ML20095B024
| ML20095B024 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 04/13/1992 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20095B017 | List: |
| References | |
| NUDOCS 9204210268 | |
| Download: ML20095B024 (179) | |
Text
, . . - . . . . . . . . . . . . . . -. . .. . - . . . . . . . . _ . - . . _ . . - - ~ . . .
. . ~ . - . . ,
I l
i I
I i'
I I
I d
1
, FILOAD FIPJRT 4
Catawba Unit 1 Cycle 7 t
i O l
-{
I i
l i=
i-l'
\'
.. Duke Power Ccapany l' IIuclear Generation Department
!!uclear Engineering Section 9204210268 920AtL PDR ADOCK 05000*..
P PDR
. - .- . . , . . - .- .__~ -. -._~. , - . - _ _ _ . - . - . - . - - . . - . - . - . _ . _ _ . . .
1 l
l l
1 i
c'em ents Pne
- 1. INTRODUCTION AND
SUMMARY
., . . . . ... . . . 1-1
- 2. OPERATING HISTORY , . . . .., , . . . .. 2-1
- 3. GENEhAL DESCRIPTION... ... , , .,. . 3-1 l
- 4. FUEL SYSTEM DESIGN. . . ., , , . . .. . 4-1 4.1. Fuel Asse:nbly Mechanical Design . . . . .. .. . 4-1 4.2. Fuel Rod Design. ... .. .. . . .. .. . 4-1 4.2.1. Fuel Rod Cladding Collapse... . . . 4-1 l 4.2.2. Fuel Rod Cladding Stress... . . . .. . 4-1 1 4.2.3. Fuel Rod Cladding Strain . .. . 4-2 4.3. Thermal C esign . . . .. . . ... . . 4-2 z-4.4. Material Design. . . .. . . 4-2 4.5. Ope ra t i n t; Expe r i enc e . . . . . . . . . . . . . . . . . . . 4-2
- 5. HUCLEAR DESIGN, . . .. . . ... . ... . 5-1 5,1. Physics Characteristics. . .. .. .... . .. . 5-1 5.2. Changes in Nuclear Design.. . . . .. 5-1
- 6. THERMAL-HYDRAULIC DESIGN,.. .. ... . . .. . . .. 6-1
- 7. ACCIDENT ANALYSIS., . ,, . .. ..... . . . .... . 7-1
'8 . PROPOSED MODIFICATIONS TO LICENSING BASIS DOCUMENTS. 8-1 8.1 Changes to Technical Specifications.. . . , 8-5 is 8.2 Changes to Core Operating Limits Report.. . .. 8-112 i 8.3 Changer to Final Safety Analysis Report . . . . 8-123
- 1) E E t.c 7 s.u
.,, ,t.a
., . . . .. . .. .... . . . 9 i I
l i
.1 i 1
'irt Of Tables Table Page 4-1 Mark-EA Nel Design Parameters and Dimensicns . 4-3 l 5-1 Phy s 1 : .? Farameters. Catawba 1 Cycles 6 and ' . 5-2 5-2 Shutdl.-m 'argin Calculaticn f or Catawba 1 Cycle 7 . E-4 I 6-1 Systet L:, certainties Included in the Statistical Co" Lec1;n Analysis . . .. . . . . . .. . . 6-2 6-2. IJominal Thermal , Hydraulic Design COnditicas, - ~
u a ,a. .. ,. . c.
. s,. u . e., > . ... . .. e-a 6-3 DI!ER F enalt1es ,, . . . . . . . 6-4 6-4 Fle'., An;maly Peaking Fenalties . . . . 6-5
, 9-1 Tecnnical Epecificaticns Changes . . B-3 3-2 Cole Oper n ing Limit s Feport Changes . . . B-4 4
4 tict *f Fimlrer Figure Page 3-1 Ccre _cading Pattern fcr Catawba Unit 1 Cycle 7 , '! - 2 3-2~ Enrl;htent and BOC Burr'.p Distribution for Catawba Unit i Cycle 7 . .... . ... . . . . 3-3 3-3 Catawba Unit 1. Cycle 7 Eurnable Absorber and .E;urce Assembly Locations . . . 3-4 5-1 BCC ~4 EFFO). Cycle 7 Two-Limensional Relative F0wer Distributi:n - HFP, Equilibrium Xenon . . . 5-5 i
l 2
l 11 i
i L
i F
i-1 i
l.'
i- !
.k.44 XV n dv,4
~ ~ 44 v4+
- n. , ..,n ~
D V,4 wy, Vu,14
, .. N4s i
This repert 'ustifies the Operation of the seventh cycle of Catawba l Nuclear Staticn. L' nit 1 at tne rated core power level 01 3411 MWth.
! Included are the required analyses as outlined in.the USNRC document j ' Guidance cr Propesea License Amenam^nts Relating t: Refueling," July 1975.
i
- Cycle ~ for Catawba Unit 1 is the seccnd Catawba cycle for which the reload fuel is supplied by EG*, Fuel Ccmpany (BWFC). Che incoming Batch 9 fuel assentlies are designated as Mark-BW. To support implementation l of Mark-BW f uel in the M Guire and Cat awba nuclear staticns, Duke Power
[ Ccmpany (OPC! has develcped new methods and models tc analyce the The thermal-hydraulic
- plants during normal and of f -norr.al operation.
! analytical models are documented in tcpical report OPC-NE-3000" l (Reference 21: for non-LOCA transients and BAW-10174 (Reference 13) for l LOCA. Porti:ns of the analytical methodology are documented in tcpical
! report DPC-NE-300lP (Reference 12) and CPC-NE-2004PA : Reference B),
' tnethodolocy is documented in DPC-NE-3000. The FSAR Chapter 15 LCCA system transient analysis methodology is documented in Ref erence 13, l- Approval c: t hese tcpical reports ha"e been ccmpj eted.
i-l Section 2 cf this repcrt is the operating history fcr fuel in Catawba l Unit 1. Section 3 is a general description of the reacter core. and l the fuel system design is provided in Section 4. Reactor and system
! parameters and conditions are summariced in Secticns 5, 6, and 7 L Changes to the Technical Specitications, Core Operating Limits Report
! (COLR). and Final Safety Analysis Report (FSAR) are provided in Section l S.
i All Of the a cidents analyced in the FSAR (Reference 1) have been I reviewed for Cycle 7 cperatien, and many of the FSAR Chapter 15 system o thermal-hydrauile accident analyses sensitive to relcad core physics parameters have been reanalyced using Duke Power methodology. Several bounding transients were analyzed in detail to demonstrate the capability of OPC calcularicnal techniques. The results of these analyses mre reported in OPC-NE-300lP. For the other reanalyced transients. the approved methodology is documented in DPC-NE-3002, A l further discussion ci accident analysis is presented in Secticn 7 of E this repcrt. Other reanalyzed transients are included in Section 8 of
, chis report.
( Amendment Mumber 74 (Unit 1) and Amendment Number 69 (Unit 2) to the l Catawba Muclear Statica Facility Lperating License allow-the removal of l cycle-specific core parameter limits frca Technical Specifications and require that these ilmits be included in a Core Operating-Limits Report (COLR). The Core Operating Limits Report is suhmitted to the URC upon issuance and d es not require approval pricr to implementation.
Changes t the cperating limits are made via the Core iperating Limits Repcrt.
The Technical Specifications have been reviewed, and the modifications for Cycle 7 are Justified in this report. Based on the analyses performed. It has been concluded that Catawba Unit 1 Cycle 7 can be safely cperated at a core power level of 2411 MW t .
11 l
. ._ . ._ m-. _ _ _ _ _ . _ . .. - . _ . _ - . _. _ _ -. _-. . .- _ _ . _ __
~ __ -. _ . .i j
1
, . . , u v ,, , ,,
- s. 'J I~ Do s ,. , u-,ad*A44J I44Q5U .% 1 l
l The cun ent :carating cycle f r Catawba. Unit 1 is Cycle 6, which achievea crit.cality en June G , 1991 and reached 100% f ull power en June 25. l};. Cycle 6 is scheduled to shut down in June 1992 after L l
350 EFFD Tarawha. Unit 1 Cycle 6 cperated with f resh Mark-B'A fuel
- assemblies. F
- evacus to Cycl; 6, Unit 1 cperated entirely with fuel l assemclies ;f .lestinghouse design. ;
Cataxta, in;- 1, Cycle " is the second Catawba Unit I relcad to contain a full relO5: ratch of Mark-F.- fuel assemblies (FAsi- Cycle 7 is s:heduled t: nartup in August 1992 at a rated pcwer level of 2 411 mit ;
and har a de;;gn cycle length f 250 EFPD. No crerating anonalies have '
..y. , p . , ,, .,4Ava.
y . g ,.uc .: a+ - p c.r a r_ +i 'n.."a *.V ^m' n d.. -
r - , u v6w . . n.. x .+. 2 .
-; m Wt . .at a .dw. _ . .m e i)' 3 * #. .> . ~_ w i
- c. y o 1 , n,_ ., .4 . .y . ..a~ .m a.
4 66 ,u4e
- 1 b
4 9
r I
i i
l I
I I
I r
i I
I i
l-i r
l l.
I c
h-h 7 1
- .s e.- - -w --. . , + .m+ e -ec .m,= ,
_ _ - _ --- - - - - - _ . _ _ _ _ - - _ _ _ _ _ _ _ - _ _ _ _ ~ _ - _ -
1-I i
f i I
i
?
m m. .m.a ne m.u ,e -. . w.
., ~
l ,3 1
i i
i l The Catawca U~.it i react:r ::re is described in dMala in Chapter 4 of ,
the FSAR ; E-terence 1). The core ccnsists of 193 asserblies, each cf )
which is a 17-ri-17 array ::ntaining 264 fuel rods, 24 guide tubes, and 1 incore inst: m.ent tube. ~here are 121 burned FAs in the cor e; 71 of I t he Mark- E'.' d= ngn tReference 2: and 49 of the Westinghouse Cptimiced l
Fuel Assently desig". and 72 f resn FA's consisting of the Mark-EW l
design (Eeterence 2) The :uel red outside diameters are u.360 and
, 0.374 inct li the clad thicknesses a:e 0.0225 and 0.024 inch :cr the OFA and Mark '. designs. respectively The Mark-SW fuel censists cf l dished end, ~.uindrical pellets cf ulanium dicxide. (See Table 4-1 fcr t
l- data). 7.ie rc= rage nominal fuel .0adings are 424.90, 426.41, 456.63, j and 456.20 k; :f uranium per fuel assertly in batches 6A, A , EA, and 9A, respectrieb The initialgr. rich.';ents :f' batches 6A, 7A and EA were 3.25, 4
.s. and 3.55 wtt d3'1 ,. The design enrichment of the fresh batches 7A ' M ar k- EX) i s 3 . 4 5 wt 't 35U.
The 4 tat ch { A. 45 batch TA, and 72 batch BA asse-blies will te shuf fled t c -
Iccaticns. 'ne batch 7A FA will be inserted intc *he
,. Center asse ri', iccation. The 4 batch 6A FA's are re-inserts frcm the spent fuel sc;1 discharged from Catawba 2 Cycle 5. The 72 fresh tatch 9A assemblies all be loaded into the core in a basically synnetric checkerocard pattern. Figure 3 -2 is a quarter c:re map showing the burnup and re n on reference nurter with correspcnding initial enrichment s c f each assertly at the beginning af Cycle 7 Cycle - 4ill .ce operated in a f eed and-bleed mode. Core reactivity is controlled ti 53 rod cluster ccntrol assemblies (RCCAs), 528 Mark-EW burnable abscrners, and soluble bcron shim. Figure 3-3 shows the Cycle 7 fresh f uel _; ations with the Mark-Bh EPRA clusters and number f p.ns tenriched tc 3.0 wt% E,4C-A1,0.13 in each location.,
l I
4 I
l t
l _ .. .
J'
m . . . . . . . . _ _ . . _ ._ . . . . . . .~ . ..-... _ . . .. _._ _ _ _
-r.w,L,u- r. .:-A CORE LOADING PATTERN FOR CATAWBA UNIT 1 C'.'CLE Y ,
REGION REFEREICE NUMBERS REGICN hW!SERS
- c. ,I v u,dy- tvRt o, evLn . e,i wA, wS aN.
1 3" --
- N d- N13 .J' i2
+ ~ 4 4 +
4 #
Y Y e *h 4 ,,
n,b O w '- + % 9 's0 n.A 3'. ;
' - 4
) 9 3 1 ~
~
r r.i' c r '_+ r 2. ,
4
/ Th i $ P 4 NjD e I ND) h. U M3. P L' "
) d E 9 3 4 5 -
- v. . + r- r 4
- r_.- -
J A- 4 ,e
- ,9 ,$ - 4 ' k4 ) is '
- 3 am 1. o I.
?
~eU 7 1 7 s a" 4 't 4 4 r p ^. e- c- 1 r e_% e =; i 7 r
.) J ? a ti 4 i H i ') aTJ JSV J a. " M'd MhI s- < *t 5 5 ,- -. ;d 4 ^)
. O e '
) 7 9 9 7 i 9 5
&. 6* ? if , ' i ., C' T_ T,** 7.* 7 v_ ? O j1 9.4 , r , a M.* a .- ?*
s1e MJe Ms4 e) *t>
HJf Ji 6 i ,3 7 4
- - 4 4 , O ?
G-CA A.:4 r F
. F- * ' '
- 4 '- F --14 M ."
a h2J - . . '; .4 m"J M2 i . MW 3 .t . Jts tt i Js , r14 2 k J $ 4 N k h p ' S y =
p.' es * * :r y.'C . F k.a L - j r p , r'
- j. r *- r - Mi nr 3 c .s i iJe at , :t- ., r J.
+ - R 6 f 9 7 -
?
y .m 4 , .o s A-i s H . ;: 4 r (N 0 s : (9 __ .a p._4 i s
4 8 sae H45 j
dv. + ,5 J ', J) , t 's MI. M14 ,ib 2. . . nJ4 s>4 1 4 9 4 C E 3 9 ~7 9 4 9 d
- 4.9 - ' F . ;7 r --
r r.: ,
c.a3 . ._
l' J4; , ,. 603 nic 2Jv 4 J i i' nJ4 H11 it 4 ,.' 2 d nSe Jis 7 6 s 6 4 9 9 7 ) A -4 10 4
?
r
- t J-: R-1: J-10 F H-12 F G-10 A-1C- F-~2 M-Es b
Mk kD .k4Y hk# bbD NEN Ind kD hiD
'. E ) *> d
'h .*
.4
? 9 4 7
' 4 6 W ?., A, 4.
k .+ r Y ei b . k U . F ( [. E M=I [ 9 .. 4
- s I 'I d 4% sb 15 ') Jt4 $bN La bC .b A . J bsj
) '
9 7 4 ? 9 a C E.t U 13 .'* O. '
r M.*1 r i '
As d>. . %! H:s re! J4) H54 J4; M34 m,s
- 4 A 4 S
- 4 $ 9 5 a l -
r r '_" ' ' p %^y r *' ;T C, t . .' ;
- 31. * > u -. T 4 n.
) b ) p kY = 6 . k 1. U 's 14
'l 9 , 7 ) 3 9 i
e r P_ 4 e u.!J C .34 7 QfL n'
.,S
.l" Sp eb ,
....* Nh.., _,M4
' Cf '
2
. 0,- n. 9 @ }
. .. f .
t B
i h
I XXX EEG! . < FEFEF CCE MUMBER "I .v u, .r.ca ~. w = ms%s v.w % .a rt;.
t,e.,+ m o.f..* e .y , , y%A.~,. =* e '*. e e aa y a
hM g
-*{gI .' M. {' J.t M
<'9.[M*
. - wa u, %'t,, A a .$.++g 1 r
1 i
l l
I-1.' ~
l l
I l'.
l
, , .-v ,,.m, .-- . . , . . , . . . . , _ .. _ .. . . , . . _, , . _ . . . .- - . - . - . . ,
FIGURE 3-2 E!!RICHMENT A'"O SCC EUFlIUP DISTRIBUTICU FOR CATAWBA 1 CYCLE 7 H G F E D F B A 1.'s a q.7 :ac,
..- a.n- - ..33;. 7 .g .c,,a ua.. .
, , , c;; o . ,+
s,- .
3 ., .x ., . e' .3 ,
3 33465.0 ;5959.4 20422.3 .0 31450.3 19709.4 33694.8 .0
-' A cA AA 9A 7A 8A 7A 9A 11894.2 .5757 ' .0 26706.2 .0 16997.2 .0 16602.4 9 15862.E 17913.9 .0 29975.1 .0 18635.5 .0 17979.1 BA ~A 9A 7A 9A CA 9A 9A 26932.6 .a 7455.5 10968.7 30412.5 0 17340.3 .0-10 30419.6 .0 19698.2 15349.7 32349.S ,3 19066.8 .0 6A C' A PA- DA 7A 9A AA 9A
.0 167G5.3 .0970.t 16760.5 .0 157E0.1 .0 24187.3 11
^
23974.0 15352.8 17917.3 .0 180E2.9 .0 27560.1 9A 7A 8A BA 9A 9A 9A 7A 28236.2 .0 30417.5 .0 26910.4 .0 22154.7 12 31454.3 ,,
.3 32353.4 -,
.0 30360.0
.0 26220.2 ,,
.n 98 in m .
.a fa ,n m
1,t o--..r2 t s5+ 8 0.1 .3 3 ,
sm93s,o
.--f, ,
- soJ1J., . C, 3 3
s e n. - -zu., 3
,2
.s 1 01 - 0, ',f -O .004r.3 V 3p
.sv n 0., . e. . Un 1
48i1 -~.4 . 4a 4t i / V 6 . '-)
EA ?A 9A BA 9A 9A 8A 32253. 6 .3 7342.0 .0 22156.3 14933.0 Average
-, , ,. ,,, a
.0 a 00<.4 .3 14 3 3e36.e ,
a t e,z,0 . 0 t<,9v.o- Manmum 7A GA RA 9A 7A BA Region p
.0 6605.9 .0 24186.5 15 .0 l',984.0 .0 27560.5 9A OA 9A 7A I
P ENRI^HMEIrr CYCLES NUMBER OF BOC BUPJ,"JP R E 7. !' ' wie '2?E FUPfrED ASSEM" IES MWD /MTU
'- EA :.25 2 4 26935
-. .n 2 4' D' vd TO e .1v
,'MJ,c A ^ "I ea
- .25 1 <4 ,5<36 1
9A 3.45 0 72 0 CCRE 193 12698 1
l l-l 3-3
1 FIGURE 3-3 CATAWBA UNIT 1 CYCLE 7 BURNAELE ABSCREER /d;D SOURCE ASSEMBLY LOCATIONS !
l 1
- o r
. , 4 4 m- 4m 4
., . .. 1 ., ..
4 y .- ,
. .. . ,, _s 4
- . : i; 1;
(
, . .: ;s a l
903- 2?Do 4 , 4. -. .. . ,.
4U .
b&, .m 6m
,.1 ,
. .s u,
.. 4 4 , . . . .. is 1; 4
. s , ,, , ,
MJ 1 E A& 1 1
14 4 4 4 4 b
- i M . r 4 a r E o a A
'RJMEER OF NUMBER OF MkE.'-EP PIMS/ASSE"cLY PAMPLATE-ASSEMPLIES .
,A. 85. *4 i,
- 44. ,G-All locaticas use 3.0 wti sac-Al Ot7 MkBW BP's.
t -
- . 3-4 i
I '
l
- 4. FUEL SYSTEM DESIGN 4 1 Fuel A_' erlv Merhanical Derin The Catawba 2 Cycle 7 core will include 72 fresh Mark-BW tuel assemblies.
- e re-inserted fuel assemblies in Cycle 7 will be Westinghouse ?ptimiced fuel ascemblies (49) and Mark-BW fuel assemblies (72). The 'Ork-BW 17 x 17 Zir:alcy spacer grid fuel assemoly is similar .n design o the Westinghcuse standard fuel assembly, Reterence
- 2. The fuel rcd _, uter diamet er and guide tube tcp section, dashpct dia* net e rs . and instrument t'be diameter are the same as the Westinghcuse standard 17 x 17 design. The unique features cf the Mark-BW design inc.ude the Zircalcy intermediate spacer grids, the spacer grid restraint system, and the use of Zircaloy grids with the standard lattice des gn Mark-BW fuel design dimensions and parameters for Catawba 1 Cy:le 7 are listed in Table 4-1.
L2 Fuel P i Desian Duke Fower C=pany has performed generic Mark-SW mechanical analyses using the approved methodologies described in Reference 3. The generic analyses envelope the Cycle 7 design as discussed below.
4.2.1 Fuel R:d Cladding Ccilapse The fuel rods .cere anal'fzed for creep collapse using the CROV ccmputer code, Reterence 4, and the methodology described in Reference 3.
Internal pin pressures and clad temperatures used in CROV were calculated using the. TACO 2 computer code, Reference 5. A conservative power histe y ahich envelopes the predicted peaking for the Catawba 1 Cycle 7 fuel las analy::ed. The ccllapse time was' conservatively determined tc be greater than the maximum predicted residence time for the Mark-5W fuel (Table 4-1).
I t 4.2.2 Fuel Rcd Cladding. Stress f-l l- As descrtbed in Reference 3, Duke Power Company has performed a l conservative generic fuel rod cladding stress analysis using the ASME-
! pressure vessel stress intensity limits as guidelines. The maximt.:
l cladding stress intensities were shown to be within the ASME limits.
- i. under all Icading conditions. The generic Mark-BW cladding stress analysis -includes the f ollowing ccnservatisms:
!
- C:nservative cladding dimensions.
High external pressure.
!-
- Lsw internal pin pressure.
l-~
- High radial temperature gradient through the clad.
I i
5 h
I.
t i
l s
41 l
L-
4.2.3 Fuel r::d Cladding St rain Diametral :_ adding strain resulting from a local power transient is limited t? . 0%. A genaric cladding strain analysis was performed using TACO 2 :: determine the maximum allowable 10 cal power change that the fuel ::uld experience without exceeding the 1.0% limit. The maximum cal:2.ated local pcwer change resulting frcm a worst case c:re maneuvering ;;enario was ;ompared with the maximum allowable pcmier change, 'his c:mparison dencnstrated that margin exists to the 1.0%
strain ilm:- i s
. mw,.-2 ,,
. o. c.. . ~. . i s . 2_ . . _ .
W The thermai perf ormance of the Mark-BW f uel assemblies was evaluated using TACO 2 ith the methodology given in Reference 3. The ncminal f uel parameters used to determine the generic linear heat rate to centerline i.eit (LHRTM) limits are given in Table 4-1. The LHRTM analysis in: uded the foll: wing bouading ccnservatisms:
- Dximum gap based cn as-fabricated pellet and clad data.
- aximum incere densiticaticn based on resinter test results.
The max 1 mum predicted Mark-Eh assembly burnup at EOC 3 (in Eatch e) is 33,868 MXD, r:U and the maximum predicted f uel rod burnup (in Satch 8) is 3 5,502 :'t:0'MTU. The fuel rod internal pressure has been evaluated for the hignest burnup rod using TACO 2 and a conservative pin power history. Tne maximum internal pin pressure is less than the nominal Reactor C::;2nt System pressure of 2250 psia. The reinserted OFA fuel design crite::a was evaluated sich acceptable results for the Cycle 7 predicted creraticn.
! 4.4 Materis resian i
l- The Mark-EL :uel is not unique n concept, nor does it utilice different c:~ronent materials. Thus, the chemical compatibility of all
- possible f uei-
- 1 adding-ccolant-assembly interatticns for the resh fuel is identical c -hat of the present fuel.
4.5 -nrere : 7 Exrerience
- Experience .;th the Mark-EW 17 x 17 fuel assembly design started with I the irradia
- On of four lead assemblies in McGuire 1. Cycle 5. McGuire ,
1 Cycle 7 -113 the third cycle of irradiatica for three of the
( assemblies and the maximum predicted assemnly burnup is 42,756 MxD/MTU.
The lead assemclies were' examined after cycles 5 and 6 and fuel I
assembly bcc twist, growth, and holddcwn spring set were all within
- nominal bounas. A poolside examinatica of the Mark-BW fuel is scheduled ::r late 1992. Four other Mark-BW lead assemblies underwent L their first.:ycle of irradiaticn in Trojan Cycle 13.
l
-Catawba 1 Cy:le 7 will be the f ourth complete reload batch of Mark-BW 17 x-17-fuel. The first ccmplete reload batch began operating in Catawba 1 C;;;1e 6 in June 1991. The second batch, McGuire i Cycle 8, ,
began operaticn in Decamber 1991 anc the third batch, McGuire 2 Cyicle 9, in March ;)92.
4-2
Table 4-1. Mark-BW Fuel Design Parameters and Dimensions Paten 8 Batch 9 Nominal fuel rod CD, in. 0.374 0.374 Nominal fuel red ID, in, 0.326 0.326 Nominal active fuel length, in. 144.0 144.0 uominal fuel pellet OD. in. 0.3195 0.3195 Fuel cellet initial density, % TD 96.0 96.0 Initial fuel enrichment, wt. %Um 3.55 3.45 Estimat ad residence time EOC 7, EffH 16,500 8400 Claddir collapse time, EFFH >18,100 >18,100 Nominal linear neat rate (LHR), kW/ft 5.43 5,43 Ave. fuel temperature 9 nc=. LHR, deg F 1360 1360 Minitura LHR to nelt, kW/ft 0-1000 K4D/MTU 21.5 21.5
- 1000 t&D/MTU 21.9 21.8
.4-3 i I
, . . . _ _ , _ _ . . . _~,-_-. __ _ _*
O, I W Y g
.4 WNL u W v - J w"*%n % T C, "b,-- c ~ , n+*a, s -** - - -
1.. .
I .'^y b yd'*h 7 Q FP.a9 pF* =
g
.*.Q.
h'N N.
& 1 , , 4 t'p. 9
. .- A Q -, ?.) 4 6 .- Qr r'*
%+ .T* a- -
g .%.% J F
'.,-+wJ 4 i .-*
7'
- p. t. Ws+w- -. > 6 A j
,P % .I .,e P&
. = '1
,. a { ia o +, .._._y c, ; . . .ae a_4. m m .~. g py in, ,
.-...p W. r.c e .h. +u 4. ,1 c, - .. ,. .
u4a_e r- .r ,ra m_ .4 .n.
- . .t' ."Jm' _ *u' .O n' f)n*
rtr. *tpp pr.t.._ m .s
~~ro 3. w3 :6 .- .,. 4. sr v.
- um ba %s ~ a .,s. -. ~ r.,1 a. i co .~9
- U'
.m y w 4
- rrv r_ t' tun q + - '1wn
- . , . , , rJm - , e. c, 3 & v _ye, .t u3.t . , a ,u t, . , .t , r t,1 a . ,. e'
- ,- n t I,,7 r U i . r y.4 . ,e.-- . - A 4
..aaLA w . , - -. ,
r,v a r2 .,
+e' 4.,11 m',
- 4. 44av , e .v. 4 .h.,. . > .. s.-_
. - . y. e u.h.p -
bpm
- u-.g.e~...,.* n , , ,
. ^.
L, ~ .l a
- m u.a - e,. . rp ~- .s. p .r ... d. ie 4s .us-.a W 3e
...a G~
3 s? f 3,s s} 0 tsa~,== s a :a --
- J' .t- e s m*-r
- w. k. -a ".a~*.J.s.
- O v es ..4.e p ~ n .it,s e .iwO y
).-*em
- 2 4 . .s yw..p W
^L "uO, ^J >* -e : p t-a p +
,u..s.4-u..r**, .- s y dae w'rn ~ .,,1. 's .r .. wR, n6M1 * * ; p p .Gp.,ppga **. k.
. , s , c. . . . ~ p .4
.A.
m .,,,Oy ._. r .,y;4 ~ .., - .. - f.f. - ~ .1 a ~a sm.s. .
pqa.,
me A1 4f .V. ., s ~. ;..- ~n. .y. . . ad .v . 4. q . .rl Y .k.g.
. . . w M' .'. .T 'P' .n v4 '-a ?,,*
sas
.M,'"*'. ?,,.'.
r y ...**%
2 ...
r,%
n dA4 U]
, ar
. .. -. .-*..n.
,si
- v. e. k. +a .,-,..w.v.
,*,-} ,. 3. sy - -
.s.c. u C &r w. 4 v- r.
. .'..u. ;,,.,,-
o rp, , , p4s 4 n.
.~.>p,.A A .
v .k.g _ - LT,,, T D. ,
~q -. . . . - 3e -p.A
~
.a p m p
- a. 4 .r w,
, _ _ _ . # A *Ay m. wum - r
- H2 ne ma v. k. g.., r 3. 4.k.o r p e. . w~ a t^ . -.,
4 *-
m.. t. .L. p r- r . i t t- ..~ ^ y
.] a e p r. >.+w.3..n.,*.n.
_._.r.9.a. a.* g *, p. , * - *- n. s.. c 4 'l p .r p j . '*'.h.a.-
. s} H.3 ap v v .7 .y m e
.. 2av *** cae 3l a y.. v.m- .4 -T e.b.a- .r A u- p m.w. .e e w 3 e t/n4 ... . . '* ' n W. i ta C..~-..
.s + ,
Se .r-m O. I y-, 4. .e. <64 ..n. - . av.. -
- 1"
. .b. a.
fa.h.3 ,
- A,, m.A.. .
-y .,3..,
. .~ ..
- i a w 3 A . s, , 1 3- e 3. +- .n. aa 4 r. ,.. *6s i .d. . a,.. u2 4S} v i i. . .~.. p .v- e ,q 3. .r. e ,,
. - se.. ar ,.,3 .4 ' 2&.1 -W .- p ,. - ,M - - .
w
.e9. Vf p b. .
+-~
' . ...4...,,..
. . *N. 4. g 1, ,*. . . 1.~.,1 3 4 w , i 1,wA . .,
3e-
- .b. ma t" .. 3 % ., ,y. , , . .** '
4
.1.'.
mA . 4} . i sv , nA A .. ' , -
si c,r. m. ./ .
c , .^.2.,
- M_ b._,v ,.c.
. - m
, '_a-a .4 , *. . , _ - , , n. g
&& _ a h * ** 9** b , p *
-g_jp),a'% 4f.t."'.TWe .7 t .89." .."%. u,f_.'. m h
/k+ N $* $ f% 8M Q Fk u-m Q.- m, f 7"'u9 " ' ' . . .1
~,,'4"./,"49."'.
[*
in -
. - s- m . _ . . % .a . .
.k. p * .m.-. ..% yu* ~ ) . e, "
^
er V _'=3y **p e)n e,.. v, u .r .n..m.,p
' n
_- t a r. a e . %. i .. p e nus,*
- sa . .., e.,. e- ~ .1.e m a..e.~. .m p t p r ae
.'. .m..eaw. w .
. . . . . .. . m upne
- y ar.4 m .e, .a . . ....a.-.
. w a .rpn-v-v.m .,. -.:,- r a %.-g -,.' u s. . p taa _. . v1,u wn r. .a.n. p wo a,um r a .e. - t.m np T _. m%r _ p
.r
~A.k.a v *2 - .- Au-.aa- 3,4 s o.' '- ~w u . .,A,u a W a +v e a,.~,, ge.mr -~, .p,4 .k.., ew .v. a_ype ...,w 3 .,.> uAnm, s m. s n...pe.r ut 2 3 n
w ,A'- O 4 . , .. - < ,. y
. w.p T' .m., we :,,,,
._ 3 A . w .1 t . v . /.o ,W d. y t.o ry.&tw , .,,m6 3- ...
- r,. 4, . . -
AU p..d I p r e_ . . a. . .
, . . . e s sAv
- w. - 'i u~d 4 3.
- s c r. . .,a .a.c., r O n a I' f, r ., .. p,d A .i m.t,D g. i .- . e
. . . . - .. r e. p T _ q. vv m p _ n 2 ~ d_ 1 v4v %.
em .u . .e . . t .h..r o. p n
(y..s.ba v, 4.
4y A Y & 4 :s . . = - a, n. a.
.-- -r
%= b b h ,v.6 t b 6 4 V~ ..,6 o]. -a%* N .d, e m$ -y
% L %/, f 4
4a 4 4.La e Q (b.4
-. J w
r 3 p 6 .4,.,
w44 1 1
1 '
A 4.ewa4*e 6wa f - r. *m**.. a ,. a.
- - %. v o o arc %,
- Ars
.s. . . 4 4. eu4.yb1r a..a .Iey. .d. ~s .#. ..r *...a h.4 - + iya.i ,. *j..1'I ts., r.*1.-a u I nrg
- u. p eu. u ,a A T4we- - ~. . s. p- p- .r- ., ,-,xC' - v.. .y, v. i .+a a ~.s .*v.L~ . n. .pp+_$
. . _ & e. , r t- e. ,.~ p n :_. ,, .4 4.r a ny-.,,,
s p.w 6 A#p A 4. m. r .*,. C2 .
- f n ,-
w? P R,a % v r, e 4 . c. .~.,) +
Z e ,.,.-. . .- .' . v. ...1 e .2
- w. h.a p. b. .p ~d. p a, S ,.,e.1m. . e O
- g. ., r,.a. b.e. . 4 t J A -p - . .o r. a
- w. e .+ .i a L v"t .s.Q. . . A). w., . . .m.
, . ,,.r v .w.
%r*
uy .Ia *"' ..
r ;.:
rm r n :._ 7. .
9-
.A
,a 4.. ^pe.va . .u es ;.. _
- I 1
i
N Table 5.1 Physics Parameters ( ) Catawba 1 Cycles 6 and 7 Cyrio 6 cvele 7 Design cycle length, EFPD 350 350 l
Design cycle burnup, IMD/MTU 14234 13843 l'
Design average core burnup - EOC, MWD /MTU 26371 26541 Design initial core loading, MTU 84.0899 86.4443 l Critical boron - BOC,ppmb, no Xeib)
! HZP, ARO 1664 1726 HFP, ARO 1529 1565 Critical baron - EOC,ppmb HZP, 110 Xe, ARO 567 622 1 HFP, Eq Xe, ARO O 2 Total Control Rod Worths - HZP, pcm BCC 691S 6585
- EOC(c) 7221 7238 Max ejected rcd worth (d) - HZP, pcm BOC (D12) 3 55 (e) 449 r-t- EOC(C) (D12) 473 (e) 05 l
Max stuck rcd worth - HZP, pcm EOC ( F -10 ) 992 1436
! EOC(C) (F-101 983 1386 l
l I- Power deficit - HZF to HFP, pcm BOC -1668 -1752 EOC ICI -3117 -3049 Doppler coeff - HFP, pcm/CF BOC no Xe -1.16 -1.19 l EOCi c) eq Xe
-1,46 -1.48 l
- i. Moderator coeff - HFP, pcm/0F
-2.15 -3.73 BOCgoXe EOC ,' eq Xe, O PPMB -32.27(e) -33.85 i
Baron. worth - HFP, p;m/ppmb BOC -8.27 -7 . 'i 6 EOC(C) -9.54 -8.86 I
t l.
l l
. 5-2
.. ~ _._ . _ __. - _ _ __ _ . _ _ _ _... _. _ . _ _ _ ._ _ _ _.
Table 5.1 Physics Parameters (d) Catawoa 1 Cycles 6 and 7 (cont)
Oerle 6 ccrie Equilibrium Xenon worth - HFP. pcm EDC .4 EFPO) 2677 2642 ECC 2993 2814 Effective ;+1ayed neutren iraction - HFP BCC 0.00626(e) 0.006191 ECC'2 0,00525(e) O.005229 (a) Cycle and 7 values wer2 obtained frcm Duke Power Company analyses.
(b) HOP aenotes het cero pcwer (core average 557"F Tavg); HFP denotes h:t : 11 power i590.E'? vessel Tavg).
{c) EOC frysics parameters calculated at design EOC plus 10 EFPD.
(d) Ejected red worth for tanks D, C, and B inserted to H P RIL.
(e) These */alues were generated by B&W Fuel Company. All etnet C/cle 6 valuas in this table were generated by Duke Power Ccmpany.
(.
Differences between Cycle 6 and Cycle 7 values for these parameters are more likely due to methodology differences than to differences between the cycles.
1 5
l i
b f .-
~
i i i
t r
5 l
l-5-3
. .. ._ .. _ _ ~ . .._ . . . . _ _ . . . _ _ _ _ . _ . _ _ . _ _ _ _ _ _ . _ . _ _ _ . . . . . _ _ . - . . . _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _
s i
F Table 5-2. Shutdown Margin Calculat:cn for Catawba 1 Cycle 7 I
[
Control Rod North EOC (PCM) EOCta'(PCMt I 1. All rod? Inserted (ARI), H2P 6585 7238 i
l 2, ARI less most reactive stuck rod. H2P 5149 5852 e
i i
l 3. Less 10% uncertaanty 4634 5267 i
Required Rod .iorth
{ 4. Rod Inserticn Allowance (RIA)(D) 273 356
- 5. Power aefe:t. MFP to H p(b) 2057 3354 6, Shutdown Margin (total available worth 2304 1557 minus total required worth)
NOTE: Required shutdown Margin is 1300 PCM.
(a) EOC physics paramete-:s calculated at 360 EFPD, i.e., design EOC plus 10 EFPD.
(b) The' Red Insertion Allowance and Power Cefect penalties in:the Shutdown Marg;n Calculation account f 0r the ef fects of transient xenon conditions.
i-l l ..
l l
l 54
- t' + -gve p-vy per,--f, e,y,-,gm,.-,- ,,,.y ,,,,,,y,,,w,.,, , . , , , , , _ , . , , _ _ , , . __ , , _ _ _ _ ___ _ , _
f
_ ..__ _ . .- _ _ _ _ . _ _ _. . . - - _ _ , _ _ - _ - . _ . _ . - . ~ - - _ . . . .__.
Figure 5-1:' E00 (4 EFFD), C:/cle 7 '"wo-Di:nensicnal Relative Fower ,
Distributicn +- HrP, Equilibrioun Xencn l 1
H 7 F E D C B A 1.0110 1.3007 0.9628 1.1452 0.9177 1.0654 0.5083 0.7284 a
, 8 1.0470 1.0063 1.0707 1.2593 0,9755 1.1317 0.8763 0.9811 1.j013 1.2306 .2722 0.9845 1.1618 1.1603 1.1976 0.6461 9 1.4066 1.3693 1.3893 1.0310 1.2696 1.2623 1.3655 0.9435 0.9890 1.2724 1.2642 1.2570 0.9854 1.2381 1.0753 0.7205~
10 1.0712 1.3894 1.350S 1.4041 1.0398 1.3178 1.2318 1.0438 t
~
1.1451 u.9643 1.2s66 1.2775 1.2752 1.1945 1.0397 0.3713 11 1.2593 1.0309 1.4034 1.3599 1.3722 1.33G5 1.268 0.7534 0.9174 1.1616 0.9851 1.2749 1.0j06 -1.1765 0.5797 12 0.9751 1.2695 1.0395 1.3712 1.1372 1.3482 0.9446 1 0653 1.1603 1.23E0 1.1945 1.1765 0.7564 0.3518 13 1.1315 1.2624 1.3177 1.3305 1.3423 1.0305 C.7645 0,8052 A 1975 1.0752 1.0397 0.5797 0.3518
, 0.8,0..- s , --
2.:05, , ,
1.,,<os 0.<%,4, 0 . ,< 6 4. ,4
. . .. m.., e 34 <
0.7263 0.6461 0.7204 0.3713 15 0.9310 0.9434 1.0437 0.7535 I
I l-l l
l l
l-I XUX F(AVG)
YYYY PEAK Fill 5-5
l i
. - . . . , - ~,,
3
.. .**.. rL *. n..
" Il [ W P i V e . % ,. . .D 1 Wi e . ~. s. r;.~i. ta' . c ~g r. p ~. 4 *. ~ :.*- i e v.. m v , -.
~y. r p .s a v.. .~.a. . *.-
m a m - .r w. *s. "g, -% j. , ~.. .)-e c-,
't.h.e py-.n p n c.v .. .
- h Qr c ata
" %4^
we e a4 . e .w .r 4. h. Sa.4
[^. sg--s .. a . n.s.'y
,re W -- A 4i{p s t.'h.,.,3
~~e
- r. g,g y..~y_.a u t: Pw. sa
-su,.- y,g y
- 4_ .e, va - -,'.tn=p wa . . .l -wgy
- 6 n e . 4_ r. .~~.n n a .
- w. .
, ~ .I p *
--7,
- 9
.f.. A "*e..N .M. 3 .v '.,t.'. C w E. .. {. .v .gv. w i c , &. i m*
u,s'.'.tu 3
-.'."1
- 7 a 0 .h. p 1 - = . a... , .+
a.1 s *ms ., e r 3.. m .F*nii < ,
.;..f v S e.$ u; -,.-
~ . - 1, a.3.1*p4
,v. j . . ,
.-.d K e . ,d a.
