNIFICANT HAZARDS CONSIDERATIONS i CTS: 3.1.3.4 - CONTROL ROD DRIVE COUPLING ! l "L2" CHANGE PNPP 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 l 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 proposed change would increase the time allowed to l accomplish recoupling, allow bypassing of the rod pattern :
control system to recouple, and remove the restriction for a single attempt to recouple. Inoperable rods are not in themselves considered as initiators for any accidents previously evaluated and therefore cannot increase. the probability of such accidents. Additionally, the proposed ' ACTION does not affect the ability of the systems to respond to such accidents since a number of inoperable control rods are- , assumed in the accident analyses. Therefore, the change does , not contribute to an increase in the consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different t kind of accident from any accident previously evaluated?-
The proposed change does not involve physical modification to the plant or change in the operations. Therefore, the change ; does not create the possibility of a new or different kind of
~
accident from any accident previously evaluated. j
- 3. Does this change involve a significant reduction in a margin of safety?
The proposcd change removes unnecessary restrictions which may ! , prevent an unnecessary shutdown - and is consistent with the l assumptions of the current safety analysis. Therefore, the proposed change does not involve a significant reduction in a f margin of safety. b t b PERRY - UNIT 1 22 10/1/93 :
.-- --- ,-,,e- - - - ~ _ , , , ,
NO SIGNIFICliNT HAZARDS CONSIDERATIONS CTS: 3.1.3.5 - CONTROL ROD POSITION INDICATION "L1" CHANGE PNPP 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?
The proposed change would require that control rods whose position is unknown be treated as inoperable control rods. Inoperable rods are not in themselves considered as initiators for any accidents previously evaluated and therefore cannot increase the probability of such accidents. Additionally, the proposed ACTION does not affect the ability of the systems to respond to'such accidents since a number of inoperable control rods are assumed in the accident analyses. Therefore, the change does not contribute to an 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 a new mode of operation and therefore 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 would consider any control rod whose position is unknown as inoperable. Since some inoperable control rods are assumed in the analysis, the proposed change does not allow opera tions which would involve a significant reduction in the marg of safety. PERRY - UNIT 1 23 10/1/93
NO SIGNIFICANT HAZARDS CONSIDERATIONS , CTS: 3.1.4.2 - ROD PATTERN CONTROL SYSTEM i i "L1" CHANGE PNPP has evaluated this proposed Technical Specification change and ; has detennined that. it involves no significant hazards i
~
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 tne significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
- a. The proposed change would require that SHUTDOWN MARG 1N be ,
determined when a control rod is determined to be stuck without regard for the status of the rod pattern control system. . Neither failure of the rod pattern control system nor i inoperable rods are considered as initiators for any accidents < previously evaluated. Therefore, the proposed change cannot . increase the probability of any previously evaluated accidents. l Additionally, the proposed ACTION does not affect the ability , of the systems to respond to such accidents. Therefore, the change does not r:ontribute to an increase in the consequences of an accident previously evaluated.
- b. The proposed change would allow bypassing of an inoperable I control rod without regard for the status of the rod pattern i control system. Neither failure of the rod pattern control ;
system nor inoperable rods are considered as initiacors for any accidents previously evaluated. Therefore, the proposed change ! cannot increase the probability of any previously evaluated accidents. Additionally, the proposed ACTION does not affect j the ability of the systems to espond' to such accidents. i Therefore, the change does not contribute to an increase in the consequences of an accident previously evaluated. . I
- c. The proposed change would provide some time to attempt to [
restore adjacent inoperable control rods to OPERABLE status, and allow a manual controlled shutdown rather than requiring a i scram. Inoperable rods are not in themselves considered as i initiators for any accidents previously evaluated and therefore cannot increase the probability of such accidents. The , proposed ACTION may affect the ability of the systems to ; respond to some accidents since the configuration would not be in accordance with the assumptions'of the accident analyses. However, the impact is expected to be insignificant due to the i margin inherent is the analysis methodology and the low probability of an event during the allowed time. Further, such , impact would be offset by avoiding an unnecessary challenge to the system through the initiation of a scram. Therefore, the ' change does not contribute to an increase in the probability or ; consequences of an accident previously-evaluated. , PERRY - UNIT 1 24 10/1/93
?
i
f f NO SIGNIFICANT HAZARDS CONSIDERATIONS i CTS: 3.1.4.2 - ROD PATTERN CONTROL SYSTEM l "L1" CHANGE ; (continued) {
- d. The proposed change would allow a limited time of operation with up to eight control rods out of sequence with the banked i position withdrawal sequence. The position of control rods is j not considered as an initiator of any previously evaluated i accident. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated. Additionally, the out of. sequence rods. !
are considered in the current evaluation of accidents and ; therefore the change does not contribute to an increase in the ! consequences of an accident previously evaluated. ; i
- 2. Does the change create the possibility of a new or different l kind of accident from any accident previously evaluated? ;
- a. The proposed change does not involve a physical modification to :
the plant or a new mode of operation and therefore does not create the possibility of a new or different kind of accident , from any accident previously evaluated. l t
- b. The proposed change does not involve a physical modification to l the plant. The chhwie in operation provides flexibility to l facilitate rod insertion and is consistent with the safety {
analysis assumptions for inoperable control rods. Therefore, , the proposed change does not create the possibility of a new or ! different kind of accident from any accident previously evaluated. , I
- c. The proposed change does not involve a physical modification to i the plant or a new mode of operation and therefore does not !
create the possibility of a new or different kind of accident ! from any accident previously evaluated. l
- d. The proposed change does not involve a physical modification to f the plant and the change in operation is considered in the !
current safety analysis. Therefore, the proposed change does , not create the possibility of- a new or different kind of accident from any accident previously evaluated. ; I
- 3. Does this change involve a significant reduction in a margin of safety?
- a. The proposed change would continue to provide the required !
! SHUTDOWN MARGIN during operation of'the plant. Therefore, the - proposed change does not allow operations which would-involve ! a significant reduction in the margin of safety. l t i PERRY - UNIT 1 25 10/1/93 , t
NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3 1.4.2 - ROD PATTERN CONTROL SYSTEM-
"L1" CHANGE !
(continued)
- b. The proposed change would continue to provide the required SHUTDOWN MARGIN during operation of the plant. Therefore, the proposed change does not allow operations which would involve a significant reduction in the margin of safety.
- c. The proposed change would allow operation in modes which may represent a minor reduction in the margin of safety. However, !
the impact is expected to be insignificant due to the numerous , conservatisms inherent is the analysis methodology and the low i probability of an event during the identified. condition. Further, such impact would be offset by avoiding an unnecessary challenge to the system through the initiation of a scram, t Therefore, the proposed change does not allow operations which ! would involve a significant reduction in the margin of safety. I
- d. This change may involve a minor reduction in the margin of f safety by allowing operation with fewer restrictions on the out !
of sequence rods. However, this reduction is offset by the high probability that the out of sequence rods would . be returned to their correct position in a short period of time and a reactor shutdown transient would be avoided. Therefore, the proposed change does not allow operations which would .! involve a significant reduction in the margin of safety ; I
?
1 l 1 1 s e PERRY - UNIT 1 26 10/1/93 i
t i NO SIGNIFICANT HAZARDS CONSIDERATIONS i CTS: 3.1.4.2 - ROD PATTERN ~ CONTROL SYSTEM i l "L2" CHANGE I PNPP has evaluated this proposed Technical Specification change and i has determined that it involves no significant hazards j consideration. This determination has been performed in accordance ; with the criteria set forth in 10 CFR 50.92. The following j evaluation is provided for the three categories of the significant j hazards consideration standards- <
- 1. Does the- change- involve a significant increase in the probability or consequences of an accident previously ,
evaluated? , Control rod patterns that: )
- 1) conform to the' requirements of BPWS while below the ,
low power setpoint, and l 2)- result in operation within the limitations of the I fuel thermal limits presented- in Technical Specification section 3.2, while greater than 25 % , power, ; are not assumed to be the initiator of any analyzed accident or ; transient. The existing restriction of "not more than 3 i inoperable control rods in any RPCS group" is not an assumption j of BPWS, or of any thermal limits calculation. Therefore, i eliminating this restriction will not result in unanalyzed i control rod patterns, and will.not significantly increase the l probability of an accident previously evaluated. l
+
When below the low power setpoint, the control rod pattern is 'I of concern in only the CRDA. Above the low power setpoint, i control rod patterns are not specifically limited. No accident l or transient previously evaluated limits the number- of } inoperable control rods in any one - group. Therefore,-this- j proposed change does not involve a significant increase in the ; consequences of an accident previously evaluated. i 1 ii 4 PERRY _- UNIT 1 27 10/1/93 {
=ye- e mer-.e. p .-
NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.1.4.2 - ROD PATTERN CONTROL SYSTEM ,
"L2" CHANGE ,
(continued)
- 2. Does the change create the possibility of a new or different !
kind of accident from any accident previously evaluated?- Tlie proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. Operation will be allowed with more than 3 inoperable control rods in an RPCS group, however separation requirements remain. The separation requirements force the inoperable control rods to be evenly distributed which adequately minimizes the consequences of the out-of-sequence ; rods. Thun this change does not create the possibility of a new or dif :rm ' kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of !
safety? The proposed change will not reduce a margin of safety because i it has no impact on any safety analysis assumptions. - In addition, the separation requirements and thermal limits i surveillances remain the same as the existing Technical Specifications. Therefore, the proposed change will' not l significantly reduce any margin of safety. 8 l f i l l 1 PERRY - UNIT 1 28 10/1/93 i i
NO SIGNIFICANT HAZARDS CONSIDERATIONS ! CTS: 3.1.5 - STANDBY LIQUID CONTROL SYSTEM i "L1" CHANGE PNPP has evaluated this proposed Technical Specification change and has determined that it involves. no significant hazards i consideration. This determination has been performed in accordance i with the criteria set forth in 10 CFR 50.92. The following i evaluation is provided for the three categories of the significant : hazards consideration standards: ! i
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously i evaluated?
The proposed change deletes the requirements for standby liquid j control _ system (SLCS) OPERABILITY during refueling. The SLCS is not assumed to initiate any previously evaluated events and .i therefore the proposed change will not affect the probability i of a previously analyzed accident. The SLCS is not assumed to l operate in the mitigation of any previously analyzed accidents !' which are assumed to occur during refueling. Therefore, the i proposed change does not contribute to an - increase in the consequences of an accident previously evaluated. l
- 2. Does the change create the possibility of a new or different ,
kind of accident from any accident previously evaluated? i The proposed change does not involve a physical modification to i the plant or a nev mode of operation and therefore does not ; i 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 l safety? l t
The proposed change would remove a backup to the available ! systems for reactivity control. However, this backup is not l considered in the margin of safety when determining the ; required reactivity for refueling events. Therefore, the proposed change does not allow operations which would involve ; a significant reduction in the margin of safety. i t PERRY - UNIT 1 29 10/1/93 I i
i NO SIGNIFICANT HAZARDS CONSIDERATIONS i CTS: 3.1.5 - STANDBY LIQUID CONTROL SYSTEM i 9 "L4" CHANGE ~ PNPP has evale ed this proposed Technical Specification change and ' has determined that it involves no significant hazards '
-consideration This determination has been performed in accordance r 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: S
- 1. Does the change involve a significant increase in the l probability or consequences of an accident previously .
evaluated? !
-t The proposed change would allow a valve in the SLCS to be in a i' position other than its required position as long~as it is capable of being manually aligned. The SLCS is not assumed to ;
initiate any previously evaluated events and therefore the , proposed change will not affect the probability of a previously j analyzed accident. The SLCS is not assumed to automatically initiate in the mitigation of any previously analyzed accidents' and the proposed change provides an equivalent level of assurance of manual initiation. Therefore, the proposed change does not contribute to an 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 a new mode of operation since the system is manually initiated by design. Therefore, the proposed change ; does not create the possibility of a new or different kind of accident from any accident previously evaluated. i
- 3. Does this change involve a significant reduction in a margin of safety? 1 i
The proposed change would allow-some system valves to be in positions other than the required safety position. However since they will remain capable of alignment _ in a manner ; consistent with initiation of the system, the proposed change ; does not allow operations which would involve a significant reduction in the margin of safety. ! I a PERRY - UNIT 1 30 10/1/93 1
l NO SIGNIFICANT HAZARDS CONSIDERATIONS : CTS: 3.1.5 - STANDBY LIQUID CONTROL SYSTEM i
"L5" CIIAFGE PNPP has evaluated this proposed Technical Specification change and l has determined that it involves. no significant hazards ;
consideration. This determination has been performed in accordance j with the criteria set forth in 10 CFR 50.92. The following I evaluation is provided for the three categories ot the sionificant i hazards considerction standards:
- 1. Does the change involve a significant increase in the l probability or consequences of an accident previously !
evaluated? The proposed change would delete a redundant portion . of a- ! specific surveillance. The remaining surveillance of the SLCS ! provides sufficient assurance that the system will operate as . designed. The SLCS is not assumed to initiate any previously ; , evaluated events and therefore the proposed change will not - affect the probability nf a previously analyzed accident. The- .; deletion of the survei' ince does not affect the capability of ? the system to operate in the mitigation of any previously ; analyzed accidents since another surveillance provides ; assurance of that capability. Therefore, the proposed change l does not contribute to an 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? l The proposed change does not involve a physical modification to. .f the plant or a new mode of operation and therefore does not t create the possibility of a new or different kind of accident j from any accident previously evaluated. ;
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed change would remove a backup verification of an l alternate method to determine the availability of the system. " However, this backup is not considered in the margin of safety i when determining the required reactivity for refueling events. j l Therefore, the proposed change does not allow operations which -; would involve a significant reduction in the margin of safety.
?
i i PERRY - UNIT 1 31 10/1/93
'i
l HO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.1.5 - STANDBY LIQUID CONTROL SYSTEM "L6" CHANGE l PNPP has evaluated this proposed Technical Specification change and f has determined that it involves no significant hazards l consideration. This determination has been performed in accordance i with the criteria set forth in 10 CFR 50.92. The following i evaluation is provided for the three categories of the significant . hazards consideration standards: i
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously '
evaluated? The proposed change would delete a redundant portion of a specific surveillance. Verifying the available weight of the sodium pentaborate is equivalent to determining that the sodium pentaborate solution concentration is within the limits of the figure. Thus, the remaining surveillance of the SLCS provides sufficient assurance that the system will operate as designed. The SLCS is not assumed to initiate any previously evaluated events and therefore the proposed change will not affect the , probability of a previously analyzed accident. The deletion of the part of the surveillance does not affect the capability of the system to operate in the mitigation of any previously analyzed accidents since the remaining surveillance provides assurance of that capability. Therefore, the proposed change ! does not contribute to an 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 a new mode of operation and therefore 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 would remove a redundant verification of a method to determine the availability of the system. However, this is not considered in the margin of safety when determining the required reactivity for refueling events. Therefore, the proposed change does not allow operations which would involve a significant reduction in the margin of safety. - PERRY - UNIT 1 32 10/1/93 ,
_ . . .- . .- -. - - ~- r NO SIGNIFICANT HAZARDS CONSIDERATIONS ! CTS: 3.1.5 - STANDBY LIQUID CONTROL SYSTEM ! l "L7 CIWJGE PNPP has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards ' consideration. This detemination has been perfomed 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 l evaluated? The proposed change would delete a redundant portion of-a specific surveillance by changing the requirement to test both subsystems every eighteen months to the new requirement to test ! the subsystems every eighteen months on a STAGGERED TEST BASIS. .; The remaining surveillance of the SLCS provides sufficient - assurance that the system will operate as designed. The SLCS is not assumed to initiate any previously evaluated events and ;
^
therefore the proposed change will not affect the probability of a previously analyzed accident. The modification to the [ test frequency of- the surveillance does not affect the ; capability of the system to operate in the' mitigation of any !' previously analyzed accidents since the surveillance still provides assurance of that capability. Therefore, the proposed i change does not contribute to an increase in the consequences : of an accident previously evaluated. : l
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? l The proposed change does not involve a physical modification to the plant or a new mode of operation and therefore does not create the possibility of a new or different kind of accident from any accident previously evaluated.
i T
- 3. Does this change involve a significant reduction in a margin of i safety?
The proposed change would revise the test frequency for ! initiating the SLCS subsystems and actually transmitting water ! to the reactor vessel. However, this test frequency is not ! considered in the margin of safety when determining the ! required reactivity for events requiring SLCS initiation. l Therefore, the proposed change does not allow operations which I would involve a significant reduction in the margin of safety. ; f PERRY - UNIT 1 33 10/1/93
ATTACHMENT 2 ITS - PSTS COMPARISON DOCUMENT - i 2A: MARKUP OF ITS 2B: DISCUSSION OF CHANGES
ATTACHMENT 2A l ITS - PSTS - COMPARISON DOCUMENT MARKUP OF ITS i i l
'1 1
I
'SDM l
3.1.1 : I 1 3.1 REACTIVITY CONTROL SYSTEMS : 3.1.1 SHUTDOWN MARGIN (SDM) LCO 3.1.1 SDM shall be: ! e
- a. a)0.387% M/k, with the highest worth control rod '
analytically determined; or Og, w
- b. m [0.28}v% M/k, with the highest worth control rod. i determined by test.
APPLICABILITY: MODES 1, 2, 3, 4, and 5. l ACTIONS ! i CONDITION REQUIRED ACTION COMPLETION TIME ; A. SDM not within limits A.1 Restore SDM to within 6 hours ; in MODE 1 or 2. limits. : i B. Required Action and 8.1 Be in MODE 3. 12 hours associated Completion i Time of Condition A i not met.
- d 3
C. SDM not within limits C.1 CI K lif~ T Ad @a ll insert h. zurg;tte.(3.)\ w.
~
in MODE 3. nsertable control > rods. D. SDM not within limits D.1 gu ochM all Im3.f}1Ty7hsert el ne.dkkep t4reura l i l in MODE 4. insertable control - rods. j AND (continued) ( PERty ; BWR/6-STS- \ 78[ 3.1-1 Rev. O, 09/28/92 W i I
SDM ., 3.1.1 ' ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Q r. - , & (continued) D.2 Initiate acTlon to I hour restore {:::e-h y&
- 08l containment}So OPERABLE status..
O - ,-) AND f 0.3 nitiate action I hour
\> d nstore one st dby gas treatmen (SGT) subsystem t OPERABLE status. '
AND D , Initiate action to I hout..
~
t rest o i 1 . e d ocp & lob'. "' atieff to/ ce.pa b ile'I3 OfE LE,fa:a in
@ eaccitseseademy & (rimo,e f containment}A j penetration flow path i g not isolated.
E. SDM not within limits E.1 Suspend CORE Isumediately in MODE 5. ALTERATIONS except for control rod insertion and fuel assembly removal. . E!2 (continued) i
- CO -~
ie O# 050 f, aq aJ ~a u Q
, - .. . . u. s
,J n . ..,kJ o < o J <A o
. a g,jpA le d, pr y, f f })hf u f da
[/f
~i,,....--
- s. a. .s J e A '
lhw /g ,.p s occu n e du' Q~ . ,
)
a 9.h tm &m s.9, .~ BWR/6 STS 3.1-2 Rev. O, 09/28/92 l
' S[nV ;
3.1.1 i i
=l ACTIONS o
-CONDITION REQUIRED ACTION' COMPLETION TIME E. (continued) E.2 Initiate action to Immediately i fully insert all insertable control rods in core cells containing one or i more fuel assemblies. :
t AND r,m st E.3 Initiate act on to i I hour restord*'g:;;r.trjd OP containment}Lto l OPERABLE status. ? Q% / E/ Ini ate action to re tore one-SGT I hour y bsystem to OPE '$L_ LE tatus. t- - 6N,,Q E Initiate action to Lhour restore p :c ::!;;;;;; 9m , I Ni "'b..YI.I.s.. .. , m..... . . isc/adles 3 a;ig
"**LE& stetus-in f -
0^E.R esin --"4 y Qi,m,
/pg/</ red containment}q % Bf .;
penetration flow path i d/D not isolated.
. e.%oJed
,,t ,,, , , ,,1, J
- ~ ~
Ika A Q_,' ,, g-jf
# /
r r.1,.t. o4 A 6dm n oteuca hos' " I- IS A abptg &W
,a v . -. +: ~a c ,4 g
/ >
a . , \ . ,h . - BWR/6 STS 3.1-3 Rev. O,-09/28/92 ' l
SDM- l 3.1.1 ; SURVEILLANCE REQUIREMENTS t SURVEILLANCE-FREQUENCY i i SR 3.1.1.1 Verify SDM is: Prior to each , a in vessel fuel ! g3 a. t 10.38]q,% ak/k with the. highest worth movement during . control rod analytically ' determined; fuel.Ioading or sequence
- b. t)0.28}IAk/k with the highest worth AND control rod determined by test.
-{
Once within: ; 4 hours after criticality following. fuel ; movement or ; control rod '
-- reactor-f pressure vessel F
h h l r l l i BWR/6 STS 3.1 4 O, 09/28/92 Rev. l i
Reactivity Anomalies 3.1.2 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 Reactivity Anomalies LCO 3.1.2 t u 1 The reactivity ldifferenceT between the [rMt:r;d ;;r; ';,g .!
~*:-d th; pr;dicted ca e E,11] shall be within e 1% Ak/k. ~!
'%re_.kh st t) Cod)(nd a a-> '
-.-. ~ 3 W prid Ahel r ) hm Ap; e I APPLICABILITY: MODES I and 2. '
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME : A. Core reacti y A.1 Restore core Q 9di fference not within limit. reactivity 9 difference}90 72 hours
~!
within Ifmit. ' B. Required Action and 8.1 Be in MODE 3. 12 hours associated Completion Time not met. ! o s i 5 1 i ( I t l BWR/6 STS 3.1-5 Rev. O, 09/28/92 ! I i i 1
1 Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE ' FREQUENCY i SR 3.1.2.1 Verify core reactivit difference} tween Once within
,i the-{*onitored care Left or.d th: p dicted 24 hours after
. -' ~ -- l e c N7;}- is within
- 1% &/k.
g reaching ' { 7
-f -- --
w,__. equilibrium m e bie2 f*J '^5A % ' ~) % conditions
' following
[ 3 r eclic ked I o .1 ) < -fA UfF. startup after
~ " ~
fuel movement ^q N or control rod
\ eplacement g _[( d i f00E i~ m '9 iof"[* , ithin the j sreactor N
@ ] pressure vessel'1 AND 1000 MWD /T thereafter BWR/6 STS 3.1-6 Rev. O, 09/28/92
Control Red OPERA 81LITY 3.1.3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE. , APPLICABILITY: MODES : and 2. ACTIONS
-------------------------------------NOTE------------------------------- -----
Separate Condition entry is allowed for each control rod. CONDITION REQUIRED ACTION COMPLETION TIME A. One withdrawn control ------------NOTE----.-------- rod stuck. A stuck rod may be bypassed in the Rod Action Control System (RACS) in accordance with SR 3.3.2.1.ju if reruired M*- . P4 to allow continued operation. A.1 Disarm the associated 2 hours control rod drive (CRD). MQ (continued) l BWR/6 STS 3.1-7 Rev. O, 09/28/92
\
Control Rod OPERABIL2TY 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 --------HOTE--------- - Not applicable when less than or equal to the low power setpoint (LPSP) of the Rod Pattern Control System (RPCS). Perform SR 3.1.3.2 24 hours and SR 3.1.3.3 for each withdrawn OPERABLE control rod. 6.Np, A.3 Perform SR 3.1.1.1. 72 hours B.
