ML20101H610
| ML20101H610 | |
| Person / Time | |
|---|---|
| Site: | Catawba, McGuire, Mcguire  |
| Issue date: | 03/15/1996 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20101H607 | List: |
| References | |
| NUDOCS 9604010011 | |
| Download: ML20101H610 (106) | |
Text
ATTACHMENT 2 Catawba Units 1 and 2 Marked-Up Technical specification Pages l
I l
9604010011 960315 PDR ADOCK 05000369 P PDR
TABLE 2.2-1 (Continued) .
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
((
E
" TRIP SETPOINT ALLOWABLE VALUE TUNCTIONAL UNIT i
EE 12. Steam Generator Water a
~
Level Low-Low h 10.~??, e4 nstrow %e 2. Jy, & na rr-ow mye 70^h
" 217% of span fry 5f*h 215.3% of span
- a. Unit 1 0% to 30% "
" 0% to 30% R 7 inearly increas inearly increas to .0% of span to . % of span 30% to 100% RTP* 30% to 100% RTP*
236.8% of narrow 235.1% of narrow
- b. Unit 2 range span range span
- 13. Undervoltage - Reactor 177% of bus voltage 276% (5016 volts) 7* Coolant Pumps (5082 volts) with a
"' O.7s. response time .
256.4 Hz with a 255.9 Hz
- 14. Underfrequency - Reactor Coolant Pumps 0.2s response time
- 15. Turbine Trip 2550 psig 2500 psig
!! a. Stop Valve EH EL EL Pressure Low RR 21% open 11% open EL El b. Turbine Stop Valve Closure ,
2z N.A. N.A.
PP 16. Safety Injection Input from ESF 9 E3 ,
CC D3 .
?% *RTP - RATED THERMAL POWER IS : ,
_ _ . . _ _ _ _ _ . _ _ _ _ . _ _ _ _ . _ . _ _ _ . _ . _ _ . . _ . _ _ _ _ _ _ _ m __.z - ___- _ - -
h TABLE 2.2-1 (Continued) g TABLE NOTATIONS (Continued)
NOTE 1: (Continued)
(595.1*p O u M I) a d ,
3; T' s '590.8'F (Nominal T y allowed by Safety Analysis);
y g(untf 2)
K3 = Overtemperature AT reactor trip depressurization setpoint penalty coefficient as presented in the Core Operating Limits Report; P = Pressurizer pressure', psig; P' = 2235 psig (Nominal RCS operating pressure);
i S = Laplace transform op'erator, s;
and f,(AI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:
[ (i) For q -q between the " positive" and " negative" f,(AI) breakpoints as presented in the Corebpera,tingLimitsReport; f (AI) = 0, where q and q are percent RATED THERMAL POWER in the top and bottom halves of t e core respectiv y, an q, + q, is total THERMAL POWER in percent of RATED THERMAL POWER; gg EE .- q, is mee negame dan ne f W) i (ii)
ForeachpercentAIthatthemagnitudeofq,OperatingLimitsReport,theATYripSetpoint
" negative" breakpoint presented in the Core na shall be automatically reduced by the f (AI) " negative" slope presented in the Core i
g:g: Operating Limits Report; and 6^ - q, is more positive than the f (iii) For
" positive" breakpoint presented in the Core q, Operating Limits Report, the each percent AI that the magnitude of ip AT lr(AI)
Setpoint gg shall be automatically reduced by the f 3(AI) " positive" slope presented in the Core
- 3. 3, Operating Limits Report.
en NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 4.5% of Rated Thermal Power.
\
n TABLE 2.2-1 (Continued) y TABLE NOTATIONS (Continued) c .
5 NOTE 3: (Continued) i c- K6 - Overpower AT reactor trip heatup setpoint penalty coefficient as presented in
- the Core Operating Limits Report for T > T" and K6 - O for T s T", l m
T - As defined in Note 1, n.
m T" - Indicated T a instrumentaffon,t s RATED THERMAL POWER (Calibration temperature for AT
- Fg S - As defined in Note 1, 585. f *F(udt t) %J and f2 (AI) is a function of the indicated differences between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
7 (i) for q, - q, between the " positive" and " negative" f 2(AI) breakpoints as presented in the Core Operating Limits Report; f (AI) 2 = 0, where q, and q, are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + q, is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent
" negative" breakpoint AI that presented the magnitude in the Core of q, Operating Limits Report, the2 AT Trip- q, is mo Setpoint shall be automatically reduced by the f z(AI)." negative" slope presented in the Core Operati'ng Limits Report; and for each percent AI that magnitude of q - q, is more positive than the f (AI) g (iii)
" positive" breakpoint presented in the bore I)perating Limits Report the $T Trip i mk ER Setpoint shall be automatically reduced by the f 2(AI) " Positive" slope presented in 88 the Core Operating. Limits Report.
EE kk NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.0% (Unit 1) and 3.3% (Unit 2) of Rated Thermal Power.
W 22 S. S.
se b
k -- _ _ - _ _ _ ___ _ _ _ _ _ _ _ _ _ _ __ _ __ _ _______ _ __
TABLE 3.3-4 (Continued) l
@ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ;
3D =
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE l
[ 4. Steam Line Isolation ,
z t
] a. Manual Initiation N.A. N.A.
~
- b. Automatic Actuation Logic N.A. N.A.
- and Actuation Relays
~ ~
- c. Containment Pressure-High-High s 3 psig s 3.2 psig
- d. Steam Line Pressure - Low 2 775 psig 2 744 psig
- e. Steam Line Pressure- s 100 psi s 122.8 psi ** ,
Negative Rate - High w 5. Feedwater Isolation ,
1 w a. Automatic Actuation Logic N.A. N.A.
4, Actuation Relays e
- b. Steam Generator Water ,
level-High-High (P-14) 8392 SS Gl.
- 1. Unit I s J2d% of s 34r2% of narrow ,
narrow range range instrument
>> instrument- span :
jj span SE 2. Unit 2 s 77.1% of s 78.9% of narrow
$$ narrow range range instrument
((' instrument span I o ,o span kh c. Tavg-Low 564*F 2 561*F r-
- d. Doghouse Water Level-High 11 inches 12 inches above 577' above 577'
= floor level floor level
- e. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
C~ _ _ _ _ _ _ _ - _ _ - - _ _ _ _ _ _ - - _ . _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ - - - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .- _ - - _ - _ _ _ _ _ _ _ _ _ _ - _ _ - - - - _ - _ _ _ - _ _ _ _ - - _ _ _ _
TABLE 3.3-4 (Continued) l ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ALLOWABLE VALUE h FUNCTIONAL UNIT IRIP SETPOINT C 6. Turbine Trip 2
- a. Manual Initiation N.A. N.A.
3
- b. Automatic Actuation N.A. N.A.
- Logic and Actuation Relays tQ
- c. Steam Generator Water Level-High-High (P-14) g
- 1. Unit I s .Sh-4% of s 34 @ of narrow arrow range range instrument instrument span span s* 2. Unit 2 s 77.1% of narrow range s 78.9% of narrow range instrument Y instrument span 5 span
- d. Trip of All Main N.A. N.A.
Feedwater Pumps
- e. Reactor Trip (P-4) N.A. N.A.
Safety injection See Item 1. above for all Safety Injection Setpoints and Allowable FF f.
Values.
88
- o. a
((
ee
- 7. Containment Pressure Control System 22 Start Permissive s 0.4 psid s 0.45 psid P.o a.
85
-~ b. Termination e 0.3 psid 2: 0.25 psid CC
- 3. 3. 8. Auxiliary Feedwater ne
- a. Manual Initiation N.A N.A.
ru -
Automatic Actuation Logic N.A. N.A.
and Actuation Relays S 3
___ _ __. _ _~_._x__
_ _ _ . _ _ - _ . __-..-u_____._-.u_-_..__----___w__
___2_m.______-._ _m__.-.______ _- _ _ _ _ - -_.ui-v--M ---
TABLE 3.3-4 (Continued) n
$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E TRIP SETPOINT ALLOWABLE VALUE g FUNCTIONAL UNIT h 8. Auxiliary Feedwater (Continued) g [c.7 7, of nArr W N3e Syn 5
d c. Steam Generator Water. 2 % d mrrow r*A3e spa n
. Level - Low-Low
- 1) Unit 1 2 17% of s (215.3%o g .from 0% span fro 0% to a 30% RT ncreasing 30% RTP m incre ing line y to line ly to 2 3 .3% of span 2 .0% of frja30%to100%
s
)
n from 30%
100% RTP y
- 2) Unit 2 2 36.8% of 2 35.1% of narrow narrow range range instrument span span Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
5 d.
- e. Loss-of-Offsite Power 2 3500 y 2 3242 V l
- f. Trip of All Main Feedwater Pumps N.A. N.A.
k g. Auxiliary Feedwater Suction EE Pressure-Low 2 9.5 psig 2S l 1) CAPS 5220, 5221, 5222 2 10.5 psig 55 2) CAPS 5230, 5231, 5232 2 6.2 psig 2 5.2 psig zz
- a. Unit 1 2 6.2 psig 2 5.2 psig
?? b. Unit 2 2 6.0 psig 2 5.0 psig L ~"
""9. Containment Sump Recirculation ,
EE a. Automatic Actuation Logic N.A. N.A.
- 3. 3. and Actuation Relays
- b. Refueling Water Storage 2 177.15 inches 2 162.4 inches Tank Level-Low Coincident .
See item 1. above for all Safety Injection Setpoints and Allowable Values.
With Safety Iniection
BfACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) i
- 1) All nonplugged tubes that previously had detectable wall l penetrations (greater than 20%), ;
- 2) Tubes in those areas where experience has indicated potential !
problems, and l
- 3) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does j not permit the passage of the eddy current probe for a tube 1 inspection, this shall be recorded and an adjacent tube shall be selected and. subjected to a tube inspection.
- c. Fer Unit 1, ia ddition to the 3% := ale, :11 tubes for which th:
- lternste plugging criteri: S:: b :: previcutl-y-appHed hhall Sc in:pected in the tube:h::t regien.
l
@ The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a I
i partial tube inspection provided: I
- 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and
- 2) The inspections include those portions of the tubes where imperfections were previously found.
- e. For Unit 1, imp,leliIentation oJdhe interim steafn generator }ube/ tube support plate elevationJ lifgging limit fop 4ycle 9 requjr6s a 100% -l ;
bobbin p, robe inspection for all hot leg 4ube support plate ;
intersections and a l' cold leg inter,sfctions down the lowest cold '
legtfubesupportpl}atewithouterdiameterstress[corrosioncracki f
4 (00 SCC) indications. The determination of tube' support plate
/intersectionyIiaving OD SCC ijidications shalVbe based on t)e i
.performancy of at least 20 percent random sa'mpling of tubps inspected !
over theit full length, fn inspection usifigj the rotating pancake coil (RPC) probe is required in order tp show operabil,i'ty of tubes with, flaw like bobbinfd oil signal amplitudes greater than 1.0 volt but less than 2.7 vpits. The RPC re its are to b evaluated to l blish that thp' principal indicqtions can be q aracterized ag The results of each sample inspection shall be classified into one of the following three categories:
Cateaory Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
CATAWBA - UNITS 1 & 2 3/4 4-13 Amendment No. 130 (Unit 1)
Amendment No. 124 (Unit 2)
]
REACTOR COOLANT SYSTEM OO ObSA$e5 SURVEILLANCE REQUIREMENTS (Continued)
Category Inspection Results C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes _ inspected are degraded tubes.
C-3 More than 12% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
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l CATAWBA - UNITS 1 & 2 3/4 4-13a Amendment No.102 (Unit 1)
Amendment No. 96 (Unit 2)
qNer Steam 3ene rdo r REACTOR COOLANT SYSTEM [ re.p kce ment on W N /}
SURVEILLANCE RE0VIREMENTS (Conti )
4.4.5.3 Insoection Frecuenc es - The bove required inservice inspections of ,
steam generator tubes shall be performe at the following frequencies: '
xt least
- a. The first inser ice inspection shall be performed after(6 Effective Full Power M hs but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the preservice inspection, result in all inspection results l falling into the C-1 category or if two consecutive inspections ;
demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval l may be extended to a maximum of once per 40 months;
- b. If the results of the inservice inspection of a steam generator .
conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall ,
apply until the subsequent inspections satisfy the criteria of '
Specification 4.4.5.3a.; the interval may then be extended to a -
maximum of once per' 40 months; and
- c. Additional, unscheduled inservice inspections shall be performed on
( each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
- 1) Reactor-to-secondary tubes leak (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or
- 2) A seismic occurrence greater than the Operating Basis' Earthquake, or
- 3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or
- 4) A main steam line or feedwater line break.
i 1
CATAWBA - UNITS 1 & 2 3/4 4-14 Amendment No.111 (Unit 1)
Amendment No.105 (Unit 2)
J
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria
- a. As used in this specification:
- 1) Imperfection means an exception to the dimensions, finish or contour of a tube (cr s!::vej from that required by fabrication drawings or specifications. Eddy-current testing indications ,
below.20% of the nominal tubeLor s:::v;l wall thickness, if I detectable, may be considere,d as imperfections;
, 2) Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube (4e-s4eevy;
- 3) Degraded Tube means a tube (er :leeve containing imperfections greater than or equal to 20% of the nominal tubefor el se wall thickness caused by degradation;
- 4) % Degradation means the percentage of the tubeldr :1:evelwall thickness affected or. removed by degradation;
- 5) Defect means an imperfection of such severity that it exceeds the r:p:f r- limit. A tube Mr-::::v;3containing a defect is ive de ecpwes;',n3 9 [p[gg
- 6) Repa h Limit means the imperfection depth at or beyond which the i ng er rep: fred by tube shall be sleev-ing. removed from It-also-seans the service by plugg#th :t er beyond=
i perfection-d
-which-a-sleeved tube ch:1' be plugg 6 The M limit is equal to 40% of the nominal tube -er 510 ve wall thickness. Jer Ur.it 1,-
.this-definition-does-not-apply-te the region-of-the-tub :ebject-
.to-the-al te rn ate-tube-plugg i ng - c ri te r4a, If a tube i any defects /s sfeeved found in the ubedue belowto jegradation the s eve will in notthtf ne F* distance 6ssi-tate plugging.
/
The Babcock &' Wilco process descri ed in Topical Rep rt BAW-2.045(P)-A, Rev. Vwill be used fo - sleeving. /
l / /
' For Unit 1 als'o, this definiti n does not apply /for tubes experiencing ,
outer diamet'er stress corros,fon cracking confitmed by bobbin probe inspection'to be within the' thickness of th9ptube support plates.