.'w
- 3 y . a 3. m t a ., y . s i s .s
. y m .r 3,n p r, e .r a - e w .v..g
~ mmC c p a, .r .+ .adiaw
- a. . . r sa
..t e vv .h vu ?.5 . .1 n .
jy -.n. ., rp . +v y. q. ~ ., ,.... :.
. o. G .5 + t* -e e L ,
u.re -e -- b $ .- -i s.-w = ~.
- p m.u .,b dg aw a, 4"' , ., c b . c 4a .i
- 9.
- t, ":M
..r . , e. i Oase t.s i i
3 ObGL-e., , c .u .. ..} % s . ..u . e. 4 t* a v w . . .
er La kn.,, s,
- s. .o. :
m--
- w. r .n, A1m
- %.-.r ... n. a +. c. o -. n. u e Q . 4.1 . 9, .y n t3 r. .4 - .e,ru . .
.=v < ta cs u 3. 1 v. -. 1. .,1. .:.
. < . m ,,..am .., ,
a u _i_
p* -.
m.-..
4 d o, .r .,. ~- s, m. . g_ +rt
- 3 ., n. . *., 4 4 mD g 4. p n ., .. n. _.
-. 3
. . , m_ ... .s..l.. .. :: p ..
- .4h a. q *y* c e- w _. gm. y*
g . .i *e--J ..P._ p r !w.i ..m.
- n. p , *. d.* n. . > - .- ..."7
,.rpA >, 4. .6 *S " Nwa., N e. . A 7a "' G b } {1' d. . s ?.*41.* .- ' p *.1 9D. 4 Table 6-1 N r.: thu t.ncert rnt ;es spec. :ically alculated f cr t tawra .
.*-v %j . .i g. - Cb p .% m1 . g gj e3 .c nu .i ., ,%e a. es.
-- w . e.
f"'s,3 - ,i w- L . -[. . ~..w,j, , , ,
. . . . . . -v
. - -.c w c,e au3 m* a s.4
.~c. p * , #,v ** L'./
. . .- 6 e 1 . C_ fs; . '*' b. t g
.2 . * .1 e *< n e .m.. s. ,. e. 3 2 I
- ..h.p . .r ...c a ." m% w m, 4 , , i 1.., _ , 4_9 . = =
ary..
=v...-e,
~*p jaw * ,'...e.C v.
~. .
y-q 4 T. .n.
. n 4ea "e" 3s.hI C 4g yv , e. .
mi , . . ... ~m _m_ _ . . -
_v n_ . m, w. ., 4 . - ~ , . , .. .;. . n , a p.u .
- m- w _. e.e _- r.
c . . m_ a. m i,m,n 3 m. .% c , ,., . . m_ : . . ....., . ,
i m.~ ._ ) r ., ,, o n.s.
g .r ., - v.3
_ . .e. . a. ,. u,o e q, m u m+ 3 . ,.
n... ,. _ ~o
.. - y
.n a~m.% ,.e , w% m.m>
-m, , , - - , , - r
- ana ., ,y. -m- _ . . . a, , 2 . c_ c_ - sw . 1- m...a.v .c . .s.e v. m n.
- . , ,s, , .,,. . -g e t,.n u, ,- ,.,y . -u .a,- _..a- - w,
,e m n a, ., , ,. = a ~. ~ a c.-..n r a .. a- .... .
.m .
n ,aese.t%
. - .4 ,~. -.
m . a.
.. w . .es m _, r g ,. . ,, c ,,,. o. o- .r.. .wi_n 4
. u.m ;> _ 3.
.g. o. q3 r.y. x._ o- . ' - a ~.y .me .e .a,~,. .,. .e 'o ",_u' ."m a P.','.d a " .' .'." .' O' ,' - - -. ~g a' "m' .M. . ".
1 . _m . a3 3 e.. .%
with .lest:n7n nse optimized fuel OFM . EWFC has pertermea a :.eries of si-..
$ A v .= ,w..
m ... .. a.. .r a - 2 .e . , .u . %. y ,. . . a- - . .., .3 4 s. 1u s . # t .M. .
- n. . . d e ca.' w~.a .
. . . '". .b. o. "aw. s . a"
.c p n t a; { e ,a. a. a
.- m e aaa t. .b a_ -.n
(,7 9 . t.t p { ~-.
6 4
't e o., yo rv h p .r t .. tb.m
- .- W
. h A.s.d e3 q +; j~. r p M. ..g _7 . 8.
q . . . . . _ .e i. _
E 3.n C . ETC 3' r9 TCP aCIC53 the Mark-Eh fuel, Reference .2 /s generic _ rans ;:::n care analy.,: 2 aas pertc = ed to determ:ne -he e ma i n...p aL ( e. a.
. ., . s.' s. L> .rwp. . o. t,. .
. w ew .n. .,. ,s .
m .. re r a ' .' r* p v. (3aa " c.l i v p Auv. %*ny e. s' T tM' n
O 1rs. 4*w e t.
b..a. M. . ., v s _ ~ * ,e o .i , w= e~. I. h. r10-
. u, .' m.-r A r* . .. s .b. w . * ** l.m %.. -
dE3 - , m,d.f., 2 u u. a. p .,.. .n_ * . 3 a . p.r. ,
~.y r 1 . , .o e a u ..u,. . . gd w v, r. p. ,w".' 1 *e"d os
--' b ua. wL2 ",..O
' ', e.^,
..v W
y s r.e. r p _ ssp u ,,, 3.r r.,g.. , n. +..H.,3D ...p
....euy.H., v . . ., k. .e_. ,
.. , m ,.m-
. a 9 s.t... ,, , y pw .1,, y . . .m ._
en,,,e a$.Sem .
..,y .
a _ i y , .$ .10%. W1' .' I. b_ p .v 4 r s e.3 va.
aj,> n4 , L d.e. d 1' a
+ 5 18 :. r .g. , 0. '.
- u. . - . .c e . ,'."..rm.,
. w p.p .,mp 7..
& u 4.6 d* 'n*
ar h.a. ". . A... d, uC s'. ae ' ' %d*te d u"'.'t'M ^p t .". a .it 'y' f.^.*
- ~
- f. ,u, e .i Au ._
, ~..,v..;, mv.. .
~. , r n e .h ua ne, .
- . s. . c, yt .; ..a
- ... naqo.r um i n. r r ,-.3 ,, e p. -' o" C .'".^ e ".vPwP p t^n9 a .I " "y "n'a S u4_ "m t".~.'.'. .' ". e dyH '
.n w d._ t .sv .n. .a ,l' t . . u n. e Lv. M aaa u. s.:is, y asc *u.b.ta.
b
%,m Aat umi.4, .~ , a, ra -- . 4..yu. ---a-4 _--.2 rwue.,3 r s <, rn 1
- s
-m.-
hot assemcli The rc.st cf t he nre was modeled as Mark-BW fuel. A number O f Ct a',9pCints and peaking CCnditiOnS Were analy;*ed, yielding a
- c. 3X .....u.u ,, ,- .4 n. . p- .; . a .1 . se-A
.,O,.
s 3 c .. .
w
...o., eM f n.u e l' .
u
[%7 anZ3. W'- 1.;W CCnditiCn h3G tsen Observed in Several Westin;n:ase
.,. , dn es v. a, . ... 1,.,.v.,,
.u_. l,~n v Hvu t u- u c _, u w - a... . a- ,,n;we. u n . u .a .hpu. ou v,n, v.u a. 1. . ;
.y 4a a; .wre. < . va X. . u. .a ue
. r. .O *s.s.t,. cero..
- #1 ;
e-v.
ev
..e. -w e .a . . .
. t%r 4 .s.v p .r. e.
. *. 3 - m a . . ..a. n . A u .; w w .r , hi
.vs
- o % .c.a + l. . a %w.
w . .
...a .
a,-k I" aa p&a -., - , r OA n t an.r, A .,%
unA s A ,.2
-e.
- m. .u.a ..a u v
- .e ns ..
y w-.8,T
- w. .v ,er.d. -.d
( a. > wJ r
... ,_ . m. m
.r o. 3.ou a p .
...y a .. - - .. - w..n%*n ..n. .+.h.a,. c. . .m a .sa 1,d3, we bw .
- - . c,
.w.A h o. .
. 1
.sd d .r txa b, v v r w.a_ . .~. , a l' ,y. ,
.-m a "' yea" a .'. * ~; "t Fs .". ". P 9. a "aa" P. . - aa c ~ e'd * ^w a ~. .'. N.;'. . a r " A .
- ^ y ." ' '~' d a " u."*"w', * . s."r'.^..
- h **^ wa-npM FQ y* r. k u.f.e...a'*
(*t 1 YM. m a
-? "
D P'
. .E..M.9~.. P 1 v {.s.."_-,
.m .T'"
Q h.*.*.F a eu .P#uj T
.k (-
a . T')
u , vs b. .. .'. P. .h.
peaki:.g anait; ind a DNER genahy and is applied to both the Catawba
~ na.' ~uj .-a 'J .- . ee . i .i '.'. ~. au" ' e D - 4 .
. . .p.va. ., %. . -3 m .. 2,.,..g , y s The two ;&ain' ng DNER peanalties assessed against the DNB margin .
ac w 7w9 *w . . ~ . ."a,.- .m. sd ". *- ai a - _ a ,
's ^ r. n' . .m .- -su' A y ') , p .' a e.u 4 n* "a t . m. '- a... n *.. a *. .. v- n. m v W*
1 pn d - '. . r rep._. 42 . . . ,v. +,,ni
. .- n. . ,g ow gn n. .e_. , .u u 3. r y. _ m. L . cn!,,J
.u ,
6-1 l
, . ~.. . .. _. .
- - - . . ~ .. -. . - - . . . . . - . . . . - - -
Table 6-1 System Uncertainties Included in the Statistical Ccre Desian Analysis Reference 8 Par-ceter Uncertainty Distributien Ccre' Power +/- 21 Normal RCS Flow +/- 2.2 % Normal Ccre Sypass Flow +> .5% Uniform Pressure- +i- 30 psi Unitorm
' Inlet Temperature +/- 4 deg F Uniform i.
T e
t l:~
v:
r-i 62 ..
--d- -- 8'+ - --,--r , ,m_o 7.g g. , , , _ ,_
-._.-- _ - . .. .--.... ._. . . . . _ . . . . - . . _ . . . . ~ . - ~ . . - . . _.. . ~ . . - _ . ~ -.. .- ~... _. ~ ~.. ~ _
a E-Table 6+2. Nominal Thermal-Hydraulic Cesign conditions Catawba 1 Cycle 7 3 +
J
- are Fewer, !!dt .3411 c:re Exit Pressure, psia 2280 f
i
/essel Ave. Temperature, Deg F 590.8 RCS Flow, gpm j55,000 l-I Care Sypass Flow, t ~.5 I
l '- Eeterence Design F.iH -,50
- i. Reference Design Axial Shape . 55 Cosine
~ur Ccrrelation 3'4CMV l
1.40 l
Statistical DNER Limit
' Design DNBR Limit 1.55 1^
I a
P b
6-3
If t
d i
a i
f, l.
a i.
4 Table 6-3. DNBR Penalties l '.
F Statistical DNBR Limit 1.40 i
- . Desic DNBR Limit 1.55 I
J DNBR Margin 10.7 1 i
l:
I'
'" E P Penalt'? Man-PW E-l l-
. .ransition Core 0% 3.8 1
. Instr =entation/ Hardware 5.6 % 2.8 %
Rod Sce 0% 3.5 %
Flow'Ancmaly 0 E 't nR -e Tctal ;UER. Penalty 6.1 % 10.6 %
Available DNBR Margin 4.6 % 0.1 %
i j . ..
~ ()-.i 1
....-.._-. =..- . .. .-.. -._ , - ._. .. . = . . - - . . . _ _ _ - . . . . . . . ._
Table 6-4. Flow Ancmaly Peaki.ng Penalties The fol10wina penalties are applied to the Catawba Maximum A11ewable Total Peakin:; Limits to ac: cunt for the RCS flow ancmaly. The ,
penalties apply to all peak magnitudes. ;
1 1
Position Penalty X4L % Peak 4
0.01 1.5 0.1 1, 5, i n ,
V.4 3.4 4
0.3 0.9 0.4 0.6 0.5 0.3 0.6-1.0 0.0 I
i t
l I
l 6
t I-I-
l l
l 65
, . - m, . . . . ,
- m. ~ a. m .A. c a. e
.en e - . r ip
.,.- m.. - a. -.,.~.- ... %.- ~.+n, ..-m e
--t -
.a. ~, e v c. m , ,3 a a..A.
- e m, .m ,- - - ..3 m -.%. -
w 4.r t y .mi. ma i.
m 1 ,- r . e- ,
..h.p.r ...a.' p,-,._ . m .. p 4
..n, .. _4-4 a . . *~e
-r__ --. .a * .f..,
3
_ . 7 .4,,r.,,.,
- . , .m. ,.v, r .o u a a- _ ha
.r--o u; - . a .z . .
. .. a.
. 1,y~. e , - um n- a . .y
.f m. ,a.4 u s ,- 4.....c,3omo _ c ea a .-a g .m - u. n. ~
&v. . o. . a.. . . .. v m. .
. m.
. . 0..o.. o. .s.,. s o.s. -
c.m , r .-. %._ . .
. . . _ . . . . . ...o
- c. ~, ~..s. s o .pem3 , . .. a..~ ,
.. . ,m g.... ., .4 m3 --m s_ e,~ a
.p.......m.. ~ , , . . . .s. . a
. ~ + _- r o .w , , m, - , , . ~ , .
.~~.a. . %,..,.
w
. co.r. v.m..e ns,- ..* N a .r .,...a ., _ .v. . . ... . . 1..
,o.. .~....~,g,..
a . . . -
c
. -.L.,. -.- m n a e... e.y,.q ~ e a .,~4 , .. , .-
ma .r ~. r . , n. a e. .r ., ee
. n =. ~ a~ e .- e , ?na e e o.. .. c>., e _o .s v , w1
- e. .a .'
y ; 5n.yee- ?.y -mes. Tau. emp + 1 p- m.
T a.
.Tmm n- n .. - A ,ua 9 . s.
.,.~_*a_ e. s v p d .m a . *. G.*..
si
.r n 3..-- e - r.
. . 1. .
- e_. . e. m _ ..
. w2.. _- r
~. amm 3 -- .e-- . . , cm.w.a. s. . m _-,,rn. ,. . ~, p.ga .r , , - . .
- m. mcm.1 a1d >c au x 1 w ,.., k. m 2i , e a .'
- ( y
. _. .. .>. iiv a
~ .r4. F.,. m- '
.r .4... ro ,n .e .y #- a .
n n.,-+ c_ - .,
m a ., . , . a v
._m."._-. -
H m 'ad. ".
. mW. , , '. *. ."_ 2 . a .' x= *. r> w.ne.r
. * , ._- rmr a. -4 . r mm.-4
. -.-.-4 r m '.- -, e. .x:
,.v , aa,u. ,- m + .~..u ,.a.a.w ed .ra.. ,
. - n #.
. m w ., m. . ~a .r a.r..a i
-_m
. . . a.- g. , . . _ , c . .
. -: -o_ , . n.c xot n. a _ _~ .r m +n t .c~e ura m - . ~. . w= n~
u a ,,v a ., ,.s.
, ,e .A
.he an,...., 3.- - ., ..mt . 2 a.,.,-; # w- u. .o r. es, ..,.,x.m 3. . m .r m
._ os . - ..er.w. _ -ut f a, e ; A pe. ... .. r. a. . e. .u_A .u. a ew
.r.,,, Aa e n e r n. ,, a~A m..v ~n , a r 1
r
. p+v .r e r,a, "e e.. ._ ~.~ - n -c a
+ p .v o n. .; . mea s.
r e ,-
- m. - . = a" o r =. . ~a .d . . i . ', n u. e. . c . a c a.' r. .q .l a , <J e.n a r .i . u n o'.
.p. w: , ,".-
, %r ,a s _ .x. : , - a,,a a 1y ua
- , ue ,- e_. n g r _#- ,- A,A .a ,xg , .w ~. w w>, . - ., , , -~A ,. n e t.g c.3,a. sm ." 3 &~ ~4
- c
..+u .
1 t a,.,.d .n . _ w- A. . ~. ,,..5,, v a. v ; t e. 4 m o - 1 i - a,, 4 . ' . . o,r.~e . - a cu 4 ng .1 e- cae o- - m 4 r, n.. v. 4.m-a, e a,e e_, p.e.-o r c , ~ u..a .r o e~
,,Tec
~ a-r t' 3 e ayk. 3 v, u _;. .. . o c . a a_ n.
n - p .r G n .n...a A e .s.~.v a mnaa x m .-
- . e_ m -. - ~- . - -
are .; e...._,,~ - . . -.w_, w q. a '
4- - r o,,. t,d '.m. ' ..i .v e- - t"y _:- ;-
.. .- %.alr- n- a . . . o. age .
a- a .rw,_.r_, .;; pia _v.aca- e '~. u . . 3 sA
' ~
.re 1 .~ ~..+~a c a> -.+.,a +
. u..r.Le ~i r. d. ~ r.-a er .v.,ir e , - c parameter calues . ave been reviewed , nth respect to the assumptions
~ k
,m . e,d .4 . . . ~.a . - c a_ : ..,, a..' .,s,e v
- m. ta. ~..k.e a wv a Tyo u nss. e -v v .+.m A m 'u. gie 4.mm r. - . .o c o_ a n. a - > c.> - - .. ,
- m ,.e-r ..
. -m%.m n
o a_2, .m ~ . 3- t e.,
01 ...c ,mn os .i - ,.+ a ., ., v ., e , _ ha._ A,v # -
o.d r.wiv-s m..vm v:m ann ,
.. m,tm . o, . , uao u e wenn a~y er v , eg ,m., .w~ _a .r p .e x.ouce .. a: ~- a 1.o.
r .m> a w- -
- a. ~a .s.
-r, m A minor :nany has been made to the crerator action time value cf 120 seconds presented in the feedwater line break analysis, Secticn 3.4.2.4
- a. w. . w,.M e,,.,,. . im, . .c
-...s 2 ,- aQ 2 L. u V,6 e -o
~
O .r , . .r.a w ..,
Ym v b #4O.A .y b
,a ;
r u. .g' ,WV6 L .
. , 1. ,(4m, c.. .% v %64w M% }m *k W
- s. npU e a n e .e.ar w f *a
,--.m. _ em a" w, M k -= w,..n-m u w .S.
.e
..h mop 7 n ae 3 i .eo ).b m e .fe.gPhw o v.$o mC)y . u e . t h"* , , a y 38 f'.
. , f eL.k.a .v. 3 .' f s.2 :af .o n. .
%. . .; . i 3 .I .1 a ne, s
- ew e e. w. .a__ - e .v .4m~ a. r ., 3 . .m 1d, a: e.a ? ryo.s .
e, , c
- _ M
- m. ._v .. ,d n_y_i ,,. . . -
w m .u.a_ c
~ e 2 a_ n... e y . . u- ea,.,
... r. ;
e e ... e , ,, g .e 2 ,1.3 . . ,
_ , r m p , r%u A.rwepA
- m. .v . .q . r
- T V._ nA', e a p. w e ,-. b'3; ~0 q p e. ,.., .s r a .r., _ tv.>
z .Vs 2L
- v. w.e. .A. e V. 4 - , , v.:._ 4 _, - c. m, p e_. r w + . .4%a C2e.. + c e m .e..* Ai i..4n.
- a. ey.+. .,. e. .i..
in-nA 1".
Onr > b;
. v. u. .v p 3 - ..
m.u3 . n -. T M"
,'a6ane>pr _.
- :~ .
.s y2p., - . . eu b.o_ g..a y 14 h.
.y r d.en..w - i ,1. o 3--4A
.% . m.n e.
p s
a.c a l' yaa ,.~p e ae.~e..-...n_ +** . e- m, y. g 1' e a-aj - .r o - e G' F ya. + " i. '-
- 1r e g+r y..a P wmg, pe p.q.
* ,( . {A -
.v -n-e..2 , mm.A. a 4 L.s
- i ms .r. p -+. %. r ! .y%%.;
d e .r wa o . 4, - Di e q ;
-..r w . ..
..=
,2.
..*1
.r. N 6 .cJ> wn
.d .4 e.- e .j, 4 .%. . 3 %..a. p . ' *b b
-.a*a" .h -, ." *+ . y v.P a ."j a . a- p ... ..~ b~ . -,.,.',a"%
m.hD h M k wk abw a - a'. .P. . 2b , .
,s'"*, ,
agd ,=y, u..f.
us,'*.T'. 7a ,". A. 9' d'.
9*
y.. 9.' Q f. .y ,
.w'*M.
- r w a. .n.e,* ao m Apreri k -a_d 4 n.. u S t,p en p c -.r .* mr ..m.mp , % ,,- "vu}e p z . . o _ .r. ('mmy nm ta --
.k.av. r.y p .r r. a' rmA e.w,y s y t a.v .c, . -n ~-y- , n_ - *w ., , a p} , -
- < f. ".".a w u d' a s'" u a P ~
.u t'W' , N .' a " a.n"~ a . ' .
LL u. u
". "s r P.
Catawba . Mle ~ parameter values hwe also been reviewed nth respect t .s r, .%u. ;--,.
a o - g _. .
, a mA -a
... tM m
- ~-.'. ' a .' 4 "y' a- o" 3 ' '/' ~" '* a^ .
.a
.b. a Q F
. ,"48".9.*".*.4"39
- * *"9 wu 3 J
- '.',.h.#'"
.. a hT, 3 _. pM mmm ["' Q (
^
bu.b.Q .v' S 'M f*A
.9 is p
- j' p TF,. m ~f d'* a .D'- M
- .r Q.. 2'.
,8
. y "- Q.
,I ag hi g . - - ' ' -
.ty.
e e v.
.s = .
_ .e.
g e.- 4ea. e-,- . y $* t , g g +.
.. -,e - - .
>.C y
".,6
- $.- . = . . . 'ka-n q
. ~[41 . a . a.
-'.P.
7-1
1 product cere'incentories, changes in the thermal-hydraulic analysis results, as .e). as changes in the' dose analysis methodology.
. Reactor coolant pump shaft sel:ure (locr.ed rotor)
. Single rod withdrawal
. Ecd ejection
- . All of the abc*ce dose analyses are described in Section 3 of this report.
Cata'./ba 1 Cycle ' reload core physics parzaieters were f ound to be bounded by the a:cident analysis assumptions far all accidents which are sensitree t: core physics parameters, thus demonstrating conservatrze r esults f or the operation of Catawba 1. Cycle 7.
4 1
H 4
1 1
1 4
72
- . . .- - . -. - - . - - . - - - - . - .- _. _ - - ~ - - - . _ - ~ . _ ~ . . -
I 1 l B, s i :PCEED MD0!FICATICI;S TO L!CDICII.G EASIZ D<XL'MCITS 3
Revisi:n: t= the >chnical Speciticaticns and C;re Operating Limit s :
Repott RCLR! ? e onen prcptsed for Cycle 7 operaticn to acco:anedat a .
the inf.luence ci sne Cycle 7 core design en pcwer peaking reactivit'f.
an.: : sntrel . : d wn: ths. The Technical Specificaticn limits and COLR lit . ;s al v. r e1 R:t chrgen in reload analysis methcdology beginning with this mae. The Cycle design analysis tasis includes a low- l leakage t w rycle design and a mixed ccre containing both D&'d Mark-BW
- and tiestingn use CFA fuel assemblies.
A cycle crect ic power distributien analysis c: the final ccre design 7 was conduct ed to genet ste the f ( A1) limits for the C"/erpower AT and Overterperat ure AT trip f unctions and the Limiting C nditions f or Operati:n :::nt rol bank inserticn and axial flux di*f erence) . The f(AI) limit s ;aeserve the centerline fuel melt and steady-state DUER limits. The Limiting C ndit: ;ns for Operation preserve the raximum j
allcwable LDCA and initial condition Dt!B peaking limits, ejected rod worth reactivity limit::. and the shutdown margin reactivity limit.
These limits .fre devel ped hased on the IIRC-approved methodO1cgy descrited in Feference ~. A reaking penalty f or quadrant power tilt was taken in the analysis so that the resulting limits acccmmodate quadrant p;xer tilt ratios up to a value of 1.02.
The maximum allowable LOCA peaking limits sh in Figure 4 of the COLR are based cn the EWFC ECOS evaluation (Ref ert s 13 and 14). A composite F. C limit was-develtped based on botn large and small break analyses. Separate ccmpcsite limits applicable to Mark-BW and CFA fuel were used in the p;wer distribution analysis, and are specified in the COLR. These limits were used directly in determinaticn of the ccntrol rod inserticn and axial flux dif f erence cperating limits given in Technical Crecificaricns 3.1.3.6 and 3.2.1. Technical Specification 3.2.2 provides the nuclear heat flux hot channel (Fg) peaking limit.
The initial N ndition D:!B maximum allowable peaking (MAP) limits shown in Table 4 ot the COLR are based on core reierence design peaking factors. The MAP limits provide allowable combinatiens of peaking factors that preserve E!!ER performance equivalen" :o the design power distributien for a limiting loao of coolant flow transient. The
- initial condit i;n MAPS are used ac described in Ref erence 7 to calculate CI!E peaking margins for determinaticn of the control rod position and axial flux difference Operating limits given in Technical
, specifications 3.;.3.6 and 3.2.1. Technical Specifiscation 3.2.3 provides the nuclear enthalpy rise hot channel (F3 g) peak.sg limit.
i i
I The methodolcgy fer surveillance for eit core hoc channel peaking factors is described in Reference 7 .a this application of the methodology. peaking margin calculations are performed '.;henever an incore flux map ., taken for surveillan % monitoring.
Specifications . 2.2 and 4.2.3 have i -itten in a form that provides this cm *ility, and the pa -set ;s required by this application cf cure mcnitoring are p ' m in the COLR. The core operating limits are prcvided in the Jperating Limits Report, in accordance with ::RC Generic Letter 88 .6 and Technical Specification l 6.9.1.9. Table 3-1 lists the Technical Specification changes required for Cr ie ~ M
- nse :unces are identical r c *hcre ruhmity:d in the 8-I T TV' T i+;sw wvgw ae..pr w ,m ww ,,m,%em. 4 .
- I I
d I
'I 1
1 1
1-approved M: tire 2 Cycle 's :eload report (Reference 17). except those :
identified ty an aste11sk ir Table 8-1. Table 3-2 lists the changes to l the Core Crerating Limit: E+;crt. These changes are being surnitted to i the NRC under separate cover 2 Parameters related to renitcring the I core powel distributien are defined in Ref erence 7, and are used by the I plant cc:t:puter software. !!.:se parameters will be supplied for l inclusien in the COLR.
l ,
l Eased cn the an.11ysis and : e ;siens to the Technical Cpe:1ii:ations and COLR descrihed in this rept:- Cycle 7 of Catawba Unit I will crerate within the 10 FR 50.46 ECC3 1 ceptance criteria. The following pages !
COntain the reTalred Technt:1. 2pecificatien revisions and the revisicn3 t0 the COLR. l i
e 4
i i .
I e I
e 3
e i
4 i
h i
i R
T a
p i
1 k
s t
F
- g.
8-2
. -.- . ~ . - .. . _ _ . -_--. .- - . . - -..-.- - - . - - . - - - . .. . - -
r Table 8-1 Techaical Specification changes rt.ecificatien Ds crirtien of Chnnae ;
1 2.1.1 decreased FA g for Mark-BW fuel ,
2.2.1 decreased fag for Mark-BW fuel l removed power range neutron flux negative rate reactor trip
- removed Total Allowance, Z valum and Sensor Error terms 3.1.3.1 included all accident analyses that would require l reevaluation in the event that one full length RCCA is I ineperable 3/4.2.2 changed Fn mftthodology to reflect Duke nomenclature j quantified surveillance requirements j 3/4.2.3 changed F 3 g methodology to reflect Duke nomenclature l quantified surveillance requiremento c nect a n m requirement i 3/4.2.5 3/4.3.3.1 removed power range neutron flux negative rate reactor trip- '
removed items associated with RTD Bypass System 3/4.3.3.2 increased low steam line pressure r' point increased feedwater isolation respo.se time increased steam line isolation response time
- removed Tctal Allowaace, Z value, and Sensor Error terms removed items associated with RTD Bypass System remaved steam line pressure dynamic compensation 3/4.4.1.2 changed reactor coolant loop operation requirement t 2/4.4.2.1
- increased pressurizer safety valve lift setpoint tolerance 3/4,4.2.2
- increased pressurizer safety valve lift setroint tolerance 3/4.5.1.c changed required cold leg accumulator boron concentration 3/4.5.2 chan'ged ECCS pump surveillance re']uirements ,
3/4.6.2
- reduced allowable primary to secondary leakage rate 3/4.6.3 -
- changed feedwater isolation valve, main steam isolation valve, and main steam isolation tr/ pass valve stroke' time from 5 seconds to Not Applicable.
3/4.7.1.4 increased main steam line isolation valve stroke t.ime 6.9.1.9 reflected change to DPC core operating limit methodology
- The proposed Technical Specification change .was not . included in the
, _ approved McGuire 2 Cycle 8 reload report, Reference 17.
X-3 m e- --a-i:se 94+twa+w w.*w-iwww-r grt ge u~ eM-**vW t 4we--"'r u apqp6 ve-iemt-t.+ywm- 1-4
. .- .. . . . _ ., __. _ _ _ - . - _ . . . _ . . _ . _ . _ . _ . . . _ _ _ _ ~ . . _ . . . _ . . _ _ _ . . _ . .
5 Table 8-2 Core Operating Litits Report Changes crecificaticn Descrirticn of Ch 422 ,
3/4.1.3.5 revised safety bank insertion limits to reflect a minimum rod '
withdrawal limit of 222 steps and '
a maximum rod withdrawal limit of -
230 steps 3/4.1.3.6 revised control bank insertion limits to reflect a minimum rod withdrawal limit of 222 steps and a maximum rod withdrawal limit of 230 steps 3/4.2.1 revised AFD limits for Cycle 7 operation 3/4.2.2 revised for Cycle 7 operation to reflect a change in the heat flux hot enannel factor Fn methodolcoy 3/4.2.3 revised for Cycle 7 operation to reflect a change in the nuclear enthalpy rise hot channel factor Fg methodology y
n y
i 4
4 r
a l
I i
i 84 i
. . . . ~ _ . ~ . . . _ . _ _ _ . . . . _ . . . . _ _ _ . _ _ . _ . _ _ . _ _ . _ . . , _ . . . _ . _ _ _ _. ___. _ ._ _ _
J
I 9.1 Changes to Technical Specificaticns 8-5
o i A ,, ; G . i.
d i .~'
j N eS0i /
\ Unac:ecta:!e j s :: Cperatien j
- /
650 \ 2400 :s:a /
645-i N / ~
,/
- 2.250 :sia /
640- j/
- /
635, /
- /
630i /
C
/
- /
g8625 :oeg ;,:, ,
d *
/
H :
ca 620-0 -
l C: :
615 - l
/ I 610 / 1915 rs;a
/
- / -
605 - /
SOO-i /
- Ac,ce/
ptacle -
5952 Crperatien
/
i /
590 - /
- /
1 !
585.f/ 4 i
580'F O.0 0.2 0.4 0.6 0.8 to 12 Fraction of Rated Thermal Power - -
FIGURE 2,1-la REACTOR CORE SAFETY LIMIT 5 - FOUR LOOPS IN OPERATION, UNIT 1 CATAWBA - UNITS 1 & 2 2-2 -Amenoment-No-86-{Enit-&
h en e ent "c, 30 (Unit t
665 .
Q TOTAL FLOW = 385000 GPM 660h -
6555
- 2455 psia UNACCEPTABLE OPERATION 650 -
645 -
2400 psia
}
640 -
635 E N' 2280 paia 3
630 5 e -
O 2 E" 625 -
3 -
2100 paia D
620 -
615 2 u45 psia 610 605 -i 600 2 595 - (
4 ACCEPTABLE 500 h OPERATION Sai i_ ,
580 i i , . ,
0 0.2 0.4 0.6 08 1 1.2 Fraction of Rated Thermal Power FIGURE 2.1-la REACTOR CORE SAFETY LIMITS - FOUR LOOPS IN OPERATION, UNIT 1 K-7
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REAC"R TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlocks Setpoints shall be set consistent with the Trip Setpoint values snown in Table 2.2-1.
APPLICABILITY: As snown for each channel in Table 3.3-1 ACTION:
- a. With a Reactor Trip System Instrumentation or Interlock Setpoint
.less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value Column of Table 2.2-1, adjust the Setpoint consistent with the Trip 5etpoint value,
- b. With the Reactor Trip System Instrumentation or Interlock Setaint less conservative than the value shown in the Allowable Values column of Table 2.2-1, W beleb 1. Adjust the Setpoint consistent with the Trip Setcoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 was satisfied for the affected channel, or d
% NeclarethechannelinoperableandapplytheapplicableACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2-1 Z + R + 5 < TA
, }
6 Where: .
I= The value from Column I of Table 2.2-1 for the affected channel, R8 The "as measured" value (in percent span) of rack error for the affected channel, 5: Either the ":is measureo" value (in cercent sNn) of the sensor error, or the value from Column 5 (Sensor Errar) of Table 2.2-1 for the affected channel, and TA = The value f rom Column TA (Total Allowance) of Table 2.2-1 for the affected channel.
_/
CATAWBA - UNITS 1 & 2 2-3 "
OfJr T I TA8tE 2.2.-1 REACIOR TRIP SYSlEH INSTRl! MENTATION IRIP SEIP0lNIS V TENSORl E g' g 101AL A ERROR Id TRIP SETPOINT AtLOWABLE VALUE J
f (ll0WANCE TA) Z, (5) j7 fitNCTIONAL UNIT N.A. N.A. N.A.
Manual Reactor Trip N.A. N.A. '
g 1.
Power Range, Neutron flux 3 2.
- a. liigh Setpoint
}
I 7. 5 5.92 0 $109% of RIP * $110.9% of RIP
- 8.3 5.92 0 <25% of RIP * $27.1% of RIP' L- b. Iow 5etpoint 0.5 0 <5% of RTP* with <6.3% of RTP' with Power Range, Heutron Flux, 1.6 i time constant L i time constant liigh Positive Rate 3 2 seconds 3 2 seconds
$g4, _
0.5 0
<5% of RTP" with
<6.3% of RIP
- with
! 1. Power Range, Heutron flux. I.6 i time constant i time constant -
liigh Negative Rate ~ -
732 seconds 32 seconds .
<31% of RIP
- Intermediate Range, 17.0 8.4 0 $25% of RIP
- 1F p Neutron flux
" 0 $105 cps $1.4 x 105 cps 54 Source Range, Neutron flux 17.0 10 2.12 for See Hate 1 See Note 2 61 Overtemperature of 6.98 for 3.ts (or Unit i and linit I and Unit I and 8.9 for 7.3 for 2.7 for e lini t 2 Unit 2 Uni $ 2 1.7 See Note 3 See Note 4 Overpower AT 4.9 1.24 74 31938 psig***
-a 4 B g. 4.0 2.21 1. 5 31945 psig Pressurizer Pressure-Low 0.71 0.5 $2385 psig $2399 psig
?)d. Pressurizer Pressure-liigh 7. 5 5.0 2.18 1. 5 (92% of instrument <93.8% of instrument S K Pressurizer Water level-liigh span span
, {# / >90% of loop >88.9% of loop 2.92 1.48 0.6 f /lR. Reactor Coolant Flos-Low / Einimem measured sinimum measured flov'
%fo flow **
j I ea2
- RTP = RATED Ti1ERMAL POWER
- loop minimum measured flow = 96,900 gpm (Unit 2), 96,250 gpa (Unit 1) h"e *** lime constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 2 seconds for lead lr and I second for lag. Channel calibration shall ensure that these time constants are adjusted to these values.
t}SJET \
9.-4 lABLE 2.2-1 (Continued) g REAC10R IRIP SYSTEH INSTRUMENTATION TRIP SEIP0lNIS I ~DeleN TOTAL SENSOR '
c All0WANCE ERROR if FUNCTIONAL UNIT (TA) Z (S) IRIP SETPOINT All0WABLE VALUE d
e 12)T. Steam Generator Water e.
m tevel Eow-Low
- a. Unit 1 17 14.2 1.5 >l1% of span >15. 3% of span f roin from 0% to 30% 0% to 30% RTP*
- increasing increasing linearly linearly to to >38.3% of spar.
> 40.0% of span from 30% to 100% RIP
- from 30% to 100%
b RIP
- 7
- b. Unit 2 11.8 1.7 2.0 >36.8% of narrow >35.1% of narrow range span range span l
l $34'. Undervoltage - Reactor 8.57 0 1. 0 >7/% of bus >/6% (5016 volts)
E Coolant Pumps voltage (5082 volts) with a g 0.1s response time g / 4A5. Underfrequency - Reactor 4.0 0 1. 0 > 56. 4 11: with a ->55.9 Itz a
3 Coolant Pumps 0.2s response tire s $ /f16. Iurbine Trip h a. Stop Vaive Eil N.A. N.A. N.A. ->550 psig >500 psig
?? Pressure Low
- b. Turbine Stop Valve H.A.
"hZ o
Closure N.A. H.A. ->1% open ~>l% open c
- 23. /6)K Safety Injection Input t N.A. N.A. N.A. N.A.
o
} from ESF 1 H. A. -
- EIP = RAl[D THfRMAt POWI1:
~. . . _ _ _ _ _ - _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ -
v 4.I T l g TABLE 2.2-1 (Continued)
Y RFACIOR 1 RIP 5YSTEM INSTRLIMENTA110N 1 RIP SLIPOINIS 7 Deleke 10TAL SENSOR '
ALtOWANCE ERROR E fuNCIl0NAL tlNIT (lA) Z_ {S) TRIP SEIP0lNT All0WABLE VAtilE
[ 17)tf. Reactor Trip System
, Interlocks
" 36 x 10 88 amps
- a. Intermediate Range H.A. H.A. H.A. 31 x 10 80 amps Neutron Flux, P-6
- b. tow Power Reactor Trips Block, P-7 I) P-10 input N.A. N.A. N.A. $10% of RTP* <12.2% of RIP *
- 2) P-13 input N.A. N.A. N.A. (10% RIP
- lurbine <12.2% RIP
- Iurbine Impulse Pressure Impulse Pressure d Equivalent Equivalent ;
4
- c. Powr Range Neutron N.A. N.A. N.A. 548% of RIP
- 150.2% of RIP
- Flux, P-8 7 d. Power Range Neutron N.A. N.A. H.A. 169% of RIP
- 1 10% of RIP
- C Flux, P-9
- e. Power Range Neutron N.A. N.A. H.A 310% of RIP
- 37.8% of RIP
- Flux, Not P-10
- Iurbine qp zdz Pressure, P-13 Impulse Pressure Impulse Pressure Pl? Equivalent Equivalent a% sb17. Reactor Trip Breakers H.A. N.A. N.A N.A. H.A.
q s
9 N.A.
Th 20' Automat.ic Trip and N.A. N.A. N.A. N.A.
$$ Interlock Logic bd M *liiF = RATED INERMAL POWER ' N ,- l
~ . . _ . . , _ _ _
U BJ c T I k TABLE 2.2-1 (Cantinued}
TABLE NOTAllONS s
NOTE 1: OVERIEMPERATURE AT aT y (n
I (y , ,sg) $ AT, (K i - Kz [T(3f,,3)-T']+K(P-P')-f(a!))
3 i a
g Where: a1 = Heasured AT by Loop Harrow Ranga RIDS; l
" 1 + t'5 m p,,g= lead-lag compensator on measured al; ti, 12 = Time constants utilized in lead-lag compensator for al, 1 3 = 12 s. l 12 = 3 s;
~
l y , g,3 = Lag compensator on measured AT; 13
= Time constant utilized in the lag compensator for AT, 13 = 0; 4 al, = Indicated AT at RATED THERMAL POWER;
JA8';1.1953 Ki l C K2 = A 02401/*F; o.o 3163 I*
r g, The function generated by the lead-lag compensator for T,yg jp dynamic compenshtlon;
}*
14, is =
Time constants utilized in the lead-lag compensator for I,yg, ts = 4 s;
- t. = 22 s, [
g'y T = Average temperature. *f; gm y, 3
=
Lag compensator on measured I,yg; E '
=
Is TimeconstantutijizedinthemeasuredI,yg lag compensator, is = 0;
~~
Y
V M.T T \
~>
k IABIE 2.2-1 TABLE NOTATI5ii(Continued) s (Continued) !
