ML20059B035
| ML20059B035 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/15/1993 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20059B017 | List: |
| References | |
| RTR-NUREG-1434 GNRO-93-00069, GNRO-93-69, NUDOCS 9310280011 | |
| Download: ML20059B035 (118) | |
Text
Ertel 2 to GNRO-03/00109 Chapter i Pcge1 GRAND GULF CHAPTER 1
5
~ h 931b280011 931015 PDR ADOCK 05000416 P
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i ATTACHMENT 1 CTS - PSTS COMPARISON DOCUMENT
(
1A:
MARKUP OF CTS 18:
DISCUSSION OF CHANGES 10: NO SIGNIFICANT HAZARDS CONSIDERATIONS
i End. 2 to GNRO-9300109 a
1 Chapter 1 Page 3
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t ATTACHMENT 1 A 1
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l CTS - PSTS 1
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COMPARISON DOCUMENT i
MARKUP OF CTS 1
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Enct 2 to GNRO.9300100 Chapter 1 Page 4
- 1. 6 D UIN1TIONS g
y-o s
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,.,, m_...___ _._. D. Z P.{ ~ d. _._n.,.._.,._./_: ::ifiseti....
, a x;.t:n:.- The enfined termsIap,.._..".D.. u_ _ _
- N
(
Sb pear in captulf and ;." 11 '
applicable throughest these Technical specificattee@qEcses j
re N
,y-hgE_ ha te that part of a Specification RegEer$Ac$cns s
3 prescribes
--eets r.,
r.giM under designated gendittens igin ggc;@
-to h e oc.
fiWErdfLANAREXP05DREl M
[ The A LANAR E shall licable cific 7
teight is equal sus of sure o the fuel in the fled bundt the speciff isht divi by the -- ;r of fuel s in ;
A 'the fuel
- e. F
.gAtAGE PLANAR LINEAR NEAT GE1tERATION RAT ((APL HG D The AVEf.:4iE ^i ^ J i.;E"" ;;OT Ex_CM.. 'APUGA b
hall be applicable to a specific planar height and is equal to the sus of the 1, #GR MGEs
-'f'T *:#
Tie ^ TES for all the fuel fees in the specified Dundle at the J
4 speciffed height divided by the number of fuel rods in the fuel hundle.
CHANNEL CALIBRATION M
wiNIe D A CMAMEL CALI8tATION shall be the adjustaant, as eccessary, of the p
channel output such that it responds the necessary range end accuracy known values of the parameter the chevel sentters. The L
R '" "'
OA5 LI8tATION thall encespass the entire chaf wl including the ensorgans alare, trip functions, and shall include the C$44fetEL FialCTIONAL TEST.
The L CAL!lRATION any be perfereed b 4any series of seguential, ove yMg
(
or tatal channel steps appin that the entire channel is calibrated.
mead e4 Jgg CMANNEL CMECK h
D A CHANNEL CBECK shall be_the[ qualitative asseseashof channel behavio during operationfly absorvatioW This detafsination stell. inclues, where possible cations camparison of the channel indication d @ staths other indi-tAs same paramstar M gpr status derived fres independent [nstrgme y
6 g CMANNEL FUNCTIONAL TEST D A CN4felEL RSCTIONAL TEST shall erac
f*irh. -ame:;;
- --d e injection of a simulated signschannel as eiese telhe senser as e
8 y
th into the -
g including!alaKj p functions,and channel fal!vre'tFripsN 0_
si.
- cw W. ;
h
. w -non a a an med4s wifvvel*flukierhmHasilers-mad / ene! r:.u O'2.
^:-h-8uneti;w (n}&ean iD The CM4NNEL FUNCTIONAL ft$T egr be perferend EPartr series of seguentf al, everlappingiar total channel steps that tAs estire channel is tasted.
0 GRAfC GULF-UNI'
- b 1-1 l
_-_m...m..___..m__-
Encl. 2 b GNRO-93AJ0109 Chaptsr 1 Pags5 INSERT 1A 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.
INSERT M
S4 VMRA2t?
Enct 2 to GNRO-93K10109 I
Chapter 1 Pegs 6 DEFINITIONS CORE ALTERATION h-cmpaeds I
i @ fuel, sources, d5R : CORE ALTERATION shall be the
-@-~---
n;--
- a movement of]on in: r a g or reactivity control $/within the reactor
+
F;wat vessel with the vessel head removed and fuel in the vessel. @;rel -
T
~
J ovement of GblRMi~lIALTERATION.IRMs. LPRMs. TIPh or special movable detectorsd consideredaHPBe COR Suspension of CORE ALTERATIONS shall not j
preclude completion of (fik movement of a component to a safe r easer.etivsg
/
position.
a
%g g'
2 18 sw vene\\
CORE OPERATING LIMITS REPORT (COLR) y pwa e A3 p
u nit
@reece renuLR is the frend O" %. dear stetjiiy/ specific documen The CO h
si limits for the current reload cycle. These cyclefspecific m n uno limits shall be determined for each reload cycle in accordance with j
Specification 5.0._;.IP Plant operation within these usereti74 limits is
/
addressed in individual Specifications.
g, Jg(;[ ga.W G",9. I, la (CRITICAL POWER RATIO q
1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the' assembly which is calculated by application of the ANFB correlation to cause some point in the assembly to experience boiling transition, divided by the j s tual assembly operating power.f DOSE E0VIVALENT I-131 DOSE ALENT I-131 shall be that concentration of I-131f g gram) SEMD alone would produce the same thyroid dose as the[microcuries/
quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-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."
A cc, t%23 GRAND GULF-UNIT 1 1-2 Amendment No. M. M,102.106
End. 2 to GNRO-9300109 Chapter 1 Pags 7 INSERT 2A i
In addition, control rod movement with other than the normal control rod drive is not considered a CORE ALTERATION provided there are no fuel assemblies in the associated core cell.
1 i
INSERT GRA!!D GULF 1-2 10/1/93
Enct 2 to GNRO-93/C0109 Chapter 1 Page8 DEFINITIONS
(
[bRYWELLINTEGRITY N
1.10 YWELL INTEGRITY shall exist when:
a.
All drywell penetrations required to be closed during accident onditions are either:
i 1.
Capable of being closed by an OPERABLE drywell automatic isolation system, or 2.
osed by at least one manual valve, blind flange, or desctivated automatic valve secured in its closed position, t
exc'ept for valves that are opened under administrative controls as p itted by Specification 3.6.4.
b.
The drywell quipment hatch is closed and sealed.
c.
The drywell a riock is in compliance with the requirements of Specification 6.2.3.
d.
The drywell leak e rates are within the limits of Specification 3.6.2.2.
e.
The suppression poo is in compliance with the requirements of Specification 3.6.3..
f.
The sealing mechanism ssociated with each drywell penetration; e.g., welds, bellows o 0-rings, is OPERABLE.
$ ll l
1 GRAND GULF-UNIT 1 1-2a Amendment No.106
Enct 2 b GNRO-9300109 1
Chapter 1 Pamp 9 DEFINITIONS I-AVERAGE DISINTEGRATION ENERGY
@ DI shall be the average (weighted in proportion to the I
of the average beta and gamma energies per disintegrationf(in MeV) for isotopes,9with half lives gr;;t; ther215 minutes, making up at least 95% of the total. nom,tM in. activ-in the coolant.
0%
W w flv m io n es EMERGENCY CORE ~ COO [ING lYSTE (ECCS) RESPONSE TIMEm&hS D The CC0CNCY40RE--f^^ LINO S707C% (ECCS7 RESPONSE interval from when the monitored parameter exceeds its ECCS F=':0 setpoint at the channel sensor until the ECCS equipment is capable of performing its S
safetyfunction7i.e.
discharge pressur(es re,ach their required values, etc.. Times s diesel generator starting and sequence loading delays),where applicable.
The response time may be measured bycany series of sequential, overlapping or ltotalsteps that the entire response time asured.
JEND/OffCYCLE (Coc Rt 5
IRCULATION PUMP TRIP SYSitm KtsPO E TIME
- M QD The ENC-OFCY"' C "EC:"C"LATIO!? ^L*'" T"!"fSYSTEM RESPONSE TIME shall be that time interval 3o complete suppression of the electric arc 3etween thJ Lully open contacts of the recirculation pump circuit breakerhfrom initial movement of the associatedfe-)
MC urbine stop valv-. :d ? R
\\b/
'necontrolvalvegd
g
(
The response' total steps Eveh that the entire response time is ocasured, except fo
' C ; 55 ~ 4 to conform to the manufacturer's design value. brea A-MACTION OF LIMITING _pnwEn--.OfMSTTY L A'
T iven location divided by the limiting LHGR for that bundl h*
Ag?
RACTION OF RATED T3E E r k b
CTION OF RATED THERMAL POWER (FRTP) shall be the measure
c divided by tee RATED THERMAL POWER.
'1]REQUENCY NOTAT1nh REQUENCY NOTATION specified for the performance of Surveillance i
wirements shall correspond to the intervals defined in Table 1.1.
M DELETED C 1
l l
t l
GRAND GULF-UNIT 1 1-3 Amendment No. Z8, 87
Encl. 2 to Gt4RO-93M09 Chapter i Page 10 DEFINITIONS O LEAKAGE
@ :Tn dIFifb tEAKAGE shall be:
NMd w
m,,.
. s g.,,< trms t--
f 3
1p.
Leana c
Lt r#
int.oicollection systeet, such as(pump s*ealser valve packin
, that is captured and conducted to a sump ofcollecting tank s
l/akag/intothedrywellatmospherefromsources.thatareboth Z y.
tion of 4By leakage detection systems er not t
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hDO Exe1 LEAKAGE.
% a %eteA La
_ ISOLATION SYSTEM RESPONSE TIME c ex ut tek.y. (ap 4 p,,n,, y.s, L.c.
D The ISOLATION SYSTEM RESPONSE TIME g
the monitored parameter exceeds its isolation qvrft etpoint at th channelsensoruntiltheisolationvalvestrav7totheirrequiredpositIons e
Times shall include diesel generator starting and sequence loadin applicable.
The response time may be measured by any serie
@ overlapping 3 e ays where er total steps hat the entire of sequential, y
k)~l MITING CONTR0TTcDMR sponse time is asasured, mas s
'" ' g ' ' p'C'* -h
- 1. O A ING CONTROL ROD PATTERN shall be a pattern which results c
eing on a thermal hydraulic limit, i.e., operating on a limiting val or\\APLHGR.g_ or MCPR. J
(
}
_ LINEAR HEAT GENERATION RATEluded
..f
{
a
' @ length of fuel LINEA
- h:AT :N:"ATION "!,T (LHGRNhall be the heat generation per unit It is the integral of the heat flux over the heat rod transfer area asscc.iated with the unit length.
LOGIC SYSTEM FUNCTIONAL TEST ps c\\ cst b % scMar _
= pmch @
6, bgl nd D., A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of a11+1ogic componen s o
- i. e.
l r(etc.) of a logic circuit, fros e@= trip units, solid state logic elem a11+ relays and contacts,
.t
.thm;t end-including the actuated to verify OPERA 8ILITY.
a The LOGIC SYSTEM FUNCTIONAL TEST may be y
performed by, series of sequential, overlapping er total systes steps c
tha the nti sos stas is tasted.
g epip*<a b
ac.
MJKIMUM FRALilui Cl PrnING POWER DENSI '
1.23 The FRACTION OF LI
/d (a
the FLPD which s 'et.
POWER DENSIJL-(MFEPD) shall bo-the1Tg n the core.
1 REMBER(5)0FTHEPUBtfC{
/
- 1. 4 MEMBE ally associated with the plant.OF THE PUBLIC shall include C
oc upat 1
- of the ytility, its contractors or vendors.This category does not include ca cry are persons who enter the site to service equipment or to make d i
e.
tional, occupational or other purposes not ass c
GRAND GULF-UNIT 1 ant.
14
End. 2 to GNRO 93/co109 Chapter 1 Page 11 DEFINITIONS MINIMUM CRITICAL POWER RATIO (MCPb
&%\\ heaer r&
The MINILM CRITICAL POWCR RATIO (MCPRfshall be the smallest @f exists in the corezs q
- t
/0P/SITEOb5ECALCULATIONMANUAL(00
( 9
.1 1.
Th 0FFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology 0/.l$ and pa ameters used in the calculation of offsite doses resulting from radio-i acti gaseous and li uid gffluents, in the calculation of gaseous and liquid effl nt monitorino and onito(Trip /etpoints, and in the conduct of the/ Environ-i ment U Radiological, ring Program.
The ODCM shall also contain (1) the Rad o etive Effluent Controls and Radiological Environmental Monitoring Programs re i d by Section 6.8.3 and (2) descriptions of the information that should be i
lud d in the Annual Radiological Environmental Operating and Semiannual R dioac ive Effluent Release Reports required by Specifications 6.9.1.6, 6.9.1.7, l
.9.1.8 nd 6.9.1.9. c
~
PERABLE - OPERABILITY t,
d.o,s,em h i
A system, subsystem,t
' component or device shall be OPERABLE or have Ato OPERABILITY when it is capable of performing its specified(function (s) and leg l ggi hen all necessary attendant Instrumentation, controls felectrical power, i
_ coolinole seal water, lubricationfger other auxiliary Cequipment that are w rM d required for the system, subsystem, f T 'D,7 component or device to perform its
(
,tign } g also capable of performing t
rlatedsupportfunction(s).emco,u/
i Svison ERATIONAL CON 3NfJN - CONDITION]
<k
- 1. B A-ERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive nation of mode switch position and average reactor coolant temperature as to ecified in Table 1.2. f P4iYSICS TESTS) 2 1.29 PHYC. 5 TESTS shall be those tests performed to measure the fundamental i
~'
~
pt:ief paracte-is. :3 of the reactor core and related instrumentation.and p) cescribed in Chapter 14 of the FSAR, 2) authorized under the provisions of
-10iCFR 5D.59; or 3) nthervise approved by the Commission.(
f+K-
- y)
@itE55UREBOUNDARY. LEAKAGE g4 l
Q.D )JE55ditE-40UNDMY3 LEAKAGE lihell t,; leek &M through a nonIsolable fault i
l inapeactorfoolantf)st ponent body, pipe wall or vessel wall.