See 4 3/5.4.a.13 for the,dilugging limit fof use within the thickness of the tube support plate. [
\
CATAWBA - UNITS 1 & 2 3/4 4-15 Amendment No.102 (Unit 1)
Amendment No. 96 (Unit 2) ,
i i
/
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
I
- 7) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-l rity in the event of an' Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c., above;
- 8) Tube Insnection means an inspection of the steam generator tube l from the point of entry (hot leg side) ccoletely around the U-l bend to the top support of the cold leg;
! [ForUnit1 or a tube 1 M ich the tu W support p te elev ion interi ugging (IPC imit has b 3en' applied, inspe on will nelude all hot leg in fsections all col eg .
l 1 ersections n to and in ding, at. t, the el of the ast track i ication for ich the in im plugg hg criteria (limitist be applied l 9) Preserv' ice Insnection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of l the tubing. This inspection shall be performed prior to initial
- POWER OPERATION'using the equipment and techniques expected to
! be used during subsequent inservice inspections.
ube Ro W Exoansion is hat portion of a tube which has been incre ed in diamet by a rolling 4rocess such tha o crevice
! exi s between th outside dia tier of the tube an he i t esheet.
11 F* Distance s the minimum ength of the rol,1 expanded porti l of the tu which cannot ntain any defec,ts in order to e e l the tub does not pull The F* dis nce is
- 1.60 i ches and is meas,o uredt offrom the the tubesheet.
bot' tom of the roll l
! expapsion transitior/or the top of th,e' tubesheet if t)ie bottom l . of the roll expans 6n is above the top of the tubes det.
I luded in this istance is a safe'y t factor of 3 us a 0.5 {
inch eddy curre vertical measu e, ment uncertain . i l ) Alternate tub oluaaino criteri does not require the tube to be removed from' service or repaired when the tube' degradation exceeds the' repair limit sofl ong as the degrjdation is in that portion ol' the tube from F* to the bottom of the tubesheet:
This de inition does not apply to tubes with degradation (i.e.,
l indica ons of cracking) 'in the F* distaa'ce. i i
I l CATAWBA - UNITS 1 & 2 3/4 4-15 a Amendment No.130 (Unit 1) l Amendment No.124 (Unit 2) l l
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued)
- 13) The Tube Suonort Plateitdim Pluaaina Criteria limit is used)
' continued service for disposition of a s m generator tube that is experiencing ter diameter init ed stress corrosion cracking confined thin the thickness the tube support plates. For app cation of the tube pport plate interim plugging criter a limit, the tube's isposition for contin d service will e based upon standa bobbin probe signal pli-
~
tude of fla like indications. he plant specific gui lines used for 1 inspections shal e consistent with th eddy curren idelines in Appen 'x A of WCAP-13854 as propriate to acco date the additiona information needed to valuate tube sup rt plate signals wi respect to the vol't e parameters as {
s cified in Specific on 4.4.5.2. .
- 1. A tube can re in in service if the gnal amplitude of a crack indic ion is less than or e al to 1.0 volts, re rd-less of th depth of tube wall p etration, if, as a sult, the pro ted end of cycle dis ibution of crack in cations is ver ed to result in tot primary to secondar leakage less an 17.5 gpm (include operational and ac ' dent leak-age
- 2. tube can remain in rvice with a bobbin oil signal amplitude greater th 1.0 volt but less an 2.7 volts provided a rotatin pancake coil (RPC) spection does not detect degradati .
- 3. Iidications o degradation with a aw type bobbin coil signal ampl} ude of equal to or eater than 2.7 volts will bepluggedforrepaired.
)
l l
CATAWBA - UNITS 1 & 2 3/4 4-16 Amendment No. 130 (Unit 1)
Amendment No. 124 (Unit 2)
REACTOR COOLANT SYSTEM SURVEILLANCE RE0UTREMENTS (Continued)
N.
- 4. If as a r uit of leakape due to a me anism other than OD SCC at e tube suppopt plate inter tion, or some other caus , an unschedu 'd mid-cycle i ection is pe ormed, the fo owing repair riteria apply nstead of 4.4 4.a.13.2.
bobbin volt e is within e ected limits indication j can remain i service. The xpected repair imits are determined rom the follo /ng equation:
}
U(v3 -vy)+vy
+ (.2) ( A t) w re:
V -
easured volta Vooc = voltage at eginning of cyc (80C)
At - time per'dd of operatio o unscheduled out e L = cycle ength (full o Mrating cycle leng where oper ingcycleisJ1etimebetweent scheduled ste m generator i spections)
V,t =
.5 volts for 3/4-inch tubes Certain tub s as identifi d in WCAP-13494, V.1, will be excluded f om applicati9 of th.e. Interim ugging Criteria Limit as it has een determi deform f lowing a pop' ped LOCA tulated that these
+ Etu es may collapse or Event. -
plu%[n3
- b. The steam generator shall be determined OPERABLE after comple ng the corresponding actions (plug er repair all tubes exceeding the .epair limit and all tubes containing through-wall cracks) required by Table 4.4-2. For Unit 1, tubes-with-defects-below-F* fall under the -
-aMercate-tube-plugg4ng-cr-iter-la-and-do-not-have te be plugged,.
4.4.5.5 Reports f(%ed
- a. Within15daysfollowingthecompletionofea(hinserviceinspection of steam generator tubes, the number of tubes ' red in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2;
~
CATAWBA - UNITS 1 & 2 3/4 4-16a Amendment No. 130 (Unit 1)
Amendment No. 124 (Unit 2)
.- _ _ _ _ _ _ _ - )
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued)
- b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
- 1) Number and extent of tubes inspected,
- 2) location and percent of wall-thickness penetration far each indication of an imperfection, and Ph pcl Identification of tubes rep.y
- 3) ..ed.
- c. rer Unit.2 1fesults of steam generator tube inspections, which fall into Category C-3, shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior.
to resumption of plant operation. This repor.t shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
or Unit 1, tpe results of inspections for all tubes for which the alternate tutie plugging cri ria has been applied shall be reported to the Nydear Regulatory ommission in acc6rdance with 10 CF 0.4, -
prior tc restart of th th6 inspection. Thi report shal f include: nit following/
Identificati of applicable o s, and
- 2) location nd size of the degradation.
/
- e. For impig entation of the vditage-based repairjcriteria to tube support / plate intersections, notify the NRC staff prior'to returni g the s 'eam generators tofservice should any 'f the following {
con tions arise:
. If the estimate leakage based o he actual measured nd-of-cycle voltage' distribution would ave exceeded the leak limit (for the postulated main steam line break utilizing' licensing basis assdmptions) during the previous operating 7 cycle.
- 2. If cir,cumferential crack }Ike indications are/ detected at the tube / support plate inter, sections.
- 3. If /
the indications are/ id entified that e tend beyond the donfines of the tube / support plate, i
- 4. If the calculated onditional burs probability exceeds 1 X 10-2 ,
notify the NRC 36d provide an assp'ssment of the safety significanceoftheoccurrence./
CATAWBA - UNITS 1 & 2 3/4 4-16b Amendment No. 130 (Unit 1)
Amendment No. 124 (Unit 2)
J
l l
Table 4.4-1
, n MINIMUM NUMBER OF STEAM GENERATORS TO BE 5 INSPECTED DURING INSERVICE INSPECTION 2,
N, 2
'Preservice Inspection No Yes E
% Four
- No. of Steam Generators per Unit Four s .
First Inservice Inspection <c__ ,, All , Two g
" Second & Subsequent Inservice Inspections Onel Onem
"" **"0*^** "'"'" "
TABLE NOTATIONS
- 1. The inservice inspection may be limited to dne steam generator on a rotating C#nif / )
soiedule encompassing 3 N % of the tubes ( here N is the number of steam w generators in the plant) if the results o the first or previous inspections 2; indicate that all steam generators are per orming in a like manner. Note e
that under some circumstances, the operatin conditions in one or more steam those in other steam generators.
?" generators may Under such circumstances the sample sequence s be found to be more severe th (ha I be modified to inspect the '
most severe conditions.
- 2. Each of the other two. steam generators not inspected uring the first inservice inspections shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described in 1 above. ,
Tabl ,4.4-2 STEAM GENERATOR IUBE IN5PECTION n
$ 15I SAMPLE INSPECil0N 2ND SAMPLE INSPECTION 3RD SANPIE INSPLt.f!ON g Sample Sire Result Action Req'tred u Result Action Required Result Action Required 3=
A minimum of $ Tubes per 5.G. C-1 None If. A. N.A.
M.A. N.A.
g C-2 Plugdefectivetubehandinspectaddi- C-1 None N.A. N.A.
tional 25 tubes in this S.G.
."-g M
C-2 Plugdefectivetubeh C-1 None and inspect additional p 45 tubes in this S. G. C-2 Plugdefectivetubehl o'
Perform action for y ,
C-3 C-3 result of first sample Perform action for C-3 C-3 result of first N.A. N.A.
sample C-3 Inspect all t in this S.G., plug All other S.G.s are defective tube and inpsect 25 tubes C-1 None N.A. N.A.
In each other S. G.
Some 5.G.s C-2 but no additional S.G. Pprform action for C-2 N.A. N. A.
are C-3 result of second sample Motification to NRC pursuant to $50.72 *
(b)(2) of 10 CFR Part 50 Additional 5.G. Is Inspect all tubes in each g C-3 5. G. plug defective
% tubes b Notification to N.A. N.A.
NRC pursuant to 550.72 3
t (b)(2) of 10 CFR Part 50 y 5=3N g Where N is the number of steam generators in the unit, and n is the number of steam generators inspected durIng an inspection
.uomt+defou.tubemi f an_onder_t%:ternatmu,,i,-. sten: e: =t s= : 3: m;;1 aa 33
- Q. O.
99 to a 33 rt rt ZZ oo ON
^^
O 33 M+
ft rt
_______.s_ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- -- - - - ~ ^-
l l
REACTOR COOLANT SYSTEM l f BASES RELIEF VALVES (Continued) reactor coolant system pressure except for limited periods where the PORV has been isolated due to excessive seat leakage and except for limited periods where the PORV and/or block valve is closed because of testing and is fully i capable of being returned to its normal alignment at any time, provided that this I evolution is covered by an approved procedure. This is a function that reduces '
challenges to the code safety valves for overpressurization events. 5) Manual l control o*f a block valve to isolate a stuck-open PORV. Testing of the PORVs '
includes the emergency 2N supply from the Cold Leg Accumulators. This test demonstrates that the valves in the supply line operate satisfactorily and that the nonsafety portion of the instrument air system is not necessary for i proper PORV operation.
l 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the Reactor Coolant System will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to main-tain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manu-facturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
hheB&Wprocess or method equivylent) to the i shection method described in i Topical Repo BAW-2045(P)-Af ev.
R 1, will used. InserviceinJp[ctionof I ensureRCSintegfriy'. Because the steam gepe sleevys' atorchanjes introduce sleevesiniythealso required)ickness wall,th and diamet3r, they reduce th sepsitivity of eddy,is described it Topical Report BAWy2045(P)-A, Rev. L with used. A methpd s /
f(upportingvalidationdatathatdemonstratestheipspectabilityofthe' sleeve andunderlyJng, tube. As regui' red by NRC for lic repairprpcess,Catawbac9mmitstovalidatetpe',s6seesauthoriz/edtousethis adequacy of any system that is used fyr' periodic inseryice inspections of,the sleeves, andpill evaluate and, as emed apprcpriate by Duke Power Comp y, implement teft'ing methods as better -
(me ods are develope and validated for commercial use. ) i The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely
[ result in stress corrosion cracking. The extent of cracking during plant opera-l tion would be limited by the limitation of steam generator tube leakage between l the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 150 gallons per day per steam generator). Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 150 gallons per day per steam generator can readily be detected. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and repaired.
CATAWBA - UNITS 1 & 2 B 3/4 4-3 . Amendment No.102 (Unit 1)
Amendment No. 96 (Unit 2)
BASES STEAM GENERATORS (Continued)
Wastage-type defects are unlikely with proper chemistry treatment of the However, even if a defect should develop in service, it q\@i@ secondary coolant.
will be found during scheduled inservice steam generator tube examinations. p/yy0
' Depair- will be required for all tubes with imperfections exceeding the cepa4+
limit of 40% of the tube nominal wall thickness. F:r 'Jait 1, defa;tive th aich f:1' under the !!!eraate tebe p!e;;ie; critert: 6 20t b;; t: b r:p:fr;d. Defectiv: : tete generater tube: can be re;: fred by th: in:t:11;tia Of :10 ve: bhich span the area of d e radatien. and serve as a replac: :rt-pressure beendary for the degraded pertion of the tube, albing tktebe t:
rema4r b :erefte. Steam generator tube inspections of operating piants have -
demonstrated the capability to reliably detect wastage type degradation'that has penetrated 20% of the original tube wall thickness.
Tube: experiencing eeter di- eter stress corresten cracking withi- the l .thickae s ef the tube :eppert p!:te: tre plugged er r:;:f r:d by th: crit:ri: l
-of 4. 4. 5. 4. ; .13-.
Whenever the results of any steam generator tubing inservice inspection '
fall into Category C-3, these results will be reported to the Commission pur-suant to Specification 6.9.2 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, addi-tional eddy-current inspection, and revision of the Technical Specifications, l if necessary. If : tube 1: :!::ved due te degradation in the r* distaace.
4 hen-any-4+f+ct: in th: tube below-4he-s-leave will r-emaia ia service witkeut-
+ spa 4r-, .
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.
l These Detection Systems are consistent with the recommendations of Regulatory l Guide 1.45, " Reactor Coolant Pressure Boundary leakage Detection Systems," May
! 1973.
l l
l l
CATAWBA - UNITS 1 & 2 B 3/4 4-3a Amendment No. 111 (Unit 1)
Amendment No. 105 (Unit 2)
DESIGN FEATURES i
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment vessel is designed and shall be maintained for a maximum internal pressure of 15 psig and a temperature of 328'F.
5.3 REACTOR CORE FUEL ASSEMBLIES i 5.3.1 The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of cylindrical zircaloy clad fuel rods with an -initial composition of natural or slightly enriched uranium dioxide as fuel material.
Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod ;
configurations, may be used. Fuel assemblies shall be limited to those fuel i designs that have been analyzed with applicable NRC staff-approved codes and !
methods, and shown by tests or analyses to comply with all fuel safety design ;
bases. A limited number of Tead test assemblies that have not completed representative testing may be placed in non-limiting core regions. Reload '
fuel shall be similar in physical design to the. initial core loading and shall have a maximum nominal enrichment of 5.0 weight percent U-235 with a maximum
, tolerance of i .05 weight percent U-235.
I CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material of which 102 inches shall be 100% boron carbide and remaining 40-inch tip shall be 80% silver,15% indium, and 5% cadmium.
For Units 1 and 2, all control rods shall be clad with stainless steel tubing, except for Unit' 2, a maximum of one Rod Cluster Control Assembly may have Inconel clad control rods, i
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE i
5.4.1 The Reactor Coolant System is designed and shall be maintaine'd:
- a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, For a pressure of 2485 psig, and i b.
- c. For a temperature of 650*F, except for the pressurizer which is 680*F.