)2:
Two or more withdrawn B.1 Disarm the associated L )Fhours control rods stuck. CRD. ^ cy AND B.2 Be in MODE 3. 12 hours . (continued) BWR/6 STS 3.1-8 Rev. O, 09/28/92
Centrol Rod OPERABILITY . 3.1.3
- ACTIONS (continued)
CONDITION REQUIRED ACTION. . COMPLETION TIME l C. One or more control. C.1 -------. NOTE--------- 1 rods inoperable for Inoperable control reasons other than rods may be bypassed ' Condition A or B. in RACS in accord with SR 3.3.2.1.,r_ ,Aanc required, to allow if . l insertion of i inoperable control , rod and continued l operation. ! Fully insert 3 hours l inoperable control ' rod.
.y A!!D C.2 Disarm the associated 4 hours ,
CRD. h D. ---------NOTE--------- D.1 Restore compliance 4 hours ; Not applicable when with BPWS. ! THERMAL POWER... Of5l *...................... M \RTL /gj.) QB D.2 Restore control rod 4 hours 1 or more inoperable to OPERABLE status. ;
@. ,_compliance control rods not in-with banked l
position withdrawal i 1 sequence (BPWS) and not separated by two . or more OPERABLE < 4 control rods. > l (continued) t i BWR/6 STS 3.1-9 Rev. O, 09/28/92 ! i ____ _ ~ _ - , -
Control Rod OPERABILITY 3.1.3 i ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. ---------NOTE--------- E.1 ' Restore the control 4 hours ' Not applicable % .. rod to OPERABLE O when THERMAL POWER N -x Ogstatus. _ .. I.......... i One or more groups N N l with four or more ' p ' inoperable '
,contril rods. ~,.
s t (k -F-- Required Action and associated Completion
;-+ 8e in H00E 3. 12 hours Time fC dition A, '
-4zq[,o C er-E__
'noton met.
- 93 g>. 9g
-~
Nine or more control rods inoperable. , SURVEILLANCE REQUIREMENTS ! 1 SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod. 24 hours i (continued) , t BWR/6 STS 3.1-10 Rev. O, 09/28/92 1
Control Rod OPERABILITV i 3.1.3 SURVEILLANCE REQUIREMENTS (continued)
- SURVEILLANCE FREQUENCY SR 3.1.3.2 ....---------------NOTE--------------------
Not required to be performed until 7 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RPCS. . Insert each fully withdrawn control rod at 7 days least one notch. SR 3.1.3.3 -------------------NOTE-------------------- Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RPCS. Insert each partially withdrawn control rod 31 days at least one notch. , SR 3.1.3.4 Yerify each control rod scram thfrom In accordance fully withdrawn to notch position 1131"1s with
.' s W conds. SR 3.1.4.1, i 7 SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)
BWR/6 STS 3.1-11 Rev. O, 09/28/92 i
r-Control Rod OPERABIL1TY. 3.1.3 r SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY , SR 3.1.3.5 Verify each control rod does not go to the Each time the withdrawn -overtravel position. control rod is withdrawn to '
" full out" position -
AND-Prior to declaring ! control rod OPERABLE sfter work on control rod or CRD System that could affect ' coupling f 5 E 6 l BWR/6 STS 3.1-12 Rev. O, 09/28/92 ' I i 1
Control Rod 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' ' OPERABLE control rods shall be " slow," in accordance with Table 3.1.4-1; and
?>
- b. No b-e0PERABLE control rode that are " slow" s shall occupy adjacent locations.
1 APPLICABILITY: MODES I and 2. ; ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i A. Requirements of the A.1 Be in MODE 3. 12 hours
'.C0 not met. -
SURVEILLANCE REQUIREMENTS i
-------------------------------------NOTE-------------------------------------
During single control rod scram time Surveillances, the control rod drive ' (CRD) pumps shall be isolated from the associated scram accumulator. SURVEILLANCE FREQUENCY ; SR 3.1.4.1 Verify each control rod scram time is Prior to within the limits of Table 3 1 4-1 with exceeding. 73g reactor steam dome pressure t7950}Ysig. 40%'RTP after ; fuel movement i within the reactor pressure vessel AlfD (continued) BWR/6 STS 3.1-13 Rev. O, 09/28/92 i ~. ,, - - - - <
Centrol Red Scran Times 3.1.4 i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 (continued) Prior to i exceeding , 40% RTP after each reactor shutdown a 120 days - SR 3.1.4.2 Verify, for a representative sample, each 120 days tested control rod scram time is within the cumulative limits of Table 3.1.4-1 with reactor steam operation in dome pressure 29950}%sig. MODE I SR 3.1.4.3 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with declaring any reactor steam dome pressure. control rod OPERABLE after work on control rod or CR0 System that could affect scram time SR 3.1.4.4 Verify each affected control rod scram time Prior to is within the limits of Tabl 3.1.4 I with exceeding reactor steam dome pressure t 950]dpsig. 40% RTP after work on control rod or CRD System that could affect scram time BWR/6 STS 3.1-14 Rev. O, 09/28/92 '
Control Rod Scram Times 3.1.4 Table 3.1.4-1 Control Rod Scram Times
- 1. NOTES--------...-------------------......
OPERABLE control rods with scram times not within the limits of this Table artionsidertCilm> Rreired Al. Eder agI;a.ue enAns a.d ms of L9 3.1.3, ,,(whffg!fGPLillTy " Q 04 ' 2. ([ inoperable 6fitro'I rodi WIh scram times re%7]Ssec
, in accordance with SR
....................................,/.3.1.3.4,andarenotconsidere ow."
(5) f ., 7hece n,t4 tol rr<k SCRAM TIMES (a)
) (seconds) ' ~ ;h . ,
REACTOR REACTOR REACTOR
' STEAM PRESSURE W1 00g O
STEAM PRESSUR h STEAM DOME PRESSURE 1h
)
NOTCH POSITION O ps ig f950]hig 1050]Tsig N43[ ( 0.30h N0.31] 29}#-- () -[0.78]* N [0.84] > D13}L ] [1.40] N [1.53] b (a) Maximum scram time from fully withdrawn position, based on de-ener;12ation_ of scram pilot valve solenoids as time zero.
, jb) For(Q-diate%umeJ.6-Q eactor steam done pressures, the scram time criteria are t
(c) determined w. linear interpolation. (cf-For-reaTuram= d=: preu ure s [nej psig, oi.iy nm;> yosii.ict, [131) gr Tscram-time-Hait eppiies,f 0.~) Scro ss %ea as a Nnchen O( reac k rt s demn Ocv16 p re ss o rt I 1+ n
< k o prig are wi111sh 46h6/,Alad /dnify,
~
- a th ,
t T {nq eu \ in Aki t ,h;wf a. (g a t y
. . n e' . ' - ~
s s .
-+t%4- C( \ t t,, q, i~.. p3 L.jp
~
1- 0 , , . -
.. )
BWR/6 STS 3.1-15 Rev. O, 09/28/92
Control Rod Scran Accu:ulators 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5- Control Rod Scram Accumulators LCO 3.1.5 Each control rod scram accumulator shall be OPERABLE. APPLICABILITY: MODES I and 2. ACTIONS
....................................-NOTE------------------------------------.
Separate Condition entry is allowed for each control rod scram accumulator. CONDITION REQUIRED ACTION COMPLETI0ft TIME A. One control rod scram A.1 --------NOTE-------- accumulator inoperable Only applicable if with reactor steam the associated dgme pressure control rod scram ap007~psig. time was within the b 4c0 limits of Omi Table 3.1.4-1 during the last scram time Surveillance. Declare the 8 hours i associated control ~ rod scram time l
' slow." .
A.2 Declare the 8 hours associated control rod inoperable. (continued) I 1 l a BWR/6 STS 3.1 16 Rev. O, 09/28/92 1
Control Rod Scran Accumulators 3.1.5 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. Two or more control B.1 Restore charging 20 minutes from > rod scram accumulators water header pressure discovery of inoperable with to e71520T"'psig. reactor steam dome Condition B qg - pressure it[9001'psig. concurrent with charging water
< 20 s AND B.2.1 --------NOTE--------- '
Only applicable if the associated control rod scram time was within the limits of ' Table 3.1.4-1 during the last scram time Survefilance. Declare the 1 hour associated control rod scram time
" slow."
QB B.2.2 Declare the I hour associated control rod inoperable. C. One or more control C.1 Verify all control Immediately upon rod scram accumulators rods associated with discovery of inoperable with inoperable charging water reactor s cas done accumulators are header pressure pressure < 900}Rpsig. fully inserted. 4"[1520}Rpsig O31 LN. D (continued) BWR/6 STS 3.1-17 O, 09/28/92 Rev.
Centrol Rod Scram Accurulators 3.1.5 l l ACTIONS ) CONDITION REQUIPED ACTION COMPLET10N TIME i C. (continued) C.2 Declare the I hour associated control rod inoperable. r- i D. Required Action and D.1 --------NOTE--------- associated Completion Not applicable if all Time not met. inoperable control rod scram o y, accumulators are q .. lg.)or%a,4d Adlc, associated with fully . C,\ inserted control : rods. Place the reactor Immediately mode switch in the > shutdown position. SURVEILLANCE REQUIREMENTS SURVEILLANCE . FREQUENCY SR 3.1.5.1 Verify each Egntrol pd scram accumulator 7 days pressure is a [1520]7sig. i k BWR/6 STS 3.1-18 Rev. O, 09/28/92
Rod Pattern Control 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control s LCO 3.1.5 OPERABLE control rods shall comply with the requirements of
/ the .
J bankedpositionwithdrawalsequence(BPWS)]\ So APPLICABILITY: MODES 1 and 2 with THERMAL POWER s , RTP.
~
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more OPERABLE A.1 --------NOTE--------- control rods not in Affected control reds J/ compliance with may be bypassed in NBPWS]4 Rod Action Control System (RACS) in accordance with SR 3.3.2.1., . ? Nove associated 8 hours control rod (s) to correct position. QR A.2 Declare associated 8 hours , control rod (s) inoperable. (continued) l 1 l l l l l l BWR/6 STS 3.1 19 Rev. O, 09/28/92 l l
Rod Pattern Control 3.1.6 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. Nine or more OPERABLE B.1 --------NOTE--------- control rods not in Affected control rods O(ilp' M {BPWS} compliance ( with may be bypassed in RACS in accordance with SR 3.3.2.1. for SI$. $$.$$ .. O Suspend withdrawal of Immediately control rods. AND B.2 Place the reactor I hour mode switch in the shutdown position. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Veri {j with*(r all OPERABLE control rods comply 24 hours
,BPWS}(
BWR/6 STS 3.1-20 Rev. O, 09/28/92 l
SLC Sys%m 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System
- LCO 3.1.7 Two SLC subsystems shall be OPERABLE.
APPLICABILITY: MODES I and 2. ACTIONS CONDITION REQUIRED ACTION
- COMPLETION TIME
~
s A. Concentrat &n of A.1 Rectore concentration 72 hours boron in solu A not within limits o' boron in solution o within limits. h) A ANp but > [ ]. _ k J 10 days from iscovery of fa to meet the LCO h h One SLC subsystem 1 Restore SLC subsystem inoperable W 7 days to OPERABLE status. s reasons--ether than A Gondhfon-A}%.a \3 i 10 da from discove of failure t meet the LC Y hflf Two SLC subsystems inoperablfffer r 1 Restore one SLC 8 hours
'g reasons-other44n subsystem to OPERABLE status.
L Condit4en-A]( e, Required Action and Qg 1 Be in H00E 3. 12 hours associated Completion Time not met. BWR/6 STS 3.1-21 Rev. O, 09/28/92
~ - -
SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 24 hours OB1 GN pentaborate AL CXTB solution isj ~ [ C C] {tVerify pli;n;g]. available volume of so 3f r t 1] 7-L
._D s SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours solution is M thinifi2 1: . d d
- L G F45Frf T 1.7 !). 'a 3p7
% SR 3.1.7.3 Ver' ... temperature of_ pu suction piping OP.; \ 24 hours is si hin the lititr -e. pFig= 3.1.' 4.j 4
) no.p -
SR 3.1.7.4 Verify continuity of explosive charge. 31 days SR 3.1.7.5 Verify the concentration of boron in 31 days Lt' solution is"{within the limits of Figure 3.1.7-1 W E Once within 24 hours after . water or boron is added to . solution ' M Once within 24 hours after solution temperature is restored fiiM Nc
,fhe:HMh-of ;
q U l [T4 ure i, t 73# f 9
! '4 1 7_1 V ~] -
s (continued) I BWR/6 STS 3.1-22 Rev. O, 09/28/92 l
SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued) ' SURVEILLANCE ' FREQUENCY SR 3.1.7.6 Verify ea h SLC subsystem manual, power 31 days - operated, and automatic valveF-in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the ' correct position. @ ' SR 3.1.7.7 Vyi fy a JL2,gch pump In accordance gpm at a develops a flow rate discharge pressure - 2 with the t I/t300: psig. Inservice A , im Testing , P ram ec.
~
2 SR 3.1.7.8 Verify flow through one SLC subsystem from 18 months on ' pump into reactor pressure vessel. a STAGGERED TEST BASIS SR 3.1.7.9 Verify all heat traced piping between 18 onths s storage tank and pump suction is ur. blocked.
, 9._ AND Once within A 24 hours after p w p.s u h a .
.- w 1 iging temperature is restored 6 witMn-the ;
Limi4s-o
~
~
[Eigure h!70 f)
- W] _
(continued) I BWR/6 STS 3.1-23 t-Rev. O, 09/28/92
SLC System 3.1.7 SURVEILLANCE REOUIREMENTS (continued) SURVEILLANCE FREQUENCY b3.1.7.10
' Verify sodium pentaborate enr1Sheent is Prior'to ' $ > [60 4,0] atom percent B-10. N addition'to 7q '(x - _- . . - . -. .
SLC tank
\)!
i 1 i BWR/6 STS 3.1-24 Rev. O, 09/28/92 i i
SLC System 3.1.7
\
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,/'
/ /
}k // /,/ /, /
i
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TEMPER ATURE (*F) ' E fsp['C,y,;~6 Figure 3.1.7-1 (page 1 of 1) Soclum Pentaborate Solution Temperature / Concentration Requirements BWR/6 STS 3.1-25 Rev. O, 09/28/92 l l
t SDV Vent and Drain Valves 3.1.8 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Yalves LCO 3.1.8 Each SDV vent and drain valve shall be OPERABLE. 3 l 1 APPLICABILITY: MODES I and 2. ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each SDV vent and drain line. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SDV vent A.1 Restore valve to 7 days ' or drain lines with OPERABLE status. one valve inoperable. B. One or more SDV vent B.1 --------NOTE--------- or drain lines with An isolated line may both valves be unisolated under inoperable. administrative control to allow draining and venting- ' of the 50V.
..................... i Isolate the 8 hours associated line. i C. Required Action and C.1 Be in MODE 3.
associated Completion 12 hours Time not met. , t I 4 BWR/6 STS 3.1-26 Rev. O, 09/28/92 s v - -. , . . - - , -
SDV Vent and Drain Valees 3.1.8 t SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 -------------------NOTE-------------------- Not required to be met on vent and drain valves closed during performance of SR 3.1.8.2. , Verify each SDV vent and drain valve is 31 days vpen. , SR 3.1.8.2 Cycle each SDV vent and drain valve to the 92 days fully closed and fully open position. SR 3.1.8.3 Verify each SDV vent and drain valve: 'T18]Lmonths
- a. Closes in U[30bconds after receipt d of an actual or simulated scram signal; and
- b. Opens when the actual or simulated scraa signal is reset.
BWR/6 STS 3.1-27 Rev. O, 09/28/92
SDM B 3.1.1 l B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM) BASES BACKGROUND SDH requirements are specified to ensure:
- a. The reactor can be made subcritical 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
- c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the 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 w&ter temperature changes experienced during all operating condi^ ions. APPLICABLE The control rod drop accident (CRDA) analysis (Refs. 2 SAFETY ANALYSES and 3) assumes the core is suberitical with the highest worth control rod withdrawn. Typically, the first control rod withdrawn has a very hig' metivity worth and, should the core be critical during the 1,
'thdrawal of the first control rod, the consequences of a CRDA could exceed the fuel damage limits fer a CRDA (see Bases for LCO 3.1.6, " Rod Pattern Control"). Also, SDM 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 OPERABLE when the reactor is in the refueling mode of operation. These interlocks prevent the withdrawal of more than one control rod from the core during refueling.
(Special consideration and requirements for multiple control rod withdrawal during refueling are covered in Special Operations LCO 3.10.6, " Multiple Control Rod 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 SDN has been demonstrated. (continued) BWR/6 STS B 3.1-1 Rev. O, 09/28/92
5DM B 3.1.1 BASES APPLICABLE Prevention or mitigation of reactivity insertion events is SAFETY ANALYSES necessary to limit energy deposition in the fuel to prevent (continued) significant fuel damage, which could result in undue release of radioactivity j(sce Base: fr LCO 2.1.7. s te,.&y L w u] 5 --(C -trai ;5LCl Sntcf jf Adequate SDM '--"--'-inadvertent criticalities and potential CRDAs involving high worth control rods (namely the first control rod ithdrawn) will ( not cause significant fuel damage. % ;6 mg M - SDM satisfies Criterion 2 of the NRC Policy Statement. h i-LC0 The specified SDM limit accounts for the uncertainty in the i demonstration of SDM by testing. Separate SDM limits are ' provided for testing where the highest worth control rod is
; (i.e., Yo codem detemined analytically or by measurement. This is due to
#, #'M4- -
the reduced uncertainty in the SDM test when the highest worth control rod is determined by measurement. When SDM is MA \*d%
~
sepene), demonstrated additional by calculations margin N! Be ea cnot r n:associated na M :d m with am test g=g ' OO b ' , T to account for uncertainties in the calculation. pc e re
' *b ded ace aty5pn da EgtTe signa rocess a cesi ma is
,. i p d to ount f uncertaintic f. es alculati s (Ref ..f
~ . .
APPLICABILITY In MODES 1 and 2, SDM must be provided because suberiticality with the highest worth control rod withdrawn ' is assumed in the CRDA analysis (Ref. 3). In MODES 3 and 4, SDM is required to ensure the reactor will be held suberitical with margin for a single withdrawn control rod. : SDM is required in MODE 5 to prevent _an inadvertent criticality during the withdrawal of a single control rod from a core cell containing one _or more fuel assemblies. , ACTIONS L1 r With SDM not within the limits of the LCO in MODE 1 or 2 SDM sust be restored within 6 hours. Failure to meet the specified SDM may be caused by a control rod that cannot~ be inserted. The 6 hour Completion time is acceptable, considering that the reactor can still be shut down, (continued) BWR/6 STS B 3.1-2 Rev. O, 09/28/92
i SDM l B 3.1.1 I BASES l ACTIONS M (continued) assuming no additional failures of control rods to insert, and the low probability of an event occurring during this interval. M If the SDM cannot be restored, the plant must be brought to MODE 3 within 12 hours, to prevent the potential for further reductions in available SDM (e.g., additional stuck control rods). 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. U ,
/p{tm must With SDM not within limits in HODE 3, the operator must fully insert all insertable control rods within I hour.
Cedhet LtuWOll >This action results in the least reactive condition for the msMakk Ccnfro l core. The allowed Completion Time of I hour is acceptable, ryds aRblQ considering the reactor can still be shut down, assuming no mscWat - h failures of additional control rods to insert. D.1. D.2. D.3. and D.4
& [mnydjatt nifiatG a A 4- _
With SDM not within limits in MODE 4, the erator must insert all insertable control rods.i- I h This action results in the least reactive condition for the core. Be-4-4wur C -apletie- -Ti=e ; ;. -: des asfficient ti;; 0: ick
-;carrccti s c" :- d -: _y t.si c , ;;nsit '"; +F- nrrbr l e2- - I' be ,am de n u wm ; .3 ..
T:i kre ; --
-:e !
Mm l rc-d ;-- t . . . . c i-i; . Actions must also be initiated - within I hour to provide means for control of potential radioactive releases. This includes ensuring 42- :d-V l containment (LCO 3.5 4 ' . %- nnd'ry "--t:i . nM) is T r-etem u i. (CCT)- OPERABLEgyat [1(LC0 -ita;t ene Standby C :Ocs Treatment .D 3.6. " .3, "Stendbf 4MT)M Sy:ter C.G sub;yster i: Or W.0LR and f at least one : c:nd;i., 60 %1 r-%
\ S entai w McontainmentMn!ctir- '! i :: isolatiofr'v~aTve (SCin)-) and associated -(t ^^. ' ' A A.
Npinstrumentation u w mmu_m m
\r m W % tu d is ab cofdllh UCs < (continued) 4WR/6-STS B 3.1-3 Rev. O, 09/28/92 i f6RP4L4 NIT 1
SDM B 3.1.1 BASES
/J56TLT 6 d ACTIONS 0.1. D.2. 0.3. and 0.4 (continued) 45e ktisn 'n +rs--++ 4nn="r arel0PERABLE/in each associated Ll ;s asso vm{ ) penetration flow path not isolatedi This may be perfomed as an administrative check, by examining logs or other O '. M i fo d iged e f 5
information, to detemine if the components are out of service for maintenance or other reasons. It is not ed;c n cW.h necessary to perform the SRs needed to demonstrate the
" ' *
- J OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored.to I OPERABLE status. In this case, SRs may need to be perfomed C 10 ,
to restore the component to OPERABLE status. -kti on: me +g (9 ; c_entinue until a!1 requir:d co ;enent are OPEP/3LE cr obe caupMie od.A@shcWeDo\s 4t, anure 'v.dhepdidRy E .1. E . 2. E . 3. 4 E . 4 / 6 n d1 F2tr-*- k "Y With SDM not within limits in H0DE 5, the operetor nwst -
@ (isonediately suspend CORE Re :,uwenHons-eve-o@ ALTERATIONS insertion that core of fuel in the withdrawal of control rods. Suspension of these activities couldorle reduce s ($DM shall not preclude completion of movement of a component to a safe condition. Inserting control rods or removing fuel from the core will reduce the total reactivity and are therefore excluded from the suspended actions.