I ec i
lAl$)
fTLR
/A/ScfT 5C GRAND GULF-UNIT 1 1-5 Amendment No. 87
Encl. 2 to GNRo.9300109 Chrpter 1 Prge 12 i
i INSERT 5A 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 5B 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.
e INSERT SC PRESSURE AND The PTLR is the unit specific document TEMPERATURE that provides the reactor vessel pressure LIMITS REPORT and temperature limits, including heatup (PTLR) and cooldown
- rates, for the current reactor vessel fluence period.
These 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 addressed in LCO 3.4.11, "RCS Pressure and Temperature (P/T) Limits."
i i
INSERT GRAND GULF 1-5 10/1/93
Enct 2 to GNRO-93S0109 Chapter 1 Pags 13 DEFINITIONS g
fPRIMARY CONTAINMENT INTEGR 1.31 PRIMA CONTAINMENT INTEGR11Y shall exist when:
a.
All containment penetrations required t e closed during accident y
conditions ar either:
1.
Capabit of being closed by a OPERABLE containment au atic iso ation system, or h0 2.
' osed by at least one anual valve, blind flan
, or deactivated automatic valve secured in its closed positio, except for valves j
that are opened under administrative contro as permitted by i
Specification 3 4.4.
b.
The containment uipment hatch is closed d sealed.
\\
\\
c.
Each contairpnent air lock is in compli ce with the requirem 'ts of Specifica fon 3.6.1.3.
1 d.
The co ainment leakage rates ar within the limits of ecification i
- 3. 6.)(2.
i I
e.
e suppression pool is in ompliance with the re irements of Specification 3.6.3.1.
/
/
\\
f.
The sealing mechanism associated with each primary containment i
g enetration; e.g., welds, bellows or 0-rings, is OPERABLE.
P50h55CONTROLPROGRAM(PCP)
\\
\\
1.32 Ke PROCESS CONTROL PROGRAM (PCP( shall contain the current formulas, singandh, analyses,tkt,anddeterminaionstobemade30ensurethatproces-sampling ackaging of sAlid radioactive stes based on de nstrated 'proces sing of actual \\r simulated w solid wastes wi be accomplished such a way as s
to assure c pliance with CFR Parts 20, 61, and 71, State re l ations, ' burial
, ground requir ents, and oth requirements gove ging the disposal of solid (radioactivewa e.
PURGE - PURGING e[
1.33 PURGE PURGING is'tIIe coptrolled process of scharging air from a
' confine t to maintain temperature, pressure midity, con ion or other' oper ng condition, i ch a manner tha, eplacement ai gas is regu Wed to pu fy the confine f
RATED THERMAL POWER (RTP)
RTP WJ"O TMRMAL POWID e a total reactor core heat transfer rate to the reactorcoolantof3833MWg GRAND GULF-UNIT 1 1-6 Amendment No. 87,102
Enci 2 to GNRO-93.t)0109 Chapter 1 Page 14 j
DEFINITIONS (R Ps)
{
V REACTOR PROTECTION Sv5 TEM RESPONSE TIME De RPs$
M T-25PRTO =Pmtu:oN 575TEMRESPONSE TIME shall be @ time interval from when the monitored parameter exceeds its* trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.
The response time may be measured bysany series of sequential, overlapping,or total steps @
that the entire response time is measured.
A so REPORTABLE EVENT [
OW
/
0 AB of p onditio y cified y ROD DENSITY Cp?1.37R000 ITY shall be
' number of ontrol rod notches inserted 3s a fractio of the total n er of con rod notcheMll rods f Mfinserted 3
{
ivalent to 100%
D DENSITY.
TCONDARYCONTAINMENTINTEGRITY{
1.38 SECONDARY CONTAINMENT INTEGRITY shall exist when:
a.
All Auxiliary Building and Encio g penetrations
(
required,to be closed during ident co tions are either:
1.
able of being sed by an OP BL secondary containment automatic isola on system, or f.
Closed by a least one manual e, blind flan
, rupture disc or deacti ted automatic valve or damper, as plicable, secured in its osed position.
l 1
b.
All Auxil' ry Building and Enclosure Buildi equipment hatches and blowout anels are closed and sealed.
i c.
The andby gas treatment system is i compliance with the r uire-k men s of Specification 3.6.6.3.
\\d.
} e door in each access to the xiliary Building any nelosure tuilding is closed, except normal entry and exit.
e.<
The sealing mechanism associated with each Auxiliary Building and Enclosure Building penetration, e.g., welds, bellows or 0-rings, is OPERABLE.
GRAND GULF-UNIT 1 1-7 Amendment No. 44,102
End. 2 to GNRO-93/Co109 Chapter 1 Pagt 15 OEFINITIONS
(
SHUTDOWNMARGIN(5Btt) kt; c, CG EUTD"Xd FMRGi_R(shall be the amount of reactivity by which v
subtritical or would be subcritical assumin all control rods are fully inserted except for the single control rod of highest reactivity worth,which is assumed to be fully withdrawn.a o
68 F; and xenon free { 4 fheg
="2"=='
is P th:'h GoAeinbr %QqQ q ms ggt
- hutfr m -+ i= i n i & +. :j *-
4 rA }
'TJ/IE BOUNDARY { Th G Fe o.cAe r D The SITE BOUNDARY shall be that line beyond which the land is neither p L.o.ne nqr leased, nor otherwise controlled by the licensee.
d.41 CELETED
-1. G ::LETE5 '
STAGGERED TEST BASIS
/N56tr 26 M
D 5TAGGERED "ST S?. SIS sha!' :casist ef.
f a.
At schedule for n systems, subsys s, trains or other desi i
c aponents obtained by dividing th pecified test interval i equal subintervals.
The testing of one system, s system, train [or other der :f ated r~
(
pga;3t 3t
+8. s ginn4 ; nr mes ensin+. rum)
THERMAL POWER Q THERMAL POWER shall be the total reactor core heat transfer rate to t reactor cool t,
b eg oc N by 4
@S~ AI::NT r!
a, OfLEAKAGE 6C k ::1' M9e;e Ti: Mis not'{DEOTIh'EDjlEAKAGE; etNkiESTRICTED AREAT
- 1. 6 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY ac-to which is not controlled by the licensee for purposes of protection of ces MEMB RS OF THE PUBLIC from exposure to radiation and radioactive materials, or a within the SITE BOUNDARY used for residential quarters or for indus-(any a
- trial, I
omercial. institutional, and/or recreational purposes.
I 1.47 CELETE"-
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,c,
Encl. 2 to GNRO-92fC0109 Chapter 1 Page 16 INSERT 8A With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.
INSERT BB 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 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.
l INSERT rsWNNLl@DW R-fA SUMADD
Encl. 2 to GNRO-93/00109 Chapter 1 Peg 317 DEFINITIONS hNTIN g
- 1. 4 ENTING is the controlled process of discharging air or gas from a A !,I
'nement to maintain temperature, pressure, humidity, concentration or con et e operating condition, in such a manner that replacement air or gas is not p ovi d or required during VENTING.
Vent, used in system names, does not imply gVENT G process.
~
GRAND GULF-UNIT 1 1-9
Encl. 2 to GNRO-93/00109 Chapter 1 Pags 18 k
'13 NTABLE1.1 SURVEILLA E FREQUENCY NOTATION NOTATION FREQUENCY S
At least nce per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
O At least on per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W At least once p 7 days.
M At least once per days.
Q At least once per 92 s.
A At least once per 184 days A
At least once per 366 days.
R least once per 18 months (550 ays).
5/U Pri to each reactor startup.
N. A.
Not ap icable.
('
P Completed rior to each release.
i GRAND GULF-UNIT 1 1-10
End 2 to GNRO-0100109 i
Chapter 1 Paga 19
/,/-/ [ pose I ok l)
~
TABLE W i
noE8"!cHLC0"0TI0xs> MODES L
t REAc70R
/V7006 777tg MODE SWITCH AVERAGE REACTOR r CQN&fDGN POSITION,
COOLANT TEMPERATUR( 0 0[ERATIOg 7
e y t:- :-etu 4 Q p
5,7A Starmp/Hof standby My tn="t"rT 'h e-J S[UTD0
/
Shutdow8'***
>200%
3 S[U Shutdowr # '" * ***
1 200 h C
5 REFUELI Shutdown or Refue **'#
-i 100*P
~"h j
/%+@, b O
l
// rene(or vessel eAd closute cl4.s he g 3', egi,
(
i t ed*S The reactor mode switch may be placed in the Run or Startup/ Hot Standby to LCD position to test the switch interlock functions provided that the control 3 io.f i rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.
k S a i " The reactor mode switch may be placed in the Refuel position while a single to Lk '
control rod drive is being removed from the reactor pressure vessel per gdkSpecification3.9.10.1.
3.i (b)(W O ne o r m o re. re a.e 4os-Fel ir the 7 ::t:r ^::ee' \\ th the vessel head closure bolts less than H
fully tensioned.er $ 2 th: 5 :d :::ved.
f M
gh. : e::m Tm _.,.m.,.
- . m.1.,~ :. a. :.
ARA gag The reactor mode switch may be placed in the Refuel position while a single 4* te of control rod is being moved provided that the one-rod-out interlock is l
[.sc.3s.4 4 h
t Add Js6 ET 12 A ) Ionedig k'is pap.
GRAND GULF-UNIT 1 1-11 Amenchent No. 63
Enci 2 to GNRO-9300109 Chapter 1 Page 20 l
INSERT 12A Insert new Sections 1.2,
" Logical Connectors,"
1.3,
" Completion Times" and 1.4, " Frequency" as shown in the markup 1
of the Improved Technical Specifications, NUREG-1434.
1 INSERT GRAND GULF 1-L9 SutDLEtt
Enct 2 to GNRO-934)0109 Chapter 1 Pcge 21 i
ATTACHMENT 1B CTS - PSTS COMPARISON DOCUMENT DISCUSSION OF CHANGES
(
(
l Enci 2 to GNRo-93X)0109 Chapter 1 Pagi 22 DISCUSSION OF CHANGES CTS: 1 - DEFINITIONS ADMINISTRATIVE A.1 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 understandable, by plant operators as well as other users.
During this reformatting and renumbering j
- process, no technical changes (either actual or j
interpretational) to the Technical Specifications were made unless they were identified and justified.
In the specific j
case of the Definitions Section, no individual numbering of each definition is made.
A.2 The sentence is deleted.
It serves only as background or basis material and is not incorporated in the BWR Standard Technical Specifications, NUREG-1434.
A.3 The format of the Actions in the BWR Standard Technical i
Specifications, NUREG-1434, contains specific fields which are more accurately presented in this revised wording of the definition.
I 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 display function.
This is the intent of the existing wording, l
and therefore the revised wording is proposed to more accurately reflect this intent; consistent with the BWR I
Standard Technical Specifications, NUREG-1434.
Since the list of equipment functions is intended to provide examples of attributes which must potentially be OPERABLE, dependent on whether it is " required" or not, the list can be applied to both analog and bistable channels, and the separate listings can be combined.
A.5 Usage of the terms "and/or" has been changed to "and" or "or",
as appropriate.
The BWR Standard Technical Specifications, NUREG-1434, Writer's Guide recommends the use of "and/or" be avoided.
The intent of the definitions is not changed.
A.6 Editorial rewording is made consistent with the BWR Standard Technical Specifications, 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.7 The words " addition, removal, relocation" have been deleted since " movement" encompasses these words, and " reactivity controls" are now described as " reactivity control components."
No change in intent or interpretation is made with this 4
i proposal.
The change is consistent with the BWR Standard Technical Specifications, NUREG-1434.
I
_ - _ _ ~.. -
I End. 2 to GNRo-93A)0109 Chapter 1 Pagi 23 1
r DISCUSSION OF CHANGES CTS: 1 - DEFINITIONS ADMINISTRATIVE (continued)
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" 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 Specifications, NUREG-1434. Eliminating the existing requirement to also be a " conservative" position avoids potential confusion and perhaps overly restrictive interpretation.
Since there is no reference on which to base the conservative evaluation (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 chage is 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, PRIMARY CONTAINMENT INTEGRITY and SECONDARY CONTAINMENT INTEGRITY have been deleted from the proposed 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, Primary Containment and Secondary Containment, along with the remainder of the LCOs in the Containment Systems chapter.
Therefore the change is an administrative presentation preference adopted by the BWR Standard Technical Specifications, NUREG-1434.
A.12 This comment number is not used for this station.
A.13 The definition of FREQUENCY NOTATION has been deleted since the abbreviations in the existing Table 1.1 are no longer used.
All Surveillance Requirement Frequencies in the proposed Technical Fpecifications are directly specified.
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 leakage" is consistent with this proposed definition.
Enci 2 to GNRO-9300109 Chapter i Psge 24 1
DISCUSSION OF CHANGES CTS: 1 - DEFINITIONS ADMINISTRATIVE (continued)
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 irtended to be those for j
collection of leakages into the drywell space.
This proposed change is a clarification of the term, and therefore the i
revised wording is proposed to accurately reflect this intent; consistent with the BWR Standard Technical Specifications, i
A.16 The definition of LOGIC SYSTEM FUNCTIOFAL TEST (LSFT) has been modified to not include the actuated device.