VOLUME 13,0501 100 edic fed for M I d 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,040 i 100 cubic feetg a nominal T, of 525'F.
- kr M + 1.
5.5 METEOROLOGICAL TOWER LOCATION .
5.5.1 The meteorological tower shall be located as shown in Figure 5.1-1.
CATAWBA - UNITS 1 AND 2 ' 5-6 Amendment No.135 (Unit 1) 2 Amendment No.129 (Unit 2)
ADMINISTRATIVE CONTROLS
- CORE OPERATING LIMITS REPORT (Continued)
, 5. DPC-NE-2011P-A, " Duke Power Company Nuclear Design Methodology for Core
- Operating Limits of Westinghouse Reactors," March, 1990 (DPC Proprietary).
,(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise
, Hot Channel Factor.)
- 6. DPC-NE-3001P-A, " Multidimensional Reactor. Transients and Safety Analysis 2
Physics Parameter Methodology," November 1991 (DPC Proprietary).
- (Methodology for Specification 3.1.1.3 - Moderator Temperature Coeffi-
. cient, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Chaniiel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
- 7. DPC-NF-2010A, " Duke Power. Company McGuire Nuclear Station Catawba Nuclear l Station Nuclear Physics Methodology for Reload Design," June 1985 (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, Specification 4.7.13.3 - Standby Makeup Pump Water Supply Boron Concentration, and Specification 3.9.1 - RCS and Refueling Canal Boron Concentration, and Specification 3.9.12 - Spent Fuel Pool Boron Concentration.)
, Ret. l,
- November-199&. SER dde ( becember ,1995 l (Methodology used in the system thermal-hydraulic analyses which i
determine the core operating limits) 1 (R ev. I,
- 9. DPC-NE-3000PA ,*" Thermal-Hydraulic Transient Analysis Methodology,"
Augus-t4994. SFA ade.d becembe.r ,199 5 (Modeling used in the system thermal-hydraulic analyses) i i
i i
i i
CATAWBA - UNITS 1 & 2 6-19a Amendment No.138 (Unit 1)
Amendment No. 132 (Unit 2)
I 1
ATTACIDENT 3 Catawba Units 1 and 2 Typed Technical Specification Pages I;
TABLE 2.2-1 (Continued) ;
b! REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E '
gg FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE
- 12. Steam Generator Water 5
D; Level Low-Low
- a. Unit 1 2: 10.7% of narrow 2: 9% of narrow n3 range span range span
- b. Unit 2 2: 36.8% of narrow 2: 35.1% of narrow range span range span
- 13. Undervoltage - Reactor 2: 77% of bus voltage 2: 76% (5016 volts)
Coolant Pumps (5082 volts) with a
~
0.7s response time K2 14. Underfrequency - Reactor at 56.4 Hz with a 2: 55.9 Hz Ei Coolant Pumps 0.2s response time ;
- 15. Turbine Trip
- a. Stop Valve EH 2: 550 psig 2: 500 psig Pressure Low NN
@@ b. Turbine Stop Valve Closure 2: 1% open 2: 1% open SS
- 16. Safety Injection Input from ESF N.A. N.A.
"["[
E E.
E. 2.
oe
TABLE 2.2-1 (Continued)
TABLE NOTATIONS n
2,
(
to NOTE 1: (Continued)
T' s 585.1*F (Unit 1) and 590.8*F (Unit 2) (Nominal T,g allowed by Safety Analysis); l E K3 = Overtemperature AT reactor trip depressurization setpoint penalty coefficient as 3 presented in the Core Operating Limits Report; P = Pressurizer pressure, psig;
[
" P' = 2235 psig (Nominal RCS operating pressure);
S = Laplace transform operator, s-1; and f3 (AI) is a function of the indicated difference between top and bottom detectors of the '
power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:
(i) For qt - 9b between the " positive" and " negative" f3 (AI) breakpoints as presented in the
'? Core Operating Limits Report; e
f, (AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + gb is total THERMAL POWER in perceit of RATED THERMAL POWER; pg (ii) For each ' percent AI that the magnitude of q, - gb is more negative than the f, (AI)
(( " negative" breakpoint presented in the Core Operating Limiss Report, the AT Trip gg Setpoint shall be automatically reduced by the fi (AI) " negative" slope presented in the A5 Core Operating Limits Report; and 22
.P (iii) For each percent AI that the magnitude of qt - 9b is more positive than the fi (AI)
" positive" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the f3 (AI) " positive" slope presented in the EE Core Operating Limits Report.
- 5. 5.
NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more SO than 4.5% of Rated Thermal Power.
t m_. --_ _ _ _ _ _ _ _ _ - _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ . _ . _ . _ _ _ _ _ _ . _
? .
l TABLE 2.2-1 (Continued)-
0 NOTE 3: (Continued) l E y K, = Overpower AT reactor trip heatup setpoint penalty coefficient as presented in the Core Operating Limits Report for T > T" and K, = 0 for T s T",
i'i a T = As defined in Note 1,
' " T" = Indicated Tyg at RATED THERMAL POWER (Calibration temperature for AT
" instrumentation, s 585.1*F (Unit 1) and 590.8"F (Unit 2)), l S = As defined in Note 1, and f2 (AI) is a function of the indicated differences between top and bottom detectors i
of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for gt - 9b between the " positive" and " negative" f2 (AI) breakpoints as presented in the Core Operating Limits Report; f2 (AI) = 0, where gt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER; t
pg (ii) for each percent AI that the magnitude of qt - 9b is more negative than the /2 (AI)
" negative" breakpoint presented in the Core Operating Limits Report, the AT Trip
((
gg Setpoint shall be automatically reduced by the f2 (AI) " negative" slope presented in EE the Core Operating Limits Report; and (iii) for each percent AI that the magnitude of qt - 9b is more positive than the f2 (AI)
" positive" breakpoint presented in the Core Operating Limits Report the AT Trip Setpoint shall be automatically reduced by the /2 (AI) " positive" slope presented in
&& the Core Operating Limits Report.
m- NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more
~
than 3.0% (Unit 1) and 3.3% (Unit 2) of Rated Thermal Power.
TABLE 3.3-4 (Continued) n ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE 5 4. Steam Line Isolation b a. Manual Initiation N.A. N.A.
2 Z b. Automatic Actuation Logic N.A. N.A.
[ and Actuation Relays
- c. Containment Pressure-High-High s 3 psig s 3.2 psig
- d. Steam Line Pressure - Low 2 775 psig 2 744 psig
- e. Steam Line Pressure- s 100 psi s 122.8 psi **
Negative Rate - High
- 5. Feedwater Isolation
- a. Automatic Actuation Logic N.A. N.A.
w Actuation Relays a
w b. Steam Generator Water h Level-High-High (P-14)
- 1. Unit I s 83.9% of s 85.6% of narrow l narrow range range instrument instrument span span EE ,
gg 2. Unit 2 s 77.1% of s 78.9% of narrow
&g narrow range range instrument gg instrument span
- span EE *
- c. T,yg-Low 2 564*F 2 561*F
- d. Doghouse Water Level-High 11 inches 12 inches
~~ above 577' above 577'
[y, floor level floor level
(( e. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
e
TABLE 3.3-4 (Continued) -
n ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS h
r FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE 5 6. Turbine Trip b a. Manual Initiation N.A. N.A.
2 Z b. Automatic Actuation N.A. N.A.
[ logic and Actuation Relays a- c. Steam Generator Water N Level-High-High (P-14)
- 1. Unit 1 s 83.9% of s 85.6% of narrow l narrow range range instrument instrument span span
- 2. Unit 2 s 77.1% of s 78.9% of narrow w narrow range range instrument 2 instrument span w span b d. Trip of All Main N.A. N.A.
Feedwater Pumps
- e. Reactor Trip (P-4) N.A. N.A.
- f. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable gg Values.
oo
&& 7. Containment Pressure Control 2g System en
- a. Start Permissive s 0.4 psid s 0.45 psid
- b. Termination 2 0.3 psid 2 0.25 psid
- 8. Auxiliary Feedwater y a. Manual Initiation N.A N.A.
- b. Automatic Actuation Logic N.A. N.A.
M and Actuation Relays i
TABLE 3.3-4 (Continued) c3 ENGINEERED S/IETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS D
g FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE 5
- 8. Auxiliary Feedwater (Continued)
C 5 c. Steam Generator Water gj Level - Low-Low
(( 1) Unit 1 at 10.7% of 2: 9% of no narrow range narrow range span span
- 2) Unit 2 2: 36.8% of 2: 35.1% of narrow narrow range range instrument span span
- d. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
$b e. Loss-of-Offsite Power 2: 3500 V 2: 3242 V b$
- f. Trip of All Main Feedwater Pumps N.A. N.A.
- g. Auxiliary Feedwater Suction Pressure-Low - r l[j(
aa
- 1) CAPS 5220, 5221, 5222 2: 10.5 psig 2: 9.5 psig Et El 2) CAPS 5230, 5231, 5232 2: 6.2 psig 2: 5.2 psig
$$ a. Unit 1 2: 6.2 psig 2: 5.2 psig
[. ['y b. Unit 2 2: 6.0 psig 2: 5.0 psig
?? 9. Containment Sump Recirculation !
- a. Automatic Actuation Logic N.A. N.A.
and Actuation Relays EE
- 5. 5, b. Refueling Water Storage 2: 177.15 inches 2: 162.4 inches r* r* Tank Level-Low Coincident c3:: With Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
l REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) j
- 1) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
- 2) Tubes in those areas where experience has indicated potential problems, and
- 3) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be
, selected and subjected to a tube inspection.
- c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
- 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and i 2) The inspections include those portions of the tubes where l imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Cateaory Inspection Results l
l C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
l C-2 One or more tubes, but not more than 1% of the i
total tubes inspected are defective, or between 5%
and 10% of the total tubes inspected are degraded
- tubes.
l C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
CATAWBA - UNITS 1 & 2 3/4 4-13 Amendment No. (Unit 1)
Amendment No. (Unit 2)
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 4.4.5.3 Inspection Frecuencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
, a. The first inservice inspection (after steam generator replacement on Unit 1) shall be performed after at least 6 Effective Full Power Months but within 24 calendar months of initial criticality (after steam generator replacement on Unit 1). Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has ]
1 occurred, the inspection interval may be extended to a maximum of !
once per 40 months; j
- b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months The increase in inspection frequency shall apply until the subseque..nt inspections satisfy the criteria of i Specification 4.4.5.3a.; the interval may then be extended to a 1 maximum of once per 40 months; and
- c. Additional, unscheduled inservice inspections shall be performed on j each steam generator in accordance with the first sample inspection ~
specified in Table 4.4-2 during the shutdown subsequent to any of the ,
following conditions: 1
- 1) Reactor-to-secondary tubes leak (not including leaks originating i from tube-to-tube sheet welds) in excess of the limits of l Specification 3.4.6.2, or I
- 2) A seismic occurrence greater than the Operating Basis Earthquake, or
- 3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or
- 4) A main steam line or feedwater line break.
CATAWBA - UNITS 1 & 2 3/4 4-14 Amendment No. (Unit 1)
Amendment No. (Unit 2)
REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria
- a. As used in this specification:
- 1) Imperfection means an exception to the dimensions, finish or !
contour of a tube from that required by fabrication drawings or l specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be l considered as imperfections;
- 2)
Dearadation means a service-induced cracking,
wastage, wear or l
. general corrosion occurring on either inside or outside of a [
tube;
- 3) Deoraded Tube means a tube containing imperfections greater than l or equal to 20% of the nominal tube wall thickness caused by degradation;
- 4) % Dearadation means the percentage of the tube wall thickness affected or removed by degradation; l i
- 5) Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective; l i
- 6) Pluoqina Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging. The plugging limit is equal to 40% of the nominal tube wall thickness.
- 7) Unserviceable describes the condition of a tube if it leaks or '
contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c., above;
- 8) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg;
- 9) Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
CATAWBA - UNITS 1 & 2 3/4 4-15 Amendment No. (Unit 1)
Amendment No. (Unit 2)
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.
4.4.5.5 Reports l -a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; l
l b. The complete results of the steam generator tube inservice inspection !
- shall be submitted to the Commission in a Special Report pursuant to j Specification 6.9.2 within 12 months following the completion of the l inspection. This Special Report shall include:
, 1) Number and extent of tubes inspected,
! )
l 2) Location and percent of wall-thickness penetration for each '
! indication of an imperfection, and l l I
! 3) Identification of tubes plugged. l l c. Results of steam generator tube inspections, which fall into Category I i C-3, shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of ,
plant operation. This report shall provide a description of '
investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
i I
i I
CATAWBA - UNITS 1 & 2 3/4 4-16 Amendment No. (Unit 1) l Amendment No. (Unit 2)
4
! Table 4.4-1 O MINIMUM NUMBER OF STEAM GENERATORS TO BE i y INSPECTED DURING INSERVICE INSPECTION 5
[ Preservice Inspection No Yes No. of Steam Generators per Unit Four Four-
- First Inservice Inspection (after Steam All Two
- a. Generator Replacement on Unit 1)
" 2 Second & Subsequent Inservice Inspections One' One 1
TABLE NOTATIONS 1
The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number R*
of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing
- in a like manner. Note that under some circumstances, the operating C conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
gg 2 Each of the other two steam generators not inspected (after steam
$$ generator replacement on Unit 1) during the first inservice inspections Ei shall be inspected during the second and third inspections. The fourth E$ and subsequent inspections shall follow the instructions described in 1 l' above.
e 55 r+ r+
l
___.----a _ ----.-.----_---.--_.--_---_a --m a._.- - -,.._._. --u ---,,.-a.--,.---,,,,s.,- -. _ - _ _ - _ - - - _ _ - - , . _ - - - - _ _ . - - - - - - - - - _ - . - _ _ . - - - --,----- <
TABLE 4.4-2 n
E STEAM GENERATOR TUBE INSPECTION E
y 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Ation Required Result Action Required E Result Action Required 3 A minimum of C-1 None N/A t.n S Tubes per N/A N/A N/A S.G. C-2 Plug defective tubes C-1 None N/A N/A
" and inspect additional N 2S tubes in this S.G. C-2 Plug deflective tubes and C-1 None inspect additional 4S tubes in this S.G. C-2 Plug defective tubes C-3 Perform action for C-3 result of first sample C-3 Perform action for C-3 N/A l N/A result of first sample R* , C-3 Inspect all tubes in this All other None N/A N/A S.G., plug defective S.G.s are C-1
? tubes and inspect 2S 5 tubes in each other S.G. Some S.G.s l
Perform action for C-2 N/A N/A C-2 but no result of second sample Notification to NRC additional pursuant to S.G. are C-3 550.72(b)(2) of 10 CFR 20 >
Part 50. Additional Inspect all tubes in each 55 S.G.is C-3 N/A N/A 55 S.G. and plu0 defective 5$ tubes. Notification to NRC pursuant to 150.72 g 55 ,
(b)(2) of 10 CFR 50.