Action must also be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies havt been fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and therefore do not have to be inserted. pg gg , Action must also be initiated within I hour to provide means P'"*3 for control of potential radioactive releases. This
- n. includestensuringNE9Hiery' containment M-7 6 A1)'is m
' ' if OPERABL EN et=Jm+-**Qil= = t = (' M A d P. _
l 4 E:7 - and+at least one 6 n isa;Iy containment isolation valve 4tCO 3.5 A27 and associated instrumentation il (100 3.2 1 2) areOPERABLEfineachassociatedpenetratio L flow patn riot isolateds 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 (continued) BW9/5 575 8 3.1-4 Rev. O, 09/28/92 PERM - WJIT 1
E INSERT B4A ; i-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 door can be opened intermittently for entry and exit. This , allowance is acceptable due to the need for contaf.nment , access and due to the slow progression of events which may result from inadequate SDM. Inadvertent reactor criticalities would not be expected'to result in the immediate release of appreciable fission products to the containment atmosphere. Actions must continue until all requirements of this Condition are satisfied. . i i f INSERT PERRY - UNIT 1 B 3.1-4 10/1/93 i
-er 4 -
SDM B 3.1.1 BASES ACTIONS E.1. E.2. E.3. E.4. and E.5 (continued) necessary to perform the SRs needed to demonstrate the OPERABILITY of the components. If, however, any required
=
component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be perfomed g trJ5c7tr 6EA to restore the couponent to OPERABLE status. =Actieu> ww 21.
-weentir,ue oul oil , yo., s cou.pr, cat:; arc Ort =L . -
SURVEILLANCE SR 3.1.1.1 REQUIREMENTS Adequate SDM must be demonstrated to ensure the reactor can be made suberitical from any initial operating condition. Adequate SDM is demonstrated by testing before or during the first startup after fuel movementt control rod replacement, C n (or shuffline within the reactor pressure vessel #j Control a I rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location. Since cere reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle. Therefore, to obtain the SDM, the initial measured value must be increased by an adder, "R", which is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated BOC core reactivity. If the value of R is negative (i.e., BOC is the most reactive point in the cycle) N (Ref.f)no4 correction to the BOC measured valuc is required DMRT E5pm Pg u.M-The SDM may be demonstrated during an in sequence control rod withdrawal, in which the highest worth control rod is analytically detensined, or during local criticals, where the highest worth control rod is determined by testing. Local critical tests require the withdrawal of out of sequence control rods. This testing would therefore require bypassing of the Rod Pattern Control System to allow the out of sequence withdrawal, and therefore additional requirements must be met (see LCO 3.10.7, " Control Rod Testing-Operating") . (continued) BWR/6 STS B 3.1-5 Rev. O, 09/28/92 l
l I INSERT BSA ' In addition, at least one door in each primary containment air lock must be closed. The closed air i lock door completes the boundary for control of , potential radioactive releases. With the appropriate , administrative controls however, the closed 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- - 2 may result from inadequate SDM. Inadvertent reactor I criticalities would not:be expected to result in'the ; immediate release of appreci,able fission products to the containment atmosphere. Actions must continue until all , requirements of this Condition are satisfied. ; t 4 e a i e T 9 9 I INSERT PERRY . UNIT 1 B 3.1-5 10/1/93 i
SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued) REQUIREMENTS
~ ~ ~ - - -
The Frequency of 4 hours after reaching criticality is 7kesc bed 63 allowed to provide a reasonable amount of time to perfom o ui3m meldc , the required al la ions nd appropriate verification. oMiflun t rm7w " During MODE , adequ' ate S is also required to ensure the reactor does 'not reach criticality during control rod 5 Se COM l.'"d b withdrawals ' An evaluation of each in vessel fuel movement acuad 6 Ac t durin core)g fuel loading (including shuffling fuel within (slaTTtie cerfomed)to ensure adequate SDM is the oueclahd toorb. M i maintained during refueling. This evaluation ensures the intemediate loading patterns are bounded by the safety b' analyses for the final core loading pattern. For example, __ E 3 I'gl bounding analyses that demonstrate adequate SDM for the most f 6c to ilqL reactive configurations during the refueling may be
--,3 nis!RT Bh j performed to demonstrate acceptability of the entire fuel r
4 - movement sequence.M [Tir the SDM demonstrations that rely ' solely on calculation, additional margin (0.10% Ak/k) must , be added to the SDN limit of 0.28% Ak/k to account for uncertainties in the calculationi f$piral offload or reload
)
cd {# / sequences inherently satisfy the SR, provided the fuel klyrd "g f assemblies are reloaded in the same configuration analyzed es,c\ rod. for the new cycle. Removing fuel from the core will always (N result in an increase in SDN. REFERENCES 1. 10JFR 50, Appendix A, GDC 26. h 2. / SectioD15.4.9]h y
- 3. NEDO-21231, " Banked Position Withdrawal Sequence
- Se Lion 4.1, January 1977.
b Ih 4. Nk Sectio k 15.4.1.1 k l - {5. FSAi!. Svction-4 3.? > 1 % [. 5
-P-A M 'GE Standard App AILfn Fuel, ten .i.i.C i-. Le g. 19 (TPR g(ctor_leauQ h ', _ _
p l 4 l BWR/6 STS B 3.1-6 O, 09/28/92 Rev. l
1 Reactivity Anomalies t B 3.1.2 , B 3.1 REACTIVITY CONTROL SYSTEMS ! B 3.1.2 Reactivity Anomalies i BASES i BACKGROUND In accordance with GOC 26. GDC 28, and GDC 29 (Ref. 1), reactivity shall be controllable such that subtriticality'is maintained under cold conditions and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Reactivity an9maly is used as a measure of the predicted versus measured core reactivity during power operation. The continual confirmation of core reactivity is necessary to ensure that i the Design Basis Accident (DBA) and transient safety , analyses remain valid. A large reactivity anomaly could be i the result of unanticipated changes in fuel reactivity, t control rod worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SDM i demonstrations (LCO 3.1.1, " SHUTDOWN MARGIN (SDM)") in ensuring the reactor can be brought safely to cold, t suberitical conditions. . When the reactor core is critical or in normal power operation, a reactivity balance exists and the net , reactivity is zero. A comparison of predicted and measured ' reactivity is convenient under such a balance, since ! parameters are being maintained relatively stable under ! steady state power conditions. The positive reactivity 4 inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable absorbers, producing zero net reactivity. 4 1 In order to achieve' the required fuel cycle energy output, the uranium enrichment in the new fuel loading and the fuel ' loaded in the previous cycles provide- excess _ positive ' reactivity beyond that required to sustain steady state operation at the beginning of cycle (BOC). When the reactor ' is critical at RTP :nd :;;:reting we.ius Lyeretsi the OC3 excess positive reactivity is compensated by burnable-i (continued) BWR/6 STS B 3.1 7 'Rev. O, 09/28/92 l l
Reactivity Anomalies , B 3.1.2 ' BASES
\
~ BACKGROUND absorbers (if any), control rods, and whatever neutron (continued) poisons (mainly xenon and samarium) are present in the fuel.
-^
The redicted core reactivity, as represent d c fc e
, is calculated by a 3D core simulator code as a -
4,%c oe.ru r
& Qs) function of cycle exposure. This calculation is performed , ss dd u -~ J h m d for projected operating states and conditions throughout the te ch r.\ r.J J usm s cycle 4 The-montered k,,, is celculated by the m e m n.;cc w g y 5i s for actual plant conditions and is then compared to the predicted value for the cycle exposure.
APPLICABLE Accurate prediction of core reactivity is either an explicit SAFETY ANALYSES or implicit assumption in the accident analysis evaluations. very a ' dent evaluatioJ1 (Ref. 2)fi s,Jnerefore. depenaent ' ( Jon curate evaluatfefi of core reafgivityC In particular, Sys ano reactivity transients, sucn as control rod withdrawal accidents or rod drop accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on cceputer codes that have been qualified against available test data, operating plant data, and analytical benchmarks. Monitoring reactivity anomaly provides additional assurance that the nuclear methods provide an accurate representation of the core reactivity. L The comparison between measured and predicted initial core reactivity provides a normalization for the calculational 1 C o 'i If the measured and
\JtMmg . _ modelsJsed to predict core reactivity.predicte3%' tor identical core c gI w_ x reasonably agree, then the assumptions used in the reload t M yclrJiesign analysis or the calculation'models used to ,
predictb aay not be accurate. If reasonable agreement t between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured value. Thereafter, any significant deviations in the that develop during measured fuel depletion prom maythe predicted be an indicat @ ion that the assumptions of the DBA and transient analyses are no longer valid, or that an unexpected change in core conditions has occurred. Reactivity anomalies satisfy Criterion 2 of the NRC Policy , Statement. l 1 (continued) BWR/6 STS B 3.1-8 Rev. O, 09/28/92
Reactivity Anomalies B 3.1.2 BASES (continued) LCO The' reactivity anomaly limit is established to ensure plant operation is maintained within the assumptions of the safety analyses. Large differences between monitored and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the o uncertainties in the Nuclear Design Methodology are larger than expected. A limit on the difference between the
- 02. ,
bk ,' monitored drMa~nd~thipredictedM%f 1% Ak/k ~ has been established based on engineer"ing~ judgment. A > 1% - deviation in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated. APPLICABILITY In HODE 1. most of the control rods are withdrawn and steady state operation is typically achieved. Under these conditions, the comparison between predicted and monitored core reactivity provides an effective measure of the reactivity anomaly. In MODE 2, control rods are typically being withdrawn during a startup. In MODES 3 and 4, all control rods are fully inserted, and, therefore, the reactor is in the least reactive state, where monitoring core reactivity is not necessary. In HODE 5, fuel loading results in a continually changing core reactivity. SDM requirements (LCO 3.1.1) ensure that fuel movements are performed within the bounds of the safety analysis, and an SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling). The SDM test, required by LCO 3.1.1, provides a direct comparison of the predicted and monitored core reactivity at cold conditions; therefore, reactivity enomaly is not required during these conditions. i ACTIONS 6,d Should an anomaly develop between measured and predicted ' core reactivity, the core reactivity difference must be restored to within the limit to ensure continued operation is within the core design assumptions. Restoration to within the limit could be performed by an evaluation of the core design and safety eralysis to determine the reason for the anomaly. This evaluation normally reviews the core (continued) L BWR/6 STS B 3.1-9 Rev. O, 09/28/92
Reactivity Anomalies B 3.1.2 BASES ACTIONS M (continued) conditions to determine their consistency with input to design calculations. Heasured core and process parameters are also nomally evaluated to determine that they are within the bounds of the safety analysis, and safety I analysis calculational models may be reviewed to verify that they are adequate for representation of the core conditions. The required Completion Time of 72 hours is based on the low probability of a DBA during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis. M If the core reactivity cannot be restored to within the 1% Ak/k limit, 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. 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. SURVEILLANCE REQUIREMENTS SR 3.1.2.1 ho ~ NM - Ie Verifying the reactivity difference between the monitored and predicted M h.g is within the limits of the LCO h ,hl'! p provides further as3urance that plant operation is l maintained within the assumptions of the DBA and transient g- anal ses. The Core Monitoring System calculates the cue - or the reactor conditions obtained from plant / (t o) d mh{'the predicted h@ atxposure m - nstrymentation. A comparison of the monitored terD,'pto t the issameused cyc to calculate the reactivity difference. The comparison is required when the core reactivity has potentially changed by i a significant amount. This may occur following a refueling i in which new fuel assemblies are loaded, fuel assemblies are ! shuffled within the core, or control rods are replaced or shuffled. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control (continued) l l BWR/6 STS B 3.1-10 O, 09/28/92 Rev. _ _ _ _ _ - _ _ - - - _ _ _ - _ - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ^ ^ - - - - - - - - - - - - - -
h i
' Reactivity Anomalies ,
B 3.1.2- l BASES i t SURVEILLANCE SR 3.1.2.1 (continued) REQUIREMENTS ' rod from another core location. Also, core reactivity changes during the cycle. The 24 hour interval after > reaching equilibrium conditions following a startup is based i on the need for equilibrium xenon concentrations in the 1 core, such that an accurate comparison between the monitored Q , d - ~..N' 'and p7elicted f:r: E,y values'can be made. For the s' 3 ,h purposes of this SR, the reactor is assumed to be at . equilibrium conditions when steady state o i control rod movement or core flow changes)perations at t 75% RTP (no have been obtained. The 1000 MWD /T Frequency was developed, considering the relatively slow change in core reactivity with exposure and operating experience related to variations in core reactivity, d , i REFERENCES 1. 10 CFR 50, Appendix A -GDC 26, GDC 28, and GDC 29. h 2. Chapterh15@ k Tkis termpulsen rega:res At cese to be ,pgrows <d power , fpyp k p>h$c M EtWIT C if uitterdEiMir$ ord p4rdsureMf hd f rroNI , g C1
- s' "l" 4" M*i" """"Y"I ""E' 4"' b"> 0
- i
, ampar;s ce issh kncnkow AbtL. , " l i
.I
~
.l
)
)
i BWR/6 STS B 3.1-11 Rev. O, 09/28/92
Control Rod OPERABILITY B 3.1.3 8 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Control Rod OPERABILITY BASES BACKGROUND Control rods are components of the Control Rod Drive (CRD) System, which is the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes to ensure that under conditions of normal operation, including anticipated operational occurrences, specified acceptable fuel design limits are not exceeded. In addition, the control rods provide the capability to hold the reactor core suberitical + under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System. The CRD System is designed to satisfy the requirements of GDC 26, GDC 27, GDC 28, and GDC 29, (Ref. 1). t,, The CRD System consists o lockin piston control rod drive mechanisms (CRDHs) and hydrau ic control unit for each drive mechanism. The locking piston type CRDM is a double acting hydraulic piston, which uses condensate water as the operating fluid. Accumulators provide additional energy for scram. An index tube and piston, coupled to the control rod, are locked at fixed increments by a collet ' mechanism. The collet fingers engage notches in the index tube to prevent unintentional withdrawal of the control rod, but without restricting insertion. This Specification, along with LCO 3.1.4, " Control Rod Scram Times," and LCO 3.1.5, " Control Rod Scram Accumulators," ensure that the performance of the control rods in the event ' of a Design Basis Accident (DBA) or transient meets the assumptions used in the safety analyses of References 2, 3, 4, 5, and 6. APPLICABLE The analy*.ical methods and assumptions used in the SAFETY ANALYSES evaluatior s involving control rods are presented in References 2, 3, 4, 5, and 6. The control rods provide the i primary means for rapid reactivity control (reactor scram), for maintaining the reactor subcritical, and for limiting i (continued) i BdR/6 STS B 3.1-12 Rev. O, 09/28/92 F
Control Rod OPERABTLITY B 3.1.3 BASES APPLICABLE the potential effects of reactivity insertion events caused l' malfu SAFETY ANALYSES (continued) Cao he CRD rapfes assuyct stem. g e capabil y of inserting e ontrol rods egsures that the assumptions for scram reactivity inThe DEX and g transient analyses are not violated. Since the SDH ensures the reactor will be suberitical with the Erh;w control rod withdrawn (assumed single failure), the additional failure of a second control rod to insert could invalidate the demonstrated SDM and potentially limit the ability of the CRD System to hold the reactor subcritical. 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 OPERABLE ensures the CRD System can perform its intended function. The control rods also protect the fuel from damage that could result in release of radioactivity. The limits protected are the MCPR Safety Limit (SL) (see Bases for LCO 3.2.2, ' MINIMUM CRITICAL POWER RATIO (MCPR)"), the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLGHR)," and LCO 3.2.3, " LINEAR HEAT GENERATION RATE (LHGR)'), and the fuel damage limit (see Bases for LC0 3.1.6, " Rod Pattern Control") during reactivity insertion events. 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 LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6 discuss in more detail how the SLs are protected by the CRD System. Control rod OPERABILITY satisfies Criterion 3 of the NRC Policy Statement. LCO OPERABILITY of an individual control rod is based on a combination of factors, primarily the scram insertion times, the control rod coupling integrity, and the ability to determine the control rod position. Accumulator OPERABILITY is addressed by LCO 3.1.5. The associated scram accumulator status for a control rod only affects the scram insertion l (continued) BWR/6 STS B 3.1-13 Rev. O, 09/28/92
i Control Rod OPERABILITY B 3.1.3 BASES LCO I times and therefore an inoperable accumulator does not (continued) innediately require declaring a control rod inoperable. Although not all control rods are required to be OPERABLE to satisfy the intended reactivity control requirements, strict control over the number and distribution of inoperable control rods is required to satisfy the assumptions of the DBA and transient analyses. APPLICABILITY In MODES 1 and 2, the control rods are assumed to function C3 ~ hr I during a DBA or transient and are therefore required to be x OPERABLE in these MODES. In MODES 3 and 4, etntrol rods are j ~ (N u A tuuker p3icoMi. 2ilo-en w.4, to be withdrawn (und e spec ~ 'l opefa11ons 7 mde wibL is, M ,9. ngle 9e'ntro Hof ithdr wal-Hot' Shut wn VI a 0 3 1B4, "S)<fgle ntr61 Rod ithdrawil-Col b
# "' " ^ / ,
dowri/' whichArovi f adequate requirements for control ' covhcA rod bbel d L rod OPERABILITY during these conditions. Control rod requirements in MODE 5 are located in LCO 3.9.5, " Control pred d Rod OPERABILITY-Refueling." ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each control rod. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable control rod, Complying with the Required Actions may allow for continued operation, and subsequent inoperable control rods are governed by subsequent Condition entry and application of associated Required Actions. A.1. A.2. and A.3 A control rod is considered stuck if it will not insert by either CRD drive water or scram pressure. With a fully inserted control rod stuck, no actions are required as long as the control rod remains fully inserted. The Required Actions are modified by a Note that allows a stuck control rod to be bypassed in the Rod Action Control System (RACS) -~. to allow continued operation. SR 3.1.5.f provi M i ' 3 3,7,I additional requirements when control rods are bypassed {in~~ RACS to ensure compliance with the CRDA analysis. With one withdrawn control rod stuck, the control rod must be disarmed within 2 hours. The allowed Completion Time of 1 I 1 (continued) i l BWR/6 STS B 3.1-14 Rev. O, 09/28/92 l
i l I Control Rod OPERABILITY B 3.1.3 1 BASES 1 i ACTIONS A.I. A.2. and A.3 (continued) 2 hours is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable amount of time to p~erfonn the Required Action in an orderl manner. ,' isolating tee ~
~
r ked c5 fcont'foi r'od f' rom's'ciam pieven'ts 'y -damage t'o the CRDH. The
~
M. p g control rod can be isolated from scram by isolating the
' hydraulic control unit from scram and normal insert and
, withdraw pressure, yet still maintain cooling water to the
/ (CRD.f ~
Monitoring of the insertion capability for each withdrawn control rod must also be performed within 24 hours. SR 3.1.3.2 and SR 3.1.3.3 perfona periodic tests of the control rod insertion capability of withdrawn control rods. Testing each withdrawn control rod ensures that a generic problem does not exist. The allowed Completion Time of OC6t 24 hours provides a reasonable time to test the control rods, considering the potential for a need to reduce power 7HER pl Pou]CR to perform the tests. Required Action A.2 is modified by a Note that states the requirement is not applicable when
/3 /rss {La or >t@Dthe actual low power setpoint (LPSP) of the rod pattern controller (RPC), since the notch insertions may not qdh he compatible with the requirements of rod pattern control (LCO 3.1.6) and the RPC (LCO 3.3.2.1, " Control Rod Block Instrumentation").
To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within 72 hours. Should a DBA or transient require a shutdown, to preserve the single failure criterion an additional control rod would have to be assumed to have failed to insert when required. Therefore, the original SDM demonstration may not be valid. The SDM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn, i i The allowed Completion Time of 72 hours to verify SDd is adequate, considering that with a single control rod stuck , in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an l additional control rod adjacent to the stuck control rod i also fails to insert during a required scram. Even with the (continued) BWR/6 STS B 3.1-15 Rev. O, 09/28/92 l
Control Rod OPERABILITY B 3.1.3 BASES ACTIONS A.I. A.2. and A.3 (continued) postulated additional single failure of an adjacent control red to insert, sufficient reactivity control remains to reach and maintain MODE 3 conditions (Ref. 7). B.1 and B.R With two or more withdrawn control rods stuck, the stuck control rods should be isolated from scram pressure within MTuD and the plant brought to MODE 3 within 12 hours. /The allowed Completion Time off1 nasr4 is acceptable, considering
~ ~ ~
Og/ ~ ~ 'tfie low'p'rbba'bWity ofTdDA during this interval. The occurrence of more than one control rod stuck at a withdrawn position increases the probability that the reactor cannot be shut down if required. Insertion of all insertable 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. i C.1 and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully inserted within 3 hours and disanned (electrically or hydraulically) within 4 hours. Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. < The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disanned by closing the drive water and exhaust water isolation valves. Electrically, the control rods can be disarmed by disconnecting power from all four directional control valve solenoids. Required Action C.1 is modified by a Note that allows control rods to be bypassed r in the RAC5 if required to allow insertion of the inoperable control rods and continued operation. SR 3.3.2.1. provides additional requirements when the control rods are b ass to ensure compliance with the CRDA analysis. 4 > The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide (continued) BWR/6 STS B 3.1-16 Rev. O, 09/28/92
Control Rod OPERABfLITY B 3.1.3 BASES ACTIONS C.1 and C.2 (continued) time to insert and disarm the control rods in an orderly manner and without challenging plant systems. D,1 and D.2
~'hvh Out of sequence control rods may increase the potential iy r.eactivity worth of a dropped control rod during a CRDA.
-16% RTP, the generic banked position withdrawal sequence (BPWS) analysis (Ref. 7) requires inserted control OC(, - //Y g rods not in compliance with BPWS to be separated by at least two OPERABLE control rods in all directions, including the diagonal.