5 1
, NolE 1: (Continued)
I C
$ l' $ 590.8"E (Nominal T, allowed by Safety Analysis); '
- d y K ., = -0:00Ii89; o. oo 34i4 i
e.
m P = Pressurizer pressure, psig;
- . P' = 2235 psig (Nominal RCS operating pressure);
5 = Laplace transform operator, s 8;
! and f (al) is a function of the indicated difference between top and bottom detectors of the i
- 1. power-range neutron ton chambers; with gains to be selected based on measured instrument a response during plant SIARIUP tests such that:
> - 39. 9 ~/. t 10 % '
- j. ry (i) for q t
gbbetween -J.2di% and .-f($%,
f,(al) = 0, where q and q are percent RA M IH N N in the top and bottom b
- p halves of the core re pectively, and qg*gb is total IllERHAL POWER in percent of RAIED IllERHAL' POWER; AI -319 %
, (ii) for each percentAthat the magnitude of qt gbis a re negative than _22rST, the L >,
gg alTripSetpointshallbeautomaticallyreducedbyMSI%offugueatRAIEB 414RMAt-POWER; and 3A IO A hE g! Ar u.o %
oo (iii) for each percentAthat the magnitude of q t 9b is more p sitive than -Je5%, the AI Irip
- . Setpoint shall be automatically reduced by L6(I% for-4) nit-i-end,2-4t44-for-Unit-? l o of its vsluc ei. RATEB-fttfRMAt-POWER. L 3t(o 7.
n
' Y o b3, NOIE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by l
, more than 3.0% (ec-48ait I r.d L3% for Unit 2c
' I A111E 2.2-1 (Coatinued}
c, I ABITloTATINi5 (continued 1 1401E 3: OVERPOWER AI I I 1 ' K li l i" i A^ }i
[ AC (l_'__!U ) ( 4
-Ks(([i i 1[S ** S I
= (1 + 12 S) Il ' la bI < Al" (K Y
Where: AI = As detined in flote 1
= As delined in tiote 1,
, {
si, 1 2
- As defined in flote 1 l l
I 3
, g - As rief ined in ilot e 1, i i
.d i3
- As defined iritiote 1, 7
u>
AI " = As defined in flote 1, l ? I E l K4 - Lfdet~ i.081*1 l
- 0.02/"f f or increasing average temperature and 0 for decreasirig average K:,
l temperature, dyriamic Ilie f unction generated by the rate-lag controller f or I, g )* g =
compensation,
,1 7 = 10 s.
= I!me constant utilized in ti.e rate-laij controller for I i7 I = As detined in tiote 1, p, g As detined in t{ote 1,
=
is
- _ - _ _ l
1 1
I i
T y
I l a
b N r o i t
n U f o p
e t s e is S t
a p r i e r
, pr T
f a
- e d 8 t e t
- 0. n u 9 o p 5 i m
t o
<- a c r
I b s i t r l i o a f C d
( e 0 e R c
= E x
} W e d s O
)e K P t d u o e n d L n -
u i n A n t a l F l i n R , l t o F E )
a n c ll f h o ( "8 I
- d s C 8 n (S 0 D .
a t n
N 9 E 0 1 O 5 , T 9 , i
- I 1 A 5 1 o 2 T > R p A e < e t 2 T T t L t e O o d ,
o S E N r N n N L o o p B E f n i n i A L i t i l r TB F I a b T A
- d t d -
i/ e d n e m 1
P n e e n H u MCr l
7 i
t e
t a u c r m i f
e e
r
- i m
x o0 d it d o a c0 d s f m .
_ . : s n n s - % ;
o0 A I i A 0 s8 l 2
= = = = - e
_ nn na
_ - ah
- I 4 ht
_ c e
_ d(M er h o
) Q I
I 1
S mt e i T m d
e u Ac
_ i n bf t
n o
A C
(
de s :
3 w 4 E T_ E T T O O N N n%$$ , 5t p ;wm .
hNI $
~., a[*Ce? _ .
h eg9 e. "
^
. _ - - . _ _ = _ . . ~ _ - _ - - - . _ - -
. - .. - - - - . . . . . . ~ . . . _ _ _ - . 1 i
I l
Attachment. I and f,(AI) is a luncticn of the indicated dif f eren:e between top and bottcm detectcrs Of a power-range neut:an icn chimcers; with gains to I
be selected ::ned on .ceasured instrument respence i; ring plant startup tests such that:
s (i) for q, - as between -35% and +35% AI: f2(AI: = 0, where qr and ,
qq are percehr TGTED THEiyAL POWER in the tcp and c:ttom halves df the I cbre respectively, and q, + qs ' '
is total THEFFAL PC'.O in percent of FATED THERMAL F 7/ER: l (ii) for each percent AI that the magnitude of q. + gb i8 *CT*
negative than -35% AI, the AT Trip Setpoint shall te automatically reduced by 7.01 cf AT , and i 4 7 11111 for each percent AI that the magnitude of q. - gb iS *C l
- posi;ive than + 3 5 % AI, the AT Trip Setpoint shall re autcmatically reduced by 7.C't Of AT 1
I I.
l l
l l-t l- i i
P X 16
tJrJ I T Z.
IABIE 2.2.-l REACIOR TRIP SYSIEH INSTRilHENTAT1014 IRIP SEIPOINIS 1
U f,101AL SENSOR -
$5 All0WANCE ERROR M #E littlC110taAl llNIT {IA) 2 (S)
IRIP SEIPolNI All0WABL E VAlllE 7
g i Hanual Reactor 1 rip N.A. N.A. N.A. N.A. fl . A .
] c. Powei itange, tieutroa flux
" <5% of RTP* with <6.3% of RIP
- with
- i. Power Range, Neutron fiux, 1.6 0.5 0 liigh Positive Rate i time constant i time constant 1 2 seconds 3 2 seconds
- with liigh Hegative Rate a time constant a time constant 32 seconds 32 securnis S Intermediate Range, 17.0 8.4 0 $25% of RIP" $31% of RIPa CL Neutron flux 6 Source Range, Neutron flux 17.0 10 0 $105 cps $1.4 x 105 cps
/ Overtemperature AI 6.98 for 3.0 for 2.12 for See Note 1 See Note 2 x Unit I and Unit I and Unit I and.
6 8.9 for 7.3 for 2.7 for Unit 2 Unit 2 Unit 2
- 11. Overpower al 4.9 1.24 1. 7 See Note 3 See Note 4 bf 9 Pressurizer Pressure-tow 4.0 2.?! 1.5 31945 psig 31938 psig***
j 10 Pressurizer Pressure-fligh 7. 5 0.71 0. 5 12385 psig $2399 psig y 11 Pressurizer Water lesel-liigh 5.0 2.18 1.5 192% of instrument $93.8% of instrument span span PP 12 Reactor Coolant flow-Low .
2.92 1.48 0.6 >90% of loop >e8.9% of loop e o, inialmum measured minimum measured ilow' flow **
e-d% ~^ KIP = RATED TilERHAL POWER o .h ** loop minimum measurett flow = 96,900 gpm (Unit 2), 96,250 Ops (Unit 1) 7 *** lime constants utilized in the lead-lag controller for Pressurizer Pressure-low are 2 seconds for lead and I second for lag. Channel calibration shall ensure that these time constants are adjusted to these values.
LJ A1.I T Z_
Q -
lABLE 2.?-1 (Continuedl
-4 g REACIOR 1 RIP SYSTEM INSTRUMENTATION TRIP SEIPOINIS bl01AL SENSOR
m
... 13. Steam Generator Water
, L~ >
m level l ow-Low
- a. linit 1 17 14.2 1.5 >l7% of span >15.3% of span trorii from 0% to 30% 0% to 301 RTP" RIP" increasing increasing linearly linearly to to >38.3% of span
> 40.0% of span from 30% to 1001 HIP" from 30% to 100%
@ RIP *
'Y
- b. Unit 2 11.8 1.7 2.0 >36.8% of narrow 335.1% of narrow range span range span J4. Undervoltage - Reactor 8.57 0 1.0 >77% of bus ~>76% (5016 volts) y Coolant Pumps voltage (5082 volts) with a 0.?s response time
[ 15. underfrequency - Reactor 4.0 0 1.0 >$6.4 Itz with a ~>55.9 Itz Coolant Pumps 0.2s response time
'"g 7
'3 16. lurbine Irip
- a. Stop Valve Eli N.A. ft. A. N.A. >550 psig >500 psig
- _ . Prer are Low
.os b. lurbine Stop Valve N.A. N.A. N.A. 31% open gl% open lZ Clnsure <
- c. '
- l. 3. II. Safety injection input N.* N.A. N.A. N.A. N.A.
'f from ESF
' 3 $' _
- RIP = RAlf D Iilf t:t4Al POWil:
~ . -. _ _ - - _ - - - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _
t> AJ3 T g lABLE 2.2-1 (Continued)
Y REACIOR 1 RIP SYSTEM INSTRtiMENTATION TRIP SEIP0lNIS ai I l01AL SENSOR h'
7 -
Q_-he1 All0WANCE ERROR I i E I tit 4C110NAL UNIT I QA) Z (5)_ ! IRIP SEiPalHI At t OWADt E VAltil
[, 1a. weact or Trip Syst. .c- j ;
Interlocks
" H.A. N.A. H.A. 31 x 10 ain amps >6 x 10 88 amps
- a. Intermeiliate lange l Neutron Flux, P-6
- b. Low Power Reai tor Trips Block, P-7 ,
- 1) P-10 input N.A. N.A. H.A. $10% of RIP * $12.2% of HIP *
- 2) P-13 input N.A. N.A. N.A. <10% RIP
- lurbine (12.2% HIF* lurbine r g Impulse Pressure impulse Pressure Equivalent Equivalent '
} c. Power Range Neutron N.A. N.A. ti. A. 148% of RIP * $50.2% of RIP ^
Flux, P-8 ti. Power Range th utron ti. A. it. A. N.A. yb'J% of RIP" 1/0% of HIP ^
I
< Flux, P ') !,
Power Range flcutron H.A. H.f. h.A 310% of RIP" >7.8% of RIP *
- e. _
py Flux, P-10 (12.2% of RIP
- Power Range Niistron N.A. H.A. N.A. $10% of HIP"
$$ t.
Flux, Not P-10
) ,ili
- Turbine impul e Chamber N.A. H.A. H.A. <10% RIP
- Turbine <12.2% RIP
- Iurbine g.
N P */
Pressure, P-13 Impulse Pressure Equivalent Impulse Pressure Equivalent Reactor Irip Breakers N.A. N.A. N.A N.A. N.A.
U l 'J .
ew H.A. N.A. N.A.
[E 20. Automatic Trip anil Interlock iogic N.A. N.A.
'r *
~
- HIP = RAILD IHERMAL 6%Wi.R
~
tJ AIIT L
.k' Continued)
IABLE2.'2-)(ITTons TA8tt TeD: ;
.g -
1
>- NOTE 1: OVERIEM* ERA 10RE AT G
ai g.
I (3 , tsg) 5 AT, (K -K 2 h U (1 1,$) - I']' + K3 (P - P') - f (u!)1 ,
[ Where: 41 = Measured AT by Loop Harrow Range RIDS; [
" 1 :S !
u p ,[g = Lead-lag compensator on measured AT; i
t ti. 22
= Time constants utilized in lead-lag covensator for AT, ta = 1r s, j r2 = 3 s; ;
- i 1
y ,- g = Lag compensator on measured AT; !
= Time constant utilized in the lag compensator for AT, r3 = 0;
& t3 !
3 AT, = Indicated AT at RATED THERMAL POWER; ;
Ki = 1.38;. l l x
b = 0.02401/*F; K2 f
=
f The function generated by the lead-lag compensator for T,yg dynamic compensation; i
= t, = 22 s, 14, 13 Time constants utilized in the lead-lag compentator for T,yg, l
, is = 4 s; 7" Average temperature. *f; I gy T =
g 3, 3
=
Lag compensator on measured I,, ;
e-gg =
. t
- Is ' Time constant utilized in the ocasured I,,g . leg compensator, is = 0; }
I w _ b
t h >_r T 2_
n
$ IABLE 2.2-1 (Continued)
@ IABLE NOTATIONS (Continued)
E
, NOTE 1: (Continued)
C l'
$ 590.8"F (Nominal I,yg allowed by Safety Analysisj; K3 = 0.001169; g
e.
P = Pressurizer pressure, psig; y
P' = 2235 psig (Nominal RCS operating pressure),
5 = laplace transform operator, s 8; and f (al) is a function of the indicated difference between top and bottom detectors of the power-range neutron len chambers; with gains to be selected based on measured instrument response during plant SIAP.IUP tests such that:
CP 7
to (i) for q t
q between -22.5% and -6.5%,
b f:(al) = 0, where q andt q areb percent RATED TilERMAL POWER in the top and bottom x halves af the core respectively, and q t *U b is total IllERHAL POWER it percent of RATED lilERMAi_ POWER; bi w
(ii) for each percent 4that the magnitude of qt gb is a re negative than -22.5%, the lwl j al Trip Setpoint shall be automatically redxed by 3.151% of 4twalue at-RAff&
g' -fttERMAL TGWER, and aT
- - Af 3f (iii) for each percentAthat the magnitude of q gb is more p sitive than -6.5%, the al Irip NI Setpoint shall be automat ically reduced by 1.St1%-fee-BMt ! =d 2.414% hr 'Jnit 2- l d of its vahc at-RATTtB-THERMAF-POWER.
2I 4c "o
13 NOTE 2: Ihe channel's maxistna Trip Setpoint shall not exceed its computed Trip Setpoint by l Ay more than 3. M fer W it I eat,1.3% e r Pait 2.
L Q
z
)
I eg .
T I O
a r
3 I I e 0 _
2 v c 1 _
N I a i m = _
U ~ j s
i s
a i
n I
i y i _
s d _
i a ,
e _
~ r _
)) c _
e 1 I 5, d -
. r r t
r o o 1
- f o f f r r
((
1 0 l e l e
di l l I
s o o I
a r -
i t t
' e r i i r
o r o u c c
~ t a g g
' r a a
)) e l l p - -
) _5, m e e d . e t t
}e 1 t a a d s i r r e n l
- e u i 0 e e n t ((
1 a h h i n r t t t o e nc )) v y n o( 5 a b i C
(s 3 g d d H 1
, , , , n e e , ,
16 S , ,
1 1 1 i t z 1 1
- T z* 1 1 1 s a i e e 2T .A
((
O'I l e t e t e l e l e t e a e
r e
n l
i t a
t o
2T o o o o o o r e
t u H H e H i N H H N c E l
~n 5 t n g n n t
l ~f l n n r, n i n i a i
, nn
, t n i i Af i i i i re oo a 18 d d or ii t d d A
- d d d d sn e e e e e e e e f u tt n n 1
n n n n n n t ca i i i i i , f a ns o i i o
i t
i t f t t t 4
- r un c t e
t e
e e e e e e 0 /e f e d d d d / 2 p p e d d d d 0 0 m em m s s
.I s s s s s s . . e ho i A A A A A A A 1 0t I c l A
))
$ = = = = = ' =
Si = = = = =
g g 1
- f i 2 1 ' , 1 I 2
I.
t
((
))
' >s I
A t
i
, I ,
, 1 3 I A
K K
3 g n i
g g
g i2 R t 1 C
W O
P 1 1 e H (( r C e V l h O a W 3
L 1
_ 02 1
P Cc f '&[3 hDE$E c 3 w H s. m
- 1 1
y l b a t
.. r n 2 f o i o
T e t p
r e T. u s t
M a p ,
O r i e r
, p I I m
" e d 8 t e
. t 0 n u 9 o p 5 i m t o
- a c r
I b s i t r l i o a f C d
( e 0 e R c
= E x
}
d .
W O
e
) e K P t d u o e n d L n u i n A n t a i t l i n R , l t o l E ) a nC
- l l f h o( 8 I d s C 8 n (S 0 D a t N 9 E 0 n 1 O 5 ,
T 9 , i
- I 1 A 5 1 o 2 T > R p
. A e 5 e t 2T 1 t t t e 6 o a ,
a s EH r l f n !
o o p Br L
Ai f
i n f.
t i n l i
r IB ~
I I a A T A " d t d 1 / e d n e l m
/ n e e n l u 0 i t m i a a.
7 f a u 'f i 1 c c r e r x 0 !
( i t d o a 0 d s f 0
A s
I n n i A s
0 mL s8 i
- = = = =
'2 l
e nn na ah
) ht l c
(
a er e
u " .
h ia
) h I 1 5 t l n d
e u
i n
t n
o C
(
3 4 E E T T O O N N
@N - :13 WP R
<$ pip o: , k AT
~
is .
f si !
g83& 7*
?sh lflllilll !l ill ill l.1l iI 1 1l 1 !Il !l ii
2.1 SAFETv L:u!*.:.
BASE 5 2.1.1 REACTCt ::EE '::R UNIT 11 i The restri:- :ns :f this Safety Limit revent overneating of the fuei anc
- ossible cis=ir; :e-f: ration nica woulc resuit in the release of fissicn crocucts t: ne ren ::r c:clant. Overneating of the fuel :',accing is crevente y restricting .e ::eration to witnin tne nucleate coiling regime wnere tne heat transfer ::ef e:ient is large anc the claccing surface temoerature is slightly acove : e :::lant saturation temcerature. .
Operati:n a::ve ne u:cer :oundary r' the nucleate coiling regime c:uld result in excess've :! adding temceratures because of the onset Of cecarture from nucleate ::: 'r; (DNB) anc ce resultant snaro recuction in neat transfer
- efficient. :NB is not a directly measuracle carameter curing operation anc therefore THERMAL :CWBR and Reactor Coolant Temeerature anc Pressure nave teen i related to ONB :nreu; . the BWCMV correlation. The BWCMV CNB correlation nas i been ceveloce: t: :recict the CNB flux and the location of CNB for axially uniform anc nonunif:r : neat flux distributions. The local CNB heat flux ratio, (DNBR), is cefinec as the ratio of the neat flux that -ould cause CNB at a particular c:re 'nati:n to tne local heat flux, anc is indicative of the margin to DNB.
The ONB cesign casis is as fo11cws: there must te at least a 95%
procability that ce minimum ONBR of the limiting roc during Condition I and II events is greater tnan or ecual to the ONBR limit of the ONB correlation being used (the BWCMV correlation in this application). The correlation DNBR limit is estaclisnec :ased on the entire applicaole experimental data set sucn that there is a 9% orecability with 95% confidence that CNB will not occur when the minimum CNBR is at the CNBR limit.
In meeting =is :esign basis, uncertainties in plant 0: erat ~ 4 l
nuclear anc :nermai :arameters, fuel f abrication parameters, anch' para::eters, the BWCMV l l
CNB correlati:n are ::nsidered statistically suen that there is at least a 9B% 1 l c nfidence that tne minimum DNBR for tne limiting roc is greater than or equal i to the DNBR limit. ~he uncertainties in the acove para::eters are used to ceter-
! mine the plant CNBR .ncertainty. This CNBR uncertainty is used to establisn a cesign DNBR value -nic . must ce met in plant safety analyses using values of input parameters incut uncertainties.
The curves of F'gure 2.1-1 show the loci of points of THERMAL POWER, Seact:r C: lant E/n: :rassure =c :ver:;; tam:eratur: M'.:w -ni:h the calculated CNBR is no less than the design DNBR value, or-tne average entnalcy
[
at the vessel exit is less than the enthalpy of saturated liquid.
0 24 CATAWBA - UNITS 1 L 2 B 2-1 Jaenoment-Nores-(4Jni t-1) laenomens-No. :: (UM4
__ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~ _ - __ . ._ _.
2.1 SAFETY u:M
5ASE5 These curves a e :ases :n 3 uele$- "m'm -'se not :nanrei '3:::e
- ),H, of f.' .
- r .e s . , r.,;nc as e C::m :~ ' ue ! 4 se-rlies t W s i m.55 fcf
.I -
c e - m. car (-5,
- vei Assem iies an a reference c: sine witn a ean. :( 1.55
- - !:- : * - N --
a ...ance is inclucea f:r an increase in F at 4 ecucac power :2sa: :n tne ex:ress;:n:
g n ,so Chg F
g = ,,J M : '. - f.Ar* ( 1- : ) ] rr
- n: a t-im r : n c m's .-
_ m
. N F
q = 1. '5 :'. - 0. 3 (1-4) :or the BWFC Marx-5W's Where P 4: tne f racti:n of RATED THERMAL F0WER.
RRB w e a totR These limit ng ntat flux c:ncitions are nigner than tnese calculated for tne range of all ::ntr:1 rc:s fully witncrawn to the maximum allewaole control rea insertion assuming the axial p wer imoalance is within :ne limits of the ft (AI) functi:n of tne Overte cerature aT trip. When the axial ;cwer imbalance is not within :ne ::lerance, tne axial power imoalance effect :n the Over-temoerature ai t-1:s will recuce tne Setpoints to provice protection consistent with core Safet, _imits.
2.1.1 REACTOR C:RE ( 0R UNIT 21 The restrictions of this Safety Limit prevent overneating of the fuel anc possible claccing perforation wnica would result in the release of fission
- rocucts to the eact:r coolant. Overneating of the fuel claccing is preventec ey restricting 'uei ::erati:n to witnin the nucleate : oiling regime wnere tne heat transfer : efficient is large and the cladding surface temperature is sligntly aoove : .e c:ciant saturation temperature.
Operation a::ve tne uc;er : uncary of the nucleate boiling regime could result in excess;ve cladding temperatures because of the onset of decarture from nucleate ::: ling (CNB) anc the resultant snarp recuction in nYat transfer
- efficient. :NB 's not a cirectly measuracle parameter caring ::eration anc inerefore THERMAL .:CWER anc Ceacter Coolant Temperature anc Pressure nave been related ts ONB trrougn the WRB-1 correlation. The WRB-1 CNB correlation nas
- een develocec : crecict the DNB flux and the lccation of CNB fcr axially uniform and nonunif:rm neat " lux cistributiens. The local DNB
- ant flux ratio, (DNBR), is cen nec as tne rat 1 of tne neat flux *.nat would cause CNB at a particular core lccation to the local heat flux, anc is incicative of the
- argin to ONS.
The ONB design basis is as follows: there must be at least a 95%
procability that tne minimum DNBR of the limiting red during Condition I and II events is greater inan or ecual to the DNBR limit of the ONB correlation teing used (the ,RB-1 correlation in this application). The correlation DNBR limit is establisned based on the entire applicable experimental data set sucn that there is a 35% crocability witn 95% confidence that CNB will not occur en tne min t- 3ER is a*. : e ONBD limit.
4 CATAWBA - UNITS 1 1 2 B 2-2
- hnc:ent-No. :: (4Jni t-l+-
5 nrent 4:. E0 (Uni W
11J:MITING 9: 7 SYSTEM S~~~: N35 jASES y..'.'
REACTCR ~:'.D "5T. M.
'..NST P.:NTATION
(. 7. :". .17 5 valueseatReact wnicn 0r -ic tri;s are Set:oint setThe Limits specifiecReact:r in Table 2 21 .
Set oints nave :een se u ctec :: for eaca functicnal unit.are tne nominal ;
System are prevented f ! ,. c. ensure inat tne c:re ano React:r C:olantthe Trip i anc cesign casts anticipatec coerational eccurrences ano to1xceecir; eration :neir neerec Safety eatures Actuatica System in mitigating th accicents. assist tne Engi-e c:nsecuences of considerec to te acjustec c0nsistent with the nominal value Teasurec" Setco: : 0ck function is wnen tne *as is within the canc allowec for calibration y. accurac -
tests and the ac:uracy to wnich Seteoints etweencan beTo acc:mme ccerational Allowable 2.2-1. Values ':r ne Reacter Tric Setcoints have bmeasurec anc calibratec, but within tneithA:'c-aole Setooints less value is acceotacle conservative than sinceOperat the r o Set:cint T ieen scecifiec en in Ta an allowance nas teen g j<J_
{Setcointmacs in the safe:v 1as ceen iiclucec for cetermining tne OPTtABILITT analysis to . acc:mmocate n cotional provision this
- trror] A .
is founc to exceed the A f a channel or,en its Tric ) <
cotion utilizes :ne "as measurec"llowaole Value. The methoJology of this point for rack anc sensor ccmconents in conjunction with adeviation frem the tion of the variable anc other the uncertainties uncertainties in of thethe calibrating instrumentati statistical c:moina-instrumon to measure the process tion 2.2-1, Z + 9 + 5 3, TA, the inte entation. In Ecua-and the senso , and the "as measurec"ractive ef fects of tne errors in the rack as specified in Tasle 2.2-1, in percent span, values of the errors are c:nsidered. I, rack drift and tne accuracy of their measurementerrors the uifference, e sensor anc assumec E
in the tion, analysis f:r Reacter trio.in cercent span, tetween t'iovalTrip Set:oint anc theta or point. in percent s:an, for the affected enannel1 from or Rack Error is the "as measurec"ue usec cevia-from its calibra:::n cint or the value deviation scan, from the analysis assumptions.
specified of the i5 or Sens:r Error i sensor sensor crift f act:r, n Table 2.2-1, in percent kvalue for REPORTABLE EVENTS.an increasec racx crift factor,Use of Ecuation anc provices a tnresnoic The methocology : ~~d of the uncertainties in the cnannels.cerive !nherent one Trip Setcoints is based ucon c:mcining Trip setpoints are :ne magnituces of these enannel uncto thethe cetermination of ertainties. Sensors anc operating within :ne allowances of these ce cacaeleuncertainty of ma in excess of the Allowaole Value exnibits me' its allowance. a Rack benavior the crift th t the racx has not i happen, an infrecuent excessive drift is expectedBeing that there i in excess of the allowance that is more Rack than oroccasional sensor crift, more serious preolems and snould warrant further investigati of , may be on.
indicative CATA'dBA - UNITS 1 L 2 3 2-3 *26
_ -E
LIMITING SAFETY SYSTEM SETTINGS BASFS REACTOR TRIP SYSTEM INSTRUMENTATIO breakers whenever a conditionThe various Reactor preset or provides monitored calculated level.
design approach aR by the Reactor Trip rp Ss aatomat capability at the In addition specified eactor Trip System to redundant tripvariables, which monitors s and trains, channel ystem reaches a therefore the prov diverse Reactor trips for which r p System functional diversit numerous
- y. system analysis to enhance the setting overall is r lirequired for those antiThe functional Reactor initiated Trip System initiatese no direct credit was cipatory assumed or in th from exces. This prevents the reactivity i ability of the Reactor a Turbine trip signal wheneverem.R The Trip e Syst acci dent actuation sive of Reactor the Engineered Coolant Safety System nsertion F cooldo that would otherwiseeactor trip is wn and thus result Manual Reactor Trio eatures Actuation System. avoids unnecessary The Reactor Trip System incl d P,ower Rance, Neutron Fig u es manual Reactor trip capabili ty.
( bistables, each with its own t iIn each of the Power Rangt setting.
i power operations tor mitigate p sattingthThe usedLow for aSetpointHigh and o independent tripLron Flux ch provides from low power, and gthe Hi h S power levels. to mitigate the e opera'tions consconsequences of a powerandexcursiprotection low etpoint trip provides protection don beginning equences of a reactivity excursi uring power The Low Setpoint trip may be on from all of approximately below the P-10 Setpoint10% of RATED THER MAL POWER) andpower is automaticallmanu level Power Range, Neutron Flux y reinstated
, Hich Rates increases Specifically which are characte iThe Power Range Positive Rat trips to ensu,re that ther crithis stic of tripacomplements rupture of the o a contr P rapidleflux s trip provides pro drop accidents.The Power teria Range are metNegative for all rod Rate ejower t Range ow Neutron Flux High ection accidents.
Negative Rate trip will preg.eaking which could cause anAt No unconservative local DNBR toent could cause local flu those control rod drop vent accidcredit this from occurring is by taken tri exist, pping for the operationPower reactor.The of thRange applicable design limit DNBR ents val foreachwhich DNBRs will beorgreate Power Rang k Qo w e( ue for fuel type.
daeAes w una k {\)e er tnan the i G e. N L W Q h05 bu CATAWBA - UNITS 1 & 2 B 2-4
1 1
TABLE 3.1-1 ACCIDENT ANALYSES REOUIRING REEVALVATION IN THE EVENT OF AN INOPERABLE Ful'.-LENGTH R00
\
l Rod Cluster Control Assembly Insertion Characteristi
- Mi l RodClusterControlAssemblyM4ASkr.4;cn 9nment Loss of Reactor Coolant from Small Rupturea Pipes or from Cracks in Large Pipes Which Actuates the Emergency Core Cooling System ,
- 01 91 Rud Cluner Centrol Asse=bly withdrawal-at-fu1Wer-
-Mafer-Reactor-tuient Systee44pe-Reptures (Lcss-of-Cochnt-Ascident4-Major Secondary Coolant System Pipe Rupture Rupture of a Control Re.ed Drive Mechanism Housing ', Rod Clut- Control Assembly Ejection)
L f
K-28 CATAWBA - UNITS 1 & 2 3/4 1-15
UNIT 1 i
POWER DISTRIBUTI:N L:MITS SURVEILLANCE REOUIEEMENT5 (Continued)
(3) Witnin 15 minutes:
(1) C:ntrol the AFD to within new AFD limits tnat are cetermined by:
recuced COLR( )
(AFD Limit) negative = f^c0 Limit) negative min
+ fg? i Margin 0P] absolute value
)
recuced COLR (AFD Limit) positive = (AFD Limit) positive
- , .r.i n
.r# ,
- Margingp] absolute value min where Margin is the minimum margin from 4.2.2.2.c.1,k$d (2) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alarm setpoints to the maaified limits of 4.2.2.2.c.2.a. or (b) ' amply with the ACTION requirements of Specification 3.2.2
- 3. Find the minimum RPS Margin of all locations examined in '
4.2.2.2.o.1 above. If any margin is less than zero, then the i following action chall be taken:
Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, reduce the Xi value for OT.iT by:
ra min K2 adjusted = K (4) - [KSLOPE' ) x Margin gp3] absolute value t
min wnere u.ARGIN 3p3 is ne minium margin f rom 4. 2.2.2.c.1.
l g gtmq % #argovio\ded w 4'O 'O d ' d ' I do< M J g6 Fq " ns U CLLDIG% $5 **H -
l
- 0) Defined and s;;ecifiec in the COLR per Specification 6.9.1.9.
l
(* K 1value from Table 2.2-1.
3/4 A2-5 MJ _/.unam n-his-4 Uni 4-1-}- k CATAWBA - UNITS 1 & 2 4ettremert Ni. 30 (t' nit 2)
UNIT 1 e n.g:s D .' .t ~ '. . UT *g' ', ..'.". ". -I SURVEILLANCE RE00'.EEMENT5 (Continuec)
L5 __ b Nb I d, Extracci ating in no m . =acenDmeasurements :: 31 Effective Full ,
Power Oays Oeyone tne most recent measurement ano if- {
i M L OP - i "F (X,v,:: (extracolatec) > [F,(X Y,Z] (extracolatec), or
- n. ,
~
- R d M L RPS Qw1Gg [Fq (X,Y.Z] (extracolatec) > (F (X,Y,2]
q (extracciatec),
either :f tne following ac.icns snail ce taken: ;
M
- 1. F,(X,Y.Z) snall be increased by 2 percent over inat scecified in 472.2.2.a, and the calculations of 4.2.2.2.c repeatec, er
- 2. A movable incere cetector power distribution mac snall be catainec, and the calculations of 4.2.2.2.c.1 snall be performec no later than the time at which the margin in 4.2.2.2.c.1 is extrapolated to be equal to zero.
- e. The limits in Specifications 4.2.2.2.c and 4.2.2.2.d are not aoplicaole in the following core plane regions as measured in percent of core heignt from the bottom of the fuel:
- 1. Lower core region from 0 to 15%, inclusive.
- 2. Upper core region from 85 to 100%, inclusive.
l 4.2.2.2 When a full ct;re power distribution map is taken for reasons otner l
l than meetir.g the recuireinents of Specification 4.2.2.2, an overail F "(X,Y,Z) 9 l
l shall be ceterminac, then increased by 3% to account for manuf acturing
! tolerances, furtner increased by 5% to account for measurement uncertainty, l
anc further increased by the radial-local peaking factor to obtain a maximum local peak. This value shall be comcared to the limit in Specification 3.2.2.
1 j U-jv m x l lb.xtrapolat:.:n of Fy f:: the initial flux map taken af ter reacn:.n; l ,
equilibrium ::ncitions is not requirec since the .n:tial flux =ap
'. esta:11shes :ne baseline ::easurement for future trencing. Also, l
extrapolat:. n :f F; limits are not valid for core cations that were
( prev:.:u: sly :::ced, :: f:: 00:e locations that were prev:.cusly within
, :2% ci -he ::re r.aignt accu: :ne cemanc position f ne :00 ::.p.
/-
~' - .
N CATA',iBA 't"dT~ 1 1 2 3/4 A2-6 *30 -Amenement-NoJ6 -{Urti t-1-)-
.Amenament - No. En (Uni
- 24--
. - . .- . - - _ - . - . . . _ _ - . - _ . - . . - - . . . . . ~ . . . - . . . . . .
for specification 4.2.2.2 Attach ent 1:
( F.? ( X , i :)) (extrapolated) 2 (F; (X, Y,2 ) ) " (extrapolated), and fF#f> Y.21) textrapolated) ( F" ( X , Y , Z ) )
's
( F;. ( A , ': , 21 ) " (extrapelated) ( F] ( X , Y , 2 ) ) "
r F](X,i,2)) (extrapolated) 2 ( FI; ( X , Y , 2 ) ) "" (extrapolated), and iF"/> .Ci) textrapolated)
( F" I X . Y . 2 ) )
. F; ( X , Y. 2 ) ) "' (extrapolated) ( Fi l X , Y,2 ) ) '"
either cf the following actions shall be taken:
l .
s I t
)
l 8-31
UNIT '. !
! i I
POWER DISTRIEUTION L:MITS 3/4.2;3 NUCLEM ENTHALPY RISE HOT CHANNEL FACTOR :1H(X Y)
LIMITING CONDITION ' R OPERATION ,
3.2.3 FM(X,Y)snali te limiteo by imposing tne fellcwing relationsnic:
- Lco Fi.f(X,Y'1 Fa d (X,Y)
J Where: :1N(X,Y)=the measu7ec racial peak.r ti n,, :: f N. a W 4 g L&O.. . %- ia --r- m e-- .. 44.u.--
- - - i. -..-.i1- s ,w-,w.
. . j .
- aFh(X,f) = the maximum allcwaole racial peak as defineo L. - in tne (COLR).
APPLICABILITY- u00E 1. (UNIT 1) ceeE c?EKAT w C3 t_IMIT s 6LLPORT ACTION.:
With Fg(X,Y) exceecing its limit:
- a. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, reduce the allowaole THERMAL POWER f rem RATED THERMAL POWER at least RRH% C ) for each 1% that Fd(X,Y) exceecs the limit, and
- b. Within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> either:
- 1. Restore FaFM(X,Y) to within the limit of Specification 3.2.3 for RATeu i ne. ".AT. r0WER, or
- 7. Reduce the Power Rangs Neutron Flux-High Trip Setpoint in Table 2.2-1 at least RRH% for each 1% that FA. (X,Y) exceeds that limit, and
- c. Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of initially oeing cutside the limit of Specificatice.
3.2.3, either:
! u i 1. Rest:re FAHR (X,Y) to within the limit of Specification 3.2.3 for RATED THERMA'E POWER, or l
i 2. Perform tne following actions:
(a) Recuce the OTAT X 1 term in Table 2.2-1 by at least TRH(23 i
for each 1% that F (X,Y) exceecs the limit, and l
(b) '/erify through incere maoping that FAHh(X,Y) is restorea t:
with" + e limit f:r the recucec 'ERMAL PCWER alicwec t'/
( ACTICN a, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
WRRH is the amount of. THERMAL POWF#
- tion required to compensate for each 1% that Fa (X,Y) exceeds (F M j provided in the COLR per
- j. Specification 6. .1.9. ty.g, L,gg $ 5p.ci9 odica 3.'.2.h
(*)TRH is the amount of OTAT K 1 setpoint recuction required to ccmoensate for each 1% tnat Fad (X,v) exceeds the limit of S:ecification 3.2.3, providec l in tne C"LR er iceci ficat1 n 6. 9.1. 9.
3/4 A2-7
" -Amendment-h-66-6! nit }-
f CATAWBA - UNITS 1 & 2 l JLtenc enc-444 Unit-U-
UNIT '
POWER 01STRIBUTICN L:u5 LIMITING CONDITICN FC; 00E:JTION ACTION (Continveci
- d. Identify anc c:rrect tne cause of the cut-0'-limit concition crice '
to increas t ;; MERFAL POWER a::ove the recuce: THERMAL F0WER limit requirec ::y ACTICN a. anc/or 1.2. , acove; sucsecuent POWER OPERA-TION may cr: eec :roviced that Fa (X v) is cemerst atec. :uc-inc:r. a....3, . ce -1.nin ...e :mit 3:ecifiec in tne 0 A.R prior :: exceeci..g the following THERMAL POWER levels:
- 1) 50% of RATED THERMAL POWER,
- 2) 75% of RATED THEPWAL POWER, and
- 3) Witnin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining greater tnan or ecual to 95%
of RATED THERMAL POWER.
SURVEILLANCE REOUI:E.ugy73 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.
4.2.3.2 FA.f(X,Y) snall be avaluated to cetermine wnether Fg (X,Y) is wiutin I its limit by:
- a. Using the movacle intore detectors to obtain a power distribution '
map at any THERMAL POWER greatar than E% of RATED THERMAL POWER.
- b. Measuring FaF[(X,Y) according to the following schedule:
q -
gm al ~i 1. Prior :: cceration above 7E% of RATED THERMAL POWER at tne bec1nning of each fuel cycle. anc the earlier of:
ArtM3
- 2. At least nce per 31 Effective Full Power Days, or
- 3. At eacn -ime the QUADRANT POWER TILT RATIO incicated by the ,
excere cetec . ors is normali:ec using incere cetector measurements,
- c. Performing tne fellcwing calculations:
- 1. For eacn location, calculate the % margin to the maximum allowacle cesign as follows:
%F. F (X,Y)y x 100%
J Margin = 1 - - hRv
[~F (X,Y) !
L. J I No a::ditjonal unegrtgi itles are required for F ,(X,Y),
because FaFg(X,Y) ft. cludes uncertainties.
~ /
CATAWBA - UNITS 1 & 2 3/4 A2-8 g' ~ Joenoment-Nm-eHUnit4)-
-Amenoment-No.-FMUni t4)-
for Specification 4.0.3.3 Attachment 1:
- 1. "rrn reachina equilibrium conditions after exceeding by 10%
c-r more of FATED THERMAL POhTR, the THEMAL POWER at which FL4 X,Y) was last determined U) . or p
- 4 i
i i
i:.
l l-i i
l l '.
l l'
i I
I l
1.
l .;
l l-(3) Durin; power escalation at the beginning of each cycle, THEMAL POWER may be increased until a power level for extended operation has been achieved and a power distribution map obtained.
i 8-34
UNIT 1 i DOWER DIST:i:EUT::N LIMITS
{
LIMITING CCNO TICN FOR OPERATION l 1
ACTION (Continued)
- 2. Fina the minimum margin of all locations examinea in 4.2.3.2.c.1 a::cv e . :f any margin is less tnan zero, comoly with :ne ACTICN 3.2. 3x;f C Fd (x,y)]'
- is 4ke 45
_ s we rements recui a s C. FA of x, y)]"o S(necification f d. Extracciating tne two mos; recent measurements to 31 Ef fective full I
Power Cays ::eyona tne most recent measurement and if:
Fahr" (extracolatea) > FaHRL (extrapolatea)
I l eitner :f int following actions snall be taken:
- 1. FAHR'M(X,Y) snall be increased by 2 percent over that specifiea in 4. 2.3.2. a, anc the calculations of 4.2. 3.2.c receatea, or '
- 2. A movaale incore detector power distribution mao shall bc cotainea, and the calculations of 4.2.3.2.c shall be performed no later than the time at which the margin in 4.2.3.2.c is extracolated to be equal to zero.
\Acf A O a c.k e en d f l
l t
l l
l 44 CATAWBA - UNITS 1 & 2 3/4 A2-9 g.35 -Amendment-NoreC5-(Unit 4)- {
-Amendment-Noree-(Urti-t-2}- ,
for Specificatica 4.2.3.2
]
Attachment 1:
- d. Extrapolatint(') at least two measurements to 31 Effective Full 4 1
)
Power Days beyond the most recent neasurement and if:
l FL. ( X , D fextrapolated) 2 ( Ph ( X ,. Y ) ) "" { extrapolated)
F" ( :- "i (extrapolated) F" ( X . Y 1 (FL(X,Y))' ' (extrapolated) ( Fi, ( X , Y ) ) *'" ,
either of the following actions shall be taken
- 1. Ft. [h , Y ) shall be increased by e percent over that specified in-4,2,3.2.a. and the calculations of 4.2.3.2.c repeated, or
- 2. A mavable incore detector power distribution map shall be obt ai v-d. and the calculations of 4.2.3.2.c shall' be pertormed-no later than the time at which the margin in
. 4.2.3.2.c is extrapolated to be equal to zero.
i (4) Extrapolation of FL for the initial flux nap taken af ter reaching equilibrium conditions is not required since the initial flux map establishes the baseline measurement for future trending.
1 4
W 1
1 d
s I
i 8.M f.