The actuated device is to be tested as part of the system functional test.
Deleting the actuated device from the definition of LSFT i
eliminates the confusion as to whether a previously performed 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 additiu..
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 administrative, with no impact of its own.
A.18 The definitions of OFFSITE DOS ' CALCULATION MANUAL and PROCESS CONTROL PROGRAM have been incorporated into the Administrative Controls Section.
Editorial wording changes are consistent with the BWR Standard Technical Specifications, NUREG-1434.
A.19 OPERATIONAL CONDITION has been deleted and a definition of MODE is added to be consistent with terminology used in the BWR Standard Technical Specifications, NUREG-1434. Since their use is interchangeable. this change is considered to be 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 comment A.25).
A.20 These additions provide clarification of the existing requirement without any modification of intent.
J Enct 2 to GNRo-93/00109 Chapter 1 Page 25 DISCUSSION OF CHANGES CTS: 1 - DEFINITIONS ADMINISTRATIVE
]
(continued)
A.21 The definition of PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) has been added consistent with the Iuproved BWR Standard 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 considered 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 considered to be editorial.
A.23 The definition of STAGGERED TEST BASIS has been modified to be consistent with its usage throughout the proposed Grand Gulf Technical Specifications.
The intent of the frequency of testing components on a STAGGERED TEST BASIS is not changed.
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 1
the current approach, which requires a determination of the Surveillance sub-interval from the test schedule based on the number of subsystems.
A.24 These footnotes are addressed by the exceptions allowed to LCO requ'.rements 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.
A.25 The intent of applying the MODE definition only when fuel is in the vessel is relocated to the definition of MODE (refer also to comment 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 considered 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 Technical Specifications, NUREG-1434.
l Enci 2 to GNRO-93,00109 Chapter 1 Pzga 23 DISCUSSION OF CHANGES CTS: 1 - DEFINITIONF ADMINISTRATIVE (continued)
A.27 The following sections are being added to the Technical Specifications.
These additions aid in the understanding and use of the new standard Technical Specification format and 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 Specifications, NUREG-1434, as approved by the NRC Staff, and the guidance is nc,t a specific deviation from anything in the existing Technical Specifications, these additions are considered to be administrative.
The added sections are as follows-i SECTION 1.2 - LOGIC CONNECTORS Proposed Section 1.2 provides specific examples of the l
logical connectors "AHR" and "QE" and the numbering sequence associated with their use.
This revision is being proposed consistent with the BWR Standard Technical i
Specifications, NUREG-1434.
SECTION 1.3 - COMPLETION TIMES Proposed Section 1.3 provides proper use and interpretation of Completion Times.
The proposed section alr o provides specific examples that aid the user in understanding Completion Times.
The proposed Completion Times Section is consistent with the BWR Standard Technical Specificaticnn, NUREG-1434.
SECTION 1.4 - FREQUENCY j
Proposed Section
.4 provides proper use and interpretation of 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 Specifications, 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.
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End. 2 to GNRo-93/00109
[
Chapter i Pags 27 P
DISCUSSION OF CHANGES I
CTS: 1 - DEFINITIONS
}
ADMINISTRATIVE (continued)
J 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 transmittar (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 i
which cannot be adjusted or changed once installed..in this
]
When a credible mechanism which can cause change or drift 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 / output relationships to the remainder of the channels and making the necessary adjustments to ensure range and accuracy.
This proposed " verification of OPERABILITY" of the sensing 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.
RELOCATED SPECIFICATIONS R.1 This comment number is not used for this station.
.___,.-,-..~.__,,-__._,,_.__.w
j a
Enci 2 to GNRo-9300109 Chapter i Page 28 I
1 j
DISCUSSION OF CHANGES j
CTS: 1 - DEFINITIONS 1
i i
TECHNICAL CHANGE - MORE RESTRICTIVE i
M.1 The intent of these changes is to provide clarity and i
completeness, and as such could be considered administrative, i
However, technically, the changes eliminate the potential to j
interpret certain plant conditions such that no MODE, or a less j
restrictive MODE, would exist.
Therefore, in proposing this change (consistent with the BWR Standard Technical j
Specifications, NUREG-1434), it will be discussed and justified j
as a "more restrictive" change.
i 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 j
corresponding MODE and could therefore be incorrectly interpreted as not requiring the application of the majority of i
Technical Specifications.
By defining this plant condition as STARTUP MODE, sufficiently conservative' restrictions will be applied by the applicable LCOs.
l 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
" Shutdown" position:
with the vessel head detensioned, both i
the definition of REFUEL as well as COLD SHUTDOWN could apply.
It is not the intent of the Technical Specification to allow an i
j option of whether to apply REFUEL applicable LCOs or to apply COLD SHUTDOWN applicable LCOs.
This proposed change precludes j
an unacceptable interpretation.
i i
The existing definition of REFUEL would cease to be j
i applicable when average coolant temperature exceeded 140 - F.
1 With the mode switch in " Refuel" a plant condition which has no j
corresponding MODE exists. This could therefore be incorrectly i
interpreted as not requiring the application of the majority of j
Technical Specifications.
By defining the REFUEL MODE as j
including plant conditions with no specific coolant temperature i
range, sufficiently conservative restrictions will be applied
{
by the applicable LCOs during all fueled conditions with the j
vessel head bolts detensioned.
I 4
TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" l
None in this section.
l d
n-.,-
1 End. 2 to GNRo-93/00109 l
Chapter 1 Paga 29 i
DISCUSSION OF CHANGES CTS: 1 - DEFINITIONS TECHNICAL CHANGE - LESS RESTRICTIVE (continued) 1
" 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 demonstrated in either case since the channel itself can not discriminate between
" actual" or " simulated."
L.2 As provided for with analog channels, the signal 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 the other logic channels.
Either method significantly increases the difficulty of performing the surveillance.
Allowing initiation of the signal close to the sensor provides a complete test of the logic channel while significantly reducing this probability of undesired initiation.
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 interpretation.
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 Specifications, NUREG-1434.
The basis for this is evident in that the Specifications that are applicable during CORE ALTERATIONS are those that protect from or mitigate a reactivity excursion event.
J Encl 2 to GNRo.93/00109 i
Chtpter i Pegs 30 1
DISCUSSION OF CHANGES t
CTS: 1 - DEFINITIONS 1
TECHNICAL CHANGE - LESS RESTRICTIVE (continued)
L.4 Consistent with the rationale presented in item L.3 above (i.e., the proposed change focuses the definition on activities that can affect the core reactivity), an additional change is proposed to allow the physical removal of a control rod to not 1
be considered a CORE ALTERATION.
In this activity the control cell must first have all the fuel bundles removed prior to this 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 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 reactivity excursion event.
After such time, sufficient margin and design features (the design of a control rod precludes its removal without all fuel assemblies in the cell removed) are in place to allow removing the Technical Specification controls during the control rod removal.
This proposed change is consistent with the BWR Standard Technical Specifications, NUREG-1434.
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End. 2 to GNRO-93/00109 Chapter 1 Paga 31 i
i ATTACHMENT 1C CTS - PSTS COMPARISON DOCUMENT NO SIGNIFICANT HAZARDS CONSIDERATIONS
(
i l
l 4
W I
Enct 2 to GNRO-9200109 Chapter 1 Page 32 NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 1 - DEFINITIONS i
"L1" CHANGE Entergy 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 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 acceptability of generating this signal.
The proposed change does not affect the procedures governing plant operations and the acceptability of creating these signals; it simply would 1
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.
Enct 2 to GNRo-93.00109 I
Quptx 1 Pgi 33 NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 1 - DEFINITIONS "L2" CHANGE Entergy has evaluated this proposed Technical Specification change and 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 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.
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 of fset by the reduction in complication which will reduce 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 will reduce the probability of personnel error during the surveillance. Therefore this change does not involve a significant reduction in a margin of safety.
Enct 2 to GNRo.93/00109 l
Chapter 1 Prp 34 NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 1 - DEFINITIONS "L3" CHANGE Entergy 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?
j 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 i
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 for CORE ALTERATIONS would be independently applied),
no analyzed event is assumed.
Therefore, this proposed 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. OPERABILITY requirements for the incore instruments remains required when necessary for monitoring and event mitigation.
Movement of the detectors involves:
- normal retraction to below the core plate via an l
installed drive system (this movement is excluded from the existing definition of CORE ALTERATIONS);
- manual retraction to the under-vessel cavity for replacement and subsequent reinsertion of a new detectors.
This activity causes no potential of changing core geometry or core reactivity;
~
Encf 2 to GNRO-9300109 Chaptsr 1 Pags 35 NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 1 - DEFINITIONS l
"L3" C5fANGE
)
(continued) j in-vessel movement of
" dummy" assemblies containing special moveable detectors (this movement 1
is excluded from the existing definition of CORE ALTERATIONS).
l Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.
)
l 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 cauce reactivity in the core to be discernibly changed. The OPERABILITY of the incore detectors to function in mitigation of analyzed svents is unaffected by this change.
The proposed change involves allowing movement of these incore detectors while n;t 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.
l
Enci 2 to GNRo-93/00109 Chapter i Page 36 NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 1 - DEFINITIONS "L4" CHANGE Entergy 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, 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 ALTERATION, involves first removing all the fuel bundles in a cell prior to this 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 the control rod movement, and 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 unique to CORE ALTERATIONS (note:
other requirements, such as those for handling loads over irradiated fuel, will remain applicable).
Therefore, this proposed 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.
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 component" while not imposing requirements unique to CORE ALTERATIONS (note:
other requirements, such as those for handling loads over irradiated fuel, will remain applicable).
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 of a new or different kind of accident from any accident previously evaluated.
Enct 2 to GNRO-03R)0109 Chapter 1 Pag 337 NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 1 - DEFINITIONS "L4" CHANGE (continued) 3.
Does this change involve a significant reduction in a margin of safety?
Since 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, not considering the proposed activity (movement of a control rod, other than with the normal control rod drive) to be a CORE ALTERATION does not involve a significant reduction in a margin of safety.
$$'a"P.#f" APPENDIX A l
GENERIC
)
i NO SIGNIFICANT HAZARDS CONSIDERATIONS and ENVIRONMENTAL ASSESSMENT i
--.-.,,,m.._.--
.4,
End. 2 to GNRO-9340109 Chrpter 1 Pags 39 GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS FOR THE CURRENT TECHNICAL SPECIFICATIONS ADMINISTRATIVE CHANGES
("Ax" Labeled Comments / Discussions)
Entergy 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 involves reformatting, renumbering and rewording of the existing Technical Specifications.
The reformatting, renumbering and rewording process involves no technical changes to existing Technical Specifications.
As such, this change is administrative in nature and does not impact initiators of analyzed events or assumed mitigation of accident or transient events.
Therefore, this 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 a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation.
The proposed change will not impose or eliminate any new or different requirements.
Thus, this 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 will not reduce a margin of safety because it has no impact on any safety analysis assumptions.
This change is administrative in nature.
As such, no question of safety is
- involved, and the change does not involve a
significant reduction in a margin of safety.
Enct 2 to GNRo-9300109 Chapter i Prge 40 GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS FOR THE CURRENT TECHNICAL SPECIFICATIONS RELOCATED SPECIFICATIONS
("Rx" Labeled Comments / Discussions)
Entergy has evaluated this proposed Technical Specification change and nas determined that it involves no significant hazards cons;deration.
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 j
hazards consideration standards:
1.
Does the change involve a
significant increase in the probability or consequences of an accident previously evaluated?
The proposed change relocates requirements and surveillances for structures, systems, components or variables which did not meet the criteria for inclusion in Technical Specifications as identified in the Application of Selection Criteria to the GGNS Technical Specifications.
The affected structures, systems, components or variables are not assumed to be initiators of analyzed events and are not assumed to mitigate accident or transient events. The requirements and surveillances for these affected structures, systems, components or variables will be relocated from the Technical Specifications to an appropriate administratively controlled document which will continue to be maintained pursuant to 10 CFR 50.59.
In addition, the affected structures, systems, components or variables are addressed in existing surveillance procedures which are also controlled by 10 CFR 50.59 and subject to the change control provisions in the Administrative Controls Section of the Technical Specifications.
Therefore, this 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 a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation.
The proposed change will not impose or eliminate any requirements and adequate control of existing requirements will be maintained.
Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
l Ertcl 2 to GNRO-93.t)0109 Chapter 1 Pa0841 GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS FOR THE CURRENT TECHNICAL SPECIFICATIONS RELOCATED SPECIFICATIONS
("Rx" Labeled Comments / Discussions)
(continued) 3.
Does this change involve a significant reduction in a margin of safety?
The proposed change will not reduce a margin of safety because it has no impact on any safety analysis assumptions.
In addition, the relocated requirements and surveillances for the affected structure, system, component or variable remain the same as the existing Technical Specifications.
Since any future changes to these requirements or the surveillance procedures will be evaluated per the requirements of 10 CFR 50.59, no reduction (significant or insignificant) in a margin of safety will be allowed.
The existing requirement for NRC review and approval of revisions, in accordance with 10 CFR 50.92, to these details proposed for relocation, does not have a specific margin of safety upon which to evaluate.
However, since the proposed change is consistent with the BWR Standard Technical Specifications, NUREG-1434, approved by the NRC Staff, revising the Technical Specificatior.s to reflect the approved level of detail ensures no significant reduction in the margin of safety.
Enct 2 to GNRo-93,00109 Chapter 1 Paga 42 GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS FOR THE CURRENT TECHNICAL SPECIFICATIONS MORE RESTRICTIVE CHANGES
("Mx" Labeled Comments / Discussions)
Entergy 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 provides more stringent requirements for operation of the facility.