Therefore. ify or more inoperable control rods Occ~.- are not in compliance with BPWS and not separated by at least two OPERABLE control rods, action must be taken to restore compliance with BPWS or restore the control rods to OPERABLE status. A Note has been added to the Condition to . Ps clarify that the Condition is not applicable when > M% RTP since the BPWS is not required to be foiiowed underints conditions, as described in the Bases for LCO 3.1.6. The allowed Completion Time of 4 hours is acceptable, considering the low probability of a CRDA occurring. s In additt control rods,to the separation requirements for inope bW assumption in the CRDA analysi . r ANF f$ fuel is that no no than three inoperable allowed in any one B trol rods are roup. Therejo , with one or more BPWS groups having four or re ' erable control rods, the control rods must be restore OPERABLE status. Required Action E.1 is modified Note Condition is not appH)eable when THEicating that the power is > 10% RTP since the BPWS s'not required to be fot conditionj s described in the Bases for L ed under these 3.1.6. The allowediompletion Time of 4 hours is acceptab A sonfidering the low probability of a CRDA occurrirlg h I If any Required Action and associated Completion Time of
; Condition A, C, i er-E- are not met or nine or more (continued)
BWR/6 STS B 3.1-17 Rev. O, 09/28/92
Control Rod OPERABILITY B 3.1.3 BASES ACTIONS h (continued) inoperable control rods exist, 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 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 nn=her.of controtrods permitted to be inoperable when EE oT~ operating aboveD RTP (i.e., no CRDA considerations) could be more than the'value 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 operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.3.1 REQUIREMENTS The position of each control rod must be determined, to ensure adequate information on control rod position is available to the operator for determining CRD OPERABILITY - and controlling rod patterns. Control rod position may be determined by the use of OPERABLE 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 SR is based on operating experience 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 red ir not stuck and is free to insert on a scram signal. These Surveillances are not required when @ the actual LPSP of the RPC since the notch insertions may not be compatible with the requirements CLhh[wa is less (9 or ch (continued) BWR/6 STS B 3.1-18 Rev. O, 09/28/92
l 1 Control Rod OPERABILITV l B 3.1.3 l BASES SURVEILLANCE SR 3.1.3.2 and SR 3.1.3.3 REQUIREMENTS (continued) of -estte n cuntruO (LC0 3.1.6) and the RPC i fj/>/<)S\ (LCO 3.3.2.1). The 7 day frequency of SR 3.1.3.2 is based on operating experience related to the changes it. CRD perfomance and the ease of performing notch testing for fully withdrawn control rods. Partially withdrawn control i i rods are tested at a 31 day frequency, based on the l potential power reduction required to allow the control rod i movement, and considering the large testing sample of SR 3.1.3.2. Furthermore, the 31 day Frequency takes inte account operating experience related to changes in CRD performance. At any time, if a control rod is innovable, a determination of that control rod's trippability , l (OPERABILITY) must be made and appropriate action taken. SR 3.1.3.4 I Verifying the s as time for each control rod to notch position 13 is
-+ _
@ /P' seconds provides reasonable assurance that the control rod will insert when required during a DBA 1
)
,p"";" 3# g4 Q 'or transient, thereby completing its shutdown function.
i This SR is perfomed in conjunction with the control rod
$p jg , (g psh i scras time testing of SR 3.1.4.1, SR 3.1.4.2 SR 3.1.4.3, and SR 3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in CLs =4r m ey h2 b, ,a LCO 3.3.1.1,and the functional testing of SDY vent and drain f valves in LCO 3.1.8joverlap this Surveillance to provide i
Qc "" g_ 5'g 3C complete testing of the assumed safety function. The associated frequencies are acceptable, considering the more I \[c 6 h h k S d frequent testing perfonned to demonstrate other aspects of control rod OPERABILITY and operatin experience, which
>Ah 6\vn,', shows scram times do not significant y change over an ;
,. operating cycle.
SR 3.1.3.5 Coupling verification is performed to ensure the control rod , is connected to the CRDM and will perform its intended ' function when necessary. The Surveillance requires verifying that a control rod does not go to the withdrawn overtravel position when it is fully withdrawn. The overtravel position feature provides a positive check on the coupling integrity, since only an uncoupled CRD can reach the overtravel position. The verification is required to be 1 (continued) BWR/6 STS 8 3.1-19 O, 09/28/92 Rev. i
Control Rod OPERABILITY B 3.1.3
- BASES i 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 i control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This includes control [ rods inserted one notch and then returned to the " full out" ' 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. l REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 27, GDC 28, and GDC 29-
;I
- 2. kA%b)Section 4.3.2.5.5h
- 3. hSection 4.6.1.1.2.5.3}b l
- 4. k)Sectioh5.2.2.2.3k
- 5. [/ Sectio 15.4.1h
- 6. A ection 15.4.9] O
- 7. NED0-21231, ' Banked Position Withdrawal Sequence," +
Section 7.2, January :.377. p l BWR/6 STS B 3.1-20 Rev. O, 09/28/92 l 1
Control Rod Scram Times B 3.1.4 i B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Control Rod Scram Times
)
BASES BACKGROUND The scram function of the Control Rod Drive (CRD) System controls reactivity changes during abnonnal operational transients to ensure that specified acceptable fuel design limits are not exceeded (Ref. 1). The control rods are scrammed by positive means, using hydraulic pressure exerted on the CR0 piston. . When a scram s;gnal is initiated, control air is vented from the scram valves, allowing them to open by spring action. Opening the exhaust valves reduces the pressure above the r main drive pisten to atmospheric pressure, and opening the inlet valve applies the accumulator or reactor pressure to the bottom of the piston. Since the notches in the index i tube are tapered on the lower edge, the collet fingers are ' forced open by c action, allowing the index tube to move upward without restriction because of the high differential pressure across the piston. As the drive moves upward and accumulator pressure drops below the reactor pressure, a ball check valve opens, letting the reactor pressure complete the scram action. If the reactor pressure is low, such as during startup, the accumulator will fully insert the control rod within the required time withcut assistnce from reactor pressure. APPLICABLE The analytical methods and assumptions used in evaluating SAFETY AN' /SES the control rod scram function are presented in References 2, 3, 4, and 5. The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the detemination of plant themal limits (e.g., the MCPR). Other distributions of scran times (e.g., several control rods scramming slower than the average time, with several control rods scramming faster than the average time) can also provide sufficient scras reactivity. Surveillance of each individual control rod's scram time ensures the scram reactivity assumed in the DBA and transient analyses can be met. (continued) BWR/6 STS B 3.1-21 Rev. O, 09/28/92
Control Rod Scram Times . B 3.1.4 ) BASES , APPLICABLE The scram function of the CRD System protects the MCPR SAFETY ANALYSES Safety Limit (SL) (see Bases for LCO 3.2.2, ' MINIMUM (continued) CRITICAL POWER RATIO (HCPR)"), and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," and LCO 3.2.3,
' LINEAR HEAT GENERATION RATE (LHGR)'"), which ensure that no fuel damage will occur if these limits are not exceeded.
Above 950 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL during the analyzed limiting power transient. Below 950 psig, the scram function is assumed to perfonn during the control rod drop accident (Ref. 6) and, therefore, also provides ; protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "RodPatternControl'). For the reactor vessel overpressure protection analysis, 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 Policy Statement. LCO The scram times specified in Table 3.1.4-1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met.
)3 To account for single failure and " slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster Q\ han those assumed in the design basis analysis. The scram o allow up to 7.5% of the control rods t%e.g.,have M3 x 7.5%a margTn
- s scram times that exceed the to have specified limits (i.e., ' slow' control rods) assuming a single stuck control rod (as allowed by LCO 3.1.3, ' Control Rod OPERABILITY") and an additional control rod failing to i scram per the single failure criterion. The scram times are 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 i 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) BWR/6 STS B 3.1-22 Rev. O, 09/28/92
Control Rod Scram Times ' , B 3.1.4
?
BASES LCO To ensure that local scram reactivity rates are maintained (continued) 2 within acceptable limits, no mor*-+han t-o ef A-=stioweA :
" slow" control rods may occupy adjacent locations.
Table 3.1.4-1 is modified by two Notes, which state control r rods with scram times not within the limits of the Table are l nsidered " slow" and that control reds with scram times ! Q 8
>R .a.3.4.
S seconds are considered inoperable as required by , 9 frasca7 823Al O - L ~/ APPLICABILITY In MODES 1 and 2, a scram is assumed to function during ' trar.sients and accidents analyzed for these plant conditions. These events are assumed to occur during l startup and power operation; therefore, the scram function of the control rods is required during.these MODES. In g : OC3 -. .) MODES 3 and 4, the control rods 'are p-j v-li;;'"to be g i withdrawn 'e r ~ cial era ^ons L J.19.J sin e I J'n/5 RT) Icon 01 d Wi draw Ho hutdo ," an a=al
- 03. .4, j ngl Cont 1 Rod ith old Sh down." icb-
}623Bf4'capabilityroy efadequate requirements for control rod scram during these conditions. Scram requirements in MODE 5 are contained in LCO 3.9.5, " Control Rod ,
OPERABILITY-Refueling." l ACTIONS Ad h-{_IAED27 623c] When the requirements of this LCO are not met!the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 : within 12 hours. 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. i SURVEILLANCE The four SRs of this LCO are modified by a Note stating that ! REQUIREMENTS during a single control rod scram tioe surveillance, the CRD : pumps shall be isolated from the associated scram i accumulator. With the CRD pump isolated (i.e., charging : valve closed), the influence of the CRD pump head does not affect the single control rod scram times. During a full .; (continued) 1 l BWR/6 STS B 3.1-23 Rev. O, 09/28/92 i
INSERT B23A This LCO applies only to OPERABLE control rods since inoperable control rods will be inserted and disarmed (LCO-3.1.3). Slow scramming control rods may be conservatively declared inoperable and not accounted for as a slow" control rods. INSERT B238 since the reactor mode switch is in Shutdown and a control rod block is applied. This provides ! INSERT B23C the rate of negative reactivity insertion during a scram-may not be within the assumptions of the safety analyses. Therefore, l l I I l i INSERT PERRY - UNIT 1 B 3.1-23 10/1/93
t Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE co. cram, the CRD pump head would be seen by all control REQUIRENENTS rods 3d would have a negligible effect on the scram (continued) insertion times. SR 3.1.4.1 The scram reactivity used in DBA and transient analyses is based on assumed control rod scram time. Measurement of the scram times with reactor steam dome pressure a 950 psig demonstrates acceptable scram times for the transients analyzed in References 3 and 4. Scram insertion times increase with increasing reactor pressure because of the competing effects of reactor steam dome pressure and stored accumulator energy. Therefore, demonstration of adequate scram times at reactor steam dome pressure greater than 950 psig ensures that the scram times will be within the specified limits at higher pressures. Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be perfonned. To ensure scram time testing is performed within a reasonable time following a refueling or after a shutdown a 120 days, all control rods n _ _; are required to be tested before exceeding 40% RTP. rc::cung (a-evtsowra This frequency is acceptable, considering the additional surveillances perfonned for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by work on control rods or the CRD System.
. Sep/e rceths *repremhhve" SR 3.1.4.2 t _
hh Stde r.rc Additional testing of a sample of controll rods is required M' r*'"E i g' to verify the continued performance of the) scram function A representative sample;contains at least (during the cycle. 10% of the control rods 4GHFno more than 20% of the b- - control rods in thew slow." If more than 20% of the 9,0,* g g Q sample is declared to be " slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this , 'o&c wq- f--shc)i l
' @ For 20% criterion 4is satisfied, or F " " " 1r " ? mt m planned testing, the control rods selected for the sample should be different for each test. Data from inadvertent scrans shoul_d be_ used whenever possible to avcid und Ae MM ady of 'tlet# ccdrol rod ((Arewped -Ae cered ~
drygvM geefde Lco LH. f _ Tcontinued) BWR/6 STS B 3.1-24 Rev. O, 09/28/92
i Control Rod Scram Times B 3.1.4 i BASES SURVEILLANCE SR 3.1.4.2 (continued) REQUIREMENTS unnecessary testing at power, even if the control rods with data were previously tested in a sample. The 120 day Frequency is based on operating experience that has shown control rod scram times do not significantly change over an operating cycle. This Frequency is also reasonable, based on the additional Surveillances done or. the CRDs at more frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5, " Control "od Scram Accumulators." SR 3.1.4.3 When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure. The scram testing must be performed once before declaring the control rod OPERABLE. The required scram time testing must demonstrate that the fftJ6CR7 affected control rod is still within Co g gj $ 1able L; .;a fw-ctv-+7mdiaan@ limits,@ pg '
-J Specific examples of work that could affect the scram times include (but are not limited to) the following: removal'of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumlator isolation valve, or check valves in the piping required for scram.
The frequers.y of once prior to declaring the i.ffected control rod OPERABLE is acceptable because of the capability of testing the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY. SR 3.1.4.4 When work that could affect the scram insertion time is performed on a control rod or CRD System, testing must be done to demonstrate each affected control rod is still witnin the limits of Table 3.1.4-1 with the reactor steam dome pressure e 950 psig. Where work has been performed at (continued) BWR/6 STS B 3.1-25 O, 09/28/92 Rev.
r i
)1 l
1 JESERT B25A , The limits for reactor pressures < 950 psig are establi__ed based on-the expected relationship to. meeting the acceptance criteria at-reactor. pressures 2 950 psig. i 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." s t l 5 t I
?
.i I
h I l 1 I I INSERT PERRY - UNIT 1 B 3.1-25 10/1/93 I i
m . . _ _ _ _ Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE SR 3,1.4.4 (continued) REQUIREMENTS ' high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 will be satisfied with one test. For a control' rod affected by work performed while shut down, however, a zero pressure and a high pressure test may be required. This testing ensures that the control rod scram performance is acceptable for operating reactor pressure conditions prior to withdrawing the control rod for continued operation. Alternatively, a test during hydrostatic pressure testing could also satisfy both criteria. The Frequency of once prior to exceeding 40% RTP is acceptable because of the capability of testing the control rod at the different conditions and the more frequent surveillances on other aspects of control rod OPERABILITY. REFERENCES 1. 10 CFR 50, Appendix A, GDC 10. l a) v561 n '
- 2. Section 14.3.2.5.5]dl-var'r
- 3. Section"{4.6.1.1.2.5.3]S-I i '
4. l af')Section15.2.2.2 ~ }^- 06l ! 5. v arsi RSectionh15.4.1]f-
- 6. kh _fL Sectionil5.4.9}S 2
b BWR/6 STS B 3.1-26 Rev. O, 09/28/92
Control Rod Scrao Accumulators B 3.1.5 l B 3.1 REACTIVITY CONTROL SYSTEMS B 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 LCO 3.1.4, " Control Rod Scram Times." APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the control rod scram function are presented in References 1, 2, 3, and 4. The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. OPERABILITY of each individual control rod scram accumulator, along with LCO 3.1.3, " Control Rod OPERABILITY," and LCO 3.1.4, ensures that the scram reactivity assumed in the DBA and transient analyses can be met. The existance of an inoperable accumulator may invalidate prior scram time measurements for the associated control rod. The scram function of the CDD System, and, therefore, the OPERABILITf of the accumulators, protects the MCPR Safety Limit (see Bases for LCO 3.2.2, " MINIMUM 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 (APLHGR)," and LCO 3.2.3,
' LINEAR HEAT GENERATION RATE (LHGR)'), which ensure that no fuel damage will occur if these limits are not exceeded (see Bases for LCO 3.1.4). Also, the scran function at low reactor vessel pressure (i.e., startup conditions) provides protection against violating fuel design limits during reactivity insertion accidents (see Bases for LCO 3.1.6,
" Rod Pattern Control").
(continued) BWR/6 STS B 3.1-27 Rev. O, 09/28/92
i Control Rod Scram Accumulators B 3.1.5 . 88"r5 APPLICABLE Control rod scram accumulators satisfy Criterion 3 of the SAFETY ANALYSES NRC Policy Statement. (continued) I LC0 The OPERABILITY of the control rod scram accumulators is required to ensure that adequate scram insertion capability , exists when needed over the entire range of reactor pressures. The OPERABILITY of the scram accumulators is ; based on maintaining adequate accumulator pressure. ( l
, i APPLICABILITY In MODES 1 and 2, the scram function is required for 'i b uc be reuh mitigation of DBAs and transients and, therefore, the scramaccum i l
f i in MODES 3 and 4, control rods are e:y C : -carto be nef L withdrawn fufufCrT.ffciai v^peration u .$ . t , = 5 ^ gi e; owe
' qg*g*
t rodo"gsw.dek ^ ' - SirJg
,s ]f*ontro(l Rdli W Con Edrawal Hot Sh own,"
and LCO drawa -Cold utdown whidU
.4
' i cedrA ted Med is pglindrJ adequate requirements for control rod scram '
OMg'j' g Ogj ~ accumulator OPERABILITY under these conditions. Requirements for scram accumulators in MODE 5 are contained (% C1f, in LCO 3.9.5, " Control Rod OPERABILITY-Refueling." ! j i ACTIONS The ACTIONS table is modified by a Note indicating that e i separate Condition entry is allowed for each control rod. This is acceptable since the Required Actions for each l Condition provide appropriate compensatory action for each ' GiiNMIM control rod.
% - yegrj Ct !aayallowforcontinuedoperationandsubsequentGaoperaUtt; Compl!
% control rods governed by subsequent Condition entry and !
application of associated Required Actions. ' t A.1 and A,2 g With one control rod scram accum.ul tor inopertble and the - reactor steam done psig, the control rod may ; be declared " slow," pressure tsince the control rod will still scram ~ at the reactor operating pressure but may not satisfy the ! OCT required scram times in Table 3.1.4-1mn = LC0h ! Required Action A.1 is modified by a Note, which clarifies (continued) { BWR/6 STS B 3.I.28 O, 09/28/92 Rev.
l Control Rod Scram Accumulators l B 3.1.5 I BASES ACTIONS A.] and A.2 (continued) I that declaring the control rod " slow" is only applicable if the associated control scram time was within the limits of Table 3.1.4-1 during the last scram time test. Otherwise, the control rod would already be considered " slow" and the further degradation of scram performance with an inoperable accumulator could result in axcessive scram times. In this event, the associated control rod is declared inoperable (Required Action A.2) and LCO 3.1.3 entered. This would result in raquiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function in accordance with ACTIONS of LCO 3.1.3. The allowed Completion Time of 8 hours is considered reasonable, based on the large number of control rods available to provide the scram function and the ability of the affected control rod to scram only with reactor pressure at high reactor pressures. B.I. B.2.1. and 8.2.2 With two or more control rod scram accumulators inoperable and reactor steam dome pressure t {UD psig, adequate pressure must be supplied to the charging water header. With inadequate charging water pressure, all of the accumulators could become inoperable, resulting in a potentially severe degradition of the scram performance. Therefore, within 20 minutes from discovery of charging water header pressure < 1520 psig concurrent with Condition 8, adequate charging water header pressure must be restored. The allowed Completion Time of 20 minutes is considered a reasonable time to place a CRD pump into service to restore the charging header pressure, if required. This Completion Time also recognizes the ability of the reactor pressure alone to fully insert all control rods. The control rod may be declared " slow,' since the control rod will still scram using only reactor pressure, but may not satisfy the times in Table 3.1.4-1. Required Action is.2.1 is modified by a Note indicating that declaring the control rod ' slow" is only applicable if the associated control scram time was within the limits of Table 3.1.4-1 during the last scram time test. Otherwise, the control rod (continued) BWR/6 STS B 3.1-29 Rev. O, 09/28/92 4 J
Control Rod Scrau Accumulators B-3.1.5 3 BASES ACTIONS B.1. B.2.1. and B.2.2 (continued) would already be considered " slow" and the further degradation of scram performance with an inoperable accumulator could result in excessive scram times. In this event, the associated control rod is declared inoperable-(Required Action B.2.2) and LCO 3.1.3 entered. This would result in requiring the affected control rod to be fully ' inserted and disarmed, thereby satisfying its intended function in accordance with ACTIONS of LCO 3.1.3. The allowed Completion Time of I hour is considered reasonable, based on the ability of only the reactor pressure to scram the control rods and the low probability of a DBA or transient occurring while the affected accumulators are inoperable. C.1 and C.2 (,00 . With one or more control rod scram accumut ators inoperable and the reactor steam dose pressure < GDpsig, the pressure supplied to the charging water header must be adequate to ensure that accumulators With the reactor steam done pressure <@[ remain charged. psig, the function of the accumulators in providing the scram force becomes much more important since the scram function could become severely degraded during : cepressurization event ~or at low reactor pressures. Therefore, immediat upon discovery of . 09 charging water header pressure 1520fpsig, concurrent with Condition C, all control rods associated with inoperable accumulators must be verified to be fully inserted. Withdrawn control rods with inoperable scram accumulators may fail to scram under these Iow pressure conditions. The associated control rods must also be declared inoperable within I hour. The allowed Completion Time of I hour is reasonable for Required Action C.2, considering the low probability of a DBA or transient occurring during the time the accumulator is inoperable. t The reactor mode switch must be immediately placed in the ' shutdown position if effy) Required Action and associated OC4- Completion Timegeannot be met. This ensures that all
- nuochded wA er \oss c{ Oc CRD pumf>
(fwecA Ad6s 8.1 ed C.13 (- 1 (c ntinued) , i A STS B 3.1-30 Rev. O, 09/28/92
j Control Rod Scram Accumulators B 3.1.5 BASES l ACTIONS M (continued) insertable control rods are inserted and that the reactor is in a condition that does not require the active function i (i.e., scram) of the control rods. This Required Action is modified by a Note stating that the Required Action is not applicable if all control rods associated with the inoperable scram accumulators are fully inserted, since the function of the control rods has been perfomed. SURVEILLANCE SR 3.1.5.1 REQUIREMENTS SR 3.1.5.1 requires that the accumulator pressure be checked every 7 days to ensure adequate accumulator pressure exists to provide sufficient scrara force. The primary indicator of accumulator OPERABILITY is the accumulator )ressure. A minimum accumulator pressure is specified, aelow which the capability of the accumulator to perfom its intended function becomes degraded and the accumulator is considered inoperable. The minimum aceuaulator pressure of 1520 psig-i is well below the expected pressure of 1750 psig @ i OP41 uveu eng (Ref. 2). Declaring the accumulator inoperable l when the minimum pressure is not maintained ensures that significant degradation in scram times does not occur. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account EW indications available in the control room. GiRh ; REFERENCES j 1. W ,' Sectio 4.3.2.5.Sh !