9
. < - . - .- ,r-y ,y,- - . . . . . - , , ,, --
. . , ..--,.~,v.r. - .m.._ , -v.-.. , , - - - x- w-- ,-
- NIT ~.
70WER DISTRIBUTI:N CMITS 3/4.2.5 DNB PAC: METERS LIMITING CCNDITI:N FOR OPERATION 3.2.5 The folio ing DNS relatec carameters shall be maintairec within the limits shown on Tacle 3.2-1:
- a. React:r ho' ant System i gg, .
- b. Pressurizer ?ressure,
- c. React:r Coolant System Total Flow Rate.
APPLICABILITY: SOCE 1. (Unit 1)
ACTION:
- a. With eitner of the parameters identified in 3.2.5a. and b. above i exceecing its ?imit, rest:re the parameter to within its limit I within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THEMAL PCWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. l
- b. With :ne ccmcination of Reactor Coolant System total flow rate and THEMAL PCWER within the region of restrictea operation specified on Figure 3.2-1, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> reduce the Power Range Neutron Flux-High Trio Setooint to below the nominal setpoint by the same amount
(% RTP) as the power reduction required by Figure 3.2-1.
- c. With tne ccmcination of Reactor Coolant System total flow rate and THERMAL PCWER within the region of pronibitec c eration specifieo on Figure 3.2-1:
- 1. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:
a) Restore the c:.noination of Reactor Coolant System total flow rate and THERMAL POWER to within the region of permissiele operation, or Mst:re the ::m:i :t*:n :f S::::- Odant ,.itaa atf flow rate and THEMAL POWER to within the r on of restricted operation and comply with acti . above, or c) Reduce THERMAL POWER to less than 50% of RATED THEMAL POWER anc reduce the Power Range heutron Flux - High Trip Setooint to less than or equal to 55% of RATED THEFAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
X-37 CATAWBA UNITS 1 & 2 3/4 A2-13 _Amencment-Norg-@n4 t -H-
%encment4er-gg-(4)ntQ
- UMI T 1 TABLE 3.3-1 -
L 9.
.Y REACTOR TRIP SYSTEM INSTRUMENTATION ' 'f E
MINIMUM E 101AL NO. CHANNELS CHANNELS. APPilCABLE IUttCTIONAL UNIT OF CHANNELS 'TO TRIP OPERABtE MODES ACTION 3
gJ
- 1. M.umal INactor Irip 2 1 2 1, 2 I .
2 1 2 3 * , 4 * , 5* 10
[
- 2. Power Range, Neutron Flux
- a. High Setpoint 4 2 -3 1, 2 2 ,
- b. Low'Setpoint 4 2 3 l###, 2 -2 4
2
- 3. Power Range, Neutron Flux :4 2 3 1, 2 .
l .
Iligh Positive Rate -)
l 1Mck e, -_
. q . i. er Range,' Neutron Flux, 4 2 3 1, 2 2
- Higti Negative Rate i gL S f Intermediate Range, Neutron Flux -2 1 2 1###, 2 3 ,
4 g E. Source Range, Neutron Flux ;
x 5- a. Startup: 2 1 2 2## 4 l
- b. Shutdown 2 1 2 32, 4*, 5* 10 -t
> / Overtemperature AT il /c Four L6op Operation 4 2 3 1, 2 6 l a ,
lA y g Overpower AT '!
AS :7 Four Loop.0peration 4 2 3 1, 2 6 l
- e z .
P? ): Pressurizer Pressure-Low 4 2 3 1 6** l
- P60 4
E*
j
., g. 7 s .. _ _ _ _ _ -
LJrJ r T l TABLE 3.3-1 (Continued)
"2=
$ REAC10R IRIP SYS1EH INSTRUMENTATION 5
'" HINIMUM E 10TAL NO. CliANNELS CilANNELS APPLICABLE FUNCTIONAL UNIT OF CllANNELS TO TRIP OPERABLE H0 DES ACTION
}
] Pressurizer Pressure-liigh 4 2 3 1, 2 6** l
[
11'. Pressurizer Water Level-liigh 3 2 2 1 6 l' 10 Jf. Reactor Coolant Flow-Low 3/ loop 2/ loop in 2/ loop in 6 3 it a. Single Loop (Above P-8) 1 any oper- each oper-ating loop ating loop
- b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop 1 6 l w below P-8) two oper- each oper-
'E ating loops ating loop B
J. JI.' Steam 9.enerator Water 4/stm 2/ste gen 3/sta gen 1, 2 6** l 11 Level--Low-Low gen in ar.y each
? operating operating a stm gen sta gen Undervoltage-Reactor Coolant 4-1/ bus 2 3 1 6 l 34'.
- p; > 13 Pumps (Above P-7) lill E 5' underfrequency-Reactor Coolant 4-1/ bus 2 3 1 6 l oo k" )f 4 Pumps (Above P-7)
[ 3 6'. Turbine Trip
? a. Stop Valve Eli Pressure 4 2 3 1#### 6 l i s' f:$; - Low
- b. Turbine Stop Valve Clusure 4 4 1#### 11 g 1 l
cc
?4 J7. Safety Injection Input 2 1, 2 9
,f, } lb froia ESF 2 1
~
Ur/r T I TABLE 3.3-1 (Continued) n .,
E 2-RfACIOR 1 RIP SYSIEH INSTRUMENTATION h HINIMUM TOTAL NO. CilANNELS - CHANNELS APPLICABLE g FUNCTIONAL UNIT OF CHANNELS TO TRIP 0PERABLE H0 DES ACTION h 17 }8'. Reactor Trip System Interlocks
~ a. Intermediate Range
- e. Neutron Flux, P-6 2 1 2 2## 8 m
- b. Low Power Reactor Trips Block, P-7 P-10 Input 4 2 3 1 8 or P-13 Input 2 1 2 1 8
- c. Power Range Neutron '
w Flux, P-8 4 2 3 1 8 k
B d. Power Range Neutron 4 2 3 1 8 4 Flux, P-9 .
- e. Power Range Neutron Flux, P-10 4 2 3 1 8 4 .
O
- f. Power Range Neutron Flux, Not P-10 4 3 4 1, 2 8 9 Turbine Impaise Chamber Pressure, P-13 2 1 2 1 8 16 Ji Reactor Trip Breakers 2 1 2 1, 2 9 2 1 2 3*, 4*, S* 10 l'f pd. Automat.ic Trip arid Interlock 2 1 2 1, 2 9 Logic 2 1 2 3*, 4*, 5* 10 t
( 's q
VNZT \
TABLE 3.3-1 (Continued)
TABLE NOTATIONS "Only if tne Reactor Trip System breakers happen to be in the closed position and the Control Rod Drive System is capable of rod withdrawal.
- Comply with the provisions of Specification 3.3.2, for any portion of the channel recuired to be OPERABLE by Specification 3.3.2.
- Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. !
- Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
- Above the P-9 (Reactor Trip on Turbine Trip Interlock) Setpoint.
ACTION STATEMENTS ACTION 1 - with the numoer of OPtRABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to CPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least MOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 2 - With the numcer of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceea orovided the following conditions are satisfied:
- a. The inoperable cnannel is placed in the tripped condition within 6 hoers,
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of otner channels per Specification i 4.3.1.1, and
- c. Either, THERMAL POWER is restricted to less than or equal te 75% of RATED THERMAL POWER and the Power Range Neutron Flux trip setooint is recuced to less tnan or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the
! QUADRANT POWER TILT RATIO is monitorea at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.2.
ACTION 3 - Witn the number of channels OPERABLE one less than the M:nimum Channels OPERABLE requirement and with the THERMAL POWEP. level:
- a. Selow the P-6 (Intermediate Range Neutron Flux Interlock) l Setuoint, restore the inoparable channel to OPERABLE I status prior to increasing THERMAL POW'ER above the P-6 Setpoint; or
- b. Above the P-6 (Intermediate Range Neutron Flux Interlock)
Setpoint but below 10% of RATED THERMAL POWER, restore the l inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.
CATAWBA vNITS 1 & 2 3/4A3-5 -AmenementMoA tHUrt&t-it M _AnanonentJoM-(Wit-&
UNa 1 g .;
TABLE 3.3-1 (Continued) j l
ACTION STATEMENTS (Continuet.Q ACTION 4 - With the numoer of OPERABLE channels one less than the Minimum Chennels OPERABLE requirement, suspend all operations involving positive reactivity changes.
ACTION 5 - celete ACTION 6 - With the numoer of OPERABL' channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following concitions are satisfied;
- a. The inoperable cnannel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperaole channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of othar channels per speci fication 4. 3.1.1.
ACTION 7 - Delete ,
ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within I hour determi.1e by observation of the associated permissive status light (s) that the laterlock is in its requireo state for the existing plant condition, er apply Specification 3.0.3.
dCTION 9 - With the numoer of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, 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 />; nowever, one channel may be bypasseo for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1, provided the otner cnannel.is OPERABLE.
ACTION 10 - With the numoer of OPERABLE channels one less tnan the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the Raactor trip creakers within the next hour.
ACTION 11 - Witn the numoer of OPERABLE channe'.s less than the Total Number of Channels,- coeration may continue provided the inocerable chanr.els are placed in the tripped condition within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> s.
f
'ATAWBA - UNITS 1 & 2 3/4 A3-6 x.42 -Amendment. No. r(tini-t 'd-henemont,--NorJMUni t 2)-
u k'I T- \
IABLE 3.3-2 h REACIOR_lRiP SYSIEH INSIRi1HENIAi!ON RESP 0flSt ilHIS B
. Ii tiCl10tlAl llH11 RtSPetlSE Ilt4E c
5 1. Manual Reactor Trip fl . A .
-i a 2. power Itange: Heistrei tiux 3 0.5 second*
e m 3. Power Range, Heutron flux, liigh Positive Rate ,
it. A.
TDekC #2L
[4. Power Range, Neutron Flux, liigh Negative Rate $ 0.5 second*
I Intermediate Rang'e, Neutron flux ti A.
4 l .ti. Source Range Neution ilux fl . A .
$ 5
't' / Overtemperature AI <
M8kseconds* l
~ s
[M.
w .;
Overpower'AI $ kseconds* l
- s. Pressurizer Pressure-Low ~< 2 seconds B
JE Pressurizer Pressuse-liigh 5 2 seconds if Y 9
@8 )1. Pressurizer Water l evel-liigli ti. A.
2a
? .*
c'E *tteutron detectors are exempt from response time testing. Response time of the neutron flux signal portion
$i'
" P of the channel shall be measured from detector output or input of first electronic component in channel.
--*Applicetde-upon-4!alation-of-flID-Bypass-System.
Sh
t/M1 T ' l '
TABLE 3.3-2 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES E
- FUNCTIONAL UNIT RESPONSE TIME C
(
- ^
J2' Iow Reactor Coolant flow 11 H a. Single Loop (Above P-8) $ 1 second 9- b. Two Loops (Above P-7 and below P-8) $ 1 second
)I.
12 Steam Generator Water Level-Low-!ow
- a. Unit 1 < 3.5 seconds
- b. Unit 2 32.0 seconds J4I undervoltage-Reactor Coolant Pumps $ 1.5 seconds Ib
$ Pl. Underfrequency-Reactor Coolant Pumps 5 0.6 second
> 14
[ 36. Turbine Trip IS x a. Stop Valve EH Pressure-Low H.A. l h b. Turbine Stop Valve Closure H.A.
J7. Safety Injection Input from ESF N.A.
Ib l[l[
- s:s J8.
l'?
Reactor Trip System Interlocks N.A.
}:sa'[
- 39. Reactor Trip Breakers N.A.
t8
(( J0. Automatic Trip and_ Interlock logic N.A.
P .* 19 it.
p
u!JE T -- \
I Alll E 4. 3-1 fi*
REACTOR 1 RIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREHENTS .
IPT!
ANALOG ACillAllNG MODES FOR CllANNEL DEVICE WillCil
$ CilANNEL OPERA 110NAl. OPERA 110HAL ACTUATION SilRVElll ANCE d CilANNE L CllECK CALIllRAIION ItST TES1 10GIC IESI IS REQtilRED g lilNtjl10 NAl llN il t-Mantial Reactor Trip N. A. N.A. H.A. R ft. A. 1, 2, 3*, 4^, 5*
m 1.
- 2. Power Ran0e, Neutron Flux
- a. liigh Setpoint' S D(2, 4), H H.A. N.A. 1, 2 H(3, 4),
Q(4, 6),
R(4, 5)
- b. Low Setpoint S R(4) M N.A. H.A. 1###, 2
% N.A. N.A. 1, 2 g 3. Power Range, Neutron Flux, N.A. R(4) H y fli o - Ic-!-e_gh Positive Rate --
~
- 4. Power Range, Neutron Flux, H.A. R(4) H N.A. H.A. 1, 2
,1; liigh Negative Rate
- 4. Interrediate Range, S' R(4, 5) S/U(1),H N.A. N.A. l###, 2 4 Neutron flux
$N Source Range, Heutron Flux R(4, 5) S/U(1),H(9) N.A. N.A. 2##, 3, 4, 5 g@ 4 5 g!}j Z Overtemperature al S R H H.A. N.A. 1, 2 y a (o E Ove. power of S R H N.A. N.A. 1, 2
$jf
- 1
$; o M. Pressurizer Pressure-tow S R H N.A. N.A. 1
' I dv s N.A. H.A. 1, 2
'y g J0. Pressurizer Pressure-liigh 5 R H
\ XX R H H.A. N.A. 1 l ,,, g ,11. Pressurizer Water level-iligh S
- - f0 W. Reactor Coolant flow-l ow S R H N.A. N.A. 1 II
.. w
LJtJE T \
TABLE.4.3-1 (Continuedl REACTOR 1 RIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENIS E
$ TRIP e ANALOG ACTilATING MOOLS FOR c- CHANNEL DEVICE WilICH
- i CllANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION SURVFILLANCE d 10NCil0NAl liNIT t.ilECK CALIBRATION lEST TEST LOGIC IESI IS RigilIRED Js." Steam Generator Water Level- S R(13) H N.A. N.A. 1, 2 m II Low-Low J4I Undervoltage - Reactor Coolant. N.A. R N.A. M N.A. 1 G Pumps J8: Underfrequency - Reactor H.A. R N.A. H N.A. I 11 Coolant Pumps 367 Turbine Trip y /r a. Stop Valve Eit Pressure N.A. R N.A. S/U(1, 10) H.A. I#
g - Low
- b. Turbine Step Valve Closure N. A. R N.A. S/U(1, 10) N.A. 1#
.17' Safety Injection Input from N.A. N.A. N.A. R** N.A. 1, 2
//, ESF
- 7 g J8' Reactor Trip Systm Interlocks 93 17 a. Intermedi te Range
{} Neutron Flux, P-6 N.A. R(4) H N. A. N.A. 2##
f> $ b. Low Power Reactor l
gg Trips Block, P-7 N.A. R(4) M(8) H.A. N.A. 1
- c. Power Range Neutron EM Flux, P-8 N.A. R;4) H(8) N.A. N.A. 1
-e d. Low Pawer Range Neutron f, Flux, P-9 N.A. R(4) M(8) N.A. N.A. 1 I f" bI
} ** This surveillance need not be performed until prior to entering STARTUP following the Unit 1 first
! refueling.
UNCT' l TABLE 4.3-1 (Continued) n
$ REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENIS E
E TRIP i ANALOG ACTUATING. MODES FOR e C!!ANNEl DEVICE WillCll 3 CilANNEl CilANNEl DPE!!AIlONAl OPERA 110HAl ACTUATION SilRVL'Ill ANCE d filNCil0NAL llNil ClllCK _ CAllBRAll0N 11 51 !LSI 10GIC 1[SI 15 RLQlllRfD
- e. J8' Reactor Trip System Interlocks (Continued) .
M l'7 l
- e. Power Range Neutron :
flux, P-10 N.A. R(4) M(8) H.A. N.A. I
- f. Power Range Neutron flux, Not P-10 N.A. R(4) 71( 8 ) N.A. N.A. 1, 2
- g. Iurbine Impulse Chamber .
W Pressure, P-13 N.A. R M(8) N.A. N.A. 1 Y J'f. Reactor Trip Breaker N.A. IL.t. N.A. M(7, 11) N.A. 1, 2, 3 * , 4 * , S
- C J8
% Automatic Trip and Interlock logic N.A. N.A. N.A. N.A. M(7) 1, 2, 3*, 4*, 5*
/9
L4JL T \
TABLE 4.3-1 (Continued)
TABLE NOTATIONS Only if the Reactor Trip System breakers happen to be closad and the Control Rod Drive System is capcole of ecd withdrawal.
- Above P-9 (Reactor Trip on Turbine Trip Interlock) Setpoint.
- Below P-5 (Intermediate Range Neutron Flux Interlock) Setpoint.
- Belos P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
(1) If not performed in prewicus 7 days.
(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power M absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(3) Single point comparison of incore to excore axial flux
- fference above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(5) Detector plateau curves shall be obtained, evaluated and compared to manufacturer's data. For the Intermediate Range and Power Range Neutron Flux cnannels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(6) :ncore - Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Specification 4.'0.4 are not. applicable for entry into
~~
MODE 2 or 1.
(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
(8) With power greater than or equal to t"a interlock setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall unsist of verifying tnat the interlock is in the required state by observing the permissive status light.
(9) Montnly surveillance in MODES 3*, 4*, and 5* shall also include verifi-cation that permissives P-6 and P-10 are in their required state for j existing plant conditions by observation of the permissive status light.
j (10) Setpoint verification is not applicable.
l (11) At least once per 18 months and following maintenance or adjustment of l the Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the Undervoltage and Shunt trips.
- l. ;(12) Deleted (13) For Unit 1, CHANNEL CALIBRATION shall ensure that the filter time constant associated with Steam Generator Water Level Low-Low is adjusted to a value l
less than or equal to 1.5 seconds.
CATAWBA - UNITS 1 & 2 3/4A3-12 _ Amendment Nos94 Unit 4) j aa ndment No. 42 (Unit 2) l-
O tJZ T 2-n TABLE 3.3-1 d REACTOR IRIP SYSTEli INSTRUMENTATION 5
HINIMUM E 10TAL NO. CHANNELS CHANNELS APPLICABLE 3 IONCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION
" 1. Manual Reactor Irip 2 1 2 1, 2 1
~ 2 1 2 3*, 4*, 5* 10
- 2. Power Range, Neutron Flux
- a. High Setpoint 4 2 3 1, 2 2
- b. Low Setpoint 4 2 3 1###, 2 2
- 3. Power Range, Neutron Flux 4 2 3 7, 2 2 1 Nigh Positive Rate t 4. Power Range, Neutron Flux, 4 2 3 1, 2 2 l g liigh Negative Rate w
d S. Intermediate Range, Neutron Flux 2 1 2 l###, 2 3 f
x 6. Source Range, Neutron Flux 5 a. Startup 2 1 2 2t# 4
- b. Shutdown 2 1 2 3*, 4*, S* 10 m - 7. Overtemperature AT
Four Loop Operation 4 2 3 1, 2 6 l 8*eO 8. Overpower AT $$ Four Loop Operation 4 2 3 1, 2 6 l $* 9. Pressurizer Pressure-Low -l 4 2 3 1 6** l NM S. 3. U a
n) ti L T 2_ TABLE 3.3-1 (Continued) REACIOR TRIP SYSTEM INSTRllMENTAlION E HINIMUM E 10TAL N0. CilANNELS CilANNELS APPLICABLE
] filNCTIONAL UNIT OF CllANNELS ,
TO TRIP OPERABLE MODES ACTION [ 10. Pressurizer Pressure-liigh 4 2 3 1, 2 6** l
- 11. Press'arizer Water Level-liigh 3 2 2 1 6 l
- 12. Reactor Coolant Flow-Low
- a. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 6 l any oper- each oper-ating loop ating loop
- b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop 1 6 l w below P-8) two oper- each oper-g ating loops ating loop w
0 13. Steam Generator Water 4/stm 2/stm gen 3/stm gen 1, 2 6** l Level--Low-Low gen t in any each x operating operating g i stm gen stm gen
- 14. Undervoltage-Reactor Coolant 4-1/ bus 2 3 1 6 l k :> Pumps (Above P-7) 8 ll 85 15. Underfrequency-Reactor Coolant 4-1/ bus 2 3 1 6 l
$@ Pumps (Above P-7) zx 16. Turbine Trip
? .? a. Stop Valve Eli Pressure 4 2 3 1#### 6 .l
$h - Low il b. Turbine Stop Valve Closure 4 4 1 1#### 11 l
!?*
-$. 17. Safety Injection Input '
from ESF 2 2 9 (( 1 1, .
. ~
JWC. T 7- ^ :' TABLE 3.3-1 (Continued) Q
> REAC10R IRIP SYSIfH INSTRUMENIATION h HINIMUM 10lAL NO. CilANNELS CHANNELS APPLICABLE c FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE H0 DES ACTION 55 d 18. Reactor Trip System Interlocks s a. Intermediate Rance
- p. Neutron Flux, P-6 2 1 2 2## 8
~
- b. tow Power Reactor Trips Block, P-7 P-10 Input 4 2 3 1 8 or P-13 Input 2 1 2 1 8
- c. Power Range Neutron w Flux, P-8 4 2 3 1 8 1 ~
27 d. Power Range Neutron 4 2 3 1 8 1 Flux, P-9 -
- e. Power Range Neutron r
Flux, P-10 4 2 3 u 1 8
~ i
- f. Power Range Neutron .
Flux, Not P-10 4 3 4 1, 2 8
- g. Turbine Impulse Chamber Pressure, P-13 2 1 2 1 8
- 19. Reactor Trip Breakers 2 1 2 1, 2 9 2 1 2 3*, 4*, 5* 10
- 20. Automatic Trip and Interlock 2 1 2 1, 2 Logic 9 2 1 2 3*, 4*, 5* 10
( i x .
JAC T Z_ TABLE 3.3-1 (Continuadj TAB _LE NOTATIONS
*0nly if the Reactor Trip System breakers happen to be in the closed position and the Control Rod Drive System is capable of rod withdrawal.
** Comply with the provisions of Specification 3.3.2, for any portion of the channel required to 'ce OPERABLE by Specification 3.3.2.
##Below the P-6 (Intermediate Range Neutron Flux Interlock) Setooint. 1
###Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
####Above the P-9 (Reactor Trip on Turbine Trip Interlock) Setpoint.
ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable cnannel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours. ACTION 2 - With the number of OPERABLE channels one less than the Total NumDer of Channels, STARTUP and/or POWER OPERATION may proceed
- provided the following conditions are satisfied
- a. The inoperable cnannel is placed in the tripped condition within 6 hours,
- b. The Minimum Channels OPERABl[ requirement is met; however, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.1.1, and
- c. Either, THERMAL POWER is restricted to less than or equal
. to 75% of RATED THERMAL POWER ano the Power Range Neutron j Flux trip setpoint is reduced to less than or equal to l 85% of RATED THERMAL POWER within 4 hours; or, the
- QUADRANT POWER TILT RATIO is monitored at least once per l 12 hours per Specification 4.2.4.2.
!' ACTION 3 - with the number of channels OPERABLE one less than the Minimum l Channels OPERABLE requirement and with the THERMAL POWER level:
- a. Below the P-6 (Intermediate Range Neutron Flux Interlock)
Setpoint, restore the inoperable cnannel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint; or D. Above the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint but below 10% of RATED THERMAL POWER, restore the
- inoperable channel to OPERABLE stctus crior to increasing
- THERMAL POWER above 10% of RATED THERMAL POWER.
l CATAWBA - UNITS 1 & 2 3/4B3-5
- 52 hancment-h,44UMt-D -
%endment No Al (Unit 2)
U ALE T g TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued) ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity cnanges. ACTION 5 Delete ACTION 6 - With the numoer of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours, and
- b. The Minimum Channels OPERABLE requirement is met; however.
the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.1.1. ACTION'7 - Delete ACTION 8 - With less than the Minimum Numoer of Channels OPERABLE, within 1 hour determine by observation of the associated permissive status light (s) that the interlock is in its required state for the. existing plant condition, or apply Specification 3.0.3. ACTION 9 - With the number of OPERABLE cnannels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANOBY within 6 hours; however, one cnannel may be bypassed for uo to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE. ! ACTION 10 - With the numcer of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable cnannel to OPERABLE status within 48 hours or open the Reactor trip breakers within the next hour. ACTION 11 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable cnannels are placed in the tripped condition within 6 hour s. l l l 8-53 CATAWBA - UNITS 1 & 2 3/433-6 3,,ng,,,t gg,39(ggg_4)_ f -Amendment-Ne-M(4 Jet 2} ( l __
LJAl.3 T~ Z-lABLE 3.3-2 I i b RLACIOR 1 RIP SYSTEH INSTRUMEHIATION RESP 0ftSE IINES E
> RESPONSE IIHE
. IlltiCIl0HAL UNIT
- 14. A.
b 1. Mariual Reactor Irip Y
~ 2. Power Range, Neutron flux $ 0.5 second*
e. n 3. Power Range, Neutron flux, N.A. liigh Positive Rate
- 4. Power Range, Neutron Flux, liigh Negative Rate 5 0.5 second*
l ti. A. S. Intermediate Range, Neutron flux N.A. R 6. Source Range, Neutron flux 1
+ w 1
to i T l. Overtemperature Al 1 4f8 3 seconds
- 1 m
$ 448k seconds
- l 7l, 8. Overpower AT
- u l
- 9. Pressurizer Pressure-tow $ 2 seconds
, 10. Pressurizer Pressuie-liigh 5 2 seconds ii ii Pressurizer Water ievel-liigh fl. A. 9!l 11. hU
- 3. oEl
.EElf UU Response time of the neutron flux signal portion
- Neutron detectors are exempt from response time testing.
l 'c?'ci of the channel shall be measured from detector output or input of first electronic component in channel. l F.3
' -4AppHesk upon delat inn of. RID-Gyp 522 Sy stem---
l l l Cd l t_.__ ____
tJ /J.E T z. 1ABLE 3.3-2 (Continued) REACTOR TRIP SYSIEH INSTRUMENTATION RESPONSE TIMES f RESPONSE TIME I
- FUNCTIONAL UNIT "x
q 12. tow Reactor Coolant flow , e a. Single loop (Above P-8) $ 1 second e- b. Two Loops (Above P-7 and below P-8) $ 1 second n
- 13. Steam Generator Water Level-tow-Low
< 3.5 seconds
- a. Ur.it I
- b. Unit 2 32.0 seconds
- 14. Undervoltage-Reactor Coolant Pumps 5 1.5 seconds U IS. Underfrequency-Reactor C elant Pumps $ 0.6 second b 16. Turbine Trip 4
- a. Stop Valve EH Pressure-Low N.A l Turbine Stop Valve Closure N.A.
G., b. v 1 Safety Injection Irput from ESF N.A. l 17. N.A. hii 18. Reactor Trip System Interlocks 8!! N.A.
@!i 19. Reactor Trip Breakers
$$ N.A.
- 20. Automatic Trip and Interlock Logic
(( n
!!O
~,
!F
{4;t I't:
tJNZT 2 TABLE 4.3-1 D V 'REAC10R TRIP SYSI[H INSTRUMENTATION SURVEILLANCE REQUIREMENTS is . I IRIP c ANALOC ACillAllNG HODES FOR 5 CilANNEL DEVICE Ml:Cil d CilANNLL CilANNEL GPERAI10NAL OPE.7AT10NAL ACTUAT10N SURVEILLANCE-g FilNC110NAL UNIT CllECK CAllBRA110N TEST TEST LDGIC 1ESI 15 REQUIRED c~ m 1. Manual Reactor Trip H. A. N.A. N.A. R N.A. 1, 2, 3*, 4^, 5*
- 2. Power Range, Neutron Flux
- a. liigh Setpoint S 0(2, 4), H N.A. N.A. 1, 2 H(3,4),
Q(4, 6), R(4, 5)
- b. Low Setpoint S R(4) H H.A. N.A. 1###, 2 M .
g 3. Power Range, Neutron flux, H.A. R(4) H N.A. N.A. 1, 2 y liigh Positive Rate e
, 4. Power Range, Neutron Flux, N.A. R(4) H N.A. N.A. 1, '
c, liigh Negative Rate
=
- 5. intermediate Range, S R(4, 5) S/U(1),H N.A. N.A. l###, 2 Neutron Flux ifiE gg 6. Source Range, Heutron Flux 5 R(4, 5) S/U(1),H(9) N.A. N.A. 2##, 3, 4, 5 YO gg 7. Overtemperature 6T S R H N.A. N.A. 1, 2 o
7
~?g g 8. Overpower al S R H N.A. N.A. 1, 2 l iS 9. Pressurizer Pressure-Low S R H N. A. N.A. 1 SfE 10. Pressurizer Pressure-liigh S R H N.A. N.A. 1, 2 h N.A.
7,g 11. Pressurizer Water ievel-liigh 5 R H H.A. 1
.l
- 12. Reactor Coolant flow-Low 5 R H N.A. N.A. 1
UNZT L TA!!if4.3-1(Continuey , R__E_ ACTOR TRIP SYSt i;4 INSTRUMENTATION $bRVEILLANCE REQUIREMENTS 3 6
> TRIP
. ANALOG ACTUAllNG MODES FOR c CilANNEL DEVICE WilICil
*i
. CitANNEL CilANNEL OPERAlI0NAL OPERATIONAL AClUATION StlRVEiLL ANCE d illNC110NAL UNil CllECK CAtIBRATION TEST TEST __
10GIC ILSI 15 REQUIRED e e- 13. Steam Generator Water Level- S R(13) H H. A. H.A. 1, 2 m 1.ow-Low
- 14. tindervoltage - Reactor Coolant N.A. R N.A. M N.A. 1 Pumps
- 15. Underfrequency - Reactor N.A. R N. A. M N.A. 1 Coolant Pumps t
= 16. Iurbine Ir.ip E a. Stop Valve Eli Pressure N.A. R H.A. S/U(1, 10) H.A. l#
y - Low
- b. Turbine Stop Valve Closure N.A. R N.A. S/U(1, 10) N. A. l#
? ~
d 17. Safety injection Input from N.A. N.A. N.A. R** N.A. 1, 2 ESF yg. 18. Reactoe Trip System Interlocks 3]
@!F
- a. Intermettiate Range Neutron flux, P-6 N.A. R(4) H H.A. N.A. 2##
em '
*3 b. Low Power Reactor
[: ec e. Trips Block, P-7 P 4. "i(4 ) M(8) N.A. N.A. I
- c. Power Range Neutron IX3 flux, P-8 N.A. R(4) H(8) N.A. N.A. 1 lQ
,2g
<* . 1ow Power Range Neutron Flux, P-9 N.A. R(4) M(8) H.A. H.A. 1
?[t ** This surveillance need not be performed until prior to entering STARIUP following the Unit 1 first refueling.
~
a i ' cla!.T T t l l:
- i 4'
IABLE 4.3-1 (Continued) i ! REACIOR 1 RIP SYSTEM INSTRUMENTAT10N SURVElltANCE REQUIREMENTS 'E IRIP l
=E ANALOG ACTUATlHG MOOLS 10R l
- l. CilANS:EL DLVICE WillCli !
4
- SilRVCILL ANCL
- 5 CllANNI L CilANNE L OPERATIONAL DPERA110NAL ACillA110N
- % Clll CK CAllBRATION IISI TEST 10GIC IEST IS REQUIRED IllNCIIONAl UNIT I' Reactor Trip System Interlocks (Continued) 3 18. ,
i y
- c. Power Range Neutron !
Flux, P-10 N.A. R(4) H(8) N.A. N.A. I I
- f. Power Range Neutron [
Flux, Not P-10 N.A. R(4) M(8) N.A. N.A. 1, 2 l Turbine impulse Chamber g. Pressure, P-13 N.A. H(8) N.A. N.A. I ; )R R
&> 1, 2, 3 * , 4 * , 5*
T l ') . Reactor Trip Breaker N.A. N.A. N.A. H(1, 11) N.A.
- i U N.A. H.A. M(7) 1, 2, 3*, 4*, 5* l I20. Automatic Trip and Interlock '
N.A. N.A. Logic ; I I
*J Ai r T 2 TABLE 4.3-1 (Continued)
TABLE NOTATIONS Only if the Reactor Trip Sjstem breakers happen to be closed and the Control Rod Drive System is capable of rod withdrawal.
# Above P 9 (Reactor Trip on Turbine Trip Interlock) Setpoint.
## Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
### Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
(1) If not performed in previous 7 days. (2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust execre channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.U.4 are not applicable for entry into MODE 2 or 1. (3) Single point comparison of incore to excore axial flux difference above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicaole for entry into MODE 2 or 1. (4) Neutron detectors may be excluded from CHANNEL CALIBRATION. (5) Detector plateau curves snail be obtained, evaluated and compared to manufacturer's data. For tne Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. (6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into
~
MODE 2 or 1. (7) Each train shall be tested at least every 62 days on a STAGGERED TEST 9tSIS. (3) With power greater than or ecual to the interlock setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying tnat the interlock is in the requireo state by observing tne oermissive status light. (9) Monthly surveillance in MODES 3", 4*, and 5* shail also include verifi-cation that permissives P-6 and P-10 are in their required state for existing plant conditions cy observation of the permissive status light. (10) Setpoint verification is not applicable. (11) At least once per 18 months and following maintenance or adjustment of the Reactor trip creakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the Undervoltage and Shunt trips. (12) Deleted (13) For Unit 1, CHANNEL CALIBRATION shall ensure that the filter time constant , associated with Steam Generator Water Level Low-Low is adjusted to a value less than or equal to 1.5 seconds. x-59 CATAVBA - UNITS 1 & 2 3/433-12 .Amandment No- c.94Un44 1)
-Amendment 40. ;,2 (4Jnit ._ _ __
INSTRUMENT A"iICN 3/4.3.2 ENGINEE"E0 S AFETY FE ATL'RES ACTUATION SYSTEM INSTRUMENTATi N LIMITING CONDITION FOR OPERATICN _._ 3.3.2 The Engineerea Safety features Actuatien System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPEPABLE with their Trip 5etpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5. APPLICABILITY: As shown in Table 3.3-3 ACTICN:
- a. vi$ an ESFAS Instrumentation or Interlock Trip Satpoint trip less
;;cntervative than the value shown in the Trip 5etpoint column but more conservative than the value shown in the Allowable Value column, of Taele 3.3-4, adjust the Setpoint consistent with the Trip 5etpoint value.
- b. With an E5FAS Instrumentation or Interlock Trip setpoint less conservative than the value shown in the Allowable Values Column of Table 3. 3-4, etpe:r:_
g p-
- n
- 1. Adjust'the Setpoint consistent with the Trip Setpoint value of ')
Table 3.3-4, ano determine within 12 hours that Ecuation 2.2-1 / as satisfied for the affected channel, or >
'K Jeclare tne channel inoperable and apply the applicable ACTION statement requirements of Table 3.3.3 until the enannel is
-estored to OPEFABLE status with its setpoint adjusted consistent with the Trip 5etpoint value.
Dj '
~ i vation 2.2-1 2 + R + 5 5, TA }
\
/ Where:
Z = The value from Column Z of Table 3.3-4 for the affected channel, l R = Tre "as measured" value (in percent span) of rack error for the
! affecteo channel, i !
5 = Either the "as reasurec" value (in cercent scan) of the sensor 1 I error, or the value from Column 5 (Sensor Error) of Table 3.3-4 fori
- ne affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the ;
- affected channel. ~- J l
! c. With an ESFAS instrumentation channel or interlock inoperable, take l the ACTION snown in Table 3.3-3. CATAWBA - UNITS 1 & 2 3/4 3-13se
U Fl%T \ I Allt i 3.3-4 9 U t INGINilHill SAflIY ll AltlRL5 ACillAll0N SYS1tH INSIRONINI All0N IRIP SilPolNIS en
# E - be\ F stN50R }
E 10lAt ERROR Q lilNCllONAl titill All0WANCE (IA) / (5) IRIP SilP0ltil All0WABil VAllit w
- l. Saf ety injection (Reactor trip, Phase "A" Isolatiosi, feedwater N
isolation, Control Room Area Ventilation Operation, Auxiliary feedwater-Motor-Driven Pump, Purge & Exhaust Isolation, Annulus Ventilation Operation, Auxiliary fluilding filtered E xhaust Operation, Essergency Diesel m Generator Operation, Component 1 Cooling Water, Turbine trip, and 3' Nuclear Service Water Operation)
- d. Hant:41 lesitiatinsi H.A. H.A. N.A. N.A. N A.
~
b Automatic Actuat. ion logic H.A. N. A. H.A. N.A. N.A. and Actuation Relays
- c. Containment Pressure-liigh 8.2 0.71 1. 5 < l.2 psig 3 1.4 psig
- d. i'ressuriier Pressure-t ow 16.1 14.4 1. 5 3 1845 psig 3 1839 psig
- e. Steam l inie Presstsie-t ow 4.6 1.31 1. 5 > 16 ps ig 3 g gisigh 7?T 744 l 2. Containment Spray
- d. Manual IHiLidl50:n fl . A . H.A. H.A. H A. N.A.
- b. Automatic Actisatinsiiogic H.A. N.A N.A. H.A. H.A.
and Actualion Relays
- c. Containment Pressiere-Higli-liigh 12./ 0.11 1.5 l
t 3 psig ! 3.2 psig Q J
*>aC T t IABif 3.3-4 (Continued}
O INGlHilitill SAILlY ll AlilRES AtillA110H SYSilH INSilttlHtHI A110H IRIP SLIP 0lHIS M ~ h DeQcf' [ 101AL SENSOR ERROR IllNCil0NAL 11 Nil All0WANCE (l'A) Z jS) \IRIPSETPalNi Atl0WABif VAluE 5
- d. 3. Containment Isoiationi g
e a. Pilase "A" Isolation m N.A. H.A. H.A. N.A. N.A.
- 1) Hanual Initiation
- 2) Automatic Actuation Logic H.A. H.A. H.A. H.A. H.A. ,
i and Actuauen Relays
- 3) Safety injection See item 1. above f or all Safety Injection Setpoints arid Allo + !.le Valises.
1-* b. Phase "B" Isolation (Nuclear gehtke s Service Water Operation) p ] [ H.A. N.A. H.A. H.A. N.A. h 1) Hanual initiatinn H.A. H.A. H.A. N.A. H.A. ; k 2)- Automatic Actuation 1.ogic and Actuation , { Helays ,
- 3) Containment Pressure' 12.7 0.71 1.5 5 3 psig i 3.2 psig liigh-High
- c. Purge and Lxhaust isolation I k
N.A. H.A. N.A. H.A. H.A. I) Hanual InitiaLi01: , I ! 4 H.A. H.A. H.A. I;. A.
- 2) Automatic Actuation H.A.
logic and Actuation j l Helays _, i
- 3) Salety injection See item 1. above for all Salety injection Setpoints aiot Allowable Values. lt i
I
v Al.IT 1 n TABLE 3.3-4 (Continued) h ENGINEERED SAFE 1Y fEAIURES ACTUATION SYSTEM INSlRtMENTA110N IRIP SEIPOINIS j\tkf_. - SENSOR f 10TAL ERROR E FUNCTIONAL UNIT Att0WAkCE (TA) Z (S) ! RIP SEIPOINI A_ll0WABLE VALUE d 4. Steam Line Isolation
- a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
e m b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. and Actuation Relays
- c. Containoent Pressure-liigh-liigh 12.7 0.11 1.5 3 3 psig $ 3.2 psig
- d. Steam Lin? Pressure - Low 4.6 1.31 1. 5 > 7Fr ps ig > M+ ps i g\
r
- 775 744 i e. Steam Iine Pressure- 8.0 0.5 0 ~ 103 psi
( < 122.8 psi ** Nega*.ive Rate - liigh , w g S. Feedwater Isolation Y a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N. A. 0; Actuation Relays
- b. Steam Generator Water f y Level-liigh-High (P-14)
. 1. Unit 1 5.4 2.18 1.5 < 82.4% of < 84.2% of narrow narrow range range instrurnent 3, 3 ins t ruenent span , d j! span a a l 2. Unit 2 21.9 1.7 2.0 < 77.1% or ( 78.Yt of narrow I
!" @@ narrew range range instrument l
{# o , instrument span span i
- c. I,yg-Low 4.0(6.0 ) 1.12(0.71 ) 1.2(0.8 ) 3 564*f 3-562f(561*fk l
i [ y$
- d. Doghouse Water Level-High 1.0 0 0.5 11 inches above Si7' 12 inches at+ove 577'
- ft i floor level floor level 1
hi e. Safety injection See item 1. at;ove for all Safety injection Setpoints and Allowable Values. l ) i
u s>z T \ n TABtE 3.3-4 (Continued)
-e 5 ENGINEERED SAFETY FEAIURES ACTUATION SYSTEM INSTRUMENIAll0N IRIP SEIPOINTS 6
c Mc\t 101AL SENSOR ERROR 5 IUNCTIONAL UNil ALLOWANCE (IA) Z (S) IRIP SLIPOINI Att0WAUll VALUE d 6. Turbine Trip [ a. Manual Initjation N.A. N.A. H.A. N.A. N.A. N b. Automatic Actuation N.A. N.A. N.A. N.A. N.A. Logic and Actuation Relays
- c. Steam Generator Water Level-High-High (P-14)
- 1. Unit 1 S.4 2.18 1.5 < 82.4% of < 84.2% of narrow narrow range range instrument y instrument span g span
, 3 2. Unit 2 21.9 1.7 2.0 $ 77.1% of $ 78.9% of narrew o narrow range range instrument ins trumerit span x f span I d. Trip of All Main N.A. N.A. N.A. N.A. N. A. Feedwater Pumps yy e. Reactor Trip (P-4) H.A. H.A. N.A. N.A. N.A. o,o "
- f. Safety Injection See Item 1. above for all Safety Injection Setpcints and Allowable Values.