These more stringent requirements do not result in operation that will increase the probability of initiating an analyzed event and do not alter assumptions relative to mitigation of an accident or transient event.
The more restrictive requirements continue to ensure process variables, structures, systems and components are maintained consistent with the safety analyses and licensing basis.
Therefore, this 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 a physical alteration of the plant (no new or different type of equipment will be installed) or changes in the methods governing normal plant operation.
The proposed change does impose different requirements.
- However, these changes are consistent with assumptions made in the safety analysis and licensing basis.
Thus, this 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 imposition of more restrictive requirements either has no impact on or increases the margin of plant safety.
As provided in the discussion of the change, each change in this category is by definition providing additional restrictions to enhance plant safety.
The change maintains requirements within safety analyses and licensing bases.
Therefore, this change does not involve a significant reduction in a margin of safety.
,~ _
End. 2 to GNRO-03/00109 i
Chrpter 1 Pags 43 GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS l
FOR THE CURRENT TECHNICAL SPECIFICATIONS l
" GENERIC" LESS RESTRICTIVE CHANGE:
RELOCATING DETAILS TO TECHNICAL SPECIFICATION BASES, UFSAR OR PROCEDURES
(" lax" Labeled Comments / Discussions)
Entergy 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 relocates certain details from the Technical Specifications to the Bases, UFSAR or procedures.
The Bases, UFSAR and procedures containing the relocated information will be maintained in accordance with 10 CFR 50.59 and are subject to the change control provisions in the Administrative Controls section of Technical Specifications.
l Since any changes to the Bases, UFSAR or procedures will be i
evaluated per the requirements of 10 CFR 50.59, no increase (significant or insignificant) in the probability or consequences of an accident previously evaluated will be allowed. Therefore, this 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 a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation.
The proposed change will not impose or eliminate any requirements and adequate control of the information will be maintained.
- Thus, this 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 will not reduce a margin of safety because it has no impact on any safety analysis assumptions.
In addition, the requirements to be transposed from the Technical Specifications to the Bases, UFSAR or procedures are the same as the existing Technical Specifications.
Since any future changes to these requirements in the Bases, UFSAR or procedures will be evaluated per the requirements of 10 CFR 50.59, no reduction (significant or insignificant) in a margin of safety will be allowed.
- - ~_ _
......-- - - - - ~.
1 Enct. 2 to GNRO-9300109 Chapter 1 Page 44 j
l GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS I
FOR THE CURRENT TECHNICAL SPECIFICATIONS l
" GENERIC" LESS RESTRICTIVE CHANGE:
i RELOCATING DETAILS TO TECHNICAL SPECIFICATION BASES, UFSAR OR j
PROCEDURES
(" lax" Labeled Comments / Discussions)
(continued)
The existing requirement for NRC review and approval of revisions, in accordance with 10 CFR 50.92, to these details proposed for relocation, does not have a specific margin of safety upon which to evaluate.
. However, since the proposed Standard Technical change is consistent with the BWR.
Specifications, NUREG-1434, approved by the NRC Staf f, revising 1
the Technical Specifications to reflect the approved level of detail ensures no significant reduction in the margin of safety.
1 i
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1 Enct 2 to GNRO-93/00109 Chtpter 1 Paga 45 1
t i
j i
GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS FOR THE CURRENT TECHNICAL SPECIFICATIONS i
" GENERIC" LESS RESTRICTIVE CHANGE:
]
EXTENDING INSTRUMENTATION STIs AND AOTs i
("LBx" Labeled Comments / Discussions) 4
]
Entergy 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 changes increase the Surveillance Test Intervals
]
i (STIs) and Allowed Out-of-service Times (AOTs) for instrumentation supporting a number of TS functions. There are l
no actual related modifications to any of the affected systems, i
However, the changes are expected to reduce the test related plant scrams ar.d test induced wear on the equipment.
Therefore, there is no significant increase in the probability of occurrence of a previously evaluated accident.
General j
Electric topical reports GENE-770-06-1 and GENE-770-06-2 showed l
that the effects of these extensions of STIs and AOTs, which produced negligible impact, are bounded by previous analyses.
Further, the NRC has reviewed these reports and approved the conclusions on a generic basis. Therefore, 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 design and functional operation of the affected equipment i
are not changed by the proposed revisions.
The proposed changes affect only the STIs and AOTs and will not impact.the function of monitoring system variables over the anticipated ranges for normal operation, anticipated operational occurrences, or accident conditions.
Further, the proposed changes do not introduce any new modes of plant operation, make any physical modifications, or alter any operational setpoints.
Therefore, the possibility of a new or different kind of accident from any previously evaluated is not created.
4 i
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Encl. 2 to GNRO-93.00109 i
Chapter 1 Paga 46 GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS FOR THE CURRENT TECHNICAL SPECIFICATIONS
" GENERIC" LESS RESTRICTIVE CHANGE:
j EXTENDING INSTRUMENTATION STIs AND AOTs
]
("LBx" Labeled Comments / Discussions)
(continued) 3.
Does this change involve a significant reduction in a margin of j
safety?
The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined.
The impact of reduced testing, other than as addressed above, is to allow a longer time interval over which instrument uncertainties (e.g., drift) may act.
The current affected instrumentation setpoints already account for the effects of drift and include a sufficient allowance to tolerate extensions of the STIs.
Implementation of the proposed changes is expected to result in an overall improvement in safety due to:
1.
Reduced testing which results in fewer inadvertent reactor
- trips, less frequent actuation of ESF components, and greater equipment availability.
11.
Improvements in the effectiveness of the operating staff in monitoring and controlling plant operation resulting from less frequent distraction to attend to testing.
Therefore, the proposed changes do not significantly reduce the margin of safety.
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Enct 2 to GNRO-93/00109 Chapttei Page 47 GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS FOR THE CURRENT TECHNICAL SPECIFICATIONS
" GENERIC" LESS RESTRICTIVE CHANGE RELOCATION OF INSTRUMENTATION ONLY REQUIREMENTS
("LCx" Labeled Comments /Discussionst Entergy 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 relocates instrumentation only requirements l
(which provide no post-accident function) from the Technical Specifications to the Bases, UFSAR or procedures.
These requirements are part of the routine operational monitoring and I
are not considered in the safety analysis.
The Bases, UFSAR and procedures containing the relocated information will be l
maintained in accordance with 10 CFR 50.59 and are subject to l
the change control provisions in the Administrative Controls section of Technical Specifications.
Since any changes to the
- Bases, UFSAR or procedures will be evaluated per the requirements of 10 CFR 50.59, no increase (significant or insignificant) in the probability or consequences of an accident previously evaluated will be allowed. Therefore, this 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 a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation.
The proposed change will not impose or eliminate any requirements and adequate control of the information will be maintained.
- Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
i 1
Encl. 2 to GNRO-93tX)109 Chrpter 1 Page 48 GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS FOR THE CURRENT TECHNICAL SPECIFICATIONS
" GENERIC" LESS RESTRICTIVE CHANGE:
RELOCATION OF INSTRUMENTATION ONLY REQUIREMENTS
("LCx" Labeled Comments / Discussions)
(continued) 3.
Does this change involve a significant reduction in a margin of safety?
The proposed change will not reduce a margin of safety because it has no impact on any safety analysis assumptions.
In addition, the requirements to be transposed from the Technical Specifications to the Bases, UFSAR or procedures are the same as the existing Technical Specifications.
Since any future changes to these requirements in the Bases, UFSAR or procedures will be evaluated per the requirements of 10 CFR 50.59, no reduction (significant or insignificant) in a margin of safety will be allowed.
The existing requirement for NRC review and approval of revisions, in accordance with 10 CFR 50.92, to these details proposed for relocation, does not nave a specific margin of safety upon which to evaluate.
However, since the proposed change is consistent with the BWR Standard Technical Specifications, NUREG-14 34, approved by the NRC Staf f, revising the Technical Specifications to reflect the approved level of detail ensures no significant reduction in the margin of safety.
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Enct 2 ts GNRO-93KJ0109 Chapter 1 Paga 49 GENERIC NO SIGNIFICANT HAZARDS CONSIDERATIONS FOR THE CURRENT TECHNICAL SPECIFICATIONS ENVIRONMENTAL ASSESSMENT This proposed Technical Specification change has been evaluated against the criteria for and identification of licensing and i
regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21.
It has been determined that the proposed change meets the criteria for categorical exclusion as provided for under 10 CFR 51. 2 2 (c) (9).
The following is a discussion of how the proposed Technical Specification change meets the criteria for categorical exclusion.
Although the proposed change involves changes to requirements with respect to inspection or surveillance requirements:
(i) the proposed change involves no Significant Hazards Consideration (refer to the Significant Hazards Consideration section of this Technical Specification Change Request),
(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite since the proposed change does not affect the generation of any radioactive effluents nor does it affect any of the permitted release paths, and (iii) there is no significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51. 22 (c) (9).
Based on the aforementioned and pursuant to 10 CFR 51.22 (b), no environmental assessment or environmental impact statement need be prepared in connection with issuance of an amendment to the Technical Specifications incorporating the changes proposed in this request.
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Enct 2 to GWRO 93/00109 Chapter 1 Page 50 i
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ATTACHMENT 2 I
i i
j ITS - PSTS i
COMPARISON DOCUMENT i
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i i
2A:
MARKUP OF ITS i
2B:
DISCUSSION OF CHANGES 1
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End 2 to GNRO-9300109 Chapter 1 Page 51 f
ATTACHMENT 2A ITS - PSTS COMPARISON DOCUMENT MARKUP OF ITS
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Enct 2 to GNRC-93/o0109 Chapter 1 Pege s2 Definitions 1.1 1.0 USE AND APPLICATION L
(
1.1 Definitions
....................................-N0TE-----..-.....--------
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.
A AV AGE BUNDLE EXPO 5URt TL; A" GE BUNDLE EXPOSURE s j
sum of the axi e--
cd exposure of the fuel l
rods in the specified b
^ ' ed by the number l
gj of fuel rods in the fuel bundle.
-WK 2-AVERAF PL NE The AVERAGE PLANAR EXPOSURE shall be a i "is to l
N i
a wccific nian ua to the sum l
'N gi cxposure o a ds in the l
N specified bundle at the specified heig ad
\\
by the number of fuel rods in the fuel bundle.
j l
/
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) p=lHGRsf hc;t gcncretica rate p;r uni; length of B) f ! cd' for all the fuel rods in the specified
~
bundle at the specified height divided by the number of fuel rods in the fuel bundle -[at tha -
I-h;ighi].
CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as l
necessary, of the channel output such that it responds within the necessary range and accuracy l
to known values of the parameter that the channel monitors.
The CHANNEL CALIBRATION shall encompass l
the entire channel, including the required sensor, alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST.
Calibration of instrument channels with resistance temperature
{
consist of an inplace(e m cN i Mati d of the y detector (RID) or thermocouple sensors G E Q reg
(,9 '~
stwitm: elemen m and normal calibration of the ya,Aue amset, gp or be}Mje (continued)
GRArJD G UL F o
- g 1.J.1-h W /S STS --
Rev.
0,-49/2G/92 I
i
Enct 2 to GNRO-93.00109 Chapter i Page 53 Definitions 1.1 1.1 Definitions t
CHANNEL CALIBRATION remaining adjustable devices in the channel.
(continued)
W evef a sensing e ent Ti~re
~ a he ne C?
' requi d inplace oss calibr on consis f
,c aring t ther sensi elements the_)
recentiv ' stalled se no elemen The CHANNEL y
CALIBRATION may be performed by meara of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.
CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation.
This detemination shall include, where possible, comparison of the char.nel 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, J a.
Ana'eq deene'lTMthe injection of a simulated i
M cr actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarm,
(
interlock, display, and trip functions, and C9,
channel failure trips.q E
istable cnannels (e.g.,, pressure switches and 3swi ontacts)-the injee ion of a simulated or actua
' nal into the cha 1 as close % '
i d
the sensor as p icable to veri b
M BILITY, includi uired alam td trip
&nctTas.J
-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 tested.
CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources,4 reactivity control componentsp3W Pdnif f C
picer;:nentp effecting react-twityl within the reactor i
vessel with the vessel head removed and fuel in the vessel. Movement of source range monitors, local power range monitors, intermediate range (continued)
BWR/6 STS 1.1-2 Rev.
O, 09/28/92
Encl. 2 to GNRO-93/00109 Chapter 1 Pags s4 Derin m ons 1.1 1.1 Definitions
(
CORE ALTERATION monitors, traversing incore probes, or special (continued) movable detectors (including undervessel replacement) is not considered a CORE ALTERATION.
In addition, control rod movement with other than the nomal control rod drive is not considered a CORE ALTERATION provided there are no fuel assemblies in the associated core cell.
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 re_ load cycle in accordance with Specification C.9.1.0 Plant operation within these limits is addressed in individual Specifications.
gf,g.@
l DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration l
of I-131 (microcuries/ gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132, I-133,1-134, l
f and I-135 actually present. The thyroid dose t
conversion factors used for this calculation shall be those listed in/fTable III of TID-14844, AEC, 1962, " Calculation of Distance factors for Power and Test Rsactor Sites" Or the c listcd in f Nui a i.v., Lids 1.1G^, 6. 1,-
c 1
E-AVERAGE E shall be the average (weighted in proportion DISINTEGRATION ENERGY to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the 4
sum of the average beta and gama energies per p-disintegration (in MeV) for isotopes, other than iodines, with half lives Q15fminutes, making up at least 95% of the total noniodine activity in the coolant.
l EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval l
SYSTEM (ECCS) RESPONSE from when the monitored parameter exceeds its ECCS TIME initiation setpoint at the channel sensor until the ECCS equipment is capable of perfoming its safety function (i.e., the valves travel to their (continued)
BWR/6 STS 1.1-3 Rev.