S = 3 (N/n)% Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspectio EE S. E.
c> n
REACTOR COOLANT SYSTEM BASES RELIEF VALVES (Continued) reactor coolant system pressure except for limited periods where the PORV has been isolated due to excessive seat leakage and except for limited periods where the PORV and/or block valve is closed because of testing and is fully capable of being returned to its normal alignment at any time, provided that this evolution is covered by an approved procedure. This is a function that reduces challenges to the code safety valves for overpressurization events.
- 5) Manual control of a block valve to isolate a stuck-open PORV. Testing of the PORVs includes the emergency N2 supply from the Cold Leg Accumulators.
This test demonstrates that the valves in the supply line operate satisfac-torily and that the nonsafety portion of the instrument air system is not necessary for proper PORV ope, ration.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the Reactor Coolant System will be maintained. The program for inservice inspection of steam gen-erator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to main-tain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manu-facturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
I The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant I chemistry is not maintained within these limits, localized corrosion may )
likely result in stress corrosion cracking. The extent of cracking during i plant operation would be limited by the limitation of steam generator tube ;
leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 150 gallons per day per steam generator).
Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads '
imposed during normal operation and by postulated accidents. Operating plants i have demonstrated that reactor-to-secondary leakage of 150 gallons per day per steam generator can readily be detected. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking i tubes will be located and repaired.
l l
CATAWBA - UNITS 1 & 2 B 3/4 4-3 Amendment No. (Unit 1)
Amendment No. (Unit 2)
REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued)
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. . However, even if a defect should develop in service, it
- will be found during scheduled inservice ~ steam generator tube examinations.
l Plugging will be required for all tubes with imperfections exceeding the l plugging limit of 40% of the tube nominal wall thickness. Steam generator
! tube inspections of operating plants have demonstrated the capability to l reliably detect wastage type degradation that has penetrated 20% of the
- original tube wall thickness. ,
l 1 l Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pur-suant to Specification 6.9.2 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, addi-tional eddy-current inspection, and revision of the Technical Specifications, ,
if necessary.
I 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE I'
3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.
l These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
1 i
CATAWBA - UNITS 1 & 2 B 3/4 4-3a Amendment No. (Unit 1)
Amendment No. (Unit 2)
DESIGN FEATURES l
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment vessel is designed and shall be maintained for j a maximum internal pressure of 15 psig and a temperature of 328'F.
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies. Each assembly shall i consist of a matrix of cylindrical zircaloy clad fuel rods with an initial l composition of natural or slightly enriched uranium dioxide as fuel material.
l Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed l
representative testing may be placed in non-limiting core regions. Reload fuel shall be similar in physical design to the initial' core loading and shall I
have a maximum nominal enrichment of 5.0 weight percent U-235 with a maximum tolerance of i .05 weight percent U-235.
CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material of which 102 inches shall be 100% boron carbide and remaining 40-inch tip shall be 80% silver,15% indium, and 5% cadmium.
For Units 1 and 2, all control rods shall be clad with stainless steel tubing, except for Unit 2, a maximum of one Rod Cluster Control Assembly may have Inconel clad control rods. ;
5.4 REACTOR COOLANT SYSTEM i DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
- a. In accordance with the Code requirements specified in Section 5.2 of !'
the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
- b. For a pressure of 2485 psig, and
- c. For a temperature of 650 F, except for the pressurizer which is 680*F.
VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 13,050
- 100 cubic feet for Unit 1 and 12,040 1 100 cubic feet for Unit 2 at a l nominal T,yg of 525 F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown in Figure 5.1-1.
CATAWBA - UNITS 1 & 2 5-6 Amendment No. (Unit 1)
Amendment No. (Unit 2)
l l
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT.(Continued) l 1
- 5. DPC-NE-2011P-A, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March, 1990 (DPC Proprietary) .
l (Methodology for Specifications 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.1.3.5 - Shutdown Rod Insertion Limits, l 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, l 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
i
- 6. DPC-NE-3001P-A, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," November 1991 (DPC Proprietary).
l (Methodology for Specification 3.1.1.3 - Moderator Temperature Coeffi-cient, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank l Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot '
Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.) j l
- 7. DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear l Station Nuclear Physics Methodology for Reload Design," June 1985 (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, Specification 4.7.13.3 - Standby Makeup Pump Water Supply Boron Concentration, and Specification 3.9.1 - RCS and Refueling Canal Boron Concentration, and Specification 3.9.12 - Spent Fuel Pool Boron !
Concentration.)
Methodology," SER dated December, 1995. )
(Methodology used in the system thermal-hydraulic analyres A ich determinethecoreoperatinglimits)
- 9. DPC-NE-3000P, Rev. 1, " Thermal-Hydraulic Transient Analysis Methodology,"
SER dated December, 1995.
(Modeling used in the system thermal-hydraulic analyses)
CATAWBA - UNITS 1 & 2 6-19a Amendment No. (Unit 1)
Amendment No. (Unit 2)
'1 ATTACHMENT 4 McGuire Unit 1 Marked-Up Technical Specification Pages 1
l l
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 2:.14 #+ SP* h 2- IS % *f stan
- 12. Steam Generator Water 2: 1247ofT.span from 0 t
- 2: 11% of span fro . f Level--Low-Low RATED THERMA , ncreasing RATED THE , increasing linea 40% of span at to 3 span at 100% of
, of RATED THERMAL POWER THERMAL POWER.
- 13. Undervoltage-Reactor : 5082 volts-each bus 2: 5016 volts-each bus Coolant Pumps
- 14. Underfrequency-Reactor 2: 56.4 Hz - each bus 2: 55.9 Hz - each bus Coolant Pumps
- 15. Turbine Trip
- a. Low Trip System Pressure 2: 45 psig at 42 psig
- b. Turbine Stop Valve Closure at 1% open 2: 1% open
- 16. Safety Injection Input N.A. N.A.
from ESF
- 17. Reactor Trip System Interlocks
- a. Intermediate Range Neutron Flux, P-6, 2: 1 x 10 40 amps 2: 6 x 10 41 amps Enable Block Source Range Reactor Trip
- b. Low Power Reactor Trips Block, P-7
- 1) P-10 Input 10% of RATED THERMAL POWER at 9%, s 11% of RATED THERMAL POWER
- 2) P-13 Input s 10% RTP Turbine s 11% RTP Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent McGUIRE - UNIT 1 2-5 Amendment No.
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)
NOTE 1: (Continued) to = Time constant utilized in the measured Tavo 189 C mPensator, as presented in the Core Operating L,imi.ts Report, 385.1 T' = 5884 F Reference Tavo at RATED THERMAL POWER, K3 Overtemperature AT reactor trip depressurization setpoint penalty coefficient as presented in
=
the Core Operating Limits Report, ;
P = Pressurizer pressure, psig, P' = 2235 psig (Nominal RCS operating pressure),
L S = Laplace transform operator, sec'1,
. [
and fi (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for qt -9b between the " positive" and " negative" fi (AI) breakpoi,nts as pre.sented in the Core Operating Limits Report; fi (AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER-(ii) for each percent imbalance that the magnitude of qt -9b is more negative than the f, (AI) " negative" ;
breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be
, automatically reduced by the f, (AI) " negative" slope presented in the Core Operating Limits Report; r I
and (iii) for each percent imbalance that the magnitude of qt - 9b is more positive than the f, (AI) " positive" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoir.t shall be automatically reduced by the f, (AI) " positive" slope presented in the Core Operating Limits Report.
McGUIRE - UNIT 1 2-8 Amendment No. [
t
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)
T = As defined in Note 1, 585.I T" =
s 58fML} Reference T,yg at RATED THERMAL POWER, S = As defined in Note 1, and
[2 (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for qt -9b between the " positive" and " negative" f2 (AI) breakpoints as presented in the Core Operating Limits Report; f2 (AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent imbalance that the magnitude of qt -9b is more negative than the f2 (AI)
" negative" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the [2 (AI) " negative" slope presented in the Core Operating i Limits Report; and (iii) for each percent imbalance that the magnitude of gt -9b is more positive than the [2 (AI)
" positive" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the [2 (AI) " positive" slope presented in the Core Operating Limits Report. ,
NOTE 3: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 4.4% of Rated Thermal Power.
NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.0% of Rated Thermal Power.
McGUIRE - UNIT 1 2-10 Amendment No.
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 4. Steam Line Isolation
- a. Manual Initiation N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A.
and Actuation Relays
- c. Containment Pressure--High-High s 2.9 psig s 3.0 psig
- d. Negative Steam Line s 100 psi with a s 120 psi with a Pressure Rate - High rate / lag function rate / lag function time constant time constant a 50 seconds 2 50 seconds
- e. Steam Line Pressure - Low 2 775 psig a 755 psig
- 5. Turbine Trip and Feedwater Isolation
- a. Automatic Actuation Logic N.A. N.A.
and Actuation Relays y 83.97. es. c,7
- b. Steam Generator Water level-- sMto,f arrow range s
narrow range High-High (P-14) ns_truffient span each steam strument span each steam generator generator
- c. Doghouse Water Level-High 12" 13" (FeedwaterIsolationOnly)
- 6. Containment Pressure Control System Start Permissive / Termination 0.3 5 SP/T 5 0.4 PSIG 0.25 5 SP/T 5 0.45 PSIG (SP/T)
McGUIRE - UNIT 1 3/4 3-29 Amendment No.
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- a. Manual Initiation N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A.
and Actuation Relays
- c. Steam Generator Water Level--Low-Low > II 7 2 #f .5P h 2 IS N' ** S f"U
- 1) Start Motor-Driven Pumps 2: 12% of span frop,Sato 2: 11% of span from 30% of RATE iAL POWER, 30% of RATED TH , POWER, increas' inearly to increasin arly to 24 4 of span at 100% 1 39 span at 100%
ATED THERMAL POWER. TED THERMAL POWER.
Elb.7% ofSp n 3- 15 7 of span
- 2) Start Turbine-Driven Pumps 12% or span from 2: 11%ofspanfromj,he 30% of RATED L POWER, 30% of RATED _ POWER, increasi inearly to increasi nearly to 3 40 . of span at 100% 23 , of span at 100%
TED THERMAL POWER. ATED THERMAL POWER.
- d. Auxiliary Feedwater 2: 2 psig 2: 1 psig Suction Pressure - Low (Suction Supply Automatic
. Realignment)
- e. Safety Injection - See Item 1. above for all Safety Injection Trip Setpoints Start Motor-Driven Pumps and Allowable Values McGUIRE - UNIT 1 3/4 3-30 Amendment No.
l
?
4 SURVEILLANCE REQUIREMENTS (Continued) i 1 1) All nonplugged tubes that previously had detectable wall i penetrations (greater than 20%),
- 2) Tubes in those areas where experience has indicated potential problems, and i i l i 3) A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall ;
be performed on each selected tube. If any selected tube does a not permit the passage of the eddy current probe for a tube i
- inspection, this shall be recorded and an adjacent tube shall i
be selected and subjected to a tube inspection. ;
- c. != additier to the 3t sa'ple, 21' F* tubes will be in;pceted.
! l c.f The tubes selected as the second and third samples (if required by j
- Table 4.4-2) during each inservice inspection may be subjected to a '
j partial tube inspection provided:
i
- 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with
- imperfections were previously found, and J
l
- 2) The inspections include those portions of the tubes where f
imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between I 5% and 10% of the total tubes inspected are i degraded tubes. l C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected I tubes are defective.
Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
McGUIRE - UNIT 1 3/4 4-13 Amendment No.
-. . _~ . . _ - - _
REACTOR COOLANT SYSTEM M r Y'eg l* c e.m e nt SURVEILLANCE REQUIREMENTS (Co tinued) 4.4.5.3 Inspection Frea encies - The a ve required inservice inspections of steam generator tubes hall be performed t the following frequencies:
a.t leas +
- a. The firs inservice inspection hall be performed afte d Effective Full wer Months but within 24 calendar months of initial critical-ity. Subsequent inservice inspections shall be performed at inter-vals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following ser-vice under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if
, two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months;
- b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of i Specification 4.4.5.3a; the interval may then be extended to a l maximum of once per 40 months; and l
- c. Additional, unscheduled inservice inspections shall be performed on l each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of 4 the following conditions:
- 1) Reactor-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2,
- 2) A seismic occurrence greater than the Operating Basis Earthquake,
- 3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, and
- 4) A main steam line or feedwater line break. ,
l l
McGUIRE - UNIT 1 3/4 4-14 Amendment No.
r i
l REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria
- a. As used in this specification:
- 1) Imperfection means an exception to the dimensions, finish or contour of a tube (^r cleg from that required by fabrication drawings or specifTcations. Eddy-current testing indications below20%ofthenominaltubegrsleev3e-all thickness, if detectable, may be considered as imperfections; l 2) Degradation means a service-induced cracking, wastage, wear or l general corrosion occurring on either inside or outside of a
! tube Qr leeve p i 3) DeoradedTubemeansatube6r!!ee@containingimperfections
~
greater than or equal to 20% of the nominal tubeq :le@ wall thickness caused by degradation;
- 4) % dearadation means the percentage of the tubeQr slec@ wall
- thickness affected or removed by degradation; i
1
- 5) Defect means an imperfection of such severity that it exceeds h limit. A tube @ !!eevp containing a defect is g -P kSS AS I l 6) 6 epairlLimit means the imperfection depth at or beyond which l the tube er :leeve shall be removed from service by plugging w-r4 paired by sleev'ag and is equal to 40% of the nominal tube ee sleeve wall thickness. This definition-does--not-apply te the area-ef the tubesheet region below the F* distance provided the tube is not--degraded (i.e., no indications of cracking) within tac F* d4+tance. If 3-tebe i 'leeved-due--to-degradatica in I
the F* distance, then- any defest4-in-the-t-ube-below the :lceve w4 LL-ramain in ser" ice wi-thout--r+pa4+.
l The-Babcock & Wilcox-process-(or-method)-equ-i4alent to the-method descr4 bed in Topical Repor-t-BAW-20454P-)-A uil' be used.
- 7) Unserviceable describes the condition of a tubeCer sieevlif it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c, above;
- 8) Tube Inspection means an inspection of the steam generator tube
- from the point of entry (hot leg side) completely around the U-bend to the top suppcrt of the cold leg; and i
McGUIRE - UNIT 1 3/4 4-15 Amendment No.
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- 9) Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed af ter the pro, r to field hydrestatic test and prier te initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
F
- 10) F* Distance is the distance into the tubesheet from th face of the tubesheet or the top of the Alasrd .
/ whichever is lower (further into utresheet) that has been conservatively chosen to nches.
- 11) F* TUBE i u e with degradation equal to or greater than 4 0'- ow the F* distance and not degraded (i.e., no indications of cracking) in the F* distance.