%["
4 7. Containment Pressure Control oy System 9A <c\-c. ( e t - a. Start Permissive N.A. N.A. N.A. $ 0.4 psid $ U.45 psid g1 b. Termination N.A. N.A. N.A. 3 0.3 psid 3 0..?5 psid 4 -
- 8. Auxiliary Feedwater I g a. Manual Initiation N.A N.A. N.A. N.A N.A.
I
- b. Automatic Actuation Logic it. A. N. A. N.A. N.A. N.A.
and Actuation Relays
vozT t c, IABLE 3.3-4 (Continued) TNGINEEl:EU SAFELY FEA1URES AClllATION SYSTEM INSTRUNENTAIION IRIP SETPOINE E SENSG C N h g "10lAL ERROR [FUNCflHNALUNIT ALLOWANCF (TA) Z fS) TRIP SEIP0lHT A_LLGWABLE VAluE
%8. Auxiliary Feedwater (Continued)
- c. Steam Generator Later Level - Low-low [
m 1) Unit 1 17 14.2 1.5 > 17% of span > 15.3% of from 0% to ipan from 0% to 30% RTP 30% RTP increasing increasing linearly to linearly to > 38.3% of sprin
> 40.0% of Trom 30% to 100%
ipan from 30% RIP w s to 100% RIP
" 2) Unit 2 11.8 1. 7 2.0 > 36.8% of > 35.1% of narr ew l g, ,
'riarrow range Fange instrtiwant span spaa
- d. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
- e. Loss-of-Of fsite Power IN.A. N.A. N. A.) > 3500 V > 3200 V
- f. Trip of All Main feedwater' Pumps N.A. N.A. N.A. N.A. N.A.
Nj g. Auxiliary Feedwater Suction Qgk fg Pressure-Low
- 1) CAPS 5220, 5221, 5222 > 10.5 psig > 9.5 psig 3f zg 2) CAPS 5230, 5231, 5232 N.A.
N.A. N.A. N.A. H.A. N.A. > 6.2 psig > 5.2 psig
- a. Unit 1 N.A. N.A. N.A. i 6.2 psig i 5.2 psig
$$ b. Unit 2 N.A. N.A. N.A. [6.0psig [5.0psig h9.
i EL 3. Containment Sump Recirculation
- a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
l and Actuation Relays i~ b. Refueling Water Storage {N.A. N.A. N.A. > 177.15 inches > 162.4 inches l Tank Level-Low l Coincident with safety < In ject ion See item 1. above for all Safety injection Setpoints and Allowable Values. <
u d.r T \ i
' TABLE 3.3-4 (Continued)
S Y ENGINEERED SAFETY FEATURES ACIllATION SYSTEM INSTRtMENTATION TRIP SETPOINTS g I___ _ SENSOR ,
' O-(-' TOTAL ERROR 1 RIP SETPOINT Att0WABLE VAlUE j s) p -
illNCTIONAL llHIT ALLOWANCE (TA) Z I
- 10. Loss of Power
[ N.A. '" a. 4 kV Bus Unilervoltage-Loss .N.A. N.A. -> 3500 V -> 3200 V '" of Voltage
- b. 4 kV Bus Undervoltage- N.A. N.A. N.A. t 3685 V 3 3611 V Grid Degraded Voltage
- 11. Control Room Area Ventilation I Operation
- a. Automatic Act4ation Logic '
H.A. N.A. H.A. N.A.
& and Actustion Relays N.A.
O N.A. N.A. 1 3200 V
" b. Loss-of-Of f site Power 3 3500 V (N.A.
x
- c. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable t
Val ue s.. 1
- 12. Containment Air Return and Hydrogen Skimmer Operation g{gh 7 N.A. N.A. N.A. N.A. N.A.
- a. Manual Initiation N.A. N.A. N.A. N.A.
- b. Automatic Actuation Logic N.A.
and Actuation Relays
- c. Containment Pressure- 12.7 0.71 1.5 -< 3 psig -< 3.2 psig liigh-High t
/
tinz T I I Allt 1 3. 3-4 (Cunt inued) (NtilN11Hl0 Salt lY ll AIURLS ACillA110H SYSilH INSlHUHlHI A110N IRIP S[lPulNIS 5 1 ,
- > %g4C SINSOR
^ 10lAl I RROR lilNCil0NAl llHil All0WANCf (IA) / (S) IRIP SIIP0lNI AL10WAlil[ VALUL ) , d i 1. Aninatus Ventilation Ope ation s e
- e. a. Hanual IniLiation N.A N.A. N.A. j 11. A. N.A.
- to
- b. Automatic Actuation logic H.A. fl. A. N.A. N.A. N.A i
and Actuation Relays L j j
- c. Safety injection See Item 1. above for all Sately injection Setpoints and Allowable Values.
- 14. Nuclear Service Water Operation
. T)clc b E Manual Initiation N.A. H.A. N.A. H.A. H.A.
R a.
$ Automatic Attuation logic H.A. N.A. N.A. H.A. !
T b. N . A.. O and Actuation Relays
- c. Ioss of-0ffsite Power , N.A. fl. A. H. A. i >
_ 3S00 V > 3200 V b u _ _ _._ --- 1
- d. Containment Spray See item 2. above for all Containment Spray Setpoints and Allowable Values.
- e. Phase "B" Isolationi See item 3.b. above for all Pliase "B" Isolation Setpoints avid Allowable Values.
- f. Safety injection See item 1. above for all Safety injection Setpoints and Allowable Values.
NA >[1. 554.4 ft. >El. 552.9 ft. Suction Iransfer-Low Pit level i
- g. NA NA
- 15. Emergency Diesel Generator Operation (Diesel fluilding Q e_. i Ventilation Operation, Nuclear Service Water Operation) j H.A. N.A. H.A. H.A.
- a. Mdilual Iailiatioll H.A
- 1
v aJIT l I Alll t 3. 3-4 (Continueil}
$ [NGlHi[RLD SAFELY llAltlRLS ACillAll0N SYSIIH INSIRilHINIAll0N 1 RIP SEIPOINIS i5 SENSOR MC 101AL ERROR ~
E filNCIIONAl IINil All0WANCE (IAl 2 (S)_ IRIP SEIPOINI Att0VABIE VA!UE ,
.-4 !
- 15. Imergency Diesel Generator I
" Operalion (Diesel Building " Ventilation Operation, Nuclear- ,
Service Water Operation) (Continued)
- b. Automatic Actuation Logic H.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- c. loss of-Offsite Power N.A. H.A. H.A. 3 3500 V 3 3200 V it. Safety injection See item. I above for all Safety injection Setpoints and Allowable Values.
o y if>. Auxiliary Building IiItered Exhaust Operation - Q.g , ; x a. Hanual Initiation N.A. H.A. H.A. H.A. N.A. S
- b. Automatic Actuation Logic H.A. H.A. H. A. k H.A. H.A.
and Actuation Relays , j ! See item 1. above for all Safety injection Setpoints and Allowable values.
- c. Safety Injecticn II. Diesel liuililing Ventilation Operation gM e
- d. Hanual Initiation N.A. H.A. N.A. H.A. N.A.
ti . Automatic Actuation logic ti. A. N.A. H. A. H.A. N.A. . { and Actuation Relays i ;
- c. Emergency Diesel Generator See item 15. above'foi all Emergency Diesel Generator Operation Operation Setpoints add Allowable Values.
(
v iJ r T I
-IABLE 3.3-4 (Continued) n D ENGINEERED SAFETY FEAltlRES ACTUATION SYSTEM INSTRUMENIATION TRIP SETPOINIS E SENSOR
$ g\ckt 10TAL ERROR AttOWANCE (IA) Z (5) IRIP SEIPOINT AtLOWABlL VAltll g iUNCil0NAL UNIT w 18. Engineered Safety Features e Actuation System Interlocks j s~ 1955 psig 31944 psig Pressurizer Pressure, P-ll H.A. N.A. H.A.
1 m a. 3 H.A. N.A. 1955 psig <1966 psig
- b. Pressurizer Pressure, not P-11' N.A.
- c. N.A. N.A. N.A. 3553*F >551"F(550Ek Low-Low T,yg, P-12 N.A. N.A. N.A. N.A.
- d. Reactor Trip, P-4 N --
L. A.
- e. Steam Generator Level, P-14 See item S. above for all Steam Generator Water Level Irip detraints w
and Allowable Values. n [
?
El e 2:n t+ il# f.x ta b q c____. . _ _ _ _ - n
~
l UU.I T \ TABLE 3.34(Continu_eg TABLE NOTATIOf45
- Time con:stants utilized in the lead-lag controller for Steam Line Pressure-Low are i t 1 50 seconds and 121 5 seconds. Channel calibration shall ensure tnat these time constants are acjusted to these values.
**The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate-High is greater than or equal to 50 seconas.
Channel calibration shall ensure that inis time constant is adjusted to this value.
--#AD,+ estrie-timm-celetiofr-ttf RTD O f p655 y5temr-s l
l l l 1 l l l CATAWBA - UNITS 1 & 2 3/M3-36 haandmentJor--o-(Uni t -1}-
.Ame noment4k>t-;+(Unit-2 F
L)fdXY L (Ault 3.3-4 9 . . O, _litGINii Rib sal t lY l( Allil4LS ACitlAl_ l3H SYSIEH INSIRUH.lNI A!!ON IR_IP SEIP_0lNI_ S en C C- SlHSOR E !GIAl LRROR Q lilHCil0NAi ilNil All0WANCE (IA} I (S) _ IRIP SLIPOINI All0WABIL VAltll
}
" 1. Safety injection (Reactor trip,
" Phase "A" Isolation, feedwater .
" Isolation, Control Room Area Vent ilatio:i Operat ion, Auxiliary feedwater-Hotor-0 riven Pump, Purge & [xtiaust Isolation, Annulus '
Ventit. . ion Operation, i Auxiliary liuilding filtered Extiaus*. Operation, Emergency Diesel u Generator Operation, Comiponent 2 Cooling Water, Turbine Trip, and P/ Huclear Service Water Operation) E
" a. Hasiual latitiatieri N.A. H.A. N.A. H.A. H.A.
, b b. Automatic Actuation Logic H.A. N.A. N.A. H.A. H.4. dild Actuation Relays
- c. Contaisuperit Pressure-liigti 8.2 0.71 1. 5 5 1.2 psig $ 1.4 psig
- d. Pressuriter Prest,ure-Iow 16.I 14.4 1.S 3 1845 psig 3 Ill39 psig
- e. Steom line Pressuse-low 4.6 1.31 1. 5 3 125 psig 3 694 psig a
- 2. Coiitaisuiiesit Spray
- a. Haintial Iris tiatioin N.A. H.A. H.A. H.A. N.A.
t
- b. Autosaatic Actuatiori logic H.A. N.A. N.A. N A. N.A.
I and Actuation Relays
- c. Contaisitnerit. P essiere-liii.jli-liigti 12.7 0.11 1., 5 $ 3 psig $ 3.2 psig l
i
UA T ~Z_ I Alll E 3. 3-4 (Continued} INGINiililD sal [lY ll AltlHES ACI,IlAll0N SYSDH INSiltilHtHIA110N 1 RIP $[lPOINIS S[HSOR h De kC ' ERROH
, 101AL flINCliONAL UNil Atl0WANCE (TA} Z _{S} IRiP SFIP0lHI Atl0WAlllE VALUE g ,
h 3. Contairunent isolationi r a. I'liase "A" isolation m
- 1) Hanisal Initiation H.A. H.A. H.A. H.A. H.A.
- 2) Automatic Actuation logic H.A. H.A. H.A. H.A. N.A.
and Actuation Relays _
- 3) Salety injection See Ites 1. above for all Safety injection Setpoints and Allowalite Valines.
R' b. Phase "!!" Isolation (Nuclear yddC- _ & Service Water Operation) T H.A.
- 1) Hanual initiaLion i N.A. H.A. H.A. H.A.
E$ 5 H.A. N.A. N.A. H.A. H.A.
- 2) Automatic Actisation logic and Actuation
- Itelays
- 3) Centainment Pressure- l?.7 0.71 1.5 1 3 psig 5 3.2 psig liigh-iligh
- c. Purgeand[xtiainstIsojation I) Haliinal liiit iat ioni H.A. H.A. N.A. H.A. H.A.
- 2) Automatic Acttialion H.A. H.A. it. A. H.A. H.A.
logic anti Acttiation j Hejays #
- 3) Safety injection See item 1. above fan all Salety injection Setpoints asid Allowable Values.
J
UdIT q_ IABLE 3.3-4 (Continued) 9 d ENGINEtRE0 SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION IRIP SEIPOINIS 6 2 ~
- h I c~ 10TAL SENSOR LRROR FUNCTIONAL UNIT Att0VANCE (IA) Z- (S) IRIP SEIP0lHi AtLOWABtE VAiHL
_E d 4. Steam Line isolation tr=
- a. Manual Initiation N.A. N. A. N.A. H.A. H.A.
1 m b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. and Actuatio'n Relays
- c. Containment Pressure-High-liigh 12.7 0.71 1.5 5 3 psig < 3.2 psig
- d. Steam Line Pressure - Low 4.6 1.31 1. 5 > 725 psig > 694 psig" ,
- e. Steam Line Pressure- 8.0 0.S 0 < 100 psi < 122. 8 ps i" w
Negative Rate - liigh ~ ~ k 5. Feedwater Isolation
- a. Autornatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
i3 Actuation Relays
- b. Steam Generatcr Water i g Level-liigh-High (P-14)
- 1. Unit 1 5.4 2.18 1. 5 < 82.4% of < 84.2% of narrow inarrow range range instrument
- ,:n a n instrument span
, <><> span au R 2. Unit 2 21.9 1. 7 2.0 < 77.1% of g[ 3 l narrow range
< 78.9% of narrow l range instrument TT instrument span j ??T span
- c. # # #
T avg . w 4.0(6.0 ) 1.12(0.71 ) 1.2(0.8 ) > 564 i > -%2SFf 561*ik m d. Doghouse Water Level-liigh 1.0 0 0.5 11 inches 12 inches above 577' above 577* gh - i floor level floor level d4 e. Safety Injection See item 1. above for all Safety injection Setpoints and Allowatile Values. I l w _ _ _ _. . - _ _ _ _
vuIT 7 n I ABL E 3. 3- 4 (Continued) U g ENGINEERED SAFE 1Y FEATURES ACIUAll0N SYSIEM INSIRUMENIA110N 1 RIP SEIPOINIS
--- _. ~
, gge StNs0R c 101AL ERROR z It!NCIl0NAL UNil Att0WANCE (IA) / (S) IRIP SEIPOINI At10WAHtt VAlut v1 6. Turbine Trip
[ a. Manual initiation il 1. H.A. N.A. N.A. N.A. N b. Automatic Actuation N.A. N.A. N.A. H.A. N.A. ! Eogic and Actuation Relays
- c. Steam Generator Water Level-Nigh-High (P-14)
- 1. Unit 1 5.4 2.18 1.5 < 82.4% of < 84.2% of narrow narrow range fange instrument y ins t ruaient span g span
[ 2. Unit 2 21.9 1. 7 2.0 $ 77.1% of $ 18.9% of narrow o narrow range range Instrument instrument span
?
w span
- d. Irip of All Main N.A. N.A. N.A. N.A. N. A.
Feedwater Pumps 2b e. Reactor Trip (P-4) N.A. N.A. N.A. N.A. N. A.
~ __-
ll gg f. Safety injection See Item 1. above for all Safety injection Setpoints and Allowable gg Values. ,[ [U /. Containment Pressure Control 9c\cke. '? System 7 '$ O a. Start Permissive N.A. N.A. N.A. 5 0.4 psid 1 0.45 psid
- b. Termination N.A. N.A. N. A. ' > 0.3 psid 3 0.25 psid
{.4 8. Auxiliary feedwater I j my a. Manual Initiation N.A N.A. N.A. N.A N.A.
- b. Automatic Actuation togic N.A. N.A. N.A. H.A. N.A.
j and Actuation Relays j
V 9J I 'T ~2. n IABLE 3.3-4 (Continued) lh ENGINtikl0 SAFELY FEATURES ACTUATION SYSTEM INSTRUM[NIAIION 1 RIP SEIPOINIS
.h QQF 101AL 5ENSOR D ERROR
[FUNC110NALUNIT All0WANCE (TA) Z (S}_ TRIP SETPOINT ALLOWABLE VALUE i (8. Auxiliary feedwater (Continued) i '
- c. Steam Generator Later l Level - Low-Low f
} m 1) Unit 1 17 14.2 1. 5 > 17% of span > 15.3% of ' I Trom 0% to- span from 0% to r j 30% NTP 30% RIP increasing increasing linearly to i linearly to > 38.3% of span !
> 40.ist of Trom 30% to 30nt span from 30% RIP j -g to 100% RIP !
I d5 2) Unit 2 11.8 1.7 2.0 > 36.8% of > 35.1% of narrow
. range instrument l l [ , .
narrow range i s '- ~ span span .; i d. Safety injection See Ites 1. above for all Safety Injection Setpoints and Allowable Values.
.h e. loss-of-Offsite Power "N.A. N.A. h. A. ) 3'3500 V > 3200 V I
- f. Irip of'All Main feedwater l , j, Pumps N.A. N.A. N.A. .N.A. H.A.
l-l g. Auxiliary feedwater Suction Pressure-Low Qdg j g. '{g 1) CAPS 5220, 5221, 5222 N.A. N.A. N.A. > 10.5 psig > 9.5 psig ) e:- 2) CAPS 5230, 5231, 5232 N.A. N.A. N.A. > 6.2 psig > 5.2 psig i 'F a. Unit 1 H.A. N.A. N.A. i 6.2 psig i 5.2 psig !Z b. Unit 2 N.A. H.A. N. A. i 6.0 psig i 5.0 psig p 9. Containment. Sump Recirculation
- lE a. Automatic Actuation Logic-~ N.A. N.A. N.A. N.A. N.A.
i (( and Actuation Relays 4
- u. Refueling Water Storage' Tank Level-tow i N.A. N.A. N A.
3 177.15 inches > 152.4 inches i Coincident With Sifety , injection See item 1. above for all Safety injection ~5etpoints and Allowable Values.
unrT 2. i
'IABIE 3.3-4 (Continued)
C
$ ENGINEERED SAFEIY FEATURES AClllAll0N SYSTEH INSIRUMENIATION TRIP SE1POINIS
~
lg"
' Ck I
10TAL SENSOR ERROR All0WANCE (TA) Z (S)_ j 1 RIP SEIPlHT All0WABtE VAtHE E lilHCIIONAL UNIT I 4 ye
- 10. toss of Power I
- 4 kV Bus Undervoltage-Loss N.A. N.A. N.A. 3 3500 V 3 3200 V a.
' " of Voltage
- b. 4 kV Bits Undervoltage- N.A. N.A. N.A. 1 3685 V 13611 V Grid Degraded Voltage
- 11. Control Room Area Ventilation ;
Operation !
- a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
9? and Actuation Relays c ',
" Loss-of-Offsite Power N.A. N.A. N.A. -> 3500 V -> 3200 V
- b. ,
- c. Safety Injection See Item 1. above for all Safety injection Setpoints and Allowable 5
- Values.
- 12. Containment Air Return and Hydrogen Skimmer Operation gjghg m -7 N.A. N.A. N.A. N.A. N.A.
- a. Manual Initiation Automatic Actuation Logic H.A. H.A. H.A. N.A. H.A.
- b. i and Actuation Relays 12.7 0.71 1.5 < 3 psig < 3.2 psig
- c. Containment Pressure- - ~
liigh-High l i l W- - - - -- - - - - - - . _ . - - . -
vca.rT L IA0lt 3.3-4 (Continued) b (NGINfililli SAllIY l[ AltlRLS AClllA110N SYSilH lilSIRIMINI A110N IRIP SEIPOINIS
~
$ pe\g\c SINSOR
. -10lAl t ilROR illNCil0NAl. IINil All0WANCE (1,A_} I (S) IRIP SilPollll All0WAlli t VAltit ;
g (,4 I 1. Annulus ventilation Operation !
- a. HanuaI initiation N.A. N.A. H.A. N.A. H.A.
N.A. ;
- b. Automatic Actuation Logic N.A. N.A. N.A.
and Actuation Relays [ H.A. ' See item 1. above for all Safety injection Setpoints and Allowable Values.
- c. Salety injection
- 14. Nuclear Service Water Operation T)clrI C R a. Hanual Initiatiun N.A. N.A. N.A. N.A. H.A.
n
@. h. Automatic Actuation logic N.A. iL A. H.A. N.A. N.A.
U and Actuation Relays ,
- c. Ioss ol-0Iisite Power H.A. N.A. N. A. 1 > 3500 V > 3200 V t __ .-
- - - - _ 1
- d. Contaisunent Spray See item 2. above for all Containment Spray Setpoints and Allowable Values. ,
- e. Phase "ll" Isolation See item 3.b. above for all Phase "B" Isolation Setpoints aiid Allo,<able Values,
- f. Safety injection See item 1. above for all Safety injection Setpoints and Allowable Values. i 9 Suction Iransfer-tow Pit level NA NA NA gli. $$4.4 ft. >Lt. SS2.*J tt. t
- 15. Imergency Diesei Generator Q g-k-e__
Operation (Diese! Building l Ventilatina Operation, Nuclear Service Water Operation)
- d. Na4Hai Initiation N.A N.A. N.A. H.A. h.A.
~ ;
k _
vAJIT 7 1AlliL 3.3-4 (ConLinned} 20
$ f MyHtf Hill SAFE 1Y fl All!RES AClllATION SYSl[H INSIRllHlHIAll0N 1 RIP SEIP0lHIS g 10lAL SENSOR ERROR gl0WANCE(LA) I _{ S ) _ IRIP SEIPOINT All0WABLE VAllit 5telCIIDHAL llNii
- Imergency Diesel Generatur !
1peration (Diese! Suilding
" Ventilation Opera Won, Nuclear
" Service Water OpeNtion) (Continued) .
- b. Automatic Actuation Logic H.A. H.A. H.A. H.A. H.A.
and Actuation Relays f
- c. loss of-OfIsite Power H.A. H.A. H.A. 1 3590 V 3 3200 V :
w d. Safety injection See Ites. I above for all Safety injection Setpoints asid Allowable Values. t w if>. Auxiliary lluilding i iltered {C aa Exhaust Operation 7 oc a. Hanual initiaLios H.A. H.A. H.A. N.A. H.A. i M
- b. Automatic Attuation Logic N.A. 11. A. H. A. { H.A. H.A.
and Actuation Relays _ j
- c. Safety injection See item 1. above for all Safety injection Setpoints and Allowable Values. ,
ll. Diesel ibailding Ventilation Operation f Q
- d. Hanual initiatinn H.A. H.A. N.A. H.A. N.A.
- b. Automatic Actualion logic N.A. H. A. N. A. H.A.
and Actuation Relays H.A. ( L -
- c. Imergency liiesel Generator See item 15. above for all Emergency Diesel Generator Operation Operation Setpoints add Allowable values.
i E
vs.r T v._ IABLE 3.3-4 (Continueo) n D w ENGINEERED SAFEIY FEAlllRES AClVATION SYSTEM INSIRUMENIATION 1 RIP SEIPOINIS e g(M 10TAL SENSOR ERROR g FUNCTIONAL UNIT ALIOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE
- 18. Engineered Safety Features i
e Actuation System Interlocks u. eu a. Pressurizer Pressure, P-11 H. A. N.A. N.A. 1955 psig >1944 psig
- b. Pressurizer Pressure, not P-ll N.A. H.A. H.A. 1955 psig <1966 psig N.A. H.A.
4
- c. Low-Low T,yg, P-12 H.A. 3553*F 3%1 fiS50*F A
- d. Reactor Trip, P-4 N H.A. N.A. N.A. _
N.A. L . A. -
- e. Steam Generator Level P-14 See item 5. above for all Steam Generator Water Level Trip Setpoints w and Allowable Values.
D 2l 5 in a 88 88 i "7 E .! ah$ 4 ! EE l 3;; 1 et l l
LJ u r T 2. TABLE 3.3-4 (Continued) TABLE NOTATIONS
- Time constants utilized in the lead-lag controller 'or Steam Line Pressure-tow are It > 50 seconos and T2 < 5 seconds. Channel calibration shall ensure that these time constants are adjusted to these values.
**The time constant utilized in tne rate-lag controller for Steam Line Pressure-Negative Rate-High is greater than or equal to 50 seconds.
Channel calibration shall ensure that this time constant is adjusted to this value.
-#ADP44&le 40n deletien tif RID Bypass Oj;tc.T l
l l 8-80 CATAWBA - UNITS 1 & 2 3/463-36 44mendment No. - 40 (Unit 1)
-Amenement- Nor-- 3 3 -(Uni t-- 2 )
LJ U 1 T~ l TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES, INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS
- 1. Manual Initiation
- a. Safety jection (ECCS) N.A.
- b. ' Containment Spray N.A.
- c. Phase "A" Isolation N.A.
- d. Phase "B" Isolation N.A. ,
- e. Purge and Exhaust Isolation N.A.
- f. Steam Line Isolation N.A.
- g. Diesel Building Ventilation Optration N.A.
- h. Nu' clear Service Water Operation N.A.
- f. Tureine Trip N.A.
- j. . Component Cooling Water N.A.
- k. Annulus Ventilation Operation N.A.
- 1. Auxiliary Suilding Filtered Exhaust N.A.
Operation
- m. Reactor Trip N.A.
- n. Emergency Olesel Generator Operation- N.A.
- o. Containment Air Return and Hydrogen Skimmer Operation N.A.
- p. Auxiliary Feedwater N.A.
l 2. Containment Pressure-High
- a. Safety-Ir Nction (ECCS) < 27(1)/12(3) l 1) Reactor Trip 12 L -2) Feecwater Isolation . < % l'2.
- 3) Phase "A" Isolation (2) 18(3)/28C#)
- 4) Durge and Exhaust Isolation 56
~5) auxiliary Feeawater(5) N.A.
6)( Nuclear Service Water Operation .__1 65(3)/76(#)
- 7) Turbine Trip N.A.
8)- Comoonent Cooling Water i 65(3)/76(4) ! 9) Emergency Diesel Generator Operation 1 11' 10). Control Room trea Ventilation 00eration N.A. 8-81 CATAWBA --UNITS 1'& 2 3/4A3-37
U A).I T l TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITTATING SIGNAL AND FUNCTION RESPONSE TIME IN SECON05
- 2. Containment Pressure-High (Continued) u) Annulus Ventilation Operation i 23
- 12) Auxiliary Building Filtered N.A.
Exhaust Operation
- 13) Containment Sump Recirculation N.A.
- 3. Pressurizer Pressure-Low
- a. Safety Injection (ECCS) 1 27(1)/12(3)
- 1) Reactor Trip i2
- 2) Feeowater Isolation < 1 l'2
- 3) Phase "A" Isolation (2) {1g(3)/2S(4)
- 4) Purge and Exhaust Isolation 16
- 5) Auxiliary Feedwater(5) y,3,
- 6) Nuclear Service Water Operation 5 65(3)/76(#)
- 7) Turbine Trip N.A.
- 8) Component Cooling Water 1 65(3)/76I#)
- 9) Emergency Ofesel Generator Operation i 11
- 10) Control Room Area Ventilation N.A.
Operation
- 11) Annulus Ventilation Operation 1 23
- 12) Auxiliary Building Filtared N.A.
~ Exhaust Operation
- 13) Containment Sumo Recirculation N.A.
4 Steam Line Pressure-Low
- a. Safety Injection (ECCS) 3 12(3)/22C#)
- 1) Reactor Trip i2
- 2) Feoowater Isolation < K 12.
3)- Phase "A" Isolation (2) 18I )/2S(#)
- 4) Purge and Exhaust Isolation <6
- 5) Auxiliary Feeowater(5) 60
- 6) Nuclear Service Water Operation 65(3)/76(#)
- 7) 7urcine Trip N.A.
- 8) Comoonent Cooling W'ater 5 65(3)/76(#)
- 9) E=ergency Diesel Generator Operation 1 11 8-82 CATAWBA - UNITS 1 & 2- 3/4A3-38
. . _ _ _ . , ~ _ -_ , _ . _ . . - . . _ . ._ ._._ _ . - - .-_
vALIT l TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 4. Steam Line Pressure-Low (Continued)
- 10) Control Room Area Ventilation N.A.
Operation
- 11) A7nulus Ventilation Operation i 23
- 12) Auxiliary Building Filtered Exhaust Isolation N.A.
- 13) Containment Sump Recirculation N.A.
- b. Steam Line Isolation il 10
- 5. Containment Pressure-High-High
- a. Containment Spray < 45
- b. Phase "B" Isolation k65(3)/76(4)
Nuclear Service Water Operation N.A. ~
- c. Steam Line Isolation i K 10
- d. Containment Air Return and Hydrogen < 600 Skimmer Operation i
- 6. Steam Line Pressure - Negative Rate-High Steam Line Isolation i 110
- 7. Steam Generator Water Level-High-High
- a. Turbine Trip <3 ,
- b. Feedwater Isolation 1 117.-
I
- 8. T,yg-Low Feedwater Isolation N.A.
- 9. Doghouse Water Level-High Feedwater Isolation N.A.
10 Start Permissive Containment Pressure Control System N.A. l 11. Termination Containment Pressure Control Systen: N.A. 8-83 CATAWBA - UNITS 1 & 2 3/443-39
L)/ JCT \ TABLE 3.3-5(Continuedl ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION FSPONSETIME,INSECONDS
- 12. Steam Generator Water Level-Low-Low
- a. Motor-Driven Auxiliary Feedwater Pumps 5 60
- b. Turbine-Oriven Auxiliary Feecwater Pump i 60
- 13. Less cf-Offsite Power
- a. Motor-0 riven Auxiliary Feedwater Pumps 1 60
- b. Turbine-Oriven Auxiliary Feedwater Pumps < 60
- c. Control Room Area Ventilatien Operation N.A.
- d. Emergency Diesel Generator Operation i 11
- 1) Diesel Building Ventilation Operation N.A.
- 2) Nuclear Service Water Operation 1 65(3)/76(4)
- 14. Trip of All Main Feedwater Pumps
- a. Motor-0 riven Auxiliary Feedwater Pumps 1 60
- b. Turbine Trip N.A.
- 15. Auxiliary Feedwater Suction Pressure-Low Auxiliary Feedwater (Suction Supply -< 16(6)
Automatic Realignment)
- 16. Refueling Water Siorage Tank Level-Low Coincident with Safety Injection Signal (Automatic Switchover to Containment Sump) 1 60
- 17. Loss of Power
- a. 4 kV Bus Undervoltage - -< 8.5 Loss of Voltage
- b. 4 kV Bus Undervoltage- -< 600 Grid Degraded Voltage
- 18. Suction Transfer-Low Pit Level Nuclear Service Water Operation N.A.
l l l l 8-84 I CATAWBA - UNITS 1 & 2 3/4A 3-40 Amenemeat-No-60-(Unit-1$ )
-Amendment-No. 54 (Unit-f)
v u,I T 7. TABLE 3.3-5 ENGINEERED SAFETY FEATURE 5 RESPONSE TIMES INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECCNDS
- 1. Manual Initiation
- a. Safety Injection (ECCS) N.A.
- b. Containment Spray N.A.
- c. Phase "A" Isolation N.A.
- d. Phase "B" Isciation N.A.
- e. Purge and Exhaust Isolation N.A.
- f. Steam Line I:olation N. A.
- g. Diesel Building Ventilation Operation N.A.
- n. Nuclear Service Water Operation N.A.
- i. Turcine Trip N.A.
- j. Comconent Cooling Water N. A.
- k. Annulus Ventilation Operation N.A.
1 Auxiliary Building Filtered Exhaust N.A. Operation
- m. Reactor Trip N.A.
- n. Emergency Diesel Generator Operation N.A.
- o. Containment Air Return and Hydrogen Skimmer Operation N.A,
- p. Auxiliary Feedwater N.A.
- 2. Containment Pressure-Hign
- a. Safety inject m (ECCS) < 27(1)/12(3)
- 1) Reactor Trip 52 l
- 2) Feeawater Isolatiu <7 l
i 3' Phase "A" Isolat, ion' ) - 18( )/28(4) l: a) Purge and Exhaust' Isolation 16
- 5) Auxiliary Feecwater(5) g , 4,
- 5) Nuclear Service Water Operation i 65(3)/76(#)
- 7) Turaine Trip N.A.
- 8) Comconent Coolir.g Water 1 55(3)/76(4)
- 9) Emergency Diesel Generator Operation 5 11
- 10) Contro' Room Area Ventilation Ooeration N. A.
CATAWBA - UNITS 1 & 2 3/483-37 8-85
v A.C T 1-TABLE 3.3-5 (Continued) ENGINEERED Sv ETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECCN05
- 2. Containment Pressure-High (Continued)
- 11) Annulus Ventilation Operation 1 23
- 12) Auxiliary Building Filtered N.A.
Exhaust Operation
- 13) Containment Sumo Recirculation N.A.
- 3. Pressuri:or Pressure-Low
- a. Safety Injection (ECCS) 1 27(1)/12(3)
- 1) Reactor Trip <2
- 2) Feecwater Isolation <7
- 3) Phase "A" Isolation (2) yg(3)/28( )
- 4) Purge and Exhaust Isolation <6
- 5) Auxiliary Feedwater(5) 3,4,
- 6) Nuclear Service Water Operation 1 65(3)/76(#)
- 7) Turcine Trio N.A.
- 8) Component Cooling Water 1 65(3)/76(#)
') Emergency Diesel Generator Operation 1 11 1G) Control Room Area ventilation N.A.
Operation 1 Annulus Ventilation Operation 1 23
;J: Auxiliary Building Filtered N. A.
Exhaust Operation
- 13) Containment Sumo Recirculation N.A.
4 Steam Line Pressure-Low
- a. Safety Injection (ECCS) 1 12(3)/22C4)
- 1) Reactor Trip <2 l 2) Feeawater Isolation <7
- 3) Phase "A" Isolation (2) ~
18(3)f;g(4) l
- 4) Purge and Exhaust Isolation <6
- 5) Auxiliary Feecwater(5) 60
- 6) Nuclear Service Water Operation 65(3)/76(#)
- 7) Turoine Trio N.A.
- 8) Component Cooling Water 1 65(3)/76(')
- 9) Emergency Diesel Generator Operation i 11 CATAW8A - UNITS 1 & 2 3/483-38 8-86
f t/MS T 2_ TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 4. Steam Line Pressure-Low (Continued)
- 10) Control Room Area Ventilation N. A. !
Operation
- 11) Annulus Ventilation Operation < 23
- 12) Auxiliary Building Filtered Exhaust Isolation N.A.
l
- 13) Containment Sump Recirculation N.A.
- b. Steam Line Isolation <7
- 5. Containment Pressure-High-High
- a. Containment Spray < 45
- b. Phase "B" Isolation 65(3)/76(4)
Nuclear Service Water Operation E.A.
- c. Steam Line Isolation
_7
- d. Containment Air Return and Hydrogen Skimmer Operation -< 600
- 6. Steam Line Pressure - Negative P. ate-High Steam Line J. solation
_7
- 7. Steam Generator Water Level-High-High
- a. Turbine Trip <3
- b. Feecwater Isolation 7
- 8. T avg
- Feecweter Isolation N.A.
- 9. Doghouse Water Level-High Feedwater Isolation N. A.
- 10. Start Parmissive Containment Pressure Control System N.A.
- 11. Terminatinn Containment Pressure Control System N.A.
CATAFBA - UNITS 1 & 2 3/463,897 l
1 V N .Z T ~2-TABLE 3.3 5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES l l INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 12. Steam Generator Water Level-Low-Low
- a. Motor-Driven Auxiliary Feedwater Pumps 1 60
- b. Turbine-Oriven Auxiliary Feeoweter Pump i 60
- 13. Loss-of-Of f site Power
- a. Motor-Driven Auxiliary Feedwater Pumps 1 60
- b. Turbine-Driven Auxiliary Feedwater Pumps 1 60
- c. Control Room Area Ventilation Operation N.A.
- d. _ Emergency Diesel Generator Operation i 11
- 1) Diesel Building Ventilation Operation N.A.
- 2) Nuclear Service Water Operation 1 65(3)/76(4)
- 14. Trip of All Main _Feedwater Pumps
- a. Motor-Driven Auxiliary Feedwater Pumps 1 60
- b. Turbine Trip N.A.
- 15. Auxiliary Feedwater Suction Pressure-Low Auxiliary Feedwater (Suction Supply 1 16(6)
Automatic Realignment)
- 16. Refueling Water Storage Tank Level-Low Coincident with Safety Injection Signal (Automatic Switchover to Containment Sump) 1 60
- 17. Loss of Power
- a. A kV Bus Undervoltage - -< 8.5 Loss of Voltage
- b. 4 kV Bus Uncervoltage- -< 600 ~
Grid Degraded Voltage
- 18. Suction Transfer-Low Pit Level ,
Nuclear Service Water Operation N.A. t CATAWBA - UNITS 1 & 2 3/483-40 8-88 f uncment No. 60 (UniL 1) f :c a t m e ,t H v. !+ (Unii.2)
._ _ . . . - - ... - - = - - , REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least 'three of the reactor coolant loops listed below shall be OPERAELE and et-4eesrt-two-of-thse-reactee-coolant-4eops-sham-bein operation:"
- a. Reactor Coolant Loop A and its associated steam generator and reector coolant pump,
- b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,
- c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, and
- d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.
APPLICABILITY: MODE 3. ACTION: s
- a. With less than the above required reactor coolant loops OPERABLE, restore ,the required loops to OPERABLE st,atus;within 72 hours or be
^
in A0 T5HUTbOW M ithTrN 4e-nE t 1P nourif. less C bo v e.loorieeperatitrn, t hct n %ce(sant recgu i rK restore e-t icu two b.- W1 AfoMFoneyeacter loops'to operation within 72 hours or open the Reactor T(rip System breakers. O
- c. Withnoreactorcoolantloepsjnoperation,suspendalloperations involving a reduction in borten concentration of the Rear +.or Coolant System and immediately initiate corrective action to - _ turn the required reactor coolant loops to operation SURVEILLANCE REQUIREMENTS bM i.2.1 At least the above required reactor coolant pumps, if not in
-% . operation, shall be determined OPERAB's once per 7 days by verifying j correct breaker alignments _and_ indicated _p ~ower availaci1itv.
~
/-
4.4.1.2./ lThe recuired steam generators shall be determine OPERABLE by verifying secondary side water level to be greater than or equal te 12% at least once per 12 hours.
/
a.4.1.273A At least-two reactor coolant locos shall be verified in operation and circulating reactor coolant at least once Der 12 hours.
"All reactor coolant pumps may be deenergized for up to 1 hour orovidad:
- (1) no operations are permitted that would cause diiution of the Reacr.or Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.
CATAWBA - UNITS 1 & 2 3/4 4-2 8 49
. . . . .= - . -- . . _ . . - . -
REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SHUTOOWN LIMITING CONDITION FOR OPERATIGN 46< a;k 2. 3.4.2.1 A minimum of one pressuri:er Code safety valve shall be OPERABLE with a lift setting of 2485 psigt L%." L 4 3 % '- 7. 7. A r v-;4 i %d APPLICABILITY: MODES 4 ana 5. ACTION: ( With no pressuri:er Code safety valse OPERABLE, immeciately suscend all operations involving positive reactivity enanges anc place an OPERABLE resi-dual heat removal 1000 into operation in the snutdown cooling moce. l SyRVEILLANCE REOUIREMENTS l 4.4.2.1 No additional requirements other than those required by
- Soecification 4.0.5.
1 "The lift setting pressure snall correscond to ameient conditions of the valve at nominal operating temoerature and pressure. CATAWBA - UNITS 1 & 2 3/4 4-7 590
1 [ REACTOR COOLANT SYSTEM , CPERATING LIMITING CONDITION FOR OPERATION 4;r se;4- IL 3.4.2.2 All pressurized Code safety valves snall be OPERABLE with a lift setting of 2485 psig g . Z yg,,_z g 3 , a i , a APPLICABILITY: MODES 1, 2, and 3. ACTION: With one pressurizer Code safety valve inoperable, either restore the inoparable valve to OPERABLE status within 15 minutes or be in at least HOT 5TANDBY within 6 hours anc in at least HOT SHUTDOWN within the following 5 nours. SURVEILLANCE REQUIREMENTS 4.4.2.2 No acditional recuirements otner than those requireo by Scecification 4.0.5.