O, 09/28/92 i
End. 2 to GNRO-93/o0109 Chapter 1 Pagi s5 DeTinitions 1.1 1.1 Definitions
(
EMERGENCY CORE COOLING required positions, pump discharge pressures reach SYSTEM (ECCS) RESPONSE their required values, 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.
g END Of CYCLE The E0C-RPT SYSTEM RESPONSE TIME shall be thatj RECIRCULATION PUMP TRIP time interval from initial he;===:- n (EOC-RPT) SYSTEM RESPONSE A the associated turbine stop valve & n r d or TIME
,4fccm Sge turbine control valve Map 4tiMeaH)-
91,
bett5cl ci, p re 5 5 u rc :'r :: Ocie IIIe we55_ure [
hets 540-intb~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, Mexceptforthebreakerarcsuppressiontime, OB1 / which is not measured but is validated to conform to the manufacturer's design value)?'
ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that l (
RESPONSE TIME
. ;#y# time interval from when the monitored parameter eAceeds its isolatio9 setpoint at the channel sensor until the isolation valves travel to their required positions.
Times shall 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
- 1. 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) l I
l BWR/6 STS 1.1-4 Rev.
O, 09/28/92
Enct. 2 to GNRO-9100109 Chapter 1 Page 56 Definitions 1.1 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.
All LEAKAGE into the drywell that is not identified LEAKAGE; c.
Total LEAKAGE Sum of the identified and unidentified LEAKAGE; t
d.
Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.
LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per RATE (LHGR) unit length of fuel rod.
It is the integral of k
the heat flux over the heat transfer area A -
associated with t e unit length.
LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCT 0 AL TEST shall be a test TEST of all logic components (i.e., all+ relays and 0,4 contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, l
ed Ti the actuated (dw+ee2 to verify OPERABILITY.
The r
LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or i
cad total system steps so that the entire logic system is tested.
\\MAXIMU The MFLPD shall be the large g u R LJHITING 1miting p~..;.
uensity in the core.
P0 DENSITY (MFLPD)
The fr
- - - vi i er density shall be e LHGR existing at a given o
ivided by the specified LHGR limit for that bundle 1
i (continued) i BWR/6 STS 1.1-5 Rev.
O, 09/28/92 1
Encl. 2 b GNRO.93/0olo9 Chapter 1 Paga 57 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'54for each Ogg class of fuel}V 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 i
Table 1.1-1 with fuel in the reactor vessel.
(or_Gye cesRAstLo1 duis eQ--@
OPERABLE-OPERABILITY
/ A system, subsystiem,
, component, or device g
shall be OPERABLE w en it is capable of performing its specified safety function (s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary A
equipment that are required for the system, L ivisicui subsystem,L W, component, or device to perform its specified safety function (s) are also capable e
of performing their related support function (s).
(
M ICS TESTS ~
PHYSICS TESTS shall those tests'perf ed to k
measure the fund ntal nuclear cha teristics of
^
the reactor c and related ins umentation.
These tests re:
a.
D ribed in Chapter 14, Initial Test
(
rogram] of the F.;
b.
Authorizedunf(rtheprovisions 10 CFR 50.59d, or i
/
\\
c.
Otherwise approved by the clear 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)
{
BWR/6 STS 1.1-6 Rev.
O, 09/28/92
End. 2 to GNRo.934X)1o2 Chapter 1 pag, sa Definitions 1.1 h [. 7, [. 7.)
1.1 Definitions 1
PRESSURE AND shall be detennined for each fluence period in TEMPERATURE LIMITS accordance with Specification w y.1. Q Plant REPORT (PTLR) operation within these operating limits is (continued) addressed in LCO 3.4.11, "RCS Pressure and Temperature (P/T) Limits."
RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 383 MWt.
REACTOR PROTECTION 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 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.
SDM shall be the amount of reactivity by which the reactor is subcritical or would be suberitical 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.
H Oc))
With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of 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 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 STS 1.1-7 Rev.
O, 09/28/92
~ _ ~
i End 2 2 GNRO-93/o0109
)
.i Chapter 1 Page 59 Definitions l
1.1 l
1.1 Definitions (continued)
\\
4 THERMAL POWER THERMAL POWER shall be the total reactor core heat 4
transfer rate to the reactor coolant.
I i
/
TURBINE BYP 55 SYSTEM T
TURBINE BYPASS SYSTEM R PONSE TIME consists 1
1
RESPONSE
ME f two components:
l a.
The time for initi movement of the ma' l
turbine stop val or control valve un il 80%
r\\
of the turbine ypass capacity is es blished; and b.
The time f r initial movement o the main
/
turbine op valve or control alve until initia movement of the turb' e bypass valve.
The res nse time may be meas ed by means of any serie of sequential, overla ing, or total steps so at the entire respons time is measured.
(
1 9
2 i
BWR/6 STS 1.1-8 Rev.
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.- w
Encl. 2 to GNRO 9100109 Chapter 1 Page6o Definitions 1.1 Table 1.1-1 (page 1 of 1) t MODES REACTOR MODE AVERAGE REACTOR MODE TITLE SWITCH POSITION COOLANT TEMPERATURE
(*F) 1 Power Operation Run NA 2
Startup Refuel (a) or Startup/ Hot NA Standby Oh 3
Hot Shutdown (a)
Shutdown 4
Cold Shutdown (a)
A A
3 _
0}
fShutdown s
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.
i l
i BWR/6 STS 1.1-9 Rev.
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Enct. 2 to GNRO-9300to9 Chapter i Psga 61 Logical Connectors 1.2 1.0 USE AND APPLICATION I
1.2 Logical Connectors PtJRPOSE The purpose of this section 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 QR.
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 indentions of the logical connectors.
Ocy.
When 1 ical connectors are u d to state a Conditimi only the rst level of logic is sed,andthelogicay6cnn,ector is eft justified with th Condition statement./f j
When logical connectors are used to state afCompletion Time d Surveillance, or Frequency, only the first level of logic is used, and the logical connector is left justified with the statement of the Completion Time, Surveillance, or Frequency.
h{ bN>
EXAMPLES The following examples illustrate the use of logical connectors.
(continued)
G kJoGtttp
'o M /S STS 1./-10 Rev.
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Enct. 2 to GNRO-0300109 Chapter 1 PQe 62 Logical Connectors 1.2 1.2 Logical Connectors
(
EXAMPLES EXAMPLE 1.2-1 (continued)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
LCO not met.
A.1 Veri fy...
AND A.2 Restore...
1 In this example, the logical connector 6ND is used to indicate that, when in Condition A, both Required Actions A.1 and A.2 must be completed.
(
(continued)
BWR/6 STS 1.2-11 Rev.
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i End. 2 b GNRO-9300109 Chipaw 1 Pans 63 j
Logical Connectors 1.2 1.2 Logical Connectors
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EXAMPLES EXAMPLE 1.2-2 (continued)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
LCO not met.
A.1 Trip...
98 l
A.2.1 Veri fy...
AND A.2.2.1 Reduce...
QB A.2.2.2 Perform...
QB A.3 Align...
(
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 QB and the left justified placement. Any one of these three Actions say 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 QB 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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BWR/6 STS 1.2-12 Rev.
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Enct 2 to GNRO-93/colo9 Chapter i Ptoe 64 Completion Times 1.3 1.0 USE AND APPLICATION l
1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.
f[lin1HQ Cndi%s km Opew%d LCOs)specifyfiEinimum requireinents for ensuring safe BACKGROUND operation of the unit. The ACTIONS associated with an LCO CW state Conditions that typically describe the ways in which the requirements of the LC0 can fail to be met. Specified with each stated Condition are Required Action (s) and Completion Time (s).
DESCRIPTION The Completion Time is the amount of time allowed for 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 Condition remains in effect and 1
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. _
d Once a Condition has been entered, subsequent t ni 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 continue to apply to each additional failure, with Completion Times based on initial entry into the Condition.
(continued)
BWR/6 STS
- 1. -13 Rev. O,09/28/92
Ercl 2 to GNRO-93/o01o9 Chapter 1 Prm 65 Completion Times 1.3 1.3 Completion Times m
I i(i15Joh g
DESCRIPTION However, when a subsecuent teiffr( subsystem, component, or (continued) variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time (s) may i
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, i
and b.
Must remain inoperable or not within limits after the j
first inoperability is resolved.
i The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be 1
limited to the more restrictive of either:
i a.
The stated Completion Time, as measured from the initial entry into the Condition, plus 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.
TheaboveCompletionTimeextension!donotapplytothose
(
Specifications that have exceptions that allow completel gf separate re-entry into the Condition (for each train, di g.s, on subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based i
on this re-entry.
These exceptions are stated in individual i
Specifications.
The above Completion Time extension does not apply to a Completion Time with a modified " time zero."
This modified
" time zero" may be expressed 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 />," where 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 c,f 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.
I (Continued)
BWR/6 STS 1.3-14 Rev.
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w
Err) 2 to GNRO-93,00109 CompletInMtes"*"
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 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 />.
(continued)
(
BWR/6 STS 1.3-15 Rev.
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... ~.
End 2 to GNRO-93/Doio9 r
Chroter 1 Paas 67 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.
h 5
When a pump is declare inoperable, Condition A is entered.
(
O If the pump is not res red to OPERABLE status within 7 days, Condition B is ntered and the Completion Time clocks for Required Actions B.1 and B.2 start.
If the CondM.
inoperable pump is restored to OPERABLE status after 4 uf B arc Condition B is entereds the Required Actions of Condition B gg may be terminated.
Muubb When a second pump is declared inoperable while the first pump is still inoperable, Condition A is not re-entered for the second pump. LC0 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 LC0 3.0.3 is entered, but continues to be tracked from the time Condition A was initially entered.
While in LC0 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)
(
BWR/6 STS 1.3-16 Rev. O,09/28/92
Encl. 2 to GNRO-9M)0109 Che mr1 Pt w 68 Completion Times 1.3 1.3 Completion Times j
(
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 operation continued in accordance with Condition B.
The l
Completion Time for Condition B is tracked from the time the Condition A Completion Time expired.
On restoring one of the pumps to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first pump was declared inoperable. This 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 does not result in the second pump being inoperable for
> 7 days.
i
)
(
l i
i (continued) 4 BWR/6 STS 1.3-17 Rev.
O, 09/28/92
4 Enct 2 to GNRO-93/C0109 Qupert Pge 60 Corpletion Times 1.3
\\
r 1.3 Completion Times t
EXAMPLES EXAMPLE 1.3-3 (continued)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Restore 7 days Function X Function X subsystem subsystem to AND inoperable.
OPERABLE status.
10 days from discovery of failure to meet the LCO l
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 ANQ inoperable.
OPERABLE status.
)
10 days from
(
discovery of failure to meet the LCO 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 QR 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.
(continued)
BWR/6 STS 1.3-18 Rev.
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Enct 3 to GMRO-9300109 Chapter 1 Pago 70 i
Completion Times 1.3 l
1.3 Completion Times t
EXAMPLES EXAMPLE 1.3-3 (continued) subsy skm When one Function X te m and ne Function Y are inoperable, Condition A and ondition B are concurrently applicable. The Completion Times for Condition A and Condition B are tracked eparately for each L.. / starting from the time each was declared inoperable and the Condition was entered.
A separate Completion Time is established for Condition C and tracked from the time the J l
second tr;in'.;as declared inoperabTe (i.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 i
Cg remaining Completion Time in Condition A is measured from i
g the time the affected trairr was declared inoperable (i.e.,
gh slun initial entry into Condition A).
The C mnletion 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 indefinite continued operation while not meeting the LCO.
This Completion Time allows for an exception to the normal
" 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.
1 (continued)
'i BWR/6 STS 1.3-19 Rev.
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End. 2 to GNRO-93/00109 Chapter 1 Pags 71 Completion Times 1.3 1.3 Conpletion Times
\\
EXAMPLES EXAMPLE 1.3-4 (continued)
ACTIONS 1
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more A.1 Restore valve (s) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> valves to OPERABLE l
- status, i
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 l
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.
l
(
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 /> (8:
extension )
V expires while one or more valves are still inoperable, i
Condition B is entered.
(continued)
(
j BWR/6 STS 1.3-20 Rev. O,09/28/92
Enct 2 6 GNRO-921oo109 Chapter 1 Page 73 Completion Times 1.3 i
1.3 Completion Times t
EXAMPLES EXAMPLE 1.3-5 (continued)
ACTIONS
--...........--........-----NOTE----------------------------
Separate Condition entry is allowed for each inoperable valve.
i CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more A.1 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
(
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.
b The Note above the ACTIONS table is a method of modifying
'g how the Completion Time is tracked.
If this method of e
modifying how the Completion Time is tracked was applicable
/
Q WLCML only tiWCondition g the Note appear in Condition c o h=.
out tha i
rofby M d g g p.( 4}g The Note allows Condition A to be entered separately for each inoperable valve, and Completion Times tracked on a per ACT1005 TOM
- 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)
BWR/6.STS 1.3-21 Rev.
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t Encl. 2 to GNRO-93/00109 Chapter 1 Page 73 Completion Times 1.3 1.3 Completion Times 1
EXAMPLES EXAMPLE 1.3-5 (continued)
If the Completion Time associated with a valve in Condition 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.
Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times, Completion i
Time extensions do not apply.
EXAMPLE 1.3-6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME f
A. One channel A.1 Perform Once per l
(
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 /> 08 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.
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 Completion Time not met.
l (continued) 1 BWR/6 STS 1.3-22 Rev.
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End. 2 to GNRO D300109 Qupter 1 Pgn 74 Completion Times 1.3 1.3 Completion Times
(
EXAMPLES EXAMPLE 1.3-6 (continued) l Entry into Condition A offers a choice between Required Action A.1 or A.2.