- b. The steam generator shall be determined OPERABLE after completing
, the corresponding actions (plug 6r rep:13alltubesexceedingthe hw"O g limi nd all tubes containing tnrough-wall cracks) required 4.4.5.5 Reports ;
l
- a. Within 15 days following the completion of each inservice inspection j of steam generator tubes, the number of tubes plugged in each steam l generator shall be reported to the Commission in a Special Report l pursuant to Specification 6.9.2;
- b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the
- l completion of the inspection. This Special Report shall include:
- 1) Number and extent of tubes inspected,
- 2) Location and percent of wall-thickness penetration for each indication of an imperfection, and
- 3) Identification of tubes plugged (fr repsired.
- c. ' The results of inspections of F* tubes shall be repotted_to-the-Commission in a report, prior to theptertTf'he unit following the inspection. This r hah include: y
- 1) IdentificatitmIF* tubes, and -
/
-)
2 ' focation and size of the degradation. .-
- y McGUIRE - UNIT 1 3/4 4-16 Amendment No.
TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection No Yes No. of Steam Generators per Unit Two Three Four Two Three Four ,
First Inservice Inspection After Stea m Gner.6e Refine-4f All One Two Two l l 2 3 Second & Subsequent Inservice Inspections One One One One TABLE NOTATION:
1 The inservice inspection may be limited to one si.eam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the !
first or previous inspections indicate that all steam generators are performing in a like manner.
Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
2 The other steam generator not inspected during the first inservice inspection sha 1 be inspected.
The third and subsequent inspections should folloy the instructions described in 1 above.
ecfiv r s'tco.m se.sen1ae rep (aceme,e 3 Each of the other two steam generators not inspected during the first inservice inspection}shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described in 1 above. ,
McGUIRE - UNIT 1 3/4 4-17 Amendment No.
_ _ ._ m _ _._ _ __ .. _ . _ _ _ _ . _ . _ . _ _ _ _ _ _ _
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
- a. No PRESSURE BOUNDARY LEAKAGE, l b. 1 gpm UNIDENTIFIED LEAKAGE, I
o.I7
- c. ggpm total primary-to-secondary leakage through all steam generators and.,500' gallons per day through any one steam generator, l 13 5
- d. ,10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,
- e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of i
2235 1 20 psig, and l
- f. 1 gpm leakage at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
- a. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- b. With any Reactor Coolant System leakage greater than any one of the i above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from i Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY l within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following i 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- c. With any Reactor Coolant System Pressure Isolation Valve leakage l greater than the above limit, isolate the high pressure portion of I the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
McGUIRE - UNIT 1 3/4 4-20 Amendment No.
l i
REACTOR COOLANT SYSTEM BASES 3/4.4.4 RELIEF VALVES (Continued) reactor coolant pressure boundary leakage. 3) Manual control of the block valve to unblock an isolated PORV to allow it to be used for manual control of RCS pressure and isolate a PORV with excessive leakage. 4) Automatic control of PORVs to control RCS pressure. This is a function that reduces challenges to the code safety valves for overpressurization events. 5) Manual control of a block valve to isolate a stuck-open PORV.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-l tained. The program for inservice inspection of steam generator tubes is bas'd e on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical l damage or progressive degradation due to design, manufacturing errors, or l
inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube dec radation so that corrective measures can be taken. [The B&W pro- _[
htfc ) e to the inspgetion method described in Topi
' cess Report(or me)2045(P)quivale B W- -A be used. I sirvice inspect
- ff of steam g or sleev is also re i ed to ensur CS integrity. ecause the si es intro-du changes in e wall thick ss and diamete they reduce t sensitivity of dy current ting, theref e, special insp ction methods ust be used. 3/
method is cribed in To cal Report BAW 045(P)-Awith pporting vali ion data tha emonstrates e inspectabil' of the sleev and underlyin ube.
As reg red by NRC f licensees au rized to use t s repair pro s, McGuire
. commits to validat theadequacyoftany system tha is used for riodic inser-l vife inspection of the sleeves, nd will evalua) and, as dee d appropriate by Duke Power, Company, impleme testing methojM as better m hods are developed arnf validated for mmercial use.
The plant is expected to be operated in a manner such that the secondar f3g coolant will be maintained within those chemistry limits found p esult negligible corrosion of the steam generator tubes. yf the M condar olant chemistry is not maintained within these limi trcalized cyo r-o ion may likely result in stress corrosion cracking.,Jh xtent of cracjdng during plant operation would be limited by the41mitation of steapfenerator tube leakage between the Reactor Cool t ystem and the Seconfary Coolant System (reactor-to-secondary leakage = 60 gallons per day pp(steam generator). Cracks having a reactor-to-secondary leakage less than/hls limit during operation will have an adequate margin of safety to withst nd the loads imposed during normal operation and by postulated acciden . Operating plants have demonstrated that reactor-to-secondary leakage of 00 gallons per day per steam generator can readily be detectedOby-ra TMon-numors-ob-st+am-oenerater MowdowQ Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it McGUIRE - UNIT 1 B 3/4 4-3 i
REACTOR COOLANT SYSTEM e BASES l -e-Q STEAM GENERATORS (Continued) will be found during scheduled inservice steam generator tube examinations. i u Repair will be required for all tubes with imperfections exceeding the ecpairfl4 3 9 limit of 40% of the tube nominal wall thickness. Installed sleever with imper = .
fections exceeding 40'< ef the sleeve nomina! wa!' thicknes: ""' be plugget i Defective steam-generator tubes can be repaired by the installatica of lecyc; l which span the area of degradatien, and serve a 2 replacement pre ee.
beundary for the degraded pertien of the tube, a!! ewing the tube te re:21- i-ser" ice. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage type degradation that has penetrated 20% of the original tube wall thickness, cer tube 'ith degradatie- belew the F1-d-i-stance, and-net degraded within the r* distance, repair is act required.
If a tube is sleeved due-to degradation in the r* distance, then any defect: in the-tube beh the sleeve will remai" in service "ithout repair.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commis-sion pursuant to 10 CFR Sections 50.72 and 50.73 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE l l
l 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS l
The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.
The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with l
the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.
McGUIRE - UNIT 1 8 3/4 4-4
)
BASES ;
OPERATIONAL LEAKAGE (continued) l The CONTROLLED LEAKAGE limitation restricts operation when the total flow d supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating g valve in the supply line fully open at a nominal RCS pressure of 2235 psig.
This limitation ensures that in the event of a LOCA, the Safety Injection flow will not be less than assumed in the accident analyses.
ud. W 155 pA le.aka3e iWt per generabr- o.z7 o total steam generator tube leakage limit of g gpm for all steam D generator net iceleted # rem the 9CS ensures that the dosage contribution from M he tube leakage will be limitea e small fraction of 10 CFR Part 100 dose guideline values in the event of either-a-steam generater tube rupture er stra.
g line break. The g gpm limit is consistent with the assumptions used in the f ,
l analysis of thes'e accidents. The gpd leakage limit per steam i i
generator ensures that steam enerat r tube integrity is maintained in the event of a main steam line rup ure o under LOCA conditions. f 5.,. tl pg,4p. I o*nd135 S e d b M s a.c e. /35 (cupter 15 +rans!cn+g I l PRESSURE B0UNDARY LEAKAGE of any magnitude is unacceptable since u may i be indicative of an impending gross failure of the pressure boundary. There-fore, the presence of any PRESSURE B0UNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reac-tor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life i of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued opera-tion within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the I
Steady State Limits.
The Surveillance Requirements provide adequate assurance that concentra-tions in excess of the limits will be detected in sufficient time to take corrective ACTION.
3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appro-priately small fraction of Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 10'gpm.
4 The values for the limits on specific activity represent limits based upon a parametric (0.27 McGUIRE - UNI- 1 B 3/4 4-5
l l
l DESIGN FEATURES l
l FUEL ASSEMBLIES Continued) to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
CONTROL R00 ASSEMBLIES l
l 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material for Unit 1 control rods shall be 80% silver, 15% indium, and 5% cadmium. All i control rods shall be clad with stainless steel tubing.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
- a. In accordance with the Code requirements specified in Section 5.2 of .
l the FSAR, with allowance for normal degradation pursuant to the )
l applicable Surveillance Requirements, j t
- b. For a pressure of 2485 psig, and
- c. For a temperature of 650*F, except for the pressurizer which is l 680*F.
1 VOLUME
/ 3,o S o i 5.4.2 The total water and steam volume of the Reactor Coolant System is 42,040- ,
100 cubic feet at a nominal T yg of 525 F. {
5.5 METEOROLOGICAL TOWER LOCATION l !
5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE CRITICALITY l
5.6.1 a. The spent fuel storage racks are designed and shall be maintained with:
- 1) k ,,5 0.95 if fully flooded with unborated water as d,escribed in Section 9.1 of the FSAR; and
- 2) A nominal 10.4" center to center distance between fuel
' assemblies placed in Region 1; and
- 3) A nominal 9.125" center to center distance between fuel asse.ablies placed in Pegion 2.
McGUIRE - UNIT 1 5-7 Amendment No.
, 1 l
l ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)
- 8. DPC-NE-3002/,I'iS$RNkapter15SystemTransientAnalysisMethodology,"
Novc;;ici 1991. DER clM eol .bec em be r (9 95 l (Methodology used in the s thecoreoperatingmits)ystemthermal-hydraulicanalyseswhichde l
- 9. DPC-NE-3000P4,bThermal-HydraulicTransientAnalysisMethodology,"Augud !
4994. SEA dahd heeder 1995 )
(Modeling used in the system thermal-hydraulic analyses)
- 10. DPC-NE-1004A, " Nuclear Design Methodology Using CASM0-3/ SIMULATE-3P,"
November, 1992.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
l 11. DPC-NE-2004P-A, " Duke Power Company McGuire and Catawba Nuclear Stations
- Core Thermal-Hydraulic Methodology using VIPRE-01," December 1991 (DPC l Proprietary) .
(Methodology for Specifications 2.2.1 - Reactor Tri tion Setpoints, 3.2.1 - Axial Flux Difference (AFD)p, System and 3.2.3Instrumenta-
- Nuclear Enthalpy Rise Hot Channel Factor FNi(X,Y).)
i 12. DPC-NE-2001P-A, Rev.1, " fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints.)
- 13. DPC-2005P-A, " Thermal Hydraulic Statistical Core Design Methodology,"
February 1995 (DPC Proprietary).
l (Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation i Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 - '
Nuclear Enthalpy Rise Hot Channel Factor) .
l 14. BAW-10162P-A, TAC 03 Fuel Pin Thermal Analysis Computer Code, B&W Fuel Company, November 1989.
(Methodology used for Specification 2.2.1 - Reactor Trip System Instru-mentationsetpoints).
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated February 1994.
(Used for Specification 2.2.1, Reactor Trip System Instrumentation Setpoints).
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident
- analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supple-ments thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
McGUIRE - UNIT 1 6-22 Amendment No.
ATTACHMENT 5 McGuire Unit 1 Typed Technical Specification Pages I
i I
l l
I
--- - _ __-._. ~ _ _ _ _ _ - - _ - - - _ _ - - - - - . - - - - -_ _ - - --__ _--- _ _.--. --__ - - - - - _ _ - - - - . . - - - . - . - - - - - - _ _ _ _ _ _ _ _ _ - - - _ - - . - . - - _ - - _ - _ _ - . _ _ _ _ _ _ _ - - . - _ - _
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS l
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 12. Steam Generator Water 2 16.7% of span 2 15% of span .
Level--Low-low t
- 13. Undervoltage-Reactor 2 5082 volts-each bus 2 5016 volts-each bus L Coolant Pumps
- 14. Underfrequency-Reactor 2 56.4 Hz - each bus 2 55.9 Hz - each bus .
Coolant Pumps l I
- 15. Turbine Trip
- a. Low Trip System Pressure 2 45 psig a 42 psig
- b. Turbine Stop Valve Closure 2 1% open 2 1% open I
- 16. Safety Injection Input N.A. N.A.
from ESF
- 17. Reactor Trip System Interlocks Intermediate Range !
a.
Neutron Flux, P-6, 21x 1040 amps 2 6 x 10~' amps ,
Enable Block Source Range Reactor Trip :
l i
i McGUIRE - UNIT 1 2-5 Amendment No.
m.-_. m.. _ . . _ _ __ _ . _m-_____. . _ . _ . - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ --_ _ ___-- _ _ _ _ _ _ _ _ _ _ _
- - - _ . - ..- ~.
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS ,
NOTATION (Continued)
NOTE 1: (Continued)
Te = Time constant utilized in the measured T3g lag compensator, as presented in the Core Operating Limits Report, T' = s 585.1*F Reference Tag at RATED THERMAL POWER, -
Ka = Overtemperature AT reactor trip depressurization setpoint penalty coefficient as presented in the Core Operating Limits Report, ,
P = Pressurizer pressure, psig, P' = 2235 psig (Nominal RCS operating pressure),
S = Laplace transform operator, secd, and f3 (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: ,
(i) for gt -9b between the " positive" and " negative" f3 (AI) breakpoints as presented in the Core Operating Limits Report; f3 (AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER ;
(ii) for each percent imbalance that the magnitude of gt - 9b is more negative than the f3 (AI) " negative" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be '
automatically reduced by the f3 (AI) " negative" slope presented in the Core Operating Limits Report; and (iii) for each percent imbalance that the magnitude of qt -9b is more positive than the f3 (AI) " positive" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the f3 (AI) " positive" slope presented in the Core Operating Limits Report.
McGUIRE - UNIT 1 2-8 Amendment No.
TABLE 2.2-1 (Continued)
REACT 00 TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)
T = As defined in Note 1, T" =
s 585.1 F Reference T,yg at RATED THERMAL POWER, S = As defined in Note 1, and fi (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion cha-hers; with gains to be selected based on measured instrument response during plant .
startup tests such that:
(i) for gt -9b between the " positive" and " negative" fh (AD breakpoints as presented in the Core Operating Limits Report; fi (AI) = 0, where. qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent imbalance that the magnitude of qt -9b is more negative than the fi (AI)
" negative" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the fi (AI) " negative" slope presented in the Core Operating Limits Report; and (iii) for each percent inbalance that the magnitude of qt -9b is more positive than the f( (AI)
" positive" breakpoint presented in the Core Operating' Limits Report, the AT Trip Setpoint shall be automatically reduced by the fi (AI) " positive" slope presented in' the Core Operating Limits Report.
NOTE 3: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 4.4% of Rated Thermal Power.
NOTE 4: The channel's maximum Trip Setpoint shall not exceed its competed Trip Setpoint by more than 3.0% of Rated .
Thermal Power. !
McGUIRE - UNIT 1 2-10 Amendment No.
.___-___-__-________--_____ - - _-________--_________-_________ - _-___ -__ _ ____ - --____ - _-____ - ___-_=-___
t TABLE 3.3-4 (Continued) i ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETP0 INT ALLOWABLE VALUES l t
- 4. Steam Line Isolation !
- a. Manual Initiation N.A. N.A.
- b. -Automatic Actuation Logic N.A. N.A. l
'- and Actuation Relays !