- 2. ~M)
AR did,Sectionk4.6.1.1.2.5.3]L L 3. wAt / Section)5.2.2.2[.3]L gg . i
- 4. fSAR Section 15.4.1 .
BWR/6 STS B 3.1-31 Rev. O, 09/28/92
)
l Rod Pattern Control i 6 3.1.6 l l B 3.1 REACTIVITY CONTROL SYSTEMS ' B 3.1.6 Rod Pattern Control l l BASES BACKGROUND Control rod patterns during startup conditions are 7 controlled by the operator and the rod pattern controller (RPC) (LCO 3.3.2.1, " Control Rod Bleck Instrumentation"), so DPo46 b that only specified control rod sequences and relative 0 3 1low - - Peer positions are allowed .over the operating range of all 2 4pu (tMP), control roas insertedte ;-;c;i "PA- The sequences effectively limit the potential amount of reactivity addition that could occur in the event of a control rod drop accident (CRDA). This Specification assures that tia control rod patterns are OQ\ consistent References,yw'1 2, and 3.h_the-ass.umptions o'; the CRDA analyse D,a n d 2 s APPLICABLE \arJD The analytical methods and assumptions used in evaluating - SAFETY AN S the CRDA are summarized in References 0, 2, and L CRDA Qt analyses assume that the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis. The RPC (LCO 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 sitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage, which could result in undue release of radioactivity. Since the failure consequences for U0, have been shown to be
- insi cant below fuel energy depositions of 300 cal /gm (Ref the fuel damage limit of 280 cal /gm provides %dS marg f safetyJrom_significant core damage, which would ' '
@ result in re)fase af radioa tivity (Refs.-C and4 evaluationsT8eff.'b+M CRDA resultingN n-a7eak Generic of a design basis CRDA (i.e., a el energy deposition of 280 cal /gm) have shown that if the peak fuel enthalpy remains below 280 cal /gm, then the maximum reactor ssure will be less than the required ASME Code limits (R Jg) hnd the calculated offs es will be well within t #
required limits (Re (continued) BWR/6 STS B 3.1-32 Rev. O, 09/28/92
Rod Pattern Control B 3.1.6 BASES 3 APPLICABLE @t Control rod patterns analyzed in eferencPifollowthe SAFETYANALYSESlbankedpo' .on withdrawal sequenc (BPWS) escribed in (contin . Reference he BPWS is applicabl from the cond1 ion of all contro s fully inserted to RTP (RefQ19 For the BPWS, the control rods are requi ed to be move in groups, with all control rods assigned to a specific group T required to be within specified banked positions (e.g., 9 l) between notches 08 and 12). The banked positions are
! defined to minimize the maximum incremental control rod lworths without being overly restrictive dyr[ng normal plant operation. The generic BPWS analysis (Ref also evaluated the effect of fully inserted, in@ operable control rods not in compliance with the sequence, to allow a limited number (i.e., eight) and distribution of fully inserted, q
inoperable control rods. Rod pattern control satisfies the requirements of Criterion 3 of the NRC Policy Statement. LCO Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the BPWS. This LCO only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3, " Control Rod OPERABILITY," consistent with the allowances for inoperable centrol rods in the BPWS.
~
(p 0 Ao APPLICABILITY In MODES I and 2, when THERMAL POWER is s % RTP, the CRDA q fisaDesignBasisAccident DBA) and, there ore, compliance
'I with the assumptions of safety analysis is required.
If When 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 g during a CRDA (Ref In HDDES 3, 4, and S, since the reactor is shut do d only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDN ensures that the consequences of a CRDA are acceptable, since the reactor will remain suberitical with a single control rod withdrawn. (continued) BWR/6 STS B 3.1-33 Rev. O, 09/28/92
Rod Pattern Control B 3.1.6 1 BASES (continued) ACTIONS A,1 and A.2 l With one or more OPERABLE 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. I Noncompliance with the prescribed sequence may be the result 5 of " double notching," drifting from a control rod drive b cooljngmitr4ransient, leaking scram valves, or a power [ reduction to 5 DQ% RTP before establishing the correct control rod pattern. 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) their to allow correct the affected control rods to be returned to 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. OPERABILITY of control rods is determined by compliance with i LCO 3.1.3; LCO 3.1.4, ' Control Rod Scram Times"; and j LCO 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 i during the time the control rods are out of sequence. l B.1 and B.2 ! If nine or 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 beyond their allowed position is allowed since, in general, insertion of ! control rods has less impact on control rod worth than s withdrawals have. Required Action 8.1 is modified by a Note (continued) BWR/6 STS ! B 3.1-34 Rev. O, 09/28/92
Rod Pattern Control B 3.1.6 BASES ACTIONS B,1 and B.2 (r.ontinued) g that allows the affected contro rods to be bypassed in RACS in accordance with SR 3.3.2.1 to allow insertion only. With nine or more OPERABLE control rods not in compliance with BPWS, the reactor mode switch must be placed in the shutdown position within I 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 I 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 SR 3.1.6.1 REQUIREMENTS The control rod pattern is verified to be in compliance with the BPWS at a 24 hout Frequency, ensuring the assumptions of the CRDA analyses are met. The 24 hour Frequency of this Surveillance was developed considering that the primary check of the control rod pattern compliance with the BPWS is performed by the RPC (LCO 3.3.2.1). The RPC provides control rod blocks to enforce the required control rod , Ob se nce and is required to be OPERABLE when operating at Fl s RTP. REFERENCES 1. Cericer Cycie 5aiety-Analy44e.
'I (M " Modifications to the Requirements for Control Rod 3I 1h-J Drop Accident Mitigating Systems " BWR Owners Group,
<s Se ion 15.4.9. hs //
94 %
" Dj NUREG-0979, "NRC Safety Evaluation Report Ofc GESSAR !! BWR/6 Nuclear Island Design, Docket
_ No. 50-447," Section 4.2.1.3.2, April 1983. D SJ, NUREG-0800, " Standard Review Plan," Section 15.4.9,
' Radiological Consequences of Control Rod Drop Accident (BWR)," Revision 2, JuIy 1981.
(continued) BWR/6 STS B 3.1-35 Rev. O, 09/28/92
Rod Pattern Control B 3.1.6 BASES REFERENCES N (continued) 10 CFR 100.11. " Determination of Exclusion Area Low Population Zonevand Population Center Distance." l (t NEDO-21778-A, " Transient Pressure Rises Affected pf - Fracture Toughness Requirements for Boiling Water Reactors," December 1978. 7% ASME, Boiler and Pressure Vessel Code. li3% NEDO-21231, " Banked Position Withdrawal Sequence'" January 1977 l l i l l i BWR/6 STS B 3.1-36 Rev. O, 09/28/92
]
SLC System B 3.1.7 l B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Centrol (SLC) System BASES l BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive xenon free state without l taking credit for control rod movement. The SLC System l satisfies the requirements of 10 CFR 50.62 (Ref. 1) on ' anticipated transient without scram (ATWS). The SLC System consists of a baron solution storage tank, two positive displacement pumps, two explosive valves, which are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged through the high pressure core spray system sparger. APPLICABLE The SLC System is manually initiated from the main control ; SAFETY ANALYSES room, as directed by the emergency operating procedures, if the operator believes the reactor cannot be shut down, or kept shut down, with the control rods. The SLC System is used in the event that not enough control rods can be 1 inserted to accomolish shutdown and cooldown in the nomal manaer. The SLC System injects borated water into the , reactor core to compensate for all of the various reactivity effects that could occt.r during plant operation. To meet , this cbjective, it is necessary to inject a quantity of C. . gg boron that oroduces a concentration oP660 ppe of natural l boron in the reactor core at 68'F. To allow for potential l 1eakage and imperfect mixing in the reactor system, an ! additional amount of boron equal to 25% g,Afe amount cited ,3 above is added (Ref. 2). The temperature versu M vo W h--d G@are concentration limits in Figure 3.1.7-1 Ha-tac accusenv4#w calculated such that the required concentration is 9 achieved accounting for dilution in the RPV with nomal ' water level and including the water volume in the residual heat removal shutdown cooling piping and in the recirculation loop piping. This quantity of borated solution is the amount that is above the pump suction 1 (continued) BWR/6 STS B 3.1-37 Rev. O, 09/28/92 4
SLC System B 3.1.7 BASES k APPLICABLE shutoff level in the boron solution storage tank. No credit SAFETY ANALYSES is taken for the portion of the tank volume that cannot be (continued) injected. The SLC System satisfies the requirements of the NRC Policy Statement because operating experience and probabilistic risk assessment have generally shown it to be important to public health and safety. LCO The OPERABILITY of the SLC System provides backup capability for reactivity control, independent of normal reactivity control provisions provided by the control rods. The OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the RPV, including the OPERABILITY of the pumps and valves. Two SLC subsystems are required to be OPERABLE, each containing an OPERABLE pump, an explosive valve and associated piping, valves, and instruments and controls to ensure an OPERABLE flow path. C2' -(no1E cme) APPLICABILITY In HODES 1 and 2, shutdown capabi itylis rec uired. In
= -- --
MODES 3 and 4, contrcl rods are Con',~y~~.'iicau to be withdrawn rhv c b teccIor anner 5 ciai Opwreuen i.C0 3.1 5, 25i gie Co trol a
'"gE *4?('O
s With wal t' Shut ," a 03. .4, " ' nia-a nte
,R Withdrgal-Co Shutd . ' wh* orovir eladequate SO,a od a l controls to ensure the reactor remains su'bcritical. In MODE 5, only a single control rod can be withdrawn from a cndec\ re1 ded- core cell containing fuel assemblier. Demonstration of s
M T1P J. 'Ig' g adequate SDN (LCO 3.1.1, " SHUTDOWN MARGIN (SDN)") ensures that the reactor will not become critical. Therefore, the previdrs
- SLC System is not required to be OPERABLE during these conditions, when only a single control rod can be withdrawn.
1 ACTIONS k_ ; N IftheborynsolutionconcbtrationislessthFthe i k) gP required 11gits for ATWS mit4ation but greater an the concentratic required for cold shutdown (origina licensing j basis),thec entration must bk restored to withi limits ! in 72 hours. I is not necessaryNnder these condit ons to ; u _ . _ _ _ _ _ ~ _ . _ . . (continued) BWR/6 STS B 3.1-38 Rev. O, 09/28/92 l l
SLC System B 3.1.7 BASES [ f - [ ACTIONS A.1 (continued) ~ N t l declare both capable SLC subsystems of erforming their orig inop)erable, since they areal design b Because of t low probability of ATWS event an that the 1 h
\ l SLC System cap bility still exists r vessel injec ion under these cond'tions, the allowed C letion Time 72 hours is accep ble and provides ade ate time to r tore
- concentration to wi in limits. The maxi m Completion .ime of 10 days ie. 511oue for this LC0 in the t. ent of multip q- ,
(W -. Condition e :tv ' l 1 O !f one SLC Gys+edubsystem (f7&t-Mitdition-fEfthe inoperable is inoperableQor subsystem must be rcoscra restored ctTdN 1gi < to OPERABLE status within 7 days. In this condition, the remaining OPERABLE subsystem is adequate to perform the shutdown function. However, the overall reliability is reduced because a single failure in the remaining OPERABLE i subsystem could result in reduced SLC System shutdown ; capability. The 7 day Completion Time is based on the availability of an OPERABLE subsystem capable of performing the intended SLC System function and the low probability of ( a Design Basis Accident (DBA) or severe transient occurring concurrent with the failure of the Control Rod Drive System , to shut down the plant.]I aximum pletion o'f ll - is & Tor Ini LCO in yt of ltipl3 If both SLC subsystems are inoperable @Ueasons-ether-ttiaiBj pq (CuiiditioR at least one subsystem must be restored to i , OPERABLE status within 8 hours. The allowed Completion Time of 8 hours is considered acceptable, given the low . probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor. ;
.l If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LC0 (continued) tWRffrMSkur3.V,,c\n B 3.1-39 Rev. O, 09/28/92
SLC System B 3.1.7 BASES ACTIONS (continued) does not apply. To achieve this 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 MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.7.1. SR 3.1.7.2. and SR 3.1.7.3 REQUIREMENTS SR 3.1.7.1 through SR 3.1.7.3 are 24 hour Surveillances, verifying certain characteristics of the SLC System (e.g., h .c the volume and temperature of the borated solution in the storage tank), thereby ensuring the SLC System OPERABILITY 04 4 ~ without disturbing nonnal plant operation. These tu h i6 P P p* Surveillances ensure the proper borated solution and w temperature, including the temperature of the pump suction piping, are maintained. Maintaining a minimum specified borated solution temperature is important in ensuring that the boron remains in solution and does not precipitate out in the storage tank or in the pump suction piping. The 24 hour Frequency of these SRs is based on operating experience that has shown there are relatively slow variations in the measured parameters of volume and temperature. SR 3.1.7.4 and SR 3.1.7.6 SR 3.1.7.4 verifies the continuity of the explosive charges in the injection valves to ensure proper operation will occur if required. Other administrative controls, such as those that Ifuit the shelf life of the explosive charges, must be followed. The 31 day Frequency is based on operating experience that has demonstrated the reliability of the explosive charge continuity. SR 3.1.7.6 verifies each valve in the system is in its correct position, but does not apply to the squib (i.e., explosive) valves. Verifying the correct alignment for - manual, power operated, and automatic valves in the SLC System flow path ensures that the proper flow paths will exist for system operation. A valve is also allowed to be in the nonaccident position, provided it can be aligned to (continued) BWR/6 STS B 3.1-40 Rev. O, 09/28/92
SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.4 and SR 3,1.7.6 (continued) ccdrefs, REQUIREMENTS the accident position from the control om, or local y by a dedicated operator at the valve / This is acceptable ' the SLC System is a manually initiated system. This Surveillance does not apply to valves that are locked, % sealed, or otherwise secured in position, since they were verified to be in the correct position prior to locking, sealing, or securing. This verification of valve alignment ' does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require ' any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct positions. The 31 day Frequency is based on engineering judgment and is consistent with the procedural controls governing valve operation that ensure correct valve positions. SR 3.1. 7.ft This Surve llance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure the proper concentration of boron exists in the storage tank. SR 3.1.7.5 must be performed anytime boron or water is added to the storage tank solution to establish that the boron solution concentration is within the specified limits.) This Surveillance must be___ performed anytime thevtemperzturr 2 70 7 is restored toAritnin the li=W M m:gre 3.1.7 F, to O(>9 " ensure no significant boron precipitation occurred. The 31 day Frequency of this Surveillance is appropriate because of the .latively slow variation of boron concentration between surveillances. SR 3.1.7.7 Demonstrating each SLC System pump evelops a flow rate a 41.2 gpa at a discharge pressure a psig ensures that pump performance has not degraded during the fuel cycle. This minimum pump flow rate requirement ensures that, when i combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test (continued) f BWR/6 STS B 3.1-41 Rev. O, 09/28/92 I
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SLC System , B 3.1.7 BASES SURVEILLANCE SR 3.1.7.7 (continued) ! REQUIREMENTS 1 confirm: one point on the pump design curve, and is ' indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failure < by indicating abnormal performance. The Frequency r,T this Surveillance is kin accordance with the Inservice Testing Program;eA 92 dy:], SR 3.1.7.8 and SR 3.1.7.9 These Surveillances ensure that there is a functioning flow path from the boron solution storage *.ank to the RPV, including the firing of an eglesive valve. The replacement charge for the explosive i:alve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. The pump and explosive valve tested should be alternated such that both complete flow paths are tested every 36 months, at alternating 18 month intervals. The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump desineralized water from a test tank through one SLC subsystem and into the RPV. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating - experience has shown these components usually pass the Surveillance test when performed at the 18 month Frequency; , therefore, the Frequency was concluded to be acceptable from ' a reliability standpoint. Demonstrating that all heat traced piping between the boron solution storage tank and the suction inlet to the injection pumps is unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution. An , acceptable method for verifying that the suction piping is unblocked is to pump from the storage tank to the test tank. The 18 month Frequency is acceptable since there is a low probability that the subject piping will to blocked due to precipitation of the boron from solution in the heat traced piping. This is especially true in light of the daily 1 (continued) BWR/6 STS B 3.1-42 Rev. O, 09/28/92
t SLC System B 3.1.7 BASES f [ SURVEILLANCE REQUIREMENTS SR 3.1.7.8 and SR 3.1.7.9 (continued)j g..'3 : SR 3.1.7.3. temperature verification of this piping quired by - detemined that the temperature of thisHowever, if, , below the specified minimum, tMe E: =; piping has fallen E x.c =ust be perfomed is restored once a +bwithin 14=4+<24 hours after the piping tempi 2th ef ";; : 3.1.:'-1. ; f (fi E 10 *F. ;
,M ' ' ' o ~sL i Enriched sodium _- \
b_a granular, enriched sodiumrate solution is made by mixing 'x k orate with water. Isotapic '\
\ tests on the granular sodium penta
\ actual B-10 enrichment must be performed pto verify the i r o addition
\perce
%ntage is t,eing used.to the SLC tank0 in atom order REFERENCES 1. j 10 CFR 50.62.
un
~
g 2. F5AR Sectionk9.3.5.3[ I I
=s=
=
r 5 t 1 BWR/6 STS B 3.1 43 Rev. O, 09/28/92
SDV vent and Drain Valves B 3.1.8 ' B 3.1 REACTIVITY CONTROL SYSTEMS I B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves BASES BACKGROUND The SDV vent and drain valves are normally open and discharge any accumulated water in the SDV to ensure that > sufficient complete volume scram. is available at all times to allow a valves close to contain reactor water.During a scram, the S The SDV consists of header piping that connects to each hydraulic control unit I (HCU) and drains into an instrument volume. There are two headers and two instrument volumes, each receiving approximatel discharges. y one half of the control rod drive (CRD) The two instrument volumes are connected to a common drain line with two valves in series. Each header is connected to a common vent line with two valves in series. The header piping is sized to receive and contain all the , water discharged by the CRDs during a scram. functions of the SDV are described in Reference 1.The design and ! APPLICABLE } SAFETY ANALYSES The Design Basis Accident and transient analyses assume all the control rods are capable of scramming. function of the SDV is to limit the amount of reactorThe primary coolant discharged during a scram. The accestance criteria for the SDV vent automatically to: and drain valves are that taey operate a. Close during scram to limit the amount of reactor coolant discharged so that adequate core cooling is maintained of 10 CFR 100 and offsite (Ref. doses remain within the limits 2); and ; b. Open on scraa reset to maintain the SDV vent and drain i path open so there is sufficient volume to accept the reactor coolant discharged during a scram. l 1
@---F+eut= Isolation of the SDV ca, also be accomplished by manual
- .at q D elosurt of the SDV valves.
the .11scharge of reactor coolant to the SDV can beAdditionally, terminated by scram reset or closure of the HCU manual isolation valves. For a bounding leakage case, the offsite i doses are adequate corewell within cooling the limits(Ref. is maintained of 103).CFR 100 (Ref. 2) an The SDV vent (continued) i BWR/6 STS B 3.1-44 Rev. O, 09/28/92
SDV Vent and Drain Valves B 3.1.8 BASES APPLICABLE I SAFETY ANALYSES (continued) during nomal plant operation to ensure the SD sufficient capacity to co during a full core scram.ntain the reactor coolant discharge To automatically ensure this capacity, a reactor scram (LCO 3.3.1.1, " Reactor Protection System (RPS) Instrumentation") is initiated if the SDV water level exceeds a specified setpoint. The setpoint is chosen such that volume insufficient all control rodsa are to accept inserted before the SDV has full scram. SDV vent Policy and drain valves satisfy Criterion 3 of the NRC Statement. LCO Theduring that, OPERABILITY a scram of all SDV vent and drain valves ensure the SDV vent and drain valves will Since the vent and drain lines are provided with in series, the single failure of one valve in the open system. will not impair the isolation function of the position Additionally, the valves are required to be open to ensure that freely at other a path is available for the SDV piping to drain times.