Required Action A.1 has a "once per" Completion Time, which qualifies for the 25% extension, per i
SR 3.0.2, to each performance after the initial performance.
If Required Action A.1 is followed and the Required Action
- s d
is not met within the Completion Time (p o ur m the @lH h b y extension allowed by SR 3.0.2), Condition B is entered.
If ~ ~
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.
j 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 1
i l
i i
(continued) i BWR/6 STS 1.3-23 Rev.
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End 2 to GNRO-91Co109 CNemr1 pyr, 75 Completion Times 1.3 1.3 Completion Times
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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 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.
Required Action A.1 has two Completion Times. The I hour Completion Time begins at the time the Condition is entered 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 Rtquired Action A.I.
If after Condition A is entered, Required Action A.1 is not met within either the initial 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or any subsequent 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval from the previous C4 ~]
M extension allowed by SR 3.0.2) performance (en:::::^0 the Condition B is entered.
The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered.
If Required Action A.1 (continued) i BWR/6 STS 1.3-24 Rev.
O, 09/28/92 7
i..
Enct 2 to GNRO-9300109 Qupkr1 Prp 76 Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued) is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expi red. _$ince e second Compl n Time of Requir ction A.1 a modified " '
zero" (i.e., a the
$ 9
- initi hour, not fr ime of Conditio ry), thg i
a ance for a etion Time ex on does not Epply.
IMMEDIATE When "Immediately" is used as a Completion Time, the COMPLETION TIME Required Action should be pursued without delay and in a controlled manner.
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i i
I.
BWR/6 STS 1.3-25 Rev.
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Encl. 2 to GNRO-93/00109 Chapter 1 Page 77 Frequency 1.4 1.0 USE AND APPLICATION t
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 in which the Surveillance must be met in order to meet the associated man understanding of the correct application j
C,g of the specified) Frequency is necessary for compliance with the SR. 3Q; Qg p gggJM]Hd j
u
~-
The "specified Frequency" is referred to throughout this section 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 LC0 is within its Applicability, represent i
potential SR 3.0.4 conflicts.
To avoid these conflicts, the l
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 4
restriction.
The use of rfomed" in these instances conveys specified meanings. A urveillance is " met" only when the acceptance criteria are satisfied.
Known failure of the requirements of a Surveillance, even without a Surveillance specifically being "perfomed," constitutes a Surveillance not " met."
" Performance" refers only to the requirement to specifically detemine the ability to meet the acceptance (continued) i BWR/6 STS
.1.f-26 '
W,)
1 Rev.
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l}
^
Enct 2 to GNRO-9100109 chapter 1 Paga 78 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 Surveillance is not required to be perfomed; and 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 examples, the Applicability of the LCO (LCO not shown) is MODES 1, 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 i
specifies an interval (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) during which the associated C4 Surveillance must be perfomed at least one time.
perfomance of the Surveillance initiates the subsequent 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
[~~ m,
extension of the time interval to 1.25 times the otated)g l
( s' 'p*e c o 8
- Frequency is allowed by SR 3.0.2 for operational b
flexibility.
The measurement of this interval continues at
,g 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 MODE or other specified condition in the Applicability of the LCO, and the perfomance of the Surveillance is not otherwise modified (continued)
BWR/6 STS 1.4 27 Rev.
O, 09/28/92
t 6
End. 2 to GNRO-93/001of f
Chapter 1 Pago 79 Frequency 1.4 l
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 MODE or other specified condition in the Applicability of the LCO for which performance of the SR is required, the Surveillance must be performed 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 FREQUENCY 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 e 25% RTP
(
l m
l 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 e 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.
Ct/
(continued)
(
BWR/6 STS 1.4-28 Rev.
O, 09/28/92
Enct 2 to GNRO.920010E Chapter 1 Pege Bo Frequency 1.4 1
1.4 Frequency
\\
EXAMPLES EXAMPLE 1.4-2 (continued)
"Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified i
condition is first met (i.e., the "once" performance in this example).
If reactor power decreases to < 25% RTP, the l
measurement of both intervals stops. New intervals start l
upon reactor power reaching 25% RTP, EXAMPLE 1,4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l
.____............-N0TE------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter a 25% RTP.
l Perfom channel adjustment.
7 drys
(
The interval continues whether or not the unit operation is
< 25% RTP between performances.
I As the Note modifies the required perfomance 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 /> af ter power reaches a 25% RTP to perfom the Surveillance.
The Surveillance is still considered to be within the "specified frequency." Therefore, if th_e Surveillance were not i
gq _j perfomed within the 7 day BM 255 pr = :.".2)) interva 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 the 7 day Frequency not met, provided operation does not 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 edense aga % se s.o.ej (Continued)
BWR/6 STS 1.4-29 Rev.
O, 09/28/92
.~.
Enct. 2 to GNRO-93,00109 Chapter 1 Page 81 j
l' Frequency 1.4
(
1.4 Frequency t
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 perfonned 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__ gerfonn a Surveillance within the specified Frequen gf....
m.....r.,
um. m u utw._ o o OCll g (::::ld apply.
r:r.:: MS SP 3.^."
d the provisions of SR 3.0.3 wou EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i
__________________ NOTE--_---------------
Only required to be met in MODE 1.
l l
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 /> l
(
~
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.
g4 Therefore, if the Surveillance were not performed within the 4
24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (P e m p the @ extension allowed by SR 3.0.2) f interval, but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no 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.
~
4 BWR/6 STS 1.4-30 Rev.
O, 09/28/92
Enct. 2 to GNRO-93/00109 Chapter 1 Page 82 a
ATTACHMENT 2B ITS - PSTS COMPARISON DOCUMENT DISCUSSION OF CHANGES
(
1 i
i a
i d
i i
o l
A
-,-----,--,-----.,,..-.n
-,--.~._,--.,_..n,<-
..w-,-,.-n.w~,--e---.-r,-
.-.....,~4
End 2 to GNRO-93.00109 Chapter 1 Pags83 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 1 - USE and APPLICATION BRACKETED ADMINISTRATIVE CHOICE B.1 Brackets removed and optional wording revised 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, Testing") and the definition is not
" Recirculation Loops required.
P.2 Consistent with the plant specific safety
- analyses, the appropriate initiation signals assumed are reflected in 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 BASES, 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.5 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.
C.6 These changes are editorial corrections of typographical or grammatical errors.
e
,.y-w
-n-.--
s'~
a-
tv w
w
4 Enct. 2 to GNRo-9100109
?
Chapter 1 Page 84 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 1 - USE and APPLICATION 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 j
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 devicas, if they are developed.
C.8 Including the terms " interlock" and " display" in the list of attributes to verify OPERABILITY for an analog channel i
introduces confusion for bistable channels which have an interlock function.
Since the list is intended to provide d
examples of attributes which must potentially be OPERABLE, l
dependent on whether it is " required" or not, the list can be i
applied to both analog and bistable channels.
This revision will remove any perception of an intended difference.
a 4
C.9 "Other compor.unts affecting reactivity" could easily be misinterpreted to include the manipulation of components which i
cause parameter changes that affect reactivity.
For example, any activity that causes a moderator temperature change would j
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 j
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.
j C.10 The term
" equipment" is used in place of
" device" for l
consistency with the terminology from IEEE-308.
I C.11 In the example being discussed, the unit is in MODE 1, (e.g.,
1 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.
j
t l
Ertet 2 6 GNRO-03/00109 l
Chapbr 2 Pago1 1
i, 1
l aRAND GULF i
i, l
CHAPTER 2 1
2 ii 1
a 1
i 4
i
)
k 1
l 1
i
_. _ _ _ _.. _ _ _ _... _, _.. _ _ _ _,... _..... ~. _.,, - _..... _.. _ _ _ _ _. _
i I
Encl. 2 to GNRO-9300109 Chapter 2 Pags 2 I
I i
I ATTACHMENT 1 k
i 1
i CTS - PSTS 4
I COMPARISON DOCUMENT i
1 A:
MARKUP OF CTS 1 B:
DISCUSSION OF CHANGES 1C:
NO SIGNIFICANT HAZARDS CONSIDERATIONS
._m.._my
---.-.~r,
I End. 2 to GNRO-9300109 i
Chapter 2 Page 3 i
I 1
1 i
l J
ATTACHMENT 1 A i
CTS - PSTS COMPARISON DOCUMENT l
)
l MARKUP OF CTS 4
l i
l
Enci 2 to GNRO-9300109 Chapter 2 Prge 4 i
{
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l
2.1 SAFETY LIMITS THERMAL POWER. Low Pressure or low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the 2.03 reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of.
rated flow.
Q (APPLICABILITY: OPERATIONALCONDITIONSIand2.)
FACTION:
O With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel i
steam dome pressure less than 785 psig or core flow less than 10% of rated fl ow.
Ibe in at least HOT SHUTDOWW 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
,5pecification 6.7.1.
44 lTHERMALPOWER.HiahPressureandHiahFlow
- .l L1 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 during two loop operation and 1.07 during single loop operation with the l
reactor vessel steam done pressure (greater thanJ785 psig and core flowfgreateD 2
@l0% of rated flow.
QPPLICABILITY: OPERATIONAL CONDITIONS I an d ACTION:
N WithMCPR1essthantheaovelimits'tandthereactorvesselsteamdomepressure
.creater r.hanl785 ssig and core flow (greater thanJ10% of rated flow,f be in at3 Seast H07 SHUTDOWt within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />sland comply wRh the requirements of Specifi'-
~
' cation 6. 7.1.
[
Z.2.2) 3 REACTOR COOLANT SYSTEM PRESSURE
~
2.0 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.
g@PPLICABILITY: OPERATIONAL CONDITIONS 1, 2. 3 and 47 ACTION:
6 1.2.Q 2,1 With the reactor coolant system pressure, meas W in the reactor vessel
_ steam dome, above 1325 psia. Ge in at least HOT SHUTDOWNTwith re_ actor coolant]
lsystem pressure less than or equal to 1325 nsin within 2~hoursIand comply with the requirements of Specification 6.7.1.
[
GRAND GULF-UNIT 1 2-1 Amendment No. M, 99,106
Enci. 2 to GNRO-9100109 Chapter 2 P:gs 5 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS (Continued)
REACTOR VESSEL WATER LEVEL 2.1 \\*3 2.1.4 The reactor vessel water level shall be above the top of the active ig eradiated fuel.
OPPLICABI'ITY:
OFfRATIONALCONDITIONS3,4and5]
ACTION:
With the reactor vessel water level at or below the top of the active irradiated fuel,fmanually initiate the ECCS to restore the water leveQ 17*7,)kQepressurize the reactor vessel as necessary for Ects eneration.J Cosg with the requirements of Specification 6,7.1.
4 GRAND GULF-UNIT 1 2-2
Enct 2 to GNRO-0300109 i
Chapter 2 Pag 36 SAFETY LIMIT 5 AND LIMITING $AFETY $YSTEM $tTTI ES i
'2.2 LIMIT!NG $AF(TY $YST[M $tTT!W 5 REACTOR PROTECTION SYSTEM INSTRUMtwfATION SETPOINTS j
2.2.1 The reactor protectica systes instrumentation setpoints shall be set ~
consistent with tne Trip 5etpoint values shown in Table 2.2.1-1.
APPLICABILITY: As shown in Table 3.3.1-1.
ACTION:
LsSS i cldlwg Tale
- 2. 7.1 - )
With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, esclare the channel inoperable and apply the applicable ACTION statement requirement 1
of Specification 3.3.1 unt11 the channel is restored to OPERABLE status with its setpoint acjustee consistent with the Trip Setpoint value.
i l
h fL60 3.3.1. I l
(O
End. 2 to GNROMCD109 Chapter 2 Page 7 l
s
)
NL-90/06 ADMINISTRATIVE CONTROLS ACTIVITIES (Continued)
Proposed tests and experiments which affect plant nucir
.i c.
are not addressed in the Final Safety Analysis Repor+
t by an individual / group other than the individual /p-
.m d the proposed test or experiment.
ared d.
Events reportable pursuant to Section 50.73 +
shall be investigated and a report prepare'
,o and which provides recommendations to pr' es the event shall be approved by the General Manac
.4.
Such report
.tions, Individuals responsible for review" e.
- 6. 5. 3.1. a. 6. 5. 3.1. b
- 6. 5. 3.1. c -
4ccordance with x
the qualification requirements
.aall meet or exceed as previously designated by
,of ANSI 18.1, 1971, General Manager, Plant Op' h
- licable.sent, Operations GGNS or shall include a determi' s
Each such review j{t(
.er or not additional, cross-disciplinary review i
.f deemed necessary such review shall be performed discipline, personnel of the appro,priate
-k determination of whether or not an f.
Each review "'
unreviewed g
a is involved.
g.
Recorde Mans'
.ctivities shall be provided to the General 4tions, PSRC and/or as necessary for required 9
re-6 g
6.6 REP ACTION 6.F
.ng actions shall be taken for REPORTA8LE EVENTS:
)
, Commission shall be notified pursuant to the requirements of section 50.72 to 10 CFR Part 50, and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and Each REPORTABLE EVENT shall be reviewed by the PSRC and submi o.
to the SRC and the Vice President, Operations GGNS.
t 6. 7 SAFETY LIMIT VIOLATION 2.1 6.7.1 The following actions shall be taken in the event a Safety Limit i violated:
m-Plad Ma%
- 2. 2. j '.
p The NRC Operations Center shall be notified w telephone as soon as g,g,3 -Q' -)42perations uwinand the SRC shall be notified within n all mts within nam hour.J he5vice Presteeng T
/ ' Cl g/jb.