- c. Containment Pressure--High-High s 2.9 psig s 3.0 psig
- d. Negative Steam Line s 100 psi with a s 120 psi with a Pressure Rate - High rate / lag function rate / lag function i time constant time constant !
2: 50 seconds me50 seconds !
- e. Steam Line Pressure - Low 2: 775 psig 2: 755 psig '
i
- 5. Turbine Trip and Feedwater Isolation i I
- a. Automatic Actuation Logic N.A. N.A.
and Actuation Relays ;
- b. Steam Generator Water level-- s 83.9% of narrow range , s 85.6% of narrow range High-High (P-14) instrument span each steam instrument span each steam l generator generator
- c. Doghouse Water Level-High 12" 13" (Feedwater Isolation Only)
- 6. Containment Pressure Control System Start Permissive / Termination 0.3 5 SP/T 5 0.4 PSIG 0.25 5 SP/T 5 0.45 PSIG (SP/T) i t
McGUIRE - UNIT 1 3/4 3-29 Amendment No.
- . . - - _ _ _ - - - . - - _ _ ---- - ,___ ~,- - u c- -a---a ~ - ---s-,-
- e
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- a. Manual Initiation N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A. ,
and Actuation Relays
- c. Steam Generator Water Level--Low-Low
- 1) Start Motor-Driven Pumps 2: 16.7% of-span 2: 15% of span
- 2) Start Turbine-Driven Pumps 2: 16.7% of span 2: 15% of span ;
- d. Auxiliary Feedwater 2: 2 psig 2: 1 psig Suction Pressure - Low (Suction Sup Realignment) ply Automatic .
- e. Safety Injection - See Item 1. above for all Safety Injection Trip Setpoints Start Motor-Driven Pumps and Allowable Values 1
McGUIRE - UNIT 1 3/4 3-30 Amendment No.
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- 1) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
- 2) Tubes in those areas where experience has indicated potential problems, and
- 3) A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
- c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection.may be subjected to a partial tube inspection provided:
- 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and !
- 2) The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the I following three categories:
Cateaory Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
- Note
- In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
i l
i l
McGUIRE - UNIT 1 3/4 4-13 Amendment No.
l
.- - - . - . . - . . . . _ _ _ .- ~ . .
SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Freauencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
- a. The first inservice inspection after the steam generator replacement shall be performed after at least 6 Effective Full Power Months but within 24 calendar months of initial criticality after the steam generator replacement. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; i
- b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall !
in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a; the interval may then be extended to a maximum of once per 40 months; and .
1
- c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of ,
the following conditions: 1 l
- 1) Reactor-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2,
- 2) A seismic occurrence greater than the Operating Basis Earthquake,
- 3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, and
- 4) A main steam line or feedwater line break.
i l
l l l 1
McGUIRE - UNIT 1 3/4 4-14 Amendment No.
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) ,
l l
4.4.5.4 Acceptance Criteria I
- a. As used in this specification:
- 1) Imperfection means an exception to the dimensions, finish or i' contour of a tube from that required by fabrication drawings or l specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be consid- l ered as imperfections;
- 2)
Dearadation means a service-induced cracking,
wastag.e, wear or general corrosion occurring on either inside or outside of a tube; l
- 3) Dearaded Tube means a tube containing imperfections greater than or equal to 20% of the nominal tube wall thickness caused by degradation;
- 4) % dearadation means the percentage of the tube wall thickness l affected or removed by degradation;
- 5) Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective; l
- 6) Pluaaina Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging and is equal to 40% of the nominal tube wall thickness.
- 7) Unserviceable describes the condition of a tube if it leaks or l contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c, above;
- 8) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and 1
1 i
l McGUIRE - UNIT 1 3/4 4-15 Amendment No.
l l
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- 9) Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition ;
of the tubing. This inspection shall be performed )rior to !
initial POWER OPERATION using the equipment and tec1niques expected to be used during subsequent inservice inspections.
- b. The steam generator shall be determined OPERABLE after completing ,
the corresponding actions (plug all tubes exceeding the plugging l l limit and all tubes containing through-wall cracks) required by ,
Table 4.4-2. {
l 4.4.5.5 Reports ;
- a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2;
- b. The complete results'of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the corapletion of the inspection. This Special Report shall include:
- 1) Number and extent of tubes inspected,
- 2) location and percent of wall-thickness penetration for each indication of an imperfection, and
- 3) Identification of tubes plugged.
McGUIRE - UNIT 1 3/4 4-16 Amendment No.
i TABLE 4.4-1 -
MINIMUM NUMBER OF STEAM GENERATORS TO BE. INSPECTED DURING INSERVICE INSPECTION Preservice Inspection No Yes ,
No. of Steam Generators per Unit Two Three Four Two Three Four First Inservice Inspection After Steam Generator All One Two Two Replacement l l 2 3 Second & Subsequent Inservice Inspections One One One One i
TABLE NOTATION: ,
1 The inservice inspection may be limited to one steam generator on a rotating schedule encompassing .
3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner. !*
Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
2 The other steam generator not inspected during the first inservice inspection after steam generator replacement shall be inspected. The third and subsequent inspections should follow the '
instructions described in 1 above.
3 Each of the other two steam generators not inspected during the first inservice inspections after '
steam generator replacement shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described in 1 above. l I
i McGUIRE - UNIT 1 3/4 4-17 Amendment No. ,
I
_- _ _ - - _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ - _ . - - ._. . _ _ _ _ _ _ _ _ _ = _ . _ - _ _ _
o l
OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
1
- a. No PRESSURE B0UNDARY LEAKAGE,
- b. 1 gpm UNIDENTIFIED LEAKAGE, l
- c. 0.27 gpm total primary-to-secondary leakage through all steam generators and 135 gallons per day through any one steam generator,
- d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,
- e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235
- 20 psig, and l
- f. 1 gpm leakage at a Reactor Coolant System pressure of 2235
- 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
- a. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,
- b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
McGUIRE - UNIT 1 3/4 4-20 Amendment No.
l
BASES 3/4.4.4 RELIEF VALVES (Continued) reactor coolant pressure boundary leakage. 3) Manual control of the block valve to unblock an isolated PORV to allow it to be used for manual control of
% RCS pressure and isolate a PORV with excessive leakage. 4) Automatic control of PORVs to control RCS pressure. This is a function that reduces challenges to the code safety valves for overpressurization events. 5) Manual control of a block valve to isolate a stuck-open PORV.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1,83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 135 gallons per day per steam generator). Cracks having l a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 135 gallons per day per steam generator can readily be detected. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects m. nnlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it McGUIRE - UNIT 1 B 3/4 4-3
_ BASES STEAM GENERATORS (Continued) will be found during scheduled inservice steam generator tube examinations.
Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage type degradation that has penetrated 20% of the original tube wall thickness.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commis-sion pursuant to 10 CFR Sections 50.72 and 50.73 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These Detection Systems att consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.
The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.
I McGUIRE - UNIT 1 B 3/4 4-4
REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (continued)
The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating i valve in the supply line fully open at a nominal RCS pressure of 2235 psig. '
This limitation ensures that in the event of a LOCA, the Safety Injection flow will not be less than assumed in the accident analyses.
The total steam generator tube leakage limit of 0.27 gpm for all steam generators and the 135 gpd leakage limit per generator ensures that the dosage contribution from the tube leakage will be limited to the applicable fraction of 10 CFR Part 100 dose guideline values for all FSAR Chapter 15 transients. '
The 0.27 gpm and the 135 gpd limits are consistent with the assumptions used in l the analysis of these accidents. The 135 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.
PRESSURE B0UNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. There-fore, the presence of any PRESSURE B0UNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
3/4.4.7 CHEF!ISTRY i
The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reac-tor Coolant System leakage or failure due to stress corrosion. Maintaining the i chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of
.the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued opera-tion within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.
The Surveillance Requirements provide adequate assurance that concentra-tions in excess of the limits will be detected in sufficient time to take corrective ACTION.
3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appro-priately small fraction of Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 0.27 gpm. The values for j the limits on specific activity represent limits based upon a parametric McGUIRE - UNIT 1 B 3/4 4-5
DESIGN FEATURES FUEL ASSEMBLIES Continued) to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
CONTROL R00 ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod
, assemblies. The full-length control rod assemblies shall contain a nominal 142
! inches of absorber material. The nominal values of absorber material for l Unit 1 control rods shall be 80% silver,15% indium, and 5% cadmium. All control rods shall be clad with stainless steel tubing.
, 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and.shall .tue maintained:
- a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, l
- b. For a pressure of 2485 psig, and i i c. For a temperature of 650'F, except for the pressurizer which is 680 F.
VOLUME .
J 5.4.2 The total water and steam volume of the Reactor Coolant System is 13,050 1 100 cubic feet at a nominal T ava of 525 F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE CRITICALITY !
5.6.1 a. The spent fuel storage racks are designed and shall be maintained with:
- 1) k ,,5 0.95 if fully flooded with unborated water as d,escribed in Section 9.1 of the FSAR; and i
- 2) A nominal 10.4" center to center distance between fuel I assemblies placed in Region 1; and
- 3) A nominal 9.125" center to center distance between fuel assemblies placed in Region 2.
McGUIRE - UNIT 1 5-7 Amendment No.
1 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)
j Hethodology," SER dated December 1995.
! (Methodology used in the s the core operating limits)ystem thermal-hydraulic analyses which determine
- 9. DPC-NE-3000P, Rev. 1, " Thermal-Hydraulic Transient Analysis Methodology," i SER dated December 1995. l (Modeling used in the system thermal-hydraulic analyses)
- 10. DPC-NE-1004A, " Nuclear Design Methodology Using CASM0-3/ SIMULATE-3P," ,
November, 1992.
l (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
- 11. DPC-NE-2004P-A, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," December 1991 (DPC Proprietary) .
(Methodology for Specifications 2.2.1 - Reactor Tri tion Setpoints, 3.2.1 - Axial Flux Difference (AFD)p ,System Instrumenta-and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor FAH(X,Y).)
- 12. DPC-NE-2001P-A, Rev. 1, " Fuel Mechanical Reload Analysis Methodology for i Mark-BW fuel," October 1990'(DPC Proprietary). l (Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation l l Setpoints.)
l
- 13. DPC-2005P-A, " Thermal Hydraulic Statistical Core Design Methodology,"
February 1995 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 -
Nuclear Enthalpy Rise Hot Channel Factor).
(Methodology used for Specification 2.2.1 - Reactor Trip System Instru-mentation setpoints).
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated February 1994.
(Used for Specification 2.2.1, Reactor Trip System Instrumentation Setpoints).
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS i
limits, nuclear limits such as shutdown margin, and transient and accident
- analysis limits) of the safety analysis are met.
i The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supple-l ments thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
McGUIRE - UNIT 1 6-22 Amendment No.
ATTACHMENT 6 l
McGuire Unit 2 Marked-Up Technical Specification Pages
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 2IG7Aafspan 215 8 of SP* n
- 12. Steam Generator Water 2 12% of span fro 2 11% of span fro f Level--Low-Low RATED THER , increasing RATED THER , increasing line o 2 40% of span at to -
span at 100% of 4 of RATED THERMAL POWER THERMAL POWER.
- 13. Undervoltage-Reactor 2 5082 volts-each bus 2 5016 volts-each bus Coolant Pumps
- 14. Underfrequency-Reactor 2 56.4 Hz - each bus a 55.9 Hz - each bus Coolant Pumps
- 15. Turbine Trip
- a. Low Trip System Pressure 2 45 psig 2 42 psig
- b. Turbine Stop Valve Closure 2 1% open 2 1% open
- 16. Safety Injection Input N.A. N.A.
from ESF
- 17. Reactor Trip System Interlocks
- a. Intermediate Range Neutron Flux, P-6, 21 x 10-10 amps 2 6 x 10-11 amps Enable Block Source Range Reactor Trip McGUIRE - UNIT 2 2-5 Amendment No.
.I t
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS t NOTATION (Continued) :
NOTE 1: (Continued) to =
Time constant utilized in the measured T,yg lag compensator, as presented in the Core Operating ;
Limits- eport, !
SBS.l T' =
s 588 2*F eference T,yg at RATED THERMAL POWER, Ka = Over emperature AT reactor trip depressurization setpoint penalty coefficient as presented in the Core Operating Limits Report, j P = Pressurizer pressure, psig, :
i P' = 2235 psig-(Nominal RCS operating pressure),
S = 1.aplace transform operator, sec-1, and f, (AI) is a function of the indicated difference b'etween top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup '
tests such that:
(i) for qt - % etween b the " positive" and " negative" fi (AI) breakpoints as. presented in the Core ;
i Operating Limits Report; f, (AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent imbalance that the magnitude of qt - qb is more negative than the f, (AI) " negative" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the fi (AI) " negative" slope presented in the Core Operating Limits Report;
. and (iii) for each percent imbalance that the magnitude of qt -9b is more positive than the fi (AI) " positive" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be i automatically reduced by the f, (AI) " positive" slope presented in the Core Operating Limits Report.
McGUIRE - UNIT 2 2-8 Amendment No.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____ __ _ _ _ _ _ _ _ _ _ _m, - .,- _ _ --- - - , . _ _ . . __ _ - _ _ . _ _ _ - ,_____- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)
T = As defined in Note 1, S85 13 T" = s 3Bfha* Reference T,yg at RATED THERMAL POWER, S = As defined in Note 1, and f2 (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for qt -9b between the " positive" and " negative" /2 (AI) breakpoints as presented in the Core _
Operating Limits Report; f2 (AI) = 0, where. qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent imbalance that the magnitude of qt - 9b is more negative than the f2 (AI)
" negative" breakpoint presented in the Core Operating Limits Report, the AT 1 rip Setpoint shall be automatically reduced by the f2 (AI) " negative" slope presented in the Core Operating Limits Report; and (iii) for each percent imbalance that the magnitude of gt - 9b is more positive than the [2 (AI)
" positive" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the f2 (AI) " Positive" slope presented in the Core Operating Limits Report.
NOTE 3: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 4.4% of Rated Thermal Power. .
NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.0% of Rated Thermal Power.
McGUIRE - UNIT 2 2-10 Amendment No.
TABLE 3.3-4 (Continued) i ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 4. Steam Line Isolation -
- a. Manual Initiation N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A.
and Actuation Relays
- c. Containment Pressure--High-High s 2.9 psig s 3.0 psig
- d. Negative Steam Line s 100 psi with a s 120 psi with a Pressure Rate - High rate / lag function rate / lag function time constant time constant 2: 50 seconds 2: 50 seconds
- e. Steam Line Pressure - Low 2: 775 psig 2 755 psig
- 5. Turbine Trip and Feedwater Isolation
- a. Automatic Actuation Logic N.A. N.A and Actuation Relays 854?o 8S. fof.