- ~
APPLICABILITY In MODES 1 and 2, scram &-- not cme) 9.uc he rtac4cr m \ SDV vent and drain valves must be OPERABLE.may be required and 4, control rods are e" > : u In MODES 3
" g gy W a j gSpecia Wit. awal-perati s Lev Im s)to be withdrawn
'iu.3,pingle Lojttroi #$oo i
fbWW M ^ Withd Shu wn," arjd LCO 3.19A l al-Co d Shutdown
- 44evnrov,idf)ade'entate'SJagf P
- ml/
1 "9 A rd Nk is rols to smA } co(hErawnf~ATso~,q that snly a ogtrg] singlNecodjhde/ c4Ph d m j can 3LlFliig MODE 5;~only~a~ sin _gl~e control rod e withdrawn from a core cell containing fuel assemblies. g/ Therefore, the SDV vent and drain valves are not required to be OPERABLE in these MODES since the reactor is scram. to suberitical and only one rod may be withdrawn and subject ACTIONS The ACTIONS table is modified by a Note indicating that a k separate Condition entry is allowed for each SDV vent and ~j drain line. This is acceptable, since the Required Actions 1 (continued) BWR/6 STS B 3.1-45 Rev. O, 09/28/92 1
SDV Vent and Drain valves < B 3.1.8 BASES ACTIONS (continued) for each Condition provide appropriate compensatory actions for each inoperable SDV line. Complying with the Required Actions may allow for continued operation, and subsequent inoperable SDV lines are governed by subsequent Condition entry and application of associated Required Actions. ; A.1 When one SDV vent or drain valve is tperable in one or more 7lines, days.the valves must be restored to OPERABLE status within The Completion Time is reasonable, given the level of redundancy in the lines and the low probability of a I scram
,,] o ._ occurring L m-- during the time the valves)are inoperable
- mrm. The SDV is still isolable since the redundant valve in the affected line is OPERABLE. '
During these periods, the single failure criterion not be preserved, and a higher risk exists to allow reactor . water out of the primary system during a scram. Dd @[3 If both valves in a line are inoperable, the line must be isolated to contain the reactor coolant during a scram. When a line is isolated, the potential for an inadvertent scram due to high SDV level is increased. Required Action B.1 is modified by a Note that allows periodic i nddm]we draining of the SDV when a line is isolated. During these periods, the line may be unisolated under administrative C " 4. t5 N8j ) ce control. This allows any accumulated water in the line b vaivef u b s s u e r M to be drained, to preclude a
. reactor scras on SDV high level.. This is acceptable, since Ct, theveha as xratee-km :h: :=t ci r=P_.w can be
__A closed quickly4tfa a scram occurs with the valve open. , t
- 63cdeditc,4d The 8 hour Completion Time to isolate the line is based on of4rder s the low probability of a scram occurring while the line is
-- not isolated and unlikelihood of significant CRD seal 1eakage. ,
Ed If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LCO (contifiued) BWR/6 STS B 3.1-46 Rev. O, 09/28/92 '
SDV Vent and Drain Valves i B 3.1.8 i i BASES ; ACTIONS [21 (continued) does not apply. i brought to MODE 3 within 12 hours.To achieve this status, the i 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. : SURVEILLANCE SR 3.1.8.1 REQUIREMENTS During normal operation, the SDV vent and drain valves should be in the open position (except when performin i SR 3.1.8.2) to allow for drainage of the SDV piping. g ' Verifying that each valve is in the open position ensures that the SDV vent and drain valves will perform their intended function during normal operation. This SR does not require any testing or valve manipulation rather, it involves position. verification that the valves are;in the correct : judgment and is consistent with the procedural con - governing valve operation, which ensure correct valve positions. Improper valve position (closed) would not affect the isolation function. , SR 3.1.8.2 i During a scram, the SDV vent and drain valves should close Cycling each valve through'its complete range of (closed and open) ensures that the valve will function i properly during a scram. JThe 92 day Frequency is based on ! operating experience and takes into account the level of redundancy in the syst=m design. , 4 SR 3.1.8.3 ! SR 3.1.8.3 is an integrated test of the SDY vent and drain valves to verify total system performance. After receipt of a simulated or actual scram signal, the closure of the SDV d/ vent and drain valves is verified. The closure time of
' {301"leconds after a receipt of a scram signal is based on $
the bounding leakage case evaluated in the accident ) j (continued) 1 BWR/6 STS B 3.1-47 Rev. O, 09/28/92
.I SDV Vent and Drain Valves :
B 3.1.8 8ASES l SURVEILLANCE SR 3.1.8.3 (continued) REQUIREMENTS l analysis. Similarly, after receipt of a simulated or actual- i scram reset signal, the opening of the SDV vent and drain. ; valves is verified. The LOGIC SYSTEM FUNCTIONAL TEST in
' g,J, j bd LCO 3.3.1.1 and the scram time testing of control rods in LCO 3.1.3+ overlap this Surveillance to provide complete._ l OnRAtitlTVh testing of the assumed safety function. The.18 month '
Frequency is based on the need to perform this Surveillance i OCC'
/
under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance j
- were performed with the reactor at power. Operating ;
experience has shown these components usually pass the , Surveillance when performed at the 18 month Frequency; : therefore, the frequency was concluded to be acceptable from a reliability standpoint. I t REFERENCES 1.
}
Section 4.6.1.1.2.4.2 . -
- 2. 10 CFR 100.
3. i NUREG-0803, " Generic Safety Evaluation Report- i, Regarding Integrity of BWR Scram System Piping," August 1981. : i i 1 l l l I BWR/6 STS B 3.1-48 Rev. O, 09/28/92 i i
i ATTACHMENT 2B 1
)
i ITS - PSTS r COMPARISON DOCUMENT : DISCUSSION OF CHANGES i l r t i i l
'i
l l DISCUSSION OF CHANGES TO NUREG-1434 ; 3.1.1 - SHUTDOWN MARGIN i BRACKETED ADMINISTRATIVE CHOICE B.1 Brackets removed and optional wording preferences revised to. > reflect appropriate plant specific requirements. l t PLANT SPECIFIC DIFFERENCE P.1 The safety aw lysis report for this station is identified as the Updated Safety Analysis Report and is correctly referred to l as the USAR. j P.2 This comment number is not used for this station. ; I P.3 This comment number is not used for this station. l t P.4 The intent of the ITS ACTIONS is to establish a fission product . f boundary. The Perry specific design requires including the l primary containment air locks to ensure an adequate boundary. : Also, this in conjunction with restoring primary containment to , 3 OPERABLE status, and restoring required primary containment isolation capability, precludes the need for requiring AEGTS to j be operable. With the Perry specific primary containment boundary, adequate fission product containment is achieved. .l l P.5 For the Perry specific l-icensing basis, neither Reference 5 or -(; other references reviewed confirmed this " design process" bases. ; i CHANGE / IMPROVEMENT TO NUREG STS , C.1 The Required Actions and Completion Times for Conditions C and ! D hav3 been revised for consistency with other similar Actions- 'j such as: E.2 in this Specification; H.1 - of proposed , Speci fication 3.3.1.1; C.2 of proposed Specification 3.3.2.1; j and throughout Sections 3.9 and 3.10. The BASES are also revised to reflect these consistency changes. C.2 This editorial revision is made for consistency with other references to this document, e.g., B 3.1.3 Ref. 7 and B 3.1.6- ; Ref. 8. ; I C.3 This editorial correction is made to appropriately identify the. + reference. ! i PERRY - UNIT 1 1 10/1/93 , i
. .-. . . .~ . . . ._ -
i i DISCUSSION OF CHANGES TO NUREG-1434 ! 3.1.1 - SHUTDOWN MARGIN ' i CHANGE /IMPROVFAENT TO NUREF STS l (continued) C.4 This editorial revision is made to place the descriptive phrase with the context of the discussion which it describes. The intent of the wording is to exclude fuel movement outside of' ; the reactor pressure vessel. The phrase does not modify ?
" control rod replacement" since this activity can only occur i within the vessel. The Bases description is also revised to ,
reflect this revision. t C.5 This reference to the Bases for LCO 3.1.7 is omitted since the reference Bases do not discuss the topic of "significant fuel damage resulting in undue release of radioactivity." -; a C.6 These revisions clarify when the requirement for demonstration .! by calculation is acceptable, and that the additional margin ; being discussed has already been incorporated into the limits- ! and is not an activity that must be accomplished when in the ! identified condition. C.7 The sentence is revised to clarify that the identified l suspensions are examples and may not include all CORE l ALTERATIONS that must be suspended. ; C.8 This is an editorial revision for consistency within the Bases. I "Shall be performed" is wording that is appropriate for a ! requirement, but not as a description of that requirement. i C.9 Just as during power operations, some unisolated penetrations 1 may be administratively open with _ capability for closure. These isolation valves would not have instrumentation required : to be OPERABLE. Other penetrations may utilize isolation-devices other than " valves." To accommodate these variations i more generic wording is editorially chosen. C.10 Not all primary containment penetrations are assumed to be isolated (or isolatable) . Only those penetrations that are
" required" to perform the assumed function are intended to be ;
addressed. This is an editorial clarification of this issue. C.11 Inadvertent criticalities are inherently criticalities which I occur inspite of intended protection against them. Given this, SDM can not " ensure" no damage from inadvertent criticalities; SDM can " provide assurance" against significant damage. a i t PERRY - UNIT 1 2 10/1/93 : 1
e DISCUSSION OF CHANGES TO NUREG-1434 3 1.1 - SHUTDOWN MARGIN ch CHANGE / IMPROVEMENT TO NUREG STS (continued) C.12 The ACTIONS associated- with these revised Bases invoke j OPERABILITY requirements because the LCOs for the systems being required OPERABLE are not (or may not be) Applicable at this ! time. Therefore, reference to the LCOs for those systems that- ! are being required OPERABLE by ACTIONS, is inappropriate. The l LCO itself still remains not Applicable. This is only ' an ! editorial enhancement to avoid the potential for mis- ; application of those LCOs. '[ 1 i t i I t B r i f i 6 r I f I b I- i PERRY - UNIT 1 3 10/1/93 i f
s DISCUSSION OF CHANGES TO NUREG-1434
- 3.1.2 - REACTIVITY ANOMALIES BRACKETED ADMINISTRATIVE CHOICE B.1 Brackets removed and optional wording preferences revised to reflect appropriate plant specific requirements. .;
l 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. 5 P.2 This editorial revision is to make the Bases wording consistent ! with the LCO for those plants using GE fuel. All plants using ; GE. fuel still perform the reactivity anomaly surveillance by comparing monitored rod densities with predicted rod densities. i The wording is consistent with NUREG-1434. I P.3 This comment number is not used for this station. P.4 This comment number is not used for this station. ! CHANGE / IMPROVEMENT TO NUREG STS , C.1 The Surveillance is modified to reflect the inability to obtain
- useful results for this comparison at very low power levels. ,
Without this modification, the frequency may require' the ; surveillance to be conducted while operating at very low power
- levels, i.e. , MODE 2, during which the comparison would provide results with questionable validity and accuracy. The NUREG-1434 Bases presently reflect the need to conduct the surveillance in MODE 1 (during POWER OPERATION). '
C.2 This editorial revision is made to place the descriptive phrase with the context of the discussion which it describes. The intent of the wording is to exclude fuel movement outside of a the reactor pressure vessel. The phrase does not modify *
" control rod replacement" since this activity can only occur ,
within the vessel. ; C.3 "
...at RTP" adequately encompasses operating temperature '
without separate mention. Editorial change only. : C.4 "Every accident" is not necessarily dependent- ' o.1 core reactivity (such as fuel handling accident). Excessive i verbiage eliminated. Sufficient detail remains. , i PERRY - UNIT 1 4 10/1/93
DISCUSSION OF CHANGES TO NUREG-1434 3.1.3 - CONTROL ROD OPERABILITY 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 number of control rods is dependent on plant reactor vessel design. Fo" PNPP the correct number of control rods is 177. P.3 ITS Condition E is not applicable to GE fuel design and analysis. The remaining Conditions are renumbered and the Bases are appropriately revised. P.4 The plant specific references are provided and renumbered as necessary. P.5 The power level for the Low Power Setpoint in the LCO is in brackets, but does not appear in brackets throughout this section of the Bases. The power level is being changed in the i Bases to make it consistent with the LCO. ~l CHANGE / IMPROVEMENT TO NUREG STS C.1 The reference to this Surveillance is corrected. The original 'l surveillance no longer exists. C.2 This terminology revised to be consistent with discussions of the SDM provided in the Bases for LCO 3.1.1. 1 C.3 This discussion of APPLICABILITY revised for consistency with general fonnat of NUREG-1434 which discusses Special Operations exceptions only in Section 3.10. C.4 The time allowed to disarm an inoperable CRD is specified as l 2 hours in Required Action A.1. This time was found acceptable based on the time required to make the containment' entry and complete the procedure without undo haste. Since " separate Condition entry" is allowed (ACTION Note), the 1 hour Completion Time of Required Action B.1 is also seen to conflict with the 2 hour Completion Time of Required Action A.1. This proposed change resolves this inconsistency. PERRY - UNIT 1 5 10/1/93
m - ._ . _ _ . _ _ _ _ __ _ - - . . __. . _ . _ . . _ . . _ . _ DISCUSSION OF CHANGES TO NUREG-1434 3.1.3 - CONTROL ROD OPERABILITY , i CHANGE / IMPROVEMENT TO NUREG STS ; (centinued)- ; i C.5 Since the Condition involves separation requirements, one
- inoperable control rod is inherently not at issue. The -i preferred presentation, therefore, is to address "two or more :
inoperable control rods". ' C.6 Correct values are provided consistent with the associated LCO + requirement. j C.7 Missing Bases discussion related to the LCO's consideration of l only OPERABLE control rods is provided. ! C.8 A more detailed reference to the specific requirements is provided by this preferred wording. C.9 The title of the LCO is provided at its first use as a. I reference in accordance with the accepted format. i C.10 Many f actors influence a statement such as " assumptions for. . scram reactivity in the DBA and transient analysis are not violated." The capability of inserting control rods alone can not ensure these assumption. Insertion capability can " provide assurance." . l i 1 PERRY - UNIT 1 6 10/1/93 l
DISCUSSION OF CHANGES TO NUREG-1434 + 3.1.4 - CONTROL ROD SCRAM TIMES j l BRACKETED ADMINISTRATIVE CHOICE l B.1 Brackets removed and optional wording preferences revised to reflect appropriate plant specific requirements. 1 PLANT SPECIFIC DIFFERENCE i P.1 The safety analysis report for this station is identified as l the Updated Safety Analysis Report and is correctly referred to as the USAR. ? P.2 This comment number is not used for this station. , i P.3 The NUREG presentation allowing "No more than 2" adjacent slow ; control rods can not be supported by the current plant specific licensing / design basis. CHANGE / IMPROVEMENT TO NUREG STS ! C.1 The "immediate" steam pressure is corrected to " intermediate" : to provide appropriate direction for use of the table. ! C.2 This phrase is redundant to the first part of the sentence _and , is potentially confusing. j C.3 This discussion of APPLICABILITY revised for consistency with { general format of NUREG-1434 which discusses Special Operations ; exceptions only in Section 3.10. : C.4 The appropriate editorial presentation for a requirement to :
" cascade" is provided with this change.
C.5 The "O psig" scram times are proposed.to be relocated to plant l procedures - note the Surveillance which requires the scram time testing that utilizes these limits remains within the TS (SR 3.1.4. 3 ) . The proposed change reflects the imposition of an additional limitation. The scram time limit does not vary linearly from 0 psig to 950 psig. The NUREG presentation format is therefore inadequate for presenting the limit. Portions of the Bases insert'are consistent with the changes proposed to ITS lable 3.1.4-1. The insert also proposes additional clarification regarding the implementation of Table-J 3.1.4-1 Note 2.in relation to satisfying the SR requirement of .
" scram time is within the limits of Table 3.1.4-1." i l
C.6 Missing Bases discussion related to the LCO's consideration of l only OPERABLE control rods is provided. ! PERRY - UNIT 1 7 10/1/93'
'i i
DISCUSSION OF CHANGES TO NUREG-1434 - 3.1.4 - CONTROL ROD SCRAM TIMES CHMJGE/ IMPROVEMENT TO NUREG STS (continued) 1 C.7 The NUREG Bases lacked a basis for the requirement. The l appropriate basis is provided. i I C.8 Clarification necessary to avoid mis-reading the statement as requiring 20% of each sample to include previously determined ;
" slow" control rods.
{ I, h 9 k i 6 i
+
-t i
i I i
+
l i l i PERRY - UNIT 1 8 10/1/93
- l DISCUSSION OF CHANGES TO NUREG-1434 3.1.5 - CONTROL ROD SCRAM ACCUMULATORS i BRACKETED ADMINISTRATIVE CHOICE i
i B.1 Brackets removed and optional wording preferences revised to i reflect appropriate plant specific requirements. l l i PIANT SPECIFIC DIFFERENCE i P.1 The safety analysis report for this station is identified as ' the Updated Safety Analysis Report and is correctly referred to l as the USAR. . P.2 This comment number is not used for this station. P.3 As indicated in USAR Sections 4 . 6.1.1. 2 . 5 . 3 and 4 . 6. 2 . 3 . 2. 3, i 600 psig reactor pressure is sufficient to ensure that the , control rods will insert without supporting accumulator : pressure. However, as indicated, the insertion times may be - slower and may not support the required scram times used in the safety analysis. Since this is the identified basis for the bracketed pressure to be used in the proposed LCO 3.1.5, this proposed pressure is in accordance with the approved safety i analysis. ; P.4 This range of expected accumulator pressure is not provided in i the referenced USAR section, j CHANGE / IMPROVEMENT TO NUREG STS l C.1 No "other" indications were previously discussed. Therefore, this is only an e646.cial correction. ' C.2 The Specification is directed to inoperable accumulators, not-inoperable control rods. Therefore, "affected control rods" i provides a more appropriate designation of the equipment of [ concern. > C.3 This discussion of APPLICABILITY revised for consistency with - general format of NUREG-1434 which discusses Special Operations exceptions only in Section 3.10. C.4 All Required Actions other than B.1 or C.1 involve the ACTION j
" declare . . . inoperable" or " declare . . . slow. " These ACTIONS' ;
can never be "not met." Therefore, the proposed revision i provides an editorial clarification of the intent. ! C.5 The Table number adequately reflects the location of the referenced information. , i
)
PERRY - UNIT 1 9 10/1/93 i
DISCUSSION OF CHANGES TO NUREG-1434 3.1.6 - ROD PATTERN CONTROL BRACKETED ADMINISTRATIVE CHOICE' B.1 Brackets removed and optional wording preferences revised to reflect appropriate plant specific requirements. PLANT SPECIFIC DIFFERENCE P.1 The reference to a specific Current Cycle Safety Analysis is-not applicable since this document has been incorporated into the USAR. The other references have been appropriately renumbered. ; P.2 The safety analysis report for this station is identified as f the Updated Safety Analysis Report and is correctly referred to ; as the USAR. P.3 The plant specific references are provided. ; P.4 The power level for the Low Power Setpoint in the LCO is in. i brackets, but does not appear in brackets throughout some parts of this section of the Bases. The power level is being ch.4ed in the Bases to make it consistent with the LCO. ; i P5 Reference title corrected. CHANGE / IMPROVEMENT TO NUREG STS None in this Section .; t l l 1 t f i h i I
)
PERRY - UNIT 1 10 10/1/93 i r
DISCUSSION OF CHANGES TO NUREG-1434 3.1.7 - STANDBY LIQUID CONTROL SYSTEM BRACKETED ADMINISTRATIVE CHOICE B.1 Brackets removed and optional wording preferences revised to - reflect appropriate plant specific requirements. l f PLANT SPECIFIC DIFFERENCE ! i P.1 The safety analysis _ report for this station is identified as' ! the Updated Safety Analysis Report and is correctly referred to i as the USAR. l P.2 The sentence is made plant specific to describe actual design-of the system. ; P.3 This comment number is not used for this station. ,
~
1 P.4 This comment number is not used for this station. P.5 This comment number is not used for this station. ! a i P.6 Wording changed to make this consistent with plant specific elements of Figure 3.1.7-1. .j i P.7 The value for the pump discharge pressure in the ICO is.in .,. brackets, but does not appear in brackets throughout some parts l of this section of the Bases. The discharge pressure is being l changed in the Bases to make it consistent with the LCO. , i ! P.8 This comment numner is not used for this station. l I P.9 The Bases are revised to be. consistent with the LCO.
. i i i P.10 This comment number is not used for this station.
i P.11 Deletion of the bracketed information of Condition A required -j the renumbering and revision of the remaining Conditions and { Bases. l CHANGE / IMPROVEMENT TO NUREG STS l
)
C.1 This change provides a consistent discussion of how multiple , entries are possible. , C.2 The operator is to be positioned at the valve controls rather than at the valve. l i PERRY - UNIT 1 11 10/1/93 : I
f DISCUSSION OF CHANGES TO.NUREG-1434 i 3.1.7 - STANDBY LIQUID CONTROL SYSTEM l i CHANGE / IMPROVEMENT TO NI"'2G STS (continued) I 1 C.3 This discussion of APPLICABILITY revised for consistency with { general format of NUREG-1434 which discusses Special Operations i exceptions only in Section 3.10. ; I C.4 The Table number adequately reflects the location of the referenced information. ! C.5 This change provides an obvious editorial correction. j C.6 This comment number is not used for this station. j i i i i. t
)
i
+
y 1 f 4 i I t I f f I i PERRY - UNIT 1 12 10/1/93 !
i DISCUSSION OF CHANGES TO NUREG-1434 l 3.1.8 - SDV VENT and DRAIN VALVES i BRACKETED ADMINISTRATIVE CHOICE' i B.1 Brackets removed and optional wording preferences revised to. j reflect appropriate plant specific requirements. I J PLANT SPECIFIC DIFFERENCE- ! P.1 The safety analysis report for this station is identified as 'I the Updated Safety Analysis Report and is correctly referred to : as the USAR. l CHANGE / IMPROVEMENT TO NUREG STS l r C.1 An obvious typographical error is corrected. l l C.2 This unnecessary delineation of the type of device for manual- l closure is deleted. Several acceptabla types may be used and j it is not important to safety analysis, i C.3 This discussion of APPLICABILITY revised for consistency with' l general format of NUREG-1434 which discusses Special Operations exceptions only in Section 3.10. l t C.4 An inoperable SDV valve may exist if the valve is closed, as ' well as when it is open and won't close. The misleading Bases I discussion is clarified. - C.5 The title of the LCO is provided at its first use. as a reference in accordance with the accepted format. ! C.6 Water accumulation can be from a variety of sources. Seal l leakage should not be an expected source, unless the scram outlet valve is also leaking. l t 4 i > t I i h L PERRY - UNIT 1 13 10/1/93 [
c#
,l ,
f ' f 4 N PERRY - UNIT 1 SECTION 3.2 3
- l 1
J e i J l J 4 4 d 1
i l ATTACHMENT 1 i 1 l l I CTS - PSTS
. COMPARISON DOCUMENT )
i r 5 1A: MARKUP OF CTS : 1B: DISCUSSION-OF CHANGES i 1C: NO SIGNIFICANT HAZARDS CONSIDERATIONS E
F . i t ATTACHMENT 1A i CTS - PSTS ! COMPARISON DOCUMENT : MARKUP OF CTS l t h p i e t t b l l l 9 I
t i i POWER DISTRIBUTION LIMITS 3/4.2 POWER DISiRIBUTION LIMITS ' 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE i LIMITING CONDITION FOR OPERATION I All AVERAGE PLANAR L'INEAR HEAT GENERATION RATES (APLHGRs) shall not N }l3.2.1exceed the limits specified in the CORL OPERATING LIMITS REPORT. IF:.MTIOJjaLCONDITION-- hen THERMAL POWER is greater than or equal to 25% g APPLICABILITY: oTRATED THEilIMOWJEE[r ACTION: C.mj0 A If at .any time during operation it is-determined that an APLHGRitexceeding
^
the limits soegified in the_ CORE OPERATING LIMITS REPORT / initiate corrective 7 ; gtion_ within 15 minutesfand restore APLHGR to withTn th'e required =fimits within i
@2 hours or reduce THERMAL POWER to less than 25% of RA Copof> the next 4 hours.
SURVEILLANCE REQUIREMENTS 58 'L4.2.1 All API Hf;Rt shal_1 he verified to be equal to or less than the above 3LLI A limitsrQLJ i. h
- a. @g 1 Et CI : he N 4 hours 9 h rco N % ,
- b. Within 12 hours after gomp'icticr?fl' RMAL POWER '------^ ^' '* '
leart :% of RATED THEPML F0'JERkcac P.;;r and !
- c. Initially and at least once per 12 hours when the reactor is
, operating with a LIMITING CONTROL ROD PATTERN for APLHGR. .