A 6E(5afet.ylimit violation RepoW)shall be orenared. [The report sha11P
~51r-g ti 'a; ""3 This rt s descr1 (1) appli e y
0L83,.J[upon circums s proc the olation
) effec of the ation 2.2.q, compo s, sy a or st
- tures, (3) co tive ac ns tak prev recur e,>
b JOSEpcr 2-3 GRAND GULF-UNIT 1 t3-Amendment No. 76
Enct 2 to GNRO-9300109 Chapter 2 P:ge 8 ADMINISTRATIVE CONTROLS SAFETY LIMIT VIOLATION (Continued) h
=
" c.
The(Safety Limit Violation Reporjshall be submitted to the Nuclear Regulatory Commission pursuant to Section 50.4 of 10 CF A* q Part 50, l the SRC and the LVice Presiaent. Operations GGN$)within M days of h
i the violation 2Z'5 Critical opera
~ ]%g g
ag d.
-by the Commission.on of the unit shall not be resumed until authorized I
- 6. 8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and mai covering the activities referenced below:
The applicable procedures recommended in Appendix "V a.
Guide 1.33, Revision 2, February 1978.
y b.
Refueling operations.
Surveillance and test activities of safety rr c.
.it.
d.
Security Plan implementation.
Emergency Plan implementation.
e.
f.
Fire Protection Program implementati s
g.
PROCESS CONTROL PROGRAM implement O,
h.
OFFSITE DOSE CALCULATION MANUT
.on.
i.
Quality Assurance Program f
,d environmental monitoring, using the guidance in Rec-4.15, February 1979.
6.8.2 Each procedure of 6.8.1 **
as required by 6.5, above, pri-
.ges thereto, shall be reviewed periodically as set forth in
. cation and shall be reviewed N
e procedures.
h 6.8.3 The following pro'
, established, implemented, and maintained:
a.
Primary Cor 8
Jutside Containment A progr a
saakage from those portions of systees outside conti 5
suld contain highly radioactive fluids during a se
or accident to as low as practical levels.
The e
a the:
/ stem outside containment containing steam or water, st the drain line to the main condenser.
. R system outside containment containing steam or water, except the line to the LRW system and headers that are isolated by manual valves.
3.
HPCS system.
4.
LPCS system.
5.
Hydrogen analyzers of the combustible gas control system.
'rOSERT 1-3 GRAND GULF-UNIT 1
+ 14 Amendment No.80
~
Encl. 2 to GNRO-93.t)0109 Chapter 2 Page 9 i
l I
4 I
ATTACHMENT 1B 1
CTS - PSTS COMPARISON DOCUMENT DISCUSSION OF CHANGES 4
Encl. 2 to GNRO-93/00109 Chapter 2 Pogs 10 DISCUSSION OF CHANGES l
CIS: 2 - SAFETY LIMITS 1
ADMINISTRATIVE A.1 hhere 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 UFSAR).
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 controlled document with specific regulatory review requirements for changes, such as the Bases, UFSAR or OQAM.
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 Specifications, NUREG-1434.
Any technical changes to this requirement will be addressed with the content of the proposed chapter location.
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.
A.4 This comment number is not used for this station.
End. 2 to GNRo-93/00109 Chapter 2 Page 11 DISCUSSION OF CHANGES CTS: 2 - SAFETY LIMITS RELOCATED SPECIFICATIONS None in this section.
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 should receive the same attention and response.
M.2 An additional reporting requirement of notification of the highest level management specifically responsible for the operation of the plant is added.
M.3 Limits on steam dome pressure and core flow are to be specified as 2.
The current SLs do not address a pressure or flow which is equal to the limit.
This proposed change will. resolve an incontinuity between current SL 2.1.1 and SL 2.1.2.
j TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 This comment number is not used for this station.
LA.2 The Required Action has been made less specific to allow 1
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.
The time frame for completion of the action is made consistent with the allowed time to restore other Safety Limit violations.
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, 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.
Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process in Chapter 5 of the Technical Specifications.
,... _ _ _ _. _ _ _ _ _. _ _ _..,... ~ _ _ _, -
End. 2 to GNRO-2300109 Chapter 2 Pcge 12 DISCUSSION OF CHANGES i
CTS: 2 - SAFETY LIMITS l
l TECHNICAL CHANGE - LESS RESTRICTIVE (continued)
" 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 current Technical Specification Section 6.5.-
_ _ _ _.... _. _ _.,. _ _,... ~,.. _.,.,.., _. _. _ _. -.. -. _ -,. _ -.
Encl. 2 to GNRO-93/00109 Chapter 2 Paga 13 i
'l ATTACHMENT 1C CTS - PSTS COMPARISON DOCUMENT i
NO SIGNIFICANT HAZARDS CONSIDERATIONS 4
,.-_--,v
,--- =
a--
Encl. 2 to GNRo-93M0109 Chapter 2 Paga 14 NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 2 - SAFETY LIMITS "L1" CHANGE Entergy has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards consideration.
This detennination 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 remove a long term review function j
from the onsite committee that is duplicated by the functions of the offsite committee.
This will allow additional time for review of short term plant conditions.
These proposed reviews are not considered as initiators for any previously evaluated accident and are not required for the mitigation of any
{
evaluated accident.
Therefore, the proposed change will not
)
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 different kind of accident from any accident previously evaluated.
3.
Does this change involve a significant reduction in a margin of safety?
1 i
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.
"c'#?"?.:?s" i
ATTACHMENT 2 ITS - PSTS COMPARISON DOCUMENT 2A:
MARKUP OF ITS 2B:
DISCUSSION OF CHANGES i
1 Enct. 2 to GNRO-9300109 Chapter 2 Page 16 J
2 a
4 i
ATTACHMENT 2A ITS - PSTS i
COMPARISON DOCUMENT MARKUP OF ITS
-. -... - - -. - -. ~.... -. -. ~.,. -.. -...
., -...,. -, - -,. -...., -...... ~.,., -.
i Erd. 2 b GNRO-93/00109 l
CMphr 2 Pag a 17 bb5 2.0 4
2.0 SAFETY LIMITS (SLs) i 2.1 SLs i
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 t 785 psig and core flow a 10% rated core flow:
- 1. O(o MCPR shall be a M 1 for two ool recirculation 4 Q operation or n [1.00 for single oo recirculation g operation.
- f. o ?
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 done pressure shall be[ :inte;ng s 1325 psig.
2.2 SL Violations i
With any SL violation, the following actions shall be completed:
i 2.2.1 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 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.
y 2.2.3 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. notify the c:xrq/ Manager, _
=- /Plantl F
+P7i':: ore:iMtT: ' u ^;;r8:::,-and the [ ffrite7evi:=rs ~
fpeci'ied 4-5)::ific;tica 5.5.2, "[^ff:it:]g":vi
- d ^;dit"]b i Ae ccrpcrake e1A4',ve aspens;h\\e y
ggg (g Ietever ab auclear sa4ef
~(continued)
C,RPdb GULY
-BWR/G STS Mb 2.0-1 Rev.
O, 09/28/92
l Enci. 2 to GNRO-93A)0109 Chapter 2 Pag a 18 3($
2.0 f
2.0 SLs 1
2.2 SL Violations (continued) ggc 2.2.4 Within 30 days, a Licensee Event Report (LER) shall be prepared pursuant to_10 CFR 50.73.
The LER shall be submitted to the NRC, thegett:it: - :
e r u i ec - - 5 ::i - ^ ---- 1r 1. 51' :"
@T&Tn;and 6-- -- r-----
- -" rp.0 the l
64 anager,M-M r ~~-t'fWe. corporea execdMve, tespems'eble for overh C-~'
plad nulear s&+g,f-
- - - ~
2.2.5 Operation of the unit shall not be resumed until authorized by the NRC.
i 2.0-2 Rev.
O, 09/28/92 l
1
End. 2 to GNRO-DC0109 Chapter 2 Pcge 19 Reactor Core SLs B 2.1.1 i
B 2.0 SAFETY LIMITS (SLs)
B 2.1.1 Reactor Core SLs BASES
,BACKGR0tJND GDC 10 (Ref.1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients and anticipated operational occurrences A00s).
damage is calculated to occur if the [init is n The fuel _ cladding integrity SL is set such that nolfuel Because fuel damage is not directly observableN,ot violated.
stepback approach is used to establish an SL, such that the MCPR is not less than the limit spes:ified in Specification 2.1.1.2.
bf
}
r f htkMhR gr a t
the 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 life of the cladding, fission product migration from source is incrementally cumulative and continuously measurable.
Fuel cladding perforations, however, can result from thermal stresses, wh: ch occur from reactor operation significantly above design conditions.
While fission product migration from cladding perforation just as measurable as that from use related cracking, the.is thermally caused cladding perforations signal a threshold beyond which still greater themal stresses any cause gross, rather than incremental, cladding deterioration.
i the fuel cladding SL is defined with a margin to theTherefore, j
conditions that would produce onset of transition boiling (i.e.,MCPR=1.00).
These conditions represent a for planned operation.significant departure from the conditio The MCPR fuel cladding iritegrity SL
{
ensures that during normal operation and during A00s, at least 99.9% of the fuel rods in the core do not experience transition boiling.
l (continued)
GRAct GutF
- E"!6 STS gf5 B 2.0-1 Rev.
O, 09/28/92
End. 2 to GNRO-9100109 Chapkr 2 Ptge 20 Reactor Core SLs B 2.1.1 i
j BASES i
1 BACKGROUND Operation above the boundary of the nucleate boiling regime (continued) could result in excessive cladding temperature because of 4
the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.
Inside the steam i
film, high cladding tem water (zirconium water)peratures are reached, and a cladding reaction may take place.
This chemical reaction results in oxidation of the fuel cladding i
to a structurally weaker fons.
This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.
4 APPLICABLE The fuel cladding must not sustain damage as a result of SAFETY ANALYSES nonsal operation and A00s.
The reactor core SLs are established to preclude violation of the fuel design Q
criterion that an MCPR4f s to be established, such that at least 99.9% of the fuel frods-in the core would not be expected to experien the onset of transition boiling.
SL The Reactor Protectio System set
" Reactor Protection System (RPS) points (LCO 3.3.1.1, Instrumentation"),in i
combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER h
level that would result in reaching the MCP b \\
bl.1.la Fuel Claddina lntecr"tv (General Electric /
O@
Corooration (Gl!) Fue' )
GE critic power correlations are applic e for all critical p r calculations at pressu a 785 psig or core flows a 10% o ated flow.
For ope ion at low and low flows, a ther basis is d, as follows: pressures i
Since the pressu d
in the bypass. region is essentially all el tion head, the core pressure i
drop at low p an lows will always be
> 4.5 psi.
yses (R. 2) show that with a j
bundle fl f 28 x 108 lb r, bundle pressure drop is arly independent o undle power and has a alue of 3.5 psi.
- Thus, bundle flow wi a 4.5 psi driving head will i
3 28 x 10 lb/hr. Full scale ATLAS t data taken at pressures from 14.7 psia to 8 sia (c
tinued)
BWR/6 STS B 2.0-2 Rev.
O, 09/28/92
Encl. 2 to GNRO-B340109 Chapter 2 Page 21 Reactor Core SLs B 2.1.1 i
BASES APPLICABLE 2 ckl. la Fuel Claddina Intecrity (GeMal Electric SAFETY ANALYSES Corooration (GE) Fuel) 4 continued)
(Q indicate t ef assembly critical power at W
this flow is ap tely 3.35 Mwt. With the design peak
- factors, s corresponds to a THERMAL ER > 50% RTP.
Th THERMAL POWER limi f 25% RTP for reactor pres re < 785 psig i conservative.
e 2.1.1.lb uel Claddina Inteority-(Aherced hies,-rei 0--h
-Cc : reti:- (f.?) F;:' ) <
The use of the correlati is valid for critical power calculations at pressures >
psig and bundle mass fluxes
> 0.25 x _10' lb/hr-f_t* (Re_f 23.
For operation at low a-pressures or Tow flow _s, the uel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:
rovided that the water level in the vessel}s down' comer is maintained above the top of the' active fugl, natural circulation is suffic'ient to ensure a mi imum bundle flow fBrsall fuel assemblies t t have a relatively'hiph' power and potentially ca pproach a critical heat flux xondition.
For t ANE 9x9 fuel desig'n,\\ or the the g
minimum bundle flow s > 30 x,ld3 lb/hr. f ANF 8x8 fuel design, miM1 mum bundle flow is 3
> 28 "x 10 lb/hr.
For a'l esigns, the coolant f
minimu's bundle flow an such thak the mass f is a(sum flow area are \\
lways
>0.25x18'ilb/hr-Jt. Full s'cale critical power tests taken p ssures down to'14 7 psia indicate that e fuel assembly critical power at 0.25 x 10' I hr 'fta is a 3.35 Mwt. Atq5%RTP,a bundle p r ofi3.3'59hrt corresponds to a bundle radial aking factor of > 3.0, which is cantlyhighertha'ntheexpectedpea\\
signi king fact r.
Thus, a THERMAL POWER limit of 25% RTP for reactor pressures < 785 psig is conservative.)
A yjasm eu (continued)
BWR/6 STS B 2.0-3 Rev.
O, 09/28/92
)
Ercl 2 to GNRo-93,00109 Chapter : Page 22 INSERT B3A j
Since the pressure drop in the bypass region is i
essentially all elevation head, the core pressure drop at low power and flow will always be 5
> 4.5 psi.
Analyses show that with a bundle flow 3
l 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 4.5 psi 3
driving head will be > 28 x 10 lb/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly 1
critical power at this flow is approximately 3.35 MWt.
With the 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.