- b. Steam Generator Water level-- s kF4 o arrow range sit %'s of rrow range High-High (P-14) ment span each steam nt span each steam generator generator
- c. Doghouse Water Level-High 12" 13" (Feedwater Isolation Only)
- 6. Containment Pressure Control System Start Permissive / Termination 0.3 5 SP/T 5 0.4 PSIG 0.25 5 SP/T 5 0.45 PSIG (SP/T)
McGUIRE - UNIT 2 3/4 3-29 Amendment No.
i
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
& ONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 7. Auxiliary Feedwater l a. Mantal Initiation N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A.
and Actuation Relays
- c. Steam Generator Water Level--Low-Low 3 S 72 vfSPAn 2 / 6 N oY Sf Ah
- 1) Start Motor-Driven Pumps 2 12% of span from W 2 11% of span fron g 30% of RATED T _ POWER, 30% of RATED AL POWER, increasin early to increasi inearly to 2 40 f span at 100% 2 39 of span at 100%
TED THERMAL POWER. TED THERMAL POWER.
k H. 7 7 cd- Sfsn 2157. of 5pa.n
- 2) Start Turbine-Driven Pumps 12% of span from 30% of RATE W
- 4AL POWER, 2 11% of span from 30% of RATED L POWER, increas' inearly to increasi nearly to 2 4 of span at 100% 2 39 of span at 100%
RATED THERMAL POWER. TED THERMAL POWER.
- d. Auxiliary Feedwater 2 2 psig 2 1 psig Suction Pressure - Low (Suction Supply Automatic Realignment)
- e. Safety Injection - See Item 1. above for all Safety Injection Trip Setpoints Start Motor-Driven Pumps and Allowable Values McGUIRE - UNIT 2 3/4 3-30 Amendment No.
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) ,
l 1
- 1) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%)
- 2) Tubes in those areas where experience has indicated potential l problems, and
- 3) A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
- c. != addition te the 3t campic, all F* tube: .ill be i m pcet d.
(' % The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a l partial tube inspection provided:
- 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and l 2) The inspections include those portions of the tubes where l imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. l C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
l Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
McGUIRE - UNIT 2 3/4 4-13 Amendment No.
REACTOR COOLANT SYSTEM erNb"WSM" ;
cep 1 ce.m e Ct- l SURVEILLANCE RE0VIREMENTS (Co inued)
/
4.4.5.3 Inspection Fre ncies - The a ve required inservice inspections of steam generator tubes hall be performed t the following frequencies:
- a. The fir at Icast m inservice inspection hall be performed after"6 Effective Full wer Months but within 24 calendar months of initial critical-ity. Subsequent inservice inspections shall be performed at inter-vals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following ser-vice under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months;
- b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a; the interval may then be extended to a maximum of once per 40 months; and
- c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
, 1) Reactor-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the
~
limits of Specification 3.4.6.2,
- 2) A seismic occurrence greater than the Operating Basis Earthquake,
- 3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, and
- 4) A main steam line or feedwater line break.
I McGUIRE - UNIT 2 3/4 4-14 Amendment No.
3 REACTOR COOLANT SYSTEM l 1
l SURVEILLANCE REQUIREMENTS (Continued) !
4.4.5.4 Acceptance Criteria
- a. As used in this specification: ,
- 1) Imperfection means an exception to the dimensions, finish or contour of a tube /5r slee@from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tubecer-::ccve> wall thickness, if detectable, may be considered as imperfections;
- 2) Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tubeQr:1cevep
- 3) Dearaded Tube means a tubeGr slec3containing imperfections greater than or equal to 20% of the nominal tubeDeev3 wall thickness caused by degradation;
- 4) % dearadation means the percentage of the tube Qr :le M wall thickness affected or removed by degradation;
- 5) Defect means an imperfection of such severity that it exceeds the(sQepaw limit. A tubeCcr slec@ containing a defect is defective;\ ,
- 6) t ans e imperfection depth at or beyond which the IIibe er sleeve shall be removed from service by plugging er-r-spaired by cleev g and is equal to 40% of the nominal tube .ev.
i eAeeve wall thickness. .This def4-itica does not apply te the area-of-tAe-tubesheet regie" belev the r* distance previded the-tube is net degr-aded (i .e., ne indications of cracki g) "!thi-the-F*-distance . I' a tube i cleeved-due te degr-adatier in the-f*-di+tance, then any-defect in--the-tube-belew the :!ceve wi44-remain- in ser" ice without repair.
The Babcock a Wilcox process (or-method) equivalent te the
. method descrited in Topical Rapor-t-BAW-20d5(P)- ^ will be used.
- 7) Unserviceable describes the condition of a tubeCee-sieemp if it leaks or contains a defect large enough to affect its structural integrity.in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c, above;
- 8) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and McGUIRE - UNIT 2 3/4 4-15 Amendment No.
l REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- 9) Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed :fter the pec'or%
f4 eld hydre:t: tic to:t = d prier te initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
&) F*faceDistance 10 of i the distance into tubesheet or th tubesheet from t p of the last hard ,
op whi er is lower (fur into the tubeshee hat has been servatively chose o be 2 inches.
- 1) F* TUBE is a e with degradation al to or greater than 40%, belo e F* distance and n degraded (i .e., no indica 'ons of cracking) in th F* distance,
- b. The steam generator shall be determined OPERABLE after completing l the corresponding actions (plug 6r r= m all tubes exceeding the I
f % ,n$ limit and all tubes containing through-wall cracks) required le 4.4-2.
4.4.5.5 Reports
- a. Within 15 days fo11'owing the ccmpletion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam <
generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2;
- b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
- 1) Number and extent of tubes inspected, q
- 2) Location and percent of wall-thickness penetration for each indication of an imperfection, and
- 3) Identification of tubes plugged er repaired.
- c. The results of insifecIions of F* tubesfail be orted tS the Commiss nd n a report, prior to e ' restart of the unit'following the)npection. This report,sh 1 include: [ ,/p
/ /
/) 1 Identificationpf'F* tubes, and /
- 2) Location, arid size of the dpg adation McGUIRE - UNIT 2. 3/4 4-16 Amendment No.
TABLE 4.4-1 ,
MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection No Yes No. of Steam Generators per Unit Two Three Fotar Two Three Four First Inservice Inspection $ffer 51% 6uer.Ar ferixement All One Two Two l l 2 3 Second & Subsequent Inservice Inspections One One One One TABLE NOTATION:
1 The inservice inspection may be limited to one steam generator on a rotating schedule encompassing .
3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the !
first or previous inspections indicate that all steam generators are performing in a like manner.
Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
2 The other steam generator not inspected during the first inservice inspection all be inspected.
The third and subsequent inspections should follow the instructions described in 1 bove.
a % .c s fe w 9 as ea c .f. c tepla emc & -
3 Each of the other two steam generators not inspected during the first inservice inspectionsk'shall be inspected during the second and third inspections. The fourth and subsequent inspections'shall
- follow the instructions described in 1 above.
i McGUIRE - UNIT 2 3/4 4-17 Amendment No.
i REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
I
- a. No PRESSURE B0UNDARY LEAKAGE,
- b. 1 gpm UNIDENTIFIED LEAKAGE, 0 27 l c. #gpm total primary-to-secondary leakage through all steam generators l
and .500* gallons per day through any one steam generator, 13 5
- d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,
- e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 i 20 psig, and
- f. I gpm leakage at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in l
Table 3.4-1.
APPLICABILITY: MODES 1, 2, 3, arid 4.
! ACTION:
l a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within i
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage l
rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
1 l
l l
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McGUIRE - UNIT 2 3/4 4-20 Amendment No.
I
I REACTOR COOLANT SYSTEM BASES 3/4.4.4 RELIEF VALVES (Continued) l reactor coolant pressure boundary leakage. 3) Manual control of the block l valve to unblock an isolated PORV to allow it to be used for manual control of RCS pressure and isolate a PORV with excessive leakage. 4) Automatic control i of PORVs to control RCS pressure. This is a function that reduces challenges l to the code safety valves for overpressurization events. 5) Manual control of a block valve to isolate a stuck-open PORV.
l 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause
_of any tube degradation so that corrective measures can be taken. JT e B&W pro-l he inspect n metnoa aje crioea in pical r
e cess (or metnog(P)quivaient Report BAW- 045 -A will t used. Inse ce inspect on of stea generator sleeves 's also requir o ensure R i ntegri ty. ecause th sleeves int -
duc je- anges in the 1 thickness nd diameter hey reduc the sensitjvity of pdycurrenttes'g,therefore special ins o must be u 6d. A method is des bed in Topic Report BAW-2) (P)-A ionmethfpportin wit su alidatio data that onstrates the nspectabilit /of the sleeye)and under ing tube As requ' ed by NRC for 'censees autho zed to use ttfis repair rocess, uire com i s to validate adequacy of,a y system that'is used f r period' inser-v' e inspections o the sleeves, pfd will evalupt'e and, as eemed ap. opriate pany, impleme methods re kyDukePowerC eveloped and lidatedforcobt'testingmethdsasbette ercial use.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation.of-steam generator tube ~ eakfgi " I M between the Reactor Coolant ystem and the Secondary Coolan p ystem reactor-to-secondary leakage = lions per day per steam generator). Cracks having a reactor-to-secondary kage less than this an adequate margin of safety to withstankthe' loads imposedLimit during 1iuring operation normal will hav operation and by postulated accid t Operating plants have demonstrated that l reactor-to-secondary leakage of 50 gallons per day per steam generator can i by--radiation-moni. tor +-of-steam-generaur-b4wdoD9 Leakage
' readily in excess beofdetecte(limit this Wilhequire~prafit-shTitdown and an unscheduled l inspection, during which the leaking tubes will be located and plugged.
i i Wastage-type defects are unlikely with proper chemistry treatment of the l secondary coolant. However, even if a defect should develop in service, it
! McGUIRE - UNIT 2 8 3/4 4-3 l
l
1 1
l l REACTOR COOLANT SYSTEM BASES l STEAM GENERATORS (Continued)
I will be found during scheduled inservice steam generator tube examinations.
-Repa4s will be required for all tubes with imperfections exceeding the repak limit of 40% of the tube nominal wall thickness. Insta!!ed sleever with imper-fections exceeding 40% of the sleeve nominal wall thicknet: "4'1 be plugged.
Defective steam generater tubes can be repaired by the i stallatien of sleeve:- l l d ich span the area of degradation, and serve as a replacement precture:
l boundary for the degraded pertier of the tube, allow ng the tube te re=i 4-i l 4sov4ce, Steam generator tube inspections of operating plants have demonstrated l the capability to reliably detect wastage type degradation that has penetrated i 20% of the original tube wall thickness. r er tube "ith degradatier bele" the r* distance, and net degraded withi- the F* distance, repair i: net required.
i 44-a-tabe i: cleeved due te degradatier 4" the r* distance, then any defect: in the tube belew the sleeve ~i11 remai" 4" service "ithcut repair.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commis-l sion pursuant to 10 CFR Sections 50.72 and 50.73 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case l basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These Detection Systems are consistent with the reco:nmendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
l 3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.
l The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure
. isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of
( the allowed limit.
The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.
t McGUIRE - UNIT 2 B 3/4 4-4 i
l REACTOR COOLANT SYSTEM BASES l OPERATIONAL LEAKAGE (Continued) l l
@ The CONTROLLED LEAKAGE limitation restricts operation when the total flow i 4
5 supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating N: valve in the supply line fully open at a nominal RCS pressure of 2235 psig.
I h This limitation ensures that in the event of a LOCA, the Safety Injection flow I
d will npt be less than assumed in the accident analyses.
l v ond.the135 3p4 (c k Se t imit pec ener~fo c e o.27 D total steam generator tube leakage limit of X gpm for all steam
( generators .et ise!ated # rem the RCS ensures that the dosage contribution from the tuDe leakage will be limited toi smaM fraction of 10 CFR Part 100 dose guideline values 4n-the-event of either a steam gen uator tube rupture or--
steam H ne break The,ggpmlimiti+consistentwiththeassumptionsusedin o.27 the analysis of these accidents. The M G gpd leakage limit per steam generator ensures that steam ge rator ube integrity is maintained in th event of a main steam line rupt re or nder LOCA conditions. ,4ar all BM M A (55 y d li mt-f.s a.re. /35 (cha Pf er is franslen-ty PRESSURE BOUNDARY LEAKAGE of any magnitude i? unacceptable since it may be indicative of an impending gross failure of the pressure boundary. There-fore, the presence of any PRESSURE B0UNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTOOWN.
3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reac-tor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continted opera-tion within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the l
Steady State Limits.
The Surveillance Requirements provide adequate assurance that concentra-tions in excess of the limits will be detected in sufficient time to take corrective ACTION.
l 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appro-priately small fraction of Part 100 dose guideline values following a steam l generator tube rupture accident in conjunction with an assumed steady state
- primary-to-secondary steam generator leakage rate of p0'gpm. The values for l the limits on specific activity represent limits based upon a parametric
'O.27 McGUIRE - UNIT 2 B 3/4 4-5
DESIGN FEATURES FUEL ASSEMBLIES (Continued) to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
CONTROL-R00 ASSEMBLIES l 5.3.2 The core shall contain 53 full-length and no part-length control rod l assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material for Unit 2 control rods shall be 100% boron carbide (8 C) 4 for 102 inches and 80%
silver,15% indium, and 5% cadmium for the 40-inch tip. All control rods shall be clad with stainless steel tubing.
5.4 REACTOR CCIOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
- a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the
~
applicable Surveillance Requirements,
- b. For a pressure of 2485 psig, and
- c. For a temperature of 650*F, except for the pressurizer which is 680*F.
VOLUME
/ 3,0S 0 5.4.2 The total water and steam volume of the Reactor Coolant System is t&;640 1 100 cubic feet at a nominal T,yg of 525'F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE CRITICALITY 5.6.1 a. The spent fuel storage racks are designed and shall be ,
maintained with: j
- 1) k ,, 5 0.95 if fully flooded with unborated water as
' d,escribed in Section 9.1 of the FSAR; and ,
- 2) A nominal 10.4" center to center distance between fuel I assemblies placed in Region 1; and j
- 3) A nominal 9.125" center to center distance between fuel I assemblies placed in Region 2.
i 1
McGUIRE - UNIT 2 5-7 Amendment No.
1 l
ADMINISTRATIVE CONTROLS l
CORE OPERATING LIMITS REPORT
- 8. DPC-NE-3002/,kSkRC"hapter15SystemTransientAnalysisMethodology,"
l Novembee-B91. SER alde A becember R95 (Methodology used in the s the core operating limits)ystem thermal-hydraulic analyses which determine R e v. 6
- 9. DPC-NE-3000P.#,I" Thermal-Hydraulic Transient Analysis Methodology," August.
M94. SE R elo dget Deegg bgc (y)$
l (Modeling used in the system thermal-hydraulic analyses)
- 10. DPC-NE-1004A, " Nuclear Design Methodology Using CASM0-3/ SIMULATE-3P,"
November, 1992. I (Methodology for Specification 3.1.1.3 - Moderator Temperature .