"The lift setting pressure snall coriespond to amuient conditions of the valve at nominal ocerating temperature and cressure.
3/4 4-8 841 CATAWBA - UNITS 1 & 2
l l REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE J LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
- a. No PRESSURE BOUNDARY LEAKAGE,
- b. 1 gpm UNIDENTIFIED LEAKAGE,
- c. gpm total reactor-to-secondary leakage through all steam generators and.500* gallons per day through any one steam generator, 2.c o
- d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,
- e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant Systein pressure of 2235 1 20 psig, and
- 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 POESSURE BOUNDARY LEAKAGE, be in at least HOT STAND 8Y within 6 hours and in COLD SHUTDOWN within the following 30 hours.
- 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 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. l
- 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 hours by use of at least two closed manual or' deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
l l l l l CATAWBA - UNITS 1 & 2 3/4 4-20 Amenoment-N&.-37(Uni t-1-) __ Amendment No-29(Unit-2)
i 3/4.5 EMERGENCY ::RE COOLING SYSTEMS 3/4.5.1 ACCUMULaTCPS COLD LEG INJECTI:N LIMITING CONDIT*:N CR OPERATION 3.5.1 Each c:lc leg injec' tion accumulator snall be OPERABLE with: !
- a. The cisenarge isolation valve open,
- b. A containec torated water volume of between 7704 ano 8004 gallons,
- c. A boren c:ncentration of between 1900 and 2100 ppm,
- d. A nitr: gen cover pressure of between 535 and 678 psig, and
- A wa ter level and pressure channel OPERABLE.
APPLICABILITY: "CDES 1, 2, anc 3*, ACTICN:
- a. With one cold leg injection accumulator inoperable, except as a result of a c!cseo isolation valve or boron concentration less than 1900 p;m, restcre the inoperable accumulator to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in HOT i SHUTDOWN within the fcilowing 6 hours.
i l b. With one cold leg injection accu =ulator inoperable due to tho l isolation valve being closed, either immeniately coen the isolatien l valve or be in at least HOT' STANDBY within 6 hours and in HOT SHUT CWN l within :ne following 6 hours.
- c. shth cne accumulator inoperacle cue to boron concentration less than 1900 ::m ano:
, 1) The volume weightec average baron c:ncentration of the +hace l
>'44; accumulators 1900 p;m or greater, restore the incoera:ie accumulator to OPERABLE status within 24 hours of the low boren determination or ne in at least HOT STANDBY within the next l 5 hours and reduce Reactor Coolant System pressure to less than l l * *n0 esig within the ic11owing 6 hours.
l
- 2) 7he volume weighted average boren concentration of the 1+w11r
2-i
/ fed _;&Bb#ngaccumulatorslessthan1900ppmbutgreaterthan p;m, restore the inoperaole accumulator to OPERABLE status or return the volume weighted average baron concentration of tne tnree limiting accumulators to greater than 1900 ppm and
" Reactor Cociant System pressure above 1000 psig.
3/4 5-1 CATAWBA - UNITS
- L 2 w ___Amencment-Nors&(Mtat amena.a. mun l
EMERGENCY CORE C: CLING SYSTEMS LIMITING CCNDITICN :CR OPERATICN (Continueo) ACTION: (Continue ) enter ACTICN c.1 sitnin 6 hours of the low ocron cetermination or :e in MOT STAN05Y witnin :ne next 6 hours and recuce Reactor Cociant System pressure to less tnan 1000 psig within tne fol-lowing 6 hours. /780
- 3) The v:lume weightec averag core concentration of the saw e f ; accumulators '.s opm od less, return tne volume
.eigntec average coron concentrat [on of the tnree limiting accumulators to greater tnan.a&67 ppm and enter ACTION c.2 witnin 1 hour of the low coron cetermination or 0a in HOT STANDBY within :nc next 6 hours anc recuce Reactor Coolant System cressure to less than 1000 psig witnin tne following l
$ hours.
SURVEILLANCE REOUIREMENTS . 4.5.1 Each cold leg injection accumulator snall be cemonstrated OPERABLE: l
- a. At least once per 12 hours by:
- 1) Verifying, by the aosence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and i
- 2) Verifying that eacn cold leg injection accumulator isolation valve is open.
- b. At least onca per 31 days anc within 6 hours after eacn solution volume i crease of greater inan or ecual to 75 gallens cy verifying the coron concentration nf the accumulator solution:
- c. At least :nce per 31 days when tne Reactor Coolant System cressure is acove 2300 psig cy verifying that power is removec from the isolation valve coerators on Valves NI54A, NI655. NI76A',' anc NISSS
, anc tnat tne respective circuit breaxcrs are paolockec; sno ! d. At least once per 13 montns by verifying tnat eacn cold leg injection accu =ulator isolation valve coens automatically under eacn of :na icilowing c:noitions:"2
- 1) When an actual or a simulateo Reactor Coolant System pressure signai exceeds tne P-11 (Pressurizer Pressure Block of Safety Injection) Setpoint, and
- 2) Upon receipt of a Safety Injection test signal.
l ** This surveillance need not be performed until prior.to entering HOT STANDBY following tne Unit I refueling. t CATAWBA - UNITS 1 1 2 3/4 5-2 lug ?menement-Nor66-(Uni t-19 4,mencment-Nor 60-(Unit 2)
EFERGENCY CORE COOLING SYSTEd5 SURVEILLANCE RECUIREMENTS (Continued) b) With a simulated or actual Reacter C:clant System pressure signal less than or ecui' to 660 esig the inteti cers will cause the valves to au'.:. 'ically close.
- 2) A visual inscection of the containment sume anc verifying inat the suosystem suction inlets are not restricted by cecris anc that the sump ccmocnents (trasn racks, screens, etc. ) snow no evicence of structural distress or acnormal corrosion.
- e. At least once per 13 months, during shutcown, by:"
- 1) Verifying,that each automatic valve in the flow patn actuates to its correct position on Scfety Injection and Containment Sump Recirculation test signals, anc *
- 2) Verifying that each of the following pu:.1ps start automatically upon receipt of a Safety Injection test signal:
a) Centrifugal charging pump, b) Safety Injection pump, and
- ) Residual heat rataval pump.
- f. By verifying that each of the following pumps develcos the incicatec differential pressure wnen tested pursuant to Specification 4.0.5:
m3
- 1) Centrifugal charging pu::p > J/,6tf osid, Safety Injecticn pumo nul
- 2) 3, .M00 p s i c , and
- 3) Residual heat removal pump 3,165 psid.
1
- g. By verifying the correct position of each electrical anc/or mecnanical stop for the following ECCS throttle valves:
l 1) Within 4 hours following completion of each valve stroking operation or maintenance on the valve wnen the ICC5 suesystems l are required to be OPERABLE, and 1
- 2) At least once per 13 months.
Centrifucal Chargino Pumo Iniection inrettle Safety Iniectien Throttle Valve Numcer Valve Numcer l NI-14 NI-164
- NI-16 NI-166 l NI-18 NI-168 NI-20 NI-170
" ints surveillance need nce be performed until prior to entering HOT SHUTCCW
- following tne Unit One first refueling.
CATAWBA - UNITS 1 AND 2 2/4 5-7 M-95
EHERGENCY CORE CCOL.',NG SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
- n. By performing a flow balance test, during snutdown, following com-pletion of modifications to the ECCS subsystems tnat alter tne suosystem flow characteristics and verifying that:
- 1) For centrifugal cnarging pump lines, with a single pump running:
a) The sum of tne injection line flow rates, excluding the highest flow rate, is greater than or equal to 345 gpm, l and CLO b) The total pumo flow rate is less than or equal to ,565 gpm,
- 2) For Safety Injection pump lines, with a single pumo running:
a) Ine sum of the injection line flow rates, excluding the nighest flow rate, is greater than or equal to 450 gpm, [ and 675 b) The total pump flow rate is less than or equal to fic gpm.
- 3) For residual heat removal pump lines, with a single pump running, the sum of the injection line flow rates is greater than or equal to 3648 gpm.
i t , l l CATAWBA - UNITS 1 & 2 3/4 5-8 Amendment No.~e6 ---(Uni t 1) ! x-% . Amandment-Hor BF(Unit-2)
IAull 3.6-2a , llNil 1 CONTAINHfNT IS01A110H VALVES , s:.o
- HAXIHilH
@ VALVE NilMillij FUNCTION IS0tA110N IIME (s}
U
,, 1. 1%se "A" Isolation
[ 08-571!# Steam Generator lA Blowdown Containment Outside Isolation $10 88-218# Steam Generator IB Blowdown Containment Outside~ Isolation <10 BB-61[l# Steam Generator IC Blowdown Containment Outside Isolat'on 210 B8- 1011# ' Steam Generator 10 Blowdown Containment Outside Isolation 210 BiF56A# $ team Generator IA' Blowdown Containment inside isolation 210 BB-19A# Steam Generator IB Blowdown Containment inside isolation 310 88-60A# Steam Generator IC Olowdown Containment inside Isolation $10 BB-8A# Steam Generator ID'lltowdawn Containment inside Isolation <10 m 811-1488# Steam Generator IA Blowdown Containment isolation Bypass .jl0 2 BB-150D# Steam Generator IB Blowdown Containment isolation Bypass $10 m 88-1498# Steam Generator IC Blowdown Containment isolation Bypass $10 08-1478# Steam Generator 10 Blowdown Centainment it.olation Bypass $10 CA-149# Steam Generator IA Hain feedwate.r to Auxiliary feedwater Nozzle Isolation g rJIl
$ CA-ISO # Steam Generator ID Hain feedwater to Auxiliary feedwater Nozzie Isolation g' PV/4 CA-151# Steam Generator IC Main feedwater to Auxiliary feedwater Nozzle Isolation p n//),
CA-152# Steam Generator ID Hain feedwater to Auxiliary feedwster Hozzle Isolation ff n/4-CA-1BS# Auxiliary Nozzle Temper SGIA ff TJ A
-CA-186# Auxiliary Nozzle Temper SGIB y /1//l CA-187# A>xiliary Nozzle leieper 5Glc p rF4 CA-180# Auxiliary Nozzle Temper SGID <,5' iY /t Cf-60# Steam Generator ID feedwater Containment isolation >5 M A l Cf-51# Steam Generator 10 feedwater Containment isolation 38'oA Cf-42# Steam Generator 10 feedwater Containment isolation gAJ A Cf-33# Steam Generator IA feedwater Containeent Isolation J8/JA
( Cf-90# Steam Generator l A i cedwater Purge Valve 55 l CI-89# Steam Generator 10 feeilwater Purge Valve $5 l Cf -Bfl# 5 team Generator IC feedwater Purge Valve $5 Cf-8/# Steam Generator 11) l'eedwater Purge' Valve $5
- w. . ,
- t L ll ! irl ; [' :: j , ' -{ ! i
] .
s ( WA4Ag4gg 4 ddUo9gg ME 0 0 0 00 UM 0000 1 1 f'5'B 555511 MI I T 6661
<23<~
1
< < < pJi~ g _'5 f1'5 <23<23 X
AN MC I T A n n L O i o i o S I ; n t t ro a a noit l l of t n o o . it ae s s t al m I I al on osi t t l osI a n n n sI t e e no I t n m m oi . t n o' n n it t a tneC i i al - nec a a l e emns t n t n os mnit sI niai o o I iat n C C t at nU t n S t no s s ne E V noC n oC o i t t i em mn L C ei n n n ni A edt U U o i a V edi a - i at - ) disl n n t t n - d N isti o o a no e O snut i i l oC u I tI On t t o C n T u e a a s
. . . . e i A OnnV l l I li1i ed t L rr i i rrrr . di n O yuuy t t e i s o S l ttl n n d ttit CCCC st C I peep e e i nu
( pRR p V V s . T u u t ssss iG SUUS t t u ssss . a N n n O aaaa ss 2 E N VV pppp nn
- M 0 UCCt e e yyyy ii 6 N l VLL n m m t aa I i C e n n n BiBBl rr A C Lddm i i e 3
N nn: a a m nnnnnnnn DD E T N U daai t t n oooooooo tt . L O F n a) n n i a iiiiiiii tt tttttt ii o N B A T C 1 asst eo smmoi ppndC puuCsr C r e t: C i c aaaaaaaa ll lll ll l oooeoooo ssssssss VVYVUU RRRR OOOOtt nn . _ I ePP rup t e p IiIIIiI I PPPP nn T u p' ee N PCC eOp r A'8C0VV U N N p(U U i 0CBA0Ci4 t 11I1 C p A 1II11Il1 tt _ ) NmmUnm m) mmmmnn
. d e
u oo orroirerd tfftt fdfi oo) oe t n e m'm a eeeeeeee m m m m aaaaee aaaaaa mm eeeemm n a i s m ttttttt t tttt nn . i y n n yl nsnt u SSSSSSSS SSSSii aa _. t l rrl ornru r - n puupsuI uO t s nnnnnnnn iiiiiiii n n n iiii nn r'. t t o pl t pI t(t( n aaaaaaaa aaaa6u C ueeu e e MMMMMMMM HMMMCC ( SCRS R R I n i o t l a o s I
"B ### ##~ A6 R " 8A88 A 8 ##79 E 7474 9 2 B ##### 66 B e 3880 2 3 7 012 93 117188 M s 4444 4 4 7 13579111 U a - - - - - - - - - - - - - - - - - - - - -
N h NNNN N N I HMMMMHMM VVVVLL P RRRR R R V SSSSSSSS SSSSWW E V L A . V 2 nDEE5 5C e89 mA g
'i 4 i i1
i l PLANT SYSTEMS l
!!AJyl STEAM LINE ISOLATION VALVES ;
LIMIT!NG CONDITICN FOR OPERATICN 3.7.1.4 Each main steam line isolation valve (MSLIV) shall be OPERABLE. APPLICABILITY: M00ES'1, 2, anc 3. ACTION: MODE 1: With or,e MSLIV inoperable but open, POWER OPERATION may continue provided the inoperaole valve is restored to OPERABLE status within 4 hours; otherwise be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. MODES 2 anc 3: With one MSLIV inocerable, suosecuent coeration in MODE 2 or 3 may proceec provided the isolation valve is maintainea closeo. The provisions of Specification 3.0.4 are not applicaule. Otherwise, be in HOT STANCBY within the next 6 hours anc in HOT SHUTDOWN witnin the following 6 hours. SURVEILLANCE REOUIREMENTS ., 6 s ecc4 s -C.c va4 \ eh 4.7.1.4 Each MSLIV shall be demonstrated OPERABLE by verifying full closure within 5 secones wnen tested oursuant to Specification 4.0.5. The provisicns of Specification 4.0.4 are not applicable for entry into MODE 3. k,r ve. h iL i l l i i CATAWBA - UNIT 5 1 & 2 3/4 7-8 8-99 l
3/4.2 COWER CISTcISUTICN LIMIT 5 (Unit 1) ! SASES The s:ecifications of this secti:n provice assurance of fuel integrity curing C:ncition I (Normai Coeration) and !! (Incicents of Mccerate F ecuency) events cy: (1) maintaining tne talculatec ONBR in the care greater inan or equal to cesign limit CNER curing normal operation and in snort-term transients. anc (2) limiting the fissien gas release, fuel pellet temperature, anc c!accing mecnanical preterties to witnin assumec design criteria. In accition, li,miting the peak linear 00-er censity during Condition I events previces assurance that the initial concitions assumec for the LOCA analyses are met inc the ICC5 accectance criteria are not exceecec. I The cefinitions of certain not cnannel anc ;eaking factors as usec in these scocificatiens are as follows: Fg (X,Y,2) Heat F'ux Hot Channel Factor, is definec as the maximum local heat i flux on :ne surf ace of a fuel roc at core elevation Z diviced by tne ' average fuel rod heat flux, allowing for manufacturing tolerances en fuei ellets and rocs;
~"
(X,Y) Nuclear Enthalpy Rise Hot Channel Facter, is definec as the ratio of j the integral of linear power along tne rod with the nignest integrated power to the average red pcwer. ((:) is cefinec as the normalizec Fq (X,Y,Z) limit for a given core neignt. l 3/4.2.1 AXIAL FLUX OI_FFERENCE-Unit 1 , The limits en AXIAL FLUX OIFFERENCE (AFO) scecified in the CORE OPERATING ! LIMITS REPORT (00LR) ensure that the9F (X,Y,Z) anc the FAH(X,Y) limits are not exceeced curing either normal c;:eration or in the event of xenon redistrib-ution following pcwer changes. The AFD envelope s;:ecified in the COLR has been l acjustec for measurement uncertainty. -- 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR. AND NUCLEAR ENTHALPY RISE n0I CHANNEL FACTOR ~(Unit 1) The limits on neat flux hot channel factor, and nuclear enthalpy rise hot l channel factor ensure that: (1) the design limits on peak local power censity I anc minimum DNBR are not exceeded and (2) in the event of a LOCA the ECCS ac-ceotance criteria are not exceeded. The peaking limits are specified in the CORE OPERATING LIMITS REPORT (COLR) per Specification 6.9.1.9. The heat flux hot channel factor and nuclear enthalpy rise hot channel factor are each measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided: . CATAWBA - UNITO 1 & 2 B 3/4 2-1 -Amencment-No-- 56 (Unit -1)
- 3. im _AmenczentJo. 50 (Alnit4)
u :- _.: .:. __.v... ' 5ASEE HEAT : LUX w0T 09 ANNE; ;4CTOR. AND NUCLEAR ENTWAL V OISE HCT CHANNEL raCTCR (un:t ;; ( :nt:nce:, '
- a. Centr:i ::s in a single grouo mcve togetner witn no incivicua; :e inserti:n ci'fering cy more tnan : 12 ste:s. incicate:, ' :e e
;r u: eman: :: sit::n;
- c. Contr:1 :e ;rcues are secuenced wita overlaccing grcues as cescr :ec in 5:e:::ation 3.1.3.5;
- . The c:ntr:! roc insertion limits of Speciff:stions 3.1.3.5 anc 3.1.3. 5 are mainta.nec; anc
- c. The axial : wer cistribution, excressec in terms of AXIAL FLUX CIFFERENCL, is maintainec witnin the limits.
FaH(X,Y) -ili ce maintainec witnin its limits ;roviced Concittens a. I througn d. acove are maintained. The limits on :ne nuclear enthaley rise not enannel f actor, F2H(X,Y), are specified in tne CCLR as Maximum Allowaole Radial Peaking limits, Octainec ty divicing tne Maximum Allowacle Total Peaking (MAP) limit by the axial peak (AXIAL (X,Y)] for location (X,Y). By cefinition, the Maximum Allowable Racial Peaking Hent thatlimits will . f:r to is ecuivalent Marx-SW fuel, tne ONBR result inwith calculatta a ONBR forFaH(X,Y) a desicn the limiting ea tranbia hb i
,y of .5 anc a limit rc reference axial ocwee snace./'ine Mark-5W MAP limits mayl *## '
Oe a: itec :: CFA f uei, provicea an accrocriate acjustment 'arter j s aanl_iec to Arnuch provice ecuivalence to a 1.49 cesign FaH(X,Y), for tne CFA. This is reflecir-i.n tne MAP limits s ecifiec in the COLS/r Tne relaxac1:n of FAH(X,Y), as a func- , fion :: i n c. nA t run:R aliows enanges in the racial power for all permissicie I control ban ( insertion limits. This relaxation is imolemented by the a: lication of tne following fa:::rs: k = (1 * (1/RRH) (1 - ?)] where k = power f actor multiplier a: plied to the MAP limits t P = THERMAL POVER / RATED THERMAL POWER I RRH is given in tne COLR thd k coM co PC- . da,5g,M g ( FQ'q(X,Y,Z) anc Fa. (X,Y) are measurec periodically, an ccmoarisons to tne allewaole limit are made to provide reasonable assurance 4that the limiting criteria will not be exceeced for operation within tne Technical Specification limits of Sections 2.2 (Limiting Safety Systems Settings), 3.1.3 (Movable Control Assemolies), 3.2.1 (Axial Flux Difference), and 3.2.4 (Quacrant Power Tilt Ratic). A poaxing margin calculation is performed to provide a basis fer decreasing the width of the AFD and f(AI) limits anc for recucing THERMAL F0WER. 1
- CATAWBA - UNITS 1 & 2 8 3/4 2-2 x. jog -Amencment-NorE64 Uni t-1)
-Amencment No. 50 (Unit 2)
for Power. Distribution Limits Bases e Attachment 1: For transitlen ccres, MAP limits may be applied to both Mark-BW and optimited fuel types provided allowances for differences in CNER are accounted for in the generation of MAP limits. The MAP limits specified in the COLR include allowances for mixed ccre CNBR effects. e + l' s Y t-e r 4 l 8 102
POWER DISTRIBU'::N _ MITS BASES HEAT- LUX' HOT CHA n :' raCTCR AND NUCLEAR ENTHALPY RI5E WOT HANNEL FACTOR (Unit 1) (Continue:) i
, -- U\ eerrto q j When an FO(( <,Z) measurement is catainea g in accorcance witn eng surveil-lance' recuiremenu :f Scect ficatien 4.2.2, no uncertainties are accliec : :ne l
- m orea cear 'tne requirec uncertainties are inclucec in the peaking limi:. .
M 44Cd - 3 l When FQ^(X,Y,Z) is ,easurec for reasons otner than meeting tne recuiremerns of i Specification 4.2.2. the measured ceak is increasec by the racial-local peaking factor-t: :: :n : x: :::1-ee-aM. Allowances of 5% for measurement uncer-tainty anc 3% for aqufacturing *.:lerances. r e - :r :::,' 9 *s + -e -e ec .
- N u Lo rs a %*
WhenanF3.Hg(4,Y) measurement is obtainec, %c.. a m _f ;% eason, no uncertainties are'a clied to tne measurec ceakQ,the recuirea uncertainties are ' includec in .ne tea (ing limit. 3/4.2.4 OUADRANT ::WER TILT RATIO (Unit 1) The QUADRANT :0WER TILT RATIO limit assures that the racial power distribu-tien satisfies .e cesign values usec in the power cacability analysis. Radial power cistri:ution measurements are made during STARTUP testing anc
- periodically curing pcwer cperation.
The limit of 1.02, at which corrective action is recuired, provides CNS , anc lineur heat generation rate protection with x y plane ccwer tilts. A , peaking increase inat reflects a QUADRANT POWER TILT RATIO of 1.02 is inclucec in the generation :f the AFD limits. The 2-heur time allowance for coeration with a tilt concition greater tnan ~.02 but less tnan 1.09 is proviced tc allow ider 'fication anc correction of a :!roppec er misalignec control red. In the event :.ucn action coes not correct the tilt, :ne margin fer uncertainty on F (X,Y,Z) q is reinstated by I l reducing the maximum allowec ocwer by 3% for each percent of tilt in excess ; o f 2*.'. i For pur;cses Of monitoring QUADRANT POWER TILT RATIO wnen one excere detector is incoera:le, the movacle incore detectors are used to c:nfirm tnat
.,...._.u... ..__.. 4_ ....:.......2.. ..
. 7:i: nat ica :ne QUADRANT POWER TILT RATIO. ~he incere cetecter monitoring _is done with a full incere flux man y two sets of four symmetric thimolesd The two sets of four_symmetrip g4gnimatesisaunicuesetofeigntcetectorlocationsj L 3/4.2.5 ONB PARA,uETERS-(UNIT 1),
- The limits en the DNS-related parameters assure that each of the parameters are maintainea witnin the normal steady-state envelope of operation assumec in the transient and a
- :ident analyses. The limits are consistent with the initial
~M2 sisum :'- s 1 : ave teen analytically :emonstr:*.ec acecuate t: maintain 2 -
! CATAWBA - UNITS 1 & 2 B 3/4 2-3 # Amencment-Nor56-(Uni t -1) f - Amcacment-No. 90-(Uni t-2 }
_ :s v,,a: .:.- ..
- . . u r.
EASES 3/4.2.5 ON3 OaiaVE*ER$-(UNIT 11 (C ntinyec) design limi ;NER rougneut eacn analy:ec transient. As notec :n Figure 3.2-;, Reactor C00iant ;, stem flow rate anc THERMAL F0WER may be "tracec ef f" against one another i'.e. , a low measurec Reacter C olant System flow rate is ac:ecta:!e i f the THER.v L ::aER i s al s: lew) :: ensure ina tne calculatec CNBR ii' no ce celow t i :es ;- DN5R value. The relat:ensn1 cefineo en Figure 3.2-; rema: s valid a ' ng as tne limits :lacec en tne nuclear enthalpy rise n:t enannei factor, _ 1 i:eci#ication 3.2.3 are maintainec. The indicated 7 value anc tne inc::stec :ressurl:er f ressure value c:rresconc to analytical li=tts of 594.8'F and 2 05.2 asig respectively, witn allowance for measurement '.ccer-tainty. When React:r C0olant System flow rate is measured, no additional allowances are necessary prior to ccmcarisen ith tne limits of Figure 3.2-1 since a measurement err:r of 2.1% for Reacter C:olant System total flow rate nas ceen all wec ':r in cetermination of tne cesign DNBR value. l The measurement errer for React:r Ccclant System total flow ratG is casec upon performing a recision neat balance anc using the result to calibrate tne - Reactor Cooiant System flow rate indicators. Potential fouling of tne feecwater - venturi whien mign not be detected could bias the result frem tne precision i heat balance in a noncenservative manner. Therefore, a penalty of 0.1% fer uncetected f:uling of the feecwater venturi is included in Figure 3.2-1. Any fouling whien mignt bias the Reactor Coolant System flow rate measurement greater l l than 0.1% can te cetected by monitoring and trencing various plant performance l ! parameters. If cetected, action snall be taken before performing subsecuent ! precision hea aiance measurements, i.e., either tne effect of tne fouling i snall be cuantifiec and ccmcensatec for in tne Reacter Coolant System fi
- rate ,
measurement :r ne venturi shall be cleanec to eliminate tne fouling. ! The 12-nour :eriodic surveillance of these carameters througn instrument reacout is sufficient to ensure inat the parameters are restorec within neir limits folle.ing ::ac cnanges and Otner expected transient oceration. incica-tien instrumentat :n measurement uncertainties are accountea for in the limits proviced in Tacle 3.2-1. l I CATAWBA - UH!T5 1 L 2 3 3/4 2-4 _ Ameriement No. 86 (Unit 1)
*HM Amencment No. 80 (Unit 2) 1 i
. - - - . _. ~ _ . . . - _ _ . . .- . . -- - - _ _ . . _ _
3/4.3 !NS :UMEN'ATION SASES 3/4.3.1 ano 3/4.3.2 REACTOR TRIP $YSTEM and ENGINEERED SAFETY FEATURES ACTUATICN 5Y$IEM !NST;bMENTAIIGN The OPERABILITY of the Reactor Trio System and the Engineerec Safety Features Actuation System instrumentation anc interlocks ensures tnat: (1) ne associatea ACTION ana/or Reactor trip will be initiated when the parameter monitorea Oy eacn enannel or comoination thereof reacnes its Setooint, (2) the soecified c:1ncicence logic and sufficient recuncancy is maintainea to permit a cnannel to te out-of-service for testing or maintenance, consistent witn main-taining an accrocriate level of reliacility of the Reactor Protection and Engi-neerea Safety Features instrumentation anc (3) suf ficient system functional capacility is available from diverse parameters. The OPERASIL*TY of these systems is recuirea to provice the overall reliacility, recunaancy, anc civersity assumed available in tne facility cesign for tne cratection anc mitigation of accicent anc transient concitions. The integratea coerstion of eacn of these systems is consistent witn tne assumotions usea in the safety analyses. The Surveillance Requirements soeci-fiea for taese systems ensure that the overall system functional capanility is maintainec ::moaracle to the original design stancaras. The cericcic surveil-lance tests :erformeo at tne minimum frequencies are sufficient to camonstrate this capacility. Specifiea surveillance intervals and surveillance and maintenance outage times nave :een cetermined in accorcar.ce with WCAP-10271, " Evaluation of Sur-veillance F eauencies and Out of Service Times for the Reactor Protection In-strumentati:n System", and sucolements to that recort. Surveillance intervals anc out of service times were aeterminea cased on maintaining an acercoriate level of reiiacility of tne Reactor Protection System and Engineerec Safety Featurts instrumentation. (Implementation of auarterly testing of RTS is ceing postooned until after acoroval of a similar testing interval for ESFAS). The NRC Safety Evaluation Report for WCAP-10271 was proviced in a letter cated Feoruary 21, 1985 from C. O. Thomas (NRC) to J. J. Shepparc (WOG-CP&L). The Engineerec Safety Features Actuation System Instrumentation Trio i 5etooints 5:ecifiea in Table 3.3-4 are tne nominal values at wnicn :ne oistacles l are set for eacn functional unit. A Setooint is considered to ce acjustec l consisten: J tn :ne nominal value onen the "3s measured" Setcoint : witnin I tne cana a .cwea f:r calioration accuracy. To acc mmocate the instrument drift assumed to occur' cetween coeration'al tests anc :ne accuracy to wnicn Setooints can ce measured anc calibrated, . Allowaole Values for tne Setooints nave oeen specifiec in Table 3.3-4 Ocera-tion with Setooints less conservative than the Trio setooint cut witnin tne Allowaole Value is acceptable since an allowance nas been maae in ne safety acccmmocate tnis error f An optional provision nas oeen inciucea ' 3De\[Oanalysist: for cetermin:ng tne 0PERABILITY of a cnannel wnen its Trip Setcoint is f auna to exceea tne Allowaole Value. The metnocology of this cotion utilizes tne
'as ressure:" reviation from tne s:eci#f ec calibration scint f:r 2:n anc CATAWEA - UN!TS 1 L 2 9 3/4 3-1 8-105
- NSTRUMENTATICN l
BASE 5 l REACT 00 TRIP SYSTEM anc ENGINEERED SAFETY FEATURES ACTUATION SYSTEM IN5TRUMENTATION (Cont 1nuec) ' D \Wsensor components in conjunction witn a statistical comoination of the otner uncertainties of tne instrumentation to measure the process variaole anc the uncertainties in calibrating tne instrument: tion. !n Ecuation 3.3-1. ! I + R + 5 < TA, tne interactive effecte of the errors in the rack and tne sensor, anc the "as measurec" values of the errors are ccnsiderec. I, as soecifiec in Taoie 3.3-4, in percent span, is the statistical summation of errors assumeo in the analysis excluding those associated with the sensor and rack drift and the accuracy of their neasurement. TA or Total Allowance is the cifference, in percent span, R or Rack Error is the "as measurec"
, deviation, in the cercent span, for the affected channel from the specifiec Trip Setooint. $ or Sensor Error is either the "as measurec" deviation of i
the sensor from its calibration point or the value soecif No in Taole 3.3-4, l in cercent scan. f*cm the analysis assu. motions. Use of Ewation 3.3-1 allows f or a sensor cr1f t f actor, an increasec rack crif t f actor, and provices a
~j nresnold value for REPORTABLE EV The metnceology to cerive the Trip Setooints is based upon comoining all of the uncertainties in the enannels. Innerent to the cetermination of tne Trip Setpoints are tne magnituces of these channel uncertainties. Sensor anc rack instrumentation utilitec in these Channels are exoectea to be C3paole of cperating within :ne allowances of these uncertainty magnituces. Rack crift in excess of the Allowaole Value exhibits the benavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessiva drift is expected. Rack or sensor crif t, in excess of tne allowance-that is more than 6ccasional, may be indicative of more sericus proclems and should warrant furtner investigation.
l-The measurement of response time at the soecified frecuencies orovides ! ass'urance tnat tne Reactor trip and the Engineered Safety Features actuation I associated with eacn channel is comoletec within the time limit assumec in the safety analyses. No credit was taken in the analyses for those cnannels witn rescanse times indicated as not applicable. Resconse time may be cemonstratec cy any series of secuential, overlaoping or total channel test measurements j orovided that sucn tests demonstrate the total channel response time as cefinea. I 5ensor rescorse time verification may be demonstratea by either: (1) in place, onsite, or offsite test measurements, or (2) utiliting reolacement
. : . a ., . 2 . ;.. .=Pti"# dc resoonse i.ime, i
The Engineerec Safety Features Actuation System senses selected plant carameters anc cetermines wnetner or not credeterminea limits are ceing exceecee. ! If they are, ne signals are comoined into logic matrices sensitive to comoina-tions incicative of various accidents, events, anc transients. Once the l I recuired logic como1 nation is comoleted, the system sencs actuation signals to nose Engineerec Safety Features components whose aggregate function cest serves the recuirements of the concition. As an examole, tne following actions may be initiatec cy tne Engineered Safety Features Actuation System to mitigate ! - e corsecuences of a steam 1 9e creax or icss-c d
-coolant accice r T Sat ection cumos star *. anc automatic valves cosi'. ion, (O Reac*.or t-'o , ( 3 ) 'eec-
-ater isolaticn. (a) s*.artuo of tne emergency diesel generators, (5) containment
'_ATAWBA - UNIT 5 ; !. 2 5 3/4 3-2 g, g g
REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 GPERATIONAL LEAVAGE i l PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an 1mpenaing gross f ailure of tne pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTCOWN. Industry experience has snown that while a limitea amount of leakage is expected from the Peactor Coolant System, the unidentified portion of this leakage can be recuced to a tnreshold value of less than 1 gpm. This thres-fep\ac t hold value is sufficiently lcw to ensure early detection of additiCnal leakage, W W The total steam generator tube leakage limit of 1 gpm for all steam
;AP^5 generators not isolated from tne Peactor Coolant System ensures that the dosage
\ centribution from the tube leakage will be limitec to a small fraction of 10 CFR
) Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. The 1 gpm limit is consistent with the assumptions used in the analysis of these accicents. Tile 500 gpd leakage limit per steam generator ensures that ste.m generator tube integrity is maintained in the event of a main steam line rupture or under LOCA ccnditions.
The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage frcm known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Cetection Systems. The CONTROLLED LEAKAGE limitation restr1 cts operation when the total ficw suppliec to the reactor ccoiant pump seals exceeds 40 gpm with the moculating valve in the supply line fully open at a ncminal Reactor Ccolant System pres-sure of 2235 psig. This limitation ensurr s that in tne event of a LOCA, the safety injection flow will r.ct be less than assumea in the safety analyses. The 1 gpm leakage from any Reactor Coolant System pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure. It is apparent that wnen pressure isolaticn is provided by two in-series check valves and when failure of one valve in the pair can go undetecteo for a substantial length of t1me, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECC5 low pressure piping which could result in a LOCA that bypasses containment, these valves sr.cuid be tested periccically to ensure low orocability of gross failure. The Surveillance Recuirements for Reactor Coolant System pressure isolation valves provide added assurance of valve integrity thereby reducing the proo-ability of gross valve failure acc consecuent intersystem LOCA. Leakage from i the pressure isolv'.on valve is IDENTIFIED LEAKAGE anc will be considered as a portion of the allowed limit. l l l , CATAWBA - UNITS 1 & 2 5 3/4 4-4 amenoment No. 84 -(Uni t 1) g 9.enoment40r7&dUnit Ei l
.. .~ -~ . -. . . . . - . . - . ._ - . - .. - - . . .. . . . ...- . - .. ~.
I 1
^ttachment 2: i t
The total steam generator tube leakage limits of 0.5 gpm for all steam generators and.200 gpd maximum leakage into cne steam generator ensures that the dosage contributien from the tube leakage will be limited to the applicable fracticn of 10 CFR Part 100 dose guideline values in the event of any FSAR Chapter 15 transient or accident. The 0.5 gpm and 200 gpd limits bound the assumptiens used in the analysis of these accidents. The 200 gpd Icakage limit per steam generator also ensures that steam generator tube integrity is maintained in the event of a main steam 2ine rupture or under LOCA conditions. f s' I' i s 1 i i i d x-108 7 - v - , -e.-... , . , - - w .w.e,w...r-.,.. - - . , - ,. -,,w-.-, . - - . , - - . - - -
.-r - ,, - ,r
ADMINISTRATIVE C:N CLS SEMI ANNUAL RADIC A."IVE EF LUINT RELEASE REPORT (Continuec) The Racicac- .e Ef fluent Release Recorts snall inciuce a list anc descri:- , tion of unclannec releases from ne site to UNRESTRICTED AREAS of racicactive materials in gase:us anc licuic ef fluents mace curing tne reporting perica. The Racicact ze Ef fluent Release Reports snall i cluce any enanges mace during tne re ert- ; :erice :: tne PROCESS CONTROL PROGP.AM (PCP) anc :: tne OFF5ITE 00SE CAL ^.'.ATICN MANUAL (ODCM), as well as a listing of new locations for dose calculat'.:ns anc/cr environmental monitoring icentified oy tne lano use census pursuan to Sceci fication 3.12.2. .
.%NTHLY OPERATING REPORTS 6.9.1.3 Routine ecor.s of c:erating statistics anc snutcewn excerience in-clucing cocumentati:n of all cnallenges to the PORVs :r safety valves, snal; be suomittec on a mentnly casis to tne NRC in acccreance with 10 CFR 50.4, r.o later tnan tne 15th of eacn mentn following the calencar montn coverec_by the report.
CORE OPERATING t:u!TS REPORT
- 6. 9.1. 9 Core c era .ing limits snall be establisnec anc documentec in the CCRE OPERATING LIMIT 5 REPORT before eacn reloac cycle or any remaining part of a reloac cycle for tne folicwing:
- 1. Moderater 'emeerature Coefficient SQL and EOL limits and 300 p;m
- surveillance li
- it f:r Specification 3/4.1.1.3,
- 2. Shutccwn San ( Insertien Limit for Specification 3/4.1.3.5, l 3. Control Sanx. Insertion Limits for Sperification 3/4.1.3.5, 1 4 Axial Flux Ci#ference Limits, target Danc", and APL'ND* for [
l Specificati:n 3/4.2.1, t
- 5. hea Q Het Channel Fact:r, F E
, <(Z), W(Z)"", APL ND" pq Qyy a3. {
W(Z W ;r 5:ecification 3/4.2.g., anc .... 1 l
, *ns TP I
- 5. Nuclear Ent.uicy Rise Hot Channel mmx, Facter, Fiq(y,y r, F2Ji ,
I and Pcwer Fact:r Multiplier, MFiH , limits for 5:ecifica-l tion 3/4. 2. 3. The analytical metaccs used t: cetermine tne core coerating limits snall be those previously reviewec anc a:crevec by NRC in:
- 1. WCAP-9272-P-a , '"4ESTINGHOUSE RELOAD S AFETY EVALUATION METH000 LOGY,"
July 1995 ] corietary). (Metncccicgy for Specifications 3.1.1.3 - Moderator Temperature l Coef ficient, 3.1.3.5 - Shutdown Bank Inserti:n Limit, 3.1.3.5 - Control Bank Insec- lcn Limits, 3.2.1 - Axial Flux NO l " Reference 1
" Reference y ,9e[rNapplicaole to totarget a 'e not applicaole W(2), and band APLand APL. NO,andWCQ.
*** Reference :s n - a:plicaule to F^ 4
***=qeference 5 : :t acclica:ie t:
- I. anc . F" _m..
i I CATA'dBA - UNITS 1 12 6-19 8-109 Asensentj e. y fU % f
~c. .w l e ment mio
ACHINISTRATIVE C:N~ROLS CCRE OPERATING '.:u:T3 REPOEI (C:ntinued) Differe .ce, 3.2.2 - Heat Flux Hot Channel Fac:ce, anc 3.2.3 - Nuclear intnaloy Rise not Channel ~acter.)
- 2. WCAP-10215-:-A, "RELAXATICN OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILL;NCE ECHNICAL SPECIFICATICN," June 1983 (W Proprietary). ,
(Mein:cciogy for 5:ecifications 3.2.1 - Axial Flux Oifference (Relaxec Axial Offset Control) anc 3.2.2 - Heat Flux Mot Channel Fac :r , ,(Z) surveillance equirements far F Metnocology.) q
- 3. WCAP-10265-P- A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATICH MuuEL USING BASH C:CE," Maren 1937, (W Proorletary).
(Metn:cciegy for Specification 3.2.2 - Heat Flux Hot Channel Facter.) 4 BAW-10152-A. 'N000LE - A Multi-Dimensicnal Two-Group React:r \ Simulator,"\ l June 1985. ; kh (Metn:cciogy for Specification 3.1.1.3 Coeffi:ient.) Moderator Temperature
, g,d # 1
! ; B9'. AW-10163P-A. " Core Operating Limit Methodology for Westingnouse-l Designee PwR's," June 1989. I d (Metncc:!:gy for Scecifications 3.1.3.5 - Shutcown Rod Insertion , Limits. 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial l ,- Flux Di'f erence, 3.2.2 - Heat Flux riot Channel Factor, anc 3.2.3 - . l Nuclear Enthalpy Rise Hot Channel Factor.) !