Because of the design j
thermal hydraulic compatibility of the reload fuel designs with the cycle 1 fuel, this justification and the associated low pressure and low flow limits remain applicable for future cycles of cores containing these fuel designs.
i i
l INSERT GRAND GULF B 2.0-3 10/1/93
i i
}
End 2 to GNRO.93m09 Reactor Core SLs Chapter 2 Page 2 B 2.1.1 i
BASES l
i i
APPLICABLE 2.1.1.2a MCPR (GE Fuel)
SAFETY ANALYSES (continued)
Th uel cladding integrity SL is set, such that n sign ficant fuel damage is calculated to occur if isnothiolated.
Since the parameters that res t in fuel he limit e
damage a not directly observable during rea or operation, 9
the therma and hydraulic conditions that r uit in the onset of tr sition boiling have been use o mark the baginning of e region in which fuel d ge could occur.
2 Although it is cognized that the ons of transition boiling would no result in damage t R fuel rods, the 1
]
critical power at ich boiling tra ition is calculated to occur has been ado ed as a conve ent limit. However, the 1 certainties in non toring the ore operating state and in a
the procedures used t calcul the critical power result in an uncertainty in th val of the critical power.
Therefore, the fuel clad integrity SL is defined as the 1
critical power ratio in t limiting fuel assembly for which more than 99.9% of the f el ds in the core are expected to
)
avoid boiling transiti
, con idering the power distribution within the core and I uncert nties.
The MCPR SL is de ruined using a tatistical model that combines all th uncertainties in erating parameters and the procedure used to calculate cri cal power.
The probability the occurrence of boili 1 transition is determined sing the approved General E tetric Critical Power cor lations.
Details of the fuel K adding integrity SL cale ation are given in Reference 2.
I ference 2 also inclu s a tabulation of the uncertainties u d in the
' det ination of the MCPR SL and of the nomina values of th parameters used in the MCPR SL statistical alysis.
m 2.1.
MCPR f~T N 11
The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an A00 from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition.
The i
margin between calculated boiling transition (i.e.,
MCPR = 1.00) and the MCPR SL is based on a detailed statistical procedure that considers the uncertainties in monitoring the core operating state.
One specific uncertainty included in the SL is the uncertainty inherent 1
(continued)
BWR/6 STS B 2.0-4 Rev.
O, 09/28/92
Enci 3 to GNRO-%0010 chapter a p.g. 24 Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 NCPR 4'"r P h - (continued) 94 SAFETY ANALYSES AoFB OP2 in the cri cal JRFwer correlation.
ReferenceI2.
describe he methodology used in detemining the
.R SL.
critic (al power correlation is based on a A A) F L 1
The significant body of practical test data, providing a high degree of assurance that the critical power, as evaluated by the correlation, is within a small percentage of the actual 1
critical power being estimated. As long as the core pressure and flow are within the range of validity of the Apps correlation, the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. Still further conservatism __
is induced by the tendency of the Qaycorrelation to (At#8]
h overpredict the number of rods in boiling transi on.
These conservatisms and the inherent accuracy of the correlation provide a reasonable degree of assurance that there would be no transition boiling in the core during sustained operation at the MCPR SL.
If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not be compromised. Significant test data 4
accumulated by the NRC and private organizations indicate that the use of a boiling transition imitation to protect against cladding failure is a very conservative approach.
Much of the data indicate that BWR fuel can-survive for an extended period of time in an environment of boiling transition.
2.1.1.3 Reactor Yessel Water Level During MODES 1 and 2, the reactor vessel water level is required to be above the top of the active fuel to prov.ide 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 Od water level becomes <
of the core height.
The reactor vessel water level SL has been established at the top of the an 420o-(continued)
~
BWR/6 STS B 2.0-5 Rev.
O, 09/28/92
4 Encl. 2 to GNRO-93/00109 Chap'sr 2 Paga 25 Reactor Core SLS g
BASES APPLICABLE 2.1.1.3 Reactor Vessei Water Level (continued)
SAFETY ANALYSES active irradiated fuel to provide a point that can be i
monitored and to also provide adequate margin for effective 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 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.
N i
APPLICABILITY SLs 2.1.1.1, 2.1.1._2,and 2.1.1.3 are applicable in all w
su t
o g
tolatedU SAFETY LIMIT L211 i
VIOLATIONS If any SL is violated, the NRC Operations Center mJst be notifie 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 (Ref.
').
4 4
i i
Exceeding an SL may cause fuel damage and create a potential for radioactive releases in cess of 10 CFR 100, " Reactor i
M Site Criteria," limits (Ref VO Therefore, it is required U
to insert all insertable con roTrods 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 also ensures that the probability of an accident occurring j
during this period is minimal.
1
-1 4
1 (continued)
BWR/6 STS B 2.0-6 Rev.
O, 09/28/92 i
1 m
4-
Nptw3 Reactor Core SLs P
-B 2.1.1 BASES SAFETY LIMIT W
(
g keview bmwiNee srb VIOLATIONS (continued)
IfanySLisviolated,theh;;r:grict: ::.ivi manau.
nt of) de nucle:r ;H:nt end the~eti'ityl shall 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 _isusediate action and assess the condition of the unit before reporting to the i
senior management.
)
b Gener&l Flaggn[Plaw) Opem+jons ad % '/he. Predde4 Opsmbs 661[)
3 W
If any SL is violated, a Licensee Event Report shall be preparedandsubmittedwithin30daystotheNRC{t.icaep*-
I*T
+
i accordance with 10'CFR 50.73_ (ReffG) to e4 aq]p/
Q, p o r t skall eds o be subm Aed e SR 2.2,5 p
If any SL is violated, restart of the unit shall not cosamence 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 nomal operation.
REFERENCES 1.
10 CFR 50, Appendix A, GDC 10.
-2.
NE0E-24011-F-A, ('etest appiu.ed. e,ii.un). E
'2. 4.
XN-NF524(A), Revision p, 90 ;2 r 1^Z.%
3g.
M "O 2 4I.
S'/.
BWR/6 STS B 2.0-7 Rev.
O, 09/28/92
l
(([ [ pYg Y.
i B 2.0 SAFETY LIMITS (SLs) i B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam done pressure protects thehRCS 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 i
preventing the release of fission products into the atmosphere.
Establishing an upper limit on reactor steam done pressure ensures continued RCS integrity. According to j
10 CFR 50, Appendix A, GDC 14, " Reactor Coolant Pressure Boundary," and GDC 15, " Reactor Coolant System Design" i
(Ref. 1), the reactor coolant pressure boundary (RCPB) shall i
be designed with sufficient margin to ensure that the design
{
conditions are not exceeded during normal operation and anticipated operational occurrences (A00s).I ring MO 1
"nd 2, reac vesse ater M vel is quired e
C bo e to _ f the ivajtfel to p ide cor ooling
<c 111
._ J Ouring normal operation and A00s, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2).
To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the cor Any further OCS hydrostatic testing with fuel in the core done under [ *d LCO 3.10.1, " Inservice Leak and Hydrostati Testing s
Operation." Following inception of unit operation, RCS Oct-componentsshallbepressuretesteginaccordancewiththe requirements of ASME Code, "ection II (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 products could enter the _
L-Niieb5n1dCFR1bbhREa
^
Site Criteria" (Ref. 4).
redu' N nader ok profeellve barrie deslped +o prevent re&saiHve en\\ eases & )
eneedQ &
(continued)
BWR/6 STS B 2.0-8 Rev.
O, 09/28/92
~..
i End. 2 to GNRO-9ND1m RCS Pressure SL i
Chapter 2 Page 28 B 2.1.2 BASES (continued)
APPLICABLE The RCS safety / relief valves and the Reactor Protection System Reactor Vessel Steam Dome Pressure-High Function j
SAFETY ANALYSES have settings established to ensure that the RCS pressure SL y
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, g
Section III,91971 Editionff including Addenda through the 41nter of 1972P(Ref. 5), which permits a maximum pressure transient of 1104, 1375 osig, of design pressure 1250 psig.
OM]
J The SL of 1325 psis, as measured iS the reactor steam domeg 8"
(press. e in::::t @ is equivalent to 1375 psig at the lowe'It elevation of tne Rcs.
The RCS is designed to ASME Code,Section III,1974 Edition (Ref. 6), for the reactor recirculation piping, which permits a maximum pressure transient of 110% of design pressures of 1250 psig for
@-] pressure SL is selected to be the lowest tra suction pipingg 1650 psig for discharge piping Th
')
g verpress_ure allowed by_the applicable codes.
J
~,
( Leben A pmp w k discWy vdgoA tops.'s beyed &e disAuq4@e, MST SAFETY LIMITS The maximum transient pressure allowahl in t e RCS pressure vessel under the ASME Code,Section III, is 110% of design The maximum transient pressure allowable in the pressure.
RCS piping, valves, and fittings is I of design pressures
,of1250psigforsuctionpipingbese p --
4EF psig for discharge piping ( The most limitin allowances is the 110% of design pressure; g of t therefore, t e SL on maximum allowable RCS pressure is established at psig a 3
i Aesuhm p Q 132s' @
APPLICABILITY SL 2.1.2 applies in all MODESj t:af, m
=5 :. :;ec[eu;M-0e inn;r OCA7 g;,;,,,,,,,;an;lT.30c;esereLel;;;.r; net'.11
- , ;,,,; ; ; ; ; j ;,u = ~ ;; g g _. = 1$;;,,
f--
-~
~
j SAFETY LIMIT W
4 5 ** '"
A e reach dem /g, ~
VIOLATIONS If any SL is violated, the NRC Operations Center must be notified within I hour, in accordance with 10 CFR 50.72 (Ref. 7).
1 (continued)
BWR/6 STS B 2.0-9 Rev.
O, 09/28/92
End 2 to GNRO-93/00109 gq3 ppgggg7, g(
Chepter 2 Pags 29 g
g BASES SAFETY LIMIT W
VIOLATIONS (continued)
Exceeding the RCS pressure SL may cause immediate 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 ensuresJ hat the operators take prompt remedialfictionyco%Jte s'+GL I
frc60kof ca accbF occurra 6 urb @is cer:o 2 B
)
g _
Ee ico Cemhe hR.b)
-h hMg If any SL is violated, thefQ:l-ut; eenie, r..... at ef] '-
p [24 hours.
tt.e nec;eet ui ni ne tr.: sti.:jrshall be notified within i
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 insediate action and assess the condition of the unit before reporting to the
_m._
_ senior management.
tCrenera r, Pfo4DpembsM Ae Vjee221,4, ofym 3 su s 9
h If any SL is violated, a Licensee Event Report shall be
{
prepared and submitted within 30 days to the NR fmanagement ef the =;;;;r ii;;;;, =d th: ;,;tmty "i--
Pn:' d;ntdecleer 0;;r;t' : _
i ::: M -t !iTn' accordance with 10 C'R 50.73 (Ref. 8).fA cop 3 o O(,1 Qpor t skall o\\ so. be.s ab mWed to Ac s R.c,
.s L.L.i 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 l
nomal operation.
REFERENCES 1.
10 CFR'50, Appendix A GDC 14, GDC 15, and GDC 28.
2.
ASME, Boiler and Pressure Vessel Code,Section III, h
@c:: re-7Q A
r (continued)
BWR/6 STS B 2.0-10 Rev.
O, 09/28/92
End. 2 to GNRO-932109 I
Chapter 3 Page M BASES REFERENCES 3.
ASME, Boiler and Pressure Vessel Code,Section XI, (continued)
Article IW 5000, 4.
5.
ASME, Boiler and Pressure Vessel Code, 1971 Edition Addenda,} winter of 1972}.
6.
ASME, Boiler and Pressure Vessel Code,$1974 Edition 7.
8.
1 i
l l
l BWR/6 STS B 2.0-11 Rev.
O, 09/28/92
-~-
End. 2 to GNRO-93/00109 Chapter 2 Page 31 ATTACHMENT 2B ITS - PSTS COMPARISON DOCUMENT DISCUSSION OF CHANGES
---_------_.,__.m.
,,,,,-_.._._,y,,,,,,,..,,,,,,,,,
_.m.,_,,
,.,g..-,,7,, _,. _,,,,,,...,,
Enct 2 m GNRo-93/00109 Chapter 2 Page 32 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 UFSAR Table 5.2-3, the recirculation system discharge piping design pressure is 1650 psig from the recirculation pump to the discharge valve and 1550 psig beyond the discharge valve.
Additionally, the piping with the most limiting design pressure, the suction piping, is identified.
P.2 The applicable critical power correlation for the current fuel cycle is "ANFB" rather than "XN-3".
This revision was identified and approved in Amendments 73 and 99 to the GGNS Operating License.
P.3 This comment number is not used for this station.
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 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 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 UFSAR).
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 controlled i
document with specific regulatory review requirements for changes, such as the Bases, UFSAR or OQAM.
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
Enct 2 to GNRO-93/00109 Chapter 2 Pegt 33 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 2 - SAFETY LIMITS PLANT SPECIFIC DIFFERENCE (continued) 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 I
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.
P.7 This comment number is not used for this station.
P.8 The current plant specific Bases are retained.
Complete applicability of the generic Bases could not be confirmed.
P.9 The reference is made more generic to include other applicable articles of ASME Section III.
P.10 This comment number is not used for this station.
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.
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 specific plant and utility management.
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 the SL and clarified.
C.5 The verb is revised to reflect that the use of the LCO is optional.
C.6 The terminology is revised to be consistent with the terminology used in TS Section 3.4 and in plant documents.
C.7 An editorial correction is made to provide for presentation that is consistent with the other requirements.
Erd 2 to GNRO-93/00109 Chapter 2 Paga 34 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 2 - SAFETY LIMITS CHANGE / IMPROVEMENT TO NUREG STS (continued)
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."
C.20 The ANF discussion of the 3.35 MWt is revised to match the GE discussion (use of "approximately").
Additionally, "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 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 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 (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.
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