Coefficient.)
- 11. DPC-NE-2004P-A, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," December 1991 (DPC Proprietary).
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumenta-tion Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3 . Nuclear Enthalpy Rise Hot Channel Factor FSi(X,Y).)
- 12. DPC-NE-2001P-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints.)
- 13. DPC-2005P-A, " Thermal Hydraulic Statistical Core Design Methodology,"
February 1995 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 -
l Nuclear Enthalpy Rise Hot Channel Factor).
l
(Methodology used for Specification 2.2.1 - Reactor Trip System Instru-mentation setpoints).
l
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated February 1994.
(Used for Specification 2.2.1, Reactor Trip System Instrumentation Setpoints).
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supple-ments thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
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McGUIRE - UNIT 2 6-22 Amendment No.
i
. -. . . ._. . . - - ._. . - ~ . . - ~ . .. . . . . - . . _ . _ _ . _ _ _ - . _ _ ._
ATTACHMENT 7 McGuire Unit 2 1
Typed Technical Specification Pages l
l l
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 12. Steam Generator Water 2 16.7% of span 2 15% of span Level--Low-Low
- 13. Undervoltage-Reactor 2 5082 volts-each bus a 5016 volts-each bus Coolant Pumps
- 14. Underfrequency-Reactor a 56.4 Hz - each bus a 55.9 Hz - each bus Coolant Pumps l 15. Turbine Trip
- a. Low Trip System Pressure 2 45 psig 2.42 psig
- b. Turbine Stop Valve Closure 2 1% open 2 1% open
- 16. Safety Injection Input N.A. N.A.
from ESF
- 17. Reactor Trip System Interlocks i
- a. Intermediate Range Neutron Flux, P-6, 21 x 10-10 amps a 6 x 10~" amps Enable' Block Source Range Reactor Trip L
McGUIRE - UNIT 2 2-5 Amendment No.
TABLE 2.2-1 (Continued) -
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)
NOTE 1: (Continued) te =
Time constant utilized in the measured T,yg lag compensator, as presented in the Core Operating Limits Report, T' =
s 585.1 F Reference T,yg at RATED THERMAL POWER, l K3 = Overtemperature AT reactor trip depressurization setpoint penalty coefficient as presented in the Core Operating Limits Report, P = Pressurizer pressure, psig, P' = 2235 psig (Nominal RCS operating pressure),
S = Laplace transform operator, sec,
and f3 (AI) is a function of the indicated difference b'etween top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for gt -9b between the " positive" and " negative" f, (AI) breakpoints as presented in the Core Operating Limits Report; f, (AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent imbalance that the magnitude of qt - 9b is more negative than the fi (AI) " negative" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the fi (AI) " negative" slope presented in the Core Operating Limits Report; and (iii) for each percent imbalance that the magnitude of qt -9b is more positive than the f, (AI) " positive" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the f3 (AI) " positive" slope presented in the Core Operating-Limits Report.
McGUIRE - UNIT 2 2-8 Amendment No.
J TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS ,
NOTATION (Continued)
T = As defined in Note 1, T" =
s 585.1 F Reference T ,g at RATED THERMAL POWER, l S = As defined in Note 1, and f2 (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: !
(i) for qt -9b between the " positive" and " negative" f2 (AI) breakpoints as presented in the Core Operating Limits Report; f2 (AI) = 0, where.qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of i
RATED THERMAL POWER; (ii) for each percent imbalance that the magnitude of gt -9b is more negative than the f2 (AI)
" negative" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the /2 (AI) " negative" slope presented in the Core Operating Limits Report; and (iii) for each percent imbalance that the magnitude of qt - 9b is more positive than the f2 (AI)
" positive" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the f2 (AI) " positive" slope presented in the Core Operating Limits Report.
NOTE 3: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 4.4% of Rated Thermal Power.
NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.0% of Rated Thermal Power.
McGUIRE - UNIT 2 2-10 Amendment No.
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
~
- 4. Steam Line Isolation
- a. Manual Initiation N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A.
and Actuation Relays
- c. Containment Pressure--High-High s 2.9 psig s 3.0 psig
- d. Negative Steam Line s 100 psi with a s 120 psi with a Pressure Rate - High rate / lag function rate / lag function time constant time' constant 2 50 seconds 2 50 seconds
- e. Steam Line Pressure - Low 2 775 psig 2 755 psig
- 5. Turbine Trip and Feedwater Isolation
- a. Automatic Actuation Logic N.A. N.A.
and Actuation Relays
- b. Steam Generator Water level-- s 83.9% of narrow range s 85.6% of narrow range l High-High (P-14) instrument span each steam instrument span each steam generator generator
- c. Doghouse Water Level-High 12" 13" (Feedwater Isolation Only)
- 6. Containment Pressure Control System Start Permissive / Termination 0.3 5 SP/T 5 0.4 PSIG 0.25 5 SP/T 5 0.45 PSIG (SP/T)
McGUIRE - UNIT 2 3/4 3-29 Amendment No.
t TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS j FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES l
- 7. Auxiliary Feedwater f
- a. Manual Initiation N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A. l and Actuation Relays j
- c. Steam Generator [
Water Level--Low-Low
- 1) Start Motor-Driven Pumps 2 16.7% of-span 2 15% of span [
t
- 2) Start Turbine-Driven Pumps 216.7% of span 2 15% of span
- d. Auxiliary Feedwater 2 2 psig 2 1 psig Suction Pressure - Low (Suction Supply Automatic Realignment) , l
- e. Safety Injection - See Item 1. above for all Safety Injection Trip Setpoints i Start Motor-Driven Pumps and Allowable Values l i
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McGUIRE - UNIT 2 3/4 3-30 Amendment No.
l 1
T REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- 1) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
- 2) Tubes in those areas where experience has indicated potential problems, and
- 3) A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
- c. Tile tubes selected as ' ' second and third samples (if required by Table 4.4-2) during eaco inservice inspection may be subjected to a partial tube inspection provided:
- 1) The tubes selected for these samples include the tubes from those creas of the tube sheet array where tubes with imperfections were previously found, and c 2) The inspections include those portions of the tubes where
- imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Cateoorv Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
McGUIRE - UNIT 2 3/4 4-13 Amendment No.
i I
REACTOR COOLANT SYSTEM l l SURVEILLANCE RE0VIREMENTS (Continued) 4.4.5.3 Inspection Frecuencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
- a. The first inservice inspection after the steam generator replacement shall be performed after at least 6 Effective Full Power Months but within 24 calendar months of initial criticality after the steam generator replacement. Subsequent inservice inspections shall be l performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive .
inspections following service under AVT conditions, not including l the preservice inspection, result in all inspection results falling !
into the C-1 category or if two consecutive inspections demonstrate that periously observed degradation has not continued and no additkn.,1 degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; i
l
- b. If the results of the inservice inspection of a steam generator i conducted in accordance with Table 4.4-2 at 40-month intervals fall
'in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency ;
shall apply until the subsequent inspections satisfy the criteria of I Specification 4.4.5.3a; the inte m ; may then be extended to a maximum of once per 40 months; ar.d l
- c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 durino the shutdown subsequent to any of the following conditiors:
- 1) Reactor-to-secondary tubes leaks (not including leaks i originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2,
- 2) A seismic occurrence greater than the Operating Basis Earthquake,
- 3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, and
- 4) A main steam line or feedwater line break.
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l McGUIRE - UNIT 2 3/4 4-14 Amendment. No.
f REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria
- a. As used in tnis specification:
- 1) Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or l specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be consid- l ered as imperfections;
- 2)
Dearadation means a service-induced cracking,
wastage, wear or general corrosion occurring on either inside or outside of a .
tube;
- 3) Dearaded Tube means a tube containing imperfections greater than or equal to 20% of the nominal tube wall thickness caused by degradation;
- 4) % dearadation means the percentage of the tube wall thickness l affected or rem'oved by degradation;
- 5) Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective; l
- 6) Pluccina Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging and is equal to 40% of the nominal tube wall thickness.
- 7) Unserviceable describes the condition of a tube if it leaks or l contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c, above;
- 8) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and 4
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1 McGUIRE - UNIT 2 3/4 4-15 Ae ndment No.
i
1 I !
l l REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) l ,
1
- 9) Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current ,
techniques prior to service to establish a baseline condition l of the tubing. This inspection shall be performed prior to 1 initial POWER OPERATION using the equipment and techniques l expected to be used during subsequent inservice inspections. l l
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceed:ng the plugging l limit and all tubes containing through-wall cracks) required by Table 4.4-2.
4.4.5.5 Reports
- a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; i
- b. The complete results'of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
l 1) Number and extent of tubes inspected, ,
- 2) Location and percent of wall-thickness penetration for each l
indication of an imperfection, and l
- 3) Identification of tubes plugged.
i l
4 i
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P 4
, McGUIRE - UNIT 2 3/4 4-16 Amendment No.
1 4
w .. .-- _
TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection No ,
Yes No. of Steam Generators per Unit Two Three sour Two Three Four First Inservice Inspection After Steam Generator All One Two Two Replacement l l 2 3 Second & Subsequent Inservice Inspections One One One One TABLE NOTATION:
P 1
The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner.
Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
2 The other steam generator not inspected during the first inservice inspection after steam generator replacement shall be inspected. The third and subsequent inspections should follow the instructions described in I above.
~
3 Each of the other two steam generators not inspected during the first inservice inspections after steam generator replacement shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described in 1 above.
McGUIRE - UNIT 2 3/4 4-17 Amendment No.
l LIMITING CONDITION FOR OPERATION
- 3.4.6.2 Reactor Coolant System leakage shall be limited to
l
- a. No PRESSURE BOUNDARY LEAKAGE,
- b. 1 gpm UNIDENTIFIED LEAKAGE,
- c. 0.27 gpm total primary-to-secondary leakage through all steam l generators and 135 gallons per day through any one steam generator, l d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,
- e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of
, 2235 1 20 psig, and i
- f. 1 gpm leakage at a Reactor Coolant System pressure of 2235 1 20 psig
! from any Reactor Coolant System Pressure Isolation Valve specified in l
Table 3.4-1.
l
! APPLICABILITY: MODES 1, 2, 3, and 4.
[ ACTION:
- a. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or bo in at least il0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by i
use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD
- SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
i i
4 McGUIRE - UNIT 2 3/4 4-20 Amendment No.
l l
l BASES 3/4.4.4 RELIEF VALVES (Continued) reactor coolant pressure boundary leakage. 3) Manual control of the block valve to unblock an isolated PORV to allow it to be used for manual control of RCS pressure and isolate a PORV with excessive leakage. 4) Automatic control of PORVs to control RCS pressure. This is a function that reduces challenges to the code safety valves for overpressurization events. 5) Manual control of j a block valve to isolate a stuck-open PORV.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main- I tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 135 gallons per day per steam generator). Cracks having l ;
a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 135 gallons per day per steam generator can readily be detected. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it i
l l
l l i i
l i
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McGUIRE - UNIT 2 B 3/4 4-3
(
REACTOR COOLANT SYSTEM BASES I
STEAM GENERATORS (Continued) l will be found during scheduled inservice steam generator tube examinations. !
Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably l detect wastage type degradation that has penetrated 20% of the original tube wall thickness.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commis-sion pursuant to 10 CFR Sections 50.72 and 50.73 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection ~ Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE Industry exper'ence has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.
The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.
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McGUIRE - UNIT 2 B 3/4 4-4 l
REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (continued)
The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig.
This limitation ensures that in the event of a LOCA, the' Safety Injection ficw will not be less than assumed in the accident analyses. l The total steam generator tube leakage limit of 0.27 gpm for all steam generators and the 135 gpd leakage limit per generator ensures that the dosage contribution from the tube leakage will be limited to the applicable fraction of 10 CFR Part 100 dose guideline values for all FSAR Chapter 15 transients.
The 0.27 gpm and the 135 gpd limits are consistent with the assumptions used i.n the analysis of these accidents. The 135 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.
PRESSURE B0UNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. There-fore, the presence of any PRESSURE B0UNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reac-tor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued opera-tion within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.
The Surveillance Requirements provide adequate assurance that concentra-tions in excess of the limits will be detected in sufficient time to take corrective ACTION.
3/4.4.8 SPECIFIC ACTIVITY 4
The limitations on the specific activity of the reactor coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appro-priately small fraction of Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 0.27 gpm. The values for the limits on specific activity represent limits based upon a parametric McGUIRE - UNIT 2 B 3/4 4-5
DESIGN FEATURES FUEL ASSEMBLIES Continued) to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-
! limiting core regions.
CONTROL R00 ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material for Unit 1 control rods shall be 80% silver, 15% indium, and 5% cadmium. All l control rods shall be clad with stainless steel tubing.
. 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
l
- a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, i
- b. For a pressure of 2485 psig, and
- c. For a temperature of 650*F, except for the pressurizer which is 680*F.
1.
VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 13,050 1 100 cubic feet at a nominal T,yg of 525*F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE j CRITICALITY l
i 5.6.1 a. The spent fuel storege racks are designed and shall be
! maintained with:
- 1) kg 5 0.95 if fully flooded with unborated water as described in Section 9.1 of the FSAR; and
- 2) A nominal 10.4" center to center distance between fuel assemblies placed in Region 1; and
- 3) A nominal 9.125" center to center distance between fuel assemblies placed in Region 2.
McGUIRE - UNIT 2 5-7 Amendment No.
~
( 1 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)
- 8. DPC-NE-3002, Rev 1, "FSAR Chapter 15 System Transient Analysis Methodology," SER dated December 1995.
(Methodology used in the s the core operating limits)ystem thermal-hydraulic analyses which determine
- 9. 'DPC-NE-3000P, Rev. 1, " Thermal-Hydraulic Transient Analysis Methodology,"
SER dated December 1995.
(Modeling used in the system thermal-hydraulic analyses)
- 10. DPC-NE-1004A, " Nuclear Design Methodology Using CASM0-3/ SIMULATE-3P,"
November, 1992.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
- 11. DPC-NE-2004P-A, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," December 1991 (DPC Proprietary) .
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumenta-tion Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor FAH(X,Y).)
- 12. DPC-NE-2001P-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 -(DPC Proprietary) .
(Methodology for Spec 1fication 2.2.1 - Reactor Trip System Instrumentation Setpoints.)
- 13. DPC-2005P-A, " Thermal Hydraulic Statistical Core Design Methodology," l February 1995 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 .
Nuclear Enthalpy Rise Hot Channel Factor). !
(Methodology used for Specification 2.2.1 - Reactor Trip System Instru-mentation setpoints).
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved j by SER dated February 1994. '
(Used for Specification 2.2.1, Reactor Trip System Instrumentation Setpoints).
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS REPORT, inciading any mid-cycle revisions or supple- I ments thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Ariministrator and Resident Inspector.
McGUIRE - UNIT 2 6-22 Amendment No.
1
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