- 6. BAW-1016SP, :ev.1, "S&W L:ss of-Coolant Accicent Evaluation Macel for f Recirculating Steam Generator Plants." Seotemoer,1989. ;
(Metn:ccicgy for Specification 3.2.2 - Heat Flux Hot CM nnei Fact:r.; i t _ The core operating limits shall te determined so that all applicable limits
.(e.g., fuel ther-al-mecnanical limits, core thermal-hydraulic limits, ECC5
-its, auclear mits sucn a3 3.utc: n targin, anc tr nsient anc ac:ident analysis limits) .f the safety analysis are met.
The CORE OFERATING LIMITS REPORT, including any mid-cycle revisions or ! . supplements thereto, snall be provided upon issuance, for each reload cycle, to the NRC in accercance with 10 CFR 50.4. H-110 . CATAWBA - UNIT 3 ' ?. 2
- 6-19a ~ Amendment No. :6 (Unit 1) a . ., . e un n nIn6 71 ]
. ,.J
.. -- - - . - . . . ,- ., .. =.-~. . .- . ~ .
=_
for Specification 6.9.1.9 Attachment 1:
- 4. BAW-lC16BPA, Rev. 1, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," January 1991
-(B&W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factcr4 ) 5~. DPC-NE-20llP-A, " Duke Power Company Nuclear Design Methodology for Core Operating. Limits of West'inghouse Peactors," March, 1990 (DPC Proprietary). 1 (Methodology for Specification 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
- Muclear Enthalpy Rise Hot Channel Factor.)
- 6. DPC-NE-300lP-A, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," November 1991 (DPC Preprietanr).
(Methodology.for Specification 3.1.1.3 - Moderator Temperature , Coefficient , 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - I Control Bank Insertion Limits, 3.2.1 - Axial Flux Di f f erence, 3.2.2 Heat' Flux Hot Channel Facter, and 3.2.3 - Nuclear Enthal'yp Rise Hot Channel Factor.) 7 DPC-NF-2010P-A, " Duke Power.Cenpany McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design,* June 1985'(DPC Proprietary). (Methodology for Specificatien 3.1.1.3 - Moderator Temperature
' Coefficient.)
- 8. DPC-NE-3002A,*FSAR Chapter 15 System Transient Analysis Methodology," November 1991.
(Methodology used in the system thermal-hydraulic analyses which
. determine the core operating limits)
- 9. DPC-NE-3000P-A, Rev. 1. " Thermal-Hydraulic Transient Analysis g . Methodology," November 1991.
I '. . . ti (Modeling used in the system thermal-hydraulic analyses) F I i.
~
8 111 I \
m - 3.2 Changes.co Core Operating Limits Report a I.- a Y d I A l-4: I. 4 f' f 4 T N E [.- 1 :- i a i l f l- 8-112 f ~.
. . . . _ . . . . - . . . . - - . . ~ . . - - - . . _ _ .- - -
4 4 Carawba 1 Cycle 7 Ccre Operating Livits Report 1.0 ccre crernin't L imi.t " Fenert l This C0re Operating Limits Eeport (COLR) for Catawba Unit 1 Cycle i 7 has teen prepared in acccrdance with the revirements of Technical Specificaticn 6.9.1.9, i The technical specifications affected b/ this report are listed I belcw: 4 3/4.1.1.3 floderatcr Temperature Cceificient l 3/4.1.3.5 Shutdown Eod Insertio.. Limit I
.1/4.1.3.6 Con-14 -1 Rod Insertion Limit
! 3/442.1 Axial Flux Difference . 1 3/4.2.2 Heat Flux Hot Channel Factor i 3./ 4. 2. 3 Muci : Enthalpy Rise Hot _ Channel Facter ! I ( f i
- i. h I b I p I
1 I l- h t T
, 8-113 ..- - .- - - - ~ . _ . - . _ - . - _ _ . . _ . . _ . . . , . _ . . . _ _ . . . _ , . . _ . . . . - . - - . - _ - , , _ _ , . - - _ . . , ,
Cat mcr.a 1 cy c:l e ~ c at c pe: M in ; L :'.i t s F.epc r t 2.2 T h > : t u r..r :d In r ett: '
'_ : r i t .;pe:1ficaticn 2 4.1.3.5)
- 2. 2.1 This chutd:wn .O. nhh11 be withir.un to at least 222 :ters
.' . 3 c:ntr 1 P'd i n c e r t i '_p i_. ! m i t : (Specification 1/4.1.3.E.)
6 2.3.1 "he 'cntrol rod tanks chall te limited in phyr:ical insertion as sh:wn in Figure 2.4 Af121.11ux Difference r pecificati:n 3/4,2.1)
- 2. 4. ' The AXI AL FLUX DIFFEPl2JCE ( AF D) Limits are provided in Figure
( AFD Limit).,jk,.. is t he negative AFD limit from Figure 3. (AFD Limit),:;hn, is the positive AFD limit from Figure 3. O X.114
Catawba Control Roa insertion L!mits for 4 t con Oceration Fully Withdrawn 1 (Meximum-230) g 230 w........ /. (.s.% 2..). . . . .. ..... (
. 80.% 230.) . .
Fully Withdrawn (Minimum-222) 200 BANKB
~
130 _(0%,164) (100%,161)
? ~
it m 140 BANK C E
~
j 120 E # o ~
@ 100 c
~
80 2 BANXD
~
60 . 3
* ~
(0%,48) 40
~
20 (30%,0) Fully inserted I I I I I\ l I l _ _ . 0 40 100 0 20 60 80 Relative Power (Percent), MT Figure 2 Control Rod Bank Insertion Limits Versus Percent Rated Thermal Power Bank insertion limits are to be set at; a function of reactor power, as measured by delta T.
*100%AT - Indicated AT Normal Full Power Concitions 8 115
110 - - ( 20.100il l (+ 10,1001l tnac m .n o m mon ,. . , 8 UnacceptaNe Operman j j :D - -
~4 70 .
b 2 rum .onuwon 60 - - 3 l j. w,50,l (+21.501l 4 $0 - =
~l
=- -
30 - - 20 * - 10 - -
, , . , e e t
, , , . , , i. . .
so ao 30 20 to . o 10 :o 30 40 50 Akhi Flux Dinersace % Detta 16 , Figure 3 Percent of Rated nermal T wer Versus Axial Flux Difference Limits
Catawba 1 Cycle 7 Core CT.erating Limits Report
- 4. 5 He r F1ux Het Channal Fsitrr - F; ( X , Y , Z ) (Specification 3/4.2.2)
- 2. 5.1 F ' = 2.32 2.5.2 F.(Z) is provided in Figure 4 for Mark-EW fuel.
2.5.3 F.(Z) is provided in Figure 5 for OFA fuel. The following parameters are Iequired for core monitoring per the Surveillance Requirements of .:pecification 3/4.2.2: 2.5.4 ( F- ( X , Y, Z ) ) " = Fi ( X , Y., Z )
- M, ( X , Y , Z ) / ( UMT
- MT
- TI LT )
> where ' (Fj (X,Y. Z ) ) # = cycle dependent maximum allowable ~ design peaking factor wnich ensures that the ( F$ ( X , Y, Z ) ) limit will be preserved for operatir.1 within the LCO limits. ( Fs ( X , Y, ))" includes allowances for calculational and measurement uncertainties, F;(X,Y,Z) = the .iesign power distribution for Fg. Fj (X , Y, Z) is provided in Table 1.
!((X,Y,Z) = the margin remaining in core location X,Y,Z to t'.:e LOCA limit in the transient power distribution. !%(X,Y,Z) is provided in Table 2.
!Jote: (FI(X,Y, Z) )" is the parameter identified asF7"(X,Y, Z) in DPC-!JE-10llFA.
2.5. 5 (F;(X, Y, Z ) )"# = F[(X, Y, Z) * (H:(X, Y , Z ) / (UMT*!ff *TI LT) ) where (FI(X, Y, Z ) ) *" = t'/cle dependent maximum allowable design peaking factor which ensures " that the centerline fuel melt limit will be preserved for operation within the LCO limits, (Fj (X, Y, Z) )"' includes allowances f or calculational and meesurement uncertainties. F$ (X, Y ,2) = the design power distribution for F g. F$ ( X , Y , Z) is provided in Table 1. M:(X,Y,2) = the margin remaining in core location X,Y,Z to the CFM limit in the transient power distribution. M:(X,Y,Z) calculations parallel the
!((X,Y,Z) calculation described in DPC-!JE-20llPA, except that the LOCA limit is replaced with the CFM limit. ft. (X , Y, Z ) is provided in Table 3.
8 117
l l 1 Cat a ):,a 1 Cycle 7 Core Cperating Limits Report Um
- Measurement Uncertainty (UMT = 1.05).
MT = Engineering hot channel factor (MT = 1.03). . TILT = Peaking penalty that accounts for allcwable quadrant power tilt ratio of 1.02, 2.5.6 f. SLOPE = 0.07B* where KSLOPE = Adjustr*~
- t o the Ki value from OT.iT '
requir A to '^= e,cate f or each l't that F; ( X , Y, ) ;. : .ts init. , typical value; actual values will k c..fu '.ad when tionitoring inputs are computed. ; l. l l l 8-118 i
I l Catawba 1 Cycle 7 Core Operating Limits Report , 1.2 (0,0. l .0m} (8.0. I .00) g (l0.8.0.94) 0.8 - h0.6 - (l2.0.0.647) 0,4 - 0.2 - 0 0 1 2 3 4 5 6 7 8 9 to 11 12 Core lleight (ft) Figure 4 K(Z), Normalized Fn(X,Y,Z)' as a Function of Core Height for Mk!3W Fuel l 6 8 119
.. .--..._. -....-_ - . . . - . . - - . - _ . . - . . . - - - . - . - . . - - - . = . . - . - . - . . . - . - . . - . -
O Catawba 1 Cycle 7 Core Operating Limits Report 1.2 -
- (0.0.1 fxnl (6.0.1 AX4! '
( l0.H.O.94 ) 1 0.8 - W - tir 0.6 l (12.0.0.647) 0,4 -
~
0.2 9 l 0 l 1 0 1 2 3 4 5 6 7 8 9 10 Ii 12 Core lleight titi l l l Figure 5 l
- l. K(Z), Normalized FniX,Y,Z) as a Function of Core Height for GFA Fuel 9
4. 8 120 1 n.._ . _ . . _ .....__. __ ._ _ . _ , _ _ . . _ , _ , _ _ _ ,_ _ _ _ _ . _ , _ ,
1 Catawba 1 Cycle ~ Core Crerating Limits Report 2.6 INelear Enthairy Rica H:t channel Farter - Fy (X,Y) ( Fia ( X , Y ) ) ~ = MAF#fX,Y)*(1.0 + ( 1/ RRii) * ( 1. 0 - F ) ] 2.6.1 Catawba 1 Cycle 7 Cperating Limit Maximum Allowable Radial Feaks (MARPfX,Y)) are provided in Table 4. The f311cwing parameters are reTaired for core monitoring per the surveillance requirements er 3/4.2.3: (FL (X, Y ) ) *'" = FL (X, Y)
- My (X, Y) / t UMR* TILT)
Note: (FL (X,Y) )""is the parameter identified as FL'(X,Y) in DPC-ME-20llPA. where UMR = Uncertainty value for measured radial peaks, (UMRrl.04). TILT = Peaking penalty that accounts for allowable quadrant power tilt ratio of 1.02.
- 2. 6.2 FLIX, Y) is provided in Table 5.
- 2. 6.3 Mw (X, Y) is provided in Table 6.
2 . 6. 4 EPR = 3 . 3 4 where FPS = Tnermal Pcwer reduction required to compensate for each 1% that Fy ( X , Y) exceeds its limit. Thermal Pm or P " Rated Thertal .cwer -
- 2. 6. 5 TPR u 0.04 where TPS = Reduction in CTAT K: setpoint required to compensate for each 1% that Fy(X.Y) exceecs its limit.
NOTE: Tables 5 and 6 will be supplied when monitoring inr,ats are computed. i 8-121
Table 4. Maximum A11ntable Radial Feak (MARP) Values
, 1 tv4,' '
, s v 4 .31 2 3 Axi31 1,4 Avi 31 Elr/at1:n (ft) Peak Peak Peak Pe a e.
MAP.P FiM.E MAF.P M1J P
.12 1.5809 1.6266 1.6722 1.7113 1.2 1.5806 1.6259 1.6677 1.7025 2.4 1.5836 1.6265 1.6663 1.7025 3.6 1.5859 1.6263 1.6635 1.6960 4.8 1.5371 1.6240 1.6571 1.6751 6.0 1.5878 1.6196 1.6470 1.6303 7.2 1.5864 1.6130 1.6265 1.5B48 9.4 1.5731 1.5956 1.5773 1.5327 9.6 1.5655 1.5612 1.5208 1.4815 10.8 1.5459 1.5152 1.4717 1.4292 12.0 1.5133 1.4693 1.4274 1.3878 1.5 Ax131 1.6 Axini 1. ~7 Axial 1.9 Axial Eletntien (ft! Peak M Peak E.ast MAP.P = 122E 12.F.E
.12 1.7477 1.7331 1.7054 1,6438 1.2 1.7433 1.7029 1.6789 1.6193 2.4 1.7126 1.6616 1.6433 1.5869 3.6 1.6735 1.6211 1.6011 1.5504 4.8 1.6313 1.5811 1.5622 1.5121 6.0 1.5268 1.5415 1.5238 3.4763 7.2 1.537P 1.4913 1.4766 1.4344 8.4 1.4886 1.4450 1.4296 1.3880 9.6 1.4399 1.'013 1.3882 1.3490 10.8 1.3883 1.3526 1.3433 1.3081 12.0 1.3500 1.3140 1.3078 1.2749 1.9 Axia], 2.1 Axial Ele *>atirn ('Q M M MIJ.P MAP.P
.12 1.5839 1.5401 1.2 1.5624 1.5154 2.4 1.5328 1.4801 3.6 1.5013 1.4395 4.8 1.4626 1.4030 6.0 1,4291 1.3619 7.2 1.3920 1.3271 8.4 1.3485 1.2824 3.6 1.3126 1.2501 1D.8 1.2726 1.2091 12.0 1.2443 1.1890
.M Tpl alves, +. be e 9 elecl t Aen c t c.p 7 un eavec tg oulysc,s i< complede.
l 8-122
x,3 Changes : the Final Safety Analysis Report k 123
-. . - - -. . __ . .- - _ = . . -.
Catawba Nucicar Statkm 15.3 Decrease in Reactor Coolant Sptem flow Rate 0 Table 15-22 (Page I of 2). Pars. meters for Postulated locked Rotor OfNte Dose Analpis Consen atise
- l. Data and and assumptions used to estunate radioactive source from postulated accident
- a. Power J evel (MWt) 3565
- b. Percent of fuel defected P'
- c. To'.al steam generator tube leak rate during oA
/'gpm accident and initial 8 hours O d. Actisity released to reactor coolant from failed fuel 10% of gap inventory
- e. Offsite Power Not available
- f. Reactor coolant activity prior to accident Primary and Secondary Activity During Normal Operations (Table Table 114 on page 111i)
- 2. Data and assumptions used to estimate activity released 0 a. Iodine partition factor 0.01 b.
-MWls 6 team release & cA-W;chiwr GM;24Hirto+hry- 1Dg,y[
,ht4&,91ftfrrtt+he,,gco em egrw c
. W 87dy. PAW.mu
--Ntaten-cu D4' 5 vuolduw u v, acco r~,ary- sys.mo (o -h.cg, M.O en-awidentr(hr5)~
- 3. Dispersion data
- a. Distance to exclusion area boundary (m) 762
- b. Distance to low population zone (m) 6096 r
- c. x/Q at exclusion area boundary (sec/m3 ) 4.5504,.74 64 g
- d. y/Q at low population zone (secim') --EEE fth
- 6CE G
. se data S ,q g 3,,g 3,4 gyg
- a. Method of dose calculations Regulatory Guide 1.4
- b. Dose convenion assumptions th W-30 AWO Regulatory GuidenMr+
1.109 Case 1 (No iodine spike) l Exclusion area boundaru ' l 0 Whole body 44r4F E.53E-l > 0 Thyroid -M- tG,6 low population zone 0 Whole body 12fbO2- GG'E O D yroid W b.~1h I i l REV: (01 OCT 1990) 15 159
15.3 Decrease in Reactor Coolant System flow Rate Catanba Nuclear Station 0 Table 1522 (Page 2 of 2). Parameters for Postulated locked Rotor OITsite Dose Analysis l Conscn atis e 1 Case 2 (With preexistir.g iodine spike) Exclusion area boundary 0 Whole body 4-4E-Gi- 7., .G 66- I O Thyroid & z 5. '7 i Low population zone 0 Whole body J-mmh 3- 6.I6-? O Thyroid & b,74 1 0 l 0 1 1 8 15-160 REY: (0) OCT 1990)
. - . . - _ . . . _ = . - . . - . . - . - _ _ _ _ ~ . . _ _ . - . . - . -
Cata=ba Nuclear Station 15.4 Reactivity and Power Distribution Anomalier,
~
0 TaWe 15-26 (Page i of 2). Parameters l' roPostulated Rod Election Offsite Dose Analysis Conservatise Realistic Data and assumptions used to estimate 1. radioactive source from postuhted accidents
- a. Powr level (MWt) J565. 3565.
A o.e--g er r. i uret_.r1 L g.n
- c. Steam generator tube lest rate prior to and ,Y O' 4+Wm a.a ..
a . .. _.- _ , _.a - pea,W
- d. Failed fuel
)m 6 p:< cent of fuel same rods in core
- e. Activity released to reactor coolant from failed fuel and available for release Noble gases 10 percent of core same gap m' ventory lodines i0 percent of core same gap m' ventory f "e:dL' _ aS _ '
n . e c_ Act_ivity relas*=d te N2Nor enn1mnt from enahe d - fual -hikbk for release 4o-tenummenr-PW ;;.w ^467c; bf 'l 2 4A 44 as a ;
!eh--
0 '25 7;..:- ef - -- 0-- ms ~ ,.m -
- h. lodine Fractions (organic, elemental, and ReFulatory Guide same particulate) 1.4
- 2. Data and assumptions used to estimate t:tivity released a.
Contamment Free volume (ft)) 1.015E + 06 same
- b. Contamment leak rate 0.3 percent of
' 0.05 percent of containment containment volume per day, volume per day, Osts24 hr Osts24 hr 0.15 percent of 0.025 percent af contamment contamment i ' volume per day, volume per day, t > 24 hr t > 24 hr "
- c. Bypass leakage fractio'n 0.07 0.07 A
.O 1"~ ; =h : != : '- "- _ _ ' = - ,! _-
[ l
- e. Offsite power Lost .
{ I i l i RET:(el OCT 1990) 15 211
. % L-. . - . hv~,e.~..~...~.._. . - - ,
ISA Reactiiity and Power Distribution Anomalies Catawba Nudcar Station O Table 15-26 (Page 2 of 2). Parameters for Postulated Rod Ejection Offsite Dose Analysis Consen atis e Realistic
- f. h duu,y im. .n'.:' d: - (!b; -
- +ovu. -
I - E .a,e m uf d uu n,f s , K R f d e,(: r ) i U. [ pf3. p Disper-don data
- a. Distance to exclusion area boundary (m) 762. 762.
- b. Dis mce to low population zone (tu) 6096. 6096.
- c. z/Q at exclusiot area boundary (sec/m3) j 0 2 hrs -5.5E=04 M
'l d. r/Q at low population zone (sec/m3) g h 0 8 hrs ME 03 t^" ; - B&hn :.2E-Of - - - - 5 4E-00 m I 4-.(-day,-- a 1E-% 2.!E_00
.44 days -- L2r 06 o E 0-
- 4. Dose data
- a. Method of dose calculation Regulatory Guide same ,
1.77
- b. Dose conversion assumptions Rg.r,cgfr-30
- a. r: Gu:b same 4-4-and 1.209
- c. Doses (Rem)
Prunary side Exclusive area boundary 0 Whcie body 7.4Gi
.-6.9E-2 0 Thyroid W ,6 3 Loci population zone 0 Whole body TTE p 1. 06M 0 Thyroid M i9b Secondary side Exclusion area boundary 0 Whole body .;.5E-g 3N 0 Thyroid -5.3E-e i46
. Low population zone -
Whole body N-O f t.0EP-0 Thyroid 53E 7. I7* t%^Z PW 15-242 REY: (01 OCT 1900)
, , -! l- l;;li;! !t - 'j4 4 )!! I, ,i ' l,,;
S <m 2 E e Oo:- [ C 1 ',l.! g; -
!q
!i
' oU:li :l
[ 'i:!, 3y} g p,, b= ' i ds [o=6 :J o_ t' l-Oxl 2,
.,z ocg3u > am.
\ . ]
f . s tu ; . ri
%u _E7}%, !h
. lii s
i
! N4 O1 gfp3e}4 n t
o t - 7Zo g,uw[>,- !1 _L
, I.
s,.
- ,Ii
.gl!_N,Iii Ij;,p
. ; t i1;}I, ,ti h, y,!1r,l, l! -
.I1
~
iw2,W_ L m! ' 1 , 4I; I , - i! Ii, s 4p lb,w,g t J f,.1 3 , i!4Ij
* * -:;l .
+
a !1lIi[: . T!' 3l!p 4>w
;!;li t k
. }4 i i I .
np_ 3{4 b+2g3 36.I hpa2. t - , I-
..12 d,_! g
. ,l ',
' it 3.i,
- i
-. ~
jpi;!- !j;
!7;rl! !g1ilJ ; -
. g(4yq! i -; dj2!t
.{ . , ,I1
_ 4 ,. . @ { b
,..gt@ e.!1 .iH Q,!!._
.i
. I1 y!j
.tiii i: t i i& l,c4 i,( i 1e 2_.,g pW
':j -
2 e2 M L-
~ i 1
.e M b'
- 6 Rw ,
- j!h 1 i i i ',i,! ;. ,l
.ir ti7 !,!: ; 4: i. + j !4 . ' : ,!4 . . .
4.g,p i ig! :g 7-rI ,! -!; -l + 4t. i - ,:l' 't!;:1 :1 t li l,.{:
.li:i!!i ,:
j . !, g Wn,1 q ti? -!; 't +
- 1pl;
,i! + ,: i ' ' j:. '1 .
p d j
*i ;i; 'i - !; . !! ii ' - e ;1l r: :. ,ii i f!.{ !)! - i . '} ii.- , - , ,!
Oa [1 1 {jt,
,{, +I}
[l1ii
-i
!t f;
1
- +
!3 4 ,
t (Q-gg
, i ;: -{ . 4 - ,
- . 3 ;.. ; ,' i . q w
1, jl, it! .l ' .; : 1 ,
, . i, ,i; q J ,
*i,f+!, ;l ;; ~iI ;l . ,1 ii I!,j 2, , ,}j!;It!:;I; - ,
t, , , , . ,i; ;4 .j l l, '1!;! i 1 .
', ': 4 ' ? ' ; , . . : ; ji , !,j 1 1! ;!i, , ! . i r ,! .1; 14 q
4r 1, l i.,j$t! ', -l, !i;. ii 4 l, t ,j . i i; . i i) !7 jl, t5,! < , , ; ' i1' !, 1 ', j- ;!f1.!4 i; j [,
; !{- ;Iti ,; !i; ! iil3 ' , tit ' ' 4 . 'i, .
i! ti;! i jti; 1l :
.f jr!i{
1 "; t
. 2: I4 7k1+p q
tjEi t:lti m 2!i 4i, !-
+ i bIi ,{Iii 3[ - 1 O ___.. __
q _ - 1 t!ii9_. lt . 1l 3 lm t!
~
_ - ~ i, , ], i 'I ,
- L:.
- _ . . _ -. . _ . _ _ _ . . _ . _ _ . _ _ . ~ . .
l
)
- 9. TIFERENCES
- 1. Catawba Nuclear Station, Final Safety Analysis Report. Docket Nos. 50 - 413/414.
- 2. BAW 10172P-A. Mark-EW Mechanical Design Report, Babcc:k & Wilcox, Lynchburg, Virginia, December 19, 1989.
- 3. DPC-hE-20 DIP-A, Eev. 1, ruel Mechanical Reload Analysis Nethodolcgy for Mark-EW Fuel, Duke Power Company, Oct:ber 1990.
- 4. BAW-10084A. Rev. 2, Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, Babcock & Wilccx, October 1978.
- 5. BAW-10141P-A, Rev. 1, TACO 2 - Fuel Performance Analysis, Babcock
& Wilcox, June 1983.
- 6. DPC-NF-2010A, McGuire Muclear Staticn/ Catawba Nuclear Station Nus: lear Physics Methodology for Reload Design, Duke Pcwer Company, June 1985.
7 DPC-ME-2011P-A, Nuclear Design Methodology for Core Operating Limits of Westingncuse Reactors, Duke Power Company, :' arch 1930.
- 9. DPC-ME-2004P-A, McGuire and Catawba Nuclear Stations Core Thermal- Hydraulic Methodology using VIPPI-01, Duke Power Ccmpany, Dece:rber 1991.
j 9. BAW-10159P-A, EWCMV Correlation of Critical Heat Flux in Mixing l Vane Grid Fuel Assemblies, Babcock & Wilcox, July 1990. I
- 10. BAW-10173P-A, Mark-BW Reload Safety Analysis for Catawba and McGuire, Babcock & Wilcox, Revision 2, February 20, 1991.
- 11. DPC-ME-3000P, Duke Power Company, Thermal-Hydraulic Transient Analysis Methodology, Revision 2, February 20, 1990.
l- [
- 12. DPC-NE-3001P, Duke Power Company, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology,
! Revision 2, November 1991,
- 13. BAW-10174-A., Mark-BW Reload LOCA Analysis for the Catawba and McGuire Units, Eabcock & Wilccx, Revision 1, McVember 1990.
- 14. BAW-10168-A, B&W Loss-of-Coolant Accident Evaluation Model For
- Recirculating Steam Generator Plants, Babcock & Wilcox, Lynchburgh, Virginia, January.1991.
- 15. DPC-UE-1003A, Revision 1, McGuire Nuclear Station / Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testing, December 1986.
l 91
_ . . . _ - _ _ . _ . _ . _ . . _ _ _ _ _ . _ . _ . - _ . . _ . _ . . _ _ . . _ _. - ___ _ _ .__ _ ..m. _ .. _.m._ . _ ~ . . . _ _ _ _ _ _ . . - . ~ _ 4 l i
- 16. DPC-!JE-3 002, McGuire !Juclear Staticn/ Catawba IIuclear Station FSAR i Chapter 15 Syste:n Transient Analycis Methodology, Revision 1, IJovember 1991. -
17 McGuire !!uclear Station Unit 2, Docket !Ju:rter 50-370, Cycle B Reload Submittal, Duke Power Co:tpany, Decerter 18,1391. 1 B i i 92 _ . . . _ , . . . , _ _ . . _ . . _ ~ . _ . . _ . _ _ , _ _ _ _ , . _ . _ , . _ _ _ ___ . . _ . _ _ _ . _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _____ _ _
_.- .- . . . . . - _ . ... -- -- . ..-~. . . . . . . - . _ . . . . ~ . . - . - - . . . - . . . . . - - . . . _ . _ i t Attachment 2 i i t a b 6
?
i h r i e I t I (- l 1
, ,. -. -. -:-,.. re._r, , . ~ ,,. _ , . , , . -. . .. . , ,,y,_. - _ ~ - . , , , . - , . . . . . _,_._._.,._,,,,..,,,m~._~,,,,,. ,,e...
. _ __=. _ ..._- - . ._ . . _ . . _ __... _ . - -.. . _ _ = _ _ _ _ . . . . _ _ . . . . _ . - . . _ _ _ . _ _ _ . .
Technical Justification 4 I i t E e s l I P O f l I r I
..m .m'.w.-s , . . . - ~.-. -.. . . . % w _ , , , c ..,m. , . , ,. ...-e. .w,,, ,- ,< . . . , g v .w. . .p.w....,,,w..r-*
_ . _ _ _ _ _ . _ _ _ . _ _ . . _ _ - - - . - _ - _ _ . _ . _ _ , _ . _ _ _ _ . _ . ~ . _ _ . . ~ _ . Pronared Favirirn tS Technical frecificatirn Piqure 2.1-la This change is the same as that approved fcr McGuire Units 1 and 2. This proposed Technical Speci raticn J.~ 3 ) revisicn changes Figure 2.1-la to reflect use of the ENCM. NF correlation and Duke Power Company'c Statistical Ccre Design (SCD) mathodology with a 1.55 thermal deaign DNBR limit. Tertnical Jurtificatirn Duke Power Company (DPC) has recalculated the Catawca reactor core safety limits using the BWCMV CHF correlation (Reference 1) along with its Statistical Core Design (SCD) methodology, Ref erence 5. With the implementaticn of these design methodologies, the nuclear enthalpy rise hot channel factor. FL, was decreased from 1.55 to 1.50. The proposed changes to Figure 2.1-1 reflect the use of this new design limit as well as the use of BWCMV and the Duke Power Company SCD. The reactor core safety limits provided on Figure 2.1-1 depict the i combinations of thermal power, Reactcr Coolant System pressure, and average temperature within which the calculated DNBR is no less than l the design limit DUBR value, or the average enthalpy at the vessel exit ' is less than the anthalpy of saturated liquid. The analysis which I defines these -limits is based en a full core of Mark-BW asaemblies with a thermal design flow rate that bounds the minimum measured flow at Catawba. The DNB limited portions of these curves are defined using the BWCMV CHF correlation with a design DNBR limit of 1.55. This design DNBR limit provides 10.7 percent thermal margin to the 1.40 BWCMV statistical design limit which is defined for the Catawba core using the DPC SCD methodology. The safety limits are based on a design peaking distribution with a nuclear enthalpy rise hot channel factor, FL, of 1.50 and a reference chopped cosine axial power shape with a peak of 1.55. To verify that this design peaking distribution is . conservative on a cycle-specific basis,. maximum allowable peaking (MAP) limits, which provide DNB equivalence to the design distribution at various safety limit statepoints, are defined. To verify that margin was available, these MAP limits were compared to predicted cycle-specific peaking distributiens, Reference 2. As part of the safety i limit / MAP limit analyses an evaluation was perform 9d which showed that, if power were reduced below 100 percent, peaking could be increased according to the following relationship: k = 1 + 0.3(1 - P) where k = the f actor by which the MAP limits are adjusted to define reduced power limits P = the frcction of rated power while maintaining the core within the aforenentioned thermal limits. L Comparison of the Mark-BW safety limits and the Westinghouse OFA safety j limits shows that at all points the Mark-BW limits are outside the OFA limits. Mixed core studies have shown that'the Mark-BW safety limits are applicable to the Westinghouse OFA fuel if a DNBR penalty is included for those assemblies. The DUBR penalty for OFA fuel is applied against the 10.7 % margin included in the design DNBR limit. l 1 1
.~, . - . . ..~.- -- ....-. - - _. -.-. . _ _ - - _ . - - - - . _ . . . .
l ' 1 l l Preecred Reviricn t c Terhni c al Ereci ficat ion 2. 2.1 .I This proposed Technical cret:fication (TS) change deletes ACTION 2.2.1.b.1 and equation 2.2-;. This change-is consistent with the proposed change to Technica. Specification 3.3.2, and the removal of the Total A11cwance (TA) c:lumn, the column, and the Sensor Error (S) column f rom Tables 2.2-1 and 3.3-4 Te dnical Justificaticn ACTION 2.2.1.b.1 provides tr.e cption of declaring an instrumentation channel operable by use of equation 2.2-1 when the Reactor Trip System Instrumentation or Interic:k Setpoint is less conservative than the Allowable Value. Use of this cptien necessitates the inclusion of the ! Total Allowance (TA), the value Z. and the Sensor Error (S) terms in TS = 2.2.1 and Table 2.2-1. Ecth ACTION 2.2.1.b.1 and 2.2.1.b.2 require that the Setpoint be adjusted consistent with the Setpoint value given in Table 2.2-1 however, deletion of ACTION 2.2.1.b.1 and equation 2.2- ! I makes TS 2.2.1 more restrictive, in that the channel must be declared inoperable with the Rea.ccr Trip System Instrumentation or Interlock Setpoint less ccnservative than the Allowable Value. The deletion of ACTION 2.2.1.b.1 and equati:n 2.2-1 improves the saf e operation of the plan: by reducing-the comp;exity of the Technical Specifications. This change also prcvides increased conformity with the ccmparable Technical Spccification f or McGuire Ihclear Station. Prcrcred Fevinien to Techni?31 Frecification Table 2.2-1 This proposed Technical Spe:ification (TS) revision changes the K values f or the overtemperature and overpower AT trip functions to reflect the use of the BWC!r! CHF correlation and Duke Power Company's Statistical Core Design (SCD) methodology with a 1.55 thermal design DNBR limit. In addition, an axial imbalance penalty, f;bil), is applied to the OPAT reacter trip. The power range neutron flux negative rate reactor trip s deleted from the Reactor Protection System. In addition, the table is revised to delete the Total Allowance (TA! column, the Z column, and the Sensor Error (S) column from Table 2.2-1. This change is consistent with the proposed change to Technical Specifications 2.2.1, 3.3-2.and Table 3.3-4. Technical Justification The overtemperature and overpower AT reacter trips are designed to protect the reactor from CIER and centerline fuel melt (CFM). Due to the.new-DNBR. methodology, the allowable operating region is modified. Therefore, new K values are calculated to censervatively bound this . operating region. The purpose of the OPAT trip function is to prevent center-line fuel melt (CFM) during normal peration and condition II transients. The-OPAT trip function is designed to trip the reactor when the measured AT exceeds 118% of normal full power AT. The f 2 ( AI) portion of the trip function is designed to lower the trip setpoint when axial flux differences (AFDs) exceed predetermined limits. Since the limiting margins to CFM occur as tne result of highly skewed power distributions, a fahiI) trip reset function can be developed to prevent 2
- - . - - . - - - - - ... - . . -- . - - . _ . - _ . _ . _ . - --_a
.__m____ . _ _ _ _ _ _ _ _ _ _ _ _ - _ _ . _ . _ . _ . - _ . _ _ . _ . _ . _ . _
CFM limits f rcm being exceeded, er to increase the available margin to the CFM limit. Analysis of the CIC7 core ndicates that .io f L1I) trip function 1. required to protect the ccre frem CFM. However, from an operati nal and design standpoint, it is desirable to prevent pcwer distributions correspondinc to high AFDs. For the ClC7 core, the f 3 b1I) trip reset , function was developed to produce a reactor trip cn high AFDs for credible everpower events rrctected by the OPAT trip function. The implementation of this function prevents power distributions for highly skewed conditions and afferas increased margin to the CFM limit. This margin is important in that t will reduce the probability that adjustments to the OTAT trip function will be required as the result of CFM surveillance requirements. A reactor trip on negative flux rate is not assumed in any of the-licensing basis accident analyses. The analysis of the dropped rod
. accident (Reference 4), for which the negative flux rate trip was designed to provide protection, assumes that reactor trip occurs on low pressurizer pressure, if at all. For cases in which no trip occurs, i.e. for low drcpped rod worths, this analysis shows that none is needed. The removal of the negative flux rate trip frcm the Reacter Prctection System will eliminate both unnecessary reactor trips resulting from such Icw worth rod drop events and spv- ous trips from maintenance / surveillance activities.
In addition to the technical justification provided for the preposed revi ion to Technical Specification 2.2-1, the removal of the TA, Z, and S columns from Table 2.2-1 improves the safe operation of the plant by improving the readability of the table. Furthermore, retention of these terms necessitates a change in nany of the values in order to be consistent with the licensing basis accident analyses performed by Duke L Power Company. It should be noted that the Trip Setpoints and Allowable Values remain unchanged unless specifically addressed as a l i proposed revision. Prer ered Pe'/irien to Technical 9recifica*ien Table 3.1-1 The table is revised to include all accident analyses that would i tequire reevaluation in the event that one full-length control rod is l inoperable. Terhnical Ju s t i f i c at ien This change is the same as that approved fcr McGuire Units 1 and 2. The existing Table 3.1-1 listed the rod cluster control assembly misalignment event when, in fact, the broader-scoped event of rod miscperation should have been given. This accident includes the static
^
misalignment as well as single rod withdrawal, dropped rod and dropped bank events, which might.all be impacted by the inoperable rod. The large break LOCA analysis, which was listed in the table, does not take credit for any control rod insertion and should therefore be removed. 3
. _ - . _ . _ . _ _ _ _ . _ _ __~ _ . . _ _ . , . . _ _._ _ ._, _ __ -_ _ - ,
. . - . . - . . . . . - . _ . - _ - . - - . _ _ _ . . _ . _ ~ . ~ . . - - . . . . . - ~ - . -
1 Prencrad Revirien to Technical 9recificatirn 3/4.2.2 i Tne change to SR 4.2.2.2.c.2.a and SR 4.2.2.2.c.2.b is to reflect Duke l Power vs. vender (B&W) methods for surveillance (i.e. remove NSLOPE and l PSLOPE frco SR'4.2.2.2.c.2.a) and provide clarification of SR ; 4.2.2.2.c.2.b action requirements. The change to SR 4.2.2.2.d surveillance method is to provide clarification of 4.2.2.2.d. i i I Technicni Justificatirn The changes were made to provide surveillance requirements consistent ! with Duke Power Company methodology for core power distributien control and surveillance of the heat flux hot channel factor, as discussed in DPC-NE-2011PA (Reference 2). The nomenclature in SR 4.2.2.2.c.2.a is changed to reflect Duke Power vs. Vender (B&W) nomenclature. SR 4.2.2.2.c 2.b provides clarification of.which margin is to be used for this action. SR 4.2.2.2.d requirec projections of the measurements to be made to determine at what point FQ(X,Y,Z) would exceed the allowable limit if the current trend continues. However, for the case when the projection of the current trend would indicate that the margin is > increasing, the actions f or the current negative margin (f rom SR 4.2.2.2.c.2 and 4.2.2.2.c.3), would be. sufficient since this margin is less than the extrapolated margin. In this case it would not be necessary to penalize the current margin based on the results of the extrapolation, since the extrapolation indicates an improvement over time. Prcncred Pevisien to Technical Grecificatien 3/4.2.3 Specification 3 /4.2.3 was revised to' reflect the power peaking surveillance method described in DPC-NE-20llPA. These revisions are summarized as follows:
- 1. The statement-Specification 3.2.3 LCO was revised to reflect <
Duke's nogenclature fer the nuclear enthalpy rise hot channel factor _[Fug(X,Y)L and related parameters required by the methodology of DPC-NE-2011PA vs. vendor (B&W) nomenclature and used throughout the Reload Report.
- 2. There are no changes to Specification 3.2.3 Actione a, b, c and d other than nomenclature as mentioned above.
- 3. -SR 4.2.3.2.b addresses the frequency for confirming that FDH(X,Y) is within its limit. In addition to performing.the surveillgnce at le _ once per 31 EFPD, the revised surveillance requires measui- ent of the-peaking factor whenever the excore quadrant power tilt ratio is ncrmalized using incore detector measurements. This ensures that the impact of any core tilt on FDH(X,Y) will be-determined and reflected in the margin
-calculation. -This-is comparable to SR 4.2.2.2.b in the Fo(X,Y,-Z)-
specification. SR-4.2.3.2.b.1 requires-a surveillance to be performed upon reaching equilibrium conditions after exceeding by 10% or more of. RATED THERMAL POWER at whichMF DH(X,Y) was lost determined.
- 4. ThepurposeofSR4.2.3.2.c.1istoperformmargincalcugggions based on the measured radial peak. The limit [FdH(X,Y)]* to which the measurement is. compared is based on the allowable design MARP limit, increased by a factor that represents the maximum amount that the power at the given assembly location can r
I.
-4
1 l l 1 increase abc.c the design value before the measured value may become limit;ng. Part c.: uses the amount of margin determined by this proce ure t o f orm the basis tor the amount of power level reductica and tne reduction in the high flux and OTDT K1 trip I setpoints reTaired in the AC'ICN statements fcr the i specificaticn. t
- 5. SR 4.2.3.2.d has been changed to reflect Duke's nomenclature and to allc.. fcr the case wnen margin may te increasing over time.