Q
~
h .y The provisions.chec1f_ication 4.0.4 are not applicaliTCA l
\
I J PERRY - UNIT 1 3/4 2-1 Amendment No. 20,33 I
- o _ ._ - - . . . .. _
POWER DISTRIBUTION LIMITS 3/4.2.2 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION Mo ' I 3. 2. 2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or ter e 3.1.1 than the MCPR. limits specified in the CORE OPEPAUNG LIMITS REPQRIJat the findica ted ;s. ficw, T"E.W.L - POWC, ;T and cer; cvci e ve eipusui u ces , i p C..u v i ;jJ. e Ewusui m ( EOGE4". ..f Q APPLICABILITY: equal to 25% of RATED D ESTIO"AL THEMifL'~ POWER. CONDITIO"A1 HERMAL POWER is greater than or ACTION: C o /JD A With MCPR_less than the limits specified in the CpRE OPERATING LIMITS REPORT, iti;:: arrectiec action ithS 15 "=t:;Erestore MCPR to within the required limit within 2 hours or reduce THEPJiAL POWER to less than 25% of capt, B RATED THEPJ4AL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS
'4.2.2 MCPR shall be determined to be equal to or greater than the limits 31d specified in the CORE OPERATING LIMITS REPORT:
a.' At least once per 24 hours, ^ - -
")
i 1 M h- n
- b. , Q - Q TED THERMAL POWER Q e.ic h @ ndWithin 12 Aours j
~
. (c Initi$liknMeasf once per 12 hour +-wheTr t!Trreadu[ o h iMO N b th-a-L.IMI-TING-00ftTROL--R00-PAWERN--f-oWRR, wi- '
. === .. ::::. ._ _,. ..-..+:.w.- ~=~- ! '
The pmviriens-of--Specifitet-ion-4r0v4-are aot-.appHeabi j l ( @ i TT~ refers to the planned reduction of rated feedwat temperature from - t ,.td.
\
nominal rated feedwater temperature (420 F), such as prolonged removal of feedwater heater (s) from service. ;<(
'**End of Cycle. Exposure (EOCE) is defined as'l) the core average exposures at which there is no longer sufficient reactivity,to achieve RATED THE MAL l
/
\
POWER with rated core flow, all control rods withdrawn, all feedwate s / heaters in service \and equilibrium Xenon, or 2) as'specified by the fu'el
'yendor l
l PERRY - UNIT 1 3/4 2-2 Amendment No. 20, 33 l 1 1 l
POWER DISTRIBUTION LIMIT 3/4.2.3 LINEAR HEAT GENERATION RATE 11 0 3,7,3 LIMITING CONDITION FOR OPERATION ^ 3.2.3 -The LINEAR HEAT GENERATION RATE.(LHGR) shall not exceed the limit ' specified in the CORE OPERATING LIMITS REPORT. eh% h APPLICABILITY: U k"'"M N"nf m " n henw THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTION: CODL A
With the LHGP, of any fuel rod-exceeding the limit, 6tiatocor =+%e ?ctien) 8tM_M _T ^ur; an?) restore the LHGR to within the limitswithin 2 hours or Mb reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours.
lSURVEILLANCEREQUIREMENTS ER 3 % 2.3 LHGR's shall be determined to be equal to or less than the limit: a'. At least once per 24 hours,
- b. Within 12 hours after f^mpiction of
- ~~
f;;n !%% RATED THERRAL POWER end no"@JTHERMAL W and POWER $iWor-ea Initially and at least once per 12 hours when the reactor is t [~ c. operating on a LIMITING CONTROL R0D PATTERN for LHGR. y The provisions'of specifWtion 4.0.T are not applicab" PERRY - UNIT 1 3/423 Amendment No. 20,33 l 1
F ATTACHMENT 1B CTS - PSTS COMPARISON DOCUMENT DISCUSSION OF CHANGES 1 l l l l
DISCUSSION OF CHANGES CTS: 3.2 - POWER DISTRIBUTION LIMITS ADMINISTRATIVE I A.1 With the THERMAL POWER 2 25% RTP, the unit will always be in MODE 1. Therefore, it is unnecessary to state in the Applicability. . A.2 The provisions of the statement on SR 4.0.4 applicability are incorporated into the new surveillance frequency and no longer , need to be explicitly stated. , RELOCATED SPECIFICATIONS I R.1 This comment number is not used for this station. TECHNICAL CHANGE - MORE RESTRICTIVE None in this section. , TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" l LA.1 The requirement to " initiate corrective action within l 15 minutes" is relocated to the Bases in the form of a ,
discussion that " prompt action" should be taken to restore the l parameter to within the limits. Immediate action may not ; always be the conservative method to assure safety. The 2 hour : completion time allows appropriate actions to be evaluated by ; the operator and completed'in a timely manner. 1 LA.2 These details are being relocated to the COLR report. The [ relocated words deal with specifics more appropriate to be excluded from the improved Technical Specifications, and located in the specific program, in this case the COLR.- LA.3 This comment number is not used for this station.- [ LA.4 This comment number is not used for this station. 0 6 9 PERRY - UNIT 1 1 10/1/93 i
DISCUSSION OF CHANGES CTS: 3.2 - POWER DISTRIBUTION LIMITS TECHNICAL CHANGE - LESS RESTRICTIVE (continued)
" Specific" L.1 This change eliminates confusion as to how often the current surveillance is required (e.g., af ter every 15% power change or at the end of any single power increase greater than 15%.)
Verifying the parameter within 12 hours of reaching or exceeding 25% RTP will generally require that the surveillance be performed sooner than "after completion of a 15% power-increase," but would also reduce the number of times the surveillance must be conducted during a startup if it is currently conducted after every 15% power change. A single verification is considered sufficient during initial startup considering the large inherent margin to operating limits at low power levels. Following the initial verification, the surveillance is performed every 24 hours to identify any trends in these parameters that may lead to long term noncompliance. L.2 Since a Limiting Control Rod Pattern is currently defined as operating on a power distribution limit such as APLHGR or MCPR, the condition is extremely unlikely and the surveillance would almost never be required. Additionally, the initial surveillance is superfluous as it would not be evident that a limiting control rod pattern has been achieved until the surveillance is performed. 4 : A I PERRY - UNIT 1 2 10/1/93
V ATTACHMENT 1C CTS - PSTS COMPARISON DOCUMENT NO SIGNIFICANT HAZARDS CONSIDERATIONS r a
NO SIGNIFICANT HAZARDS-CONSIDERATIONS CTS: 3.2 - POWER DISTRIBUTION LIMITS "L1" CHANGE I PNPP 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: j
- 1. Does the change involve a significant increase in the ,
probability or consequences of an accident previously evaluated? i The change to the surveillance Frequency will ' require ' the , verification of power distribution limits only once during low- : power operations with periodic reverification to identify : trends. The power distribution limits are used to verify the ! unit is operating within the initial assumptions of the safety ' analyses. Significant changes in these_ parameters are ; indicative of unanticipated operation, but .are not, in themselves, identified as . initiators of any previously analyzed ( accident. Therefore, the change in Frequency of the- : surveillance will not significantly increase the probability of' an accident previously identified. At low power, there.are las.ge inherent margins to these operating limits and during ; normal operation, change in the power distribution parameters , is slow. Therefore, the proposed frequencies are sufficient to ! assure the parameters remain within limits and the change does not significantly increase the consequences of a previously ; evaluated 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 ! nor does it require physical modification to the plant. Therefore, the change does not create the possibility of a new .i or different kind of accident from any accident previously i evaluated. !
- 3. Does this change involve a significant reduction'in a margin of safety? j This change has no impact on any safety analysis assumption since the verification of operation within the parameter limit is still required and is consistent with those assumptions. !
The proposed Surveillance Frequency has been determined through i engineering judgement. to be adequate for assuring the
.[
parameters do not exceed the limits. Therefore, the change : does not involve a significant reduction in a margin' of safety. j PERRY - UNIT 1 1 10/1/93 l J
i NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 3.2 - POWER DISTRIBUTION LIMITS : f "L2" CHANGE , PNPP has evaluated this proposed Technical Specification change and r has determined that it involves no significant hazards i 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: I
- 1. Does the change involve a significant increase in the ;
consequences probability or of an accident previously- ' evaluated? j i The deletion of the surveillance when operating with a LIMITING ' i CONTROL ROD PATTERN will have minimal ef fect on the probability ; or consequences of an accident since operating at the parameter limit does not invalidate safety analysir assumptions. Additionally, it would not be evident that a LIMITING CONTROL ROD PATTERN had been achieved until the 24_ hour Frequency i surveillance was performed. As a result, the 24 hour Frequency surveillance serves to assure the parameter does not exceed the 's limits. This Frequency has been demonstrated through operating j experience to be adequate. The change, achieves consistency with the BWR Standard Technical Specifications. Therefore, no ; significant increase in the probability or consequences ! previously evaluated is involved with this change. ! l 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 nor does it require physical modification to the plant. Therefore, the change does not create the possibility of a new ! or different kind of accident from any accident 'previously ! evaluated. i
- 3. Does this change involve a significant reduction in'a margin of safety?
t This change has no impact on any safety analysis assumption ' since operating at the parameter limit is consistent with those ; assumptions. The existing 24 hour surveillance Frequency is maintained and has been demonstrated through' operating .i experience to be adequate for assuring the parameter does not exceed limits. Therefore, the change does not involve; a . significant reduction in a margin of safety. l
)
PERRY - UNIT 1 2 10/1/93 I i
l ATTACHMENT 2 ITS - PSTS ! COMPARISON DOCUMENT , I 2A: MARKUP OF ITS 2B: DISCUSSION OF CHANGES
I ATTACHMENT 2A ITS - PSTS COMPARISON DOCUMENT . MARKUP OF ITS
-/ ,
APLHGR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.I AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) i LCO 3.2.1 All APLHGRs shall be less than or equal to the limits specified in the COLR. APPLICABILITY: THERMAL POWER t 25% RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1 A. Any APLHGR not within A.1 Restore APLHGR(s) to 2 hours I limits. within limits. i t i B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion to < 25% RTP. -, Time not met. ! SURVEILLANCE REQUIREMENTS ! SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal Once within i to the limits specified in the COLR. 12' hours after a 25% RTP ! i AND I 24 hours ! thereafter ' I l i PE R kY- truiT 1- i
-BWR/6-STS ~mo 3.2-1 Re N 9/28/92~
MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS ! 3.2.2 MINIMUM CRITICAL POWER RATIO (MCFR) LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in. the COLR. I APPLICABILITY: THERMAL POWER a 25% RTP. ACTIONS , CONDITION REQUIRED ACTION COMPLETION TIME i A. Any MCPR not within A.I Restore MCPR(s) to 2 hours limits. within limits. B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion to < 25% RTP. Time not met. > SURVEILLANCE REQUIREMENTS ; SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal Once within to the limits specified in the COLR. 12 hours after a 25% RTP AND 24 hours thereafter BWRM-STS hrq -VJ 1 3.2-2 --Rc v . O, 09/28/92-
HGR (Opt..'eu M 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) (Optic .;l) A LCO 3.2.3 All LHGRs shall be less than or eaual to the limits specified in the COLR. APPLICABILITY: THERMAL POWER t 25% RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 2 hours limits. Restore LHGR(s) to within limits. B. Required Action and 8.1 Reduce THERMAL POWER 4 hours associated Completion to < 25% RTP. Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to Once within the limits specified in the COLR. 12 hours after a 25% RTP AND 24 hours thereafter CC/; SidtM M- Eb\ 3.2-3 Rc '. . O,-03/20/9E
l APRM Gain and Setpoints (Optional) [ 3.2.4 POWER DISTRIBUTION LIMITS ; i 3.2.4 verage Power Range Monitor (APRM) Gain and Setpoints (Optional) ! t LCO 3.2.4 a. MFLPD shall be less than or equal to Fraction of RTP; or l i b Each required APRM setpoint specified in the COLR shall ., e made applicable; or j
- c. Eac required APRM gain shall be adjusted such that the
[ APRM adings are = 100% times MFLPD. i APPLICABILITY: THERMAL POWER a 25 TP. ! h ACTIONS k CONDITION REQUIRED A COMPLETION TIME
\
A. Requirements of the A.1 Satisfy the 6 hours .! LCO not met. requirements of the i LCO. - . l i
- 8. Required Action and B.1 Reduce THERMAL POWER 4 hours -
associated Completion to < 25% RTP. r Time not met. ! N '
~
-l l
l l i BWR/6 STS 3.2-4 Rev. O, 09/28/92 { ;
APRM Gain and Setpoints (Optional) 3.2.4 N SURVEILIANCE REQUIREMENTS SURVEILLANCE FREQUENCY - SR 3.2.4.1 Verify MFL in limits. Once within 12 hours after
@ a 25% RTP kk'
} AND 24 hours Thereafter r ,
s 1 i ( BWR/6 STS 3.2-5 Rev. O, 09/28/92
APLHGR B 3.2.1 B 3.2 POWERDISTRIBUTI0hLIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ( BASES
=
BACKGROUND rods in a fuel assembly at any axial location.The the APLHGR are specified to ensure that the fuel Limits desic-on limits identified in Reference 1 are not exceeded during anticipated operational occurrences (A00s) and that the peak cladding temperature (PCT) during the postulated design basis loss ofin coolant limits specified 10 CFR 50.46.accident (LOCA) does not exce APPLICABLE SAFETY ANALYSES Pi he fuel design limits are presented atin in t Chapters 4, 6, and 15, and in ReferencesThe ., 1 and 2. u stW analytical methods and assumptions used in evaluating Desig V Basis Accidents (DBAs), anticipated operational transients, and nomalFMR presented i o erations apten that determine 6, a 15, and in APLHGR limits are References 1,2,a#32fpo4y P5 Fuel design evaluations are performed to demonstrate that the 1% limit on the fuel cladding plastic strain and other fuel design limits described in Reference 1 are not exceeded
, limit LP M APduring A00s for operation with LHGR up to the *
~ fo,r eath f ro mts are egy.tvalentpe~DiGR)mi y
# fuel a mbi P vided tuc.-tire ~ local-pfakJg factfr off exposure and Thiits are oeveloped as a function et 9 % )q e various operating core flow and power ,
limiting A00s (Refa. states to ensure aJd erence to fuel design lim are dete i' r.d 4)._ Flow dependent APLHGR limits h code (Ref 5' ed using the three dimensional BWR simulator {
' to analyze slow flow runout transients. The :'
flow depen ent multiplier, MAPFACr, maximum core flow runout capability.is dependent on the MAPFACr curves are ' provided based on the maximum credible flow runout transient for Loop Manual and Non Loop Manual operation. The result of a single failure or single operator error during Loop Manual operation is the runout of only one loop because both recirculation loops are under independenc control. Manual operational modes allow simultaneous runout of bothNon Loop ' loops because a single controller regulates core flow. (continued) BWWfr=STS 9ERRi- $1T 1 B 3.2-1 hc Gr 09f2&M2 ht PM6
APLHGR B 3.2.1 ' BASES APPLICABLE SAFETY ANALYSES Based on analyses of limiting plant transients (other than (continued) core flow increases) over a range of power and flow conditions, power dependent multipliers, MAPFAC,, are also generated. to initial core flow levels at power levels below th which turbine stop valve closure and turbine control valve fast closure scram signals are bypassed, both high and low core flow HAPTAC, limits are provided for operation at power levels level. power between 25's RTP and the previously mentioned bypass The exposure dependent APLHGR limits are reduced by MAPFAC, and MAPFACg at various operating ' conditions to ensure that all fuel design criteria are met for normal operation and A00s. A complete discussion the analysis code is provided in ReferencQ-dj.3 pg. LOCA analyses are then performed to ensure that the above determined APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A comple g discussion of the analysis code is provided in Referenc %'7 The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly. The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA 4( g analysis divided by its local peaking factor. A M. eEcb is conservative multiplier is applied to the LHGR assumed in . Speciiedlu the LOCA analysis to account for the uncertainty associated with the measurement of the APLHGR. . b ' For single recirculation loop operation,' the MAPFAC
~
multiplier is. limitsi to a maximunicR.0C IRefvtW This W +Qmit is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe cladding heatup during a LOCA. The APLHGR satisfies Criterion 2 of the NRC Policy Statement. LCO The APLHGR limits specified in the COLR are the result of fuel design, DBA, and transient analyses. For two (continued) BWR/6--AL Iu 3 - ). } \ B 3.2-2 Rev,- -0 r-09/28/ 92-
APLHGR B 3.2.1 BASES LCO (continued) recirculation loops operating, the limit is determined by multiplying the smaller of the MAPFACe and MAPFAC, factors times the exposure dependent APLHGR limits. With only one recirculation loop in operation, in conformance with the requirements of LCO 3.4.1, " Recirculation Loops Operating," the limit is determined by multiplying the exposure dependent APLHGR limit by the smallest of MAPFACe, MAPFAC , i and Q.05. Mere-ev6Jl has been determined by a specific singleTrecirculationloopanalysis(Ref.2). _ %
- 9 vaheYel$ed Sts'sO Ne\rek% leep openhw in & MLJ, i uh APPLICABILITY The APLHGR limits are primarily derived from fuel design evaluations and LOCA and transient analyses that are assumed to occur at high power levels.
and operating experience have shown that as power isDesign c reduced, the margin to the required APLHGR limits increases. This trend continues down to the power range of 5% to 15% RTP whe'n entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor (IRM) scram function provides ' pro =pt scram initiation during any significant transient, thereby effectively removing any ADlH6R limit Compliance concern in MODE 2. Therefore, at THEisuit POWER levels s 25% RTP, the reactor operates with substantial margin to the APLHGR limits; thus, this LCO is not required. ACTIONS A.1 h If any APLHGR exceeds the required limits an assumption regarding an initial condition of the DBA and transient 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
@ withinitslimit$andisacceptablebasedonthelow probability of aftransient or DBA occurring simultaneously with the APLHGR out of specification. ~
B.1 If the APLHGR cannot be restored to within its required Q limi within the associated Completion Time, the plant must (continued) 0=/0 STS- b yu;A i B 3.2-3 E. 4-0948/92
i
?
i APLHGR I B 3.2.1 ! BASES ! i
.f
. l' ACTIONS B,d (continueo) '
i I be brought the LCO doesto nota apply. MODE or other specified condition in which POWER must oe reduced to < 25% RTP within 4 hour ' allowed Completion experience Time is reasonable, based on operatin! orderly man,ner to reduce andTHERMAL POWER toplant without challenging < 25% RTP in an systems. , I SURVEILLANCE ! SR 3.2.1.1 ; REQUIREMENTS i APLHGRs.are required to be initially calculated within 24 hours thereafter.12 hours after THERMAL POWER is a 25 They are compared to the specified i within the a,sumptions of the safety analysis. lim Frequency is based on both engineering judgment andThe 24 hour i unfer normal conditions.r0 cognition of the slowness of chang The :2 hour allowance after ; THERMAL POWER e 25% RTP is achieved is acceptable given t t large inherent margin to operating limits at low power (' h i E gwi o ti-P-b, "Gued Elad'k W A'J b tf 4 Q .c Rc.che %c\(GEso p ( ta4at appond pviser). _ j j' REFERENCES @ (~2) P j {pe ene!y:is).
@@h ,
Chapter 15, Appendix g { Chapter 15, Appendix i
! (4.
$-NF-80-19(P)(Af"~ Exxon Nuclear Methodoygy f~oq 1 bon {ng Water Re&c(ors, Neutron 1cs Methods- for Design R ;
'JDj Aitaly5is,"_ Yo1ume,1, Jun_e 1981., _
f XN-NF-80-19(A), " Exxon Nuclear M WatehReactors, ECCS'Evaluatiohn .ethodoT0gy for1giling ! Sevi_sio'n.1, June 1981 r - qdel," Volume 2, 1 S. NEO bq r,\oWis. - 3 o) 3o A , " 5 % d) % \ c.. i tb r N\t%)3 jg (. N CD o '#1159, " Qul 8. c.h r [ N t t Om Oin _.\ . C a r e. S t u s M ofo J d br b .\ , q war L A rt " O n a.- i en q 1 tWRffr5T-S 9 u e 3.wg i 1 B 3.2-4 f,0 ; . O. 00123/0 ?
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fl . gjd N Ehr0Lc,s>- :10 s et , h C J c d bu. . J3ori dadnot - n .\p
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MCPR B 3.2.2 , B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) ; BASES BACKGROUND h CPR is result inpower. ratio of the fuel assembly power that would the onset of boiling transition to the actual fuel assembly The MCPR Safety Limit (SL) is set such that 99.9fs of the fuel rods avoid boiling transiti n if the limit is not violated (refer to the Bases for SL .2.-I-2). Th-operating limit MCPR is established to ensure that no fuel damag)e results during anticipated operational occurrences (A00s . Although fuel damage does not necessarily occur if a fuel rod actually experiences boiling transition (Ref.1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion. The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., l the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). conditions and bundle power levels are monitored andBecause pl . determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur. APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in evaluating ' the A00s to establish the operating limit MCPR are presente Q irthe5.VSAR, Chapters 4, 6, and 15, and References 2, 3, 4 . , J D and To ensure that the MCPR SL is not exceeded durin - any transient event that occurs with moderate frequency,g limiting transients have been analyzed to determine the. largest reduction in critical power ratio (CPR). of transients evaluated are loss of flow, increase inThe types pressure and power, positive reactivity insertion, and coolant temperature decrease. i the largest change in CPR (ACPR).The Whenlimiting transient the largest ACPR yields is , added to the MCPR SL, the required operating limit MCPR is obtained. 4
.i (continued)
BWR/G ^rTS- 9e RAA ~ \ t B 3.2-5 Ps . G, 09/20/03 ,
MCPR B 3.2.2 BASES APPLICABLE f,S',and 0 SAFETY ANALYSES The MCPR operating limits derived from the transient analysis are dependent (continued) state (MCPRr and MCPR,, the operating core flow and power : fuel design limits during theectively) to ensure adherence to with moderate frequency (Refs. orst transient 3, ", =d 5). that occurs , Flow dependent MCPR limits are determined by steady state thermal hydraulic {g methog}using (Ref the three dimensional BWR simulator code (hf@7).r d the ultichannel ihe.mel hyd = ?i: : Ode MCPRr curves are provided based on the maximum credibleoperation. flow runout transient for Loop Manual and Non Loop Manual The result of a single failure or single operator error during Loop Manual operation is the runout of only one loop because both recirculation loops are under independent control. 'Non Loop Manual operational ~ modes allow simultaneous controller regulatesrunout of both loops because a single, core flow. Power dependent MCPR limits (MCPR,) are detemined by the three dimensional transient BWR simulator code and the one dimensi code (Ref. 8). Due to the sensitivity of the transient response to initial 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, nigh and low flow QBrr operating limits are provided for operating between mentioned bypass power level. 25% RTP and the previously ' The MCPR satisfies Criterion 2 of the NRC Policy Statement. LC0 The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The MCPR operating limits are determined by the larger of the MCPRr and MCPR, limits. 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 _itopera
,jl ptftimunH>as4 tis]b and the moderator _a+ve-m4te void ratio 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) BWRM-STS hU3 -V ~ A 1 B 3.2-6 6 0, 00/20/42 i
MCPR i B 3.2.2 BASES { APPLICABILITY _ (continued) f tpf nominal value'of the initial MCPR P_is) exDi 9 \\ cyC3.5.) Studies of the variation of limiting flow conditions. These studies $ d limiting transients.of key actual plant parameter valu The results of these studies demonstrate that a margin is expected between performanc and the power MCPRtorequirements, is reduced 25% RTP. and that margins increase as This trend is expected to continue occurs. to the 5% to 15% power range when entry into MO When in MODE 2, the intermediate range monitor 1 (IRM)increase power provides rapid scram initiation for any significant transient MCPR compliance concern., which effectively eliminates any , i
< 25% RTP, the reactor is operating with substan to the MCPR limits and this LCO is not required.
ACTIONS M [h If any MCPR is ou; side the required limit an assumption analyses may not be met.regarding an initial condition of the (C2)
Therefore, prompt action should be such that the plant remains operating within conditions.
h The 2 hour Completion Time is normally sufficient to restore the MCPR(s) to within its limitand $p DBA occurring simultaneously with the MCPR specification. M h within the associated Completion Time, the pl brought LCO doesto nota apply. MODE or other specified condition in which the must be reduced to < 25% RTP within 4 hours.To achieve this The allowed. Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems. '. I (continued) 8htlE STS ke.ny'C.\ \ B 3.2-7 Rei. O. 09f?W i
MCPR B 3.2.2 4 BASES (continued) i SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours after THERMAL POWER is a 25% RTP and then every ) 24 hours thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 12 hour allowance after THERMAL POWER reaches e 25% RTP is acceptable given the large inherent margin to operating limits at low power lev is. -- = - 1 %Eo.rc ,4.- u
-at,Su- P A," G rw m ete " wo.\ d
%( wre% A J.<
s4an,.uaredoen3. J A pbdl, ~ D ; L REFERENCES l 1.
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i 9 @NUREG-0562, June 1479ff"'t Red failuus As A (ensega,,5ce_ @; ubtau. toi11nse Det M s' 32'. { Plant specifi?cerrZnt c,c!c n%ty-anPyW]El
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, w y _
7-(l . { Appendix 15B]p/
. EftinunkTli?eload gp gg Lhew7 p, S- 5' A'. .{ Appendix 15C}[ fixi- d,1if l Relcud 3 Cyof6'I +
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a lo 7 Appendix 15D}. f 6. XN-NF-80 i9 3), "Eixon Nuclear Meth ology for 11ing Water eactors, Neutronics Metho for Design lk a Analysis," lume 1 (asisupplemented). i N x
- 7. XN-Nk- -19(P)(A), " Exxon Nucle'ar Methodology fo /
Boiling ter Reactor' Methodolog Summary Des THERMEX Thermal Limits iption," Vblume 3 s l& ' N.. I Revision 2, (inuar 1987. \\ 'N t 8. XN-NF-79-71(P)\ , y Exxon Nuclea Plant Methbdology for ng Water _ React)rs," Revisi R 2, Novemb?r3_ 1981,.,
"BWR[6'GeribYic R65~Wittfarawal Error Analysis," General
.El'ectric Standard Safety Analysis Report, GESSAR-II,
/ Appendix 15B.f ,
i (1 WEo F. -lot 3 o > , " N c. J y h g r . \ M '< s . 9 A A 6 ,, rn g ,j;l , s~~. o a x.<.t . a % < m . u ,.1 (ort t $ ~ 1 i e.A & 3J d $ ,, b .\.s g w .A(r O c J ,rt," > em.,,g 4^"M:n Re .._e zeg,2e, 2 :
h LHGR (Opti =21) %. B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) (Optix@ BASES BACKGROUND The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. Limits on the LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, including anticipated operational occurrences Exceeding the LHGR limit could potentially resu(A00s). lt in fuel damage and subsequent release of radioactive materials. Fuel design limits are specified to ensure that fuel system damage, fuel rod failure or inability to cool the fuel does not occur during the anticipated operating conditions identified in Fretergence-4h{Qgs f, _s g
. _ ~
0 fe WK, C%phers 4, is,od 15; d in APPLICABLE The analytical methodiDiind7ssGHi~ptions\used in evaluating . SAFETY ANALYSES the fuel system design are presented idJReferences 1 and 2, The fuel assembly is designed to ensure (in conjunction with the core nuclear and thermal hydraulic design, plant equipment, instrumentation, and protection system) that fuel damage will not result in the release of radioactive materials in excess of the guidelines of 10 CFR, Parts 20, 50, and 100. The mechanisms that could cause fuel damage during operational transients and that are considered in fuel evaluations are:
- a. Rupture of the fuel rod cladding caused by strain from the relative expansion of the U02 pellet; and
- b. Severe overheating of the fuel rod cladding caused by inadequate cooling. =
A value of 1% plastic strain of the E w y cladding has been defined as the limit below which fuel damage caused by overstraining of the fuel cladding is not expected to occur (Ref._3). Thf6CPR~ Safety EfEiiT-ensuresAharfue13amage , s faUsed~,b severe overheating of thp-fuel rod clJdding,i Q ',% voided.
/f9 '
Fuel design evaluations have been performed and demonstrate that the 1% fuel cladding plastic strain design limit is not exceeded during continuous operation with LHGRs up to the (continued) BWR/6 STS B 3.2-9 O, 09/28/92 Rev.
+
g LHGR MBm3.2.3 WL BASES l APPLICABLE operating limit specified in the COLR. The analysis also. SAFETY ANALYSES includes allowances for short term transient operation above : (continued) the operating limit to account for A00s, plus an allowance i for densification power spiking. ! The LHGR satisfies Criterion 2 of the NRC Policy Statement. l LCO The LHGR is a basic assumption in the fuel design analysis. The fuel has been designed to operate at rated core power with sufficient design margin to the LHGR calculated to cause a 1% fuel cladding plastic strain. The operating i limit to accomplish this objective is specified in the COLR. APPLICABILITY The LHGR limits are derived from fuel design analysis that is limiting at high power level conditions. At core thermal power levels < 25% RTP, the reactor is operating with a substantial margin to the LHGR limits and, therefore, the , Specification is only required when the reactor is operating at t 25% RTP. ' 1 ACTIONS n/ UI1Nn-Ilt ddsp A.1 B,nik ofiti fud qd i If any LHGR exceeds its required limit, an assumption regarding an initial condition of the fuel design analysis is rot met. Therefore, prompt action should be taken to C6 restore the LHGR(s) to within its required limit (s)hthat ' a the plant is operating within analyzed con.ditions4 The 2 hour Completion Time is no .. ally sufficient to restore the LHGR(s) to within its limit and is acceptable based on the low probability of a transient or Design Basis Accident occurring simultaneously with the LHGR out of specification. : B.1 h If the LHGR cannot be restored to within its required limit 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, THEEMAL POWER must be reduced to < 25% RTP within 4 hours. Thw allowed , L (continued) ,
-C'./; STS kG.f I q ~\)O I B 3.2-10 n. C. 03/2 /02 j 6
. ~ .- . . . .- .- -. .- .- - -. - . .
Q LHGR (^?!b =? B 3.2.3- < i BASES l ACTIONS 8.1 (continued) Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems. , i i h SURVEILLANCE SR 3.2.3.1 %qere i REQUIREMENTS oss The HGRs @ required to e initially calculated within C
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12 ours a ter THERMAL POWER is a 25% RTP-and then every i 24 hours thereafter. compared with the specified i limits in the COLR to ensure that the reactor is operating 1 within the assumptions of the safety analysis. The 24 hour- ! Frequency is based on both engineering judgment and i recognition of the slowness of changes in power distribution. ! under normal conditions. The 12 hour allowance after i THERMAL POWER e 25% RTP is achieved is acceptable given the ; large inherent margin to operating limits ,at lower power ! levels. > t h h 1. REFERENCES ^ MI Y=?ph[. Q ML,Ck*rk Vi5. l b b 2. Chapteh4}8-j (- p b 3. NUREG-0800,4Section4 II N ^A.2(g), Revision 2, July 1981. l I . I k i LR/0 ';TS IS' B 3.2-11 %.
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, 09/20/24 i
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APRM Gain and Setpoints . B 3.2.4 ' B 3. POWER DISTRIBUTION LIMITS B 3.2.4 erage Power Range Monitor (APRM) Gain and Setpoints s L BASES 2O&am> BACXGROUND he OPERABILITY of the APRMs and their setpoints is an i itial condition of all safety analyses that assure rod in ertion upon reactor scram. Applicable GDCs are GDC 10 "Re tor Design"; GDC 13, " Instrumentation and Control"; ft/ GDC e " Protection System Functions"; and GDC 29, "Prote ion against Anticipated Operation Occurrences" j (Ref. 1) This LCO is provided to require the APRM gain or APRM flow iased scram setpoints to be adjusted when operating u der conditions of excessive power peaking to maintain acc table margin to the fuel cladding integrity Safety Limit L) and the fuel cladding 1% plastic strain limit. The condition of e essive power peaking is determined by , the ratio of the act 1 power peaking to the limiting power peaking at RTP. This atio is equal to the ratio of the core limiting MFLPD to e Fraction of RTP (FRTP) where FRTP is the measured THERMAL WER divided by the RTP. Excessive power peaking exists when: MFL
,l, FRTP indicating that MFPLD is not dec easing proportionately to k the overall power reduction, or c versely, that power peaking is increasing. To maintai margins similar to those at RTP conditions, the excessive pow peaking is t compensated by gain adjustment on the PRMs or adjustment of '
the APRM setpoints. Either of these a 'ustments has effectively the same result as maintaini MFLPD less than or equal to FRTP and thus maintains RTP m gins for APLHGR , and MCPR. The normally selected APRM setpoints position the scram , above the upper bound of the normal power / flow perating region that has been considered in the design o the fuel rods. The setpoints are flow biased with a slope + hat approximates the upper flow control line, such tha an approximately constant margin is maintained between e flow ; biased trip level and the upper operating boundary fo core ; flows in excess of about 45% of rated core flow. In t range of infrequent operations below 45% of rated core ow, (continueo ! BWR/6 STS B 3.2-12 O, 09/28/92 Rev.
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1
APRM Gain and Setpoints : (' B 3.2.4 BASES [Q{gg7{gy l BACKGROUND he margin to scram or rod blocks is reduced because of the ! (continued) n niinear core flow versus drive flow relationship. The l no ally selected APRM setpoints are supported by the ; ana ses presented in References 1 and 2 that concentrate on ' event initiated from rated conditions. Design experience - has.sh n that minimum deviations occur within expected , margins o operating limits (APLHGR and MCPR), at rated - 3 condition for normal power distributions. However, at - other than ated conditions, control rod patterns can be established at significantly reduce the margin to themal- ! limits. Ther ore, the flow biased APRM scram setpoints may ; be reduced duri g operation when the combination of THERMAL 4 POWER and MFLPD 'ndicates an excessive power peaking i distribution. : The APRM neutron fl signal is also adjusted to more ! Olosely follow the fu 1 cladding heat flux during power -l transients. The APRM utron flux signal is a measure of : the core thermal power ring steady state operation. ! During power transients, e APRM-signal leads the actual ! core thennal power respons because of the fuel thermal time , constant. Therefore, on po r increase transients, the APRM i signal provides a conservati ly high measure of core l thermal power. By passing the PRM signal through an i electronic filter with a time e stant less than, but .l approximately equal to, that of e fuel thermal time ; constant, an APkM transient respan e that more closely i follows actual fuel cladding heat x is obtained, while a : conservative margin is maintained. e delayed response of ! the filtered APRM signal allows the f w biased APRM scram i levels to be positioned closer to the er bound of the ! normal power and flow range, without'unn cessarily causing j reactor scrams during short duration neut n flux spikes. ! These spikes can be caused by insigniff'can transients such as performance of main steam line valve su illances or momentary flow increases of only several perc nt. .; i APPLICABLE The acceptance criteria for the APRM gain or setp 'nt l SAFETY ANALYSES adjustmen'.s are that acceptable margins.(to APLHGR nd NCPR) ! be maintained to the fuel cladding integrity SL and e fuel. . cladding 1% plastic strain limit. ' FSAR safety analyses (Refs. 2 and 3) concentrate on the _ f rated power condition for which the minimum expected marg ~n. (continued) ; BWR/6 STS B 3.2-13 Rev. O, 09/28/92
APRM Gain and Setpoints B 3.2.4 BASES \ 4 DEL.ETE D > APPLICABL to the operating limits (APLHGR and MCPR) occurs. SAFETY ANAL ES LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (continued) (APLHGR)," and LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," limit the initial margins to these operating limits at rated conditions so that specified acceptable fuel design limits are met during transients initiated from rated onditions. At initial power levels less than rated levels, P t margin degradation of either the APLHGR or the MCPR du 'ng a transient can be greater than at the rated con tion event. This greater margin degradation during the trans to limi nt isat primarily offset by the larger initial margin the lower than rated power levels. However, power di ributions can be hypothesized that would result in reduced ma ins to the pretransient operating limit. When combined wi the increased severity of certain transients at other than rated conditions, the SLs could be approached. At substantial distributions c reduced power levels, highly peaked power Id be obtained that could reduce thermal margins to the mi imum levels required for transient events. To prevent or miti is adjusted upward te the such situations, either the APRM gain ratio of the core limiting MFLPD to the FRTP, or the ow biased APRM scram level is required to be reduced by the r io of FRTP to the core limiting MFLPD. Either of these djustments effectively counters the increased severity of som events at other than rated conditions by proportional increasing the APRM gain or i l proportionally lowering the low biased APRM scram setpoints dependent on the increased pe ing that may be encountered. ! The APRM gain and setpoints sati fy Criteria 2 and 3 of the NRC Policy Statement. i LCO Meeting any one of the following condit ns ensures I acceptable operating margins for events scribed above: l
- a. Limiting excess power peaking;
- b. Reducing the APRM flow biased neutron flu upscale scram setpoints by multiplying the APRM set ints by the ratio of FRTP and the core limiting valu of MFLPD; or l
1 i (continue ' BWR/6 STS B 3.2-14 Rev. O, 09/28/92 i
APRM Gain and Setpoints B 3.2.4 BASEk g MQQD ) LCO c. Increasing the APRM gains to cause the APRM to read (continue greater than 100(%) times MFLPD. This Condition is to account for the reduction in margin to the fuel cladding integrity SL and the fuel cladding 1% plastic strain limit. LPD is the ratio of the limiting LHGR to the LHGR litrit f the specific bundle type. As power is reduced, if the des' n power distribution is maintained, MFLPD is reduced in ' prop tion to the reduction in power. However, if power peakin increases above the design value, the MFLPD is not reduced in proportion to the reduction in power. Under these co itions, the APRM gain is adjusted upward or the APRM flow When the re lased scram setpoints are reduced accordingly. tor is operating with peaking less than the design value, it is not necessary to modify the APRM flow biased scram s tpoints. Adjusting the APRM gain or setpoints is eq ' valent to maintaining MFLPD less than or equal to FRTP, as stated in the LCO. For compliance with C0 Item b (APRM setpoint adjustment) or Item c (APRM gain djustment), only APRMs required to be OPERABLE per LCO 3.3.1. , " Reactor Protection System (RPS) Instrumentation,". are r ufred to be adjusted. In addition, each APRM may be allowed o have its gain or setpoints adjusted independently of her APRMs that are having their ; gain or setpoints adjusted. APPLICABILITY , The MFLPD limit, APRM gain adjust.ent, or APRM flow biased scram and associated setdowns are ovided to ensure that the fuel cladding integrity SL and t e fuel cladding 1% plastic strain limit are not violated uring design basis transients. As discussed in the Bases or LCO 3.2.1 and LCO 3.2.2 suffic.ient margin to these li 'ts exists below 25% RTP and, therefore, these requirement are only necessary when the plant is operating at a 75% RTP. ACTIONS A.I ; If the APRM gain or setpoints are not within limi while the MFLPD has exceeded FRTP, the margin to the fue cladding integrity SL and the fuel cladding 1% plastic strai limit (conti ued) ! BWR/6 STS B 3.2-15 Rev. O, 09/28 2
APRM Gain and Setpoints B 3.2.4 BASq
/ DELETE 05 ACTIONS AJ (continued) may be reduced. Therefore, prompt action should be taken to restore the MFLPD to within its required limit or make P4 a:ceptable APRM adjustments such that the plant is operating within the assumed margin of the safety analyses.
The 6 hour Completion Time is normally sufficient to restore ither the MFLPD to within limits or the APRM gain or s points to within limits and is acceptable based on the lo robability of a transient or Design Basis Accident , occu ing simultaneously with the LCO not met. B.1 If the APRM ain or setpoints cannot be restored to within their require limits within the associated Completion Time, the plant must e brought to a MODE or other specified condition in whi h the LCO does not apply. To achieve this status, THERMAL P ER must be reduced to < 25% RTP within 4 hours. The allo d Completion Time is reasonable, based on operating experi ce, to reduce THERMAL POWER to
< 25% RTP in an order manner and without challenging plant systems.
SURVEILLANCE SR 3.2.4.1 \g REQUIREMENTS The MFLPD is required to be ca ulated every 24 hours and compared to FRTP or APRM gain o setpoints to ensure that the reactor is operating within t assuniptions of the safety analysis. This SR is requir only to determine the appropriate gain or setpoint and is n intended to be a CHANNEL FUNCTIONAL TEST for the APRM ga or flow biased neutron flux scram circuitry (assuming M PD is greater than ' FRTP). The 24 hour Frequency is chosen to cincide with the determination of other thermal limits, speci 'cally those for the APLHGR (LCO 3.2.1). The 24 hour Frequ cy is based on both engineering judgment and recognition of he slowness of changes in power distribution during normal operation. The 12 hour allowance after THERMAL POWER e 25% RT is achieved is acceptable given the large inherent marg to operating limits at low power levels. i (continued) BWR/6 STS B 3.2-16 Rev. O, 09/28/92
APRM Gain and Setpoints B-3.2.4 BASB Qed) g, QQ{g.p g } REFERDiCES 1. ' CFR 50, Appendix A, GDC 10, GDC 13, GDC 20, and GD pl 2. FSAR, Sectio ].
- 3. FSAR, Section [ ].
N t b-l P i BWR/6 STS B 3.2-17 Rev. O, 09/28/92
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l ATTACHMENT 2B ITS - PSTS COMPARISON DOCUMENT DISCUSSION OF CHANGES P 1 3 h l' L i
l i DISCUSSION OF CHANGES TO NUREG-1434 { SECTION 3.2 - POWER DISTRIBUTION LIMITS BRACKETED ADMINISTRATIVE CHOICE j I
-B.1 Brackets removed and optional wording preferences revised to !
reflect appropriate plant specific requirements. l i i PLANT SPECIFIC DIFFERENCE
.Ii i
P.1 The safety analysis report for this station is identified as j the Updated Safety Analysis Report and is correctly referred to ; as the USAR. i P.2 The plant specific current cycle safety analysis has been .i incorporated into the USAR Chapter 15, Appendix 15B. [ P.3 This comment number not used in the Perry submittal. f a P.4 Deletion of the APRM Gain and Setpoints specification was previously submitted by PNPP based on analysis for the maximum l extended operating domain (MEOD) and approved as part of the i initial full power operating license to the Perry-Unit 1. i P.5 The references have been revised to account for the use of 'l General Electric fuel ~rather than the Exxon fuel. discussed in i the improved technical specifications. ! 1 This comment number is not'used for this station.
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P.6 f P.7 This comment number is not used for this station. ,
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P.8 This information could not be substantiated and is proposed to l be deleted. The reference will also be deleted since this is l 1 the only use. i i < P.9 The appropriate reference for identification of the anticipated i operating condition 1 as related to the safety analysis, j P.10 Appropriate safety analysis references are included consistent. l with the similar Bases description for Specification 3.2.1. j P.11 The reference is not specifically applicable in the context presented. The Bases statement is sufficient, based on ! historical understanding and engineering judgement, without a f detailed reference, i s P.12 This comment number is not used for this station. i i PER.RY - UNIT 1 1 10/1/93
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DISCUSSION OF CHANGES TO-NUREG-1434 i SECTION 3.2.- POWER DISTRIBUTION LIMITS ,! I i CHANGE / IMPROVEMENT TO-NUREG STS j q C.1 The addition completes and corrects'the reference. C.2 This comment number not used for this station. i C.3 The flow control valve is normally at . the maximum position ., while on slow recirculation pump speed. 'However, this is not i the important point of the discussion and the unnecessary j description of position is proposed to be' deleted. , C.4 This sentence provides only unnecessary repetition of the '! definition of APLHGR and is proposed to.be deleted. I C.5 The APLHGR multiplier is cycle dependent and proposed to be . specified in the COLR in accordance with the guidance.provided in Generic Letter 88-16 for other such parameters. C.6 APLHGR, MCPR (for a given fuel type) and LHGR each have only a
; single limit; therefore, the singular term is used to refer to-
, each limit. Other editorial corrections made for consistency in wording between each' Bases. C.7 An obvious typographi aal error is corrected. C.8 The Bases for SR 3.2.3.1 is revised to discuss the plural "LHGRs" to match the SR discussion of "all LHGRs." C.9 The' deleted sentence provides MCPR Safety Limit Bases and is already included in B3.2.2. It is not appropriate for'the LHGR _) Bases discussion and is proposed to be deleted. C.10 These Bases are discussing MCPRp. An obvious editorial oversight corrected. C.11 Fuel cladding may be "Zircaloy" or "ZIRLO" as'ic ?ntified in 4.0 .,
" Design Features." The more generic =" fuel" is used here. f C.12 Since there are more than one fuel type, there is not a single "the" MCPR; there is one MCPR for each fuel type. An editorial correction is made for clarity.
r [ I PERRY - UNIT 1 2 10/1/93 !}}