This surveillance requires projecticns of the measurements tc be made to determine at what p0 int F DH fX,Y) would exceed the allowable limit if the gurrent trend continues. In part d.1 a penalty is applied to FOHtX,Y) if the trend indicates that the measured peak would exceed the limiting peak within the 31 EFPD surveillance period, and the margin calculaticns are repeated. In part d.2, the measurement is obtained ano the margin calculations are repeated ;that apprcpriate actions can be taken before cero margin is readhed. ) 1 Technical Justific;ticn l Specification 3/4.2.3 was revised to provide required actions and surveillance requirements consistent with Duke Power Company methodology ter ccre power distribution control and surveillance of the nuclear enthalpy rise hct channel facter, es discussed in DPC-NE-2011PA (Reference 2). The nomenclature is cnanged to reflect Duke Power nomenclature vs. Vendor (B&W) ncmenclature and clarification of l surveillance requirements is provided to te consistent with Duke's methods described in DPC-NE-20llPA. SR 4.2.3.2.b.1 is changed to better clarify the surveillance , requirement and to ensure that the plant is at equilibrium conditions prior t o a measurement. The surveillance requirement also has a provision similar to the requirement it zuplaced stating that during power escalation at the beginning of each cycle, THERMAL POWER may be increased unti.1 a pcwer level f or extended cperatien has been achieved anr1 a power distribution map cbtained. SR 4.2.3.2.c is changed to reflect Duke Power vs. vendor (B&W) nemenclature. This surveillance requirement i= comparable to the SR 4,2. 2. 2. c on Fg (X , Y,2) . SR 4.2.3.2.d is changed to reflect Duke Tcwer vs. vendor (B&W) nomenclature and to provide clarification of 4.2.3.2.d. An additional check has been added to determine if the margin is increasing. This ' surveillance requirement is comparable to the new SR 4.2.2.2.d on FgtX,Y,2). Prcrcsed Revirien t- Technical Frecificatien 3'4.2.5 This-revisica-is intended to provide consistency between Duke's
-Westinghouse plants, and correct an action item.
Technical f uc t i f i c at i r-n This change is administrative in nature becaust it corrects a typograghpical error. It does not represent an actual change to the Requirement of 3/4.2.5. (. 5
- _ _ . ~._____ _ ._ ._ _ _. ~ _ . ___ _ __ _. _ _ _ _ _ . . - - _
l Prercrad Pevicien te Technir31 frecificatien Table 3.7-: 2 This channe is to delete the re tor trip en power range neutron flu" negative rate fron the React:r Protection System. Technical Justificatirn Refer to the technical justificaticn f or the prcposed :evisicn to TS Table 2.2-1. Prercred Revisien t c Tech *.iral_gecificaticn Table 3.3-2 The reactor trip cn power range neutrcn flux negative rate is deleted. The response times associated with the RTD Bypass System f or the OTAT and OPA" are deleted, as is footnote regarding RTD Bypass System. Neutrot M"tector response time exemption j a added to CPAT trip. Technical Justificatien For the deletien of the po.ter range neutrcn flux negative rate, refer to the tecnnical justificaticn for the proposed revision to TS Table 2.2-1. For the response times and footnote regarding RTD Bypass System, the RTD Bypass System has heen deleted. The neutron detector response time exemption which is applicaLle to the OTAT- trip is now applicable to the OPAT trip due to the addition of the f;(AI) : f unction to the CPAT trip. Refer to the technical justification for the proposed revision to TS Table 2.2-1. Preresed Pevisien tc Technical Frecificatien Table 4.3-1 This change is to delete the reactor trip on power range neutron flux negative rate from the Reactor Protection System. Technical Jus *ificaricn Refer to the technical justification for the proposed revision to TS 1 Table 2.2-1. Prcresad Revisien tc Technien1 Frecificatien 1.3.2 This proposed Technical Specification (TS) change deletes ACTION 3.3.2.b.1 and equation 2.2-1. This-change is consistent with the , proposed change.to Technical -specification 2.2.1,. and the removal of the' Total Allowance-(TA) column, the-Z column, and the Sensor Error (S) column from Tables 2.2-1 and 3.3-4.
- Technical Justi ficatien l3 ACTION 3.3 2.b.1 provides the option of declaring an instrumentation channel operable by use of equation 2.2-1 when the ESFAS Instrumentaticn or Interlock Setpoint is less conservative than the Allowable Value. Use of this option necessitates the inclusion of-the Total Allowance (TA), the value Z, and the Sensor Error 'S) terms in TS 6
V -
~ _ _ . _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ - _ _ _ _ . _ . _ _ _ . - - _ _ _ _
l a 1
)
J 4
! 3.3.0 and Table 3.3-4. Peth ACTI2n 3.3.2.b.1 and 3.3.2.b.2 requile ;
that the Setpoint te adjusted consistent with the Setpoint value given in Table 3.3-4 however, deleticn of ACTIOM 3.3.2.b.1 and equation 2.2-1 makes TS 3.3.2 more rer;rictive. in that the channel must be declared inoperable with the EF:ns Instrumentaticn er Interlock Setroint less chaservative than th' Allowable Value. The deletien of ACTICM 3.3.2.b.1 and equatisn 2.2-1 improves the safe cperation of the plant by reducing the ccmplexity of the Technical Specifications. This change also provides increased conformity with the comparable Technical Specification f or McGuire Maclear Station prcrm ed Peri m n tc TecMical Necifi aticn Table 3.3-4 This proposed revision changes the low steam line pressure setpcint for safety injecticn and main steam line isolation from 725 psig to 775
, psig. The-allcwable value ftr this trip functicn is changed frcm 694 i psig to 744 psig, maintaining the same 31 psig allowance for rack uncertainties, and the lead-lag ccntroller for steam line prescure-low is deleted. In addition the table is revised to delete the Total Allowance (TA) column, the Z column, and the Sensor Error (S) column f rcm Table 3. 3 -4. This change is censistent with the prcposed change to Technical Specifications 2.2.1, 3.3.2 and Table 2.2-1. Lastly. the Allowable Values associated with RTD Eypass System for the Feedwater isolation on Tavg-Low and ESFAS P-12 interlock on Low-Low Tavg are deleted.
Technical J e tificatien 9 For TS Table'3.3-4 items 1.e , 4.d, and lead-lag removal: The higher steam line pressure setpoint is consistent with all licensing basis safety analyses. The steam line break event was reanalyced (Reference 4) assuming an uncompensated low steam line pressure setpoint of 700 psig, allowing 75 psig for instrument j uncertainty and margin. This reanalysis shows this Condition IV transient does not exceed the impcsed Ccndition II acceptance criterion of no DNB. i The inadvertent cpening of a steam generator relief or safety valve, in terms of primary system overcocling, is essentially a small steaa line break. This Condition II event exhibits steam releases markedly less l than the' steam line break event. Therefore, this event is bcunded by j the steam line break event and d:es not require reanalysis. This change does not necessitate reanalysis of the peak centainment temperature analysis. The removal of the dynamic compensation of the steam line pressure signal, even with the increased setpoint, will delay main steam line isolation following certain breaks in the main steam line. Since the effect cf main steam isolaticn is to terminate
-the blowdown of the intact steam generators, postponing MSIV closure will tend to retard the b1cwdown of the faulted generator.and delay tube uncovery. Also, the extended blowdown ci the intact steam generators will reduce the Reacter-Coolant System temperature, which will further delay tube uncovery. Once uncovery does occur, this Icwer primary system temperature will effectively reduce the enthalpy of the break flow.
The removal of the dynamic compensation of the steam pressure signal, l which acccmpanies the change in the Icw pressure setpoint. sill essentially eliminate the spurious ESF actuation on minor (but rapid) l l l
O pressure decr easec in t he seccndary system, In additicn. the increase in containment temperatute following a steam line break is caused by 4 the release of steam f rom t he steam generatcrs to the containment atmosphere via the break. This steam initially comes from two sources, I the steam generatcr ccnnected to the steam line with the break and the cther three steam generatcra. In the sub.eguent discussicn. the tormer , is referred to as the faulted SG. wnlle the latter are referred t o as the intact SGs. Since the most limiting break locations are between the f ault+J SG and its MSIV, the steam line isolation function does not prevent tiowdown from that SG. However, even for a single MSIV : failure, steam line isolation terminates the blowdown from the intact SGs. Because cf the energy content of the RCS, steam line breaks occurring from an at-power initial ccndition are limiting for peak containment temperature. During power Operation, steam line isolation is actuated by low steam line pressure or high-high containment pressure. The current ccnfiguration for the steam line pressure actuation has lead / lag ccmpensation on the pressure signal. For a decreasing measured pressure in the steam line, this compensated signal will fall f aster than the measured pressure signal. Therefore, for a given transient in measured pressure and a given isolation setpoint, isolaticn en compensated pressure will occur faster than isolation on ' measured pressure. As the rate of decrease of the measured pressure slows (smaller breaks), the compensatica becomes less important, resulting in compensated pressure and measured pressure becoming more ;
~
nearly identical. For this situation the isolation setpoint as a more important factor, in determining isolaticn time, than is the presence ' of the lead / lag compensation. The proposed Catawba 1 Cycle 7 Technical Specification change removes the lead / lag compensaticn and increases the isolation setpoint from 725 psig to 775 psig, This change, as explained above, would result in later steam line isolation for larger breaks and earlier isolatien for smaller breaks. The proposed Technical Specification change results in a faster, and therefore more conservative, isolation time fcr these smaller breaks. For the three cases for which mass and energy releases are presented in the Cataaca FSAR, credit for steam line isolation on low steam line pressure is taken only for the 1.4 ft) double endel breah case. Steam line isolation for this case occurs at 11 seconds. However, high-high containment pressure is actuated at 3'psig per the Technical Specifications. A safety analysis setpoint of 3.75 psig bounds the instrument uncertainty associated with the measurement of containment pressure, and the time to reach containment pressure of 3.75 psig is 0.85 seconds. Adding 7 seconds f or the maximum allowable Technical Specification response tine for steam line isolation on high-high containment pressure gives 7.85 seccnds for completion of steam line isolatien, compared to the 11 second time given in the FSAR. Therefore, even if there were a delay, due to removal of the lead / lag compensation, in reaching th' '"v steam line pressure isolation setpoint for larger breaks 'SAR mass and energy releases would remain ccnservative becaut ation would be accomplished-en high-
- high containment pressure tior to the time that isolation is credited in the FSAR.
For the removal of the Total Allowance (TA) column, the Z column, and the Sensor Error (S) column from Table 3.3-4: In additicn to the technical justification provided for the proposed i revision to Technical Specification 3.3.2, the removal of these terms (- from Table 2.2-1 improves the cafe operation of the plant by improving 8
- .. - _. ~ __- - - - .. . - - . . _ - - . . - . _ ._. -.- - - . _ - . . -
the readability of the table. It should be noted that tne Trip Setpoints and Allowable Values remain unchanged unless specifically addressed as a preposed revision. For TS Table 3.3-4 items 5.c, 18.c, and Table Notations: The RTD Bypass System has been deleted. upposed Revivien to Technical Frecificaticn Table 3.3-5 Two. response times are modified in this prop. sed change, the feedwater isolation response time is changed from 7 seconds to 12 seconds and the ' steam line isolation time is changed from 7 seconds to 10 ceconds. lechnical Justification i The extended response times are consistent with or conservative for all licensing-basis safety analyses. These two rerponse times are
- assumptions in the steam line break analysis. The 12 second feedwater l- isolation time and the 10 second steam line isolation times have been i employed in the steam line break analysis apprcved in Reference 4.
Increasing these response times, from the current Technical Specitication values, causes the prinary system overcooling to worsen due to the extended blowdown of the intact generators and the additional mass of main feedwater delivered to the faulted generator. Reanalysis (Reference 4) shows this Condition IV transient does not exceed the imposed Conditien II acceptance criterion of no DNO, The inadvertent opening'of a steam generator relief or safety valve, in terms of prima _y system overcooling, is essentially a small steam line - break. This Condition II event exhibits steam releases markedly less than the steam line. break event. Therefore, this event is bounded by
.the steam line break event and does not require reanalysis. ,
Y The increased feedwater isolation response time also impacts the ! analysis of the excessive feedwater flow event, The effect in negligible, however, since the DNBR decrease is terminated by the turbine trip, which occurs 3 seconds after the high-high steam generator level setpoint is reached. The fact that the feedwater isolation' valves close 12 seconds after this setpoint is reached rather than 7 seconds does not affect the minimum DNBR achieved. For the-peak containment -temperature analysis (Reference 6), a lower-m li _ cui mass in the f aulted generator yields censervative results. 'This leads to an earlier uncovery of the steam generator tubes and,.thus, the advent of superhea*ed steam exiting the breah. Lengthening the feedwater-isolation response time will increase the amount of feedwater. delivered to the fa..ted generator and delay tube uncovery. Increasing the main steam line Asolation response time has a similar effect as the removal of the dynamic compensation in the steam line pressure signal
- i. discussed in the technical justification for the proposed change to TS Table 3.3-1. In addition, as. explained'in Item 7 of Section 6.2.1.4.1-of the Catawba FSAR, the mass and energy releases werc; calculated
~
beyond the point of psak containment temperature. As shown in Tables 6-47,.6-49,'and 6-51 of the Catawba FSAR. the mass and energy release.3 L from.the faulted SG are not zero at the end-of'the release calculation, indicating 'that inventory remains in the f aulted SG at that time, which
~is after the time of peak containment temperature. Therefore, the addition of feedwater beyond the amount assumed in the FSAR analysis, l due to an increase of feedwater isolation time cf 5 seconds, would l-l t
I ( i 9 l I
. . , . ~ . . - _ . _ . _ _ . . _ _ _ _ _ . _ _ _ _ _ . , - - _ _ _ - . . - _ _ _ _ _ _ . _ . _ _ > _ - .
b simply it. crease the amcunt of inventory remaining in the faulted SG at the end of the mass and energy release calculation. It would not increase the calculated. containment temperature.
- Significant' difficulty has been experienced in meeting the current specification response times for both of these ESF functions.
Increasing the allowable response times should eliminate this dif.ficulty. Prerosed Revir ien t e Terhm c31 90e ri fic ation 3. 4 :1. 2 The specification is being cnanged to require that the three operable reactor coolant locps be in creration in Mode 3. Iechnical Justification This restriction is imposed in order to make the specifications consistent with the reanalysis of the uncontrolled bank withdrawal f rom i- subcritical or Icw power startup condition. E s M 2evisien to Technical Frecificatien 3.4.2.1 & 3.4.2.2 This mcdification changes the tolerances on the pressuriner safety valve lift setpoint from 21% to +3%,-2% in all modes ot operation. Technical 'ustificatien The pressurizer code safeties are not tested in place but are removed and shipped to a testing facility. The safety concerns for removal and replacement of these valv: - are difficult access to the work area, difficulty in lifting dev._e rigging for valve removal / replacement, and valve transport to/frem the pressuricer. Since this work is performed in a radiological envircreeat, work activities are further complicated by anti-contamination clorhing. For a conservative approach, all three. valves are removed each cutage for testing. The setpoint drift seen during testing would again fall under the proposed setpoint variance change. The change would possibly reduce work in the pressuriner by , 66% by requiring only one valve to be tested per outage. In sammary, cafety henefits.would be gained by less work in a dangerous environment and-less' radiation exposure-. The larger allowable deviation from the nominal lift setting is consistent'with the licensing basis analysec. An increased pressuriter l- safety valve lift setpoint impacts the peak Reactor Coolant System ! pressure calculated for pressure increase transients. A pressure l increase is the result of a heatup in the Reactor Coolant System due a L mismatch between the her. generated in the reactor core and.the heat removed by the secondar system. The three accident categolies . involving such heat transfer mismatches are the decrease in secondary
. heat removal, decrease in Reactor Coolant System flow rate, and l reactivity and power distribution anomaly transients. The feedline break, locked rotor and rod ejection events are the limiting pressure
. increase transients in these three accident categories, respectively.
These events have all been analyzed assuming a lift setpoint 3 percent above the nominal value. These analyses show that the peak Reactor Coolant System pressure criterien, 110% of design pressure, is met for the f eedline break and 13cked rotor events. The peak Reactor Coolant System pressure critericn of 120% of design pressure is met in the rod ejection-accident analysis. - 10
l l l l The amount by which the safccy valve lift setpoint is allowed to drift j downward is restricted to 2 percent of nominal in crder to ensure that safety valve lift cannet preclude reactor trip en high pressuriser pressure. For CNB-transients in which a high pressurl:er pressure reactor trip does not prevent the lifting of the safety valves, the effect of this reduced setpoint on the transient CNBR is evaluated. Since low pressure is ccnservative for DNBE analyses, it is typically assumed that the pressuriner PORVs and sprays mitigate the pressure , increase due to the system heatup and thereby preclude safety valve lift. For the uncontrolled tank withdrawal at power and single rod , withdrawal events, hcwever, the cperation of the pressurizer pressure i ccntrol system would tend to yield an earlier reactor trip en overtemperature AT due to pressure compensation of the trip setpoint. The reanalysis et these events show that all acceptance criteria are met. Prcrosed Revision to Technic.1 Erecification 3.4.6.2.c This change reduces the allcwahle total reactor-to-secondary leakage i rate. Technical Justificatirn This revision is reqaired to limit primary-to-secondary leakage to , values which-ensure that the offsite doses for FSAR Chapter 15 transients and accidents meet the applicable f raction of the 10 CFR 100 limits. The change is necessary to account for an increase in the predicticn of the number of fuel pins exceeding the CNBR limit for the locked rotor transient. This transient was reanalyzed for C1C7 using approved Duke Power methods. The revised dose calculation also incorporates the impact of steam generator tube bunale uncovery, which results in less credit for iodine partitioning in the steam generator secondary. t Prcroped Revicien to Tachnical Snecificatien 3.5.1.1 This change raises the required average cold leg accumulator baron i concentration in ACTICNC c.2 and c.3 from 1500 to 1800 ppm, and bases this average on all foir accumulators instead of just the limiting. three. Igchnical Justification Calculating the volumetric average boron ccncentration based cn all four cold leg accumulators is valid, since, regardless of the break location, the contents of each accumulator will be emptied (either directly or 'adirectly) into the contairment sump. A volumetric average concentration of 1800 ppm will ensure long-term subcriticality folluwing a LOCA. Procesed Pevisica_;c Techni cal _ Sreci fication 4. 5. 2 If & h) T. S. 4.5.2(f) gives the ECCS pump performanca requirements. The centrifugal charging pump required developed head is decreased from 2380 to 2223 psid. The safety injection. pump required developed head is decreased from 1430 to 1341 psid. T. S. 4.5.2(h) gives the ECCS delivered flo'.. requirerents ~he centrifugal chargina pump *:tal flew l I -
_ = _ . _ _ l: i l. l rat e is decreased f rom 565 to 560 gpm. The safety injection pump total flou rate is increased from 660 to 675 gpm. These revisions are applicable to Unit 1 and Unit 2. Technical Justi ficat ien The ECCS pump required developed head and delivered f. low specificatione are being revised to provide performance test margin for periodic testing, and to be ccnsistent with revised ECCS pump and system performance. requirements. These specifications are consistent with the
. current safety analyses and with the revised analyses associated with Catawba 1 Cycle 7. The ECCS pump runout flow rates are being adjusted to reflect revised pump vender information. Performance test margin is required to enable suf ficient allowances f or instrument err ors and to permit . reasonable t est acceptance criteria. Pump performance at the new specific tion values is sufficient to meet all acceptance criteria in both the current FSIR analyses, and in t he revised analyses submitted with or ref erenced by the Catawba 1 Cycle 7 reload. The proposed revisiens are therefore acceptable.
I Prenmed Peri sinn t o Tec - iical Soecification Table 3.Gda ; This change clarifies the required maximum stroke titte of the feedwater isolation valves, main steam isolation valves, and main steam isolat ion bypass valves. The numerical value of the stroke time of these valves l is changed to NA. j Te"hnical Justification ;
. - l Tae justification f or the change in valve stroke time as it relates to :
system thermal-hydraulic-response during a steam line break event is -f presented for the change to Technical Specification Table 3.3-5. ! Although these valves are included in Table 3.6-2a, the list of i cortainment isolation valves for Unit 1, these valves do not receive a -l cont ainment isolation signal. These valves perform a containment l isolation function only to the extent that credit for their operation 1 might be taken in the dose analysis. Since these valves receive no I containment isolation signal, and credit for the operation of these ; valvus is not taken in the dose analysis, a maximum stroke time is not l
-appltcable-for these valves. ,
l Pronesed Revision to Technical Foecification 4.7. M l i The permissible stroke time for the main steam isolation valves is : changed from 5 to 8 seconi l 1 Technical Justification i, The larger allowable isolation valve stroke time is consistent with or , conservative for all licensing basis safety analyses, The valve stroke ! time, when added to the applicable instrumentation delays, yields the j overall ESF response-time. As stated in the technical justification i for the proposed revision to TS Table 3.3-4, this response time is ! input to the steam line break transient analycis. Increasing the stroke j time causes the primary system overcooling to warsen due to the j extended blowdown of the intact generators. Analysis using the approved j methodology in Reference 4 shows this Condition IV transient does not ! violate the imposed Condition II acceptance criterion of no DNB. l 1 i l i l-t 12 I
The inadvertent cpening of a steam generator relief cr safety valve, an terus of primary system overcOcling. is essentially a small steam line break. This C n.iticn d II event exhibit s steam releases markedly less than the steam line break event. Theletcre, this event is bounded by the cteam line break event and d:es not require reanalysis. For the peak ccntainment temperature analysis (Reference 6), increasing the main steam line isolatacn response time has a similar effect as the removal cf the dynamic compensation in the steam line pressure signal discussed in the tecnnical justification for the proposed change to TS Table 3.3-4. In addit ion. the issue of peak containment temperature is one cf eguipment qualification tcr the post-accident envircnment. Fcr the following reasons thIs proposed Tect.nical Specificatica change is Judged to be acceptable: Since the ef f ect of steam line isolation is to terminate the blowdcwn of the intact SGs, post; ning MSIV clcsure will tend to retard the b)cwdcwn of the taulted SG and delay tube uncovery. Also, the extended blowdown of the intact SGs will reduce the RCS temperature, which will further delay tube uncovery Once unccvery does occur, this er PSS temperature will result in a reduced break flow enthalpy It '
. uld also be noted that, whether the MSlV stroke time is 5 seconds cr 8 seccnas, the isolation f unction is ccmpleted well bef ore tube uncovery Except ter the non-lirmiting 1.4 ft 3 double-ended break, the peak containment temperature occurs at a relatively long time after the completion of isolation, even if tne strcke time is extended to B seconds:
Ereak Sice (ft2) 0.86 0.4 5 seccnd Isolaticn Time (sec) 17.5 26 8 secund Isolation Time (sec) 20.5 29 Peak Temperature (sec) 127 >150 Frca reviewing the mass and energy releases in Tables 6-50 and 6-52 of the Catawba FSAR, the following infermation can be determined: Break Sice (ft 2) 0.36 0.4 FSAR Intact SG Mass Release :lbm) 24,730 16,830 Intact SG Mass Release Rate (lbm/ sect 1490 650 Releases During 3 Additional Sec 4470 1950 (1bm) Estimated Intact SG Mass Releases 29.250 18,780 (lbm) It can be seen that, while the intact SG releases during the FSAR 0.86 ft 2 break are more than 451 larger than during the 0.4 ft 2 break (24,780 vs. 16,830, a difference of 7950 lbm), the difference in peak containment temperatures given in Figures 6-21 and 6-22 of the Catawba FSAR is barely discerr.ible (327.7oF vs. 326.30F). This small difference is expected since the containment temperature is primarily sensitive to the higher enthalpy later releases from the faulted SG. The additional intact SG releases due to extending the valve otroke time are, as shcwn above, conservatively estimated to be 4470 lbm and 1950 lbm. These values are much less than the 950 lbm difference between the intact SG releases fcr the two break sices, ahich resulted in an insignificant change in peak containment temperature. Therefore, the additional release due to increasea valve strcke time during this period of the transient will not cause a significant incrnase in the re- rc a: nman- t=~per a re. 13
. . _ _ , _ . . _ . ._ .- _ . - - _ _ _ _ _ _ _ . . _ . _ _ . ~ . . _ _. _ . . _
f There is margin-in the current equipment qualification limit of 3400F , to acccmmodate an increase of 12.3 F (above the FSAR result of 327.7 F) due to any effect of increased total mass and energy released from the intact steam generators due to selayed completion of the steam line , isolation function. In su=mant, the release of saturated steam for 3 additional seconds from the intact SGs, during a period well before the peak containment temoerature, would not cause a significant impact en this peak t en,perature . The magnitude of the additional releases is much lese than the difference between intact SG releases ter the different FSAR split break cases, which have resulting temperature differences of 0.220F. Even if such a release were to cause a nonnegligible
- temperature increase, there is approximately 12 F margin between the current FSAR result and the equipment qualification limit.
Significant difficulty has been experienced in. meeting the current specification stroke time for these valves. Increasing the allowable stroke time should eliminate this difficulty. Pretored F4visien to Terhnical Erecification 6.9.1.9 Changes to Specification 6.9.1.9 are as follows. Fk(XY,Z) was added + to the list of items for Heat Flux Hot Channel Fact 6r, note '*** was added to W(Z) BL, FAHRL was changed to FAHL(X,Y), reference numbers in note '**' were changed and the reference list was updated. Technical Justificatien TheadditionofFh(X,Y,Z) and the change of FAHRL to FAHL(X,Y) were changes in nomencaature due to a methodology change. The new nomenclature is consiscent with Duke's methods described in DPC-NE-2011rA. The *** note had previously been excluded from W(Z)en, so it was added. The references for the **** note were changed since the references changed. The reference list was updated to include references describing the new methodology. a a f d 3 14
- . - ~ . - - - - . . . . - - - - . _ - _ - - .- . . . - . -. . - . - . . - . ~ . . - . . .
Feferences
- 1. BAW-10159P-A, BWCMV C:rrelation of Critical Heat Flux in Mixing Vane Grid Fuel Assemb]ies, Babcock & Wilcox, July 1990.
- 2. DPC NE-2011P-A, Duke Power Company, Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, March, 1990.
1 BAW-10174-A, Mark-BW Reload LOCA Analysis for the Catawba and McGuire Units, Babcock & Wilcox, May 1991.
- 4. DPC-NE-3001P, Duke Power Company, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology, Revision-1, November 1991.
- 5. CPC-NE-2004P-A, Duke Power Company, McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01, December 1991.
- 6. WCAP-10988, Cobra-NC, Analysis for a Main Steamline Break in the Catawba Unit 1 Ice Ccndenser Containment, Westinghouse Nuclear Energy Systems, November 1985, 1
A l l l l { t l l 15 1 .
s 4 .A 4 JL_ le _rJ. 5 .w. M-- s-.me-. ne. .,'e.4-g .J>.h&4JimAst M a,s A'a da5 A 24-M ar 4_A d,N. i -g ,p a- 4. A M4 & 4 se de 4 c.{ .,A,, .2 k Attachinent 3 4 5 I C h J 4 I 1 l li I-i .
, - - - ., -, - .--,_ .----- ., - - ~ . - .. - --- . . -,-,-c, - - - ,
NO SIGNIFICANT IIAZARDS ANALYSIS The following analysis, required by 10 CFR 50.91, concludes that the proposed amendment will not involve significant hazards consideration as defined by 10 CFR 50.92. 10 CFR 50.92 states that a proposed amendment involves no significant ha7ards consideration if operation in accordance with the proposed amendment would not:
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2) Create the possibility of a new or different kind of accident from any previously evaluated; or
- 3) Involve a significant reduction in the margin of safety.
POWER DISTRIBUFION AND SAFETY LIMITS Catawba Unit 1 Cycle 6 was the first Duke Power Nuclear Station for which B&W Fuel Company (BWFC) supplied the reload fuel. The Catawba Unit 1, Cycle 6 Reload Report presented an evaluation that cor.cluded the core reload using Mark-BW fuel would not j adverely impact the safety of the plant. This reload report is similar, but reflects that ! Duke Power performed the analyses in support of the operation of Cycle 7 rather than l BWFC. The Catawba Unit 1, Cycle 7 Reload Safety Evaluation Report (Attachment 1) presents an evaluation which demonstrates that the core reload using Mark-BW fuel will not adversely j impact the safety of the plant. During Cycle 7, the core will contain 72 fresh fuel assemblies,72 burned fuel assemblies supplied by B&W and 49 Westinghouse supplied Optimized Fuel Assemblies (OFA). l A LOCA evaluation for operation of Catawba Nuclear Station with Mark-BW fuel has been completed (BAW 10174, Mark-BW Reload LOCA Analysis for the Catawba and McGuire Units). Operation of the station while in transition from Westinghouse supplied OFA fuel to B&W supplied Mark-BW fuel is also justified in this topical. BAW-10174 demonstrates that Catawba Nuclear Station continues to meet the criteria of 10 CFR 50.46 when operated with Mark-BW fuel. 12rge Break LOCA calculations completed consistent with an approved evaluation model (BAW-10168P and revisions) demonstrate compliance with 10 CFR 50.46 for breaks up to and including the double ended severance of the largest primary coolant pipe. The small break LOCA calculations used to license the plant during previous fuel cycles are shown to be bounding with respect 1
m _ - _ - - . _ _ _ . _ - _ _ _ _ . __ _ 1 to the new fuel design. This demonstrates that the plant meets 10 CFR 50.46 criteria when
. the core is loaded with Mark-BW fuel.
During the transition from Westinghouse OFA fuel to Mark-BW fuel, both types or fuel
~
assemblies will reside in the core for several fuel cycles. Appendix .A to BAW-10174 demonstrates that results presented above apply to the Mark-BW fuel in the transition core, and that insertion of the Mark-BW fuel will not have an adverse impact on the cooling of ; the Westinghouse fuel assemblies. Duke Power Company's Topical Reports DPC-NE-3000, DPC-NE-3001, and DPC-NE-2004 provide evaluations and analyses for non-LOCA transients which are applicable to Catawba. The scope of these analyses includes all events specified by sections 15.1-15.6 of Regulatory Guide 1.70 (Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants) and presented in the Final Safety Analysis Report for Catawba. The analysis and evaluations performed for these topicals confirm that operation of Catawba Nuclear Station for reload cycles with Mark-BW fuel will continue to be within the previously reviewed and licensed safety limits. One of the primary objectives of the Mark-BW replacement fuel is compatibility with the
~
, resident We-tinghouse fuel assemblies. The description of the Mark-BW fuel design and the thermal-hydraulics and the core physics performance evaluation demonstrate the similarity between the reload fuel and the resident fuel. The extensive testing and analysis summarized in BAW-10173P shows that the Mark-BW fuel design performs, from the standpoint of neutronics and thermal-hydraulics, within the bounds and limiting design
- criteria applied to the resident Westinghouse fuel for the Catawba plant safety analysis.
Each FSAR accident has been reviewed to determine the effects of Cycle 7 operation and to ensure that the radiological consequences of postulated accidents are within applicable regulatory guidelines, and do not adversely affect the health and safety of the public. The ; design basis LOCA evaluations assessed the radiological impact of differences between the Mark-BW fuel and Westinghouse OFA fuel fission product core inventories. Also, the - dose calculation effects from non-LOCA transients reanalyzed by Duke- Power were evaluated using Cycle 7 characteristics. The calculated radiological consequences are all within specified regulatory guidelines and contain significant levels' of margin. The analyses contained in the referenced Topical Reports indicate that the existing design criteria will continue to be met. - Therefore, the enclosed TS changes will not increase the probability or consequences of an accident previously evaluated.
- As . stated in the above discussion, normal operational conditions and all fuel-related transients have been evaluated for the use of Mark-BW fuel at Catawba Nuclear Station.
i Testing and analysis was also completed to ensure that, from the standpoint of neutronics i and thermal-hydraulics, the Mark-BW fuel would perform within the limiting design criteria. Because the Mark-BW fuel performs within the previously licensed safety limits, 2 l
the possibility of a new or different accident from any previously evaluateo is not created. The reload-related changes to the TSs do not involve a signi6 cant reduction in the margin of safety. The criculations and evaluations documented in BAW-10174 show that Catawba will continue to meet the criteria of 10 CFR 50.46 when operated with Mark-BW fuel. The evaluation of non-LOCA transients documented in DPC-NE-3001 also confirms that Catawba will continue to operate within previously reviewed and licensed safety limits. Because of this, the TS changes to suppo't the use of Mark-BW fuel will not involve a significant reduction in the ma gin of safety. An administrative change is being made to TS Tables 2.2-1 (Reactor Trip System Instrumentation Trip Setpoints), and Table 3.3-4 (Engineered Safety Features Actuation System Instrumentation Trip Setpoints). Since these tables contain values that are not identical for each unit, a separate table will be provided for each unit. The pages will be labeled " Unit 1" or " Unit 2", and there will be an "A" in the page number for Unit 1 and a "B" in the page number for Unit 2. The TS Tables will be copied on white paper for Unit I and on yellow paper for Unit 2 to further distinguish applicability. Table 3.3-4 will also have references to the R fD bypass system deleted, since the RTD bypass system has been removed, and they no longer apply. These changes are admin;strative in nature, and are being made only to clarify the TS. Since they involve no change in requirements, they involve no signi6 cant hazards. . 1 REMOVAL OF TOTAL ALLOWANCE Z AND SENSOR ERROR FROM TABLES 2.2-1 AND 3.3-4 The removal of the Total Allowance, Sensor error, and Z columns from Tables 2.2-1 and 3.3-4, along with the deletion of TS 2.2.1.b.1, 3.3.2.b.1, and equation 2.2-1, which provide for the use of these values, do not involve any significant hazards consideration.
- These speci6 cations provide the option of declaring instrumentation operable when the setpoint is less conservative than the allowable value. This is done through the use of equation 2.2-1. With the deletion of Speci6 cations 2.2.1.b.1, 3.3.2.b.1, equation 2.2-:,
and the Total Allowance, Sensor Error, and Z columns from Tables 2.2-1 and 3.3-4 the channel must be declared inoperable with the setpoint less conservative than the Allowable Value. This change is more conservative than the current requirements, and therefore involves no significant hazards. DELETION OF NEUTRON FLUX IIIGII NEGATIVE RATE TRIP The removal of the Power Range Neutron Flux High Negative Rate trip will not result in any previously-reviewed accident becoming more probable or more severe. The trip is a response to a pre-existing transient condition and would not initiate any accident. The trip is designed to provide protection from a dropped control rod. However, in the event of 3 r ~ ,- e
, ~ .. _ _ _ . _ _ _ . . _ _ _ . - . . ______ _ - . . . _ . . _ a dropped rod, the reactor is assumed to trip on low pressurizer pressure. Therefore, the protection function is retained. The consequences of a dropped rod have been analyzed and found to be within acceptable limits. Likewise, the removal of this trip will not create a new accident not previously reviewed. The removal of a response to a transient will not initiate a new transient. There are no credible unanalyzed transients which will occur as a result of a dropped rod. The removal of this trip will reduce the potential for spurious or unnecessary trips which may occur as a result of maintenance or the drop ~ of a low-worth rod. There are no other hardware ; modifications or procedure changes that will be made as a result of this deletion which could create the possibility of a new accident. No margin of safety will be reduced by this change. As noted above, if a dropped rod necessitates a trip, the trip function will be accomplished as a result of low pressurizer pressure. For those dropped rods for which no trip is necessary, the removal of this trip will provide protection against an unnecessary transient. REI)UCE ALLOWABLE PRIMARY TO SECONDARY LEAKAGE The allowable primary to econdary leakage has been reduced to limit the offsite
' radiological dose consequences due to the reanalysis of the locked rotor, rod ejection, and single uncontrolled rod withdrawal FSAR Chapter 15 events. The new limits are more y conservative.than the current TS requirements. Lowering the allowable primary to secondary leakage will not increase the probability of a previously evaluated accident, it will ensure that the dose consequences of an accident are within allowable limits. The possibility of a new or different accident from any previously evaluated is not created because there will be no physical changes to the plant operating procedures, other than to more conservatively limit leakage. There will not be a significant reduction in the margin of safety due to the fact that the allowable leakage is more conservative.
- Based on the above, it is concluded that no significant hazards are associated with this
< change. INCREASE IN OPERABLE RCS LOOPS IN MODE 3 and INCREASE COLD ' LEG ACCUMULATOR REQUIRED BORON CONCENTRATION .
=These amendments will not involve any significant hazards' consideration. The proposed changes will result in the parameter or operating condition involved becoming more conservative than the current TS requirement. The NRC's own guidance, published in the-l- Federal ' Register (48CFR 14870), states that an amendment which results b conditions 4
l
- becoming more restrictive is not likely to result in signiGeant hazards consideration as a defined by 10 CFR 50.92. Therefore, it may be concluded, with no further analysis, that
, - these amendments will not involve a significant hazards consideration. ECC3 PUMP PERFORMANCE REQUIREMENTS The proposed amendments will not involve a significant increase in the probability or consequences of an accident previously evaluated because the Loss-of-Coolant-Accident (LOCA) analysis, to which the ECCS flowrates are input assumptions, is unchanged and, therefore, continues to meet applicable acceptance criteria. The proposed amendments will not result in a significant increase in the podility of a
.new accident because the new values represent a change in required pump performan2.
The new values represent no change in the assumptions made in the LOCA analysis, or any a physical change in the plant. Enough margin exists between the flow used in the LOC # analysis and the new required pump flows that a reanalysis was not necessary The proposed changes wi'l not result in a significant decrease in a margin ol' safety, because pump performance at the new values is sufficient to meet all acceptance criteria. in both the current FSAR analysis and any analysis associated with Catawba 1 Cycle 7. Based on the above, it is concluded that no significant hazards exist.
. INCREASE IN PRESSU.RIZER _ CODE SAFETY VALVE SETPOINT TOLERANCES The proposed amendment will not result in a significant increase in the probability or consequences of any previously analyzed accident. The valve lift setting is challenged only i
- after a transient has been initiated and is not a contributor to the probability of any transient or accident. The transients'which involve' pressure increases which would potentially
~
challenge the safety valves have been analyzed td determine the consequences of delayed or premature valve actuation at the extremes of the new setpoint tolerances. These analyses show that all applicable acceptance criteria are met using the wider tolerances. The proposed amendment will not result in the creation of any new accident not previously [ evaluated. As noted above, the setpoint tolerance only affects the time at which the safety . L valve opens following or during a transient', and is not a contributor to the probability of b an accident. The proposed amendment will not result in a significant decrease in a margin of safety. The limiting transient in each accident category has been analyzed to determine the effect
. of the change in lift setpoint tolerance on the transient. In each case, the results of the analyses met all acceptance criteria.
l 5 l-l I __ __. .- ._ . - _ . . _ . ,_- _ _ . -
i l Based on the above, it is concluded that no significant hazards exist. LOW STEAM LINE SETPOINT PRESSURE CIIANGE l Changing the Low Steam Line Pressure setpoint and removal of dynamic compensation will not increase the probability or consequences of any previously-reviewed accident. The higher steam line pressure setpoint is consistent with all licensing basis safety analyses. This change, in conjunction with the removal of the dynamic compensation of the steam pressare signal, is intended to reduce or eliminate spurious Engineeced Safeguards Features (ESF) actuations which are caused by minor (but rapid) pressure decreases in the secondary system. The proposed amendment will not result in a new accident not previously reviewed. A , change in steam line pressure is a response to an existing transient condition, rather than a precursor or initiating event. A change in the steam line pressure setpoint is also not a precursor or initiating event. The proposed amendment will not result in a significant decrease in a margia of safety. The reanalysis of the steam line break accident which was performed shows that all imposed Condition II acceptance criteria are met. Based on the above, it is concluded that no significant hazards exist.
- FEEDWATER AND MAIN STEAM LINE ISOLATION VALVE STROKE TIME The proposed changes to the valve stroke times in Tables 3.3-5 and Table 3.6-2a will not significantiv increase the probability or consequences of any previously evaluated accident.
The ef% cts of the delays in isolation times on the wrious transients affected have been analyzed and found to be acceptable. Since these valves do not recieve a containment isolation signal, and no credit is taken for operation of these valves in the dose analysis for a containment isolation function, a maximum stroke time does not apply for containment isolation. - The proposed changes will not significantly increase the possibility of a new accident not previously evaluated. Feedwater and main steam isolation are responses to ongoing transients, rather than initiators or precursors of transients. No equipment or component reconfiguration will occur as a result of this change. The proposed changes will not significantly decrease any margin of safety. As noted above, the effects of the longer isolation times have been evaluated and found to be acceptable. 6
C l.
?
Based on the above, it is concluded that no significant hazards exist. REVISE LIST OF ACCIDENTS REQUIRING REEVALUATION IN TIIE EVENT OF AN INOPERABLE RCCA The proposed change to Table 3.3-1 will not change the probability or consequences of any accident or reduce any safety margin, because the table simply lists accident analyses which must be reevaluated in the event of an inoperable rod cluster control assembly (RCCA). The activities involved are analytical only, and do not introduce any operational considerations. Revision of the table to more accurately define the affected analyses is an ' administrative effort relced to activities (analyses) which are conducted offsite after the fact of a postulated inoperable RCCA. ~ Based on the above, it is concluded that no significant hazards exist. ENVIRONMENTAL IMPACT STATEMENT The proposed Technical Specification change has been reviewed against the criteria of 10 CFR 51.22 for environmental considerations. As shown above, the proposed changt does not involve any signi6 cant hazards considention, nor increase the types and amour.,s of effluents that may be released offsite, nor increase -the individual or cumulative occupational radiation exposures. Based on this, the proposed Technical Specification change' meets the criteria given in 10 CFR 51.22 (c) (9) for categorical exclusion from the requirement for an Environmental Impact Statement. V , h 7 M N}}